Text

Forwards Rept on Unmonitored Release Pathways to Assist in Closeout of NRC Open Item 50-320/85-21-05.Pathways Evaluated to Determine Release Bounds & Dose Assessments
	ML20246K685
Person / Time
Site:	Crane
Issue date:	05/10/1989
From:	Roche M; GENERAL PUBLIC UTILITIES CORP.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	4410-89-L-0005, 4410-89-L-5, NUDOCS 8905180124
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GPU Nuclear Corporation UC Mr Post Office Box 480 Route 441 South Middletown, Pennsylvania 17057 0191 717 944 7621 TELEX 84 2386 Writer's Direct Dial Number:

(717) 948-8400 May 10,1989 4410-89-L-0005/0216P US Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Dear Sirs:

Three Mile Island Nuclear Station, Unit 2 (TMI-2)

Operating License No. DPR-73 Docket No. 50-320 Closecut of NRC Open Item 50-320/85-21-05 As documented in NRC Inspection Report 50-320/85-21 and continued in 50-320/86-08, GPU Nuclear has identified potential unmonitored release pathways which may have existed during and af ter the TMI-2 accident in March 1979. These pathways were evaluated to determine release bounds and dose assessments. Attached for your information is a summary report regarding this evaluation which also serves to update the 1979 Dose Assessment reports previously submitted. GPU Nuclear is providing this information to assist in closing Inspector Follow-up Item 50-320/85-21-05.

Sincerely, i

ND M. B. Roche Director, TMI-2 EDS/ emf Attachment cc:

F. I. Young - Senior Resident Inspector, TMI W. T. Russell - Regional Administrator, Region I J. F. Stolz - Director, Plant Directorate I-4 L. H. Thonus - Project Manager, TMI Site j

I \\

8905180124 890510

{DR ADOCK0500go GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation

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REPORT ON UNMONITORED RELEASE PATHNAYS AT TMI-2

1.0 INTRODUCTION

Release pathways during the TMI-2 accident were investigated and the releases documented in TDR-THI-116. " Assessment of Offsite Radioactive Doses from the Three Mile Island Unit 2 Accident," July 31, 1979, (Reference 1).

The release pathways were evaluated and documented in TDR-55, " Pathways for Transport of Radioactive Material Following the TMI-2' Accident," July 8, 1981, (Reference 2).

Releases to the atmosphere were from Reactor Building leakage, steam released from the atmospheric steam dump valves and ventilation air from the Auxiliary and Fuel Handling Buildings which was exhausted through filtration systems via the plant vent.

An evaluation of potential unmonitored release pathways at TMI-2 was initiated due to the discovery of two unmonitored release pathways, one from the Contaminated Drain Tanks WDL-T-11A/B and a second from a ventilation duct connecting the M-20 area of the THI-2 Control Building Area Basement to the River Water Pump House (RWPH).

In addition to these two pathways, twelve other potential unmonitored release pathways have been identified:

Atmospheric Steam Releases (ASR)

Turbine Building Out-Leakage (T. Bldg)

Diesel Generator Building (DG Bldg)

Control Building Out-Leakage Door No. 10, 305' Elevation of the Autiliary Building EPICOR II Liner Door Borated Water Storage Tank (BHST)

Processed Water Storage Tanks 1 & 2 (PWSTs 1 & 2)

Condensate Storage Tank (C0-T-1A)

Sewage Tanks Storage (Waste) Module Sump Groundwater This evaluation was based on available liquid sample results, airborne sample results, water transfer records, and plant design exhaust air flow criteria.

The reported sample analysis values were used without regard to uncertainty, i.e., errors associated with the analysis, which varied from a few percent to over fifty percent. Discounting the uncertainty errors was judged to provide a more realistic estimate of the releases, whereas using the upper statistical limit of each sample result would yield an unrealistically conservative upper bound.

In addition, the sample concentrations were assumed to be constant from one sample to the next sample and instantaneously mixed with no buildup or decline.

When sample data was not available, relationships were established with activity in the plant vent at the time and exhaust flow rate.

No credit has been taken for plate-out.

The Reactor Building has a design integrated leakage rate of less than 0.13%

by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at 56.2 psig. During the first 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months of the accident, the Reactor Building was under a slight positive pressure ranging from 0 to 4.3 psig. Since Reactor Building atmospheric isolation occurred at 07:56 on March 28, 1979, at a very low leak rate, the leakage through the penetrations would not have been a significant factor in 0216P

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' the amount of radioactive material released (Reference 2).

At 13:50 on March 28,'1979, a 28 psig Reactor Building pressure spike (hydrogen burn) occurred, 4

which would have led to the release of some airborne radioactive material from the Reactor Building through penetrations into the Auxiliary and Turbine Buildings that would have been proportional to the Reactor Building leak rate.

From about 14:00 on March 28, 1979 until the controlled purge of the TMI-2 Reactor Building during June 1980, the Reactor Building was maintained at slightly negative pressure, therefore, any leakage would have been insignificant.

1 The results of this evaluation indicate that there were seven instances during 1979 when a calculated unmonitored airborne release exceeded the + 25% error l

margin which was originally reported to the NRC in MET-ED/GPU letter GQL 1129 1

dated August 30, 1979 and MET-ED/GPU letter TLL 094, dated February 29, 1980, (Refer to Table 3).

The unmonitored airborne releases occurred mainly from the M-20 area of the TMI-2 Control Building Area Basement via the RHPH (Reference 3).

Additional unmonitored releases from the Turbine Building also contributed on three occasions after 1979. All other potential unmonitored release pathways would not have represented a significant contribution to TMI-2 releases and were within the error that was reported.

The unmonitored release pathway from the M-20 area of the TMI-2 Control Building Area Basement via the RHPH was eliminated on or about September 27, 1985.

The fan has been isolated by disconnecting the ductwork at the fan and by capping the fan ductwork opening. As documented in this evaluation and for the current condition of TMI-2, the remaining pathways do not represent a potentially significant release pathway during Mode 4.

This report presents the results of the evaluations performed on thirteen potential unmonitored release pathways.

The Contaminated Drain Tank evaluation was submitted to the NRC via GPU Nuclear letter 4410-86-L-0038, dated February 28, 1986, (Reference 4) and is not included in this report.

This evaluation of potential unmonitored pathways concludes that the only significant airborne releases to the environment occurred via the plant vent and that any releases through the unmonitored pathways does not affect the previously calculated dose to the population as documented in TDR-TMI-ll6 (Reference 1). 0216P

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2.0, RESULTS AND DISCUSSION E

2.1 CONTROL BUILDING M-20 AREA VIA RIVER WATER PUMP H00!_

This unmonitored release pathway consisted of a ventilation duct connecting the M-20 area of the TMI-2 Control Building Area Basement with the River Water Pump House (RWPH)..Also, there is a pathway from the M-20 area into the Turbine Building.

The most conservative approach is that all activity in M-20 was released via the RWPH.

This potential unmonitored release pathway is considered to be the most significant unmonitored pathway to the environment (Reference 3).

The characterization of particulate, iodine and noble gas activity potentially released via the RHPH pathway to the environment is based on available M-20 area airborne radioactivity sample results (Refs. 3 & 5).

As a conservative measure, it was assumed that the concentration present in the M-20 area at the time of each sample had existed since the previous sample.

It was also assumed that a mixing effect of the recirculation fans occurred in the RHPH as the air was drawn from the M-20 area.

The evaluation considered the mixing to be instantaneous and contributed to a dampening of short-term high concentration spikes.

An exponential approach to equilibrium was also assumed in calculating airborne radioactivity concentrations within the RHPH (Reference 3).

The derived RHPH airborne concentrations were converted to release rates.

The postulated mechanism for release to the environment is-the 200 CFM (9.44E+4 cc/sec) exhaust flow rate of the fan connecting the duct in the M-20 area to the RHPH.

The release rates were compared to limits for iodines, particulate, and noble gases.

In each case, the Technical Specifications instantaneous release rate limit for TMI-2 was initially partitioned by a ratio of the exhaust volume flow rate of the RWPH and plant vent.

If this analysis indicated a Technical Specifications limit could have been reached, a further review was performed to determine if total station effluents for that time period would have exceeded Technical Specifications limits. Additionally, the total activity for each radionuclides released through the RWPH pathway was calculated by integrating the activity concentrations with time and tabulating the results by calendar quarter (Reference 3).

2.1.1 TOTAL RADI0 ACTIVITY RELEASED TO THE ENVIRONMENT The total activity released to the environment through the RWPH pathway was calculated for each radionuclides based on the calculated releases from the M-20 area into the RHPH.

All of the activity from the M-20 area was assumed to have been released to the environment using an exponential model.

It was also assumed that the concentration present in the M-20 area at the time of each sample had existed since the previous sample.

No air sample results were available for the period prior to April 6, 1979, to estimate releases from the RHPH.

Therefore, the first sample on April 6, 1979, was used to estimate potential initial releases. A spline fit of the airborne concentrations from i 0216P

L April 6,.1979 through April 21, 1979, indicates that an exponential

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extrapolation from March 28, 1979, to April 6.-1979, was a reasonable assumption for particulate and noble gas release estimates.

Releases for I-131 and I-133 were estimated based on a exponential extrapolation, decay corrected, to March 28, 1979, because the. sample data indicate that this is a reasonable, yet conservative approach.

The calculated releases from the RHPH to the environment, by radionuclides and activity for the period of March 28, 1979, to December 31, 1979, are as depicted in Table 1.

Based on all available air samples collected in the M-20 area after December 31, 1979, the radionuclides activity was less than the lower limit of detection (LLD) after that time (Reference 3).

In summary, the unmonitored releases via RWPH in addition to the previously reported values are postulated as follows:

an additional 5.76E-5 Ci of Xe-133m, 2.6E-6 C1 of Cs-134, and 5.25E-6 Ci of Cs-137 during the First Quarter of 1979 and an additional 2.7E-4 Ci of Cs-134 and 4.8E-4 C1 of Cs-137 during the Fourth Quarter of 1979.

TABLE 1 AIRBORNE RADI0 ACTIVITY' RELEASES FROM THE RWPH TO THE ENVIRONMENT FROM MARCH 28. 1979. THROUGH DECEMBER 31. 1979 Radionuclides Total Activity (Ci)

I-131 3.53E-1 I-133 6.16E Xe-133 4.50E+0 Xe-133m 1.07E-4 Cs-134 2.76E-4 Cs-137 5.15E-4 Co-60

< LLD 2.1.2' INSTANTANEOUS RELEASES RATES 2.1.2.1 10 DINE-131 AND 10 DINE-133 The airborne activity concentrations in the RHPH were derived from airborne samples taken in the M-20 area.

The derived concentrations in the RHPH were then converted to release rates. The concentrations and releases of I-133 were based on the I-131 and I-133 ratio released from the plant vent, because sample data was not available. The remaining I-133 results were less than the LLD.

The maximum instantaneous release rate occurred on March 28, 1979.

The release rates for I-131 and I-133 were 6.32E-4 pCi/sec and 3.08E-4 pC1/sec, respectively based on an exhaust flow rate of 200 CFM (9.44E+4 cc/sec). This calculation neglects any 0216P

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effects of plate-out which may have occurred in the ductwork s

The Technical Specifications release rate limits were not exceeded.

2.1.2.2

-PARTICULATE Airborne particulate concentrations in the RHPH were derived from airborne sample results of the M-20 area.

It is assumed that all activity was released to the environment.

The calculation method conservatively neglects any effects of plate-out which may have occurred.in the duct from the M-20 area to the RHPH.

It should be noted that smear samples from the impeller blade-of fan AH-E-13 in the M-20 area and radiation measurements of the filter in the explosive gas sensor in the RHPH did not indicate particulate radioactivity above LLD (Reference 3).

l The maximum release rate calculated for particulate was 2.72E-4 pCi/sec on October 5, 1979.

This value is less than the Technical Specifications release rate limit.

In no instance was the maximum permissible concentration (MPC) for unrestricted areas exceeded (Reference 3).

2.1.2.3 NOBLE GASES 2.1.2.3.1 Xe-133 The calculated Technical Specifications release rate limit for Xe-133 1s 4.5 E+4 pC1/sec.

If this release rate limit is partitioned to the RHPH, a maximum value of 51.4 pCi/sec, corresponding to a release concentration of 5.4E-4 pCi/cc.

could be released from the RHPH without exceeding the Technical Specifications instantaneous release rate limit (Reference 3).

In addition, the calculated Technical Specifications calendar quarter release rate limit and corresponding concentration of Xe-133 in the RHPH are 10.1 pCi/sec and 1.07E-4 pC1/cc, respectively.

The derived maximum concentration and maximum release rate of Xe-133 from the RHPH was 1.38E-6 pC1/cc and 1.3E-1 pCi/sec, respectively; this occurred on March 28, 1979. Therefore, the Technical Specifications release rate limits were not exceeded. Noble gas concentrations released to the environment were calculated in a manner similar to the method for I-131.

In no instance was the MPC for unrestricted areas exceeded.

2.1.2.3.2 Kr-85 A Kr-85 concentration of 5.7E-4 pCi/cc was measured in the M-20 area on July 13, 1982.

If this Kr-85 concentration was assumed to be transferred to the RHPH, the resulting concentration in the RHPH would be 4.11E-4 pCi/cc.

The corresponding release rate would be 38.8 pCi/sec and a total 0216P

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of 3.35 Ci of Kr-85 could have been-released.

This release rate is-less than the derived RWPH instantaneous release rate limit of 51.4 pCi/sec.

From the TMI-2 Third Quarter 1982 Quarterly Dose Assessment Report,- 4410-82-L-0047 dated November 19, 1982.(Reference 6),

25.9 Ci of Kr-85 were released from the plant vent and 86.4 C1 of Kr-85 were released from EPICOR II vent, for a total of 112.3 C1.

The range of error for Kr-85 reported in the effluent report is 160%. Compared to the reported.112.3 C1 of Kr-85, an additional 3.35 Ci represents an approximate 3%

increase which is well within the 160% error margin.

2.2 ATMOSPHERIC STEAM RELEASES (ASR)

During the March 1979 accident, Steam Generators A and B were periodically operated in an atmospheric dump mode to control primary system pressure.

Radiological samples indicated no elevated activity levels during this mode of operation.

At approximately 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and 55 minutes into the accident, Steam Generator B was used to eject heat from the primary system to the condenser.

In approximately 4 seconds, the condenser vacuum pump discharge gaseous radiation monitor-(VA-R-748) increased from 100 counts per minute (cpm) to 8E+5 cpm at 7:07 a.m. on March 28, 1979 (Reference 2).

Based on a Xe-133/I-131 ratio.of 80:1, this corresponds to a release rate of approximately 2.0E-2 pCi/sec of I-131.

Note that in this configuration, this release of I-131 was monitored by the station.

vent atmospheric radiation monitor (HP-R-219).

When the high alarm setpoint of 2,000 cpm, which corresponds to a concentration of 5.5E-6 pCi/cc or a release rate of approximately 1 pC1/sec of noble gas, was exceeded, Steam Generator B was isolated.

During this evolution, Steam Generator B's isolation valves were manually opened to the steam chest and, thus, to the environment via the atmosphere dump valves (ADV) for approximately 12 seconds.

Based on an activity of 8E+5 cpm on VA-R-748 during this 12 second interval and a ratto of noble gas to I-131 of 40:1 measured at the vacuum release sample panel, the release rate of I-131 was 10.8.pC1/sec for 12 seconds, i.e.,

approximately 130 microcuries of I-131 was released.

Based on losing the condenser partition factor of 7.5E-3, this relates to a total release of approximately 1.73E-2'Ci of I-131 released through the ADV to the environment.

It is concluded-that this release did exceed the. Technical Specifications particulate release rate limit of 0.3 pCl/sec.

For comparison, the monitored release rate from the plant vent was above 4 pCi/sec for a period of about 5.4E+4 seconds (Reference 1).

2.3 TURBINE BUILDING OUT-LEAKAGE (T. BLDG)

Turbine Building air samples were used to estimate releases from the Turbine Building. Source leakage into the Turbine Building and the data from the Turbine Building sump analyses were not considered individually but were assumed to be combined into the measured airborne results. 0216P l

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l The design criteria listed in the TMI-2 Final Safety Analysis Report

' (FSAR) Section 12.2.3 were used as a basis for this estimate of the Turbine Building unmonitored release pathway source term.

The FSAR lists the free air volume of the Turbine Building as 2.8E+6 cubic feet (7.93 E+10 cc) and a designed unfiltered exhaust vent flow of 230,000 CFM (1.09E+8 cc/sec).

The first positive sample of airborne radioactivity in the Turbine Building was on April 1, 1979.

The analytical results were 2.10E-10 pCi/cc I-131 and 4.3E-8 pCi/cc Xe-133. This sample was

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l conservatively assumed to be positive retrospective to March 28, 1979,

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even though analysis of the first sample taken on March 30, 1979, at 1148 hours0.0133 days 0.319 hours 0.0019 weeks 4.36814e-4 months in the east end of the Turbine Building 348' elevation revealed less than LLD.

Based on the assumption that the April 1, 1979, sample is conservative, a total of 2.97E-3 Ci of I-131 and 6.08E-1 Ci of l

Xe-133 were released in the First Quarter of.1979 from the Turbine Building.

For the Second Quarter of 1979, various air sample results from the Turbine Building were used to generate postulated releases. Based on all air sample results.and the maximum design flow rate of the Turbine Building ventilation system, Table 2 provides the total activity for each radionuclides that was released via this pathway during the First and Second Quarters of 1979.

TABLE 2 ESTIMATES OF TURBINE BUILDING AIRBORNE RELEASES, DURING THE FIRST AND SECOND QUARTERS 1979 Radionuclides Curies I-131 2.71E-1 I-133

< LLD Xe-133 6.39E+0 Xe-133m 1.37E+0 Cs-134 6.77E-5 Cs-137 2.04E-3 Co-60 3.46E-4 Of these results, only the postulated release of Co-60 is significant when compared to the previously reported value.

The originally reported value was 8.71E-5 Ci of Co-60 for the Second Quarter of 1979.

There were no positive results present in any Turbine Building air samples taken after the Second Quarter of 1979.

2.4 DIESEL GENERATOR BUILDING (DG BLDG)

Following the THI-2 accident, ingress and egress to the Auxiliary Building was through the DG Bldg.

Based on airborne sample analysis within the DG Bldg. and the designed exhaust flow rate, it is estimated that 2.69E-1 Ci of I-131 and 6.14E-3 Ci of Xe-133 were released during the Second Quarter of 1979. The estimated releases of I-133 were less than LLD.

In no instance was the MPC for unrestricted areas exceeded. 0216P

s 2.5, CONTROL BUILDXNG OUT-LEAKAGE The Control Building has three areas that may vent directly to the atmosphere.without filtration or monitoring (Reference 7).

These are:

l

1) the Mechanical Equipment Room which has a ventilation supply and

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exhaust of 1000 CFM (4.72E+5 cc/sec), 2) the Unit 2 Control Room which has a ventilation supply and exhaust ranging from 1500 CFM to 14,350 CFM (7.lE+5 cc/sec to 6.77E+6 cc/sec), and 3) the Cable Battery and l

Switchgear Room, which has a ventilation supply and exhaust ranging from 750 CFM to 29,820 CFM (3.54E+5 to 1.4E+7 cc/sec). Of these three potential contributors, only the 1st) area and the 3rd) area are considered in this evaluation.

Their combined maximum exhaust is 30,820 CFM (1.45E+7 cc/sec). The Control Room ventilation or 2nd) area is not considered to be an unmonitored pathway, because following the accident, it was placed in the-recirculation mode and there was no exhaust to the environment.

At the.present, there are no specific airborne sample results for the-Mechanical' Equipment Room or the Cable Battery and Switchgear Rooms that have been identified. Therefore, as a conservative estimate, the results of five.available airborne samples from the Control Building were used to estimate a source term.

The calculation assumes that a sample taken at the Control Building 322' elevation on April 18, 1979, which was the maximum concentration of the five air samples, is representative of all Control Building airborne activity and that this concentration existed for a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period. This sample indicated a I-131 concentration of 9.9E-11 pC1/cc. When decay corrected to March 28, 1979, the concentration would be 6.6E-10 pCi/cc.

The maximum release rate of iodine and particulate was calculated to be 1.44E-3 pCi/sec on April 18, 1979; decay corrected to March 28, 1979, the release rate would be 9.6E-3 pCi/sec.

This value is a fraction of the Technical Specifications instantaneous iodine and particulate release rate of 0.3 pC1/sec.

For Xe-133, maximum concentration of 1.8E-7 pCi/cc, which resulted in a maximum release rate of 2.54E+0 pCi/sec, occurred on March 28, 1979.

This maximum concentration is less than the MPC value of 3.0E-7 pCi/cc for Xe-133 listed in 10 CFR 20, Appendix B. Table II, Column 1.

This maximum release rate is less than 0.006% of the calculated Technical Specifications instantaneous release rate limit for Xe-133 of 4.5E+4 pCi/sec.

In summary, the maximum source term contribution from the Control Building unfiltered pathway is approximately 0.5% of the Technical Specifications release rate limit for iodines and particulate and 0.003%

of the Technical Specifications instantaneous gross gaseous release rate limit.

2.6 DOOR NO. 10. 305' ELEVATION AUXILIARY BUILDING Door No. 10 at the 305' elevation in the Auxiliary Building opens to the environment.

Under normal operating conditions and when airborne contamination is measured in adjacent areas, Door No. 10 is secured.

However, at times it is necessary to open this door to perform transfers to the Auxiliary Building. 0216P

i In order to assess'the magnitude of potential releases through this pathway, an experiment was conducted on December 29, 1986 to measure the linear flow velocity into the Auxiliary Building with Door No. 10 open.

The average of three values taken at approximately a-five (5) ft height was 583 feet / minute (ft/ min). Concurrent Auxiliary Building ventilation fans running during the test were 7 A/B and 8 C/0.

This fan arrangement provides 51,500 CFM (2.43 E+7 cc/sec) supply and 67,000 CFM (3.16E+7 cc/sec) exhaust through the Auxiliary Building. -The concurrent Fuel Handling Building exhaust flow-was 44,000 CFM (2.08 E+7 cc/sec).

Therefore, the combined Auxiliary and Fuel Handling Buildings exhaust was 111,000 CFM (5.24E+7 cc/sec).

The area of Door No. 10 in the open position was estimated to be 166 ft2 (Reference 8).

Therefore, assuming all make-up air is drawn through Door No. 10, the calculated inward flow through Door No. 10 is 96,778 CFM (4.57E+7 cc/gec) based on an average linear flow of 583 ft/ min and an area of 166 ft'.

This constitutes approximately 87% of the makeup air to the Auxiliary and Fuel Handling Buildings.

A second calculation was made with the Auxiliary and Fuel Handling Buildings ventilation at the Technical Specifications minimum value of 88,000 CFM (4.15E+6 cc/sec).

This value, when divided by the surface 2

area of Door No. 10 (166 ft ) equals 530 ft/ min inflow into the Auxiliary Building. Using the 87% value calculated above, a minimum of 461 ft/ min could be drawn into the Auxiliary Building HVAC through the open Door No. 10 at minimum exhaust flow rate conditions for the Auxiliary and Fuel Handling Buildings.

A third calculation was made with Auxiliary Building exhaust off and Fuel Handling Building ventilation at mirvimum flow.

If only 10% estimated air was being drawn from the Auxiliary Building through the Fuel Handling 2

Building (3,600 CFM/166 ft ) then air would still be drawn in through Door No. 10 at approximately 22 ft/ min, assuming all makeup air is from the Auxiliary Building.

Included in the design features of the plant is the condition that the supply fans trip before the exhaust fans trip, in order to avoid pressurizing the buildings.

Under worst conditions, where the Auxiliary Building supply and exhaust are shutdown and only 10% of makeup air is drawn from the Auxiliary Building through the Fuel Handling Building, a slight negative flow of 22 ft/ min or 3650 CFM (1.72E+6 cc/sec) into the Auxiliary Building should still be measured at Door No. 10 when it is open.

Therefore, except under conditions where the building could become pressurized, releases through this pathway would be considered insignificant.

2.7 EPICOR II LINER 000R Normal operation of the Auxiliary Building Liquid Cleanup System requires that the ventilation system and monitor be operable.

Receipt of an abnormal alarm requires that processing be terminated.

As a result, it is not reasonable to expect that significant abnormal releases would occur through the liner door at EPICOR II.

To evaluate the typical

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i airborne activity, samples from the EPICOR II Catualk and EPICOR II

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' Mezzanine AMS-3 monitors were evaluated-for 1988.

The average gross alpha concentration was 4.7E-14 pC1/cc, the average gross beta was 5.1E-13 pC1/cc, the average Cs-137 was 1.97E-12 pCi/cc, the average Cs-134 was less than LLD, and there was one positive value for Sr-90 of 9.9E-14 pCi/cc. These levels are well below MPC values for l

unrestricted areas. As a result, particulate concentrations within i

EPICOR II during normal operations do not result in unmonitored releases i

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of significant environmental impact.

l 2.8 BORATED HATER STORAGE TANK (BHST)

The BHST is not directly vented to the atmosphere, but is connected via relief valve /vacue breaker arrangement (Reference 9).

Tritium, as tritiated water vapor, is the only significant contaminant released to the atmosphere from water storage tanks. Although other radionuclides are present they remain in solution as dissolved solids. Transient t

releases of tritiated water vapor occur due to evaporation.

The average water temperature in the BHST is 79*2 'F.

The expected contribution to 7

the total releases from the BHST due to evaporation is approximately 20%

based on a 15 degree average annual heating and diurnal variation (Reference 10). The other approximately 80% of tritiated water vapor

)

occurs when a water transfer is made to the BWST.-

Tritiated water vapor in equilibrium with the aqueous phase is displaced from the tank during water transfers. This analysis assumes that the release to the atmosphere of tritiated water vapor is equal to the tank fill rate in gallons per minute.

Water transfer and radionuclides analysis records were used to estimate the release of teitiated water vapor from the BHST for the time period of 1980 through 1988.

No water transfers to the BHST were made for the year 1979 and First Quarter of 1980.

I The maximum postulated release of tritium occurred in the Second Quarter of 1980, when a calculated 3.93E-3 Ci was displaced by the introduction of 319,500 gallons of water into the BHST. A tritium concentration of i

9.37E-2 pC1/cc in solution and a partition fraction of 3.47L.-5 was assumed (Refs. 11 & 12).

The second highest release of tritium occurred in the Second Quarter of 1985 when 1.10E-3 C1 was displaced by the transfer of 100,408 gallons from the Clean Water Receiving Tank, and 38,855 gallons from the Boric Acid Mix Tank.

A tritium concentration of I

6.0E-2 pC1/cc and a vapor concentration of 2.0E-6 pCi/cc were calculated.

j The isotope effect, which is the effect due to the difference in mass i

between isotopes of an element on the rate and/or equilibrium of chemical l

transformations, or the suppression effect due-to approximately 3,000 parts per million (PPM) of boric acid in solution were not considered in these estimates. These two (2) combined effects are estimated to cause i

only a 1% reduction in the partitioning factor for tritiated water vapor i

released from the BHST.

Table 4 provides an estimats of the tritium released from the BHST for the period of 1980 through 1988.

The total unmonitored tritium released during the period of 1980 through 4

1988 from the BHST is estimated to be 1.3E-2 Ci.

I 1 0216P 1

l G

c 2. 9, PROCESSED WATER STORAGE TANKS 1 & 2 (PWSTs 1 & 2)

The PWSTs are vented directly to the atmosphere.

The driving force behind the release of tritiated water vapor from the PWSTs is volume displacement due to the addition of water to the tanks.

The initial transfer of water to PWST-1 was performed on July 25, 1981.

There were 11 subsequent water transfers to PWST-1 through August 27, 1981.

The maximum volumes of water transferred were 51,639 gallons during July 1981 and 52,037 gallons during August 1981.

There were no tritium analyses performed on the water transferred.

Therefore, it was assumed that the initial average tritium concentration was 1.05 pCi/cc in the PWSTs (References 11 and 12). A partition fraction of 3.47E-5 was also assumed.

No water transfers to PWST-1 were recorded from September 1981 until January 3, 1982.

The initial transfer of water to PWST-2 was on September 19, 1981.

Subsequent releases of tritium were based on recorded volumes of water transferred to the PWST, and tritium concentrations derived from curie values divided by the volume of each transfer.

Following the installation of recirculation and sampling capabilities in the PNSTs, the average tritium concentration was measured to be:

H-3 YEAR (p C1/ml) 1985 0.34 + 0.08 1986 0.24 1 0.03 1987 0.12 2 0.03 1988 0.11 0.04 The decrease in tritium concentration is attributed to both radioactive decay and dilution.

Table 4 lists the releases of tritiated water vapor from the PWSTs for the period of 1981 through 1988.

The only significant postulated release (5.41E-3 C1) occurred in the Third Quarter of 1985.

This release exceeded the originally reported value by approximately 0.2%, which is well within the 160% error niarsjin of that previously reported.

The total unmonitored tritium released from July 25, 1981, through December 31, 1988 from the PWSTs it estimated to be 9.83E-2 C1.

2.10 CONDENSATE STORAGE TANK CO-T-1A To estimate the release of tritium from CO-T-1A, calculations similar to those for the BWST and PWSTs were performed. When water transfer volumes are unknown, a traximum tank fill rate of 280 gallons / minute based on the l

design flow rate of pump ALC-P-5 was assumed.

The release rate of water vapor was also assumed to be equal to the maximum water fill rate.

From 1980 until October 2, 1981, the tritium concentration was taken to be 0216P

___--________----_________J

s-

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constant and equal to 2.80E-1 C1/cc.

This concentration was derived from tritium analyses of EPICOR II. Clean Tank (CC-T-2) samples.

The transfer pathway was determined to be from CC-T-2 to C0-T-1A.

A partition fraction of 3.47E-5 was also assumed, as in Sections 2.8 and 2.9.

For a CO-T-1A tank concentration of 2.80E-1 pC1/cc tritium, the derived release rate.is equal to 1.72E-1 pC1/sec.

After October 2, 1981, the tritium concentration in CO-T-1A was 1.02E-1 pCi/cc which corresponds to a release rate of 6.25E-2 pCi/sec. This release rate was used, when water was transferred into CO-T-1A, to estimate the remainder of 1981 releases from CO-T-1A.

From January 1, 1980, to December. 31, 1988, the known volumes of water transferred and known tritium concentrations were used to estimate releases.

Table 4 lists the releases of tritiated water vapor from CO-T-1A.

The total of unmonitored tritium released during the period of 1980 through 1988 from CO-T-1 A is estimated to be 1.07E+1 C1.

2.11 SEWAGE TANKS Tritium,~ as tritiated water vapor is considered the only contaminant-capable of being relea.ed to the environment from sewage tanks. At typical concentrations of less than 1E-5 pC1/cc, and without the presence of any driving force other than displacement by water pumped into these tanks, concentrations greater than 3.5E-10 pCi/cc released to the environment are not anticipated.

These levels are several orders of magnitude below MPC to unrestricted areas. Compared to the evaluations for tritium released from the water. storage tanks (Bk..,

PWSTs, and CO-T-1A), any tritium displaced by water addition to sewage tanks is not considered significant.

2.12 STORAGE (WASTE) MODULE SUMP This sump was sampled weekly until 1989 at which time it is sampled prior to being discharged to the Industrial Waste Treatment System.

The average Cs-137 concentr.ation from 1983 through 1987 was 2.2E-7 pCi/cc.

Due to a leaking container the maximum Cs-137 concentration was 2.8E-5pCi/cc.

Cs-137 is a dissolved solid in solution and is not expected to be released in significant quantities due to evaporation of the water from the sump.

l Tritium can be evaporated as tritiated water vapor.

For three months in 1984, one month in 1985, one month in 1987, and three months in 1988, the tritium concentration for this sump was approximately 1.5E-5 pCi/cc.

If this maximum value of 1.5E-5 pCi/cc is assumed for a whole year along with a Class A Pan evaporation of 46 inches per year (Reference 3) a very conservative estimate of 6.2E-5 Ci of tritium / year could be released from this sump by evaporation.

However, the sump is not directly exposed to the atmosphere; whereas, the Class A Pan evaporation rates are measured by direct exposure to atmosphere.

Thus, this 0216P

}

, estimate, which assumes a uniform concentration throughout the year, and i

direct exposure to the atmosphere, is a very conservative estimate of the releases from the Storage Module Sump.

The normally reported values of tritium are several curies / month and, therefore, 6.2E-5 C1/ year of tritium is not significant by comparison.

2.13 GROUNDWATER A groundwater monitoring program was initiated around THI-2 in 1980. A network of stations was sited so as to detect leakage of water from the THI-2 Reactor and Auxiliary Buildings.and outside storage tanks.

Tritium concentrations in the TMI groundwater resulting from small leakage of water from the TMI-2 BHST are well below Federal regulatory limits (10 CFR 20, Appendix B).

Groundwater monitoring results are reported to the NRC in annual TMI Radiological Environmental Monitoring Reports (Reference 10). Therefore, no further discussion of unmonitored releases into groundwater is warranted.

. 0216P

r 4

..1 y

3.0 COMPARISON OF CALCULATED VALUES WITH SEMI-ANNUAL EFFLUENT REPORTS AND RELEASE LIMITS 3.1 Table 3 compares the values of the postulated releases'for all isotopes other than tritium via the unmonitored airborne pathways to those previously reported values. Values _ indicated by Note 1 are those that exceed the 125% margin of error associated with the reporting requirements of that time (References 13 and 14). Tritium, as trittated water-vapor, is of primary concern when in equilibrium with the aqueous phase and the air is displaced from a tank or pool during water transfers; it is not of concern otherwise.

Table 4 accounts.for postulated tritium airborne releases from water storage tanks or pools.

L All other tritium releases were monitored via the plant vent (References 13 and 14).

3.2 Table 4 compares the values of the postulated tritium airborne releases for the water storage' tanks by calendar quarter from 1980 through 1988.

Values indicated by Note 1 are those within the 125% margin of error; values indicated by Note 2 are those within the 160% margin of error..As the concentration of radionuclides inventory decreased, the margin of error in the calculations increased.

When each tank total in Table 4 is added to the previously reported value, the new value is within the error of the reporting criteria.

In fact, the sum total of the airborne tritium released from all water storage tanks from 1980 through 1988 was less than 1% of the total airborne tritium releases previously reported dur.ing the same period.

Based on this analysis, the unmonitored release pathways for tritiated water vapor from the water' storage tanks are not considered a significant contributor to releases from TMI-2.

. 0216P L

..o TABLE 3 s

CALCULATED CURIES RELEASED VIA UNMONITORED AIRBORNE PATHWAYS VS. REPORTED CURIES (C1)

Radio-1979 1979 1979 1979 nuclide Source 1st Quarter 2nd Quarter 3rd Quarter 4th Quarter I-131 RWPH 1.86E-4 3.53E-1

ML20246K685

Text

Dear Sirs:

1.0 INTRODUCTION

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