Text

Proposed Tech Specs,Providing Allowable RCS Specific Activity Limit Based on OTSG Insp Results Performed Each Refueling Outage
	ML20154P866
Person / Time
Site:	Three Mile Island
Issue date:	10/19/1998
From:	; GENERAL PUBLIC UTILITIES CORP.
To:	;
Shared Package
ML20138L375	List: ML20154P844; ML20154P866; ML20154P894;
References
	NUDOCS 9810230093
	Download: ML20154P866 (30)
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Proposed Technical Specification Revised Pages l

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1 9810230093 981019 1 PDR ADOCK 05000289 1 P PDR l

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LIST OF FIGURES FIGURE TITLE PAGE 2.1-1 Core Protection Safety Limit TMI-1 2-4a 2.1-2 DELETED 2.1-3 C' re Protection Safety Bases TMI-1 2-4c 2.3-1 TMI-l Protection System Maximum Allowable Setpoints 2-11 2.3-2 -DELETED 3.1-1 Reactor Coolant System Heatup/Cooldown Limitations 3-Sa (Applicahic thru 10 EFPY) 3.1-2 Reactor Coolant Inservice Leak and Hydrostatic Test 3-5h (Applicahic thru 10 EFPY) 3.1-2a Dose equivalent 1-131 Primary Coolant Specific Actual 3-9h Limit vs. Percent of RATED THERMAL POWER 3.1-2h Primary Coolant Pre-Accident DEI Based on 3-9c Primary-to-Secondary Leakage During a Hypothetical MSLB 3.1-3 Limiting Pressure vs Temperature Curve for 3-18b 100 STD cc/ Liter H2O 3.5-2A - i thru' DELETED I 3.5-2M '

3.5-1 Incore Instivmentation Specification 3-39a Axial Imbalance Indication 3.5-2 Incore Instrumentation Specification 3-39h Radial Flux Tilt Indication 3.5-3 Incore Instrumentation Specification 3-39c 3.11-1 Transfer Path to and from Cask Loading Pit 3-56h 4.17-1 Snubber Functional Test - Sample Plan 2 4-67 5-1 Extended Plot Plan TMI N/A l 5-2 . Site Topography 5 Mile Radius N/A 5-3 Gaseous Effluent Release Points and Liquid Effluent N/A Outfall Locations 5-4 Relationship Between Initial Enrichment and Acceptable 5-7a Fuel Burnup (Spent Fuel Pool A - Region II) vii 1

Amendment Nos.1IrMr29r39r454###r71r196rl#9r12#r136r134r142rl5#rl44r16; 448cl#4 l

3.1.4 REACTOR COOLANT SYSTEM ACTIVITY 3.1.4.1 LIMITING CONDITION FOR OPERATION The specific activity of the primary coolant shall be limited to:

a. Less than or equal to the most restrictive DOSE EQUIVALENT l 131 limit specified on Figure 3.1-2b, l and

b. Less than or equal to 100/ 5 microcuries/ gram

3.1.4.2 APPLICABILITY: at all times except refueling 3.1.4.3 ACTION:

MODES: Power Operation, Start-Up, Hot Standby

a. With the specific activity of the primary coolant greater tinn the most restrictive DOSE l

EQUIVALENT l-131 timit specified on Figure 3.1-2b. for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ** during one continuous time interval or exceeding the limit line shown on Figure 3.1-2a, be in at least ilOT SHU'IDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Power operation may continue when DOSE EQUlVALENT l-131 is i below the mos' astrictive limit specified on Figure 3.1-2b. I

b. With the specific activity of the primary coolant greater tinn 100/ 5 microcuries/ gram be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Power operation may continue when '

primary coolant activity is less than 100/ E microcuries/ gram.

MODES: At all times except refueling.

c. With the specific activity of the primary coolarit greater than the most restrictive DOSE l

EQUIVALENT I-131 limit specified on Figure 3.1-2b, or greater than 100/ E microcuries/ gram I perform the sampling and analysis requirements of Table 4.1-3 until the specific activity of the primary coolant is restored to within its lindts.

Bases '

j The maximum limitation on the specific activity of the primary coolant of 1.0 microcuric/ gram DOSE EQUIVALENT l-131 casures that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months dose at the site boundary will b: wcil within the Part 100 limit following a steam generator tube rupture accident. The limitations on the specific activity of the primary coolant specified on Figure 3.1-2b ensure that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the exclusion area boundary and the 30 day LPZ doses will be a small fraction of the Part 100 limit following steam line break accidem with postulated accident irjuced steam generator tube leakage in conjunction with an assumed steady state prinnry-to-secondary steam generator leakage rate of 1.0 GPM (Figure 3.1-2b represents total leakage from all sources). The limits .

specified in Figure 3.1-2b represent projected 2-hour leakage and twil accident duration leakage bawd on OTSG {

tube inspection results, not to exceed 7898 gallons in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or 24,243 gallons for the accident de moa. The l

allowable activity level is determined based on the OTSG tube inspection results for both 2-hom postulated leakage l and total accident duration leakage. The most restrictive primary coolant activity level establishes the maximum l limit for the operating cycle.

5 shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energics per disintegration (in McV) for isotopes, other than iodines, with halflives greater than 15 minutes, l making up at least 95% of the total non-iodine activity in the coolant.

- The time period begins from the time the sample is taken.

3-8 Amendment No. 408;44k244

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. Se ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than the limit specified in 3.1.4.1.a, but within the allowable limit shown on Figure 3.1-2a, accommodates possible iodine spiking phenomenon w hich may occur following changes in l-TilERMAL POWER Prc:. ceding to HOT SHUTDOWN prevents the release of activity should a steam generator tube mpture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves.

The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency ofisotopic analyses following power changes may be pennissible ifjustified by the data obtained.

The NRC stalT has performed a generic analysis of airborne radiation released via the Reactor Building Purge Isolation Valves. The dose contribution due to the radiation contained in the air and steam released through the purge isolation valves prior to closure was found to be acceptable provided that the requirements of Specifications 3.1.4.1,3.1.4.2 and 3.1.4.3 are met.

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20 JQ 40 50 to 70 80 90 100 F1RCENT OF RATEh THERMAL POWER FIGURE 3.1-2a Dose equivalent I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER l i 3-9b Amendment NoA0&;467,244 a

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Primary Coolant Pre-Accident DEI Bascd on Primary-to-Secondary Leakage During a Ilypothetical MSLB 30000 l 28000 !

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24000 22000

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Ci 18000 16 16000 l14000 y 12000 l L9 10000

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2000 0

0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 DEI (uCl/a/*3 l

2 Hour Leakage

Total Leakaged 1

(a) Total iodine activity as dose equivalent I-131 l

l Figure 3.1-2b Primary Coolant Pre-Accident DEI Based on Primary-to-Secondary Leakage During a Hypothetical MSLB i

l 3-9c Amendment No.

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b TABLE 4.1-3 MINIMUM SAMPLING FREQUENCY

> .I_se_m_ Check Frequency 5

1 1. Reactor Coolant a. Specific ActisityDetermmanontocompareto At least once each 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months dunng POWER OPERAT10N, f the 100/ E pCi/gm limit HOT STANDBY, START-UP, and HOT SHUTDOWN.

k b. Isotopic Analysis for DOSE EQUIVALENT i) I per 14 days dunng power operahons.

3 I-131 Cermian b ii) One Sample between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following a THERMAL

-* POWERchange eM.g 15% of the RATED !

] THERMAL POWER within a one hour penod dunng h.

.j 1mer ar *=. startg and hot sW iii) #Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , wheneser the specifx: actisity exceeds the most restnctist DOSE EQUIVALENT I-131 lin* .pecified in Figure 3.1-2b, or 100' E pCi/ gram i aunng allmodes but refuehng.

c. Radiochenucalfor 5Determmanon 1 per 6 months

danng power wouun.

d. Chemistry (CI,Fand02 ) 5 times / week when T , is greater than 200 F.

Y e c. Boronconcentranon 2 tunes / week

f. TritiumRadioactisity Monthly 2, Borated Water Storage Baron concentrahon Weekly and aRer each makeup when reactor coolant system Tank Water Sample pressure is greater than 300 psig or T , is greater than 200 F.

3. Core Floodmg Tank Water Bomn concentrabon Monthly and aRer each make up when RCS pressure is greater Sampic than 700 psig.

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ENCLOSUREI TM1-1 Technical Specification Change Request No. 272 Safety Evaluation No Significant llazards Consideration and Proposed Technical Specification Revised Pages l

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ENCLOSURE 1 1920-98-20210 Page1of20

1. Technical Specification Change Reauest No. 272 GPU Nuclear requests that the following changed replacement pages be inserted into existing Technical Specifications:

Revised Technical Specification Pages: vii,3-8, 3-9,3-9b,3-9c,4-9 These pages are attached to this enclosure.

II. Reason for Channe The purpose of this Technical Specification Change Request is to revise the TMI-l Technical Specification limit on reactor coolant system specific activity contained in Technical Specification Section 3.1.4 and Table 4.1-3. The proposed change revises the reactor coolant system specific activity limit from 0.35 microcurie / gram dose equivalent iodine 1-131 to a cycle specific dependent value based on the once-through steam generator (OTSG) inspection results performed each refueling outage, with a maximum allowable limit of 1.0 microcurie / gram dose equivalent iodine 1-131. The proposed maximum allowable limit of 1.0 microcurie / gram (pCi/gm) dose equivalent iodine 1-131 was the TMI-l Technical Specification limit prior to Cycle 12 stanup (October 1997).

III. Safety Evaluation Justifyine Channe TMI-l Technical Specification Amendment No. 204 issued October 2,1997 revised the TMI-l Technical Specifications to decrease the maximum allowable dose equivalent iodine (I-131) limit in the reactor primary coolant from 1.0 pCi/gm to 0.35 Ci/gm. This revision was required to suppon a proposed main steam line break (MSLB) accident reanalysis for TMI-l which accounted for additional accident dose consequences resulting from postulated post-accident once-through steam generator (OTSG) tube leakage. This MSLB accident reanalysis was originally submitted to NRC for review and approval on August 14,1997. NRC approval of the postulated OTSG post-accident tube leakage analysis was based on NRC's dose analysis which assumed a primary coolant activity limit of 0.35 Ci/gm. This assumption ensured that the resulting dose consequences in the NRC's confirmatory MSLB reanalysis did not exceed a small fraction of the 10CFR Part 100 guidelines or GDC 19 control room operator dose limits.

The proposed changes provide a cycle specific primary coolant specific activity limit based on the OTSG inspection results performed each refbeling outage, with a maximum allowable limit of1.0 Ci/gm. The proposed limits are based on an updated reanalysis of the TMI-l MSLB accident using revised assumptions for atmospheric dispersion coefficients and flashing fraction from the postulated tube leak pathway. The dose consequences remain a small fraction of the 10CFR Part 100 guidelines and GDC-19 limits as described below.

ENCLOSURE 1 1920-93-20210 Page 2 of 20 Main Steam Line Break (MSLB) Description The postulated MSLB is assumed to be the result of a double-ended rupture of a 24 inch outside diameter steam line on one steam generator from 100% power consistent with the existing TMI-l Steam Line Break design basis accident described in UFS AR Section 14.1.2.9. This is the largest possible break which results in the maximum cooldown rate.

Since the once-through steam generator (OTSG) design has the maximum inventory at full power conditions, staning the event from full power maximizes the heat removal capability of the steam generator. In addition, the postulated MSLB is assumed to occur with the failure of the feedwater regulating valve to the affected steam generator, This is the worst single failure, as it maximizes the overcooling for this event by maximizing the main feedwater flow to the affected generator. The effect of maximizing the overcooling is to maximize the steam generator axial tensile tube loads, which results in the maximum leakage.

Primary-to-secondary leakage from postulated flaws in the kinetically expanded tube-to-tubesheet joint are assumed to occur during the postulated steam line break accident.

This total leakage is limited to 2763 gallons in the first two hours, and a total leakage of 8525 gallons over the duration of the accident. Currently approved values of accident induced leakage described in TMI-l UFS AR Section 14.1.2.9 are a maximum of 3228 gallons for the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and 9960 gallons for the duration of the accident. Reactor coolant leakage into the steam generator continues until the Reactor Coolant System (RCS) can be cooled down and the leakage terminated. This was calculated to take a total of 23.33 hours3.819444e-4 days 0.00917 hours 5.456349e-5 weeks 1.25565e-5 months (Reference GPU Nuclear Calculation C-1101-224-E610-060, Rev.

0), and resulted in an average primary-to-secondanfl eak rate of 23.03 gpm (hot) for the initial 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the accident, and an average leak rate of 4.50 gpm (hot) for the remaining 21.33 hours3.819444e-4 days 0.00917 hours 5.456349e-5 weeks 1.25565e-5 months of the transient. The basis for the previously approved leakage values is contained in NRC Safety Evaluation Report, dated October 2,1997, issued for TMI-l License Amendment No. 204. This accident analysis is performed using Standard Review Plan (SRP)l5.1.5, Appendix A assumptions.

GPU Nuclear submittal to the NRC, dated November 26,1997 (6710-97-2441), provided the methodology used to evaluate the total accident induced primary-to-secondary leakage from the kinetic expansion region that may be postulated to occur during a design basis MSLB. Plant operating and surveillance procedure controls will ensure proper l implementation of the cycle specific primary coolant activity level based on the projected MSLB accident-induced primary-to-secondary leak rate.

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ENCLOSUREl 1920-98-20210 l Page 3 of 20 I lodine Spiking The environmental consequences from this accident are performed using SRP 15.1.5, Appendix A assumptions (Reference GPU Nuclear Calculation C-1101-900-E000-065, Rev. 0). Iodine spiking effects are analyzed for both an accident-induced spike (AlS) and a pre-accident spike (PAS). For the AIS, it is assumed that the unit has been operating with 1.0 Ci/gm Dose Equivalent lodine (DEI). The relative isotopic distribution in the reactor coolant is assumed to be the same as Table 14.2-4 of the existing TMI-l UFSAR, and is adjusted to a mix equivalent to 1.0 Ci/g DEI as shown on Table 1, attached. It is assumed that the reactor trip and/or primary system depressurization creates an iodine spike in the primary system to a value 500 times greater than the release rate .

corresponding to the 1.0 Ci/g DEI equilibrium activity assumed prior to the accident. !

!- No spiking of noble gases is assumed to occur.

For the pre-accident spike it is assumed that a reactor transient has occurred prior to the postulated MSLB and has raised the primary coolant iodine concentration to 60 pCi/gm DEI, which is the maximum permitted at 100% power by TMI-l Technical Specification 3.1.4.3 (operation at this level is not allowed for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ). The noble gas concentrations in the RCS were assumed to be at the maximum limit of 100/E-Bar specified by Technical Specification 3.1,4.1 (speci6c activity at this level requires hot shutdown within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ). Adjustment of Table 14.2-4 of the TMI-l UFSAR results in isotopic distribution shown on Table 2, attached.

lodine Partitioning The amount oflodine (1 )2 that is released from the RCS leakage to the secondary side and ultimately to the environment is calculated in Enclosure 3, Proprietary Polestar !

Calculation No. PSAT 05653 A.04, Rev,1. This is based on the following considerations l for the transpon of radiciodine into and through the OTSG: l 12 appearance as the result of partitioning as liquid flashes. This occurs due to the elemental iodine in the liquid which partitions to the gas phase. i 12 appearance as the result of evaporation-to-dryness of unflashed liquid.

12 appearance as the result of stripping ofliquid remaining in the steam generator.

12 deposition in the steam generator, primarily on the Inconel-600 tubes.

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ENCLOSURE 1 1920-98-20210 Page 4 of 20 l A decontamination factor (DF) for 12 can be calculated based on the pH and deWtion of 12 on the OTSG tubes. The results of the decontamination factor (DF) cab c for radiciodine (as 12) are shown on the attached Figures I and 2 for the AIS e d iS cases respectively Each Figure shows the RCS pressure and temperature, the flashing fraction, and the total release fraction as functions of time. Figure I shows that for the AIS the total release fraction briefly exceeds 25% between about 7000 seconds (about 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ) and about 30000 seconds (about 8.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months ), with an average of about 15% over the 2 -23.33 hour3.819444e-4 days 0.00917 hours 5.456349e-5 weeks 1.25565e-5 months period. For conservatism, a release fraction of 50% will be used for the first 10 minutes and 25% from 10 minutes to 23.33 hours3.819444e-4 days 0.00917 hours 5.456349e-5 weeks 1.25565e-5 months . Figure 2 shows that for the PAS, the total release fraction exceeds 25% for only the first 10 minutes then decreases over the duration of the transient. The flashing fraction decreases from about 45% to about 25% in the first 10 minutes of the transient, and then further decreases approximately linearly to near zero at 23.33 hours3.819444e-4 days 0.00917 hours 5.456349e-5 weeks 1.25565e-5 months . For conservatism, a release fraction of 50% will be used for the first 10 minutes and 25% from 10 mirutes to 23.33 hours3.819444e-4 days 0.00917 hours 5.456349e-5 weeks 1.25565e-5 months . These assumed release fraction values are greater than the flashing fraction for both cases.

Radiological Analysis The STARDOSE computer program is used to calculate the control room and offsite dose consequences. STARDOSE is proprietary to Polestar Applied Technology and is fully consistent with applicable Code of Federal Regulations and Regulatory Guides. The computer program is prepared under the Polestar Quality Assurance Program and intended for use in applications covered by Appendix B to 10 CFR 50 and 10 CFR 21. A STARDOSE model schematic is shown in attached Figure 3.

The TMl-1 control room habitability evaluation methodology and assumptions, including atmospheric dispersion factors for releases to the control room, are described in GPU Nuclear's letter to the NRC dated March 24,1998. The dispersion coefficients calculated for the Maximum Hypothetical Accident in the above referenced letter are conservatively used for the MSLB evaluation. The releases occurring from the MSLB accident will be diffused first through the Intermediate Building prior to being released to the environment. Any release from the Intermediate Building will be impacted by the total building complex blockage area including reactor building, turbine building, service building and intermediate building. Released activity will disperse rapidly into the building complex wake. Site geometry shows that distances to the control building ventilation system intake (yard intake) are greater for this release than assumed in the MHA analysis (112 meters for Intermediate Building releases compared to 91 meters for reactor building releases). This increased distance will result in lower X/Q values than those calculated if actual distances are utilized. Therefore it is considered a conservative approach to use X/Q values from the MHA since actual values will be lower due to increased blockage, initial building diffusion, and increased distance to the yard intake.

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ENCLOSUREI 1920-98-20210 Page 5 of 20 The atmospheric dispersion coef6cients utilized for the UFSAR Chapter 14 MSLB accident reanalysis are determined on a directional dependent basis with 6xed Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) boundary as indicated in existing TMI-l Technical Speci6 cation Section 5.1.1. Previous atmospheric dispersion analysis assumed a variable EAB boundary.

TMI-l Technical Speci6 cation Section 5.1.1 de6nes the EAB as a 2000 R. (610m) radius determined from the minimum distance in an easterly direction from the plant to the shore of the mainland. Figure 5-1 in the Technical Speci6 cations indicates the exclusion area as a stretched circle centered equidistant between the TMI-l and TMI-2 reactor buildings with circular radius centered at each unit's reactor building centerline equal to 2000 ft. This stretched circle configuration defmes the EAB from all directions and serves as the basis for determining EAB X/Q values using meteorological data from the site and Regulatory Guide 1.145 methods. The LPZ is defmed in Technical Speci6 cation l 5.1.1 as an area with a two-mile radius and is depicted in Technical Specification Figure 5-2. The X/Q values for the LPZ have also used Regulatory Guide 1.145 methods for determination. The new X/Q values have been determined by subtracting containment radius from the boundary radii in each of the sixteen directions evaluated.

The evaluation of accident X/Q values for TMI-l has also been updated to re6ect recent site meteorological data collected during four (4) years including 1992,1993,1995,1996 with data recovery above 90% and Regulatory Guide 1.145 methodology. This data is provided in Attachment I and is the latest available data. Less than 90% of the 1994 data was recovered, therefore, this year was excluded. The proposed accident X/Q values are based on a larger data base of meteorological data than the current TMI-l UFSAR accident X/Q values. This larger and updated database provides a more accurate representation of conservative accident X/Q values than previously utilized in design basis accident analysis. Data used in the analysis were collected on the site meteorological tower located at the nonhern end of Three Mile Island. The data used were hourly values of speed and direction from the 100 R. level and vertical temperature difference between 150 ft. and 33 R. Since wind speed measurements at the site are made at the 100 fl. level, they were adjusted to the standard 10 meter (33 ft.) level utilizing a power law relationship as a function of stability. The Pasquilt diffusion class was determined using vertical temperature difTerence (delta-T) and the categories given in NRC Regulatory Guide 1.23. Values of avand azused in the Regulatory Guide 1.145 dispersion equations were determined as a function of distance and stability class using the standard Pasquill-Gifford curves (Figures 1 and 2 of Regulatory Guide 1.145).

Pickard, Lowe & Garrick's (PLG) WINDOW code was used to perform the Regulatory Guide 1.145 calculations (Reference PLG report " Accident X/Q Values for TMI-1",

dated July 23,1998). It has been used since 1969 for calculating X/Q values in support of nuclear plant site evaluations. Values ofX/Q for the EAB were determined for each hour of the 4-year data base using the Regulatory Guide 1.145 equations. Cumulative probability distributions were made for each of the 16 direction sectors for the 4-year period. An envelope was constructed around all 16 direction-dependent curves and the 0.5% probable value (i.e., the value exceeded no more than 0.5% of the time) was

ENCLOSURE 1 1920-98-20210 Page 6 of 20 determined to be 8.0E-4 sec/m'. These results are shown in Figure 4. A second criterion required by the Regulatory Guide 1.145 procedure is that the 5% probable direction-independent X/Q value must be less than the 0.5% value for the direction-dependent case.

Inspection of Figure 6 shows the 5% value is 7.8E-4 sec/m' which is less than 8.0E-4 sec/m' direction-dependent value. According to the regulatory guide, the direction-dependent value must be used since it is higher. Table 4 summarizes the EAB results.

For the LPZ, the Regulatory Guide 1.145 procedure for different averaging times was used. The direction-dependent X/Q values at the LPZ for each direction are shown in Figure 5. As shown in Figure 7, the 1-hour LPZ 0.5% X/Q (4 years) was plotted on log-log graph paper at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and a straight-line was drawn to the annual average value at 8760 hours0.101 days 2.433 hours 0.0145 weeks 0.00333 months . Values at intermediate averaging times were taken from the straight-line connecting the two points. These values can be compared with more realistic 0.5%

probable averages computed by the WINDOW code shown as the lower line on Figure 7.

This comparison showed that the Regulatory Guide 1.145 technique is conservative by more than a factor of two for intermediate averaging times beyond 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The sector average equations in Regulatory Guide 1.111 were used aner the Grst 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The 5%

probable X/Q values at the LPZ for each averaging time are plotted on Figure 8. Results are summarized in Table 4.

Radiological Consequences The radiological consequences of the bounding integrated primary-to-secondary accident induced leakage with a maximum allowable del-131 limit of 1.0 Ci/gm are shown on the attached Table 3 for the PAS and AIS cases (Reference GPU Nuclear Calculation C-l 1101-900-E000-065, Rev. 0). The results show that for an MSLB with an assumed pre-accident iodine spike the calculated doses are well within the guideline values of 10 CFR Pan 100. For the MSLB with an assumed accident initiated iodine spike, the calculated l doses are slightly higher than the existing UFSAR, but remain a small fraction of the 10 CFR Part 100 limit. For both cases, the resulting 30 day control room doses are well below the 10 CFR Appendix A, GDC 19 limits and the NRC Standard Review Plan 6.4 guideline of 30 Rem, and remain bounded by the March 24,1998 TMI-l design basis Control Room Habitability Analysis. The results are also presented graphically on the proposed Technical Speci6 cation Figure 3.1-2b, which plots allowable DEI limit based on primary-to-secondary leakage (Reference GPU Nuclear Calculation C-1101-900-E000-071, Rev. 0). The primary-to-secondary leakage values shown represent projected l 2-hour leakage and total accident duration leakage based on OTSG tube inspection l results. The most restrictive leakage value establishes the maximum primary coolant l activity level for the subsequent operating cycle. A higher leakage will result in lower l DEI limit, or conversely, a lower leakage will allow a higher DEI up to a maximum allowable limit of 1.0 pCi/gm.

Therefore the proposed change in the TMI-l Technical Speci6 cation required RCS speci6c activity level does not adversely affect nuclear safety or safe plant operation.

ENCLOSUREI 1920-98-20210 Page 7 of 20 I

l Table 1 - RCS lodine and Noble Gas Activities for AIS l

RCS lodine Activities at 1 uCilg DEI l Adjusted Adjusted FSAR RCS ICRP 30 FSAR RCS FSAR RCS FSAR RCS i Isotopic DCF DEI Isotopic DEI (uCi/g)(*) (Rem /Cl)(b) W/@ (*) W/@ (#) W/@ (*)

, 1-131 5.71 E+ 00 1.08E+06 5.71 E+00 8.37E-01 8.37E-01 l l-132 1.92E+00 6.44E+03 1.14E-02 2.81 E-01 1.68E-03 l l-133 6.07E+00 1.80E+ 05 1.01 E+00 8.90E-01 1.48E-01 l l-134 7.57E-01 1.07E+03 7.47E-04 1.11 E-01 1.09E-04 l

! l-135 3.08E+00 3.13E+04 8.92E-02 4.52E-01 1.31 E-02 Totals 1.75E+ 01 6.82E+00 2.57E+00 1.00E+ 00 FSAR Table Adjusted FSAR Table to 1 uCi/g ISOTOPE 14.2-4 RCS DEI (uCi/g) 0 (uCi/g) #

KR-83M 5.30E-01 7.77E-02 KR-85M 2.43E+00 3.56E-01 KR-85 9.75E+00 1.43E+00 l

! KR-87 1.28E+00 1.88E-01 KR-88 3.95E+00 5.79E-01 XE-131M 2.68E+00 3.93E-01 XE-133M 4.22E+00 6.19E-01 XE-133 3.92E+02 5.75E+01 XE-135M 4.85E-01 7.11 E-02 l XE-135 8.37E+00 1.23E+00 XE-138 6.92E-01 1.01 E-01 l

l (a) lodine isotopic activities from Table 14.2-4 of the TMI-1 FSAR (1% failed fuel).

(b) lodine dose convension factors from ICRP 30.

(c) lodine isotopic activities Table 14.2-4 of the TMI-1 FSAR converted to dose equivalent I lodine 131 (DEI).

(d) todine isotopic activities from Table 14.2-4 of the TMI-1 FSAR scaled down to produce a l DEI of 1.0 uCl/g.

(e) Scaled down isotopic activities converted to dose equivalent I-131.

(f) Noble gas isotopic activities from Table 14.2-4 of the TMI-1 FSAR (1% failed fuel).

(g) Noble gas isotopic activities that maintain the same relative abundance as 1% failed fuel but scaled down to coincide with the lodines at 1.0 uCi/g del.

l l

l

. . . _. ~ __ _ - _. __ _ _ _ _ . _ _ _ . . _ _- _ . _ _ .. .. _ _

t-ENCLOSURE 1 1920-98-20210 l Page 8 of 20 Table 2 - RCS lodine and Noble Gas Activities for PAS RCS lodine Activities at 60 uCilg DEI Adjusted Adjusted FSAR RCS ICRP 30 FSAR RCS FSAR RCS FSAR RCS Isotopic DCF DEI Isotopic DEI (uCi/g)" (Rem /CI)" (uCi/g)" (uCi/g)# (uCi/g)*

l-131 5.71E+00 1.08E+06 5.71E+00 5.02E+01 5.02E+01 1-132 1.92E+00 6.44E+03 1.14E-02 1.69E+01 1.01E-01 l-133 6.07E+00 1.80E+05 1.01 E+00 5.34E+ 01 8.89E+00

! l-134 7.57E-01 1.07E+03 7.47E-04 6.66E+00 6.57E-03 i l l-135 3.08E+00 3.13E+04 8.92E-02 2.71 E+01 7.85E-01 Totals 1.75E+01 6.82E+00 1.54E+02 6.00E+01 ]

RCS Noble Gas Activity l

Ci FSAR y p Ebar Calc 100/Ebar l ,

Tin . RCS Avg Gamma Avg Beta (C,)(pp) ( RCS l

(min) (uCi/g)m (MeV/ dis) (MeV/ dis) (uCi/g)"

KR-83M 109.8 0.53 0.0026 0.0382 0.021624 0.539 KR-85M 264 2.43 0.152 0.243 0.95985 2.472 KR-85 5658000 9.75 0.00211 0.222 2.1850725 9.917 KR-87 76.2 1.28 1.42 1.05 3.1616 1.302 KR-88 168 3.95 1.74 0.34 8.216 4.018 XE-131 M 16992 2.68 0.271 0.137 1.09344 2.726 XE-133M 3252 4.22 0.0567 0.177 0.986214 4.292 XE-133 7590 392 0.0497 0.146 76.7144 398.721 XE-135M 15.6 0.485 0.429 0.098 0.255595 0.493 !

XE-135 548.4 8.37 0.248 0.316 4.72068 8.513

XE-138 14.4 0.692 N/A N/A N/A 0.704

!~ Total ====> 426.387 I(C)(pp)o i 98.314 433.697 Ebar ===> 0.231 100/ebar 433.7 i (a) lodine isotopic activities from Table 14.2-4 of the TMI-1 FSAR (1% failed fuel) l.

(b) lodine dose conversion factors from ICRP 30 (c) lodine isotopic activities Table 14.2-4 of the TMI-1 FSAR converted to dose equivalent fodine 131 (DEI)

(d) lodine isotopic activities from Table 14.2-4 of the TMI-1 FSAR scaled up to produce a DEI of 60 uCi/g (e) Scaled up isotopic activities converted to dose equivalent I-131 (f) Noble gas isotopic activities from Table 14.2-4 of the TMI-1 FSAR (1% failed fuel)

(g) Calculation of Ebar to determine 100/Ebar = 433.7 uCi/g for the FSAR mix.

l (h) Noble gas isotopic activities scaled up to produce a total activity of 433.7 uCi/g while maintaining the same relative abundance as 1% failed fuel.

I l

I ENCLOSURE 1 1920-98-20210 Page 9 of 20 Table 3 - MSLB Dose Results Accident Initiated Spike (AIS) Doses (Rem) 1 Thyroid Whole Body Skin Control Room Doses 3.32E+00 5.34E-04 4.1lE-03 EAB Doses 2.97E+01 9.96E-02 N/A LPZ Doses 2.54E+01 4.85E-02 N/A Note : For the MSLB with an accident initiated spike, the dose acceptance criteria for the EAB i

and LPZ are 2.5 Rem whole body and 30 Rem thyroid. The criteria for the control room are 5

{

Rem whole body and 30 Rem thyroid. l l

i I

l Pre-Accident Spike (PAS) Doses (Rem):

Thyroid Whole Body Skin Control Room Doses 5.02E+00 1.80E-03 1.81 E-02 l

EAB Doses 4.09E+01 1.88E-01 N/A l LPZ Doses - 1.20E+01 4.44E-02 N/A Note : For the MSLB with a pre-accident spike, the dose acceptance criteria for the EAB and LPZ are 25 Rem whole body and 300 Rem thyroid. The criteria for the control mom are 5 Rem whole body and 30 Rem thyroid.

ENCLOSUREI 1920-98-20210 Page 10 of 20 TABLE 4 ,

X/Q Calculations For TMI-1 EAB and LPZ

- O M . M Q at 5*/. MQ at 0.5*/. MQ Averaging Direction ~. RG1.145 5% X/Q at

" A EAB at LPZ Direction k"IC"I*I. Time (hours) (Toward) (sedm') LPZ (sedm')

Time (hours) (sedm'3) (sedm') (sedm')

1 8.0E-4 N 7.8E-4 1.4E-4 NE 1.4E-4 1.4E-4 0-2 2 5.6E-4 N -

9.8E-5 NE 1.4E-4 9.4E-5 2-8 8 - - -

4.8E-5 N** 6.0 E-5

4.6E-5 8-24 16 - - -

8.7E-6 WNW*

3.9E-5* 8.3E-6 24-96 72 - - -

4.4 E-6 WNW*

1.6E-5

4.2E-6 96-720 624 - - -

2.0E-6 N'* 4.0E-6* 1.8E-6 Annual Avg. 8760 - - - -

NE 7.8E-7 -

Interpolated from Figure 7 !

- Values in the NE direction would be lower i

ENCLOSURE 1 1920-98-20210 Page11of20 Figure 1 - M SLB-AIS - Cooldown and Release 1200 0.3 1000 .--

0.25

RC S Te mp 7 800 h. Flash Frac 0.2

}

a \ Tot Rel Frac c o

600 -

0.15 3 400 N -- 'N

- 0.1 200 -

0.05 I O I

, 0 0 .10000 20000 30000 40000 50000 60000 70000 80000 90000 Time (sec) l

1 l

ENCLOSURE 1 1920-98-20210 Page 12 of 20 l

I Figure 2 - M SLB-PAS - Cooldown and Release 1200 0.3 1000 RCS Pres 025

. RC S Temp 5 800 Flash Frac 0.2

\

1 Tot Rel Frac E c E 600 \ '

0.15 E

.o I

C \

m l I \ E r 400 N -1.

~

0.1 ,

i 200 - ' -

0.05 4

0 0 0 10000 20000 30000 40000 50000 60000 70000 80000 90000 Time (sec) t l

l

ENCLOSURE 1 1920-98-20210 Page 13 of20 Figure 3 - STARDOSE MODEL FOR TM-1 CONTROL ROOM DOSE ANALYSIS - MSLB 10 CFM 400105610 CRW" __ 9800 CRA TO EfMROMENT v A 90% FILTER 3.08 CFM (0-2 tws) '

I.6 TM(2-23.33 hrs) -- 400DO CRW*

p.g 1'<%

,c : " CRW*

CXM ROL BUKDNG 90 %

"* N 4 FILTER CXRE N (CONTROLROOM P(RCS)

For AIS - release framon is 0.5 (010 nts) and 0.25 (10 nts - 23.33 tus)

For PAS - release framon is 0.5 (0-10 nts) and 0.25 (10 nts - 23.33 tus)

" Frst value is the Ikw rate for the finit 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months prior to manuel adion to placesystemin redaAshon.

t i

I ,: trj i!i !! ,! ! : ;,h I!;

i f!t ,.

I .b L t!;[;

1 E1 0 E E W W W W R2 U02 f o 0

2 N E N N N N E E

sE E

E sssssWWWN NN WN s sWs S

O8914

- x , e e_ -

.- o- c.

L - _. *-

C0 2 eg N 9 a E1 P 2 0-E

'I,1

_ 0 0

yi 1

)

6 I 9

5 9

3 .

9 '

2 3 9 0-(

t a .I E

a D s 0 0

I e ' ;

)

M u . 1 T

f l

a _ i m

o V c s

r

_ e a

Q s e (

Y / =

4 X ,I Q

n o B 4 0-

/

X A

d e

s a E E _

B 0 _

s l ,

0 _

A iiIl e 1 l

V u

a ,

g_ -

Q

/

g - - -

X - -

m- - -

4 ; - - -

e _

r ; _ 5 u - -

i g

_ 0-F _

_r E 0

0 0

%0 1 1 0 0 0 1 0 0 0 1 0 0 0 1 1

o,a og yeO6eD.

ENCLOSUREI 1920-98-20210 Page 15 of 20 Figure 5 - X/Q Values Based on 4 Years of TMI Data (92,93,95,96)

LPZ X/Q Values 100.00% , ,

y, i

+.

-+- N

--m- NN E D

=* 10.00 % 4 A

+ NE a -w-ENE 3 .

I

-m-E (U -*-ESE L SE g 1.00 % m SSE

'E l -+- SSW g -e-SW O , - a-- W S W

_x_ w y' 0.10 %

t1' ,

_,_wnw M j , - - Nw

-~ ~ ~

_i i NNW i

0.01 %

1.00E-05 1.00E-04 1.00E-03 1.00E-02 X/Q (sec/m )

, ' , ',! . [i.! l

,. * , l i :

1 E1 0 0 R2 2 2 U02 fo 0-S - E O8 6

0 0

L 91- 1 C0 2eg N 9 a E1 P )

t a

a D

6 9

5 9

3 y 9 l 2

9 I

(

4-B E 3

A 8 0-E

. E 7 iI t

0 e !

0 ht lw 1 t

a Q

/

Y X *

)

t n /

m e c d e s

n (

e Q p I e X d ,I I

n n

i o

t 4 c 0-e E i

r .

0 D 0 1

r u

l o

I 1

l e

b , ,

a -

b .-

o r _

P .-

5 6 .-

e 5 r 0-u g E i 0

_ F %

0 0

%0 1 1 0 0 0 1 0 0 0 1 0 0 0 1 1

b5goE eo2E i

ENCLOSUREI 1920-98-20210 Page 17 of 20 Figure 7 - LPZ X/Q Values Based on 4 Years of TMI Data (92,93,95,96) 1.00E-03 i , ;-

~

I I

l

+-

l 1.4E-4 --+-Interpolated (0.5% Calcu;ated Maximum 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Value alues 1.00E b +l 6'OE-5I in the NE plotted at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , Calculated g ,

7-+-

d 9.8E-5 l~ ,, Maximum Average Annualin the NE p lli. T l 3.9E-5 } plotted at 8760 hours0.101 days 2.433 hours 0.0145 weeks 0.00333 months cdatM

} 4'8E-SI I i 0.5% 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

~

~ \ l

& Annual 1.6E-5 l Averages 3 1.00E-05 ^ _L - -+-Calculated -

\ .0E-6 v ages -

l 4.4E-6 ll l-0.5% Probable Calculated for Each l r 0E-7l [

Averaging Time in Direction with 2.0E-6l N .

1.00E Maximum Value ;

l7 i

1.00E-07 1 10 100 1000 10000 Averaging Time (hours) i

l l ENCLOSUREl 1920-98-20210 Page 18 of 20 Figure 8 - 5% Probable Direction Independent X/Q at the LPZ (92,93,95,96 Data) 100 00% ;-- __ _- , , ,

3 n '" % : _.

I i l 1 i 10.00 % -+

+ - i 7 l 4 _

l 1.8E-6 F 14.2E-6 18se-6l 74 6E-s IN-- 1 1.4E-4 l j !

! 9.4E-5 l -

I'\ +1 hour m d- -~

+2 hour "

$ l +8 hour ;

b +16 hour j II h _ +72 hour

[ ! +624 hour 0.10% b !

-r t 9 g_._._ !

_d_

0.01 %

1.0E-06 1.0E-05 1.0E-04 1.0 E-03 XIQ (see m')

- . . ~ -- .- - .- . - . . -

ENCLOSURE 1 1920-98-20210 Page 19 of 20

! - IV. ' No Significant Hazards Consideration GPU Nuclear has determined that this Technical Speci6 cation Change Request poses no significant hazards consideration as defined by 10CFR50.92.

I. Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability of occurrence or the consequences

! of an accident previously evaluated. The proposed amendment has no affect on structures, systems or components. The existing steam line break criteria are maintained. This change only accounts for radiological consequences resulting from a revised maximum allowable reactor coolant system (RCS) speciGc activity l- limit of 1.0 pCi/gm. The new radiological consequences of the revised MSLB l

accident, which also incorporate more conservative values for atmospheric dispersion, are below 10CFR100 limits and 10CFR50, Appendix A, GDC-19

, limits for the control room. The use of revised atmospheric dispersion factors for l other TMI-l accident analyses is addressed in a separate license amendment l

request submittal. Therefore, this activity does not involve a significant increase in the probability of occurrence or the consequences of an accident previously evaluated.

2. Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any previously evaluated. The proposed amendment has no impact on any plant structures, systems or components OTSG tube structuralintegrity is maintained. Therefore, this activity does not create the possibility of a new or diflerent kind of accident from any previously evaluated.

l

3. Operation of the facility in accordance with the proposed amendment would not

involve a significant reduction in a margin of safety. The proposed amendment l has no impact on stmetures, systems, or components. OTSG tube inspection criteria are not afTected by this change. OTSG tube structural integrity is maintained. The existing TMI-l Technical Specification Section 3.1.4.1 Bases state that the limitations on the specific activity of the primary coolant ensure that

the resulting 2-hour doses at the site boundary will be well within 'the Part 100 i limit following associated design basis accidents postulated in conjunction with an assumed steady state primary-to-secondary steam generator tube leakage of 1.0 gpm. This margin of safety is preserved since resulting dose consequences incorporating more conservative values for atmospheric dispersion remain well l

within the Part 100 limit. Therefore, this activity does not reduce the margin of

safety.

e V. Imniementation GPU Nuclear requests that the amendment authorizing this change become effective immediately upon issuance.

ENCLOSUREI 1920-98-20210 Page 20 of 20 ATTACIIMENT 1 1992,1993,1995,1996 Meteorological Data (Supplied by Electronic Disk)

l ENCLOSURE 2 Certificate of Service for TMI-I Technical Specification Change Request 272 i

1 l

1

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION IN Tile MATTER OF DOCKET NO. 50-289 GPU NUCLEAR INC. LICENSE NO DPR-50 CERTIFICATE OF SERVICE This is to certify that a copy of License Amendment Request No. 272 to Appendix A of the Operating License for Three Mile Island Nuclear Station Unit 1, has, on the date given below, been filed with executives of Londonderry Township, Dauphin County, Pennsylvania; Dauphin County, Pennsylvania; and the Pennsylvania Department of Environmental Resources, Bureau of Radiation Protection, by deposition in the United Sates mail, addressed as follows:

Mr. Darryl LeHew, Chairman Ms. Sally Klein, Chairman Board of Supervisors of Board of County Commissioners of Londonderry Township Dauphin County R.D. #1, Geyers Church Road Dauphin County Courthouse Middletown, PA 17057 Front & Market Streets Harrisburg, PA 17101 Director, Bureau of Radiation Protection PA Dept. of Environmental Resources Rachael Carson State Office Building PO Box 8469 Harrisburg, PA 17105-8469 Att: Mr. Stan Maingi GPU NUCLEAR INC.

BY: hudd addj h Mce Pre'sEdent and irecfor,'TMI DATE: /6//9/99 i *

ML20154P866

Text

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