ML20198L813

Proposed Tech Specs Pages Revising TS 3.4.8, Specific Activity, Figure 3.4-1,Table 4.4-4 & TS Bases 3.4.8

ML20198L813
Person / Time
Site:	Braidwood
Issue date:	01/14/1998
From:	COMMONWEALTH EDISON CO.
To:
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ML20198L802	List: ML20198L798 ML20198L813 ML20198L818 ML20198L828
References
NUDOCS 9801160128
Download: ML20198L813 (20)

Text

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ATTACHMENT B MARKED UP PAGES FOR PROPOSEC CHANGES TO Ai'PENDIX A TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES  !

NPF-72, AND NPF-77 '

9 BRAIDWOOD STATION UNITS 1 and 2 REVISED PAGES:

3/4 4 27 '

3/4 4-28 3/4 4-29 3/4 4-30*

3/4 ~4-31 B 3/4 4-5 B 3/4 4-6*

B 3/4 4 7*

• This page has no changes but is included for continuity. ,

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9901160128 990114 POR ADOCK 05000456-P PDR . '

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• REACTOR CODLANT SYETEM 1/4.4.8 SPECIFIC ACTIVITY l LINITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited to:

a. Less than or equal to 1 microcu-le per gram DOSE EQUIVALENT ,

I-131**, and I

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b. Less than or equal to 100/E microcuries per gram of gross ,  ;

)Adioactivity. <

1 APPLICABILITY: MODES 1, 2, 3, 4, and 5.

AGI1ON:

MODES 1, 2 and 3*:

a. With the specific activity of the reactor coolant greater than i 1 Sicrocurie per gram DOSE EQUIVALENT I-131** for more than i 44 hours5.092593e-4 days <br />0.0122 hours <br />7.275132e-5 weeks <br />1.6742e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least H0T STAND 8Y with T ,

less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; and

b. With the specific activity of the reactor co614nt p eater than ,

100/E microcuries per gram, be in at least HOT STAND 8Y with T ,

less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

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' 'Nith T mgregterth 1 to 500*f. g,og

• • For Unit It TETfr coblant DOSE EQUIVALENT I-13) will be limited to er9fF microcuries per gram.

BRAIDWOOD - UNITS 1 & 2 3/4 4-27 AMENDMENT NO. 69 f

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1 REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION ACTION (Continued) l l

MODES 1, 2, 3, 4, and 5: j With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131* or greater than 100/E microcuries per gram, perform the sampling and analysis requirements of  ;

Item 4.a) of Table 4.4-4 until the specific activity of the reactor '

coolant is restored to within its limits.

SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the reactor coolant shall be deterwined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.

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• For Unit It reactor coolant DOSE EQUIVALENT I-131 will be limited to 4h46-microcuries per gram.

. BRAIDWOOD - UNITS 1 & 2 3/4 4-28 ~ AMENDMENT NO. 69 t

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20 m 40 m so 70 ao 90 100 PERC[NT or RATED THERW POWER FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR C001. ANT SPECIFIC ACTIVITY LIMIT VEP. SUS PERCENT OF RATED THERhAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >lyCI/ GRAM DOSE EQUIVALENT I-131* l

• For Unit 1, Reactor Coolant Specific Activity >0.35 pCi/Gran DOSE EQUIVALENT I-131 BRAIDWOOD - UNITS ] & I 3/4 4-29 AMENDMENT NO. 69

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20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT l-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > inci/ GRAM DOSE EQUlVALENT l.131'

• For Unit 1 through Cycle 7, Reactor Coolant Specibe Actmty > 0 05 pCvGram DOSE EQUIVALENT l-131

., CRAIDWOOD - UNITS 1 & 2 3/4 4-29 AMENDMENT NO.

TA8tE 4.4-4 ,

REACTOR COOLANT SPECIFIC ACTIVITY SAPFLE N ANALYSIS PemS_".

TYPE OF MEASUREMENT W ANALYSIS SAhPLE AND ANAL'fSIS FREQUENCY MODES IN WHICH SAMPLE AfW ANALYSIS REGUIRED

1. Gross Radioactivity Determination **- At least once per 77 'ours I, 2,-3, 4

2. Isotopic Analysis for DOSE EQUIVA- Once per 14 days I

!ENT l-131 Concentration

3. Radiochemical for E Determination"*

i Once per 6 months

• 1

4. Isotopic Analysis for Iodine including I-131, I-133, and 1-135 a) Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, if, 2f, 3f, 4f, Si whenever the soecific activity exceeds I pCl/ gram DOSE EQUIVALENT I-131****

or 100/E pCl/ gram l of gross radioactivity, and

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b) One sample between 2 1, 2, 3

and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a THElWIAL POWER change exceeding 15% of the RATED THERMAL POWER within a 1-hour period.

t BRAIDWOOD - UNITS 1 & 2 3/4 4-30 AMENDMENT NO. 69

• TABLE 4.4-4 (Continued)

TABLE NOTA 110tLi

1. Until the specific activity of the Reactor Coolant System is restored within its limits.

• Sample to be taken after a minimum of 2 EFPD and 20 days of POWER.

OPERATIO,1 have elapsed since reactor was last subcritical 'or 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.

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• • A gross radioactivity analysis shal1 consist of the quantitative  !

measurement of the total specific activity of the reactor coolant except i for radionuclides with half-lives less than 10 minutes and all i radiciodines. The total specific activity shall be the sum of the degast,ed beta-gamma activity and the total of all identified gaseous '

activities in the sample within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the sample is taken and extrapolated back to when the sample was taken. Determination of the contributors to the gross specific activity shall be based upon those energy peaks identifiable with a 95% confidence level. The latest i.vailable data may be used for pure beta-emitting radionuclides.

• • • A radiochemical analysis for E shell consist of the quantitative measurement of the specific activity for each radionuclide, except for radionuclides with half-lives less than lo minutes and all radiciodines, which is identified in the reactor coolant. The specific activities .for these individual radionuclides shall be used in the determination of E.,

for the reactor coolant sample. Determination of the contributors to E shall be based upon these energy peaks identifiable with.a 95% confidence n e,- L C ele. 7 o .oS

• • • • For Unit 1, reac or coolant DOSE EQUIVALENT I-131 will be limited to 9:$9-microCuries per gram.

BRAIDWOOD - UNITS 1 & 2 3/4 4-31 AMENDMENT NO. 69 e

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_ REACTOR COOLANT SYSTEM '

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~ OPERATIONAL LEAKAGE (Continued)-

5 The Surveillance Requirements for RCS pressure isolation valves provide l added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS pressure f isolation-valves is. IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

3/4.4.7- CHEMISTRY The limitations on Reactor Coolant System chosistry ensure that corrosion of the Reactor Coolant System is minimized and reduces-the potential for i Reactor Coolar.t System leakage or failure due to stress corrosion.

the chemistry within the Steady-State Limits ~ provides adequate corrosionMaintaining i protection to ensure the structural, integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion ,

studies show that operation may be. continued with contaminant concentration '

- levels in excess of the Steady-State Limits, up to the Transient' Limits, for the specified limited time intervals without having a significant effect on the-structural integrity of the Reactor Coolant System. The time interval .

permitting concinued operaticn within the restrictions of. the Transient Limits provides time for taking corrective actions to restore the contaminant concen-trations to within the. Steady-State Limits.

The Surveillance Requirements provide adequate assurance.that concentrations  !

in excess of the limits will be detected in sufficient time to take corrective action.

3/4.4.8 SPECIFIC ACTIVITY c

The limitations on the specific activity of the reactor coolant ensure that the resulting 2-hour doses at the SITE BOUNDARY will not exceed an 1 appropriately small fraction of 10 CFR Part 100 dose guideline values following a steam generator tube rupture accident in conjunction with an assumed steady-state reactor-to-secondary steam generator leakage rate of 1 gpm. The values for the-limits on specific activity represent limits based upon a arametric  !

evaluation by the NRC of typical site locations.

These values are conservative  :

in that specific site parameters of the Braidwood Station, such as SITE BOUNDARY l

location and noteorological conditions were not considered in this evaluation.,

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. BRAIDWOOD / UNITS-1 L 2- B 3/4 4-S: - AMENOMENT NO.10 4

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insert B For Unit 1 through Cycle 7, the limitations on the specific activity of the reactor coolant ensure that the resultng 2-hour off-site doses will not exceed an sporopriately small fraction of the 10 CFR Part 100 doss guideline values following a Main Steam Line Break accident in conjunc' ion with an assumed l steady-state primary-to-secondary steam generator leakage rate of 150 gpd from r each of the unfaulted steam generators and a maximum site allowable primary-to-secondary leakage from the fauited steam generator.

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SPECIFIC ACTIVITY (Continued) i The sample analysis for determining the gross specific activity and E can i exclude the radiciodines because of the low reactor coolant limit of 1 microcurie / I gram DOSE EQUIVALENT I-131, and because, if the limit is exceeded, the radiciodine j level is to be determined every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. If the gross specific activity level and radiolodine level in the reactor coolant were at their limits, the radioiodine contribution would be approximately IL In a release of reactor coolant with a typical mixture of radioactivity, the actual radiciodine contribution would be about 20L The exclusion of radionuclides with half-lives less than 10 minutes from these determinations has been made for several reasons. The first consideration is the difficulty to identify short-lived radionuclides in a sample that .vquires a significant time to collect, transport, and analyze.

The second consideration is the predictable delay time between the postulated release of radioactivity from the reactor coolant to its release to the environ-ment and transport to the SITE BOUNDARY, which is relatable to at least 30 minutes decay time. The choice of 10 minutes for the half-life cutoff was made because of the nuclear characteristics of the typical reactor coolant radioactivity. The radionuclides in the typical reactor coolant have half-lives of less than 4 minutes or half-lives of greater than 14 minutes, which allows a distinct window for determination of the radionuclides above and below a half-life of 10 minutes. For these reasons the radionuclides that are excluded from consideration are expected to decay to very low levels before they could be transported from the reactor coolant to the SITE BOUNDARY under any accident condition.

Based upon the above considerations for excluding certain radionuclides from the sample analysis, the allowable time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> between sample taking and completing the initial analysis is based upon a typical time necessary to perform the sampling, transport the sample, and perform the analysis of about 90 minutes. After 90 minutes, the gross count should be made in a reproducible geometry of sample and counter having reproducible beta or gamma self-shielding properties. The counter should be reset to a reproducible efficiency versus energy. It is not necessary to identify specific nuclides. The radio-chemical determination of nuclides should be based on multiple counting of the sample with typical counting basis following sampling of less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, about I day, about I week, and about 1 month.

Reducing T,yg to less than 500*F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the reactor 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 reactor coolant will be detected in sufficient time to BRAIDWOOD - UNITS 1 & 2 B 3/4 4-6 AMENDMENT NO. 10 L v

' REACTOR COOLANT SYSTEM I

i BASES  ;

SPECIFIC ACTIVITY (Continued)  :

take. corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomenon.

A reduction in frequency of. isotopic analyses following power changes may be permissible if 1 justified by the data obtained.

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- 3/4.4.9 PRESSURE / TEMPERATURE LIMITS

- The temperature and pressure changes during heatup and cooldown'are limited to be consistent with the requirements given in the ASME Boiler and

- Pressure Vessel Code,Section III, Appendix G:-

1. The reactor _ ;;alant temperature and pressure and system heatup and t

cooldown rates (with the exception of the_ pressurizer) shell be-limited in accordance with figures 3.4-2a (3.4-2b) and 3.4-3a (3.4-3b) for the service period specified thereon:

a. Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the -

limit lines r.hown. Limit lines for cooldown rates between those presented may be obtained by interpolation; and

b. Figures 3.4-2a (3.4-2b) and 3.4-3a (3.4-3b) define limits to assure prevention of non-ductile failure only. For normal .

operation, other inherent plant characteristics, e.g. , pump -

heat addition and pressurizer heater capacity, may limit the ,

heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

2. These limit lines shall be calculated periodically using methods provided below,

3. The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70*F,

4. The pressurizer heatup and cooldown rates shall not exceed .100*F/hr and 200* F/hr respectively. The spray shall not be used if the.

temperature difference between the pressurizer and the spray fluid is. greater than 320'F, and

5. System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of.ASME Boiler and Pressure Vessel Code,Section XI.

The fracture toughness properties of the ferritic materials in the reactor  ;

vessel are determined in accordance with the 1973 Summer Addenda to Sec ,

of the ASME Boiler;and-Pressure Vessel and Code.

B 3/4 4-7 AMENDMEM NO. 30 I-' BRAIDWOODfUNITS .1 & 2

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ATTACHMENT C EVALUATION OF GlGNIFICANT HAZARDS CONSIDERATIONS FOR PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-72 AND NPF-77 Commonwealth Edison (Comed) has evaluated this proposed amendment and determined that it involves no significant hazards considerations. According to Title 10 Code of Federal Regulations Section 50 Subsection 92 Paragraph c (10 CFR 50.92 (c)), a proposed amendment to an operating license involves no significant hazards considerations if operation of the facility in accordance with the proposed amendment would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or

3. Involve a significant reduction in a margin of safety.

A. INTRODUCTION Comed proposes to revise Braidwood Technical Specification (TS) 3.4.8,

" Specific Activity," Figure 3.4-1, Table 4.4 4 and Technical Specification Bases 3.4.8 for Braid'uood Unit 1. This revision willlower the Unit 1 Reactor Coolant System (RCS) Dose Equivalent (DE) lodine 131 (1-131) level from 0.35.

i microCuries per gram (pCi/gm) to 0.05 pCi/gm. This revision will also lower the RCS DE l-131 activity limit in TS Figure 3.4-1. These revisions will remain in effect for the remainde of Unit 1 Cycle 7. At the completion of Braidwood Unit 1 C:rcle 7, Comed will be replacing the original Westinghouse Model D-4 Steam Cenerators, allowing the TS RCS DE l-131 activity limit to be returned to the standard value of 1.0 pCi/gm.

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' This change is required in order to provide additional margin to the maximum allowable primary-to-secondary leakage limit. The total potential leakage includes primary to-secondary leakage from indications remaining in service in

j. the faulted steam generator due to the application of the approved Interim

- Plugginr Criteria, F* criteria, and 150 gallon per day (gpd) leakage from each of

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the three unfaulted steam generators.

B. NO SIGNIFICANT HAZARDS ANALYSIS

1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Generic Letter 95-05, " Voltage-Based Repair Criteria For Westinghouse Steam Generator Tubes Affected By Outside Diameter Stress Corrosion Cracking,"

' allows lowering of the RCS DE l-131 activity as a means for accepting higher projected leak rates if justification for equivalent 1-131 below 0.35 pCi/gm is -

provided. Four methods for determining the impact of a release of activity to the public were reviewed to provide this justification. These four methods are as follows:

Method 1: NRC NUREG 0800, Standard Review Plan (SRP) Methodology Method 2: Methodology described in a report by J.P. Adams and C.L. Atwood, "The lodine Spike Release Rate During a Steam Generator Tube Rupture," Nuclear Technology, Vol. 94, p. 361 (1991) using Braidwood Station reactor trip data.

Method 3: Methodology described in a report by J.P. Adams and C.L. Atwood, "The lodine Spike Release Rate During a Steam Generator Tube Rupture," Nuclear Technology, Vol. 94, p. 361 (1991) using normalized industry reactor trip data.

Method 4: Methodology described in a draft EPRI Report TR-103680, Revision 1, November 1995, " Empirical Study of lodine Spiking in PWR Plants".

! The effect of reducing the RCS DE l-131 activity limit on the amount of activity released to the environment remains unchanged when the maximum site allowable primary-to-second'ary leak rate is proportionately increased and the i.

iodine releas e rate spike factor is assumed to be 500 in accordance with the SRP. With an RCS DE l-131 activity limit of 1.0 pCi/gm, the maximum site allowable leakage limit was calculated, in accordance with the NRC SRP methodology, to be 6.64 gpm at room temperature and pressure. Comed has evaluated the reduction of the RCS DE l-131 activity to 0.05 pCi/gm along with the increase of the allowable leakage to 132.8 gpm at room temperature and pressure and has concluded:

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the offsite dose, including the iodine spiking factor, will be less than the l 10CFR100 limits.

A Based on the NRC SRP methodology for dose assessments and assuming the  ;

iodine spike factor of 500 is applicable at the new 0.05 pCilgm RCS DE l-131  :

activity limit, the Control Room dose, the Low Population Zone dose, and the dose at the Exclusion Area Boundary continue to satisfy the appropriately small _  ;

fraction of the 10CFR100 dose limits. ,

- An evaluation of the Control Room dose, attributed to an MSLB accident '

concurrent with steam generator primary-to-secondary leakage at the maximum site allowable limit, was performed in support of a license amendment request for application of a 1.0 volt Interim Plugging Criteria. This evaluation concluded that -

- the activity _ released to the environment during an eight (8) hour time period from an MSLB accident (812 Curies for a Pre-accident iodine spike and 888 Curies for i

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an accident-initiated iodine spike) is bounded by the activity released to the -

, environment from the Loss of Coolant design basis accident (1290 Curies).- '

Therefore, the Control Room dose, due to the MSLB accident scenario, is  ;

bounded by the existing Loss of Coolant Accident (LOCA) analysis. The .

_ maximum site allowable primary-to-secondary leakage is limited by the offsite '

dose at the Exclusion Area Boundary due to an accident-initiated spike.

The. report by J.P. Adams and C.L. Atwood, "The lodine Spike Release Rate During a Steam Generator Tube Rupture," Nuclear Technology, Vol. 94, p. 361 (19g1), concluded that the NRC SRP methodology, which specifies a release

. rate spike factor of 506 for iodine activity from the fuel rod to the RCS, is ,

conservative when the RCS DE l 131 concentration is greater than 0.3 pCi/gm.

In order to evaluate whether a release rate spike factor of 500 is conservative below 0.3 pCi/gm, actual operating data from the previous reactor trips of Braidwood Units 1 and 2, with and without fuel defects, were reviewed and analyzed using the methodology presented in Section ll.C of the Adams and Atwood report (Method 2). The same five data screening criteria described in the Adams and Atwood report were applied to the Braidwood data to ensure consistency and validity when comparing the Braidwood results to the data in the Adams and Atwood report. Of the reactor trip events at Braidwood Units 1 and 2, seventeen (17) met the five data acreening criteria.

i Seven (7) of the seventeen (17) Braidwood trips occurred during cycles with no-fuel defects. In all seven of these instances, the calculated spike factor was l

much less than the spike factor of 500 assumed in the NRC SRP methodology. l Braidwood Unit 1 Cycle 7 is currently operating with no fuel defects and an RCS 4 DE l-131. activity of approximately 3E-4 pCi/gm. The seven previous trips with  ;

no fuel defects _ had steady state iodine values that are reasonably close to the current operating conditions.' It is therefore reasonable to conclude that, y

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i assuming continued operation with little to no fuel defects, the calculated spike factors from these events would reflect an actual event for Unit 1 Cycle 7, i.e. the spike factor will be less than 500.

Since some of the Braidwood spike factors were greater than 500 when the RCS DE l-131 activity prior to the accident was less than 0.3 pCl/gm, Comed examined the conservatisms in the current release rate calculation. The primary reason for the high spiking fcctors contained in the Adams and Atwood report (up to 12,000), is not because the absolute post-trip release rate is high (factor numerator), but rather because the steady-state release rate (factor denominator) is low. The Braidwood specific data resulted in six (6) events with a calculated release rate spike factor greater than 500, it is not expected based upon the Unit 1 Cycle 7 fuel conditions that a spiking factor greater than 500 would occur. The revised RCS DE l-131 activity limit will also ensure that the operating cycle will not continue if significant fuel defects develop.

In order to evaluate the Braidwood specific data against the NRC SRP methodology, the release rate for a steady state RCS DE l-131 activity of 1.0 pCi/gm was calculated. Using the Braidwood specific data, the pre-trip steady-state release rate is 27.5 Cl/hr. Using a release rate spike factor of 500 for the accident initiated spike, the post trip maximum release rate would be 13,733 Ci/hr (SRP Methodology). The highest post trip lodine release rate from the Braidwood trip data, Event 15, was 1335 Cl/hr, it is important to remember that this number is determinor, by conservatively increasing the post-trip RCS De I-131 activity by a factor of three (3), in accordance with the Adams and Atwood report, The purpose of this amendment request is to reduce the TS RCS DE l-131 limit by a factor of twenty as compared to the original TS RCS DE l-131 limit of 1.0 pC /gm. By decreasing the TS RCS DE l 131 activity by a factor of twenty the maximum iodine release rate is 686.7 Ci/hr, (13,733 Cl/hr divided by 20). Two (2) of the seventeen (17) Braidwood data points exceed this value. Both occurred during cycles with fuel defects. Braidwood Unit 1 is currently operating with no fuel defects. Fifteen (15) of the 168 data points 'n the Adams and Atwood report exceed 686.7 Ci/hr. For the combined database of 185 data points, of which seventeen (17) exceeded 686.7 Ci/hr, only two (2) of these seventeen (17) data points had a pre-trip RCS DE l-131 activity below 0.05 pCi/gm. The 95% confidence prediction for the combined data sets bounded one (1) of these two (2) data points. This data indicates that the possibility for a post-trip iodine fuel release rate to exceed 686.7 Ci/hr, when the pre-trip RCS DE l-131 concentration is at or below 0.05 pCi/gm, is small. The conservatisms mentioned in the following sections will reduce the possibility of exceeding a small fraction of the 10 CFR100 limits should a fuel release greater than 686.7 Ci/hr occur.

. K:sgrp\mfsiiodine2. doc C-4

4 jf the Braidwood dat1 were plotted with the Adams and Atwood data, the conclusions of the Adams and Atwood report would not be compromised. Where the Braidwood data contains spike factors greater than 500, the RCS DE l-131 concentrations are below 0.05 mCilgm Since the Braidwood data includes very few data points near 0.05 pCl/gm (the requested new TS limit), it is appropriate to use the Braidwood database combined with the Adams and Atwood database near 0.05 pCi/gm to determine if a spike factor of 500 is appropriate. The combined databases contains seventy-nine (79) data points with a Pre-Trip RCS DE l-131 activity between 0.01 pCi/gm and 0.10 pCl/gm. Sixty two (62) of these seventy nine (79) data points (78 %) have spike factors less than 500. Using the entire Braidwood database combined with the Adams and Atwood database,141 of the 185 data points (70%) have an iodine spike factor less than 500.

Therefore, it is reasonable to assume thM 3 spike factor of 500 would not be exceeded for a majority of the events if an MSLB accident were to occur while the RCS DE l-131 activity is at or below 0.05 pCi/gm. The highest spike factor seen in the Adams and Atwood report near a Pre-Trip RCS DE l-131 activity of 0.05 pCi/gm was 773 (at 0.05 pCl/gm). The corresponding release rate for this event was 368 Cl/hr which is less than the calculated Braidwood maximum release rate of 686.7 Ci/hr.

The predominant factors in calculating the offsite dose are the post-trip iodine release rate from the fuel and the flowrate at which the activity is being released to the environment, not whether the spike factor is greater than or less than 500.

The post trip DE l 131 release rate will determine the level of activity in the RCS that will be released. The flowrate will determine at what rate this activity is released to the environment. Method 3, which used an approach in the Adams and Atwood report, concluded that, at a 95% confidence of a SS percentiie, the post-trip iodine release rate was bounded by 0.608 Ci/hr MWe. For Braidwood r Station, which has a MWe rating of 1175, the post-trip iodine release rate, at a 95% confidence of a 85 percentile, should not exceed 714 Ci/hr. Two (2) of the seventeen (17) reactor trips from Braidwood exceeded 714 Ci/hr. These two (2) reactor trips had post-trip iodine reicase rates of 1335 Ci/hr (spike factor of 3471) and 802 Ci/hr (spike factor of 1483). Both occurred during cycles with fuel defects. Braidwood Unit 1 is currently operating with no fuel defects.

In the fourth method, the results from a Draft Electric Power Research Institute 1

i (EPRI) Report TR-103680, Rev.1, November 1995, " Empirical Study of lodine Spiking in PWR Power Plants" were applied. The objective of the EPRI study was to quantify the iodine spiking in a postulated Main Steam Line Break /

Steam Generator Tube Rupture (MSLB/SGTR) accident sequences, in the EPRI report, an iodine spike factor between 40 and 150 was determined to match data

from existing plant trips. The maximum iodine spike factor value of 150 was 1

applied to a steady state equilibrium RCS DE l-131 activity of 0.33 pCi/gm. The

- K:sgrpimfsW@ne2. doc C3

resulting two-hour average iodine concentration for a postulated MSLB/SGTR j accident sequence was determined to be 3.1 pCi/gm. Since the EPRI report is based on industry data and the EPRI method predicted a post-accident indsne l activity, which is a small fraction of the activity predicted by the NRC SRP l

methodology, it can be expected that, for the proposed 0.05 pCi/gm limit under l an MSLB/SGTR accident sequence, the post-accident iodine activity would l

typically be a small fraction of the RCS DE l-131 activity predicted by the NRC SRP methodology. For Braidwood, using the SRP methodology with an RCS DE l 131 activity of 1.0 pCl/gm and a spike factor of 500, the Post Trip RCS activity two ho'Jrs after the event would be near 38 pCl/gm. At an RCS DE l 131 activity i of 0.05 pCi/gm, it would require a spike factor of nearly 10,000 to obtain a Post-l Trip RCS DE l 131 activity near 38 pCi/gm. With a Post-Trip RCS DE-l 131 1 activity of 38 pCi/gm, an increase in the allowable leak rate could impact the 10CRF100 limits. To accommodate for an increase in the allowable leak rate by a factor of twenty, the resultant activity would need to be below 1.9 pCl/gm. Two (2) of the seventeen (17) post trip data points from Braidwood exceeded 1,9 pCilgm. Bott occurred during cycles with fuel defects. Braidwood Unit 1 is currently operating with no fuel defects. The conservatisms mentioned below will reduce the possibility of exceeding a small fraction of the 10 CFR100 limits should the post-trip lodine exceed 1.9 pCilgm.

Based on evaluations by the four methods above, Braidwood can conclude that the current methodology (Method 1) used to predict iodine spiking is conservative. Although dose projections indicate with confidence tnat the iodine spiking factor limit will be met, the conservatisms in the offsite dose calculatior.

and current Braidwood Unit 1 operating cr ditions listed below, provide added assurance that the 10CFR100 limits, General Design Criteria (GDC) 19 criteria, and the requirements of NRC Generic Letter 95-05 will be satisfied if the iodine spike factor exceeds 500 or the pcst-trip fuel release rate exceeds 686.7 Ci/hr.

As further assurance that the 10CFR100 and GDC 19 limits are not exceeded, several conservatisms are inherent to the offsite dose calculation. These conservatisms include, but are not limited to:

1. The meteorological data used is at the fifth percentile, it is expected that the actual dispersion of the iodine would result in less exposure at the site boundary than the 30 Rem limit of 10CFR100.

2. lodine partitioning is not accounted for in the faulted SG. With the high pH of the secondary water, some partitioning is expected to occur. An iodine partition factor of 0.1 is more realistic ( per Table 15.1-3 of Reference 8) than the 1.0 valued (no partitioning) used in the offsite dose cWulation. This reduces calculated dose by 905 3, The activity in the RCS is not expected to increase instantaneously with the spike in iodine released from the defective fuel.

. K:sgrpimfsWWine2. doc C-6 y

e '

i 4." The results from the Braidwood tube pull dat, indicate that the projected .  !

Interim Plugging Criteria leak rate is conservative.  :

In addition, the current Braidwood ' Unit 1 operating conditions provide defense in depth and provide further assurance that the 10CFR100 and GDC 19 limits will l

not be exceeded:

1. Braidwood Unit 1 is currently operating with a debris resistant fuel design l!

which is less likely to develop fuel defects.

2. As evidenced by industry data, if debris related fuel failures are going to -

occur they are most likely to be occur early in the cycle. Braidwood Unit.1 has operated approximately 6 months into its current cycle and has seen  ;

no signs of fuel defects. Therefore. fuel failure prior to completion of the current cycle is not likely.

3.- The RCS DE l 131 activity is likely to be less than the TS limit. With the  ;

current Braidwood Unit 1 RCS DE l-131 activity near 3E-4 pCi/gm with no fuel defects, the spike factor is expected to be considerably smaller than- i the 500 value. .

4; it is unlikely, for the short time period this amendment is being requested (remainder of Cycle 7), that an accident-initiated lodine spike for Braidwood Unit 1 would be greater than the NRC SRP assumed value.

5. Primary-to-secondary leakage is likely to be less than the TS limit (150 i gpd) in each of the four SGs prior to the event. Currently, minimal primary-to secondary leakage (less than 5 gpd) exists at Braidwood Unit '

1.  ;

These proposed changes do not result in a significant increase in the '

consequences of an accident previously analyzed.

The RCS DE I.131 sctivity limit is not considered as a precursor to any accident.  :

Therefore, this proposed change does not result in a significent increase in the probability of an accident previously analyzed. '

2. The proposed change does not create the possibility of a new or U differant kind of accident from any accident previously evaluated.

1 The changes proposed in this amendment request conservatively reduce the -

Unit 1 RCS DE l-131 activity limit at which action needs to be taken. The changes do not directly affect plant _ operation. 'These changes will not result in

the installation of any new equipment or systems or the modification of any L existing equipment or systems. . No' new operating procedures, conditions or configurations will be created by this proposed amendment.

Q  ; K:igrpimfs\ iodine 2. doc C-7 I i

-%r' - , -c '.

- - ~.r, - w m - = = =+- y n,ev m v +f-trr m--wemw ev t- --sa w w--e r rp e uwi--'- - .a.--r+, qw w-,'v-- Uwtm -y we .-y

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

4 <

Accordingly, this proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction in a  !

margin of safety. '

NRC Generic Letter 95-05 allows lowering of'the RCS dose equivalent iodine as a means for accepting higher projected leakage rates provided justification for the RCS DE l-131 activity below 0.35 pCi/gm is provided. Four methods for determining the fuel rod iodine release rates and spike factors during an accident * ,

were reviewed. Each of these methods utilized actualindustry data, including

) Braidwood Units 1 and 2, for pre- and post-reactor trip RCS DE l-131 activities.

~ Each of the methods demonstrated that the actual fuel rod iodine release rates are a small fraction of the release rate as calculated using the NRC SRP methodology. Although these values are a small fraction of that determined by the NRC SRP Method, Braidwood is also requesting an increase in the allowable t

primary-to-secondary leak rate during MSLB. By decreasing the TS RCS DE l-131 activity limit by a factor of twenty and increasing the allowable leak rate by a factor of twenty, the activity released to the oublic would be equal to or less than the activity calculated by the SRP method fc,r each of the seventeen reactor trip events reviewed at Braidwood. The predicted end-of-cycle 7 leak rate is 122.3 gpm (Room T/P). The calculated site boundary dose due to this leakage _is 27.63 Rom. This dose meets the requirements of 10CFR100 and GDC 19. All design basis and off-site dose calculation assumptions remain satisfied. This proposed change would not result in a reduction in a margin of safety.

Therefore, based on the above evaluation, Comed has concluded that these changes involve no significant hazards considerations.

.; K:sgrp\mfs\ lodine 2. doc '

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I l

l ATTACHMENT D ~  !

ENVIRONMENTAL ASSESSMENT FOR PROPOSED CHANGES TO APPENDIX A 1 TECHNICAL SPECIFICATIONS OF l

FACILITY OPERATING LICENSES l

NPF-72 AND NPF-77  !

Commor wealth Edison Company (Comed) has evaluated this proposed license amendment request against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with Title 10, Code of Federal Regulations, Part 51, Section 21 (10 CFR 51.21). Comed has determined that this proposed license amendment request meets the criteria for a categorical exclusion set forth in 10 CFR 51.22(c)(9). This determination is based upon the following:

1. The proposed licensing action involves the issuance of an amendment to a license for a reactor pursuant to 10 CFR 50 which changes a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or which changes an inspection or a surveillance requirement. This revision will lower the Reactor Coolant System Dose Equivalent lodine 131 limit from 0.35 microCuries per gram M

- 0.05 microCuries per gram. This revision will also lower the Reactor Coolant System Dose Equivalent lodine-131 limit in Technical Specification Figure 3.4-1. These revisions will be in effect for the remainder of Unit 1 Cycle 7;

2. this proposed license amendment request involves no significant hazards considerations as demonstrated in Attachment C;

3. there is no significant change in the types or significant increase in the amounts of any effluent that may be released off-site; and

4. there is no significant increase in individual or cumulative occupational radiation exposure.

Therefore, pursuant to 10 CFR 51.22(b), neither an environmental impact statement nor an environmental assessment is necessary for this proposed license amendment request.

.- . K:sgrphfs\iodincidoc D-1

. -. .