ML20216F135

Proposed Tech Specs 3.4.8 Re Specific Activity

ML20216F135
Person / Time
Site:	Braidwood
Issue date:	09/02/1997
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20216F120	List: ML20216F114 ML20216F135 ML20216F162 ML20216F169
References
NUDOCS 9709110170
Download: ML20216F135 (20)

Text

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

l l

l ATTACHMENT 8 ,

MARKED UP PAGES FOR '

PROPOSED CHANGES TO APPENDIX A l

TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-72, AND NPF-77 l

BRAIDWOOD STATION UNI ~' Ad2 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.

9709110170.970902 i PDR ADOCK 05000456-P PDR, K:nlat>bwdstmgentrwdil31 -

, REACTOR C00LANT SYSTEN 3/4.4.8 SPECIFIC ACTIVITY I LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited to:

4

a. Less than or equal to 1 microcurie per gram DOSE EQUIVALENT i 1-131**, and I

b. Less than or equal to 100/l microcuries per gram of gross radioactivity.

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

ACTION:

MODES 1, 2 and 3*: i

a. With the specific activity of the reactor coolant greater than  :

1 microcurie per gram DOSE EQUIVALENT I-131** for more than I 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least NOT STAND 8Y with T m 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 a

b. With.,the specific activity of the reactor coolant greater than 100/E microcuries per gram, be in at least H0T STAND 8Y with Tm 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 />. -

9

• With T m g(reatTr,+h==

ty<m y Crdea3_= anal to 500*F.

c.to

• • For Unit l', reactor coolant DOSE EQUIVALENT I-131 will be limited to OrM microcuries.per gram.

BRAIDWOOD - UNIT: 1 & 2 3/4 4-27 AMENDMENT NO. 69

-,--o w y w , ,,,- - - 4,---, -, -y , m -ms,r ~ ,- ,,-,p 4y, y , , - , , , _, _4, me, e. gm,. _ , , , . , _ , , , - , . . , - , _ _ , _ _ s,

EACTOR C00LANT SYSTEM LIMITING ColmITION FOR OPERATION i ACTI0fi IContinued)

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

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, perfom the sanplinfi and analysis requirements of item 4.a) of Table 4.4-4 until the specif' c activity of the reactor coolant is restored to within its limits.

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

1

(%+ C,<lh 1

c.to

'for Unit F, reactor coolant DOSE EQUIVALEN1.131 will be limited to 045 '

microcuries per gram.

BRAIDWOOD - UNITS 1 & 2 3/4 4-28 AMEN 0 MENT NO. 69

R e p / ,. . . wi(L .rNya /2 7 A

/

23 't 1.

\ t ,

't e

250 s \\ ,

i j, 1 T ,

s 't i

g \

t l

. s ,

I \ ,

s s .--

-.a 1 20 \ Occunaw s t x \ /

f i 'l l t /

\ \

{ _T

't 1 /

/

160 s

\i ,

k -

~

\ A

\  %

( x / 1

\ \

\ ,

\ ----

5 t00

\ '

L AccerAsut opsumov'ron uuri UNACCEPTAaLE

\ '

L DONFoRUWr\

g A < \ v

\ t s

1. T '

s T h

w

\t -

.e y

s

\ (m ? tw --

y _ s ,

/ L <

h \

~~

f

\ (mitw_ __

\

/ \

0,o ,, ,, 3, ,, , ,, ,o ,,0 PCRCINT Of RAT (0 iTRWW. POWER FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT _ SPEC C ACTIVITY >lpCI/ GRAM DOSE EQUIVALENT I-131* l (Ar.v3C Cycle 7

. 6,10

• for Unit P, Reactor Coolant Specific Activity >0rS5 pC1/Gran DOSE EQUIVA NT I-131 j BRAIDWOOD - UNITS 1 & 2 3/44-29 AMENDMENT NO.

INSERT A

'L I l , -- d_

2250 \ '

t:: \

a 't _

_ l "I i ,,

1 s'

t 1.

k  :

200 (T

g - UNACCEPTABLE ._

g L OPERATION E h a i  !

$150 --,. \ -

8 0

xh

>- T g 'L

( i E100 h E --

AccEFTABLE OPERATK)N FOR 3 i A -*

UNIT 1 AFTRR CYCLE F AND UNIT 2: ( )

y _] UNACCEPTABLE OPERATION FOR UNIT 1 CYCLE 7 h UNIT 1 LIMIT v;'

p t

\i AFTER CYCLE 7; -

UNIT 2 LIMIT

g h i W T J

50  %

b \

s m

g i w i h

y _ .._

" AccEPTAats L UNIT 1 CYCLE 7 LIMIT 0 I ii# i# I iii1 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.41 DOSE EQUIVALENT l 131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > 1pCl/ GRAM DOSE EQUlVALENT l 131'

• For Unit i through Cycle 7 Reactor Coolant Specific ActMty > 0.10 pCVGram DOSE EQUIVALENT l 131.

BRAIDWOOD UNITS 1 & 2 3/4429 AMENDMENT No.

4 TABLE 4.4-4 REACTOR C00UWT SPECIFIC ACTIVITY SMFLE N MALYSIS PMERM 1

TYPE OF MEASUREMENT SAMPLE W ANALYSIS MODES IN WICH SMPLE M ANALYSIS _

FREQUENCY M ANALYSIS HEMIRED

1. Gross Radioactivity At least once per 72 hears 1, 2, 3, 4 Detemination**

i

'2. Isotopic Analysis for DOSE EQUIVA- Once per 14 days LENT I-131 Concentration I  !

3. Radiochemical for E Detemination*** Once per 6 months

• 1

4. Isotopic Analysis for iodine a) Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, Including I-131 I-133, and I-135 If,2f,38,4#,5#

whenever the specific activity exceeds 1  !'

pC1/ gram DOSE 93flVALENT I-131**** '

or 100/E pC1/ gram '

of gross radioactivity,

• • d b) One sample between 2 1,2,3 i, and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a  !

THEMAL POWER change i

exceeding 15% of the RATED THEMAL POWER within a 1-hour j j period. ,

BRAIDWOOD - UNITS I & 2 3/4 4-30 M U D 0IT NO. 69 -

, TABLE 4.4-4 (Continued)

TABLE NOTATIONS

1. Untti the s cific activity of the Reactor Coolant system is restored within its inits.

4 Sample to be taken after a minimum of 2 EFP0 and to days of POWER l OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.

• • A gross radioactivity analysis shall consist of the quantitative l 4

measurement of the total specific activity of the reactor coolant except t for radionuclides with half-lives less than 10 minutes and all

~

I

, radiciodines. The total specific activity shall be the sum of the degassed 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 955 confidence level. The latest aval able data may be used for pure beta-omitting radionuclides.

• • • A radiochemical analysis for i shall consist of the quantitative measurement of the specific activity for each radionuclide, except for radionuclides with half-lives less than 10 minutes and all radioiodines,

. which is identified in the reactor coolant. The specific activities for these individual radionuclides shall be used in the detemination of E_ ,

for the reactor coolant sample. Determination of the contributors to E shall be based upon these energy peaks identifiable with.a 955 confidence gg cgg level,

• • • • For Unit If reactor coolant DOSE EQUIVALENT l-131 will be limited to er35-microcuries per gram, i

l i

f

. BRAIDWOOD - UNITS 1 & 2 3/4 4-31 . AMEN 0 MENT NO. 69 ,

, ,w m m. - m .,m. owr ..ww, ay.m.,<,,o,,,:,--ey-w *.,,ww ,-,w,..--..,,ee n. -ve-.+ .--...-.,w --,>,E..-,,-=,u-n.,'- m5-+-.----

l i

r REACTOR COOLANT SYSTEM BASES OPERATIONAL LEAKAGE (Continued)

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

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

the chemistry within the Steady-State Limits provides adequate corrosionMaintaining protection over to ensure the life the structural integrity of the Reactor Coolant System of the plant.

chloride, and fluoride limits are time and temperature dependent.The Corrosion assoc studies show that operation may be continued with contaminant concentration levels in excess of the Steady-State Limits, up to the Transient Limir , for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting cont.inued operation 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 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 appropriately small fraction of 10 CFR Part 100 dose guideline values following asteamgeneratortuberuptureaccidentinconjunctionwithanassumedsteady-state reactor-to secondary steam generator leakage rate of 1 gpm. The values for the limits on specific activity represent limits based upon parametric 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 location and meteorological conditions, were not considered in his evaluation.

e BRAIDWOOD - llNITS 1 & 2 B 3/4 4-5 AMEN 0 MENT NO,10

h i

f 4

4 j Insed B For Unit 1 through Cycle 7, the limitations on the specific activity of the reactor _-

4 coolant ensure that the resulting 2-hour off site doses will not exceed an

, . appropriately small fraction of the 10 CFR Part 100 dose guideline values following's Main Steam Line Break accident in conjunction with an assumed

. . steady state primary to-secondary steam generator leakage rate of 150 god from each of the unfaulted steam generators and a maximum site allowable primary-to-secondary leakage from the faulted steam generator.-

d v

i 4

l 7

4 ..

s i

t' K:nla\b>bwd\stmgen\brwdil31 w- c -iF---  % + = -M v vJy--7b g ate--- e 4 +3&e rE----- * - - - .*-wp'y We- -e-- a w Y ) "9 r-jarW 1

REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued)

The sample analysis for determining the gross specific activity and E can exclude the radioiodines because of the low reactor coolant limit oi 1 microcurie /

gram DOSE EQUIVALENT I-131, and because, if the limit is exceeded, the rtdioiodine 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 radioiodine level in the reactor coolant were at their limits, the radioiodine contribution would be approximately 1%. In a release of reactor coolant with a typical mixture of radinactivity, the actual radiofodine contribution would be about 20%. The exclusion of radionuclides with half-lives less than 10 rr.inutos from these determinations has been made for several reasons. The first consideration is the difficulty to identify short-lived radionuclides in a

' sample that .equires 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 SOUNDARY, 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. Af ter 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

(

REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued) take corrective action. Information obtained on fodine 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 justified by the data obtained.

3/4.4.9 PRESSURE / TEMPERATURE LIMITS d- 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 coolant temperature and pressure and system heatup and

cooldown rates (with the exception of the pressurizer) shall 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 shown. 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.

I 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 S. 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 matrrials in the reactor vessel are determined in accordance with the 1973 Summ2r Addenda to of the ASME Boiler and Pressure Vessel and Code.

8 3/4 4-7 AMEN 0 MENT NO. 30 PRAIDWOOD ~ UNITS 1 & 2

ATTACHMENT C EVALUATION OF SIGNIFICANT 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 accordence 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," Table 3,4-1 and Technical Specification Bases 3.4.8 for Braidwood Unit 1. This revision will lower the Unit 1 Reactor Coolant System (RCS) Dose Equivalent (DE) lodine 131 (1-131) level from 0.35 microCuries per gram (pCilgm) to 0.10 pCilgm. 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 remainder of Unit 1 Cycle 7. At the completion of Braidwood Unit 1 Cycle 7, Comed will be replacing the original Westinghouse Model D-4 Steam Generators, allowing the TS RCS DE l-131 activity limit to be returned to the standard value of 1.0 pCilgm.

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 circumferential indications which may exist in the faulted steam generator, leakage from indications remaining in service in the faulted steam generator due to the application of the approved Interim Plugging Criteria and F* criteria, and 150 gallon per day (gpd) leakage

. K:nta\b>tmd\stmgen\brudil31 C-1

u a _ a- _ _ c-_a -_A4 . A _ =2 =: _ =.Ae 4

- from each of 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 4 a draft EPRI Report TR-103680, Revision 1, November 1995, " Empirical Study of lodine Spikirig 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 secondary leak rate is proportionately increased and the iodine release rate spike factor is assumed to be 500 in accordance with the SRP. With an RCS DE l-131 activity limit of 1.0 pCilgm, the maximum site allowable leakage limit was calculated, in accordance with the NRC SRP methodology, to be 9.33 gallons per minute (gpm). The 9.33 gpm allowable leakage limit was calculated for leakage at the normal operating reactor coolant temperature and pressure. This corresponds to a room temperature and pressure leakage limit of 6.63 gpm. Comed has evaluated the reduction of the RCS DE l-131 activity to 0.10 pCilgm along with the increase of the allowable leakage to 94 gpm (66.3 gpm at room temperature and pressure) and has concluded:

assuming a spike factor of 500, the maximum activity released is not K:nlaitnbudistmgen\brudil31 C-2

changed,and

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

Based on the NRC SRP methodology for dose assessments and assuming the iodine spike factor of 500 is applicable at the new 0.1 pCi/gm 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 from an MSLB accident (154 Curies for a Pre-accident iodine spike and 105 Curies for 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 ellowable primary-to-sccondary 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 (1991), concluded that the NRC SRP methodology, which specifies a release rate spike factor of 500 for lodine 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 Cilgm, 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 screening criteria.

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 much less than the spike factor of 500 assumed in the NRC SRP methodology.

Braidwood Unit 1 Cycle 7 is currently operating with no fuel defects and an RCS DE l-131 activity of approximately 3E-4 pCilgm. The seven previous trips, with no fuel defects, had steady-stato iodine values that are reasonably close to the current operating conditions. It is, therefore reasonable to conclude that, K:nlaibybudistmge:Abnsdil31 C-3

1 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 spike factors were greater than 500 when the RCS DE l-131 activity, prior to the accident, was less than 0.3 pCilgm, Comed examined the conservatisms in the current release rate calculation. The primary reason for these high ratios (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 Cilhr (SRP Methodology). The highest post-trip iodine release rate from the Braidwood trip data, Event 15, was 1335 Ci/hr. Although this value is lower than that determined by the NRC SRP Method at 1.0 pCilgm, Braidwood is also requesting an increase in the allowable primary-to-secondary leak rate. By decreasing the TS RCS DE l-131 activity limit by a factor of ten and increasing the allowable leak rate by a fa: tor of ten, the maximum iodine release rate is 1373 Ci/hr None of the Braidwood data exceeds 1373 Cifnr, although eight (8) of the 168 data points in the Adams and Atwood report exceed 1373 Cilhr. The eight (8) data points had a pre-trip RCS DE l-131 activity between 0.09 pCilgm and 0.6 pCilgm. Only one (1) of the eight (8) data points had a pre-trip DE l-131 activity below 0.1 pCilgm.

f If the Braidwood data 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 0.3 Cilgm. Since the Braidwood data does not include data near 0.1 Cilgm (the requested new TS limit), it is appropriate to use the Adams and Atwood database near 0.1 pCilgm to determine if a spike factor of 500 is appropriate. The Adams and Atwood database contains forty-two (42) data points with a Pre-Trip RCS DE l-131 activity between 0.05 pCi/gm and 0.15 pCilgm. Thirty-four (34) of these forty-two (42) data points (81%) have spike factors less than 500. Using the entire Adams and Atwood database,130 of the 168 data points (77%) have an iodine spike factor less than 500.

Therefore, it is reasonable to assume that a spike factor of 500 would not be K:nla\bytmd\stmgen\bmdil31 C-4

i 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 0.1 pCilgm. The highest spike factor .i seen in the Adams and Atwood report near a Pre-Trip RCS DE l-131 activity of 0.1 pCilgm was 1160 (at 0.093 pCl/gm). This release rate is less than the 4 calculated Braidwood maximum value of 1373 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 a different approach in the Adams and Atwood report, concluded that, at a 95% confidence of a 90 percentile, the post-trip lodine release rate was bounded by 1.09 Ci/hr-MWe.

For Braidwood Station, which has a MWe rating of 1175, the post-trip lodine release rate, at a 95% confidence of a 90 percentile, should not exceed 1280 .

Ci/hr. One (1) of the seventeen (17) reactor trips from Braidwood exceeded 1280 Cilhr. This reactor trip had a post-trip iodine release rate of 1335 Cilhr (spike factor of 3471). The second highest post trip iodine release rate from the Braidwood data was 802 Cl/hr (spike factor of 1483).

For the combined Adams /Atwood and Braidwood data sets, below 0.1 Cl/gm, all but one data point is bounded by the 1373 Cilhr release rate. This one data point is bounced by the 95% confidence. This data suggests that the possibility for a post-trip iodine fuel release rate to exceed 1373 Cl/hr, when the pre-trip RCS DE l-131 concentration is at or below 0.1 pCilgm, is small.

In the fourth method, the results from a Draft Electric Power Research Institute (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 lodine 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 lodine spike factor value of 150 was applied to a steady-state equilibrium RCS DE l-131 activity of 0.33 pCi/gm. The resulting two-hour average iodine concentration for a postulated MSLB/SGTR accident sequence was determinea to be 3.1 pCl/gm. Since the EPRI report is

- based on industry data and the EPRI method predicted a post-accident iodine

- activity, which is a small fraction of the activity predicted by the NRC SRP methodology, it can be expected that, for the proposed 0.10 pCilgm limit under an MSLB/SGTR accident sequence, the post accident iodine activity would 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 4 1-131 activity of 1.0 pCilgm and'a' spike factor of 500, the Post-Trip RCS activity two hours after the event would be near 35.5 pCilgm. At an RCS DE l-131 K:nta\b>bwd\stmgen\bntdil31 C-5 I

- activity of 0.1 pCi/gm, it would require a spike factor of nearly 5000 to obtain a

- Post-Trip RCS DE l-131 activity near 35.5 pCi/gm. With a Post-Trip RCS DE l-131 activity of 35.5 pCilgm, an increase in the allowable leak rate could impact the 10CFR100 limits. To accommodate for an increase in the allowable leak rate by a factor of ten, the resultant activity would need to be below 3.55 pCilgm.

None of the seventeen (17) post-trip data from Braidwood has exceeded 3.55 pCl/gm. The maximum Post-Trip RCS activity seen at Braidwood is 3.29 pCilgm at approximately three hours after the event.

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 that the iodine spiking factor limit will be met, the conservatisms in the offsite dose calculation 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 post-trip fuel release rate exceeds 1373 Cilhr. These conservatisms include, but are not limited to:

1. The RCS DE l-131 activity is more 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 the 500 value.

2. 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.

3. 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 Byron /Braidwood UFSAR) than the 1.0 valued (no partitioning) used in the offsite dose calculation. This reduces the calculated dose by 90%.

4. Primary-to-secondary leakage is not expected to be at the TS limit (150 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.

5. The activity in the RCS is not expected to increase instantaneously with the spike in lodine released from the defective fuel.

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

7. The results from the Braidwood tube pull data indicate that the Interim

Plugging Criteria leak rate is conservative.

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

~ K:nla\bybwistmgen\brudil31 C-6

consequences of an accident previously analyzed.

The RCS DE l-131 activity limit is not considered as a precursor to any accident.

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

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

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 existing equipment or systems. No new operating procedures, conditions or configurations will be created by this proposed amendment.

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 pCilgm is provided. Four methods for determining the fuel rod iodine release rates and spike facters during an accident were reviewed. Each of these methods utilized actual industry 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 primary-to-secondary leak rate. By decreasing the TS RCS DE l-131 activity limit by a factor of ten and increasing the allowable leak rate by a factor of ten, the activity released to the public would be equal to or less than the activity calculated by the SRP method for each of the seventeen reactor trip -

events reviewed at Braidwood. The predicted end-of-cycle 7 leak rate is 62.4

. gpm (Room T/P). The calculated site boundary dose due to this leakage is 28.2 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 K:1.la\bybwd\stmgembmdil31 C-7

changes involve no significant hazards considerations.

5 l

- K:nla\bybwGstmgen\bntdil31 C-8

ATTACHMENTD -

ENVIRONMENTAL ASSESSMENT FOR PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-72 AND NPF-77 Commonwealth 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 chan00s 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 willlower the Reactor Coolant System Dose Equivalent lodine 131 limit from 0.35 microCuries per gram to 0.10 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:nla\b3tmd4tmgembrwdil31 D-1 l

l

. _ _ _ _ . - - _ ~