ML20216F114

Application for Amends to Licenses NPF-72 & NPF-77,revising TS Section 3.4.8, Specific Activity. Rev 0 to Calculation BRW-97-0798-M & Rev 3 to Calculation 95-011 Encl

ML20216F114
Person / Time
Site:	Braidwood
Issue date:	09/02/1997
From:	Stanley H COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20216F120	List: ML20216F114 ML20216F135 ML20216F162 ML20216F169
References
NUDOCS 9709110167
Download: ML20216F114 (19)

Text

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'Isl Ml 4 4 % b4til September 2,1997 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555

Subject:

Braidwood Nuclear Power Station Units 1 and 2 Request for Amendment Facility Operating Licenses NPF-72 and NPF 77 NRC Docket Number: 50-456 and 50-457

Reference:

1. C Shirakiletter to D. Farrar dated July 26,1995, transmitting Amendment 167 for Zion Unit 1

2. J. Hosmer letter to NRC Document Control Desk, dated January 31,1997, requesting amendment to the Byron Unit 1 Technical Specifications Pursuant to Title 10, Code of Federal Regulations, Part 50, Section 90 (10 CFR 50.90), Commonwealth Edison Company (Comed) proposes to amend Appendix A, Technical Specifications, for Facility Operating Licenses NPF 72 and NPF 77, for the Braidwood Nuclear Power Station, Units 1 and 2, respectively. Comed proposes to revise Technical Specification Section 3.4.8 ' Specific Activity', Table 3.4-1 and Technical Specification Bar.os 3.4.8 for Braldwood Unit 1. These changes will reduce the allowable Unit 1 Reactor Coolant ' System Dose Equivalent lodine 131 activity from 0.35 microCuries/ gram to 0.10 microCuries/ gram for the remainder of Cycle 7. f This amendment is necessary in order to provide additional margin to the  ;

maximum site allowable primary _to secondary leakage limit during an-  !

accident.

The justification presented in the attachments to lower the dose equivalent iodine activity below 0.35 microCuries/ gram utilizes a methodology which is [yJ()/

consistent with that used for the Zion Unit 1 Technical Specification E

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A Unhum Onnpan)

Amendment 167 (Reference 1) and the Byron Unit i Technical Specification  !

requested amendment (Reference 2). i

! This package affects Braidwood Unit 1 only, but is being submitted for f Braidwood Unit 1 and Braidwood Unit 2, because the Technical Specification ,

pages are common to both units.  ;

Enclosed is:

Attachment A: Description and Safety Analysis of Proposed Changes to  :

Appendix A Technical Specifications i Attachment B: Marked Up Pages for Proposed Changes to i AppendiTechnical Specifications Attachment C: Evaluation of Significant Hazards Considerations for Proposed Changes to Appendix A Technical Specifications 1'

Attachment D: Environmental Assessment for Proposed Changes to-Appendix A Technical Specifications >

! i Comed requests that this proposed amendment be reviewed and approved .

by October 10,1997.  ;

The proposed changes, in this license amendment, have been reviewed by On Site and Off Site review in accordance with Comed procedures.  ;

! Comed will notify the State of Illinois of our application for this license '

amendment request by transm_itting a copy of this letter and its attachments to the designated State Official.

I affirm that the control of this transmittalis true and correct to the best of my knowledge, information and belief, if you have any questions concerning this correspondence, please contact  !

Terrence Simpkin, Braidwood Regulatory Assurance Supervisor, (815) 458-2801, extension 2980.

Sincerely,  !

[ , )/ tL v

. (4./

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- JJ H.- Gene St ley  :

1. Site Vice President

' Braidwood Nuclear Generating Station t

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, . . , . - .~.,...a- . . . _ . . . . . . _ , . . . _ . , _ , _ , . . , , . _ _ _ , . , _ . _ _ _ . , _ _ _ , _ , _ _ . _ , , _ . _ , . . . _ _ _ _ . , , , _ . , _ _ _ _ .

Signed before me on this b ' day of $c DY6 rn b e r' .1997 by '

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_> IN l f'~hppldgk'['jj{'[~~'^~~~^)

Notary Public ll MICHELLE A TURNBULL h ,

notAnnuanc, erAtt os ntwois i

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Attachments .

cc: C. Phillips, Senior Resident inspector - Braidwood

> G. Dick, Braidwood Project Manager NRR A. B. Beach, Regional Administrator - Rlll D. Lynch, Senior Project Manager NRR Office of Nuclear Safety. IDNS 4

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ATTACHMENT A DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSE NPF 72 AND NPF 77 DESCRIPTION OF THE PROPOSED CHANGE A.

Commonwealth Edison (Comed) proposes to revise Technical Specification (TS) 3.4.8, " Specific Activity," Table 3.41 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) activity limit frorn 0.35 microCuries per gram (pCilgm) to 0.10 pCilgm through Cycle 7. During the end-of Cycle 7 refueling outage, the original Westinghouse Model D-4 steam generators (SG) will be replaced with Babcock & Wilcox International (BWI) steam generators. Consequently, the reduced RCS DE l 131 activity limit will only be required until the original steam generators are replaced. The Unit 2 RCS DE l 131 activity hmits are unaffected by this change.

These changes are discussed in detailin Section E of this attachment. The affected TS pages showing the actual changes are included in Attachment B of this request.

B. DESCRIPTION OF THE CURRENT REQUIREMENT TS 3.4.8 requires that the specific activity of the reactor coolan', be less than or equal to 0.35 pCilgm DE l 131 for Unit 1. When in Modes 1,2, or 3 (greater than or equal to 500*F), action is required to place the unit in at least Hot Standby with T.,less than 500 F within G hours if the RCS DE l-131 activity limit has been exceeded for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or if the limits of TS Figure 3.4-1 have been exceeded. When in Modes 1,2,3,4, or 5, sampling and analysis in accordance with Table 4.4-4 is required when the RCS DE l 131 activity limit of

. 0.35 pCilgm for Unit 1 is exceeded until the specific activity of the RCS is rostored to within its limits.

C. BASES FOR THE CURRENT REQUIREMENT The limitations on the specific activity of the RCS provide confidence that the resulting two-hour off site dose will not exceed an appropriately small fraction of K:nlatytmdwtmgentnsdiDI A1

the 10 CFR Part 100 dose guidelino values. The evaluation was based on an acceptance criteria of 30 Rem thyroid dose at the Exclusion Area Boundary per NUREG 0800, the Standard Review Plan (SRP), Section 15.1.5, Appendix A for an accident initiated lodinn spike. The bounding accident is a Main Steam Line Break (MSLB)in conjunction with an assumed steady stato reactor to-secondary steam generator leak rato of one gallon por minute (gpm) and the RCS specific activity at the TS lim lt. These conditions were applied to both a pro accident and an accident initiated iodine spike. For the pre accident iodine transient, the RCS DE l 131 activity was assumed to be at the TS transient limit of 60 pCl/gm.

For the accident initiated spike, the activity was assumed to be at the standard TS steady stato limit of 1.0 Ci/gm with a post-trip iodine release rate spike from the fuel to the RCS 500 timos tho steady stato release rato. The secondary coolant DE l 131 activity was assumed to be at the TS limit of 0.1 pCilgm. Each steam generator was assumed to hevo operational leakage at the TS limit of 150 gallons por day (gpd). The accident initiated spike was determined to be the most limiting condition.

In suppori of a licenso amondment request for the application of a 1.0 volt Interim Plugging Critoria (IPC) for steam generator tube support plate indications, the maximum site allowable primary-to secondary leakage was determined. The NRC SRP methodology was used to determine this leakage limit. The resulting limit of 9.4 gpm at RCS temperature and pressure was obtained, assuming an RCS DE l 131 concentration of 1.0 pCilgm. The 1.0 Vnit IPC amendment request was approved in a May 7,1994, Safety Evaluation Report (SER)(Reference 1). The maximum site allowable leakage limit was raised to 26.8 gpm at RCS temperature and pressure by reducing the TS RCS DE l-131 activity limit from 1.0 pCilgm to 0.35 pCilgm in support of the IPC licenso amendments (References 1,2, and 3).

D. NEED FOR REVISION OF THE REQUIREMENT Comed is requesting a reduction in the Unit 1 RCS DE l 131 activity limit from 0.35 pCllgm to 0.10 pCilgm. This change is required in order to provide additional margin to the maximum site allowable primary to secondary leakage limit. The total postulated 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* critoria, and 150 gpd leakage at room temperature and pressure (Room T/P) from each of the three unfoulted steam generators. The tutal leakage, during an MSLB accident, predicted to occur at the End-of-Cycle 7, is 62.4 gpm at Room T/P. This accident leakage includes 57.1 gpm (Room T/P) from indications left inservice due to the 3.0 Volt Interim Plugging Criteria,5 gpm (Room T/P) from indications that could develop during Cycle 7 at the top of the tubesheet roll transition K:nta\bytmdsungnerudil.11 A2

- region, and 0.3 gpm (Room T/P) from the unfaulted SGs. Since the F* criteria has not been implemented on any inservice SG tubes, no leakage is included for the F* criteria.

E. DESCRIPTION OF THE REVISED REQUIREMENT The footnotes associated with the TS Limiting Condition for Operation (LCO) 3.4.8.a; Modes 1,2, and 3 Action a; Modes 1,2,3,4, and 5 Action; and Table 4.4-4 will be revised to lower the RCS DE l 131 activity limit for Unit 1 through Cycle 7 to 0.10 pCilgm. The revised footnote will read:

"For Unit 1 through Cycle 7, reactor coolant DOSE EQUIVALENT l 131 will be limited to 0.10 microCuries per gram."

TS Figure 3.41, " DOSE EQUIVALENT l 131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >1 Cl/ GRAM DOSE EQUIVALENT l 131*" will be revised to reflect the new Unit 1 RCS DE l 131 activity limit. Specifically, the curve labeled " UNIT 1 LIMIT" will be revised to

" UNIT 1 CYCLE 7 LIMIT" and will be modified to reflect the 0.10 pCilgm limit.

The curve labeled " UNIT 2 LIMIT" will be revised to " UNIT 1 LIMIT AFTER CYCLE 7, UNIT 2 LIMIT " The area under the " UNIT 2 LIMIT" curve labeled

" ACCEPTABLE OPERATION FOR UNIT 2, UNACCEPTABLE OPERATION FOR UNIT 1" will be revised to " ACCEPTABLE OPERATION FOR UNIT 1 AFTER CYCLE 7 AND UNIT 2, UNACCEPTABLE OPERATION FOR UNIT 1 CYCLE 7," The footnote to TS Figure 3.41 will be revised to read:

For Unit 1 through Cycle 7 Reactor Coolant Specific Activity > 0.10 pCl/ Gram DOSE EQUIVALENT l-131.'

An insert will be added to Bases 3/4.4.8, " Specific Activity,* to identify the bases for the reduced 1 131 limit. This insert will read as follows:

"For Unit 1 through Cycle 7, the limitations on the specific activity of the reactor 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 a Main Steam Line Break accident in conjunction with an assumed steady state primary-to secondary steam generator leakage rate of 150 gpd from each of the unfaulted steam generators and a maximum site allowable primary to secondary leakage from the faulted steam generator."

F. BASES FOR THE REVISED REQUIREMENT

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The methodology presented herein to lower the RCS DE l 131 activity below 0.35 pCilgm, is consistent with that used for the Zion Unit 1 Technical Specification Amendment No.167 (Reference 6), which lowered the RCS DE l- .

131 activity limit to 0.04 pCl/gm. The justification is also consistent with that used for the Byron Unit i amendment request (Reference 7), which is requesting to lower the RCS DE l 131 activity limit to 0.2 pCilgm. The Staff in an SER approved this methodology for Zion dated July 26,1995. Specifically, the Zion SER stated the following:

"Therefore, based on the plant specific information supplisd by the licensee, the staff considers it unlikely for the short time period of this amendment that an accident initiated lodine spike for Zion Unit 1 would be greater than the NRC SRP assumed value. The change to the RCS dose equivalent lodine concentration below 0.35 microCurles per gram, as proposed by the licensee, is acceptable for the interim period for which the TS change is requested."

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 nCl/gm, the site allowable leak limit, calculated in accordance with the NRC SRP methodology, was determined to be 6.63 gpm at room T/P. By reducing the RCS DE l 131 activity limit to 0.10 pCilgm, maintaining the iodine release rate spike fe: tor of 500, and increasing the site allowable leakage limit to 66.3 gpm (6.63 gpm divided by 0.1), the maximum activity released to the environment is not changed. Therefore, the offsite dose assessment and conclusions previously reached remain valid and continue to meet the requirements of 10CFR100.

However, the lodine release rate calculation methodology requires further evaluation when the RCS DE l-131 activity limit is reduced to values below 0.35 pCilgm. In August of 1995, the Staff issued NRC Generic Letter 95-05 (GL 95-05)" Voltage Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking." In Section 2.b.4 of to GL 95-05, pertaining to the calculation of off site and Control Room doses, the Generic Letter states, " Reduction of reactor coolant iodine activity is an acceptable means for accepting higher projected leakage rates and still meeting the applicable limits of Title 10 of the Code of Federal Regulations Part 100 and GDC 19 utilizing licensing basis assumptions." The Generic Letter also states, " Licensees who wish to take credit for reduced reactor coolant system lodine activities (below 0.35 microCuries per gram dose equivalent 1-131) in the radiological dose calculation should provide a justification supporting the request that evaluates the release rate data described in Reference 6."

Reference 6 of GL 95-05 is a report by J.P. Adams and C.L. Atwood, "The lodine Spike Release Rate During a Steam Generator Tube Rupture," Nuclear K:nla'hbudatmycnibrudil31 A-4

Technology, Vol. 94, p. 361 (1991).

Since some of the spike factors in the Adams and Atwood report 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. Comed has summarized and compared four methods postulating the effects of an MSLB accident in conjunction with primary to secondary leakage. These four methods that were evaluated are:

Method 1: NRC SRP Methodology Method 2: Calculation of site specific lodine release rates using actual Braidwood Unit i and Braidwood Unit 2 operational data. The data includes iodine release rates with and without fuel defects. The iodine release rate methodology described in Section ll.C of the Adams and Atwood report is used to perform the calculation.

Method 3: Calculation of an absolute iodine release rate normalized to plant power derived from an industry database at a 95% confidence level as described in Section lll of the Adams and Atwood report.

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

Method 1 Evaluation (NRC SRP Methodology)

In accordance with the NRC SRP, the current maximum site allowable primary-to secondary leakage dose calculation for Braidwood (Reference 4) assumes two cases:

1) an initial RCS DE l 131 activity of 60 pCilgm due to a pre-accident iodine spike caused by a reactor transient; and,

2) an initial RCS DE l-131 activity of 1.0 pCilgm with a concurrent lodine spike that increases the DE l 131 release rate from the fuel rods to the RCS by a factor of 500. The spike factor is defined as the ratio of the post-trip release rate to the steady state release rate.

To determine which case is limiting, the site allowable leak rate was calculated based on Low Population Zone (LPZ) dose, and Exclusion Area Boundary (EAB) i using the applicable 10CFR100 thyroid dose limit. Case 2 was determined to be the limiting case and resulted in a maximum site allowable primary-to secondary leakage of 6.63 gpm (Room T/P) with the RCS DE l 131 activity limited to 1.0 l pCilgm, This is more limiting than Case 1 (initial RCS DE l 131 activity of 60 j pCilgm) which resulted in a maximum site allowable primary to secondary l 1

I K:nta\bybwdutmpnibrudil31 A5 l

leakago limit of 46.1 gPm (Room T/P). The 6.63 gpm maximum site allowable primary to-secondary leakage limit includes 150 gpd from each of the throo unfaulted steam generators.

By reducing the RCS DE l 131 activity limit by the same proportion as increasing the allowable leak rato and maintaining the SRP iodine release spiko factor of 500, thoro is no not change in the total amount of Curies released during the transient. The revised RCS DE l 131 lirnit of 0.10 pCl/gm replaces the original steady stato limit of 1.0 pCilgm for Case 2 and 6.0 pCl/gm replaces the original translent limit of 60 nCl/gm for Caso 1.

A reduction in the RCS DE l-131 activity limit, while maintaining the SRP lodine release spiko factor of 500, allows for an increase in the maximum site allowable primary to secondary leakage without an increase in thyroid dose at the sito boundary. Based upon an lodino limit of 1.0 pCi/gm, the maximum sito allowable leakago was calculated to bo 6.63 gpm (Room T/P). The Unit 1 RCS DE l 131 activity limit was reduced from 1.0 pCilgm to 0.35 pCilgm in support of the Interim Plugging Critoria License Amendment Roquest that was approved for Braidwood Unit 1 in References 1,2, and 3. This reduction of the RCS DE l 131 activity limit increased the maximum site allowable leakage from 6.63 gpm to 19.0 gpm (Room T/P). By further reducing the lodino limit to 0.10 pCilgm, the maximum site allowable leakage can be increased to 66.3 gpm (6.63 gpm divided by 0.10) without a resulting increase in the thyroid dose at the site boundary. The predicted end of-cycle 7 leak rato is 62.4 gpm (Room T/P). The calculated MSLB doso due at this leakago is 28.2 Rom. This dose meets the requirements of 10CFR100 and GDC 19.

An evaluation of the Control Room doso, 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 licenso amendment request I

for application of a 1.0 volt Interim Plugging Critoria (Reference 5). This ovaluation concluded that the activity released to the environment from an MSLB accident (154 Curies for a Pre-accident iodino spike and 105 Curios for an accident initiated lodino spike) is bounded by the activity released to the environment from the Loss of Coolant design basis accident (1290 Curios).

Therefore, the Control Room doso, 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.

Method 2 Evaluation (Adams and Atwood Methodology) )

The Adams and Atwood report concluded that the NRC SRP methodology, which specDes a release rato spike factor of 500 for lodine activity from the fuel rod to l K:nta*>budistmun%mdil31 A-6 i

the RCS, is conservative when the RCS DE l 131 concentration is greator then 0.3 pCilgm. In order to evaluate whether a release rato spiko factor of 500 is conservativo below 0.3 pCilgm, actual operating data from the previous reactor trips of Draidwood Units 1 and 2, with and without fuel dofocts, woro reviewod and analyzed using the methodology proson;;d in Section ll.C of the Adams and Atwood report (Mothod 2). The samo fivo data setooning critoria described in the Adams and Atwood report woro applied to the Braldwood data to ensure consistoney and validity when comparing the Braidwood results to the data in the Adams and Atwood report. The specific data scrooning critoria applied to each Draldwood reactor trip are as follows:

1) The plant must have boon at steady stato conditions a minimum of favo days prior to the reactor trip.

2) Knowledgo of the sioady stato lodino concontration.

3) At least one post trip chomistry samplo obtained two to six hours following the reactor trip.

4) No occurrence of a post trip RCS porturbation.

5) Availability of all requisito transient information (e.g., pt:fication flow, trip dato and timo, post trip samplo dato and timo).

Of the reactor trip ovents at Braidwood Units 1 and 2, seventoon (17) mot the fivo scrooning critoria described above. The data collected and the calculated roloaso rato for each transient satisfying the scrooning critoria is summarized in Tablo A 1. The post trip maximum roloaso rato in the table is based on the bounded maximum lodino concentration (throo timos the measured post trip concentration) and an assumed timo after the trip of two hours.

The ovents in Unit 1 Cycles 3 and 4, and Unit 2 Cycles 3 and the second part of Cyclo 4 occurred during periods of no fuel defects. All remaining ovents occurrod during cycles with fuel defects. Braidwood Unit 1 Cycle 7 is currently operating with no fuel defects and on RCS DE l 131 activity of approximately 3E-4 pCilgm. Evonts 1, G,8,9,11, and 12 have steady stato iodino values that are reasonably close to the current operating conditions. It is thorofore reasonable to concludo that, assuming continued operation with little to no fuel defecte, the calculated spiko factors from thoso events would reflect an actual event for Unit 1 Cycle 7. In all six of those instances, the calculated spike factor is a small fraction of the assumod spiko factor of 500 in the NRC SRP methodolo0y.

Sinco some of the spiko factors were greater than 500 when the RCS DE l 131 activity, prior to the accident, was loss than 0,3 Cilgm, Comed examined the conservatisms in the current roloaso rato calculation. The primary reason for thoso high factors (up to 12,000)is not due to the fact that the absoluto post trip release rato is high (factor numerator), but rather because the steady stato relonso rato (factor denominator)is low. The Braidwood specific data resultod in six ovents with a calculated release rato spiko factor greater than 500 (Events 2, K nl/b tmdatmeen'bnndil11 3 A7

3,4,5,15, and 16). Each of these events had a pre accident RCS DE l 131 activity limit concentration below 0.3 pCilgm. It is not expected, based upon the Unit 1 Cycle 7 fuel conditions, that a spiking factor greater than 500 would occur.

A low RCS DE l 131 activity level would offset the effects of a spiking factor greater than 500, if one did occur, on the post trip lodine release rate. The revised lodine limit will also ensure that the operating cycle will not continue if significant fuelleakage develops.

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 pCilgm was calculated. Using the Braidwood specific data, the pre trip steady-state release rate is 27.5 Ci/hr. Using a release rate spike factor of 500 for the accident initiated spike, the post trip maximum release rate would be 13,733 Cl/hr (SRP Methodology). The highest post trip iodine release rate from the Braidwood trip data, Event 15, was 1335 Cl/hr. Although this value is lower than that determined by the NRC SRP Method at 1.0 pCi/gm, 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 maximum iodine release rate is 1373 Cl/hr. None of the Braidwood data exceeds 1373 Cl/hr, although eight (8) of the 168 data points in the Adams and Atwood report exceed 1373 Ci/hr. The eight (8) data points had a pre-trip RCS DE l 131 activity between 0.09 pCilgm and 0.6 pCi/gm. Only one (8) of the eight (8) data points had a pre-trip RCS DE

! 131 activity below 0.1 pCilgm.

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 pCilgm. Since the Braidwood data does not include data near 0.1 pCl/gm (the requested new TS limit), it is appropriate to use the Adams and Atwood database near 0.1 pCl/gm 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 pCilgm 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 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 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 pCilgm). This release rate is less than the calculated Braidwood maximum value of 1373 Cl/hr. ,

As further assurance that the 10CFR100 and GDC 19 limits are not exceeded, K:nlahlmdutmpnerudil31 A-8

several conservatisms are built into the offsite dose calculation (defense in depth). 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 i RCS DE l 131 activity near 3E-4 pCl/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 Rom 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 Reference 8) than the 1.0 valued (no partitioning) used in the offsite dose calculation. This reduces 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 iodine released from the defective fuel.

6. It 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.

Althcugh the iodine release spike factor can be postulated to be greater than 500, tne conservatisms in the offsite dose calculation would reasonably result in an exposure at the site boundary below the 10CFR100 limits. lodine spikes greater than 500 are typically associated with moderate amounts of defective fuel. Little to no defective fuel and large amounts of defective fuel typically do not result in iodine spike factors greater than 500.

Method 3 (Statistical Adams and Atwood Methodology)

The Adams and Atwood report applied a statistical analysis to the industry data to estimate the probability distribution of the normalized release rate associated with the iodine spike. The results from this statistical analysis are cumulative probability distributions that are a measure of the probability that an accident would result in an iodine spike with magnitude less than a given value. This methodology was used to determine if the Braidwood reactor trip data was consistent with the industry data. This analysis ratios the post-trip fuel rod iodine release rate to the pre-trip steady-state reactor power (megawatts-electric, MWe). This methodology was used to normalize the fuel rod iodine K:ntait9tmdwtmpennut:131 A9

l I

release rates in order to compare the fuel rod lodine release rates among the various plant sizes while eliminating the artificiality of assuming a single steady.  ;

state iodine concentration. By utilizing this normalization method and the l statistical techniques referenced in the Adams and Atwood report, a normalized 1 fuel rod lodine release rate for the 168 events, evaluated in the Adams and Atwood report, was determined to be less than 1.09 Cilhr MWe at a 95%

confidence on the 90th percentile.

Applying this fuel rod lodine release rate to Braidwood Unit 1 at full power (1175 MWe), the predicted post-accident release rate would be 1280 Cl/hr. This means that with 95% confidence, it is expected that 90% of all the accidents will result in an lodine spike with a normalized release rate less than 1280 Cilhr. As discussed in Method 2, the actual Braidwood post-trip lodine release rates ranged from 0.02 to 1335 Ci/hr. One (1) of the seventeen (17) data points from Braidwood exceeds the 1280 Ci/hr limit.

The predominant factors in calculating the offsite dose is the post trip iodine release rate from the fuel and the flowrote 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 lodine release rate will determine the level of activity in the RCS that will be released. The flowrote will determine at what rate this activity is released to the environment. 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 Ci/hr (spike factor of 3471). The second highest post-trip lodine release rate from the Braidwood data was 802 Ci/hr (spike factor of 1483).

Figure A 2 shows the post-trip iodine release rate from the fuel versus the pre-trip RCS DE l 131 concentration using the data from the Adams and Atwood report and the Braidwood data. Also shown on Figure A-2 is a 95% confidence prediction for the combined data sets. This prediction shows that, below 0.1 Cilgm, all but one data point is bounded by the 1373 Ci/hr release rate. This one data point is bounded by the 95% confidence. The 95% confidence also bounds all but one of the Braidwood data points. 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. The defense-in-depth measures mentioned above should reduce the possibility of exceeding the 10CFR100 limits should a fuel release rate greater than 1373 Cilhr occur.

Method 4 (EPRI Report Methodology)

This method presents the results from the Draft EPRI Report TR-103680, Rev.1, November 1995, " Empirical Study of lodine Spiking in PWR Power Plants." The objective of the EPRI study was to quantify the lodine spiking in postulated Main Steam Line Break / Steam Generator Tube Rupture (MSLB/SGTR) accident K:ntahtmdistmgen'brudil31 A 10

sequences. Based upon measured data from fifteen normal operational reactor i transients and two SGTR events, the EPRI empirical model shows a good correlation between the measured and the predicted RCS DE l 131 concentration when using iodine release rate spike factors of 45 to 150. This also supports the conclusion in Method 2 that the NRC SRP spike factor of 500 .

is conservative.  !

The EPRI empirical model was used to predict the fuel rod lodine release rate in postulated MSLB/SGTR accident sequences. Predictions based on the empirical model agree well with observed spiking based on industry data.

Predictions for two MSLB/SGTR accident sequences yield two-hour average iodine concentrations of 3.1 pCl/gm or less in the reactor coolant. This value is less than the value based on the NRC SRP methodology described in the EPRI i report (38 pCl/gm), indicating that the SRP methodology significantly over predicts the iodino spike. For Braidwood, using the SRP methodology with on RCS DE l 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 pCl/gm. At an RCS DE l 131 acilvity of 0.1 pCilgm, it would require a spike factor of nearly 5000 to obtain a Post-Trip RCS DE l 131 activity near 35.5 pCilgm. With a Post-Trip RCS DE l 131 activity of 35.5 pCi/gm, an increase in the allowable leak rate could impact the 10CFR100 limits. To accommodate for an increase in the allowable leak rate by a tactor 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 pCilgm. The maximum Post-Trip RCS activity seen at Braidwood is 3.29 pCilgm at approximately three hours after the event.

Conclusion -

The current Braidwood Unit 1 Cycle 7 RCS DE l-131 activity level has been relatively stable at approximately 3E-4 pCilgm over the last three months.

Braidwood Unit 1 has operated the previous two cycles with high fuel integrity. If an MSLB accident were to occur during Cycle 7 with the present Unit 1 RCS DE l-131 activity, the specific activity in the RCS would not be expected to increase substantially (much less than a spike factor of 500). Therefore, assuming continued high integrity of the Unit 1 fuel, it is reasonable to expect that for the remainder of Cycle 7, during the time period for which the requested amendment would be applicable, the accident-initiated lodine spike factor for Unit 1 would be below the NRC SRP assumed value of 500.

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 lodine spiking factor limit will be met, the conservatisms in the offsite dose calculation (defense-in-depth arguments listed above) provide added assurance that the 10CFR100 limits, General Design Criteria (GDC) 19 criteria, and the Kmla'hytmdistmgerfbrudil31 A ll

requirements of NRC Generic Letter 95-05 will be satisfied if the lodine spike factor e> Aeds 500 or the post trip fuel release rate exceeds 1373 Cl/hr, Comed  :

is required by TS 4.4.5.3.c 4 to perform an inservice inspection of the SGs in the  ;

event of a MSLB accident. This inspection is required prior to returning the SGs  ;

to service following the accident. ,

i G. IMPACT OF THE PROPOSED CHANGE The requested change in the Unit 1 RCS DE l 131 activity limit from the current 0.35 pCi/gm to 0.10 pCl/gm permits an extension to the Unit 1 Cycle 7 operating period to allow full cycle operation without predicted primary to secondary accident leakage exceeding the maximum site allowable steam generator leakage limit. Full cycle operation will permit Unit 1 operation until the scheduled steam generator replacement outage. Utilizing the current RCS DE l-131 activity of 0.35 pCl/gm, the site maximum allowable leakage limit is 19.0 gpm (Room T/P). Utilizing the proposed 0.10 pCilgm RCS DE l-131 activity limit, the revised maximum site allowable leakage limit is 66.3 gpm (Room T/P). The predicted end of-cycle 7 leak rate is 62.4 gpm (Room T/P). The calculated site

. bounday dose due to this leakage is 28.2 Rem. This dose meets the requirements of 10CFR100 and GDC 19.

Generic Letter 95 05 permits lowering the dose equivalent iodine activity as a means for accepting higher projected leakage rates provided justification for RCS DE l 131 activity below 0.35 pCilgm is given. Should the RCS DE l 131 activity increase to the proposed 0.10 pCl/gm !imit, the expected iodine spike should be less than the spike predicted by the NRC SRP. Higher lodine release factors may result with RCS DE l 131 activities less than 0,10 pCi/gm, but due to the initial RCS DE l 131 activity being lower, the resultant dose at the site Exclusion Area Boundary would not exceed a small fraction of the 10CFR100 limits. Recent operating history of Unit 1 provides added confidence that the SRP criteria are conservative.

This amendment request will not result in any changes to existing systems or equipment, nor will it result in the installation of any new systems or equipment.

Therefore, this proposed change would not result in any significant negative impact on any system or operating mode.

At the completion of Braidwood Unit 1 Cycle 7, Comed will be replacing the original Westinghouse D-4 steam generators with BWI steam generators. With the replacement of the steam generators, the RCS DE l 131 activity limit will be returned to the standard value of 1.0 pCilgm.

H. SCHEDULE REQUIREMENTS K:nla'bytmdutmgen'bntdil31 A 12

. _ _ - _ _ _.u_.--_ _ . _ _ - . _ - ___ __ _ . _ - , _

l l

4 This change permits an extension of the Unit 1 Cycle 7 operating period to allow l full cycle operation to the SG Replacement outage. In order to facilitate outage v scheduling, Comed requests that this proposed amendment be reviewed and . i approved by October 10,1997. j

l. REFERENCES i

1. R. Assa letter to D. Farrar dated May 7,1994, transmitting Amendment 50 for

Braidwcod

2. R. Assa letter to D. Farrar dated August 18,1994, transmitting Amendment  !

54 for Braidwood

3. R. Assa letter to D. Farrar dated November 9,1995, transmitting Amendment 77 for Braidwood l

4. Braidwood Calculation BRW.97 078 M, Revision 1, Site Allowable Leakrate  !

Calculation for SG interim Plugging Criteria

5. Braldwood Calculation 95-011, Revision 3, Control Room Dose Calculation 6.- C. Shiraki letter to D. Farrar dated July 26,1995, transmitting Amendment '

167 for Zion Unit 1 i 7. J. Hosmer letter to NRC Document Co,itrol Desk, dated January 31,1997, requesting amendment to the Byron Unit 1 Technical Specifications

8. Byron /Braldwood Updated Final Safety Analysis Report (UFSAR) f i

i f

I l

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- K:nla'b3 bwd6tmgen\brudil31 A 13

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Pre-Trip RCS DE I-131 Concentration (pCilgm) o Adams /Atwood Data e Braidwood Data Regression ---- 95% Prediction 1373 - - - - - - 0.1 L._._. ._ _ _ . _

_ _ _._