ML20198L798
| ML20198L798 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 01/14/1998 |
| From: | Tulon T COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20198L802 | List: |
| References | |
| NUDOCS 9801160125 | |
| Download: ML20198L798 (20) | |
Text
- _ _ _
Commonwcalth ihn 0,mpan) 3-firaidwual Generating Statkm Route e t, llow h i =
lirx esille, IL G407>X,19
- Tel Hl 5-4542801
-January 14,1998 United States Nuclear Regulatory Commission Washington, D.C. 20555 Attention:
NRC Document Control Desk
Subject:
Braidwood Nuclear Power Station, Units 1 and 2 Reouest for Amendment Facility Operating Licenses NPF-72 and NPF-77 NRC Docke_t., Number: 50-456 and 50-457
Reference:
1.
H. Stanley letter to NRC Document Control Desk, Dated September 3,1997, requesting Amendment to the Braidwood Unit 1 Technical Specifications 2.
December 11,1997, Meeting between the Commonwealth Edison Company and the Nuclear Regulatory Commission 3.
Westinghouse Letter CCE-97-305, " Comed Braidwood Unit 1, IPC Engineering CYCLESIM Runs", dated November 26, 1997 4.
C. Shiraki letter to D. Farrar dated July 26,1995, transmitting Amendment 167 for Zion Unit 1 5.
J. Hosmer letter to NRC Document Control Desk, dated January 31,1997, requesting Amendment to tb Byron Unit 1 Technical Specifications Reference 1 ' transmitted the Commonwealth Edison Company's (Comed) request to amend Appendix A, Technical Specifications, for Facility Operating Licerises NPF-72 and NPF-77, for the Braidwood Nuclear Pow - Station, Units 1 and 2, respectively. Comed proposed to revise Technical Specification Section 3.4.8 " Specific Activity", Figure 3.4-1, Table 4.4-4 and. Technical Specification Bases 3.4.8 for Braidwood Unit 1. These changes would reduce the allowable Unit 1 Reactor Coolant System (RCS) Dose Equivalent lodine-131 (DE l-131),
activity from 0.35 microCuries/ gram to 0.1 microCuries/ gram. Subsequent to that submittal, Comed and the Nuclear Regulatory Commission met to discuss the justification of the predicted end-of-cycle leak rate from steam generator tubes with outer stress corrosion cracking.
At the most recent meeting, Reference 2, Comed stated that in order to provide additional margin to the maximum site allowable primary-to-secondary leakage limit during a main steam line break, Comed would revise the pending Technical Specification to reduce the DE l-131 activity from the requested 0.1 microCuries/ gram to 0.05 microCurielgram. The decision to lower the RCS DE l-131 to 0.05 i
microCuries/ gram is based on a revised Total End-of-Cycle 7 projected leak rate.
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i NRC Document Control Desk January 14,1998 val'Je during a Main Steam Line Break with containment bypass of 122.3 gpm at rcom temperature and pressure. This includes 122 gpm leakage attributable to IPC indications and 0.3 gpm leakage from the unfaulted steam generators. The revised IPC leak rate value is based on a sensitivity study performed by Comed, Reference 3, which determined the effect on projected IPC leak rate values depending on how voltage dependent growth rates are cpplied. The study showed that the projected End-of-Cycle IPC leak rate valuec increased up to a peak of 122 gpm then began to decrease and become statistically unreliable.
Comed conservatively chose 122 gpm IPC leak rate as a bounding value for Cycle 7. Therefore, the attached amendment supersedes that previously submitted via Reference 1.
The justification presented in the attachments to lower the dose equivalent iodine activity below 0.35 microCuries/ gram utilizes a methodology which is consistent with that used for the Zion Unit 1 Technical Specification Amendment 167 (Reference 4) and the Byron Unit 1 Technical Specificat n iequested amendment (Reference 5).
This package affects Braidwood Unit 1 only, but is being submitted for Braidwood Unit 1 and Braidwood Unit 2 beceJse the Tecnnical Specification pages are common to both units.
Enclosed is:
Attachment A:
Description and Safety Analysis of Proposed Changes to Appendix A Technical Specifications Attachment B:
Marked Up Pages for Proposed Changes to Appendix A Technical Specifications t
Attachment C:
Evaluation of Significant Hazards Considerations for Proposed Changes to Appendix A Technical Specifications Attachment D:
Environmental Assessment for Proposed Changes to Appendix A Technical Specifications Attachment E:
Braidwood Calculation No. BRW-97-0798-M: " Allowable Leak Rate Calculation for Steam Generator Interim Plugging Criteria", Rev. 4, dated December 12,1997 i
Attachment F:
Braidwood Calculation No. 95-011: " Control Room Thyroid Dose from a Main Steam Line Break", Rev. 5, dated December 10,1997 i
d
.c
' NRC Document Control Desk - January 14,1998 Comed requests that this proposed amendment be reviewed and approved by.
April 1,1998. Braidwood Unit 1 is currently operating under an. operability
- evaluation which contains a requirement to administratively maintain DE l-131 below 0.05 pCi/gm throughout Cycle.7, or unt:1 such time as the Staff approves th;s amendment request.
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. licence amendment request by transmitting a copy of this letter and its attachments to the designated State Official.
-I affirm that the control of this transmittal is true and correct to the best of my knowledge, information and belief.
If you have any ouestions concerning this correspondence, please contact this office.-
Sincerely, Ti hy J. Tulon Vice President raidwood Nuclear Generating Station Signed before me on this day of Ln
.1998 by Y
Not$ry Public ll OFFICIAL SEAL l NOTARY PUBUC,8 TATE OF ILUNOl MICHELLE A TURNBULL l
b Attachments kE:..N0e".'-).7-$$[#U' cc:
C. Phillips, Senior Resident inspector - Braidwood G. Dick, Braidwood Project Manager - NRR
. A. B. Beach, Regional Administrator - Rlli D. Lynch, Senior Project Manager - NRR Office of Nuclear Safety-lDNS
ATTACHMENT A DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF FAC?LITY OPERATING LICENSE NPF-72 AND NPF-77 A.
DESCRIPTION OF THE PROPOSED CHANGE '
Commonwealth Edison (Comed) proposes to revise Technical Specification (TS) 3.4.8, " Specific Activity," Figure 3.4-1, Table 4.4-4 and Technical Specification Bases 3.4.8 for Braidwood Unit 1. This revision willlower the Unit 1 Reactor Coolant System (RCS) Dose Equivalent (DE) lodine 131 (1-131) activity limit from 0.35 microcuries per gram ( Ci/gm) to 0.05 pCi/gm 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 Intemational (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 limits 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 coolant be less than or equal to 0.35 pCi/gm 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,y 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 /> if the RCS DE l-131 activity limit has been exceeded for more than 48 h.)urs or if the limits of TS Figure 3.4-1 have beeni 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 pCi/gm for Unit 1 is exceeded until the specific activity of the RCS is restored 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:sgrp\\mfs\\iodinc2. doc A1
~
4 s
- the 10. CFR Part 100 dose guideline _ values. The evaluation was based on an acceptance criteria of 30 Rem thyroid dose at the Exclusion Area Boundary per L
NUREG 0800, the Standard Review Plan (SRP), Section 15.1.5, Appendix A for -
an accident initiated iodine spike. The bounding accident is a Main Steam Line Break (MSLB) in conjunction with an assumed steady-state reactor-to-secondary steam generator leak rate of one gallon per minute (gpm) and the RCS specific --
activity at the TS limit. These conditions were applied to both a pre-accident and an accident-initiated iodine spike. For the pre-accident iodine transient; the RCS DE I 131 activity was assumed to be at the TS transient limit of 60 pCi/om. For-the accident-initiated spike, the activity was assumed to be at the standard TS 4
' steady-state limit of 1.0 Ci/gm with a post-trip iodine release rate spike from the fuel to the RCS 500 times the steady-state release rate. The secondary coolant DE l-131 activity was assumed to be at the TS limit of 0.1 Ci/gm. Each steam-generator was assumed to have operational leakage at the TS limit of 150 gallons per day (gpd). The accident-initiated spike was determined to be the most limiting condition.
In support of a license amendment request for the application of a 1.0 volt interim Plugging Criteria (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 Cilgm. The 1.0 Volt 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 pCi/gm to 0.35 pCi/gm in support of the IPC license amendments (References 1,2, and 3).
I 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 pCl/gm to 0.05 pCi/gm. 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 L indications remaining in' service in the faulted steam generator due to the application of the approved Interim Plugging Criteria, F* Criteria, and 150 gpd leakage at room temperature and pressure (Room T/P) from each of the three L
unfaulted steam generators. The totalleakage, during an MSLB accident, predicted to occur at the End-of-Cycle 7 has been revised to 122.3 gpm at Room T/P, (Reference 10). This accident leakage includes 122 gpm (Room T/P) from indicatir.ns left inservice due to the 3.0 Volt Interim Plugging Criteria, and 0.3 gpm (Room T/P) from the unfaulted SGs.' Since the F* criteria has not been implemented o.n any inservice SG tubes, no leakage is included for the F*
c K:sgrpin fs\\ iodine 2. doc A-2
crit' ria.
e Comed decided to revise the End-of-Cycle 7 IPC proje::ted leakage during a MSLB from the original value of 57.1 gpm (Room T/P), which was submitted as
. part of the Braidwood 1 Post Outage 90 Day Report (Reference 11), to a revised End-of-Cycle 7 IPC projected leakage value of 122 gpm (Room T/P). This was due to a study performed by Comed, which demonstrated the leak rate sensitivity varied depending on how the voltage dependent growth rates were applied. The final value of 122 gpm (Room T/P) is a bounding value determined by varying the voltage growth bin lower cut-off value, Reference 10.
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 0.05 pCi/gm. The revised footnote will read:
"For Unit 1 through Cycle 7, reactor coolant DOSE EQUIVALENT l-131 will be limited to 0.05 microCuries per gram."
TS 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 >1 pCi/ 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 0.05 Ci/gm 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 UN.lT 1 AFTER CYCLE 7 AND UNIT 2, UNACCEPTABLE OPERATION FOR UNIT 1 CYCLE 7." The footnote to TS Figure 3.4-1 will be revised to read:
"For Unit 1 through Cycle 7. Reactor Coolant Specific Activity > 0.05 Ci/ 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 limk. 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 9-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
' K:sgrp\\mfsiiodine2. doc A-3
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
- The methodology presented herein to lower the RCS DE l-131 activity below 0.35 pCi/gm, 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 pCi/gm. The justification is also consistent with that used for the Byron Unit 1 amendment request (Reference 7), which requested
'owering the RCS DE l-131 activity limit to 0.2 Ci/gm. The Staffin an SER approved this methodology for Zion dated July 26,1995. Specifically, the Zion SER stated the following:
"Therefore, based on the plant-specific informatica supplied by the licensee, the staff considers it unlikely for the short time period of this amendment that an accident-initiated iodine spike for Zion Unit 1 would be greater thar. ihe NRC SRP assumed value. The change to the RCS dose equivalent iodine concentration below 0.35 microCuries per gram, as proposed by the licensee, is acceptable for the interim period for whicn 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 pCi/gm, the site allowable leak limit, calculated in accordance with the NRC SRP methodology, was determined to be 6.64 gpm at room T/P, (Reference 4). By reducing the RCS DE l-131 activity limit to 0.05 Ci/gm, maintaining the iodine release rate spike factor of 500, and increasing the site allowable leakage limit to 132.8 gpm (6.64 gpm divided by 0.05), 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 iodine 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 NPC 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 of'-site and Control Room doses, the Generic Letter states, " Reduction of reacw.cociant iodine acSvity 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 K:sgrpafs\\ iodine 2. doc A-4
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 iodine activities (below 0.35 microCuries per g~ ram 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
- 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 iod ne release rates using actual Braidwood Unit 1 and Braidwood Unit 2 operational data. The data L
includes iodine release rates with and without fuel defects. The lodine 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 111 of the Adams and Atwood report.
Method 4: Methodology described in Draft EPRI Report TR-1036SO, Revicion 1 November 1995. " Empirical Study of lodine Spiking in PWR Power Plants".
Method 1 Evaluat'on (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 pCi/gm due to a pre-accident iodine
-- spike caused by a reactor transient; and, 2) an initial RCS DE l-131 activity of 1.0 pCi/gm with a concurrent iodine 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-stcte release rate.
i K:sgrp\\mfs\\ iodine 2. doc A-5 s
41.
To determine which case is limiting, the site allowable ieak rate was calculated i
based on Low Population Zone (LPZ) dose, and Exclusion Area Boundary (EAB) 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 6.64 gpm (Room T/P) with the RCS DE l-131 activity limited to 1.0 pCl/gm, (Reference 4). This is more limiting than Case 1 (initial RCS DE l-131 activity of 60 pCi/gm) which resulted in a maximum site allowable primary-to-secondary leakage limit of 46.1 gpm (Room T/P). The 6.64 gpm maximum site allowable primary-to-secondary leakage limit includes 150 gpd from each of the bree unfaulter) steam generators.
By reducing the RCS DE l-131 activity limit by the same proportion as increasing the allowable leak rate and maintaining the SRP iodine release spike factor of 500, there is no net change in the total amount of Curies released during the transient. The revised RCS DE l-131 limit of 0.05 Ci/gm replaces the original steady-state limit of 1.0 Ci/gm for Case 2 and 3.0 pCilgm replaces the original transient limit of 60 pCi/gm for Case 1.
A reduction in the RCS DE l-131 activity limit, while maintaining the SRP iodine release spike factor of 500, allows ior an increase in the maximum site allowable primary-to-secondary leakage without an increase in thyroid dose at the site boundary. Based upon an iodine limit of 1.0 Ci/gm, the maximum site allowable leakage was calculated to be 6.64 gpm (Room T/P). The Unit 1 RCS DE l-131 activity limit was reduced from 1.0 Cilgm to 0.35 pCilgm in support of the Interim Plugging Criteria License Amendment Request 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 6.64 gpm to 19.0 gpm (Room T/P). By further reducing the iodine limit to 0.05 pCi/gm, the maximum site allowable leakage can be increased to 132.8 gpm (C.64 gpm divided by 0.05) without a resulting increase in the thyroid dose at the site boundary. The maximum predicted end-of-cycle 7 total leak rate is 122.3 gpm (Room T/P). The calculated MSLB dose due to this leskage is 27.63 Rem (Reference 4). This dose meets the requirements of 10CFR100 and GDC 19.
As required by GDC 19, an evalui tion of the Control Room dose, attributed to an MSLB accident concurrent with steam generator prima:y-to-secondary leakage at the maximum site allowable limit, was performed (Reference 5). This evaluation concluded that the activity released to the environment during an eight (c) hour time period from an MSLB accident (812 Curves for a Pre-accident iodine spike and 888 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 allowable primary-to-secondary leakage is limited by the offsite dose at the Exclusion Area Boundary due to an accident-initiated l
K:sgrpsmfs\\ iodine 2. doc A-6
^
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spike / TherAfore, the requirements of 10CFR100 and GDC 19 are met.
~ Method '2 Evaluation (Adams and Atwood Methodology)
The Adams and Atwood report concluded that the NRC SRP methodology,
< which specifies a release rate spike factor of 500 for iodine activity from the fuel
^
- rod to the RCS, is conservative when the RCS DE l-131 concentration is greater 3
~ than 0.3 pCi/gm.- In order to evaluate whether a release rate spike factor of 5001 a conservative below 0,3 pCilgm, 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 -
7 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 wnen comparing the Braidwood results to the data in the Adams and Atwood report. The specific data screening criteria applied to each Braidwood. reactor trip are as follows:
- 1) The plant must have been at steadymtate conditions a minimum of five days prior to the reactor trip.
- 2) Knowledge of the steady-state iodine concentration.
- 3) At least one post-trip chemistry sample obta!ned two to six hours following L
_ the reactor trip.
- 4) ' No occurrence of a post-trip RCS perturbatiors.
L
- 5) Availability of all requisite transient information (e.g., purification flow, trip date and time, post-trip sample date and time).
Of the reactor trip events at Braidwood Units 1 and 2,- seventeen (17) met the five screening criteria described above. The data collected and the calculated release rate for each transient satisfying the screening criteria is summarized in Table A-1. The post-trip maximum release rate in the table is based on the
[
bounded maximum iodine concentration (three times the largest actual E
measured post-trip concentration) and an assumed time after the trip of two hours.
The events in Unit 1. Cycles 3 and 4, and Unit 2 Cycles 3 and the second part of L>
Cycle 4 occurred during periods of no fuel defects. All remaining events occurred during cycles with fuel defects. Braidwood Unit 1 Cycle 7 is currently operating with no fuel defects and an RCS DE l-131 activity of approximately 3E-4 pCi/gm. Events 1,6,8,9,11, and 12 have steady-state iodine values that are
- reasonably close to the current operating conditions. it is therefore reasonable
'.to conclude that, assuming continued operation with little to no fuel defects, the calculated spike factors from these events would roflect an actual event for Unit E1 Cycle 7. In all six of these instances, the F..culated spike factor is a small l fraction of the assumed spike factor of 500 in the NRC SRP methodology.
K:sgrp\\mfs\\ iodine 2. doc :
A7
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e l
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 pCi/gm, Comed examined the conservatisms in the current release rate calculation. The primary reason for the high spiking factors contained in the Adams and Atwood report (up to 12,000), is not due to the fact that 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 events with a calculated release rate spike factor greater than 500 (Events 2, 3,4, 5,15, and 16). Each of these events had a pre-accident RCS DE l-131 activity limit concentration below 0.3 pCi/gm and occurred during cycles having fuel defects.
It is not expected based upon the L./,1it 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 iodine release rate. The revised iodine limit will also ensure that the operating cycle will not continue if significant fuel leakage 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 Cilgm was calculated (Reference 9). 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 Ci/hr. It is important to remember that this number is determined by conservatively increasing the largest measured post-trip RCS De 1-131 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 pCi/gm. By decreasing the TS RCS DE l-131 activity by a factor of twenty the maximum iodine release rate is 680.7 Ci/hr, (13,733 Ci/hr divided by 20). Two (2) of thc 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 in 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 Cilgm. These two (2) reactor trips had post-trip iodine release rates of 1335 Ci/hr (spike factor of 3471) and 802 Cilhr (spike factor of 1483). This can be seen in Figure A-1 which shows the post trip iodine release rate from the fuel versus the pre-trip RCS DE l-131 concentration using data from the Adams and Atwood report and the Braidwood data. Also shown on Figure A-1 is a 95%
confidence prediction for the combined data sets. This prediction shows that of the two (2) points which exceeded 686.7 Cilhr and were below 0.05 pCi/gm, one K:sgrp\\mfs\\iodincidoc A8
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~f.
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- f(1) of the.two (2) was bounded by the 95% confidence prediction; This datai
' : indicates that the possibility for a post-trip iodine fuel release rate to exceed 1
f 686.7 Ci/hr, whon the pre-trip RCS DE l-131. concentration is at or below 0.05
- pCilgm, is small. The defense-in-depth measures mentioned below will reduce'~
a
.the possibility of exceeding a'small fraction of the 10 CFR100 limits should a fuel,
]
' release greater than 686.7 Cilhr occur.
Plotting the Braidwood data with the Adams and Atwood data, (Figure A-2), one.
can see that the conclusions of the Adams and Atwood report are not compromised. Where the Braidwood data contains spike. factors greater than
'i 500, the RCS DE l-131 concentrations are below 0.05 pCi/gm; Since the -
1Braidwood data includes very few data points near 0.05 pCilgm (the requested
^
new TS limit), it is _ appropriate to use the Braidwood database ' combined with the 1
Adams and Atwood database near 0.05 pCi/gm to determine if a spike factor of-
~
500 is' appropriate. The combined databases contain seventy-nine (79) data ~
points with a Pre-Trip RCS DE l 131 activity between 0.01 pCi/gm and 0.10 :
pCilgm. 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 3
- Adams and Atwood database,141 of the 185 data points (76%) have an iodine L
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 0.05 pCi/gmL The highest spike factor seen in the Adams and Atwood report near a L
' Pre-Trip RCS DE l-131 activity of 0.05 pCi/gm was.773 (at 0.05 pCi/gm). The corresponding release rate for this event was 368 Ci/hr which is less than the
- calculated Braidwood maximum release rate of 686.7 Ci/nr.
e
- Defense-in-Depth As further assurance that the 10CFR100 and GDC 19 limits are not exceeded, several conservatisms are inherent to the offsite dose calculation. These g
F~
conservatisms in&:le, 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 j
boundary than the 30 Rem limit of 10CFR100.
h
- 2. lodine partitioning is not accounted for in the far NSG. 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 -
calculationc This reduces calculated dose by 90%.
t F
- 3.,The activity.in the RCS is not expected to increase instantaneously with
' the spike in iodine released from the defective fuel.
4.1 The results from the' Braidwood tube pull data indicate that the projected Interim Plugging Criteria leak rate is conservative.
s;
- K
- sgrp\\mfs\\ iodine 2. doc
' A-9 y
+
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. ~ -.,..
,, -. -. - - - -..,. - -. -...,. - - -.~.~.
. in addition, the' current Braidwood Unit 1 operating conditions provide defense in
..t
- 1. ; Braidwood Unit 1 is currently operating with a debris resistant fuel desico 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.
a 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 Ci/gm with no l.
fuel defects, the spiko factor is expected to be considerably smaller than ' '
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 iodine spike for Braidwood Unit 1 would be greater than the NRC SRP assumed value, j-
' 5. 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.
- Although the iodine release spike factor can be postulated to be greater than 500, the defense-in-depth measures listed above would reasonably result in an
- exposure at the site boundary below the 10CFR100 limits. Units operating with little to no defective fuel are expected to result in spiking factors well below 500.
Method 3 (Statistical Adams and Atwood Methodology)
The Adams and Atwood report applied a statistical analysis to the industry data to' estimate te probability distribution of the normalized release rate associated with the iodin e 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 methodolooy 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, t
' M_We)c;This methodology was used to normalize the fuel rod iodine release
' rates in order to compare the fuel rod iodine release rates among the various plant sizes while eliminating the artificiality of assuming ~a single steady-state
- iodine concentrationf By utilizing this normalization method and the statistical 41 techniques referenced in the Adams and Atwood report, a norrnalized fuel rod e
' ' K:sgrp\\mfs\\ lodine 2. doc.
a.
A-10
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. ~. _
~. _,..,
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iodine release rate for the 168 events evaluated in the' Adams and Atwood report ~
was determined to be less than 0.608 Ci/hr MWe at a 95% confidence on the{
- 85th percentile.:
1 Applying this fuel rod iodine release rate to Braidwood Unit 1 at full power (1175 :
MWe), the predicted post-accident release rate would be 714 Ci/hr. This meanst
-that with 95% confidence, it is expected that 85% of all the accidents will result in an iodine spike with a normalized release rate less than 714 Ci/hr. As discussed
- in Method 2, the release rate calculated using the SRP accounting for a factor of.
. twenty decrease in the TS RCS DE l-131 activity is 686.7 Ci/hr.
-t The predominant factors in calculating the offsite dose are the post-trip iodine -
y release rate from the fuel and the flowrate at_which the activity is being released -
l to the environment, not whether the spike factor is greater than or less than 500; The post-trip iodine 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. As discussed in Method 2, two (2) of the seventeen (17) reactor trips from Braidwood exceeded 686.7 Ci/hr. Both occurred during cycles with fuel defects. Braidwood Unit 1 is currently operating with ne fuel defects. The defense-in-depth measures mentioned in the previous,
- section will reduce the possioility of exceeding a small fraction of the 10CFR100
? limits should a fuel release rate greater than 686.7 Cl/hr 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 yeas to quantify the iodine spiking in postulated Main Steam Line Break / Steam Generator Tube Rupture (MSLB/SGTR) accident sequences. Based upon measured data from fifteen normal operational reactor transients and two SGTR events, the EPRI empirical model shows a good correlation between the measurea 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.
a '
The EPRI empirical model was used to predict the fuel rod iodine release rate in
, postulated MSLB/SGTR accident sequences. Predictions based on the
-empirical model agree well with observed spiking based on industry data.
P Predictions for two MSLB/SGTR accident sequences yield two-hour average iodine concentrations of 3.1 pCi/gm or less in the reactor coolant. This value is -
Isss than the value based on the NRC SRP methodology described in the EPRI
? report (38 pCi/gm), indicating that the SRP methodology significantly over 1 predicts the iodine spike. For Braidwood, since the RCS mass'and the clean-up L system constant are similar to that used in the EPRI model, using the SRP
~ Kisgrpimfs\\iodineldoc-A-ll 4
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methodology with an RCS DE l-131 activity of 1.0 pCi/gm and a spike factor of 500, the Post-Trip RCS activity two hours after the event would also be near 38 Cl/gm. At an RCS DE l-131 activity of 0.05 Ci/gm, it would require a spike factor of nearly 10,000 to obtain a Post-Trip RCS DE l-131 activity near 38 pCi/gm. _ With a Post. Trip RCS DE-l-131 activity of 38 Ci/gm, an increase in the allowable leak rate could impact the 10CRF100 limits. To accommodate for an i rease in the allowable leak rate by a factor of twenty, the r%oitant activity would need to be below 1.9 pCilgm. Two (2) of the severneen (17) post-trip data points from Braidwood exceeded 1.9 pCl/gm. Both occurred during cycles with fuel defects. BrLidwood Unit 1 is currently operating with no fuel defects. The defense-in-depth measures mentioned in Method 2 will reduce the possibility of exceeding a small fraction of the 10 CFR100 limits should the post-trip iodine exceed 1.9 pCi/gm.
Conclusion The current Braidwood Unit 1 Cycle 7 RCS DE l-131 activity level has been relatively stable at approximately 3E-4 pCi/gm over the last six months.
Braidwood Unit 1 has operated the previous two cycles with high fuelintegrity. 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. The corresponding spiking factor would be expected to be much less 500. Therefore, with the current 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 iodine 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 iodine spiking factor limit will be met, the conservatisms in the offsite dose calculation I
(defense-in-depth arguments listed above) provide added assurance that the 10CFR100 limits, General Design Criteria (GDC) 19 critaria, and the requirements of NRC Generic Letter 95-05 will be satisfied.
G.
IMPACT OF THE PROPOSED CHANGE The requested change in the Unit 1 RCS DE l-131 activity limit from the currerst 0.35 pCilgm to 0.05 pCilgm provides additional margin for 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 pCilgm, the site maximum allowable leakage limit is 19.0 K:sgr Amfs\\ iodine 2. doc A 12
a
_ gpm (Room T/P). Utilizing the proposed 0.05 pCi/gm RCS DE l-131 activity limit, the revisod maximum site allowable leakage limit is 132.8 gpm (Room T/P). The predicted end-of-cycle 7 total leak rate is 122.3 gpm (Room T/P).1The calculated Exclusion Area Boundary (EAB) dose due to this leakage is 27.63 Rem,-
- (Refe ence 4). 'An evaluation of the control room dose attributable to a MSLB
~
~
1 accident scenrio demonstrated that the MSLB accident scenario was bounded -
by the existing Loss of Coolant Accident (LOCA) analysis, and that the maximum -
site allowable primary-to-secondary leakage is limited by the offsite dose at the EAd due to an accident-initiated spike. Therefore, the requirements of 10CFR100 and GDC 19 have been met.
Generic Letter 95-05 permits lowering the dose equivalent iodine activity las a.
means for accepting higher projected leakage rates provided justification for..
RCS DE l-131 activity below 0.35 pCi/gm is given. Should the RCS DE l-131 activity increase to the proposed 0.05 pCi/gm limit, the expected iodine spike should be less than the spike predicted by the NRC SRP. Higher iodine spiking.
factors may result with RCS DE l-131 activities less than 0.05 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 4
SRP criteria are conservative.
l This amendment request will not result in any changes to exitning 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 t
(
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 EWI steam generators. With
[
the replacement of the steam generators, the RC S DE l-131 activity limit will be returned to the standard value of 1.0 pCi/gm.
H.
SCHEDULE REQUIREMENTS
_This change permits a, extension of the Unit 1 Cycle 7 operating period to alkw i
. full cycle operation to the SG Replacement outage. In order to facilitate outage
' scheduling,. Comed requests that this proposed amendment be reviewed and e
' approved by April 1,1998.
l.
~ REFERENCES-
' 1. R. Assa letter to D. Farrar dated May 7,1994, transmitting Amendment 50 for Braidwood K:sgrpunfs\\ iodine 2. doc A.13 -
/
- ~. -
- 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 -
- 4. Braidwood Calculation BRW-97-078-M, Revision 4, Site Allowable Leak Rate Calculation for SG Interim Plugging Criteria
- 5. Braic' wood Calculation 95-011, Revision 5, Control Room Dose Calculation
- 6. C. Shirakiletter to D. Farrar dated July 26,1995, transmitting Amendment 167 for Zion Unit 1
- 7. J. Hosmer letter to NRC Document Control Desk, dated January 31,1997, requesting amendment to the Byron Unit 1 Technical Specifications
- 8. Byron /Braidwood Updated Final Safety Analysis Report (UFSAR)
- 9. Braidwood Calculation BRW-96-530-M, Revision 1, Determination RCS Dose Equivalent lodine 1131 Release Rate and Spike Factor Based on Plant Trip.
. Data
- 10. Westinghouse Letter CCE-97-305, " Comed Braidwood Unit 1, IPC Engineering CYCLESIM Runs' dated November 26,1997
- 11. August 14,1997 letter from H. Stanley to NRC Document Control Desk, "Braidwood Station Unit 1, Steam Generator Interirn Plugging Criteria 90 Day Report" 4
K:sgrp\\mfs\\ iodine 2. doc A 14
5 l
I 1
i Table A-1 Brandwood Statfori DE M31 Reactor Trip Data I
Post-Teip Pro Trip Pro Trip Post Trip Pro Trip Post Trip StendrSasse Meeknen
]
Teh=
- ovver et lodine DE k131 Post Trip Imane DE8-131 CVFbw CV Flow Fuel Defects Release Role Rolesee lodine Spike i
)
Evers Uria Cvele Dase/Tnme Tre (%)
Dese/Thne (uCuge)
Desernme (uC3gm)
(gpm)
(gpm)
(Y/N)
(CWhr)
State (Clk)
Factor Y - 1 sod from j
1 1
2 1/12/90 1324 99 1/129010.15 6 30E-04 1/12/90 16 25 2.80E-03 118 118 detwe 0017 1 0373 82.7 1
Y - 1 rod from 2
1 2
6/8/90618 100 6/780745 410E-03 6/8/90030 2 20E41 118 116 debris 01108 88 8078 799 8 I
Y - 1 rod from 3
1 2
9/2%901735 99 9/2%'90 8 JO 2 50E-03 9~2%90 20 05 210E-01 115 64 detw6s 00659 80 8817 1227.5 Y - 1 roe from 4
1 2
12/1101647 99 9 12/1/90 7 55 2 30E-03 12/1/90 20 35 2 26E41 117 116 detwis 0 0616 912315 1480 i
Y - 1 rod from l
5 1
2 12/30/90 0 21 99 7 12/30907.50 2 20E43 12/30r9011.30 1.30E41 75 75 debne 00385 50 4946 1312.1 N - possidy 1 rod but couhs 6
1 3
11611017 99 9 115917:35 8 89E44 11/8/91 2:45 187E44 114 110 not Ihd 00232 0 0142 06 N - possitdy 1 sed but could 7
1 4
1/7/93 13 14 100 1/7/93 7.30 2 67E43 1/7/9316'JO 2.34E-03 113 90 not fhd OD892 08292 9.1 a/116412.30 4.41 E-03 8/11/94 16 30 7.05E43 Y - 1 rod 2.7015 231.2 8
1 5
8/11/94 10 28 100 8/11447:30 4 47E-04 8/11/94 20 15 5.39E43 114 75 found 0 0117 4/9/95 8:10 2.70E43 41/951040 2.74E-03 Y - 1 rod 1.085 98.5 9
1 5
4S95311 98 9 4%95800 4 07E44 4/%9513-40 2.42E-03 116 116 found 00100 1
6 None 1
7 None nm w.s w
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n Y - 1 rod due 10 2
2 8/1/91 23 54 87.5 8/1/91840 2 21E43 8/2/91236 3 99E-03 118 118 to debris 0 0607 1 2573 21.1 11 2
3 2/25/92 14 45 100 2/25/92 7.25 7 91E44 2/25/92 17:45 7.00E44 121 92 N
O0219 0.1997 87 i
__ 2 2
3 3/15e92 6 20 100 3/14/92 7.50 9 40E-04 3/15/92 8.30 8.10E-04 122 75 N
O0262 0.2103 8
13 2
3 9/10/92 21 28 100 9/10/92 7.3/
1.2 7E-03 9/11/92 005 5 55E44 122 120 N
0 0354 0 0884 25 l
14 2
3 11/14/92 18 52 99 11/14/92 7.40 1.22E43 11114/92 22 20 7.04E 04 122 121 N
0034
(*.1541 45 10G93 300 1.59E+00 10/3/93 5 05 3.29E+00 1334 9 3471.7 15 2
4 10093150 100 10/2/93 7:45 140E-02 10r3/93 7.20 140E+00 120 120 Y - 2 rods 0 3845 4/5/94 1835 1.98E+00 802.17 1482 6 16 2
4 4/5/94 15 39 100 4/5/94900 197E-ts2 4/5/942200 159E+00 120 120 Y - 2 rode 05411 N - defect 8/2/94 20 20 1.39E43 removed 02682 67 17 2
4 8/2/94 17 49 99 8 8/2/94830 2 41E-03 8/3/94005 1.22E-03 70 71 in Apr814 0 0395 l
2 5 None 2
6 None codene2 m.
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