ML20211D691
| ML20211D691 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 09/11/1997 |
| From: | Lynch M NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20211D696 | List: |
| References | |
| NUDOCS 9709290207 | |
| Download: ML20211D691 (12) | |
Text
7590-01-P UNITED STATEF NUCLEAR REGULATORY __COMMIS$10N COMMONWEALTH EDISON COMPANY DOCKET NOS. STN 50-456 AND STN 50-457 NOTICE OF CONSIDERATION OF ISSUANCE OF AMENDMENTS TO fACjllTY OPERATING LICENSES. PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMI'4ATION. AND OPPORTUNITY FOR A HEARING The U.S. Nuclear Regulatory Comission (the Comission) is consid6 ring issuance of amendments to Facility Operating License Nos. NPF-72 and NPF-77 issued to the Comonwealth Edison Company (Comed, the licensee) for operation of the Braidwood Station, Units 1 and 2, located in Will County, Illinois.
The proposed amendments would revise Technical Specifications (TS)
Section 3.4.8, Figure 3.4-1 and Table 4.4-4 and also revise TS Bases Section 3/4.4.8. The revisions reduce the TS maximum allowable dose equivalent (DE) iodine-131(1-131) concentration in the primary coolant from 0.35 to 0.10 l
microcuries per gram for the remainder of the present Braidwood, Unit 1, l
operating cycle (i.e., Cycle 7); this operating cycle is projected to end in September 1998.
Before issuance of the proposed license amendments, the Comission will have made findings required by the Atomic Energy Act of 1954, as amended (the Act) and the Comission's regulations.
The Comission has made a proposed detemination that the amendments requested involve no significant hazards consideration. Under the Comission's regulations in 10 CFR.50.92, this means that operation of the facility in accordance with the proposed amendments would not (1) involve a D
DO K 05 56 P-PDR e
i 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. As required by 10 CFR 50.91(a),
the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:
1.
The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
l Generic Letter 95-05, " Voltage-Based Repair Criteria For Westinghouse i
Steam Generator Tubes Affected By Outside Diameter Stress Corrosion I
Cracking," allows lowering of the RCS (Reactor Coolant System)if DE I-131 activity as a means for accepting higher projected leak rates justification for equivalent I-131 below 0.35 pCi/gm [microcurtas per gram)isprovided.
Four methods for determining t1e 1mpact 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)usingBraidwoodStationreactortripdata.
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 reacto' trip data.
Method 4: Methodology described in a draft EPRI Report TR-103680 Revision 1, November 1995, " Empirical Study of Iodine Spiking in PWR Plants".
The effect of reducing the RCS DE l-131 activity limit on the amount of activity released to the environment remains unchanged when the maximum site allowable )rimary-to-secondary leak rate is proportionately increased and tie iodine release rate spike factor is assumed to be 500 in accordance with the SRP. With an RCS DE I-131 activity limit of 1.0 pC1/gm, the maximum site allowable leakage lireit 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 ana pressure. This corresponds to a room temperature-and pressure leakage limit of 6.63
e i
gpm. Comed has evaluated the reduction of the RCS DE I-131 activity to 0.10 pCi/gm along with the increase of the allowable leakage to 94 gpm i
(66.3gpmatroomtemperatureandpressure)andhasconcludedt j
not changed, pike factor of 500, the maximum activity released is i
assuming a s and the offsite dose, including the iodine spiking factor, will be 1
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 I-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 (main steamline break) 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)f Coolant design basis accident (1290 Curies).
is bounded by the activity released to the environment from the Loss o Therefors, the Control Room dose, due to the MSLB acctdent 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.
i 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 iodine activity from the fuel rod to the RCS, is conservative when the RCS DE I-131 concentration is greater than 0.3 #Ci/gm.
In order to evaluate whether a release rate spike factor of 500 is conservative below 0.3 #C1/gm, actual operating data from the previous reactor trips of Braidwood Units 1 and 2, with and without fuel defects, were reviewed and analyzed using the methodology presented in Section II.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 spf ke factor of 500 assumed in the l
l
NRC SRP methodology.
Braidwood Unit 1 Cycle 7 is currently opercting with no fuel defects and an RCS DE I-131 activity of approximately 3e.-4 pCi/gs. The seven previous trips with no fuel defects had stea6 state iodine values that are reaso,nably close to the cur, rent 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 reflect an actual event for Unit 1
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Cycle 7, i.e., the spike for. tor will be less than 500.
Since some of the. spike factors 6:ere greater than 500 when the RCS DE 1-131 activity, prior to the accident was less than 0.3 pC1/gm Co examinedtheconservatismsinthecurr,entreleaseratecalculatIon.mEdThe primary reason for these high ratios ( p to 12,000) is not because the absolute post-trip release rate is hig (factor numerator), but rather because the steady-state release rate factor denominator) is low. The Braidwood specific data resulted in si (6) events with a calculated release rate spike factor greater than 500.
It is not expected, based u>on the Unit 1 Cycle 7 fuel conditions, that a spiking factor greater tian 500 would occur. The revised RCS DE I-131 activity limit will also ensure that the o defects develop. perating cycle will not continue if significant fuel In order to evaluate the Braidwood specific data against the NRC SRP methodology, the release rate for a steady-state RCS DE I-131 activity of 1.0 pti/gm was calculated. Using the Braidwood specific data, the pre-trip steady-state release rate is 27.5 C1/hr.
Using a release rate spike factor of 500 for the accident-initiated spike, the post-trip maximum release rate would be 13,733 C1/hr (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 #C1/gm, Braidwood is also requesting an increase in the allowable primary-to-secondary leak rate.
By decreasing the TS RCS DE I-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 Ci/hr. None of the Braidwood data exceeds 1373 C1/hr, although eight (8) of the 168 data points in the Adams and Atwood report exceed 1373 Cs/hr.
The eight (8) data points had a pre-trip RCS DE I-131 activity between 0.0g pC1/gm and 0.6 pCi/gm.
Only one (1) of the eight (8) data points had a pre-trip DE I-131 activity below 0.1 pCi/gs.
-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 pCi/gm. SincetheBraIdwood data does not include data near 0.1 pCi/gm (the requested new TS limit),
it is appropriate to use the Adams and Atwood database near 0.1 pCi/ p.
to determine if a spike factor of 500 O 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 pCi/gm. Thirty-four(34)
5
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l i
of these forty-two (42) Adams and Atwood database,130 of the 500. Usin the ent're i
points (77 ) have an todine spike factor less than 500.
Therefore it is reasont le to assume that a spike factor of 500 would not be exc,eeded i
for a majority of the events if an MSLB accident were to occur while the RCS DE I-131 activity is at or below 0.1 #C1/gm.
The highest spike factor seen in the Adams and Atwood report near a Pre-Trip RCS DE I-131 activity of 0.1 #C1/gm was 1160 (d maximum va/gm).
at 0.093 pC1 This release rate is less than the calculated Braidwoo lue of 1373 C1/hr.
i 1
1 l
The predominant factors in calculating the offsite dose are the post-9 trip iodine release rate from the fuel and the flowrate at which the 3
i i
activity is being released to the environment, not whether the spike i
factor is greater than or less than 500. The post-trip DE I-131 release rate will determine the level of activity in the RCS that will be l
released. The flowrate will determine at what rate this activity is released to the environment.
Method 3, which used a different approach I
l in the Adams and Atwood report, concluded that, at a 95% confidence of a 90 percentile, the post-trip todine release rate was bounded by 1.09 Ci/hr-MWe.
For Braidwood Station, which has a MWe rating-of 1175, the i
post-trip lodine release rate, at a 95% confidence of a 90 percentile, should not exceed 1280 C1 trips from Braidwood exce/hr. One (1) of the seventeen (17) reactor eded 1280 C1/hr. This reactor trip had a post-l trip iodine release rate of 1335 Ci/hr (spika factor of 3471). The second highest post-trip iodine release rate from the Braidwood data was 802 Ci/hr (spike factor of 1483).
For the combined Adams /Atwood and Braidwood data sets, below 0.1 #C1/gm, i
all but one data point is bounded by the 1373 C1/hr release rate.- This-j one data point is bounced [ bounded) by the 95% confidence. This data suggests that the possibi19ty for a post-trip iodine fuel release rate to exceed 1373 C1/hr, when the or below 0.1 pCi/gm, is small. pre-trip RCS DE I-131 concentration is at i
In the fourth method, the results from a Draft Electric Power Research
. Institute (EPRI) Report TR-103680, Rev.1. November 1995, ' Empirical Study of Iodine Spiking In PWR Power Plants" were applied. The j
objective of the EPRI study was to quantify the iodine spiking in a postulated Main Steam Line Break / Steam Generator Tube Rupture (MSLB/SGTR) accident sequences.
In the EPRI report, an iodine spike factor between 40 and 150 was determined to match data from existing i
plant trips. The maximum iodine spike factor value of 150 was applied j-
-to a steady-state equilibrium RCS DE I-131 activity of 0.33 pC1/ge. The j
resulting two-hour average iodine concentration for a postulated MSLB/SGTR accident sequence was determined to be 3.1 #Ci/gm. Since the EPRI report is based on industry data and the EPRI method predicted a post-accident todine 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 pCi/gm limit under an MSLB/SGTR accident sequence, the post accident iodine activity would typically be a small fraction of the l
~
. ~
RCS DE I-131 activity predicted by the NRC SRP methodology.
For Braidwood, using the SRP methodology with an RCS DE I-131 activity of 4
l.0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> #Ci/gm and a spike factor of 500, the Post-Trip RCS activity two s after the event would be near 35.5 pCi activity of 0.1 #C1/gm, it would require a s/gm. At an RCS DE l-131 pike factor of nearly 5000 to obtain a Post-Trip RCS DE I-131 activity near 35.5 pC1/gm. With a Post-Trip RCS DE l-131 activity of 35.5 pC1/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 factor of ten, the resultant activity would need to be below 3.55 pCi None of the The maximum (17) post-trip data from Braidwood has ex/gm.
seventeen ceeded 3.55 pCi/gm.
Post-Trip RCS activity seen at Braidwood is 3.29 pC1/gm at approximately three hours after the event.
Based on evaluations by the four methods above, Braidwood can conclude that the current methodology (Method 1) ions indicate with conf i
used to predict icdine spiking is conservative. Although dose project that the iodine spiking factor limit will be met, the conservatisms in i
the offsite dose calculation provide added assurance that the 10CFR100 1
limits, General Design Criteria GDC 19 criteria, and the requirements of HRC Generic Letter 95-05 will(be s)atisfied if the iodine spike factor exceeds 500 or the post-trip fuel release rate exceeds 1373 C1/hr.
]
These conservatisms include, but are not limited to:
1.
The RCS DE I-131 activity is more likely to be less than the TS limit.
With the current Braidwood Unit 1 RCS DE I-131 activity l
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.
Iodine partitioning is not accounted for in the faulted SG. With the high pH of the secondary water, some partitioning is expected to occur.
An todine partition factor of 0.1 is more realistic r
(per Table 15.13 af Byron /Braidwood UFSAR) than the 1.0 valued
- value) duces the calculated dose by 90%.
'his re (no partP.ioning) used in the offsite dose calculation.
4.
Primary-to-secondary leakage is not expected to be at the TS limit j
(150gpd in each of the four SGs prior to the event. Currently, minimal p)rimary-to-secondary leakage (less than 5 gpd) exist 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.
0 7-6.
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 I 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.
These proposed changes do not result in a significant increase in the l
consequences of an accident previously analyzed.
The RC$ DE I-131 activity limit is not considered as a precursor to any l
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 I-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 propo:ed change does not create the possibility of a new or d< fferent 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 l.etter g5-05 allows lowering of the RCS dose equivalent iodine as a means for. accepting higher projected leakage rates provided justification for the RCS DE I-131 activity below 0.35 pC1/gm is provided.
Four methods for determining the fuel rod iodine release rates and spike factors during an accident were reviewed.
Each of these methods utilized actual industry data, including Braidwood Units 1 and 2, for pre-and post-reactor trip RCS DE I-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 I-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
8-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 Rem. 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 saargin of safety.
Therefore, based on the above evaluation, Comed has concluded that these changes involve no significant hazards considerations.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied.
Therefore, the NRC staff proposes to determine that the amendments requested involve no significant hazards consideration.
The Comission is seeking public coments on this proposed determination. Any coments received within 30 days after the date of publication of this notice will be considered in making any final determination.
Normally, the Comission will not issue the amendments until the expiration of the 30-day notice period. However, should circumstances change during the notice period such that failure to act in a timely way would i
result, for example, in derating or shutdown of the facility, the Comission may issue the license amendments before the expiration of the 30-day notice period, provided that its final determination is that the amendments involve no significant hazards consideration. The final determination will consider all public and State coments received. Should the Comission take this action, it will publish in the FEDERAL REGISTER a notice of issuance and i
provide for opportunity for a hearing after issuance. The Comission expects that the need to take this action will occur very infrequently.
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6 Written comments may be submitted by mail to the Chief, Rules and Directives Branch, Division of Freedom of Information and Publications Services, Office of Administration, U.S. Nuclear Regulatory Comission, Washington, DC 20555-0001, and should cite the publication date and page i
number of this FEDERAL REGISTER notice. Written coments may also be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of written comments received may be examined at the NRC Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for hearing and petitions for leave to intervene is discussed below.
By October 17. 1997
, the licensee may file a request for a hearing a
with respect to issuance of the amendments to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene.
Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's " Rules of Practice for Domestic Licensing Proceedings" in 10 CFR Part 2.
Interested persons should consult a current copy of 10 CFR 2.714 which is available at the Commission's Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at the local public document ro6m Iccated at the Wilmington Public Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.
If a request for a hearing or petition for leave to intervene is flied by the above date, the Commission or an Atomic Safety and Licensing Board, designated by the Commission or by the Chairman of the
10 Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the designated Atomic Safety and Licensing Board will issue a notice of hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene shall set 1
forth with particularity the intsrest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be pemitted with particular reference to the following factors:
(1)thenature of the petitioner's right under the Act to be made party to the proceeding; (2) the nature and extent of the petitioner's property, financial, or other interest in the proceeding; and (3) the possible effect of any order which may be entered in the proceeding on the petitioner's interest. The petition should also identify the specific aspect (s) of the subject matter of the proceeding as to which petitioner wishes to intervene. Any person who has filed a petition for leave to intervene or who has been admitted as a party may amend the petition without requesting leave of the Board up to 15 days prior to the first prehearing conference scheduled in the proceeding, but such an amended petition must satisfy the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference ccheduled in the proceeding, a petitioner shall file a supplement to the petition to intervene which must include a list of the contentions which are s'ought to be litigated in the matter.
Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted.
In addition, the petitioner shall provide a brief explanation of the bases of the contention and a concise statement of the alleged facts or expert opinion
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which support the contention and on which the petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expert opinion.
Petitioner must provide sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact.
i Contentions shall be limited to matters within the scope of the amendments l
under consideration. The contention must be one which, if proven, would entitle the petitioner to relief. A petitioner who fails to file such a supplement which satisfies these requirements with respect to at least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing, including the opportunity to present evidence and cross-examine witnesses.
If a hearing is requested, the'Comission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held.
If the final deternination is that the amendments requested involve no significant hazards consideration, the Commission may issue the amendments and make them immediately effective, notwithstanding the request for a hearing.
Any hearing held would take place after issuance of the amendments.
If the final determination is that the amendments requested involve a significant hazards consideration, any hearing held would take place before the issuance of any amendment.
r
, A request for a hearing or a petition for leave to intervene must be filed with the Secretary of the Commission, U.S. Nuclear Regulatory Comission, Washington, DC 20555-0001, Attention: Rulemakings and Adjudications Staff, or may be delivered to the Comission's Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, by the above date.
A copy of the petition should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Comission, Washington, DC 20555-0001, and to Michael I. Miller, Esquire; Sidley and Austin, One First National l
Plaza, Chicago, Illinois 60690, attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended petitions, supplemental petitions and/or requests for hearing will not be entertained absent a determination by the Comission, the presiding officer or the presiding Atomic Safety and Licensing Board that the petition and/or request should be granted based upon a balancing of the factors specified in 10 CFR 2.714(a)(1)(1)-(v) and 2.714(d).
For further details with respect to this action, see the application for amendment dated September 2,1997, which is available for public inspection at the Comission's Public Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at the local public document room located at the Wilmington Public Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.
Dated at Rockville, Maryland, this 11th day of September 1997.
FOR THE UCLEAR REGULATORY COMMISSION n>
M. D. Lynch, nior Project Manager Project Directorate III-t Division of Reactor Pro lects Office of Nuclear Reactar Regulation U