[Federal Register Volume 62, Number 180 (Wednesday, September 17, 1997)]
[Notices]
[Pages 48899-48903]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-24675]
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NUCLEAR REGULATORY COMMISSION
[Docket Nos. STN 50-456 AND STN 50-457]
Commonwealth Edison Company; Notice of Consideration of Issuance
of Amendments to Facility Operating Licenses, Proposed no Significant
Hazards Consideration Determination, and Opportunity for a Hearing
The U.S. Nuclear Regulatory Commission (the Commission) is
considering issuance of amendments to Facility Operating License Nos.
NPF-72 and NPF-77 issued to the Commonwealth 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 (I-131) concentration in the primary coolant
from 0.35 to 0.10 microcuries per gram for the remainder of the present
Braidwood, Unit 1, operating cycle (i.e., Cycle 7); this operating
cycle is projected to end in September 1998.
Before issuance of the proposed license amendments, the Commission
will have made findings required by the Atomic Energy Act of 1954, as
amended (the Act) and the Commission's regulations.
The Commission has made a proposed determination that the
amendments requested involve no significant hazards consideration.
Under the Commission'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 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.
Generic Letter 95-05, ``Voltage-Based Repair Criteria For
Westinghouse Steam Generator Tubes Affected By Outside Diameter Stress
Corrosion Cracking,'' allows lowering of the RCS [Reactor Coolant
System] DE I-131 activity as a means for accepting higher projected
[[Page 48900]]
leak rates if justification for equivalent I-131 below 0.35
Ci/gm [microcuries per gram] is provided. Four methods for
determining the impact of a release of activity to the public were
reviewed to provide this justification. These four methods are as
follows:
Method 1: NRC NUREG 0800, Standard Review Plan (SRP) Methodology.
Method 2: Methodology described in a report by J.P. Adams and C.L.
Atwood, ``The Iodine Spike Release Rate During a Steam Generator Tube
Rupture,'' Nuclear Technology, Vol. 94, p. 361 (1991) using Braidwood
Station reactor trip data.
Method 3: Methodology described in a report by J.P. Adams and C.L.
Atwood, ``The Iodine Spike Release Rate During a Steam Generator Tube
Rupture,'' Nuclear Technology, Vol. 94, p. 361 (1991) using normalized
industry reactor trip data.
Method 4: Methodology described in a draft EPRI Report TR-103680,
Revision 1, November 1995, ``Empirical Study of Iodine Spiking in PWR
Plants''.
The effect of reducing the RCS DE I-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 I-131
activity limit of 1.0 Ci/gm, the maximum site allowable
leakage limit was calculated, in accordance with the NRC SRP
methodology, to be 9.33 gallons per minute (gpm). The 9.33 gpm
allowable leakage limit was calculated for leakage at the normal
operating reactor coolant temperature and pressure. This corresponds to
a room temperature and pressure leakage limit of 6.63 gpm. ComEd has
evaluated the reduction of the RCS DE I-131 activity to 0.10
Ci/gm along with the increase of the allowable leakage to 94
gpm (66.3 gpm at room temperature and pressure) and has concluded:
--assuming a spike factor of 500, the maximum activity released is not
changed, and
--the offsite dose, including the iodine spiking factor, will be less
than the 10CFR100 limits.
Based on the NRC SRP methodology for dose assessments and assuming
the iodine spike factor of 500 is applicable at the new 0.1
Ci/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) is bounded by the activity released to the environment
from the Loss of Coolant design basis accident (1290 Curies).
Therefore, the Control Room dose, due to the MSLB accident scenario, is
bounded by the existing Loss of Coolant Accident (LOCA) analysis. The
maximum site allowable primary-to-secondary leakage is limited by the
offsite dose at the Exclusion Area Boundary due to an accident-
initiated spike.
The report by J.P. Adams and C.L. Atwood, ``The Iodine 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 Ci/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 spike factor of 500
assumed in the NRC SRP methodology. Braidwood Unit 1 Cycle 7 is
currently operating with no fuel defects and an RCS DE I-131 activity
of approximately 3E-4 Ci/gm. The seven previous trips, with no
fuel defects, had steady-state iodine values that are reasonably close
to the current operating conditions. It is, therefore reasonable to
conclude that, assuming continued operation with little to no fuel
defects, the calculated spike factors from these events would reflect
an actual event for Unit 1 Cycle 7, i.e., the spike factor will be less
than 500.
Since some of the spike factors were greater than 500 when the RCS
DE I-131 activity, prior to the accident, was less than 0.3
Ci/gm, ComEd examined the conservatisms in the current release
rate calculation. The primary reason for these high ratios (up to
12,000) is not because the absolute post-trip release rate is high
(factor numerator), but rather because the steady-state release rate
(factor denominator) is low. The Braidwood specific data resulted in
six (6) events with a calculated release rate spike factor greater than
500. It is not expected, based upon the Unit 1 Cycle 7 fuel conditions,
that a spiking factor greater than 500 would occur. The revised RCS DE
I-131 activity limit will also ensure that the operating cycle will not
continue if significant fuel defects develop.
In order to evaluate the Braidwood specific data against the NRC
SRP methodology, the release rate for a steady-state RCS DE I-131
activity of 1.0 Ci/gm 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 Ci/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 Ci/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 Ci/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 I-131 activity between 0.09
Ci/gm and 0.6 Ci/gm. Only one (1) of the eight (8)
data points had a pre-trip DE I-131 activity below 0.1 Ci/gm.
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 I-131 concentrations are below 0.3 Ci/gm.
Since the Braidwood data does not include data near 0.1 Ci/gm
(the requested new TS limit), it is appropriate to use the Adams and
[[Page 48901]]
Atwood database near 0.1 Ci/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 I-131 activity between 0.05
Ci/gm and 0.15 Ci/gm. 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 I-131 activity is at or below 0.1 Ci/gm. The
highest spike factor seen in the Adams and Atwood report near a Pre-
Trip RCS DE I-131 activity of 0.1 Ci/gm was 1160 (at 0.093
Ci/gm). This release rate is less than the calculated
Braidwood maximum value of 1373 Ci/hr.
The predominant factors in calculating the offsite dose are the
post-trip iodine release rate from the fuel and the flowrate at which
the activity is being released to the environment, not whether the
spike factor is greater than or less than 500. The post-trip DE I-131
release rate will determine the level of activity in the RCS that will
be released. The flowrate will determine at what rate this activity is
released to the environment. Method 3, which used a different approach
in the Adams and Atwood report, concluded that, at a 95% confidence of
a 90 percentile, the post-trip iodine release rate was bounded by 1.09
Ci/hr-MWe. For Braidwood Station, which has a MWe rating of 1175, the
post-trip iodine release rate, at a 95% confidence of a 90 percentile,
should not exceed 1280 Ci/hr. One (1) of the seventeen (17) reactor
trips from Braidwood exceeded 1280 Ci/hr. This reactor trip had a post-
trip iodine release rate of 1335 Ci/hr (spike 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
Ci/gm, all but one data point is bounded by the 1373 Ci/hr
release rate. This one data point is bounced [bounded] by the 95%
confidence. This data suggests that the possibility for a post-trip
iodine fuel release rate to exceed 1373 Ci/hr, when the pre-trip RCS DE
I-131 concentration is at or below 0.1 Ci/gm, is small.
In the fourth method, the results from a Draft Electric Power
Research Institute (EPRI) Report TR-103680, Rev. 1, November 1995,
``Empirical Study of Iodine Spiking In PWR Power Plants'' were applied.
The objective of the EPRI study was to quantify the iodine spiking in a
postulated Main Steam Line Break/Steam Generator Tube Rupture (MSLB/
SGTR) accident sequences. In the EPRI report, an iodine spike factor
between 40 and 150 was determined to match data from existing plant
trips. The maximum iodine spike factor value of 150 was applied to a
steady-state equilibrium RCS DE I-131 activity of 0.33 Ci/gm.
The 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 iodine activity, which is a small fraction of
the activity predicted by the NRC SRP methodology, it can be expected
that, for the proposed 0.10 Ci/gm limit under an MSLB/SGTR
accident sequence, the post-accident iodine activity would typically be
a small fraction of the 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 1.0 Ci/gm and a spike factor of 500, the Post-
Trip RCS activity two hours after the event would be near 35.5
Ci/gm. At an RCS DE I-131 activity of 0.1 Ci/gm, it
would require a spike factor of nearly 5000 to obtain a Post-Trip RCS
DE I-131 activity near 35.5 Ci/gm. With a Post-Trip RCS DE I-
131 activity of 35.5 Ci/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 Ci/gm. None of the seventeen (17)
post-trip data from Braidwood has exceeded 3.55 Ci/gm. The
maximum Post-Trip RCS activity seen at Braidwood is 3.29 Ci/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) used to predict iodine
spiking is conservative. Although dose projections indicate with
confidence that the iodine spiking factor limit will be met, the
conservatisms in the offsite dose calculation provide added assurance
that the 10CFR100 limits, General Design Criteria (GDC) 19 criteria,
and the requirements of NRC Generic Letter 95-05 will be satisfied if
the iodine spike factor exceeds 500 or the post-trip fuel release rate
exceeds 1373 Ci/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 near 3E-
4 Ci/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 iodine partition factor of 0.1 is more realistic (per Table
15.1-3 of Byron/Braidwood UFSAR) than the 1.0 valued [value] (no
partitioning) used in the offsite dose calculation. This reduces the
calculated dose by 90%.
4. Primary-to-secondary leakage is not expected to be at the TS
limit (150 gpd) in each of the four SGs prior to the event. Currently,
minimal primary-to-secondary leakage (less than 5 gpd) exists at
Braidwood Unit 1.
5. The activity in the RCS is not expected to increase
instantaneously with the spike in 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.
These proposed changes do not result in a significant increase in
the consequences of an accident previously analyzed.
The RCS DE I-131 activity limit is not considered as a precursor to
any accident. Therefore, this proposed change does not result in a
significant increase in the probability of an accident previously
analyzed.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The changes proposed in this amendment request conservatively
reduce the Unit 1 RCS DE 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 proposed change does not create the possibility
of a new
[[Page 48902]]
or different kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction in
a margin of safety.
NRC Generic Letter 95-05 allows lowering of the RCS dose equivalent
iodine as a means for accepting higher projected leakage rates provided
justification for the RCS DE I-131 activity below 0.35 Ci/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 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
margin 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 Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission 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 result, for example, in derating or shutdown of the facility,
the Commission 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 comments
received. Should the Commission take this action, it will publish in
the Federal Register a notice of issuance and provide for opportunity
for a hearing after issuance. The Commission expects that the need to
take this action will occur very infrequently.
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 Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments 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
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 room located 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 filed by the above date, the
Commission or an Atomic Safety and Licensing Board, designated by the
Commission or by the Chairman of the 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 forth with particularity the interest 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 permitted with particular reference to the
following factors: (1) the nature 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
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought 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 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. Contentions shall be limited to matters within
the scope of the amendments 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
[[Page 48903]]
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 Commission 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 determination 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.
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
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission'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 Commission,
Washington, DC 20555-0001, and to Michael I. Miller, Esquire; Sidley
and Austin, One First National 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 Commission, 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) (i)-(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 Commission'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 Nuclear Regulatory Commission.
M. D. Lynch,
Senior Project Manager, Project Directorate III-2, Division of Reactor
Projects, Office of Nuclear Reactor Regulation.
[FR Doc. 97-24675 Filed 9-16-97; 8:45 am]
BILLING CODE 7590-01-P
