[Federal Register Volume 65, Number 94 (Monday, May 15, 2000)]
[Notices]
[Pages 31017-31021]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-12129]


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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-346]


FirstEnergy Nuclear Operating Company, (Davis-Besse Nuclear Power 
Station); Exemption

I

    FirstEnergy (the licensee) is the holder of Facility Operating 
License No. NPF-3, which authorizes the operation of the Davis-Besse 
Nuclear Power Station (DBNPS). The license states that the licensee is 
subject to all rules, regulations, and orders of the Nuclear Regulatory 
Commission (NRC or the Commission) now or hereafter in effect.
    The Commission is taking an action to approve this request prior to 
publication in the Federal Register of its Environmental Assessment and 
Finding of No Significant Impact. In accordance with 10 CFR 51.13, the 
Commission has determined that emergency circumstances are present to 
support the issuance of this exemption prior to publication in the 
Federal Register in that failure to act in a timely way would result in 
prevention of resumption of plant operation.
    The facility consists of a pressurized-water reactor at the 
licensee's site located in Ottawa County, Ohio.

II

    The DBNPS is planning to implement a plant modification during the 
twelfth refueling outage, which is scheduled to end in May 2000. The 
modification will change the equipment used to prevent boric acid 
precipitation following certain loss-of-coolant accidents (LOCAs) to 
enhance the flow of water through the core, thus controlling the 
accumulation of boric acid in the core and preventing boric acid 
precipitation.
    The Code of Federal Regulations at 10 CFR 50.46 provides acceptance 
criteria for the ECCS, including long-term cooling requirements in 
50.46(b)(5) and an option to develop the ECCS evaluation model (EM) in 
conformance with appendix K requirements (10 CFR 50.46(a)(1)(ii)). 10 
CFR part 50, appendix K, Section 1.D.1, in turn, requires that accident 
evaluations use the combination of ECCS subsystems assumed to be 
operative ``after the most damaging single failure of ECCS equipment 
has taken place.'' In addition, Appendix K Section I.A.4. specifies a 
requirement to assume decay heat generation rate is equal to 1.2 times 
the values for infinite operating time in a specified ANS standard.
    The proposed action would exempt the Licensee from the single-
failure requirement for very low probability scenarios under certain 
conditions. The exemption is limited to the systems required for 
preventing boron precipitation during the long-term cooling phase of a 
LOCA. In addition, the action would exempt the Licensee from the decay 
heat generation rate assumption specified in Appendix K, Section I.A.4.
    Specifically, DBNPS requested the following exemption by its 
letters dated March 15, and April 3, 2000: \1\, \2\
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    \1\ Campbell, Guy G., ``Request for Exemption from 10 CFR 50, 
Appendix K, for Boric Acid Precipitation Control Methodology (TAC 
No. MA7831),'' Letter to NRC from Vice President, Nuclear, 
FirstEnergy, Davis-Besse Nuclear Power Station, March 15, 2000.
    \2\ Campbell, Guy G., ``Supplemental Information Regarding the 
Request for Exemption from 10 CFR 50, Appendix K, for Boric Acid 
Precipitation Control Methodology (TAC No. MA7831),'' Letter to NRC 
from Vice President Nuclear, FirstEnergy, Davis-Besse Nuclear Power 
Station, April 3, 2000.

    FirstEnergy, with respect to the Davis-Besse Nuclear Power 
Station, is exempt from the single failure criterion requirement of 
10 CFR part 50, appendix K, Section I.D.1, with

[[Page 31018]]

respect to (1) Simultaneous failure of both the primary auxiliary 
spray method and the backup decay heat removal drop line method of 
controlling boron concentration due to failure of an emergency core 
cooling component that results in inability to initiate, or continue 
to operate, an active means of controlling core boron concentration, 
and (2) Not establishing that the backup decay heat removal drop 
line method of controlling boron concentration is otherwise in 
compliance with appendix K and 10 CFR 50.46(b)(5) requirements. 
Specifically, when establishing that boron precipitation will not 
occur in the decay heat removal system cooler, the Davis-Besse 
Nuclear Power Station credited flow through hot leg nozzle gaps and 
did not include all of the specific conservatisms required by 
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appendix K.

    The staff considers that the modifications would also require an 
exemption from the decay heat generation rate requirement contained in 
10 CFR part 50, appendix K, Section I.A.4.

III

    Certain LOCAs can result in a reactor coolant system (RCS) 
configuration in which the core is covered with boiling water and decay 
heat is transported from the core by steam while makeup water is 
provided to keep the core covered. This condition can result in 
accumulation of boric acid in the core since boric acid continues to be 
added via the makeup water, but little boric acid is removed by the 
steam. If too much boric acid accumulates, some might precipitate and 
prevent water from reaching the core to keep it cooled.
    The DBNPS reactor vessel (RV) is equipped with reactor vessel vent 
valves (RVVVs). The RVVVs will cause water to flow through the core to 
control buildup of boric acid when needed for all LOCA conditions 
except for (1) some LOCAs between the reactor coolant pumps (RCPs) and 
the RV and (2) decay heat generation rate comparable to approximately a 
month following extended operation at full power for some LOCAs. Active 
means of controlling boric acid concentration are provided to address 
the case when the RVVVs are not effective.
    In licensee event report (LER) 98-008 (October 1, 1998), DBNPS 
reported that for some small-break LOCAs, initiation of its active 
method of boron precipitation control (BPC) could cause steam binding 
in the suction piping of both decay heat removal (DHR) pumps. As part 
of the corrective action for LER 98-008, DBNPS committed to address all 
issues related to long-term LOCA BPC and to complete a related plant 
modification to improve the active methods by the end of the twelfth 
refueling outage. Improved active methods of BPC and the associated 
exemption request are in response to that commitment.
    With the improved active methods, if the RVVVs are not effective, 
then (1) the primary active method of BPC is a new means of supplying 
water to the pressurizer via the auxiliary spray line and (2) a new 
backup method will take water from an RCS hot leg via the DHR system 
drop line and return water to the RV via the core flood nozzles. DBNPS 
has stated that either method will provide sufficient flow of water 
through the core to provide BPC.
    The DBNPS identified the following single failure vulnerabilities 
for situations where the RVVVs cannot be established as being 
effective:
    (1) The primary BPC method is only connected to one train of high-
pressure safety injection (HPSI) and is subject to any single active 
component failure in the flow path. Thus, a backup method is needed.
    (2) The backup BPC method is potentially vulnerable to boron 
precipitation in the DHR cooler and to certain failure modes that are 
common to both the primary and backup BPC methods.
    In its March 15, and April 3, 2000, submittals, the DBNPS requested 
an exemption from certain requirements of the criteria. DBNPS justified 
its request on the basis of improvements over the existing methodology, 
conservatisms in calculations that result in over-prediction of the BPC 
problem, and a risk evaluation.

IV

    Two new active methods are planned for BPC: (1) A primary method 
using an improved auxiliary spray path into the pressurizer and (2) a 
backup method using flow into the DHR suction pipe from an RCS hot-leg 
pipe. A new pipe and new valves are being installed to accommodate the 
primary method. This path will supply about 250 gpm to the pressurizer, 
sufficient to fill the pressurizer in approximately an hour, after 
which BPC will be achieved by flow from the pressurizer into the 
reactor vessel via an RCS hot-leg. High-pressure injection (HPI) Pump 2 
will be used with ``piggyback'' suction from DHR/low-pressure injection 
(LPI) Pump 2. A failure anywhere in the flow path could result in 
failure of this method to provide water to the pressurizer.
    A backup method is provided in case the primary method fails. This 
method will use one of the two operating DHR/LPI pumps to take suction 
from the DHR drop line and to discharge a low flow rate into the 
reactor vessel via the core flood nozzles. The second DHR/LPI pump will 
be unthrottled and will continue to take suction from the emergency 
sump. The first pump will ensure a net flow of water through the core 
by withdrawing water from an RCS hot-leg while the second pump will 
ensure that makeup water is supplied to the RCS so that core cooling is 
ensured.
    If only one ECCS train is available, the backup method is not 
available since the available ECCS train must be used to ensure the 
water makeup function. Thus, failure of ECCS Train 2 will disable both 
the primary and the backup method for BPC. DBNPS reported the results 
of a common-mode failure evaluation of this condition that identified 
several areas where a single-failure could disable both the primary and 
backup BPC methods. We briefly audited this evaluation.
    The DBNPS assumed an initial RCS boric acid concentration of 1900 
ppm for the small break LOCAs for analysis of DHR cooler performance on 
the basis that, after the first few days of operation, the actual RCS 
concentration prior to the LOCA would be 1700 to 1800 ppm. Injection 
water was included from the borated water storage tanks at about 2800 
ppm and from the core flood tanks at about 4000 ppm. For the large and 
medium LOCAs, the 1900 ppm assumption was not used because much of the 
original water is lost from the RCS prior to injection, and the core 
flood tanks and borated water storage tank were assumed to inject into 
the RCS consistent with the LOCA RCS pressure calculations. This 
approach is acceptable because the amount of boron predicted to be in 
the core will be consistent with the sources of boron.
    The DBNPS assumed 1.0 times the American Nuclear Society (ANS) 
standard infinite operation decay heat generation rate for calculation 
of the DHR cooler aspects of the backup method, whereas Appendix K 
specifies 1.2. Although using 1.0 is more realistic and is suitable for 
probabilistic risk calculations, the calculation does not include the 
conservatism required by Appendix K. The DBNPS exemption request 
therefore encompasses not complying with the Appendix K calculational 
requirement. Realistically, when considered in conjunction with a 
likely hot leg nozzle gap that provides a boron dilution path, DBNPS 
has shown that BPC will be maintained through the cooler. This, in 
conjunction with the low probability of encountering the condition (as 
discussed below), demonstrates that use of an assumed 1.0 decay heat 
generation rate does not constitute an undue risk and is therefore, 
acceptable.

[[Page 31019]]

    Traditionally, core boric acid concentration evaluations use a 
solubility limit of the actual solubility reduced by four weight 
percent, an approach the staff has accepted in past Appendix K reviews 
to account for such items as solubility uncertainty and the non-uniform 
temperatures that may result in the RV. The DBNPS stated it used 4 
percent for its core analyses, but that it used a 90 percent of the 
solubility limit for the DHR cooler analysis. This reduced margin 
approach is reasonable and is acceptable for the DHR cooler analysis 
because the complex flow patterns and potential temperature non-
uniformities associated with the RV will not be present in the DHR 
cooler.
    The DBNPS found that, when the backup method is first initiated, 
core boric acid concentration in water initially entering the DHR 
cooler could exceed solubility limits due to the low DHR cooler 
temperature. In its March 15, 2000, submittal, DBNPS addressed this for 
the break conditions of concern by assuming there would be water flow 
from above the core into the downcomer via the hot leg nozzle gaps. The 
licensee calculated that this flow would maintain the core boric acid 
concentration below a value where the DHR cooler problem would occur 
until the backup method was performing its core dilution function. In 
its April 3, 2000, submittal, DBNPS requested that the exemption cover 
the calculated initial DHR cooler response since there was insufficient 
evidence to substantiate the claimed gap flow under the requirements of 
10 CFR 50.46 and Appendix K. The staff examined the licensee's 
evaluation using more realistic assumptions with respect to initial 
boron concentration, DHR cooler flow, decay heat rate, and DHR cooler 
temperatures. The staff concurs with the licensee that boric acid 
precipitation in the DHR cooler will not occur due to the conservative 
nature of their assumptions.
    The DBNPS did not attempt to address the change in core damage 
frequency (CDF) due to the planned modifications since BPC was not 
previously addressed in its plant risk assessment. Instead, it 
addressed the total risk associated with BPC. This assessment was based 
on several conservative assumptions. The DBNPS assumed that, for 
certain break size and location combinations, active BPC failure would 
cause core damage. This is consistent with the past regulatory approach 
to prevent conditions where boric acid precipitation could occur and 
the assumed failure to do so would be a failure to prevent core damage. 
Realistically, a significant quantity of boric acid would have to 
precipitate to lead to a loss of heat transfer that could cause core 
damage. This is an unquantified conservatism.
    The CDF is directly affected by the initiation rate of accidents of 
concern to BPC failure. For the bounding calculations, the DBNPS stated 
that it used generic LOCA rates of 5 x 10<SUP>-</SUP>\6\ and 
4 x 10<SUP>-</SUP>\5\ events/reactor-year for large and medium LOCAs, 
respectively, from NUREG/CR-5<SUP>-</SUP>\5\750. DBNPS then assumed 
that an active control method was needed for breaks lower than the 573-
foot elevation in the cold-leg RCP discharge piping for medium and 
large-break LOCAs, and that the break rate of concern was 25 percent of 
the large and medium LOCA frequency, leading to an initiation rate of 
1.1 x 10<SUP>-</SUP>\5\/reactor-year for active BPC. DBNPS then 
calculated the CDF due to boron precipitation to be approximately 
1.1 x 10<SUP>-</SUP>\7\/reactor-year \3\ (i.e., the frequency of an 
accident occurring in combination with a failure that renders both 
active BPC methods inoperable). DBNPS also reported the large early 
release frequency (LERF) associated with boron precipitation to be 
1.1 x 10<SUP>-</SUP>\11\/reactor-year. DBNPS concluded that the 
proposed plant modification would not be a significant contributor to 
the total CDF or LERF of the plant (approximately 
1.63 x 10<SUP>-</SUP>\5\ and 7.3 x 10<SUP>-</SUP>\8\/reactor-year, 
respectively). Regulatory Guide 1.174, ``An approach for Using 
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis,'' July 1998, considers an 
increase in risk to be very small if CDF and LERF are less than 
10<SUP>-</SUP>\6\ and 10<SUP>-</SUP>\7\, respectively. It further 
considers decreases in CDF and LERF to be satisfactory. The DBNPS 
predictions meet the guidance and are acceptable.
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    \3\ This calculation assumes the hot-leg nozzle gaps pass water 
with respect to calculating DHR cooler response. The effect of 
excluding hot-leg nozzle gap flow is addressed below. The values 
discussed here are only changed by a small amount.
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    The LOCAs where the RVVVs are initially ineffective are those 
involving roughly the lower half of the cold-leg piping between the 
RCPs and the RV. Considering symmetry and working with one side of the 
RCS that consists of one hot leg, a SG, and two cold legs, the actual 
fraction of concern was evaluated. Each cold leg has a segment between 
the RCP and the RV and between the SG and the RCP. Assuming each 
segment has about the same likelihood of breaking, and a hot leg 
section is about 3 times as long as a cold leg segment, and since the 
breaks of concern are in the cold leg between an RCP and the RV, and 
only a break in the lower half of a cold leg at that location is of 
concern, then the fraction of big pipe breaks of concern is (\1/2\)(2)/ 
(2+2+3) = 0.14. DBNPS assumed 0.25, a conservatism of a factor of 1.8 
with respect to this example.
    The DBNPS identified several other conservatisms in its risk 
assessment calculations. For example, with the exception of the backup 
method DHR cooler calculation where DBNPS used 1.0 times the decay 
heat, it used 1.2 times the decay heat for an infinitely irradiated 
core, thus predicting a faster boric acid concentration increase rate 
than would be expected, it took no credit for operator recovery 
actions, and, with the exception of the original DHR cooler analysis, 
it took no credit for hot-leg nozzle gaps. We agree that the above 
mentioned assumptions introduced conservatisms in the BPC related risk 
estimates assessed by DBNPS.
    The DBNPS addressed a potential increase in scope to include both 
HPI trains in the primary BPC method as opposed to only having HPI 
Train 2, thus eliminating part of the failure concern. It reported a 
CDF of 1.3 x 10<SUP>-</SUP>\8\/reactor-year for two trains, which it 
compared to the CDF of 1.1 x 10<SUP>-</SUP>\7\/reactor-year for only 
having HPI Train 2. DBNPS concluded that an increase in scope would not 
achieve a significant benefit in terms of risk reduction. NUREG-1.174 
considers an increase in CDF to be very small if it is less than 
10<SUP>-</SUP>\6\. In effect, the risk in moving from two trains to one 
would increase by 10<SUP>-</SUP>\7\, well within the 10<SUP>-</SUP>\6\ 
criterion. We therefore agree with the DBNPS conclusion and we find the 
decision to remain with one HPI train to be acceptable because a 
significant benefit would not be achieved by the increased scope.
    As discussed above, the backup BPC method was not shown to be 
functional using assumptions consistent with appendix K, nor was it 
shown to be functional using more realistic assumptions unless hot-leg 
nozzle gap flow was credited. Consequently, the DBNPS assumed a nozzle 
gap failure probability of 0.1, and predicted a CDF of 
1.3 x 10<SUP>-</SUP>\7\/reactor-year. We believe that a 0.1 failure 
probability is a reasonable bound and the actual failure probability 
would most likely be smaller. This, in conjunction with other potential 
bypass paths, such as associated with the core former-downcomer-thermal 
shield region and other applicable conservatisms is sufficient for us 
to accept the 0.1 probability used in this risk assessment. The 
increase from the previously calculated 1.1 x 10<SUP>-</SUP>\7\/
reactor-year is small enough that risk-associated

[[Page 31020]]

conclusions from the original analysis remain unchanged.
    The new connection between the Train 2 HPI and LPI systems 
introduces a potential for overpressurization of the Train 2 LPI system 
if valves are misaligned. The DBNPS evaluated this potential and the 
measures it will put in place to prevent valve misalignment, and 
reported an increase in CDF of less than 10<SUP>-</SUP>\8\/reactor-year 
due to valve misalignment. This is a negligible impact on the overall 
CDF of 1.63 x 10<SUP>-</SUP>\5\/reactor-year.
    The equipment modification addresses recognized weaknesses in the 
previous response to BPC and improves the defense-in-depth and safety 
margins should such conditions be encountered. DBNPS did not provide 
the calculated CDF and LERF that existed prior to the modification, but 
we judge the modification would reduce CDF and LERF because it 
addresses recognized weaknesses. DBNPS calculated that the CDF and LERF 
due to boron precipitation with the modification would be approximately 
1.1 x 10<SUP>-</SUP>\7\/reactor-year and 1.1 x 10<SUP>-</SUP>\11\/
reactor-year, respectively. These are small when compared to the total 
CDF and LERF from all causes of 1.63 x 10<SUP>-</SUP>\5\/reactor-year 
and 7.3 x 10<SUP>-</SUP>\8\/reactor-year, respectively. Further, 
Regulatory Guide 1.174 indicates that increases in CDF and LERF are 
very small if less than 10<SUP>-</SUP>\6\/reactor-year and 
10<SUP>-</SUP>\7\/reactor-year, respectively, and that decreases 
satisfy the relevant principles of risk-informed regulation. Here, the 
total contribution is smaller than what RG 1.174 considers to be small 
as an increase. These comparisons establish that the proposed exemption 
does not present an undue risk to public health and safety.

V

    Pursuant to 10 CFR 50.12, ``* * * The Commission may, upon 
application by any interested person or upon its own initiative, grant 
exemptions from the requirements * * * which are * * * authorized by 
law, will not present an undue risk to the public health and safety, * 
* * are consistent with the common defense and security (and) * * * 
special circumstances are present * * *.'' Special circumstances are 
present whenever, according to 10 CFR 50.12(a)(2)(ii), ``Application of 
the regulation in the particular circumstances would not serve the 
underlying purpose of the rule or is not necessary to achieve the 
underlying purpose of the rule * * *.''
    The requested exemption is authorized by law and does not affect 
the systems and processes associated with common defense and security.
    As identified above, the requirements of 10 CFR Part 50 apply to 
BPC and the DBNPS exemption request. With respect to the single-failure 
aspect of this evaluation, the underlying purpose of the single-failure 
criterion requirement is to assure long-term cooling performance of the 
ECCS in the event of the most damaging single-failure of ECCS 
equipment.
    As a licensing review tool, the single-failure criterion helps 
assure reliable systems as an element of defense in depth. As a design 
and analysis tool, it promotes reliability through enforced redundancy. 
Since historically, only those systems or components that were judged 
to have a credible chance of failure were assumed to fail, the 
criterion has been applied to such responses as valve movement on 
demand, emergency diesel generator start, short circuit in an 
electrical bus, and fluid leakage caused by gross failure of a pump or 
valve seal during long-term cooling. Reactor vessels or certain types 
of structural elements within systems, when combined with other 
unlikely events, were not assumed to fail because the probabilities of 
the resulting scenarios were deemed sufficiently small that they need 
not be considered. Certain passive failures 24 hours or more after 
initiation of a LOCA, such as pipe breaks, were not addressed as single 
failures because the compounded probabilities were judged sufficiently 
small that they could be discounted without affecting overall systems 
reliability.
    The single-failure criterion was developed without the benefit of 
numerical failure assessments. Regulatory requirements and guidance 
consequently were based upon categories of equipment and examples that 
must be covered or that are exempt, and do not allow a probabilistic 
consideration during routine implementation. Hence, a single failure 
that was not judged to be incredible (exempt) during development of the 
regulations, whether or not there is a substantial impact upon overall 
system reliability, will not meet the regulatory requirements. A non-
beneficial result is inconsistent with the objective of the single-
failure criterion, which was not intended to force changes if 
essentially no benefit would accrue. This is the case with potential 
failure of the active means of BPC.
    No US plants have encountered LOCA conditions where BPC was of 
concern. BPC measures are not needed for hot-leg breaks because water 
will flow through the core, thus preventing significant boric acid 
buildup, they are not needed if excore thermocouples indicate an 
adequate subcooling margin because there is no boiling to cause 
concentration of boric acid, and they are not needed for many of the 
remaining breaks until decay heat is low because water will flow from 
the core to the upper downcomer via the RVVVs, thus providing a 
mechanism to control accumulation of boric acid in the core. Active 
means for BPC are needed in case one of the above conditions is not 
satisfied.
    The DBNPS will provide two active methods of BPC. The first does 
not meet the single-failure criterion. The second does not meet 
regulatory requirements for analyses applicable to an acceptable system 
and is susceptible to some of the same failures that cause failure of 
the first. Further, the second has a small likelihood of failing to 
function when first initiated because core bypass flow is necessary for 
a short time to prevent conditions where boron precipitation may occur. 
However, DBNPS has predicted via a conservative assessment that the 
total BPC-related CDF and LERF are about 10 <SUP>-</SUP>\7\/reactor-
year and 10 <SUP>-</SUP>\11\/reactor-year, respectively. The DBNPS has 
further described in-depth, proceduralized actions that will be applied 
to restore an active BPC method should it fail to initiate when called 
upon. These actions, in combination with the predicted failure rate 
without the actions, establish that a satisfactory defense-in-depth is 
provided such that long term cooling performance of the ECCS will 
continue to be met. Therefore, the requested exemption meets the 
special circumstances requirement of 10 CFR 50.12(a)(2)(ii) with 
respect to the single failure criterion requirements.
    With respect to the decay heat generation rate specified in 
appendix K, section I.A.4, the underlying purpose of the heat 
generation rate is to provide an appropriate value for the ECCS 
evaluation model. The DBNPS assumed 1.0 times the American Nuclear 
Society standard infinite operation decay heat generation rate for 
calculation of the DHR cooler aspects of the backup method whereas 
appendix K specifies 1.2. The staff considers the use of 1.0 to be more 
realistic and suitable for probabilistic risk calculations. Therefore, 
the requested exemption meets the special circumstances requirement of 
10 CFR 50.12(a)(2)(ii) with respect to the decay heat generation rate 
in that use of the 1.2 value is not necessary to achieve the underlying 
purpose of the rule.

[[Page 31021]]

VI

    For the foregoing reasons, the NRC staff has concluded that an 
exemption is acceptable to the requirements of appendix K, section 
I.D.1, 10 CFR 50.46(b)(5), and 10 CFR 50.46(a)(1)(ii) with respect to 
the DBNPS active methods for BPC. The NRC staff has determined that 
there are special circumstances present, as specified in 10 CFR 
50.12.(a)(2)(ii), in that application of the specific regulations is 
not necessary in order to achieve the underlying purpose of these 
regulations, which is to assure long term cooling performance of the 
ECCS in the event of the most damaging single failure of ECCS 
equipment. In addition, the staff has determined that an exemption to 
appendix K, section I.A.4 is acceptable with respect to the decay heat 
generation rate. Special circumstances exist in that use of the 1.2 
value specified in appendix K, section I.A.4, is not necessary in order 
to achieve the underlying purpose of the rule.
    Accordingly, the Commission has determined that, pursuant to 10 CFR 
50.12(a), the requested exemption is authorized by law, will not 
endanger life or property or the common defense and security, and is 
otherwise in the public interest. Therefore, the Commission hereby 
grants the requested exemption. This exemption is effective upon 
issuance.

    Dated at Rockville, Maryland, this 5th day of May 2000.

    For the Nuclear Regulatory Commission.

Suzanne C. Black,
Acting Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 00-12129 Filed 5-12-00; 8:45 am]
BILLING CODE 7590-01-P