[Federal Register Volume 63, Number 193 (Tuesday, October 6, 1998)]
[Notices]
[Pages 53730-53736]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-26745]
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NUCLEAR REGULATORY COMMISSION
[Docket Nos. 50-321 and 50-366]
Southern Nuclear Operating Co. Inc., et al.; Notice of
Consideration of Issuance of Amendments to Facility Operating Licenses,
Proposed No Significant Hazards Consideration Determination, and
Oportunity for a Hearing
The U.S. Nuclear Regulatory Commission (the Commission) is
considering issuance of amendments to Facility Operating License Nos.
DPR-57 and NFP-5 issued to Southern Nuclear Operating Company, Inc., et
al. (the licensee) for operation of the Edwin I. Hatch Nuclear Plant,
Units 1 and 2, located in Appling County, Georgia.
The proposed amendments would revise the Technical Specifications
to accommodate an increase in maximum licensed thermal power level from
2558 megawatts thermal (MWt) to 2736 MWt.
The licensee submitted the proposed changes by letter dated August
8, 1997. In processing this request, the staff recognized on September
29, 1998, it inadvertently failed to publish a notice of proposed
issuance of the amendments in the Federal Register. In the August 8,
1997, original application, the licensee requested that the proposed
amendments be issued prior to startup from the fall 1998 refueling
outage on Unit 2. Startup from the refueling outage is presently
scheduled for October 18, 1998.
Upon being informed by the staff that a notice of proposed issuance
of amendments inadvertently was not published, the licensee requested,
by letter dated September 30, 1998, that the proposed amendments be
processed on a exigent basis.
The need for exigency is based on the fact that the licensee would
be required to postpone changes to procedures, instrumentation, and
setpoints on Unit 2 until after startup and power ascension of the
plant if the amendments were not issued prior to restart. The licensee
would then be required to implement these changes while online which
would increase the possibility of a plant scram and introduce a
potential for unnecessary transients on the plant.
[[Page 53731]]
The licensee has evaluated the impact of the schedule change and
the online implementation of the extended power uprate (EPU) and
determined that receiving the amendments prior to startup will result
in a net increase in plant safety and reliability. Reliability benefits
include a reduced potential for an inadvertent reactor scram while
adjusting instrumentation online and human performance issues
associated with training and procedures. Implementation of the EPU
requires adjustment of the direct scram from the turbine stop valve and
the turbine control valve fast closure and the main steamline high flow
isolation setpoints. These adjustments place the plant in a
configuration that results in generation of a half scram signal and an
increased potential for an unnecessary full scram of the plant.
Implementation of the EPU also requires adjustment of the average power
range monitor (APRM) setpoints, including the APRM simulated thermal
power scram.
In addition, the licensee has identified approximately 20
instrumentation and controls and 30 operations procedures that would
require revisions prior to and after the issuance of the uprate
amendments if they are not issued prior to Unit 2 startup. This may
result in human factor concerns associated with procedure revisions and
operator training.
Therefore exigency is appropriate in order to allow implementation
of these amendments and will result in a net benefit in plant safety
and reliability.
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.
Pursuant to 10 CFR 50.91(a)(6) for amendments to be granted under
exigent circumstances, the NRC staff must determine that the amendment
request involves 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:
I. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated
based upon the following discussion:
A. Evaluation of the Probability of Previously Evaluated Accidents
The proposed extended power uprate imposes only minor increases
in plant operating conditions. No changes to rated core flow, rated
reactor pressure, or turbine throttle pressure are required. The
higher power level will result in moderate flow increases in systems
associated with the turbine cycle (e.g., condensate, feedwater, and
main steam). The small increase in operating temperatures for BOP
[balance of plant] support systems has no significant effect on LOCA
[loss-of-coolant accident] or other accident probabilities. The
extended power uprate evaluations confirm the higher power level has
no significant effect on flow induced erosion/corrosion. The
limiting feedwater and main steam piping flow increases were
evaluated and shown to be approximately proportional to the power
increase. The affected systems are currently monitored by the Plant
Hatch erosion/corrosion program. Continued system monitoring
provides a high level of confidence in the integrity of potentially
susceptible high energy piping systems.
When required, the occurrence frequency of accident precursors
and transients is addressed by applying the guidance of NRC-reviewed
setpoint methodology to ensure acceptable trip avoidance is provided
during operational transients subsequent to implementation of
extended power uprate. The setpoint evaluation confirmed Plant Hatch
extended power uprate does not increase the number of challenges to
the protective instrumentation.
Plant systems, components, and structures were verified as
capable of performing their intended functions under increased power
conditions with a few minor exceptions.
That is, some components will be modified prior to
implementation of the extended power uprate program to accommodate
the revised operating conditions * * *. The Plant Hatch extended
power uprate does not significantly affect the reliability of plant
equipment. In cases where plant availability could be impacted by
BOP equipment performance, modifications and administrative controls
will be implemented to adequately compensate. No new components or
system interactions that could lead to an increase in accident
probability are created due to operation at 2763 MWt [megawatts
thermal].
The probability of design basis accidents (DBAS) occurring is
not affected by the increased power level, since the applicable
criteria established for plant equipment (e.g., ANSI Standard B3 1.1
and ASME [American Society of Mechanical Engineers] Code) will still
be followed when the plant is operated at the new power level. The
extended power uprate analysis basis assures the limits prescribed
by the Code of Federal Regulations (CFR) (e.g., LOCA PCT [peak clad
temperature], SLMPCR, 10 CFR 20) will be maintained by meeting the
appropriate regulatory criteria. Similarly, factors of safety
specified by application of the CFR design rules were demonstrated
to be maintained, as have other margin-assuring acceptance criteria
used to judge the acceptability of the plant. Established reactor
scram setpoints are such that there should be no increase in scram
frequency due to the increased power level. No new challenges to
safety-related equipment will result. Therefore, the proposed
Operating License and Technical Specifications changes do not
involve a significant increase in the probability of an accident
previously evaluated.
B. Evaluation of the Consequences of Previously Evaluated Accidents
ECCS-LOCA Analysis
The Plant Hatch emergency core cooling system loss-of-coolant
accident (ECCS-LOCA) performance analysis was performed for extended
power uprate using methodology approved by the NRC for analysis
required by 10 CFR 50.46. This revised analysis utilizes the same
methodology (SAFER/GESTR) as the existing ECCS-LOCA analysis. ECCS
requirements assumed for extended power uprate are very similar to
the existing 1986 analysis. In accordance with regulatory guidance,
the Plant Hatch ECCS-LOCA analysis was performed at 102% of the new
RTP of 2763 MWt, or 2818 MWt. The licensing peak clad temperature
remains well below the 10 CFR 50.46 required limit of 2200 deg.F.
Therefore, the analysis demonstrates Plant Hatch will continue to
comply with 10 CFR 50.46 and 10 CFR 50, Appendix K at extended power
uprate conditions. Thus, the consequences of accidents are not
significantly increased at the higher power level.
Abnormal Operating Transient Analysis
An evaluation of the Plant Hatch Unit I and Unit 2 Final Safety
Analysis Reports (FSARs) and reload transients was performed for
extended power uprate to demonstrate the proposed maximum power
level will have no adverse effect on plant safety. The evaluation
was performed for a power level of 2763 MWt, with the exception of
certain event evaluations that were performed at 102% of 2763 MWt.
The transient analysis performed to demonstrate the acceptability of
Plant Hatch extended power uprate employed the same NRC-approved
methods used today.
The limiting transient events at extended power uprate
conditions, including events that establish the core thermal
operating limits and events that bound other transient protection
criteria, were evaluated. The limiting transients were benchmarked
against the existing RTP [rated thermal power] level by performance
of the event analysis at both the proposed power level and the
current RTP level. In addition, an expanded group of transient
events was evaluated to confirm these events remained less limiting
than the most limiting transients. The transient events included in
the expanded group were chosen based upon events demonstrated to be
sensitive to initial power level. This evaluation confirmed the
existing set of limiting transient events remains valid for the
Plant Hatch extended power uprate. The evaluation was performed for
a
[[Page 53732]]
representative core and demonstrates the overall capability to meet
all transient safety criteria. Cycle-specific analyses will continue
to be performed for each fuel reload to demonstrate compliance with
the applicable transient criteria and establish cycle-specific
operating limits.
The results of the limiting transients evaluation demonstrate
extended power uprate can be accomplished without a significant
increase in the consequences of the transients evaluated. The fuel
thermal-mechanical limits at extended power uprate conditions are
within the specific design criteria for the GE fuels currently
loaded in the Plant Hatch cores. Also, the power-dependent and flow-
dependent minimum critical power ratio (MCPR) and maximum average
planar linear heat generation rate (MAPLHGR) limits utilized at
Plant Hatch since the mid-1980s require only minor changes. The peak
reactor pressure vessel (RPV) bottom head pressure remains within
the ASME Code requirement for RPV overpressure protection. The
effects of plant transients were evaluated by assessing disturbances
caused by a malfunction or single failure of equipment, or operator
error, consistent with the FSARs [Final Safety Analysis Reports].
Limiting transient events tend to be slightly more severe
([approximately equal to] 1%) when initiated from the new power
level, assuming a 1.12 safety limit (SLMCPR) which was determined
using the latest NRC-approved methods. However, for the most
limiting transient, an evaluation of a representative core showed
little or no change is required to the operating limit MCPR (OLMCPR)
at extended power uprate and the integrity of SLMCPR is maintained.
The margin of safety established by the SLMCPR is not affected and
the event consequences are not significantly affected by the
proposed extended power uprate to 2763 MWt. Cycle-specific analyses
will continue to be performed for each fuel reload to demonstrate
compliance with the applicable transient criteria and establish
cycle-specific operating limits.
The transient analysis results demonstrate the Plant Hatch core
thermal power output can be safely increased to 2763 MWt without
significantly affecting the consequences of previously evaluated
postulated transient events. The results of the extended power
uprate transient evaluation are summarized as follows:
1. Events Resulting in Nuclear System Pressure Increase
a. Main Generator Load Rejection with No Steam Bypass. At
extended power uprate conditions, the fuel transient thermal and
mechanical overpower results remain below the NRC-accepted design
criteria.
b. Main Turbine Trip with No Steam Bypass. At extended power
uprate conditions, the fuel transient thermal and mechanical
overpower results remain below the NRC-accepted design criteria.
c. Main Steam Isolation Valve (MSIV) Closure. At extended power
uprate conditions, this event (with a scram initiated by the valve
closure) remains nonlimiting with respect to fuel thermal limits.
d. Pressure Regulator Failure--Closed and Slow Closure of a
Single TCV [temperature control valve]. These transients remain
nonlimiting as compared with other more severe pressurization
events.
2. Event Resulting in a Reactor Vessel Water Temperature Decrease
a. Loss of Feedwater Heating. The consequences of this event at
the extended power uprate conditions remain nonlimiting with regard
to the cycle OLMCPR. The results at low core flow conditions are
actually slightly higher than for the high core flow condition
because of increased inlet coolant subcooling into the reactor core.
The calculated thermal and mechanical overpower limits at extended
power uprate conditions for this event also meet fuel design
criteria.
b. Inadvertent High Pressure Coolant Injection (HPCI) Actuation.
For the limiting condition analyzed, both the high water level
setpoint and the high RPV steam dome pressure scram setpoints are
not reached. Based upon the peak average fuel surface heat flux
results, the HPCI actuation event will be bounded by the limiting
pressurization event with respect to delta critical power ratio
([delta]CPR) considerations. In addition, the fuel transient thermal
and mechanical overpower limits remain within the allowable NRC-
accepted design values.
c. Shutdown Cooling Residual Heat Removal (RHR) Malfunction.
This event is not affected by extended power uprate.
3. Event Resulting in a Positive Reactivity Insertion
Rod Withdrawal Error (RWE)
The current rod block monitor (RBM) system with power-dependent
setpoints was analyzed for the RWE event at extended power uprate
conditions using a statistical approach consistent with NRC approved
methods. The analysis concluded the transient is slightly more
severe with a greater [delta]CPR from the initial most limiting CPR.
However, the fuel and mechanical overpower limits remain within the
NRC accepted design criteria.
4. Event Resulting in a Reactor Vessel Coolant Inventory Decrease
a. Pressure Regulator Failure to Full Open. The results of this
transient for extended power uprate remain nonlimiting as compared
with other more severe pressurization events.
b. Loss of Feedwater Flow. This transient event does not pose
any direct threat to the fuel in terms of a power increase from the
initial conditions. Water level declines rapidly and a low water
level causes a reactor scram. Actuation of HPCI and reactor core
isolation cooling (RCIC) terminate the event. However, the loss of
feedwater flow event is included in the extended power uprate
evaluation to assure sufficient water makeup capability is available
to keep the core well covered when all normal feedwater is lost. A
plant-specific analysis performed in support of the extended power
uprate program shows a large amount of water remains above the top
of the active fuel. This sequence of events does not require any new
operator actions or shorter operator response times. Therefore,
operator actions for the event do not significantly change for
extended power uprate.
c. Inadvertent Opening of a Safety/Relief Valve (S/RV), Loss of
Auxiliary Power, and Loss of One DC System. These events remain less
severe at extended power uprate conditions.
5. Event Resulting in Core Coolant Flow Decrease
a. Recirculation Pump Seizure. The recirculation pump seizure
transient evaluation includes the assumption the pump motor shaft of
one recirculation pump stops instantaneously. As a result, core flow
decreases rapidly. The heat flux decline lags core power and flow,
and could result in a degradation of core heat transfer. At extended
power uprate conditions, the consequences of the pump seizure event
remain nonlimiting. Note the Unit 2 FSAR classifies this event as an
accident due to the low probability of occurrence.
b. RPT and Recirculation Flow Control Failure Decreasing Flow.
These transients remain nonlimiting at extended power uprate
conditions.
6. Event Resulting in Core Coolant Flow Increase
Recirculation Flow Controller Failure Increasing Flow
The results of this transient for extended power uprate remain
nonlimiting as compared with other more severe pressurization
events.
7. Event Resulting in Core Coolant Temperature Increase
Failure of RHR Shutdown Cooling
This event is not significantly affected by the increase in
licensed thermal power.
8. Event Resulting in Excess of Coolant Inventory
Feedwater Controller Failure--Maximum Demand
The CPR calculated for this event at extended power uprate
conditions is slightly higher than the corresponding value for the
current rated power. However, the trend for the feedwater controller
failure--maximum demand event is consistent--with the analysis for
the current rated power level. The fuel thermal margin results are
within the acceptable limits for the fuel types analyzed.
DBA Challenges to Containment
The primary containment's response to the limiting DBA was
evaluated at 2763 MWt, plus a 2% adder. The effect of extended power
uprate on the short-term containment response (peak values), as well
as the long-term containment response for containment pressure and
temperature confirms the suitability of the plant for operation at
the new power level. Factors of safety provided in the ASME Code are
maintained, and the safety margin is not altered by uprating power
to 2763 MWt.
Short-term containment response analyses were performed for the
limiting DBA LOCA, a double-ended guillotine break of a
recirculation suction line, to demonstrate operation at a bounding
reactor power will
[[Page 53733]]
not result in exceeding the containment design limits. This limiting
DBA LOCA event results in the highest short-term containment
pressures and dynamic loads. The analysis determined, at the
proposed reactor power level, the maximum drywell pressure values
increase only [approximately equal to] 1 psi and remain well bounded
by the containment design pressure. Extended power uprate has no
adverse effect on the containment structural design pressure.
Because increasing RTP increases residual heat, the containment
long-term response will have slightly higher temperatures. Long-term
suppression chamber temperatures remain within the design
temperature of the structure; thus, ASME Code factors of safety are
maintained and the safety margin is not affected. An analysis
confirmed ECCS pump net positive suction head (NPSH) is not
adversely affected with this temperature response, and the long-term
response does not adversely affect the containment structure or the
environmental qualification (EQ) of equipment located in the drywell
and torus. The drywell long-term temperature response is not
adversely affected for the higher reactor power; thus, the
containment long-term response for extended power uprate is
acceptable.
The impact of a reactor power increase on containment dynamic
loads was evaluated and found to have no adverse effect for
conditions that bound the proposed power level. Thus, containment
dynamic loads are acceptable for operation at 2763 MWt.
The Plant Hatch extended power uprate evaluation of the primary
containment response to DBAs confirmed the proposed power level does
not result in a significant increase in the consequences of a
postulated accident for a reactor power level [approximately equal
to] 2% greater than the proposed increase to 2763 MWt.
Radiological Consequences of DBAs
For Plant Hatch extended power uprate, the radiological
consequences of the limiting DBAs were reevaluated. The evaluations
included the effect of the proposed power level on the radiological
consequences of accidents presented in the FSARs. Reference 3
provides information on a revised radiological dose analysis for the
DBA LOCA and shows doses remain within 10 CFR 100 limits at the new
power level.
This DBA LOCA radiological evaluation was performed using input
and evaluation techniques consistent with current regulatory
guidance and appropriate plant design basis. The inputs and analysis
methods are different from those utilized in the current licensing
basis evaluation presented in the FSARs and the Atomic Energy
Commission safety evaluation report supporting the initial plant
licensing. However, the input used in the extended power uprate
radiological evaluation provides a conservative assessment of the
potential radiological consequences. The conclusions of these
evaluations are consistent with the original licensing basis
evaluations. The radiological consequences of the limiting DBA
remain within 10 CFR 100 guidelines for the proposed RTP level. For
the purpose of analysis, the new RTP level was increased by an
additional 2% in accordance with regulatory guidance.
To demonstrate the change in consequences, the evaluation of
radiological consequences using the different analysis inputs and
methods was performed for the existing licensed RTP level and the
proposed RTP level.
The impact of the proposed licensed power level on the fuel
handling accident, control rod drop accident, and main steam line
break outside primary containment was evaluated. The radiological
consequences remain well below regulatory limits.
The evaluation of DBA radiological consequences confirmed
extended power uprate does not result in a significant increase in
consequences at a power level of 2763 MWt. The results remain below
10 CFR 100 guideline values. Therefore, the postulated radiological
consequences do not represent a significant change in accident
consequences and are clearly within the regulatory guidelines for
the proposed power level increase.
Other Evaluations
1. Performance Improvements
The extended power uprate safety analysis was performed taking
into account the implementation of the following previously approved
special operational features.
a. Single-Loop Operation (SLO). The safety analysis for extended
power conditions shows the single-loop operating mode remains valid.
The current trip setpoints determined for two-loop operation (TLO)
were confirmed to be acceptable for SLO, with a correction applied
to account for the actual effective drive flow applied when
operating with a single loop. The SLO settings were conservatively
established to be consistent with the TLO settings, while ensuring
the appropriate corrections are applied to the MAPLHGR and the OLCPR
to account for SLO.
b. Maximum Extended Load Line Limit (MELLL). The safety analysis
for new power conditions shows the operating domain as analyzed is
valid for extended power uprate conditions, even with operation
permitted on a slightly higher absolute rod line.
c. Increased Core Flow (ICF). The safety analysis for extended
power uprate shows that operation at ICF conditions remains
acceptable.
d. Final Feedwater Temperature Reduction (FFWTR). The safety
analysis for extended power uprate shows operation at FFWTR
conditions remains acceptable.
e. Average Power Range Monitor/Rod Block Monitor Technical
Specification (ARTS) Improvements. The safety analysis for extended
power uprate conditions shows the ARTS improvements remain valid for
the extended power uprate conditions.
2. Effect of Extended Power Uprate on Support Systems
An evaluation was performed to address the effect of the
extended power uprate on accident mitigation features, structures,
systems, and components within the BOP. The evaluation results are
as follows:
a. Auxiliary systems, such as building heating, ventilation, and
air-conditioning (HVAC) systems, reactor building closed cooling
water, plant service water, spent fuel pool cooling; process
auxiliaries, such as instrument air and makeup water; and the post-
accident sampling system were confirmed to operate acceptably under
normal and accident conditions at the proposed power level.
b. Secondary containment and standby gas treatment system were
confirmed to be adequate relative to containing, processing, and
controlling the release of normal and post-accident levels of
radioactivity.
c. Instrumentation was reviewed and confirmed capable of
performing control and monitoring functions at the proposed power
level. As required, analyses were performed to determine the need
for setpoint changes for various functions (e.g., APRM simulated
thermal power scram setpoints). In general, setpoints are to be
changed only to maintain adequate difference between plant operating
parameters and trip setpoints, while ensuring safety performance is
demonstrated. The revised setpoints were established using NRC-
reviewed methodology as guidance.
d. Electric power systems, including the main generator and
switchgear components, were verified as being capable of providing
the required electrical load as a result of the increased power
level. An evaluation of the auxiliary power system confirmed the
system has sufficient capacity to support all required loads for
safe shutdown, maintain a safe shutdown condition, and operate the
required engineered safeguards equipment following postulated
accidents. No safety-related electrical loads were affected which
would impact the emergency diesel generators.
e. Piping systems were evaluated for the effect of operation at
higher power levels, including transient loading. The evaluation
confirmed piping and supports are adequate to accommodate the
increased loading resulting from operation at higher power
conditions.
f. The effect of the higher power conditions on a high energy
line break (HELB) was evaluated. The evaluation confirmed
structures, systems, and components important to safety are capable
of accommodating the effects of jet impingement, blowdown forces,
and the environmental effects resulting from HELB events.
g. Control room habitability was evaluated. Post-accident
control room and Technical Support Center doses at 2763 MWt were
confirmed to be within the guidelines of General Design Criterion 19
of 10 CFR 50, Appendix A. (See Ref. 3.)
h. The EQ of equipment important to safety was evaluated for the
effect of normal and accident operating conditions at the proposed
power level. The equipment remains qualified for the new conditions.
The preventive maintenance program will continue to provide
equipment maintenance or replacement to ensure equipment EQ at
extended power uprate conditions.
3. Effect on Special Events
The consequences of special events (i.e., anticipated transient
without scram (ATWS); 10 CFR 50, Appendix R; and station blackout)
remain within NRC-accepted
[[Page 53734]]
criteria at 2763 MWt. Vessel overpressure protection was analyzed
assuming a closure of the MSIVs with a neutron flux scram, Although
the peak reactor vessel bottom head pressure increases slightly at
extended power uprate conditions, it is well within the ASME Code
overpressure limit of 1375 psig. The standby liquid control (SLC)
system capability analysis illustrates the plant can still achieve
cold shutdown without dependence upon the control rods. Core
thermal-hydraulic stability was evaluated. The new power level and
modified power-to-flow map will not affect the ability to detect and
suppress limit-cycle oscillations. Extended power uprate also does
not adversely affect other special events, because the available
equipment is not changed and the input assumptions for the
evaluations are not significantly changed. Concurrent malfunctions
assumed to occur during accidents were accounted for in the safety
analyses for the proposed power level increase. The consequences of
these equipment malfunctions do not change with the implementation
of the extended power uprate program.
Conclusion
The evaluation of ECCS performance demonstrated the criteria of
10 CFR 50.46 are satisfied, thus, the margin of safety established
by the criteria is maintained. The analysis demonstrated the ECCS
will function with the most limiting single failure to mitigate the
consequences of the accident and maintain fuel integrity. Challenges
to the containment were evaluated and the integrity of the fission
product barrier was confirmed. The radiological consequences of DBAs
were evaluated and it was found the effect of the proposed extended
power uprate on postulated radiological consequences does not result
in a significant increase in accident consequences. The evaluations
provide conservative results for the proposed power level of 2763
MWt and demonstrate the proposed extended power uprate does not
result in a significant increase in accident consequences.
The abnormal transients were analyzed under extended power
uprate conditions, and the analysis confirms the power increase to
2763 MWt has only a minor effect upon MCPR and the SLMCPR results.
Thus, the margin of safety as assured by the SLMCPR is maintained.
The effect of extended power uprate on the consequences of abnormal
transients that result from potential component malfunctions is
acceptable; thus, operation at the new power level does not result
in a significant increase in transient event consequences.
The spectrum of analyzed postulated accidents and transients was
investigated and determined to meet current regulatory criteria. In
the area of core design, the fuel operating limits will still be met
at the requested power level, and fuel reload analyses will show
plant transients meet NRC-accepted criteria. The evaluation of
accident consequences was performed consistent with the proposed
changes to the plant Technical Specifications. Therefore, the
proposed Operating License and Technical Specifications changes will
not cause a significant increase in the consequences of an accident
previously evaluated for Plant Hatch Unit 1 and Unit 2.
II. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously evaluated
based upon the following discussion:
The BWR [boiling water reactor] configuration, operation, and
event response is unchanged by the higher power level. Analyses of
transient events confirm the same transients remain limiting and no
transient events will result in a new sequence of events that could
lead to a new accident scenario. The extended power uprate analyses
confirm the accident progression is basically unchanged.
An increase in power level does not create a new fission product
release path, or result in a new fission product barrier failure
mode. The same fission product barriers, such as the fuel cladding,
the reactor coolant pressure boundary (RCPB), and the reactor
containment, remain in place. Fuel rod cladding integrity is ensured
by operating within thermal, mechanical, and exposure design limits,
and is demonstrated by the extended power uprate transient and
accident analyses. Similarly, analysis of the RCPB and primary
containment demonstrates the increased power level has no adverse
effect upon these fission product barriers. The proposed Technical
Specifications changes in support of extended power uprate
implementation are consistent with the analyses, and assure
transient and accident mitigation capability in compliance with
regulatory requirements.
The effect of Plant Hatch extended power uprate on plant
equipment was evaluated. No new operating mode, safety-related
equipment lineup, accident scenario, or equipment failure mode
resulting from the increased power was identified. The full spectrum
of accident considerations defined in the FSARs was evaluated, and
no new or different kind of accident resulting from the extended
power uprate was identified. Extended power uprate analyses were
performed using developed technology which was applied assuming the
capability of existing plant equipment in accordance with existing
regulatory criteria, including accepted codes, standards, and
methods. GE has analyzed BWRs, with higher power densities and no
new power-dependent accidents were identified. In addition, this
uprate does not create any new sequence of events or failure modes
that lead to a new type of accident.
All necessary actions will be taken prior to implementation of
this program to ensure safety-related structures, systems, and
components remain within their design allowable values and also
ensure they can perform their intended functions under higher power
conditions. The extended power uprate does not increase or create
any new challenges to safety-related equipment or other equipment
whose failure could cause a different kind of accident from that
previously evaluated.
III. The proposed changes do not involve a significant reduction
in a margin of safety based upon the following discussion:
The transient and accident analyses, as well as a majority of
the plant-specific evaluations, to support the extended power uprate
were performed at 2763 MWt and increased by an additional 2% in
accordance with regulatory guidance, when applicable, for the
evaluation of accidents and transients. The analyses demonstrate
sufficient margins of safety exist. The evaluation of transient
events and instrument setpoints demonstrate sufficient margin when
compared to criteria establishing margins of safety for the proposed
increase in power level.
The Plant Hatch extended power uprate analysis basis assures the
power-dependent safety margin criteria prescribed by the CFR will be
maintained by meeting the appropriate regulatory criteria.
Similarly, factors of safety specified by application of the ASME
Code design rules are maintained, as are other margin-assuring
acceptance criteria used to judge the acceptability of the plant.
A. Fuel Thermal Limits
No change in the basic fuel design is required to achieve the
extended uprate power level or to maintain the margins as discussed
above. No increase in the allowable peak rod power is requested. The
abnormal transients were evaluated at the higher power level for a
representative core configuration. The analysis confirms the
extended power uprate has no significant effect upon the OLMCPR or
the SLMCPR. The fuel operating limits, such as MAPLHGR and the
OLMCPR, will still be met at the new power level. The analyses
confirm the acceptability of these operating limits for extended
power uprate without an adverse effect upon margins to safety. Cycle
specific analyses for each fuel reload will continue to be performed
to demonstrate compliance with the applicable transient criteria and
establish cycle-specific operating limits.
B. DBA Challenges to Fuel
Evaluation of the ECCS performance demonstrates the criteria of
10 CFR 50.46 are satisfied; thus, the margin of safety established
by the criteria is maintained. This evaluation was performed at 2763
MWt, and increased by an additional 2% in accordance with regulatory
guidance. The analysis demonstrates Plant Hatch will continue to
comply with the guidance of 10 CFR 50.46 and the margin of safety
established by the regulation will be maintained following the
increase in power level.
C. DBA Challenges to Containment
The primary containment response to the limiting DBA was
evaluated for extended power uprate. The effect of the increased
power on the short-term containment response (peak values), as well
as the long-term containment response, for containment pressure and
temperature confirms the suitability of the plant for operation at
the proposed power level of 2763 MWt. Factors of safety provided in
the ASME Code are maintained and safety margin is not affected.
Short-term containment response analyses were performed for the
limiting DBA LOCA, consisting of a double-ended guillotine break of
a recirculation suction line, to demonstrate operation at the new
reactor power will not result in exceeding containment design
limits. The analyses determined the
[[Page 53735]]
maximum drywell pressure increases only slightly and is bounded by
the containment design pressure. Extended power uprate has no
adverse effect on containment structural design pressure.
Long-term suppression chamber temperatures remain within the
design temperature of the structure; thus, factors of safety
provided in the ASME Code are maintained and the safety margin is
not affected. Analyses confirm ECCS pump NPSH is not adversely
affected with this temperature response, and the long-term response
does not adversely affect the containment structure or the EQ of
equipment located in the drywell and torus.
The impact of a reactor power increase on containment dynamic
loads was evaluated and found to have no adverse effect for
conditions that bound the proposed increase in power level. Thus,
containment dynamic loads are acceptable for extended power uprate.
The Plant Hatch extended power uprate evaluation of the primary
containment response to the DBA confirms the increased power level
does not result in the reduction in a margin of safety.
D. DBA Radiological Consequences
The FSARs provide the radiological consequences for each DBA.
The magnitude of the potential consequences is dependent upon the
quantity of fission products released to the environment, the
atmospheric dispersion factors, and the dose exposure pathways. For
the case of extended power uprate, the atmospheric dispersion
factors and the dose exposure pathways do not change. Therefore, the
only factor that will influence the magnitude of the consequences is
the quantity of activity released to the environment. This quantity
is a product of the activity released from the core and the
transport mechanisms between the core and the effluent release
point.
The radiological consequences of DBAs were evaluated and it was
found there is not a significant increase in consequences. The
results remain below 10 CFR 100 guideline values. Therefore, the
postulated radiological consequences are clearly within the
regulatory guidelines, and all radiological safety margins are
maintained for the proposed power level of 2763 MWt.
E. Transient Evaluations
The effect of plant transients was evaluated by assessing a
number of disturbances of process variables, and malfunctions or
failures of equipment consistent with the FSARS. The transient
events tend to be slightly more severe ([approximately equal to] 1%)
when initiated from the new power level, assuming a 1.12 SLMCPR,
which was determined using the latest GE methods approved by the
NRC. However, for the most limiting transient, an evaluation of a
representative core shows no significant change to the OLMCPR is
required for the new power level and the integrity of the SLMCPR is
maintained.
Cycle-specific analyses for each fuel reload will continue to be
performed to demonstrate compliance with the applicable transient
criteria and establish cycle-specific operating limits.
The fuel thermal-mechanical limits at extended power uprate
conditions are within the specific design criteria for the GE fuels
currently loaded in the Plant Hatch cores. Also, the power-dependent
and flow-dependent MCPR and MAPLHGR methods remain applicable. The
peak RPV bottom head pressure remains within the ASME Code
requirement for RPV overpressure protection.
The margin of safety established by the SLMCPR is not affected
by the proposed power level increase to 2763 MWt.
F. Special Events
The event acceptance limits for special events remain unchanged
for extended power uprate. For example, the peak RPV bottom head
pressure remains below the 1375 psig ASME Code requirement for RPV
overpressure protection. Acceptance limits for ATWS, Appendix R, and
station blackout also remain unchanged.
G. Technical Specifications Changes
The Technical Specifications ensure the plant and system
performance parameters are maintained at the values assumed in the
safety analysis. The Technical Specifications (setpoints, trip
settings, etc.) are selected such that adequate margin exists. For
instruments that initiate protective functions (e.g., reactor
protection system, ECCS, and containment isolation), proper account
is taken of inaccuracies introduced by instrument drift, instrument
accuracy, and calibration accuracy. The Technical Specifications
address equipment availability and limit equipment out-of-service to
assure the plant will have at least the complement of equipment
available to deal with plant transients as that assumed in the
safety analysis. The evaluations and analyses performed to
demonstrate the acceptability of extended power uprate were
performed using input consistent with the proposed changes to the
plant Technical Specifications.
The events (i.e., transients and accidents) that form the
Technical Specifications Bases were evaluated for extended power
uprate conditions using input and initial conditions consistent with
the proposed Technical Specifications changes. Although some changes
to the Technical Specifications are required, no NRC acceptance
limit is exceeded. Therefore, the margins of safety assured by
safety limits and other Technical Specifications limits are
maintained. The proposed changes to the Bases are consistent with
the evaluations demonstrating acceptability of the new licensed
power level of 2763 MWt.
Conclusion
The spectrum of postulated accidents and transients was
investigated and was determined to meet the current regulatory
criteria for Plant Hatch at extended power uprate conditions. In the
area of core design, fuel operating limits will still be met at the
new power level, and fuel reload analyses will show plant transients
meet the NRC-accepted criteria as specified in the plant Technical
Specifications. Challenges to fuel and ECCS performance were
evaluated and shown to meet the criteria of 10 CFR 50.46 and 10 CFR
50, Appendix K. Challenges to the containment were evaluated and the
integrity of the fission product barrier was confirmed. Radiological
release events were evaluated and shown to meet the guidelines of 10
CFR 100. The proposed Operating License and Technical Specifications
changes are consistent with the Plant Hatch extended power uprate
evaluations. The evaluations demonstrate compliance with the margin-
assuring acceptance criteria contained in applicable codes and
regulations. Therefore, the proposed Operating License and Technical
Specifications changes do not involve a significant reduction in the
margin of safety.
References
1. NRC letter from D. M. Crutchfield to G. L. Sozzi (GE),
``Staff Position Concerning GE BWR Extended Power Uprate Program,''
TAC No. M91680, dated February 8, 1996.
2. NRC letter from K. N. Jabbour to J. T. Beckham, Jr.,
``Issuance of Amendments--Edwin I. Hatch Nuclear Plant Units I and
2,'' (TAC Nos. M91077 and M91078), dated August 31, 1995.
3. SNC letter BL-5356 from H. L. Sumner, Jr., to the NRC,
``Revised Post-LOCA Doses,'' dated April 17, 1997.
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
amendment request involves no significant hazards consideration.
The Commission is seeking public comments on this proposed
determination. Any comments received within 14 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 14-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 14-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. 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 Administrative Services, Office of Administration,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and
should cite the
[[Page 53736]]
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D59, 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 November 5, 1998, 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 Appling County Public Library, 301 City
Hall Drive, Baxley, Georgia. 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 a 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 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 the amendments are issued before the expiration of the 30-day
hearing period, the Commission will make a final determination on the
issue of no significant hazards consideration. If a hearing is
requested, the final determination will serve to decide when the
hearing is held.
If the final determination is that the amendment request involves
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 amendment request involves 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 Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC,
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 amendments dated August 8, 1997, as supplemented by
letters dated March 9, May 6, July 6, July 31, September 4, September
11, and September 30, 1998, and also advanced information related to
the application dated April 17, 1998, which are 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 Appling County Public Library, 301 City
Hall Drive, Baxley, Georgia.
Dated at Rockville, Maryland, this 1st day of October 1998.
For the Nuclear Regulatory Commission.
Herbert N. Berkow,
Director, Project Directorate II-2, Division of Reactor Projects--I/II,
Office of Nuclear Reactor Regulation.
[FR Doc. 98-26745 Filed 10-5-98; 8:45 am]
BILLING CODE 7590-01-P
