[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