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

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