.SOUTHERN CALIFORNIA Peter T. Dietrich

E.DISON Senior Vice President & Chief Nuclear Officer

An EDISON INTERNATIONAL® Company

April 5, 2013

10 CFR 50.90

U. S. Nuclear Regulatory Commission

ATTN: Document Control Desk

Washington, D.C. 20555

Subject: Docket No. 50-361

Amendment Application Number 263

Steam Generator Program

San Onofre Nuclear Generating Station, Unit 2

References: (1) Letter from Mr. Elmo E. Collins (USNRC) to Mr. Peter T. Dietrich

(SCE), dated March 27, 2012, Confirmatory Action Letter 4-12-001,

San Onofre Nuclear Generating Station, Units 2 and 3, Commitments

to Address Steam Generator Tube Degradation (ML12087A323)

(2) Letter from Mr."Peter T. Dietrich (SCE) to Mr. Elmo E. Collins

(USNRC), dated October 3, 2012, Confirmatory Action Letter - Actions

to Address Steam Generator Tube Degradation, San Onofre Nuclear

Generating Station, Unit 2 (ML12285A263)

(3) Letter from Mr. Richard J. St. Onge to NRC, dated March 14, 2013,

Operational Assessment for 100% Power Case Regarding

Confirmatory Action Letter Response, San Onofre Nuclear Generating

Station, Unit 2 (ML1 3074A793)

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Southern California Edison (SCE) hereby submits license

amendment application 263 to operating license NPF-10 for San Onofre Nuclear

Generating Station (SONGS) Unit 2. Amendment Application 263 consists of the

enclosed Proposed Change Number (PCN-602).

The proposed amendment requests that Technical Specification 5.5.2.11.b. 1 be revised

to add a footnote to require that compliance with the steam generator structural integrity

performance criterion (SIPC) be demonstrated up to 70% Rated Thermal Power (2406.6

megawatts thermal) and that Facility Operating License Condition 2.C(1) "Maximum

Power Level" be revised to add a footnote to restrict operation of SONGS Unit 2 to no

more than 70% Rated Thermal Power for the SONGS Unit 2, Cycle 17.

P.O. Box 128

San Clemente, CA 92672

(949) 368-6255 PAX 86255

Fax: (949) 368-6183

Pete.Dietrich@sce.com
Document Control Desk -2- April 5, 2013

The proposed changes are in support of re-start of SONGS Unit 2 following

identification of steam generator tube wear during the SONGS Unit 2, Cycle 17 refueling

and steam generator inspection outage. The proposed changes will maintain

consistency for Cycle 17 between Technical Specification 5.5.2.11 .b and the maximum

power level in License Condition 2.C(1). This change does not relieve SCE of any of its

obligations to demonstrate and maintain the integrity and operability of the SONGS

steam generator tubes for the power level at which the reactor will operate. Separately

from this license amendment request, SCE must also demonstrate that its actions are

consistent with the NRC Confirmatory Action Letter (CAL) (Reference 1) which provides

the necessary actions for re-start.

The Enclosure to this letter provides a technical and regulatory evaluation of the

proposed amendment, including a No Significant Hazards Consideration analysis for the

proposed amendment. SCE has determined that there are no significant hazards

considerations associated with the proposed amendment and that the change is exempt

from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9).

In order to support the planned SONGS Unit 2 return to service, SCE requests approval

of this License Amendment Request (LAR) no later than May 24, 2013. SCE requests

the license amendment be made effective upon NRC issuance.

The LAR applies to SONGS Unit 2 Cycle 17. This will allow SCE to identify long-term

corrective actions, including identification of a long-term power level. Prior to Unit 2

Cycle 18 operation, SCE commits to submit and obtain NRC approval of a License

Amendment Request for long-term operation of Unit 2. In addition, as discussed in the

enclosure, SCE commits for Unit 2 Cycle 17 to revising the time requirement for

providing the Technical Specification 5.7.2.c Special Report to 60 days.

On March 14, 2013, SCE submitted an Operational Assessment (OA) for operation of

SONGS Unit 2 at 100% power (Reference 3). Due to the time needed for the NRC staff

to review the OA for 100% power, it is possible that the staffs review may not be

complete prior to May 24, 2013. As a result, SCE is submitting this license amendment

request to allow for the timely restart of SONGS Unit 2.

There are two regulatory commitments in this application in addition to those previously

made in SCE's response to the CAL (Reference 2).

In accordance with 10 CFR 50.91, SCE is notifying the State of California of this

License Amendment Request by transmitting a copy of this letter and its enclosure to

the designated State Official.

The LAR incorporates by reference a number of documents. The declaration provided

below also applies to the following documents:
Document Control Desk -3- April 5, 2013

* Letter from Mr. Peter T. Dietrich (SCE) to Mr. Elmo E. Collins (USNRC), dated

October 3, 2012, Confirmatory Action Letter - Actions to Address Steam

Generator Tube Degradation, San Onofre Nuclear Generating Station, Unit 2

(ML12285A263) including Enclosure 1 and Enclosure 2 (excluding Attachments

1-5)

" Letter from Mr. Richard J. St. Onge to NRC, dated November 28, 2012,

Confirmatory Action Letter Response - Proprietary Documents, San Onofre

Nuclear Generating Station, Unit 2 (ML12348A287) and Enclosures 5 and 6

* Letter from Mr. Richard J. St. Onge to NRC, dated April 2, 2013, Revised

Response to Request for Additional Information (RAI 11), Revision 1, Regarding

Confirmatory Action Letter Response (TAC No. ME 9727), San Onofre Nuclear

Generating Station, Unit 2

Letter from Mr. Richard J. St. Onge to NRC, dated January 18, 2013, Response

to Request for Additional Information (RAI 12) Regarding Confirmatory Action

Letter Response (TAC No. ME 9727), San Onofre Nuclear Generating Station,

Unit 2 (MLI 3022A408)

Letter from Mr. Richard J. St. Onge to NRC, dated April 2, 2013, Response to

Request for Additional Information (RAI 13), Revision 1, Regarding Confirmatory

Action Letter Response (TAC No. ME 9727), San Onofre Nuclear Generating

Station, Unit 2

Letter from Mr. Richard J. St. Onge to NRC, dated January 29, 2013, Response

to Request for Additional Information (RAl 14) Regarding Confirmatory Action

Letter Response (TAC No. ME 9727), San Onofre Nuclear Generating Station,

Unit 2 (ML13032A009)

Should you have any questions, or require additional information, please contact

Mr. Mark Morgan, Licensing Lead, at (949) 368-6745.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on ii 6• ".13

(Date)"

Sincerely,
Document Control Desk -4- April 5, 2013

Enclosures:

PCN-602 with Attachments

1- List of Regulatory Commitments

2- Existing Facility Operating License Pages, SONGS Unit 2

3- Retyped Facility Operating License Pages SONGS Unit 2

4- Clean Facility Operating License Pages, SONGS Unit 2

cc: A. T. Howell Ill, Regional Administrator, NRC Region IV

R. Hall, NRC Project Manager, SONGS Units 2 and 3

B. Benney, NRC Project Manager, SONGS Units 2 and 3

G. G. Warnick, NRC Senior Resident Inspector, SONGS Units 2 and 3

S. Y. Hsu, California Department of Public Health, Radiologic Health Branch
ENCLOSURE

EVALUATION OF THE PROPOSED CHANGE

PCN-602

Steam Generator Program

1. SUMMARY DESCRIPTION

2. DETAILED DESCRIPTION

3. TECHNICAL EVALUATION

3.1 Background

3.2 Discussion

4. REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria

4.2 No Significant Hazards Consideration

4.3 Conclusions

5. ENVIRONMENTAL CONSIDERATION

6. REFERENCES

ATTACHMENTS:

1. List of Regulatory Commitments

2. Existing Facility Operating License Pages, SONGS Unit 2

3. Retyped Facility Operating License Pages, SONGS Unit 2

4. Clean Facility Operating License Pages, SONGS Unit 2

1 of 14
1.0 SUMMARY DESCRIPTION

Southern California Edison Company (SCE) is proposing a temporary change to the

San Onofre Nuclear Generating Station (SONGS) Unit 2 Facility Operating License.

The proposed changes request Technical Specification 5.5.2.11 .b.1 be revised to

require that compliance with the steam generator structural integrity performance

criterion (SIPC) be demonstrated up to 70% Rated Thermal Power (2406.6 megawatts

thermal) and that Facility Operating License Condition 2.C(1) "Maximum Power Level"

be revised to restrict operation of SONGS Unit 2 to no more than 70% Rated Thermal

Power for the SONGS Unit 2 Cycle 17. This is consistent with the power level in

Commitment 1 in the SONGS Confirmatory Action Letter response (Reference 6.1).

This change does not relieve SCE of any of its obligations to demonstrate and maintain

the integrity and operability of the SONGS steam generator tubes for the power level at

which the reactor will be authorized to operate.

2.0 DETAILED DESCRIPTION

Existing Technical Specification 5.5.2.11.b.1 states that steam generator tubes shall

retain structural integrity over the "full range of normal operating conditions" including

retaining a safety factor of 3.0 against burst under "normal steady state full power

operation" primary-to-secondary pressure differential.

SCE proposes the following footnote be added at the end of Technical Specification

5.5.2.11.b.1:

*For the duration of Unit 2, Cycle 17, the terms "full range of normal operating

conditions" and "normal steady state full power operation" shall be based upon the

steam generators being operated under conditions associated with reactor core

power levels up to 70% of Rated Thermal Power (2406.6 megawatts thermal).

During Cycle 17, Unit 2 will not be operated above this power level.

Existing Facility Operating License Condition 2.C(1) states the following:

(1) Maximum Power Level

Southern California Edison Company (SCE) is authorized to operate the facility

at reactor core power levels not in excess of full power (3438 megawatts

thermal).

2 of 14
SCE proposes the following footnote be added at the end of Facility Operating License

Condition 2.C(1):

* For the duration of Unit 2, Cycle 17, SCE is authorized to operate the facility at

a reactor core power level not in excess of 70% of Rated Thermal Power (2406.6

megawatts thermal).

3.0 TECHNICAL EVALUATION

3.1 Background

On January 31, 2012, SONGS Unit 2 was in a refueling and steam generator inspection

outage, and SONGS Unit 3 was in Mode 1 operating at 100% power. A leak was

detected in a SONGS Unit 3 steam generator. SCE operators promptly shut down the

unit in accordance with plant operating procedures. The leak resulted in a small

radioactive release to the environment that was well below the allowable federal limits.

The steam generator tube wear that caused the SONGS Unit 3 steam generator tube

leak was the result of tube-to-tube interaction. This type of wear was confirmed to exist

in a number of other tubes in the same region in both SONGS Unit 3 steam generators.

Subsequent inspections of the SONGS Unit 2 steam generators identified this type of

wear also existed in a pair of tubes (one contact location) in SONGS Unit 2 steam

generator E-089. On March 27, 2012, the NRC issued a Confirmatory Action Letter

(CAL) to SCE describing actions that the NRC and SCE agreed will be completed prior

to returning SONGS Units 2 and 3 to service (Reference 6.2).

SCE's response to the CAL, which provides the basis for the conclusion that steam

generator tube integrity will be maintained assuming operation up to 70% Rated

Thermal Power, is provided in a letter from Mr. Peter T. Dietrich (SCE) to Mr. Elmo E.

Collins (USNRC), dated October 3, 2012, Confirmatory Action Letter - Actions to

Address Steam Generator Tube Degradation, San Onofre Nuclear Generating Station,

Unit 2 (ML12285A263) (Reference 6.1). That document is being reviewed separately

by the NRC as part of the CAL process. Final determination on the adequacy of the

actions being taken to support operation at reduced power will be made pursuant to the

CAL restart approval process.

The proposed changes maintain consistency for Cycle 17 between Technical

Specification 5.5.2.11.b.1 and the maximum power level in License Condition 2.C(1).

Following NRC approval of this change, Technical Specification 5.5.2.11 will continue to

require that steam generator tube integrity be retained over the range of power levels

which the plant will be allowed operate.

On March 14, 2013, SCE submitted an OA for operation of SONGS Unit 2 at 100%

power (Reference 6.3). Due to the time needed for the NRC staff to review the OA for

100% power, it is possible that the staffs review may not be complete prior to May 24,

3 of 14
2013. As a result, SCE is submitting this license amendment request to allow for the

timely restart of SONGS Unit 2.

3.2 Discussion

The proposed changes require that compliance with the SIPC be demonstrated up to

70% Rated Thermal Power (2406.6 megawatts thermal) for SONGS Unit 2, Cycle 17

and that reactor core thermal power level be limited to 70% Rated Thermal Power

during Cycle 17. The duration of one cycle was chosen to allow SCE to identify long

term corrective actions, including identification of a long-term maximum power level,

and obtain necessary NRC approval for those actions.

3.2.1 Proposed License and Technical Specification Changes

The current Technical Specification 5.5.2.11 requires that steam generator tube

structural integrity be retained over the "full range of normal operating conditions" and

that a safety factor of 3.0 against burst be retained for primary-to-secondary pressure

differential "under normal steady state full power operation." The proposed change

provides a footnote that states that for Cycle 17 the terms "full range of normal

operating conditions" and "normal steady state full power operation" refer to steam

generator conditions associated with operation up to and including 70% Rated Thermal

Power.

The proposed change to Facility Operating License Condition 2.C(1) reduces the

maximum power level for operation of SONGS Unit 2 for the next refueling cycle

(Cycle 17). This proposed change is consistent with the power level commitment in

Reference 6.1.

The proposed changes maintain consistency between Technical Specification 5.5.2.11

and the maximum power level in License Condition 2.C(1) for Cycle 17. All other

aspects of the Steam Generator Program, including methods and acceptance criteria

used to perform operational assessment and demonstrate steam generator tube

integrity, are unaffected by the proposed changes.

3.2.2 Impact of Reduced Power Operations on Safety Analysis

The justification for safe operation at 70% Rated Thermal Power is provided in a letter

from Mr. Peter T. Dietrich (SCE) to Mr. Elmo E. Collins (USNRC), dated October 3,

2012, Confirmatory Action Letter - Actions to Address Steam Generator Tube

Degradation, San Onofre Nuclear Generating Station, Unit 2 (ML12285A263)

(Reference 6.1). As part of the NRC review of SCE's response to the CAL, the NRC

has provided several requests for additional information (RAIs). SCE has submitted

response to most of those RAIs, including RAIs 11 through 14 that pertain to impact of

operation at 70% rated Thermal Power on plant systems and the safety analysis

(References 6.4 through 6.7). These RAI responses are being reviewed separately by

4 of 14
the NRC as part of the CAL process and the conclusions are summarized here. These

responses are also incorporated by reference into this LAR.

The response to RAI 11 (Reference 6.4) evaluates the impacts of operation at 70%

Rated Thermal Power on the Unit 2, Cycle 17 core design and reload analysis, including

the Updated Final Safety Analysis Report (UFSAR) Chapter 15 safety analyses. Power

levels evaluated range from 50% to 100% which bound 70% power operation. The

areas evaluated were:

e Safety Analysis Methodology - The NRC approved safety analysis methods, as

described in TS 5.7.1.5, are used to establish the core operating limits specified

in the Core Operating Limits Report (COLR) which encompass from Mode 6 up

to Mode 1 operation at the Rated Thermal Power. Therefore, operating at the

70% Rated Thermal Power level is within the scope of SCE safety analysis

methodology. No change to the safety analysis methodology is required for

reduced power operation.

0 Safety Analysis - All safety and reload analyses were assessed for impact. The

reload and safety analyses determined to be impacted by reduced power

operation were re-analyzed. The conclusions of the reload analyses, including

safety analyses, for reduced power operation are as follows: (1) All safety

analyses results meet the established acceptance criteria, and (2) The

radiological dose consequences for all safety analyses remain bounded by the

dose consequences reported in the UFSAR. This assessment is summarized in

Table 1 of RAI 11 (Reference 6.4).

* Technical Specifications - A review has been performed to determine the impact

of extended power operation at 70% Rated Thermal Power on SONGS Unit 2

Technical Specifications (TS). The review addressed the Limiting Condition for

Operation (LCO), applicability, action statements, and surveillances. With the

exception of the proposed change to Technical Specification 5.5.2.11, the review

determined that the current TS are applicable to extended power operation at

70% Rated Thermal Power, and that no other TS changes are required.

* Impact Assessment for Technical Specification Surveillance Requirements - The

Technical Speciation surveillance requirements were evaluated for the impacts of

reduced power operation. Two surveillance procedures related to monitoring

Reactor Coolant System (RCS) flow that had been required to be performed only

above 85% Rated Thermal Power were revised to require performance at 70%

Rated Thermal Power. No other surveillances were identified to be impacted by

plant operation at 70% Rated Thermal Power.

This assessment determined operation at reduced power is acceptable with respect to

SONGS Unit 2, Cycle 17 reload core design and safety analysis. All Technical

Specification LCO's, applicability, action statements, and surveillance requirements

under the reactor core design and monitoring program normally performed at or above

5 of 14
70% Rated Thermal Power will be performed with the plant operating at approximately

70% Rated Thermal Power.

It was identified that one Technical Specification reporting requirement is affected by the

length of the initial proposed operating period. Technical Specification 5.7.2.c requires

that "A report shall be submitted within 180 days after the initial entry into Mode 4

following completion of an inspection performed in accordance with the Specification

5.5.2.11." Given the initial operating period is 150 days and the uncertain length of

subsequent operating periods, SCE commits that for Unit 2 Cycle 17 SCE will provide

the report required by Technical Specification 5.7.2.c within 60 days after the initial entry

into Mode 4 following completion of an inspection performed in accordance with the

Specification 5.5.2.11.

The response to RAI 12 (Reference 6.5) evaluates Reactor Coolant System (RCS) flow

uncertainty as affected by power operation at 70% Rated Thermal Power. This

assessment performed an RCS flow uncertainty analysis for extended operation at 70%

Rated Thermal Power. The assessment determined that the increase in secondary

calorimetric power uncertainty and RCS flow uncertainties are accounted for in the

overall uncertainty analysis required by the reload safety analyses as detailed in the

response to RAI 11.

The response to RAI 13 (Reference 6.6) evaluates how the existing Emergency Core

Cooling System (ECCS) analysis accounts for the changes to the steam generator heat

transfer characteristics resulting from the installation of the new steam generators and

operation at 70% Rated Thermal Power. RAI 13 documents that operation at 70%

Rated Thermal Power remains bounded by the current SONGS Unit 2 ECCS

performance Analyses of Record (AORs) for Large Break Loss-of-Coolant Accident

(LBLOCA), Small Break Loss-of-Coolant Accident (SBLOCA), and post Loss-of-Coolant

Accident (LOCA) Long Term Cooling (LTC).

The response to RAI 14 (Reference 6.7) dispositions the mechanical and

instrumentation and control (I&C) design evaluations performed to support operation at

70% Rated Thermal Power. These evaluations are in addition to the evaluations

addressed in RAIs 11, 12 and 13. The mechanical design evaluations assessed the

reactor vessel internals (RVI) analyses, reactor coolant system (RCS) structural

analyses, loss of coolant accident (LOCA) hydraulic blowdown loads analyses, and

RCS natural circulation analyses of record (AORs). The results bound and support the

planned operation of Unit 2 at 70% power with approximately 3% Steam Generator

Tube Plugging. The I&C design evaluations assessed potentially impacted systems

including the main steam, main feedwater, and steam generator blowdown flow

instruments, the secondary calorimetric function of the core operating limits supervisory

system (COLSS), the steam generator narrow-range (NR) water level instruments, the

plant protection system (PPS) setpoints, the steam bypass control system (SBCS), the

digital feedwater control system (DFWCS), and the pressurizer level setpoint and RCS

reference temperature (TREF) programs. The results and conclusions of the evaluations

6 of 14
were that power operation at 70% Rated Thermal Power did not significantly affect the

systems and programs.

3.2.3 Steam Generator Tube Integrity

The proposed changes will maintain consistency between the power level referenced in

Technical Specification 5.5.2.11 for assessments of steam generator tube integrity and

the maximum power level in License Condition 2.C(1) for Cycle 17. The proposed

changes will ensure that steam generator tube integrity is demonstrated for the

maximum power level specified in License Condition 2.C(1).

The impact of operation at 70% Rated Thermal Power on steam generator tube integrity

was provided in Reference 6.1 (ADAMS Package No. 122850320). The Operational

Assessments provided in Reference 1 demonstrate that limiting SONGS Unit 2 power to

70% eliminates the thermal hydraulic conditions that cause fluid elastic instability (FEI)

in the SONGS Unit 2 steam generators by reducing the steam velocity and void fraction.

Further, at 70% Rated Thermal Power the SONGS Unit 2 steam generators will operate

within an envelope of steam velocity and void fraction that has proven successful in the

operation of similar steam generators. Operational Assessments also demonstrate that

limiting power to 70% provides assurance that loss of tube integrity due to FEI will not

occur. The impact of operation at 70% Rated Thermal Power on steam generator tube

integrity is being reviewed separately as part of the CAL process.

3.2.4 Risk Insights

Based on a review of the evaluations performed by station personnel on the potential

operational, engineering analyses, and system impacts of operating at 70% versus

100% Rated Thermal Power, SCE has concluded that reduced power operations would

have no quantifiable adverse impact on plant systems performance, operations

response, or the frequencies of initiating events. Therefore, no quantifiable adverse

impact on plant risk was identified in terms of the core damage frequency (CDF) or

large early release frequency (LERF) if the plant is operating at 70% versus 100%

Rated Thermal Power for the operating cycle.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria

There are no General Design Criteria (GDC) in 10 CFR Part 50 Appendix A that directly

address maximum power level. GDC 10, 11, 12, 13, and 14, of 10 CFR Part 50,

Appendix A, define requirements for reactivity and power control. The evaluations

discussed in Section 3.2 above demonstrate compliance with these GDC over the

expected range of operating conditions, including the entire proposed power operating

range. Therefore, the proposed change to Facility Operating License Condition 2.C(1)

7 of 14
to limit power operation to 70% Rated Thermal Power does not affect the ability of

SONGS Unit 2 to comply with these GDC.

General Design Criteria (GDC) 30, 31 and 32 of 10 CFR Part 50, Appendix

A define requirements for the reactor coolant pressure boundary (RCPB) with respect to

structural integrity. Steam generator tubes and associated components, such as plugs,

must be capable of maintaining reactor coolant inventory and pressure. Technical

Specification requirements for steam generator tubes and associated components

ensure that these GDC are met by ensuring steam generator tube integrity.

The proposed amendment requests that Technical Specification 5.5.2.11.b.1 be revised

to require that compliance with the steam generator tube structural integrity

performance criterion (SIPC) be demonstrated up to 70% Rated Thermal Power (2406.6

megawatts thermal) for SONGS Unit 2, Cycle 17, and requests that License Condition

2.C(1) be revised to limit reactor core thermal power level to 70% Rated Thermal Power

during Cycle 17. Thus, steam generator tube integrity will be required to be assessed at

the maximum power level specified in License Condition 2.C(1).

The operational assessments that demonstrate steam generator tube integrity at 70%

Rated Thermal Power are provided in a letter from Mr. Peter T. Dietrich (SCE) to Mr.

Elmo E. Collins (USNRC), dated October 3, 2012, Confirmatory Action Letter - Actions

to Address Steam Generator Tube Degradation, San Onofre Nuclear Generating

Station, Unit 2 (ML12285A263) (Reference 6.1) and are being reviewed separately by

the NRC as part of the NRC's review of the Confirmatory Action Letter (CAL) response.

Those documents are incorporated by reference.

Therefore, the proposed changes to License Condition 2.C(1) and Technical

Specification 5.5.2.11 .b.1 will not adversely affect compliance with the GDC on the

RCPB.

4.2 No Significant Hazards Consideration

The proposed amendment requires that compliance with the steam generator structural

integrity performance criterion (SIPC) be demonstrated up to 70% Rated Thermal

Power (2406.6 megawatts thermal) for SONGS Unit 2 Cycle 17 and amends License

Condition 2.C(1) to specify a maximum power level of SONGS Unit 2 to 70% during

Cycle 17.

The proposed changes are in support of re-start of SONGS Unit 2 following corrective

actions taken to address steam generator tube wear identified during the SONGS Unit 2

Cycle 17 refueling and steam generator inspection outage. The proposed changes

maintain consistency between Technical Specification 5.5.2.11 and the maximum power

level in License Condition 2.C(1), in that steam generator tube integrity must be

demonstrated assuming operation up to the maximum power level specified in the

license.

8 of 14
The basis for operation of SONGS at 70% Rated Thermal Power, which is already

authorized by the SONGS Unit 2 operating license, is provided in a letter from Mr. Peter

T. Dietrich (SCE) to Mr. Elmo E. Collins (USNRC), dated October 3, 2012, Confirmatory

Action Letter - Actions to Address Steam Generator Tube Degradation, San Onofre

Nuclear Generating Station, Unit 2 (ML12285A263)(1 ). The impact on steam generator

tube integrity is evaluated in the operational assessments provided in that letter and is

being reviewed separately by the NRC.

SCE has evaluated whether or not a significant hazards consideration is involved with

the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92,

"Issuance of Amendment", as discussed below:

1. Does the proposed change involve a significant increase in the

probability or consequences of an accident previously evaluated?

Response: No.

The proposed change-would reduce the power level for San Onofre Nuclear

Generating Station (SONGS) Unit 2 for Cycle 17.

The proposed changes do not affect the probability of any accident initiators

because there is no adverse effect on plant operations or plant conditions. The

proposed revisions will require that tube integrity will be demonstrated up to 70%

Rated Thermal Power and that operation of the SONGS Unit 2 steam generators

will be limited to a maximum reactor power level of 70%. As a result, the change

will continue to ensure that tube integrity is retained over the range of power

levels at which the plant will operate. The proposed change to reduce the power

level will not affect the probability of any accident initiators because the only

effect on plant operations is to lower the allowable power level. Operation at

reduced power is acceptable under the current licensing basis and operation of

the plant will remain bounded by the assumptions of the analyses of accidents

previously evaluated in the UFSAR.

The proposed changes do not create any new mode of plant operation. The

proposed changes do not result in a change to any design function nor do the

proposed changes affect any analysis that verifies the capability of a system,

structure or component to perform a design function.

The proposed changes maintain consistency between the conditions for which

the structural integrity performance criteria (SIPC) must be demonstrated with the

maximum power levels at which SONGS Unit 2 will be authorized to operate.

SCE will be required to perform operational assessments for steam generator

tube integrity as required by the current Technical Specifications for the range of

power levels within which SONGS Unit 2 will operate during Cycle 17. The

proposed changes do not affect the tube performance criteria for those

assessments with the exception that they will be applied over a range of

9 of 14
conditions up to and including 70% Rated Thermal Power. Therefore the

proposed changes do not involve a significant increase in the probability or

consequences of a steam generator tube failure under normal or postulated

accident conditions.

As part of the NRC review of SCE's response to the CAL(1 ), the NRC provided

several Requests for Additional Information (RAIs). The response to RAI 11(2)

assessed safety analysis methods, analysis acceptance criteria, radiological

dose consequences, applicability of Technical Specifications Limiting Conditions

for Operation, applicability, action statements, surveillances and impact on

performance of Surveillance Requirements. The assessment determined

operation at reduced power is acceptable with respect to SONGS Unit 2, Cycle

17 reload core design and safety analysis.

The assessment that evaluated how Reactor Coolant System (RCS) flow

uncertainty is affected by operation (3F at 70% Rated Thermal Power was provided

in the response to RAI 12 . This assessment determined that the increase in

secondary calorimetric power uncertainty and RCS flow uncertainties are

accounted for in the overall uncertainty analysis required by the reload safety

analyses as detailed in the response to RAI 11.

The assessment that evaluated how the existing Emergency Core Cooling

System (ECCS) analysis accounts for the changes to the steam generator heat

transfer characteristics resulting from the installation of the new steam

generators was provided in the response to RAI 13(4). The results and

conclusions of this evaluation was that operation at 70% Rated Thermal Power

remains bounded by the current SONGS Unit 2 ECCS performance Analyses of

Record (AOR) for Large Break Loss- of-Coolant Accident (LBLOCA), Small

Break Loss- of-Coolant Accident (SBLOCA), and post Loss-of-Coolant Accident

(LOCA) Long Term Cooling (LTC).

The assessment that evaluated how the mechanical and instrumentation and

control (I&C) design evaluations performed to support operation at 70% Rated

Thermal Power was provided in the response to RAI 1 4(5). These evaluations are

in addition to the evaluations addressed in RAIs 11, 12 and 13. The results and

conclusions of the RAI 14 evaluations were that power operation at 70% Rated

Thermal Power will either remain bounded or will not significantly affect the

associated systems and programs.

As a result, operation at 70% Rated Thermal Power does not result in a

significant increase in the probability or consequences of a previously evaluated

accident.

Therefore, the proposed changes do not involve a significant increase in the

probability or consequences of an accident previously evaluated.

10 of 14
2. Does the proposed change create the possibility of a new or different

kind of accident from any accident previously evaluated?

Response: No

The proposed changes do not require a change in any plant systems, structures

or components or the method of operating the plant other than to reduce power

for the duration of Cycle 17. Therefore, the proposed changes do not create the

possibility of a new or different kind of accident from any accident previously

evaluated.

The proposed changes do not adversely affect the method of operation of the

steam generators nor introduce any changes to existing design functions of

systems, structures or components that could create the possibility of a new or

different kind of accident from any previously evaluated. Also, the proposed

change will not introduce any significant changes to postulated accidents

resulting from potential tube degradation. Because SONGS Unit 2 will operate at

or below 70% Rated Thermal Power, the change will continue to ensure that tube

integrity is demonstrated over the range of power levels at which the plant will

operate. Therefore, there is no significant increase in the probability that the

tubes will fail or leak during the period.

Therefore, the proposed changes do not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of

safety?

Response: No

The proposed change to reduce the power level during SONGS Unit 2, Cycle 17,

will not reduce any margins of safety.

The effect of operation at 70% Rated Thermal Power does not result in SONGS

Unit 2 being operated outside of any currently allowable regulatory or licensing

limits. SONGS Unit 2 will operate within its current design and licensing basis.

Therefore, this change has no significant impact on any parameter that would

affect a design basis limit for a fission product barrier, and there would be no

significant impact on any margin of safety associated with such barriers.

The proposed changes will maintain consistency between the power level for the

steam generator operational assessments in Technical Specification 5.5.2.11

and the maximum power level specified in License Condition 2.C(1). The tube

performance criteria, including the margins of safety specified in Technical

Specification 5.5.2.11 are not being changed. The changes will ensure that tube

11 of 14
integrity is demonstrated against the unchanged performance criteria over the

range of power levels at which the plant will be licensed to operate.

Therefore, the proposed changes do not involve a significant reduction in a

margin of safety.

Based on the above, SCE concludes that the proposed amendment does not involve a

significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and

accordingly, a finding of "no significant hazards consideration" is justified.

(1) Letter from Mr. Peter T. Dietrich (SCE) to Mr. Elmo E. Collins (USNRC), dated

October 3, 2012, Confirmatory Action Letter - Actions to Address Steam

Generator Tube Degradation, San Onofre Nuclear Generating Station, Unit 2

(ML12285A263)

(2) Letter from Mr. Richard J. St Onge to NRC, dated April 2, 2013, Response to

Request for Additional Information (RAI 11), Revision 1, Regarding Confirmatory

Action Letter Response (TAC No. ME 9727), San Onofre Nuclear Generating

Station, Unit 2

(3) Letter from Mr. Richard J. St Onge to NRC, dated January 18, 2013, Response to

Request for Additional Information (RAI 12) Regarding Confirmatory Action Letter

Response (TAC No. ME 9727), San Onofre Nuclear Generating Station, Unit 2

(ML13022A408)

(4) Letter from Mr. Richard J. St Onge to NRC, dated April 2, 2013, Response to

Request for Additional Information (RAI 13), Revision 1, Regarding Confirmatory

Action Letter Response (TAC No. ME 9727), San Onofre Nuclear Generating

Station, Unit 2

(5) Letter from Mr. Richard J. St Onge to NRC, dated January 29, 2013, Response to

Request for Additional Information (RAI 14) Regarding Confirmatory Action Letter

Response (TAC No. ME 9727), San Onofre Nuclear Generating Station, Unit 2

(ML13032A009)

4.3 Conclusions

In conclusion, based on the considerations discussed above, (1) there is reasonable

assurance that the health and safety of the public will not be endangered by operation in

the proposed manner, (2) such activities will be conducted in compliance with the

Commission's regulations, and (3) the issuance of the amendment will not be inimical to

the common defense and security or to the health and safety of the public

12 of 14
5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement

with respect to installation or use of a facility component, the steam generators, located

within the restricted area, as defined in 10 CFR 20. However, the proposed amendment

does not involve (i) a significant hazards consideration, (ii) a significant change in the

types or significant increase in the amounts of any effluent that may be released offsite,

or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendments meet the eligibility criteria for categorical

exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no

environmental impact statement or environmental assessment need be prepared in

connection with this proposed amendment.

6.0 REFERENCES

6.1. Letter from Mr. Peter T. Dietrich (SCE) to Mr. Elmo E. Collins (USNRC), dated

October 3, 2012, Confirmatory Action Letter - Actions to Address Steam

Generator Tube Degradation, San Onofre Nuclear Generating Station, Unit 2

(ML12285A263)

6.2. Letter from Mr. Elmo E. Collins (USNRC) to Mr. Peter T. Dietrich (SCE), dated

March 27, 2012, Confirmatory Action Letter 4-12-001, San Onofre Nuclear

Generating Station, Units 2 and 3, Commitments to Address Steam Generator

Tube Degradation (ML12087A323)

6.3. Letter from Mr. Richard J. St Onge to NRC, dated March 14, 2013, Operational

Assessment for 100% Power Case Regarding Confirmatory Action Letter

Response, San Onofre Nuclear Generating Station, Unit 2 (ML13074A793)

6.4. Letter from Mr. Richard J. St Onge to NRC, dated April 2, 2013, Revised

Response to Request for Additional Information (RAI 11), Revision 1,

Regarding Confirmatory Action Letter Response (TAC No. ME 9727), San

Onofre Nuclear Generating Station, Unit 2

6.5. Letter from Mr. Richard J. St Onge to NRC, dated January 18, 2013, Response

to Request for Additional Information (RAI 12) Regarding Confirmatory Action

Letter Response (TAC No. ME 9727), San Onofre Nuclear Generating Station,

Unit 2 (ML13022A408)

6.6. Letter from Mr. Richard J. St Onge to NRC, dated April 2, 2013, Response to

Request for Additional Information (RAI 13), Revision 1, Regarding

Confirmatory Action Letter Response (TAC No. ME 9727), San Onofre Nuclear

Generating Station, Unit 2

13 of 14
6.7. Letter from Mr. Richard J. St Onge to NRC, dated January 29, 2013, Response

to Request for Additional Information (RAI 14) Regarding Confirmatory Action

Letter Response (TAC No. ME 9727), San Onofre Nuclear Generating Station,

Unit 2 (ML13032A009)

14 of 14
Attachment 1

List of Regulatory Commitments
Commitment # One-Time Only Sustainable Due Date

Submit and obtain NRC Prior to Mode 4

approval of a License X entry for Unit 2

Amendment Request for long Cycle 18

term operation of Unit 2 beyond operation

Cycle 17.

For Unit 2 Cycle 17, provide the X Within 60 days

report required by Technical after the initial

Specification 5.7.2.c. entry into Mode

4 following

completion of an

inspection

performed in

accordance with

the Specification

5.5.2.11, Steam

Generator

Program
Attachment 2

Existing Facility Operating Pages, Unit 2
-3-

(3) SCE, pursuant to the Act and 10 CFR Part 70, to receive,

possess, and use at any time special nuclear material as reactor

fuel, in accordance with the limitations for storage and amounts

required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended;

(4) SCE, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to

receive, possess, and use at any time any byproduct, source and

special nuclear material as sealed neutron sources for reactor

startup, sealed sources for reactor instrumentation and

radiation monitoring equipment calibration, and as fission

detectors in amounts as required;

(5) SCE, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to

receive, possess, and use in amounts as required any byproduct,

source or special nuclear material without restriction to

chemical or physical form, for sample analysis or instrument

calibration or associated with radioactive apparatus or

components; and

(6) SCE, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to

possess, but not separate, such byproduct and special nuclear

materials as may be produced by the operation of San Onofre

Nuclear Generating Station, Units . and 2 and by the

decommissioning of San Onofre Nuclear Generating Station Unit 1.

C. This license shall be deemed to contain and is subject to the

conditions specified in the Commission's regulations set forth in

10 CFR Chapter I and is subject to all applicable provisions of the

Act and to the rules, regulations and orders of the Commission now or

hereafter in effect; and is subject to the additional conditions

specified or incorporated below:

(1) Maximum Power Level

Southern California Edison Company (SCE) is authorized to

operate the facility at reactor core power levels not in excess

of full power (3438 megawatts thermal).

(2) Technical Specifications

The Technical Specifications contained in Appendix A and the

Environmental Protection Plan contained in Appendix B, as

revised through Amendment No. 226, are hereby incorporated in

the license. Southern California Edison Company shall operate

the facility in accordance with the Technical Specifications and

the Environmental Protection Plan.

Amendment No. 226
Procedures, Programs, and Manuals

5.5

5.5 Procedures, Programs, and Manuals (continued)

5.5.2.11 Steam Generator (SG) Program (continued)

b. Performance criteria for SG tube integrity. SG tube integrity

shall be maintained by meeting the performance criteria for

tube structural integrity, accident induced leakage, and

operational LEAKAGE.

1. Structural integrity performance criterion: All inservice steam generator tubes shall retain structural

integrity over the full range of normal operating

conditions (including startup, operation in the power

range, hot standby, and cool down and all anticipated

transients included in the design specification) and

design basis accidents. This includes retaining a safety

factor of 3.0 against burst under normal steady state

full power operation primary-to-secondary pressure

differential and a safety factor of 1.4 against burst

applied to the design basis accident primary-to-secondary

pressure differentials. Apart from the above

requirements, additional loading conditions associated

with the design basis accidents, or combination of

accidents in accordance with the design and licensing

basis, shall also be evaluated to determine if the

associated loads contribute significantly to burst or

collapse. In the assessment of tube integrity, those

loads that do significantly affect burst or collapse

shall be determined and assessed in combination with the

loads due to pressure with a safety factor of 1.2 on the

combined primary loads and 1.0 on axial secondary loads.

2. Accident induced leakage performance criterion: The

primary to secondary accident induced leakage rate for

any design basis accident, other than a SG tube rupture,

shall not exceed the leakage rate assumed in the accident

analysis in terms of total leakage rate for all SGs and

leakage rate for an individual SG. Leakage is not to

exceed 0.5 gpm per SG and 1 gpm through both SGs.

3. The operational LEAKAGE performance criterion is

specified in LCO 3.4.13, "RCS Operational LEAKAGE."

(continued)

SAN ONOFRE--UNIT 2 5.0-14 Amendment No. 140,204
-3-

(3) SCE, pursuant to the Act and 10 CFR Part 70, to receive,

possess, and use at any time special nuclear material as reactor

fuel, in accordance with the limitations for storage and amounts

required for reactor operation, as described in the Final Safety

Analysis Report, as supplemented and amended;

(4) SCE, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to

receive, possess, and use at any time any byproduct, source and

special nuclear material as sealed neutron sources for reactor

startup, sealed sources for reactor instrumentation and

radiation monitoring equipment calibration, and as fission

detectors in amounts as required;

(5) SCE, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to

receive, possess, and use in amounts as required any byproduct,

source or special nuclear material without restriction to

chemical or physical form, for sample analysis or instrument

calibration or associated with radioactive apparatus or

components; and

(6) SCE, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to

possess, but not separate, such byproduct and special nuclear

materials as may be produced by the operation of San Onofre

Nuclear Generating Station, Units I and 2 and by the

decommissioning of San Onofre Nuclear Generating Station Unit 1.

C. This license shall be deemed to contain and is subject to the

conditions specified in the Commission's regulations set forth in

10 CFR Chapter I and is subject to all applicable provisions of the

Act and to the rules, regulations and orders of the Commission now or

hereafter in effect; and is subject to the additional conditions

specified or incorporated below:

(1) Maximum Power Level

Southern California Edison Company (SCE) is authorized to

operate the facility at reactor core power levels not in excess

of full power (3438 megawatts thermal)*.

(2) Technical Specifications

The Technical Specifications contained in Appendix A and the

Environmental Protection Plan contained in Appendix B, as

revised through Amendment No. , are hereby incorporated in

the license. Southern California Edison Company shall operate the facility in accordance with the Technical Specifications and

the Environmental Protection Plan.

*For the duration of Unit 2, Cycle 17, SCE is authorized to operate the

facility at a reactor core power level not in excess of 70% of Rated Thermal

Power (2406.6 megawatts thermal).

Amendment No. XXX
Procedures, Programs, and Manuals

5.5

5.5 Procedures, Programs, and Manuals (continued)

5.5.2.11 Steam Generator (SG) Program (continued)

b. Performance criteria for SG tube integrity. SG tube integrity

shall be maintained by meeting the performance criteria for

tube structural integrity, accident induced leakage, and

operational LEAKAGE.

1. Structural integrity performance criterion: All inservice steam generator tubes shall retain structural

integrity over the full range of normal operating

conditions (including startup, operation in the power

range, hot standby, and cool down and all anticipated

transients included in the design specification) and

design basis accidents. This includes retaining a safety

factor of 3.0 against burst under normal steady state

full power operation primary-to-secondary pressure

differential and a safety factor of 1.4 against burst

applied to the design basis accident primary-to-secondary

pressure differentials. Apart from the above

requirements, additional loading conditions associated

with the design basis accidents, or combination of

accidents in accordance with the design and licensing

basis, shall also be evaluated to determine if the

associated loads contribute significantly to burst or

collapse. In the assessment of tube integrity, those

loads that do significantly affect burst or collapse

shall be determined and assessed in combination with the

loads due to pressure with a safety factor of 1.2 on the

combined primary loads and 1.0 on axial secondary loads.*

2. Accident induced leakage performance criterion: The

primary to secondary accident induced leakage rate for

any design basis accident, other than a SG tube rupture,

shall not exceed the leakage rate assumed in the accident

analysis in terms of total leakage rate for all SGs and

leakage rate for an individual SG. Leakage is not to

exceed 0.5 gpm per SG and I gpm through both SGs.

3. The operational LEAKAGE performance criterion is

specified in LCO 3.4.13, "RCS Operational LEAKAGE."

*For the duration of Unit 2, Cycle 17, the terms "full range of normal operating

conditions" and "normal steady state full power operation" shall be based upon

the steam generators being operated under conditions associated with reactor core

power levels up to 70% of Rated Thermal Power (2406.6 megawatts thermal). During

Cycle 17, Unit 2 will not be operated above this power level.

(continued)

SAN ONOFRE--UNIT 2 5.0-14 Amendment No. -1W,2O4GXXX
Attachment 4

Clean Facility Operating Pages, Unit 2
-3-

(3) SCE, pursuant to the Act and 10 CFR Part 70, to receive,

possess, and use at any time special nuclear material as reactor

fuel, in accordance with the limitations for storage and amounts

required for reactor operation, as described in the Final Safety

Analysis Report, as supplemented and amended;

(4) SCE,. pursuant to the Act and 10 CFR Parts 30, 40, and 70, to

receive, possess, and use at any time any byproduct, source and

special nuclear material as sealed neutron sources for reactor

startup, sealed sources for reactor instrumentation and

radiation monitoring equipment calibration, and as fission

detectors in amounts as required;

(5) SCE, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to

receive, possess, and use in amounts as required any byproduct,

source or special nuclear material without restriction to

chemical or physical form, for sample analysis or instrument

calibration or associated with radioactive apparatus or

components; and

(6) SCE, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to

possess, but not separate, such byproduct and special nuclear

materials as may be produced by the operation of San Onofre

Nuclear Generating Station, Units 1 and 2 and by the

decommissioning of San Onofre Nuclear Generating Station Unit 1.

C. This license shall be deemed to contain and is subject to the

conditions specified in the Commission's regulations set forth in

10 CFR Chapter I and is subject to all applicable provisions of the

Act and to the rules, regulations and orders of the Commission now or

hereafter in effect; and is subject to the additional conditions

specified or incorporated below:

(1) Maximum Power Level

Southern California Edison Company (SCE) is authorized to

operate the facility at reactor core power levels not in excess

of full power (3438 megawatts thermal)*.

(2) Technical Specifications

The Technical Specifications contained in Appendix A and the

Environmental Protection Plan contained in Appendix B, as

revised through Amendment No. , are hereby incorporated in

the license. Southern California Edison Company shall operate

the facility in accordance with the Technical Specifications and

the Environmental Protection Plan.

*For the duration of Unit 2, Cycle 17, SCE is authorized to operate the

facility at a reactor core power level not in excess of 70% of Rated Thermal

Power (2406.6 megawatts thermal).

Amendment No.
Procedures, Programs, and Manuals

5.5

5.5 Procedures, Programs, and Manuals (continued)

5.5.2.11 Steam Generator (SG) Program (continued)

b. Performance criteria for SG tube integrity. SG tube integrity

shall be maintained by meeting the performance criteria for

tube structural integrity, accident induced leakage, and

operational LEAKAGE.

1. Structural integrity performance criterion: All inservice steam generator tubes shall retain structural

integrity over the full range of normal operating

conditions (including startup, operation in the power

range, hot standby, and cool down and all anticipated

transients included in the design specification) and

design basis accidents. This includes retaining a safety

factor of 3.0 against burst under normal steady state

full power operation primary-to-secondary pressure

differential and a safety factor of 1.4 against burst

applied to the design basis accident primary-to-secondary

pressure differentials. Apart from the above

requirements, additional loading conditions associated

with the design basis accidents, or combination of

accidents in accordance with the design and licensing

basis, shall also be evaluated to determine if the

associated loads contribute significantly to burst or

collapse. In the assessment of tube integrity, those

loads that do significantly affect burst or collapse

shall be determined and assessed in combination with the

loads due to pressure with a safety factor of 1.2 on the

combined primary loads and 1.0 on axial secondary loads.*

2. Accident induced leakage performance criterion: The

primary to secondary accident induced leakage rate for

any design basis accident, other than a SG tube rupture,

shall not exceed the leakage rate assumed in the accident

analysis in terms of total leakage rate for all SGs and

leakage rate for an individual SG. Leakage is not to

exceed 0.5 gpm per SG and 1 gpm through both SGs.

3. The operational LEAKAGE performance criterion is

specified in LCO 3.4.13, "RCS Operational LEAKAGE."

*For the duration of Unit 2, Cycle 17, the terms "full range of normal operating

conditions" and "normal steady state full power operation" shall be based upon

the steam generators being operated under conditions associated with reactor core

power levels up to 70% of Rated Thermal Power (2406.6 megawatts thermal). During

Cycle 17, Unit 2 will not be operated above this power level.

(continued)

SAN ONOFRE--UNIT 2 5.0-14 Amendment No.