Text

Amendment Application No. 236 Proposed Change Number (PCN) 582 Technical Specification (TS) 5.5.2.15 Containment Leakage Rate Testing Program
	ML072681094
Person / Time
Site:	San Onofre ; (NPF-015)
Issue date:	09/24/2007
From:	Katz B; Southern California Edison Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RSC 07-05, Rev 0
	Download: ML072681094 (126)
	v • d • e

SOUTHERN CALIFORNIA Brian Katz 5 I EDISON Vice President An EDISON INTERNATIONAL,' Company September 24, 2007 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555

Subject:

Docket No. 50-362 Amendment Application No. 236 Proposed Change Number (PCN) 582 Technical Specification (TS) 5.5.2.15 Containment Leakage Rate Testing Program San Onofre Nuclear Generating Station Unit 3

Dear Sir or Madam:

Southern California Edison (SCE) hereby requests approval of Amendment Application Number 236, which consists of PCN 582. PCN 582 proposes to revise the San Onofre Nuclear Generating Station (SONGS) Unit 3 Technical Specification TS 5.5.2.15, "Containment Leakage Rate Testing Program." The request and supporting analysis is for a one-time extension from the currently approved 15-year interval since the last ILRT to a 16 year interval since the last Integrated Leak Rate Test (ILRT). As planned, the effect of the proposed change, however, will result in SONGS Unit 3 operating a few months past the current required ILRT date: September 9, 2010. This proposed change allows deferral of the next ILRT Type A Test from September 9, 2010 to prior to startup from the Unit 3 Cycle 16 refueling outage.

The last SONGS Unit 3 ILRT was performed in September 1995. The next ILRT is required to be performed by September 10, 2010. The Unit 3 Cycle 16 refueling outage is the scheduled steam generator replacement (SGR) outage for SONGS Unit 3, which is currently scheduled to start in the fall of 2010 and to end in the first quarter of 2011. The SGR activities include breaching the containment structure. The current TS requires the ILRT to be performed no later than September 9, 2010, which is about two months before the start of the SGR outage. This proposed change is based on and has been evaluated using the "risk informed" guidance in RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant -Specific Changes to the Licensing Basis."

P.O. Box 128 San Clemente. CA 92674-0128 40i'?

949-368-9275 Fax 949-368-9881

Document Control Desk September 24, 2007 SCE is following precedent set by license amendments issued by the NRC for Florida Power & Light Company (FPL), St. Lucie 2 (ML053190343) and Seabrook Station (ML060520032), and by Entergy, River Bend Station (ML060410310).

SCE requests approval of the proposed License Amendment by September 2008, as SCE will need to perform the Unit 3 ILRT during the fall of 2008 if this request is not approved.

SCE requests this amendment be effective immediately upon issuance, to be implemented within 60 days.

If you have any questions or require additional information, please contact Ms. Linda T. Conklin at (949) 368-9443.

Sincerely,

Enclosures:

1) PCN 582, Notarized Amendment Application Affidavits

2) PCN 582 Licensee Evaluation

3) San Onofre Nuclear Generating Station Probabilistic Risk Assessment, Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach, Unit 3, Rev. 0, July 2007.

cc:

E. E. Collins, Jr., Regional Administrator, NRC Region IV C. C. Osterholtz, NRC Senior Resident Inspector, San Onofre Units 2 and 3 N. Kalyanam, NRC Project Manager, San Onofre Units 2 and 3 S. Y. Hsu, Department of Health Services, Radiologic Health Branch

Enclosure I Proposed Change Number (PCN) 582 Notarized Amendment Application Affidavit

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Application of SOUTHERN CALIFORNIA

)

EDISON COMPANY, ET AL. for a Class 103)

License to Acquire, Possess, and Use

)

a Utilization Facility as Part of

)

Unit No. 3 of the San Onofre Nuclear

)

Generating Station

)

Docket No. 50-362 Amendment Application No. 236 SOUTHERN CALIFORNIA EDISON COMPANY, ET AL. pursuant to 10 CFR 50.90, hereby submit Amendment Application No. 582. This amendment application consists of proposed change No. NPF-15-582 to Facility Operating License NPF-15. Proposed change No.

NPF-15-582 is a request to revise Technical Specification (TS) 5.5.2.15, "Containment Leakage Rate Testing Program," to allow a one-time extension of the 15-year period of the performance-based leakage rate testing program for Type A tests as prescribed in TS 5.5.2.15. The 15-year interval between integrated leakage rate tests is to be extended to 16 years from the previous integrated leakage rate test.

State of California County of San Diego Brian Katz, Vice Pre ýnt Subscribed and sworn to (eF-eff~imedý before me on this 2,-+/-Jk day of

,e2007 by bV 1cc vx personally known to me orprovomto be the person who appeared before me.

Notary Pubic' L

AA DAWN A.

FAM J*__*Commimoan 0 163105 Proposed Change Number (PCN) 582 Licensee Evaluation

PCN-582 LICENSEE'S EVALUATION

1.

DESCRIPTION

2.

PROPOSED CHANGE

3.

BACKGROUND

4.

TECHNICAL ANALYSIS 4.1 Traditional Engineering Considerations 4.2 Evaluation of Risk Impact

5.

REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements/criteria

6.

ENVIRONMENTAL CONSIDERATION

7.

REFERENCES

8.

ATTACHMENTS A.

Existing Technical Specification Pages, Unit 3 B.

Proposed Technical Specification Pages, Redline and Strikeout, Unit 3 C.

Proposed Technical Specification Pages, Unit 3 Page 1 of 18

PCN-582 LICENSEE'S EVALUATION

1.0 DESCRIPTION

Proposed Change Number (PCN) 582 proposes to revise Technical Specifications (TS) 5.5.2.15, "Containment Leakage Rate Testing Program" for San Onofre Nuclear Generating Station (SONGS) Unit 3. PCN 582 proposes a one-time extension of the current interval between the Type A tests from 15 years to 16 years. The change reflects a one-time deferral of the next Type A Containment Integrated Leak Rate Test (ILRT) from September 9, 2010 to prior to startup from the Unit 3 Cycle 16 refueling outage, which is scheduled to commence in the fall of 2010 and to end in the first quarter of 2011. SONGS Unit 3 shall not operate past September 9, 2011 until the Type A Test is satisfactorily completed.

This PCN 582 includes a risk impact assessment report (Enclosure 3, Reference

23) that utilizes the methodology identified by the Nuclear Energy Institute (NEI)

(Reference 19).

2.0 PROPOSED CHANGE

Technical Specification Section 5.5.2.15 currently requires the following:

A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995 as modified by the following exception:

NEI 94 1995, Section 9.2.3: The first Type A Test performed after the September 10, 1995 Type A Test shall be performed no later than September 9, 2010.

Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," (Reference 1) endorses NEI 94-01, Revision 0, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," dated July 26, 1995 (Reference 2) and prepared by the NEI. NEI 94-01 provides methods acceptable to the NRC staff for complying with the provisions of Option B as described in Regulatory Guide 1.163. NEI 94-01 includes the criterion that Option B Type A testing be performed at a frequency of once per 10 years.

NRC to SCE letter dated August 24, 2005 (Reference 22) approved Amendment No. 189 to Facility Operating License No. NPF-15 for SONGS 3 to extend the 10 year frequency to 15 years to be performed no later than September 9, 2010.

This proposed change in the current licensing basis is a one-time extension of the test interval from 15 years to prior to startup from the Unit 3 Cycle 16 refueling outage, which is scheduled to commence in the fall of 2010 and to end in the first quarter of 2011. The approved one-time deferral of the integrated leakage rate test would be incorporated into Technical Specification 5.5.2.15 as follows:

Page 2 of 18

PCN-582 LICENSEE'S EVALUATION Unit 3

...as modified by the following exception:

NEI 94 1995, Section 9.2.3: The first Type A Test performed after the September 10, 1995 Type A Test shall be performed prior to startup from the Unit 3 Cycle16 refueling outage, which is scheduled to commence in the fall of 2010 and to end in the first quarter of 2011. SONGS Unit 3 shall not operate past September 9, 2011 until the Type A Test is satisfactorily completed."

In summary the proposed change will revise TS 5.5.2.15 entitled Containment Leakage Rate Testing Program to allow a one-time deferral of the Type A Containment Integrated Leak Rate Test (ILRT) from September 9, 2010 (current approved 15 year interval) to prior to startup from the Unit 3 Cycle 16 refueling outage. This proposed change is based on and has been evaluated using the risk informed guidance in Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" (Reference 3).

3.0 BACKGROUND

Containment leakage tests are performed to verify that Containment leakage is maintained below the acceptable limits stated in Technical Specification 5.5.2.15. The leakage tests ensure the public health and safety in the case of a design basis accident that would release radioactivity to the containment.

The leakage testing program consists of the following types of periodic tests:

(1)

Type A Test - measures the overall integrity of the containment

system, (2)

Type B Test - measures leakage rates across pressure retaining or leakage limiting boundaries other than valves, and (3)

Type C Test - measures containment isolation valve leakage rates.

This request does not modify the existing Appendix J, Type B and Type C testing programs nor does it change the Appendix J, Type A, Type B, or Type C Test methods. The change is a one-time exception to the Type A Test frequency.

This request represents a cost beneficial licensing change. Without this change, SCE would be required to perform two leak rate tests; one during the Unit 3 Cycle 15 refueling outage, and one following the steam generator replacement during the Cycle 16 refueling outage. The integrated leak rate test imposes significant expense on the station while the safety benefit of performing it at 15 years, versus 16 years, is minimal. Cost savings have been conservatively estimated at about $400,000 for actually performing the test and about $38,900 per hour for each hour of critical path outage time eliminating power replacement cost (the number of critical path hours is variable).

Page 3 of 18

PCN-582 LICENSEE'S EVALUATION

4.0 TECHNICAL ANALYSIS

The proposed changes have been evaluated to determine that current regulations and applicable requirements continue to be met, that adequate defense-in-depth and sufficient safety margins are maintained, and that any increases in core damage frequency (CDF) and large early release frequency (LERF) are small and consistent with the acceptance criteria in Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis,"

July 1998, (Reference 3).

4.1 Traditional Engineering Considerations In San Onofre Nuclear Generating Station (SONGS) License Amendment No.

135 (Unit 3) (Reference 7), Southern California Edison (SCE) committed to testing as required by 10 CFR 50, Appendix J, Option B, and in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September, 1995.

The adoption of the Option B performance-based containment leakage rate testing program did not alter the basic method by which Appendix J leakage rate testing is performed, but it did alter the frequency of measuring primary containment leakage in Type A, B, and C tests. Frequency is based upon an evaluation which looks at the "as found" leakage history to determine the frequency for leakage testing which provides assurance that leakage limits will be maintained.

The allowed frequency for testing was based upon a generic evaluation documented in NUREG-1493 (Reference 8). NUREG-1493 made the following observations with regard to decreasing the test frequency:

"Reducing the Type A (ILRT) testing frequency to one per twenty years was found to lead to an imperceptible increase in risk. The estimated increase in risk is small because ILRTs identify only a few potential leakage paths that cannot be identified by Type B and C testing, and the leaks that have been found by Type A tests have been only marginally above the existing requirements. Given the insensitivity of risk to containment leakage rate, and the small fraction of leakage detected solely by Type A testing, increasing the interval between ILRT testing had minimal impact on public risk."

The surveillance frequency for Type A testing in NEI 94-01 is once per 10 years based on an acceptable performance history (i.e., two consecutive periodic Type A tests at least 24 months apart where the calculated performance leakage rate was less than 1.0 La) and consideration of the performance factors in NEI 94-01, Section 11.3. Based on the last two consecutive ILRT Type A tests (Unit 3 -

September 10, 1995 and March 9, 1992), would be once every 10 years, however, a one-time extension to 15 years was granted by NRC to SCE letter dated August 24, 2005 (Reference 22).

A Type A test can detect containment leakage due to a loss of structural capability. All other sources of containment leakage detected in a Type A test analysis can be detected by the Type B and C tests.

Page 4 of 18

PCN-582 LICENSEE'S EVALUATION 4.1.1 Inspections 4.1.1.1 IWE/ IWL Inservice Inspection (ISI) Activities As required by 10 CFR 50.55a(b)(2)(vi), Inservice Inspection (ISI) of the SONGS Unit-3 containment building is conducted in accordance with the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Section Xl, 1992 Edition with the 1992 Addenda, as modified and supplemented by 10 CFR 50.55a(b)(2)(viii) and 10 CFR 50.55a(b)(2)(ix). The initial 120-month inspection interval for the Containment ISI began September 9, 1998, and will end on September 8, 2008. Successive 120-month intervals will comply with 10CFR 50.55a(g)(4)(ii). ASME Code Subsection IWE provides the requirements for the inservice inspection of Class MC and Metallic Liners of Class CC Components; Subsection IWL provides requirements for Class CC Concrete Components of Light Water Cooled Plants. SCE requested and received approval of the following relief requests.

1)

Relief Request No. RR-E-2-03, Seals and Gaskets of Class MC pressure retaining components, Examination Category E-D, Item numbers E5.10 and E5.20 of IWE-2500, " Examination and Pressure Test Requirements", Table IWE-2500-1.

2)

Relief Request No. RR-E-2-04, All Class MC, Subarticle IWE-2200(g), preservice examination requirements of reapplied or coated containment.

3)

Relief Request No. RR-E-2-05, All Class MC, Subarticle IWE-2500(b) visual examinations per Table IWE-2500-1 of painted or coated containment components prior to removal of paint or coatings.

4)

Relief Request No. RR-E-2-06, All Class MC, paragraphs IWE-2420(b) and IWE-2420(c) successive examination requirements for components found acceptable for continued service.

5)

Relief Request No. RR-E-2-07, Class MC pressure retaining bolting, Table IWE-2500-1, Examination Category E-G, Pressure Retaining Bolting, Item 8.20

6)

Relief Request No. RR-E-2-08, All components subject to the requirements for ISI of Class CC Concrete Components, Examination Category L-A, Concrete, Item L.1.11 as applicable to IWL-2310, Visual Examination and Personnel Qualification and IWA-2210, Visual Examinations.

In accordance with IWE-1240, Surface Areas Requiring Augmented Examination, SONGS Unit-3 has identified two locations on the steel liner exposed to substantial traffic due to scaffolding material during refueling outage.

Liner plate thicknesses at these locations were ultrasonically examined during refueling outage (RFO) -11 in January 2001, and RFO-1 3 in September 2004, for Unit-3. Measured thicknesses were greater than design required thickness in all the locations and no other degradation of liner plate noted. Liner wall thickness at these locations will be ultrasonically examined in future refueling outages as specified in Table IWE-2500-1, Category E-C, Item No. E4.12. Next scheduled examinations are for Unit-3 RFO-15 (October 2008). Refueling Page 5 of 18

PCN-582 LICENSEE'S EVALUATION outage dates are tentative and may be subject to change due to plant operating conditions.

As stated above, SCE requested and received approval for relief request RR-E-2-03, Seals and Gaskets of Class MC pressure retaining components, Examination Category E-D, Item numbers E5.10 and E5.20 of Table IWE-2500-1. Relief request RR-E-2-03 allowed the leak-tightness of the seal and gaskets to be tested in accordance with 10 CFR 50, Appendix J in lieu of Code required visual examination VT-3.

There is no separately scheduled ISI on any seal or gasket that is Type B tested per Option B of Appendix J of 10 CFR 50. Type B testing is performed on electrical penetrations, fuel transfer blind flange and fuel transfer bellows, equipment hatch, and the airlocks. Though the frequency can be as long as 10 years, the electrical penetrations are typically tested at a 60 month interval; the fuel transfer flange, fuel transfer bellow, and equipment hatch every refueling outage (about every 20 months), and the airlocks every 30 months. Should the penetrations be repaired or adjusted or opened, post maintenance testing is the Type B test and /or door seal test to assure proper operation. For example, the equipment hatch used for access to containment is Type B tested prior to being opened during outages. After the hatch is closed, the Type B test is conducted to assure that leakage is less than the administrative limit.

NRC Information Notice 92-20, "Inadequate Local Leak Rate Testing" discussed the inadequate local leak rate testing of two-ply stainless steel bellows. SCE has evaluated this notice and concluded that it is not applicable to SONGS (Reference 9). The SCE evaluation (Reference 9) was submitted to the NRC by letter dated December 2, 2004, (Reference 18).

SCE will perform the following examinations as specified in the ISI program.

i)

General visual examination of containment surfaces per Category E-A, Item No. E1.11, every period (once in 40 months) of the 10-year ISI interval as required per 10 CFR 50.55a(b)(2)(ix)(E).

Next scheduled examinations are for Unit-3 RFO-1 5 (October 2008).

ii)

Visual VT-3 examination of the containment surfaces per Examination Category E-A, Item No. E1.12 at the end of 10 year ISI interval. Next scheduled examinations are for Unit-3 RFO-1 5 (October 2008) iii)

Visual VT-3 examination of the containment surface vent system per Examination Category E-A, Item No. E1.20 at the end of 10 year ISI interval. Next scheduled examinations are for Unit-3 RFO-1 5 (October 2008) iv)

Ultrasonic examination to verify minimum wall thickness of containment surfaces requiring augmented examination per Category E-C, Item No. E4.12, every period (once in 40 months) of the 10 year ISI interval. Next scheduled examinations are for Unit-3 RFO-1 5 (October 2008).

v)

Visual VT-3 examination on Moisture Barriers per Examination Category E-D, Item No. E5.30 of Table IWE-2500-1. Next scheduled examinations are for Unit-3 RFO-15 (October 2008).

Page 6 of 18

PCN-582 LICENSEE'S EVALUATION iv)

Visual VT-1 examination on Bolted Connections Examination per Category E-G, Item No. E8.10 of Table IWE-2500-1. Next scheduled examinations are for Unit-3 RFO-15 (October 2008).

SCE completed first interval, first and second period inspections for Unit-3 in January 2001 and September of 2004 with acceptable results. The ASME Code Section XI IWE and IWL containment inspections provide a high degree of assurance that any degradation of the containment structure is identified and corrected before a containment leakage path is introduced.

In summary, the general visual examination of containment surfaces every period, visual VT-3 examination of the containment surfaces at the end of the ISI interval, visual VT-3 examination of the containment surface vent system at the end of the ISI interval, ultrasonic examination to verify minimum wall thickness of containment surfaces requiring augmented examination, visual VT-3 on Moisture Barrier, visual VT-1 examination on Bolted Connections, and the Appendix J, Option B, Type B test provide reasonable assurance the integrity of the containment pressure boundary will be maintained during the extended Type A test frequency.

4.1.1.2 Maintenance Rule Monitoring to support ILRT The condition of the Containment Building structure is monitored under the maintenance rule program to ensure that maintenance is effective and the structure is capable of performing its intended functions. SONGS procedure S0123-XXIV-20.2, "Maintenance Rule For Structures," was implemented to meet the requirements of 10CFR50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," for the containment buildings.

The first Maintenance Rule inspections were completed in RFO-9 in 05-06/1997, and documented in Calculation C-501-02.01 for Unit 3 and established the baseline for future Maintenance Rule inspections. Two inspections have been performed since the baseline: RFO-10 (03-05/1999, Calculation C-501-02.02) and RFO-12 (01-03/2003, Calculation C-501-02.03) for Unit 3. Evaluations of inspection results have concluded that the containment structure for Unit 3 continues to meet its design bases. Subsequent inspections have found no adverse trending in the containment structure. Minor degraded conditions have been identified but the degraded conditions do not affect the structural integrity of the containment structures. The degraded conditions are described and evaluated in Section 8.4 of the SONGS Maintenance Rule calculations. The conditions of the containment liner and coating are good. Equipment supports, HVAC ducts and electrical raceways are also in good condition.

The Maintenance Rule inspection is scheduled for Unit 3 RFO-16 (October 2010). However, the containment liner coating inspection is conducted every refueling outage as part of the SONGS response (Reference 15) to Generic Letter 98-04 (Reference 16).

Page 7 of 18

PCN-582 LICENSEE'S EVALUATION 4.1.1.3 Containment Visual Inspection As required by NEI 94-01 (Reference 2) and R.G. 1.163, part C.3 (Reference 1),

visual examinations are performed of accessible interior and exterior surfaces of the containment system for structural deterioration. SCE to NRC letter dated May 27, 2005 (Reference 24) committed SONGS to performing these visual examinations every other refueling outage and prior to the 15 year Type A test upon NRC approval of the 15 year extension on August 24, 2005 (reference 22).

Such visual examinations were conducted in RFO-12 (Jan - Feb 2003), RFO-13 (Sep - Dec 2004), and RFO-14 (Oct - Dec 2006) under SONGS procedure S03-V-3.12 with no adverse findings. These examinations will be performed prior to the next Type A test in RFO -16.

4.1.2 Previous Integrated Leakage Rate Test Results Previous Type A tests confirmed that the SONGS reactor containment structure has leakage well under acceptance limits and represents minimal risk to increased leakage.

The commercial operation dates for Unit 2 and Unit 3 are August 18, 1983 and April 1, 1984 respectively. Since then, SONGS has performed three operational Type A tests for Unit 2 and Unit 3. The results are well within the leakage limit and are presented in Appendix D of Enclosure 3.

The testing history and structural capability of the containment have established that SONGS has had acceptable containment leakage rates, that the structural integrity of containment is assured, and that there is negligible impact in extending the Type A test interval on a one-time basis.

4.1.3 Plant Operational Performance During power operation, instrument air leaks from air-operated valves inside containment pressurize the containment building. Containment pressure and conditions approaching the limits allowed by the Technical Specifications are monitored. Because it is routinely necessary to reduce the increase in the building internal pressure by periodic operation of the containment pressure relief, a large pre-existing leak would make it unnecessary to periodically operate the containment pressure relief (referred to as venting containment at SONGS.)

This change in operating pattern would be noticed by plant operators.

Although not as significant as pressure resulting from a Design Basis Accident, the fact that the containment can be pressurized by leakage from air-operated valves provides a degree of assurance of containment structural integrity (i.e., no large leak paths in the containment structure). This feature complements the various inspection requirements of the containment structure.

Page 8 of 18

PCN-582 LICENSEE'S EVALUATION 4.2 Evaluation of Risk Impact 4.2.1 PRA Approach 10 CFR 50, Appendix J allows individual plants to extend Type A surveillance testing requirements and to provide for performance-based leak testing. The San Onofre Nuclear Generating Station Probabilistic Risk Assessment, Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach, Revision 0, April 2005 (Reference 17) documents a risk-based evaluation of the proposed change of the integrated leak rate test (ILRT) interval for the San Onofre Nuclear Generating Station (SONGS). The proposed change would impact testing associated with the current surveillance tests for Type A leakage testing. No change to Type B or Type C testing is proposed at this time.

This evaluation for SONGS utilizes the guidelines set forth in NEI 94-01 (Reference 2), the methodology used in EPRI TR-104285 (Reference 13),

NUREG-1493 (Reference 8). The NEI guidance also considers the submittals generated by other utilities. The assessment contained in this submittal (Enclosure 3) utilizes the method set forth and utilizes metrics presented in the NEI interim guidance (Reference 19). The regulatory guidance on the use of probabilistic risk assessment (PRA) findings in support of a licensee request to a plant's licensing basis, RG 1.174 (Reference 3) is also utilized.

Using updated SONGS PRA data and post-accident dose results, Enclosure 3 evaluates the risk associated with various ILRT intervals as follows:

3 years - Interval based on the original requirements of 3 tests per 10 years.

10 years - interval based on 10 CFR 50 App J Option B requirements.

15 years - This is the current test interval required for SONGS.

16 years - Proposed extended test interval is prior to start up from the RFO-16 outage.

Note: The 10 year interval analysis is provided as additional information.

Page 9 of 18

PCN-582 LICENSEE'S EVALUATION 4.2.1 Summary of Risk Results/Conclusions The specific results from the main report (Enclosure 3) are summarized in Table 1 below. The Type A contribution to LERF is defined as the contribution from Class 3b.

Table 1 Summary of Risk Impact on Extending Type A ILRT Test Frequency Risk Impact for 3-years Risk Impact for 10-years Risk Impact for 15-years Risk Impact for 16-years (baseline)

(current requirement)

Total integrated risk (person-rem/yr) 9.685 9.689 9.692 9.693 Type A testing risk (person-rem/yr) combination of 3a and 3b 1.89E-3 6.31E-3 9.46E-3 1.01E-2

% total risk (Type A / total) 0.020%

0.065%

0.098%

0.104%

Type A LERF (Class 3b) (per year) 1.70E-8 5.65E-8 8.48E-8 9.05E-8 Conditional Containment Failure Probability (CCFP) 1.87E-1 1.94E-1 1.98E-I 1.99E-I Changes due to extension from 15 years (current)

A Risk from current (Person-rem/yr) 5.79E-4

% Increase from current (A Risk / Total Risk) 0.006%

A LERF from current (per year) 5.65E-9 A CCFP from current 0.09%

Page 10 of 18

PCN-582 LICENSEE'S EVALUATION Table 1 (continued)

Summary of Risk Impact on Extending Type A ILRT Test Frequency Changes due to extension from 10 years A Risk from 10-year (Person-rern/yr) 3.48E-3

% Increase from 10-year (A Risk / Total Risk) 0.036%

A LERF from I 0-year (per year) 3.40E-8 A CCFP from 10-year 0.55%

Changes due to extension from 3 years (baseline)

A Risk from baseline (Person-rem/yr) 7.53 E-3

% Increase from baseline (A Risk / Total Risk) 0.078%

A LERF from baseline (per year) 7.35E-8 A CCFP from baseline 1.18%

Page 11 of 18

PCN-582 LICENSEE'S EVALUATION The results are discussed below:

" The person-rem/year increase in risk contribution from extending the ILRT test frequency from the current once-per-fifteen-year interval to once-per-sixteen years is 0.000579 person-rem/year.

The risk increase in LERF from extending the ILRT test frequency from the current once-per-1 5-year interval to once-per-16 years is 5.65E-9/yr.

" The change in conditional containment failure probability (CCFP) from the current once-per-15-year interval to once-per-16 years is 0.09%.

" The change in Type A test frequency from once-per-fifteen-years to once-per-sixteen-years increases the risk impact on the total integrated plant risk by only 0.006%. Also, the change in Type A test frequency from the original three-per-ten-years to once-per-sixteen-years increases the risk only 0.078%. Therefore, the risk impact when compared to other severe accident risks is negligible.

Reg. Guide 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Reg. Guide 1.174 defines very small changes in risk as resulting in increases of core damage frequency (CDF) below 10-6/yr and increases in LERF below 10-7/yr. Since the ILRT does not impact CDF, the relevant criterion is LERF.

The increase in LERF resulting from a change in the Type A ILRT test interval from a once-per-fifteen-years to a once per-sixteen-years is 5.65E-9/yr. Guidance in Reg.

Guide 1.174 defines very small changes in LERF as below 10-7/yr. Increasing the ILRT interval from 15 to 16 years is therefore considered non-risk significant and the results support this determination. In addition, the change in LERF resulting from a change in the Type A ILRT test interval from a three-per-ten-years to a once per-sixteen-years is 7.35E-8/yr, is also below the guidance.

R.G. 1.174 also encourages the use of risk analysis techniques to help ensure and show that the proposed change is consistent with the defense-in-depth philosophy.

Consistency with defense-in-depth philosophy is maintained by demonstrating that the balance is preserved among prevention of core damage, prevention of containment failure, and consequence mitigation. The change in conditional containment failure probability was estimated to be 0.09% for the proposed change and 1.18% for the cumulative change of going from a test interval of 3 in 10 years to 1 in 16 years. These changes are small and demonstrate that the defense-in-depth philosophy is maintained.

In reviewing these results the SONGS Unit 3 analysis demonstrates that the change in plant risk is small as a result of this proposed extension of ILRT testing. The change in LERF defined in the analysis for the sensitivity study also indicates that the change in LERF is within the acceptance criterion.

5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration Southern California Edison (SCE) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 1 OCFR50.92, "Issuance of amendment," as discussed below:

Page 12 of 18

PCN-582 LICENSEE'S EVALUATION Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed revision to Technical Specifications adds a one time extension to the current interval for Type A testing (10CFR50, Appendix J, Option B, Integrated Leak Rate Testing). The current test interval of 15 years, based on past performance, would be extended on a one time basis to 16 years from the last Type A test. The proposed extension to Type A testing does not involve a significant increase in the probability or consequences of an accident since research documented in NUREG-1493, "Performance-Based Containment System Leakage Testing Requirements," September 1995, has found that, generically, very few potential containment leakage paths are not identified by Type B and C tests. The NUREG concluded that reducing the Type A testing frequency to once per twenty years was found to lead to an imperceptible increase in risk. A high degree of assurance is provided through testing and inspection that the containment will not degrade in a manner detectable only by Type A testing. The most recent Type A test at Unit 3 shows leakage to be below acceptance criteria, indicating a leak tight containment. Inspections required by the American Society of Mechanical Engineers (ASME) Code Section Xl (Subsections IWE and IWL) and maintenance rule monitoring (10CFR50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants) are performed in order to identify indications of containment degradation that could affect leak tightness. Type B and C testing required by Technical Specifications will identify any containment opening such as valves that would otherwise be detected by the Type A tests. These factors show that a Type A test extension will not represent a significant increase in the consequences of an accident.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed revision to Technical Specifications adds a one time extension to the current interval for Type A testing (10CFR50, Appendix J, Option B, Integrated Leak Rate Testing). The current test interval of 16 years, based on past performance, would be extended on a one time basis to 16 years from the last Type A test. The proposed extension to Type A testing cannot create the possibility of a new or different type of accident since there are no physical changes being made to the plant and there are no changes to the operation of the plant that could introduce a new failure mode creating an accident or affecting the mitigation of an accident. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

Page 13 of 18

PCN-582 LICENSEE'S EVALUATION

3.

Do the proposed changes involve a significant reduction in a margin of safety?

Response: No The proposed revision to Technical Specifications adds a one time extension to the current interval for Type A testing (10CFR50, Appendix J, Option B, Integrated Leak Rate Testing). The current test interval of 15 years, based on past performance, would be extended on a one time basis to 16 years from the last Type A test. The proposed extension to Type A testing will not significantly reduce the margin of safety. The NUREG 1493, "Performance-Based Containment System Leakage Testing Requirements," September 1995, generic study of the effects of extending containment leakage testing found that a 20 year extension in Type A leakage testing resulted in an imperceptible increase in risk to the public. NUREG 1493 found that, generically, the design containment leakage rate contributes about 0.1 percent to the individual risk and that the decrease in Type A testing frequency would have a minimal affect on this risk since 95% of the potential leakage paths are detected by Type C testing. Regular inspections required by the American Society of Mechanical Engineers (ASME) Code Section Xl (Subsections IWE and IWL) and maintenance rule monitoring (10CFR50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants) will further reduce the risk of a containment leakage path going undetected.

Therefore the proposed change does not involve a significant reduction in a margin of safety.

Based on the above evaluations, SCE concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10CFR50.92(c) and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements/Criteria 10 CFR 50.54(o) - "Primary reactor containments for water cooled power reactors, other than facilities for which the certifications required under 50.82(a)(1) have been submitted, shall be subject to the requirements set forth in appendix J to this part."

10 CFR 50, Appendix A, General Design Criteria (GDC) 52 - "Capability for containment leakage rate testing. The reactor containment and other equipment which may be subjected to containment test conditions shall be designed so that periodic integrated leakage rate testing can be conducted at containment design pressure."

GDC 53 - "Provisions for containment testing and inspection. The reactor containment shall be designed to permit (1) appropriate periodic inspection of all important areas, such as penetrations, (2) an appropriate surveillance program, and (3) periodic testing at containment design pressure of the leak tightness of penetrations which have resilient seals and expansion bellows."

Page 14 of 18

PCN-582 LICENSEE'S EVALUATION GDC 54 - "Piping systems penetrating containment. Piping systems penetrating primary reactor containment shall be provided with leak detection, isolation, and containment capabilities having redundancy, reliability, and performance capabilities which reflect the importance to safety of isolating these piping systems. Such piping systems shall be designed with a capability to test periodically the operability of the isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits."

10 CFR 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," Option B, "Performance-Based Requirements."

Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," September 1995.

NUREG-1493, "Performance-Based Containment Leak-Test Program,"

September 1995.

Analysis The Containment Building, Containment penetrations, and Containment isolation barriers are designed to permit periodic leakage rate testing as required by General Design Criteria (GDC) 52, 53, and 54 of Title 10 Code of Federal Regulations, Part 50, Appendix A.

10 CFR 50 Appendix J, was revised, effective October 26, 1995, to allow licensees to choose containment leakage testing under Option A "Prescriptive Requirements" or Option B "Performance-Based Requirements." In License Amendment No. 135 (Unit 3) (Reference 7), SCE committed to testing as required by 10 CFR 50, Appendix J, Option B, and in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program, dated September, 1995." Regulatory Guide 1.163 specifies a method acceptable to the NRC for complying with Option B by approving the use of Nuclear Energy Institute (NEI) 94-01 and ANSI/ANS-56.8-1994 (Reference 14) subject to several regulatory positions in the guide.

Exceptions to the requirements of RG 1.163, are allowed by 10 CFR 50, Appendix J, Option B,Section V.B, "Implementation," which states, "The Regulatory Guide or other implementing document used by a licensee, or applicant for an operating license, to develop a performance based leakage-testing program must be included, by general reference, in the plant technical specifications. The submittal for technical specification revisions must contain justification, including supporting analyses, if the licensee chooses to deviate from methods approved by the Commission and endorsed in a regulatory guide."

Therefore, this application does not require an exemption to Option B.

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 amendments will not be inimical to the common defense and security or to the health and safety of the public.

Page 15 of 18

PCN-582 LICENSEE'S EVALUATION

6.0 ENVIRONMENTAL CONSIDERATION

Southern California Edison (SCE) has determined that the proposed amendment would change requirements with respect to the installation or use of a facility component located within the restricted area, as defined in 1 OCFR20, or would change an inspection or surveillance requirement. SCE has evaluated the proposed changes and has determined that the changes do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amount of effluent that may be released offsite, or (iii) a significant increase in the individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 1 0CFR51.22 (c)(9). Therefore, pursuant to 1 0CFR51.22 (b), an environmental assessment of the proposed change is not required.

7.0 REFERENCES

1.

NRC Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Test Program," September 1995.

2.

NEI 94-01, "Nuclear Energy Institute Industry Guideline For Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 0, July 26, 1995.

3.

NRC Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," July 1998.

4.

Not used.

5.

Not used.

6.

Not used.

7.

San Onofre Nuclear Generating Station, Issuance of License Amendments 144/135 re: Use of new Containment Leakage Rate Testing Program as required by 10 CFR 50, Appendix J, Option B for SONGS Units 2 and 3 (TAC NOS. MA1778 and MA1779), dated November 6, 1998.

8.

NUREG-1493, "Performance-Based Containment Leak-Test Program,"

Final Report, September 1995.

9.

SONGS Independent Safety Engineering Group Operating Experience Evaluation, Subject, " NRC Information Notice 92-20, Inadequate Local Leak Rate Testing," dated June 22, 1992.

10.

Indian Point 3 Nuclear Power Plant, "Supplemental Information Regarding Proposed Change to Section 6.14 of the Administrative Section of the Technical Specifications", Entergy, IPN-01-007, January 18, 2001.

11.

Indian Point Nuclear Generating Unit No. 3 - Issuance of Amendment Re:

Frequency of Performance-Based Leakage Rate Testing (TAC NO.

MB0178), United States Nuclear Regulatory Commission, April 17, 2001.

Page 16 of 18

PCN-582 LICENSEE'S EVALUATION

12.

Evaluation of Risk Significance of ILRT Extension, Revision 2, Florida Power Corporation, F-01 -0001, June 2001, attached to a letter from Dale E. Young (Crystal River) to the Document Control Desk (NRC) dated June 20, 2001;

Subject:

Crystal River - Unit 3 - License Amendment Request # 267, Revision 2, Supplemental Risk-Informed Information in Support of License Amendment Request # 267.

13.

Electric Power Research Institute, TR-104285, Gisclon, J. M., et al, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals,"

August 1994.

14.

American National Standard ANSI/ANS-56.8-1994, "Containment System Leakage Testing Requirements."

15.

SCE Letter dated November 12, 1998 from A. E. Scherer to the NRC; Subject," NRC Generic Letter 98-04: Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System After a Loss-of-Coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Material in Containment. San Onofre Nuclear Generating Stations Units 2 and 3."

16.

NRC Generic Letter 98-04 dated July 14, 1998 - Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System After a Loss-of-Coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Material in Containment.

17.

San Onofre Nuclear Generating Station Probabilistic Risk Assessment, Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach, Revision 0, April 2005

18.

Letter from A. E. Scherer (SCE) to the U.S. Nuclear Regulatory Commission dated December 2, 2004;

Subject:

Docket Nos. 50-361 and 50-362, Supporting Information Regarding Amendment Application Nos.

224 and 208, Proposed Change Number (PCN) 554, Technical Specification (TS) 5.5.2.15, Containment Leakage Rate Testing Program, San Onofre Nuclear Generating Stations Units 2 and 3.

19.

Haugh, J., et al, "Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals," Revision 4, Nuclear Energy Institute (NEI), November 2001.

20.

Letter from D. E. Nunn (SCE) to the U. S. Nuclear Regulatory Commission dated June 30, 2004;

Subject:

Docket Nos. 50-361 and 50-362 Amendment Application Nos. 224 and 208, Proposed Change Number (PCN) 554, Technical Specification (TS) 5.5.2.15, Containment Leakage Rate Testing Program, San Onofre Nuclear Generating Stations Units 2 and 3.

21.

Miller, J., "San Onofre Nuclear Generating Station Probabilistic Risk Assessment Evaluation of Risk Significance of ILRT Extension," Revision 0, Ricky Summitt Consulting (RSC), Inc., RSC04-02, March 2004.

Page 17 of 18

PCN-582 LICENSEE'S EVALUATION

22.

Letter from J. N. Donohew (NRC) to H. B. Ray (SCE) dated August 24, 2005;

Subject:

San Onofre Nuclear Generating Stations Units 2 and 3 -

Issuance of Amendments on Containment Leakage Rate Testing Program (TAC NOS. MC3797 and MC 3798).

23.

San Onofre Nuclear Generating Station Probabilistic Risk Assessment, Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach, Revision 0, July 2007, Unit 3.

24.

Letter from B. Katz (SCE) to the U. S. Nuclear Regulatory Commission dated May 27, 2005;

Subject:

Docket Nos. 50-361 and 50-362 Supplemental One To Amendment Application Nos. 224 and 208, Proposed Change Number (PCN) 554, Technical Specification (TS) 5.5.2.15, Containment Leakage Rate Testing Program, San Onofre Nuclear Generating Stations Units 2 and 3 (NRC Accession Number ML051530089).

8.0 PRECEDENT 1.0 St. Lucie Unit 2 Submittal Dates: 3/31/2005, (ML050950235), 11/9/2005 (ML053190184)

NRC Approval Date: 12/23/2005 (ML053190343) 2.0 Entergy River Bend Station, Unit 1 Submittal Dates: 3/8/2005 (ML050740346), 1/17/2006 (ML060230049)

NRC Approval Date: 2/9/2006 (ML060410310) 3.0 Florida Power & Light Company Seabrook Station, Unit 1 Submittal Date 9/29/2005 (ML052770187)

NRC Approval Date 3/24/2006 (ML060520032)

Page 18 of 18

PCN 582 Attachment A (Existing Technical Specification Pages, Unit 3)

Procedures,

Programs, and Manuals 5.5 5.5 Procedures,

Programs, and Manuals (continued) 5.5.2.15 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.

This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program",

dated September 1995 as modified by the following exception:

NEI 94 1995, Section 9.2.3: The first Type A Test performed after the September 10, 1995 Type A Test shall be performed no later than September 9, 2010.

The calculated peak containment internal pressure related to the design basis loss-of-coolant accident, Pa, is 45.9 psig (Pa will conservatively be assumed to be equal to the calculated peak containment internal pressure for the design basis Main Steam Line Break (56.5 psig) for the purpose of containment testing in accordance with this Technical Specification).

The maximum allowable containment leakage rate, La, at P,, shall be 0.10% of containment air weight per day.

Leakage rate acceptance criteria are:

a.

The Containment overall leakage rate acceptance criterion is

< 1.0 La.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are

0.60 La for the Type B and Type C tests and

0.75 La for the Type A tests;

b.

Air lock testing acceptance criteria are:

1)

Overall air lock leakage rate is

0.05 La when tested at Pa_

2)

For each door, the leakage rate is

0.01 La when pressurized to 2 9.0 psig.

(continued)

SAN ONOFRE--UNIT 3 5.0-20a Amendment No. 189 1

PCN 582 Attachment B (Proposed Technical Specification Pages, Redline and Strikeout, Unit 3)

Procedures, Programs, and Manuals 5.5 5.5 Procedures,

Programs, and Manuals (continued) 5.5.2.15 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.

This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program",

dated September 1995 as modified by the following exception:

NEI 94 1995, Section 9.2.3: The first Type A Test performed after the September 10, 1995 Type A Test shall be performed no later than September 9, 2010. prior to startup from the Unit 3 Cycle 16 refuelinq outaqe, which is scheduled to commence in the fall of 2010 and to end in the first quarter of 2011.

SONGS Unit 3 shall not operate past September 9, 2011 until the Type A Test is satisfactorily completed.

The calculated peak containment internal pressure related to the design basis loss-of-coolant accident, Pa, is 45.9 psig (Pa will conservatively be assumed to be equal to the calculated peak containment internal pressure for the design basis Main Steam Line Break (56.5 psig) for the purpose of containment testing in accordance with this Technical Specification).

The maximum allowable containment leakage rate, La, at Pa, shall be 0.10% of containment air weight per day.

Leakage rate acceptance criteria are:

a.

The Containment overall leakage rate acceptance criterion is 1.0 La-During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and Type C tests and

0.75 La for the Type A tests;

b.

Air lock testing acceptance criteria are:

1)

Overall air lock leakage rate is

0.05 La when tested at

> P_.

2)

For each door, the leakage rate is

0.01 La when pressurized to Ž 9.0 psig.

(continued)

SAN ONOFRE--UNIT 3 5.0-20a Amendment No. 4-84

PCN 582 Attachment C (Proposed Technical Specification Pages, Unit 3)

Procedures, Programs, and Manuals 5.5 5.5 Procedures,

Programs, and Manuals (continued) 5.5.2.15 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.

This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program",

dated September 1995 as modified by the following exception:

NEI 94 1995, Section 9.2.3: The first Type A Test performed after the September 10, 1995 Type A Test shall be performed prior to startup from the Unit 3 Cycle 16 refueling outage, which is scheduled to commence in the fall of 2010 and to end in the first quarter of 2011.

SONGS Unit 3 shall not operate past September 9, 2011 until the Type A Test is satisfactorily completed.

The calculated peak containment internal pressure related to the design basis loss-of-coolant accident, Pa, is 45.9 psig (Pa will conservatively be assumed to be equal to the calculated peak containment internal pressure for the design basis Main Steam Line Break (56.5 psig) for the purpose of containment testing in accordance with this Technical Specification).

The maximum allowable containment leakage rate, La, at P*,

shall be 0.10% of containment air weight per day.

Leakage rate acceptance criteria are:

a.

The Containment overall leakage rate acceptance criterion is 1.0 La.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and Type C tests and

0.75 La for the Type A tests;

b.

Air lock testing acceptance criteria are:

1)

Overall air lock leakage rate is

0.05 La when tested at > Pa-

2)

For each door, the leakage rate is < 0.01 La when pressurized to Ž 9.0 psig.

(continued)

SAN ONOFRE--UNIT 3

5. 0-20a Amendment No.

San Onofre Nuclear Generating Station Probabilistic Risk Assessment, Unit 3 Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Revision 0 July 2007

Southern California Edison RSC 07-05 San Onofre Nuclear Generating Station Unit 3 Probabilistic Safety Assessment Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Revision 0 July 2007 Principal Analyst R. SuImmitt RSC Ricky Sumrnmitt Consulting, Inc.

8351 E. Walker Springs Lane, Suite 401 Knoxville, TN 37923 USA Telephone +865-692.4012 Telefax +865.692-4013 R-;.3k and Reliability Engineering

Document Revision History Document RSC Principle AnalystfPrcject RSC Internal Technical and RSC Approval for Revision Manager (name/date)

Editorial Review Complete Client Release (initials/date)

(initials/date)

Original Issue R. Summitt/04-1 2 -07 J. Jansen/04-25-07 R. Surnmitt/07-25-07 I

2 3

4 5

6 7

8 9

10 RSC Ricky Sumwni~tt Consutint~g, Inc.

8351 E. Walker Springs Lane, Suite 401 Knoxville, TN 37923 USA Telephone +865.692.4012 Telefax +865.692.4013 Risk and Rebabililvy Engineenng

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table of Contents Section Page Main Report: Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach 1.0 P U R P O S E.................................................................................................................................

I 1.1 SUM M ARY OF THE AN ALYSIS.....................................................................................

1 1.2 SUM M ARY OF RESULTS/CONCLUSIONS...................................................................

2.0 DESIGN INPUTS.....................................................................................................................

6 3.0 ASSUM PTIONS.......................................................................................................................

9 4.0 CALCULATIONS..................................................................................................................

10 4.1 CALCULATIONAL STEPS..........................................................................................

10 4.2 SUPPORTING CALCULATIONS.....................................................................................

2

4.3 REFERENCES

35 Appendices A.0 EXTERNAL EVENTS SEN SITIVITY STUD Y...............................................................

I A.1 SUM M ARY OF THE ANALYSIS.................................................................................

I A.2 SUM M ARY OF RESULTS/CONCLUSIONS..............................................................

I A.3 DESIGN INPUTS................................................................................................................

5 A.4 CALCULATIONS.......................................................................................................

8 A.5 SUPPORTIN G CALCULATIONS................................................................................

8 A.6 REFEREN CES...................................................................................................................

30 B.0 POPULATION DOSE ESTIM ATES............................................................................

.......... I B.I ESTIMATION OF INTACT CONTAINMENT POPULATION DOSE.......................

I B.I.1 M ethodology.........................................................................................................

I B.I.2 Licensing Basis Inform ation.........................................................................................

I B.1.3 Dose Scaling Factor......................................................................................

RSC 07-05 1

Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach B.1.4 C alcu lation of P opu lation D ose.....................................................................................

B. 1.5 Sensitivity Assessment of Evacuation Assumption on Intact Containment Dose E stim a tio n................................................................................................................................

4 B.2 ACCIDENT SEQUENCE DOSE ESTIMATION...............................................................

5 B.3 R E F E R E N C E S...................................................................................................................

10 C.0 RESPONSE TO USNRC REQUEST FOR ADDITIONAL INFORMATION FOR DEGRADATION OF THE EMBEDDED SIDE OF THE STEEL DRYWELL STRUCTURE.... 1 C.1 A N A L Y SIS A PP R O A C H....................................................................................................

1 C.2 A N A L Y SIS R E SU L T S....................................................................................................

I C.3 R E F E R E N C E S.....................................................................................................................

5 D.0 EVALUATION OF RELEVANT SONGS ILRT EXPERIENCE......................................

I D.1 A N A L Y S IS R E SU L T S........................................................................................................

I D.2 R E F E R E N C E S.....................................................................................................................

2 RSC 07-05 ii Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach List of Tables Table Pagye Table I Summary of Risk Impact on Extending Type A ILRT Test Frequency............................

3 T able 2 SO N G S Plant D am age States..............................................................................................

6 Table 3 Release Category Radionuclide Fraction...........................................................................

9 Table 4 Containment Failure Classifications (from Reference 13)................................................... 12 Table 5 SONGS PRA Release Category Grouping to EPRI Classes (as described in Reference 13)13 T able 6 B aselin e R isk P rofile........................................................................................................

.. 17 Table 7 Risk Profile for Once in Ten Year Testing........................................................................ 19 Table 8 Risk Profile for Once in Fifteen Year Testing 2......

1 Table 9 Risk Profile for Once in Sixteen Year Testing 23 Table 10 impact on LERF due to Extended Type A Testing Intervals.........................................

25 T able 11 Source T erm O utcom es...................................................................................................

8.....

?

Table 12 Class 3b Contributions U sing A djusted CD F......................................................................

29 Table 13 Adjustment to Small LOCA Contribution to STC Refinement Outcome 4 (from Reference 1 9 )................................................................................................................................................

3 0 Table 14 Class 3b Contributions Using Adjusted CDF..................................................................

30 Table 15 Intact Containment Release Category Tirning 31 Table 16 Radionuclide Releases by Release Category.................................................................

32 Table 17 Adjustment to Address Source Term Magnitude and Timing

..... 34 Table 18 Class 3b Contributions Using Adjusted CDF..................................................................

34 Table 19 Impact on Conditional Containment Failure Probability due to Extended Type A Testing In te rv a ls.......................................................................................................................................

3 5 Appendices Table A. I Summary of Risk Impact on Extending Type A ILRT Test Frequency.............

Table A.2 SONGS Plant Damage States Including Fire and Seismic............................................

5 RSC 07-05 iii Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table A.3 Release Category Radionuclide Fraction........................................................................

7 Table A.4 Containment Failure Classifications (from Reference 5)................................................ 9 Table A.5 SONGS PRA Release Category Grouping to EPRI Classes (as described in Reference 5)10 T able A.6 B aseline R isk Profile...................................................................................................

.. 14 Table A.7 Risk Profile for Once in Ten Year Testing

.......... 16 Table A.8 Risk Profile for Once in Fifteen Year Testing 18 Table A.9 Risk Profile for Once in Sixteen Year Testing 0....

0 Table A. 10 Impact on LERF due to Extended Type A Testing Intervals....................................

23 Table A.11 Source Term O utcom es..............................................................................................

26 Table A. 12 Class 3b Contlibutions Using Adjusted CDF.............................................................

27 Table A. 13 Class 3b Contributions Using Adjusted CDF.............................................................

27 Table A. 14 LERF Contribution Adjustment for PCS-35 and NLO-4.........................................

28 Table A.1.5 Adjustment to Address Source Term Magnitude and Timing 28 Table A. 16 Class 3b (LERF) Contributions Using Adjusted CDF...............................................

29 Table A. 17 Impact on Conditional Containment Failure Probability due to Extended Type A T e stin g In te rv a ls..........................................................................................................................

3 0 Table B.1 Predicted D ose Rates taken from Reference 1.....................................................................

?

Table B.2 Calculation Parameters taken from Reference 1.............................................................

3 Table B.3 Calculation Parameters taken from Reference I.............................................................

3 Table B.4 Comparison of Class 1, 3a and 3b Person-Rem for Baseline and No Evacuation S e n sitiv ity S tu d y............................................................................................................................

5 Table B.5 Reported Person Rem Estimates for SulTy Source Term Groups (summarized from R e fe re n c e 5 )...................................................................................................................................

7 Table B.6 Assignment of SUITy Source Term Groups to EPRI Classes..........................................

8 Table B.7 Average Person-Remn for SuITy Source Term Groups...................................................

9 Table B.8 Average Person-Remn for EPRI Classes Based on Sury Source Term Groups............ 10 RSC 07-05.

iv Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table C.1 SONGS Liner Corrosion Risk Assessment Results Using CCNP Methodology............ I Table C.2 Changes Due to Extension from 15 Years (current) to 16 Years...................................

4 Table C.3 Changes Due to Extension from 3 Years (baseline) to 16 Years.....................................

4 Table D.I SO N G S ILRT Resultant Leak R ates................................................................................

I List of Figures Fi gure Page Figiure 1. Quantified Source Term Category Diagram for SONGS Case I (from Reference 19)..... 27 Figure 2. La2O3 and CeO2 Release Timing (derived from Reference 21).......................................

33 Appendices Figure A.1. Quantified Source Tern-Category Diagram for SONGS Case 2 (from Reference. 15).25 RSC 07-05 Printed: 07/25/2007

Evaluation of Risk SiEnificance of ILRT Extension Based on the NEI Approach Main Report:

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach RSC 07-05 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach 1.0 PURPOSE The purpose of this report is to provide an estimation of the change in risk associated with extending the Type A integrated leak rate test (ILRT) interval beyond the current 15 years' to 16 years for the San Onofre Nuclear Generating Station (SONGS) Unit 3. Specifically, this report utilizes the methodology identified by the Nuclear Energy Institute (NEI).

Similar assessments of the proposed change for an extension to 15 years are documented in References 3 and 4 and serve as the bases for this document. The evaluation found in Reference 3 is consistent with similar assessments performed for Comanche Peak, Indian Point 3 (IP3) plant6

, Crystal River 3 (CR3)8 and St. Lucie) which were also approved by the NRC. References 4 and 9 are based on the NEI evaluation technique (Reference 2).

This report examines a baseline analysis using the NEI-based method and includes a sensitivity case that addresses the risk impact including external events (Appendix A).

The intact containment dose and doses for Classes 6, 7 and 8 are developed in Appendix B. An evaluation of the embedded liner degradation potential is performed consistent with that provided in Reference 3 and is documented in Appendix C. SONGS experience related to Type A testing is also provided in Appendix D.

1.1

SUMMARY

OF TEE ANALYSIS 10 CFR 50, Appendix J10 allows individual plants to extend Type A surveillance testing requirements and to provide for performance-based leak testing. This report documents a risk-based evaluation of the proposed change of the ILRT interval for the SONGS Unit 3.

The proposed change would impact testing associated with the current surveillance tests for Type A leakage, procedure S03-V-3.1211 for Unit 3. No change to Type B or Type C testing is proposed.

This analysis utilizes the guidelines set forth in NEI 94-01 2, the methodology used in EPRI TR-104285 and NUREG-149314.

The NEI guidance also considers the submittals generated by other utilities. The assessment contained in this document utilizes the method and metrics set forth in Reference 9 supported by the metrics identified in Reference 12.

The regulatory guidance on the use of probabilistic risk assessment (PRA) findings in support of a licensee request to a plant's licensing basis, RG 1.17415 is also utilized. The analysis utilizes the recent updated SONGS PRA results provided in References 16 and 17.

This calculation evaluates the risk associated with various ILRT intervals as follows:

& 3 years - Interval based on the original requirements of 3 tests per 10 years 10 years - Test interval specified by 10 CFR 50, Appendix J, Option B 15 years - This is the current test interval approved for SONGS Unit 3 16 years - Proposed extended test interval for Unit 3 RSC 07-05 I

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Evaluation of Risk Significance of ILRT Extension Based on the NE] Approach 1.2

SUMMARY

OF RESULTS/CONCLUSIONS The specific results are summarized in Table I below. The Type A contribution to LERF is defined as the contribution from Class 3b.

The detailed calculations performed to support this report were of a level of mathematical significance necessary to calculate the results recorded. However, the tables and illustrational calculation steps presented may present rounded values to support readability.

RSC 07-05 Printed: 07/25/2007

Evalutation of Risk Significance of ILRT Extension Based on the NEI Approach Table I Summary of Risk Impact on Extending Type A ILRT Test Frequency Risk Impact for 3-years Risk Impact for 10-years Risk Impact for 15-years Risk Impact for 16-years (baseline)

(current requirement)

Total integrated risk (person-rem/yr) 9.685 9.689 9.692 9.693 Type A testing risk (person-rem/yr) combination of 3a and 3b 1.89E-3 6.3 IE-3 9.46E-3 I.OIE-2

% total risk (Type A / total) 0.020%

0.065%

0.098%

0.104%

Type A LERF (Class 3b) (per year)

Taken from Table 18 1.70E-8 5.65E-8 8.48E-8 9.05E-8 Conditional Containment Failure Probability (CCFP)

Taken from Table 19 1.87E-I 1.94E-I 1.98E-t 1.99E-I Changes due to extension from 15 years (current)

A Risk from current (Person-rem/yr) 5.79E-4 17c Increase from current (A Risk / Total Risk) 0.006%

A LERF from current (per year) 5.65E-9 A CCFP from current 0.09%

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EvahlwIion of Risk Significance of TLRT Extension Based im the NEI Approach Table I (continued)

Summary of Risk Impact on Extending Type A ILRT Test Frequency Changes due to extension from 10 years A Risk from t0-year (Person-rem/yr) 3.48E-3

% Increase from 10-year (A Risk / Total Risk) 0.036%

A LERF from. 10-year (per year) 3.40E-8 A CCFP from 10-year 0.55%

Changes due to extension from 3 years (baseline)

A Risk from baseline (Person-rem/yr) 7.53E-3

% Increase from baseline (A Risk / Total Risk) 0.07 8%

A LERF from baseline (per year) 7.35E-8 A CCFP from baseline 1.18%

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Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach The results are discussed below:

" The person-rem/year increase in risk contribution from extending the ILRT test frequency from the current once-per-fifteen-year interval to once-per-sixteen years is 0.000579 person-remn/year.

The risk increase in LERF from extending the ILRT test frequency from the current once-per-15-year interval to once-per-16 years is 5.65E-9/yr.

The change in conditional containment failure probability (CCFP) from the current once-per-15-year interval to once-per-16 years is 0.09%.

The change in Type A test frequency from once-per-fifteen-years to once-per-sixteen-years increases the risk impact on the total integrated plant risk by only 0.006%. Also, the change in Type A test frequency from the original three-per-ten-years to once-per-sixteen-years increases the risk only 0.078%. Therefore, the risk impact when compared to other severe accident risks is negligible.

Reg. Guide 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Reg. Guide 1.174 defines very small changes in risk as resulting in increases of core damage frequency (CDF) below 10-I/yr and increases in LERF below 10-7/yr. Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A ILRT test interval from a once-per-fifteen-years to a once per-sixteen-years is 5.65E-9/yr. Guidance in Reg.

Guide 1.174 defines very small changes in LERF as below 10- /yr, increasing the ILRT interval from 15 to 16 years is therefore considered non-risk significant and the results support this determination. In addition, the change in LERF resulting from a change in the Type A ILRT test interval from a three-per-ten-years to a once per-sixteen-years is 7.35E-8/yr, is also below the guidance.

" R.G. 1.174 also encourages the use of risk analysis techniques to help ensure and show that the proposed change is consistent with the defense-in-depth philosophy. Consistency with defense-in-depth philosophy is maintained by demonstrating that the balance is preserved among prevention of core damage, prevention of containment failure, and consequence mitigation. The change in conditional containment failure probability was estimated to be 0.09% for the proposed change and 1.18% for the cumulative change of going from a test interval of 3 in 10 years to 1 in 16 years. These changes are small and demonstrate that the defense-in-depth philosophy is maintained.

In reviewing these results the SONGS Unit 3 analysis demonstrates that the change in plant risk is small as a result of this proposed extension of ILRT testing. The change in LERF defined in the analysis for the sensitivity studies also indicate that the change in LERF is within the acceptance criterion.

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Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach 2.0 DESIGN INPUTS The SONGS PRA is intended to provide "best estimate" results that can be used as input when making risk informed decisions.

The PRA report provides the most recent results for the SONGS PRA. The inputs for this calculation come from the information documented in the SONGS PRA and the level 2 update (References 16 and 17). The SONGS plant damage states are summarized in Table 2.

Table. 2 SONGS Plant Damage States Plant Damage Representative Sequence Frequency State

(/yr)

PDS I Transient with Loss of Secondary Heat Removal 7.7E-06 Transient with Loss of Secondary Heat Removal, and Loss of Containment PDS 2 Spray Recirculation 4.4E-09 Transient with Loss of Secondary Heat Removal, and Loss of Containment PDS 3 Heat Removal 4.0E-07 PDS 4 Transient/SSL with Loss of HPSI in Recirculation 7.2E-08 TransientlSSL with Loss of HPSI in Recirculation, and Loss of Containment PDS 5 Heat Removal 7.6E-14 Transient/SSL with Loss of 1PSI in Recirculation, Loss of Secondary Heat PDS 6 Removal, and Loss of Containment Spray Recirculation

8. IE-07 Transient with Loss of I-fPS1/LPSI Injection and Loss of Containment Heat PDS 7 Removal 5.IE-06 Transient with Loss of HPSI/LPSI Injection. Loss of Secondary Heat PDS 8 Removal and Loss of Containment Heat Removal 9.2E-07 PDS 9 Small LOCA with Loss of Containment Spray Recirculation 5.9E-09 PDS 10 Small LOCA with Loss of Containment Heat Removal 1.1E-07 PDS II Small LOCA with Loss of Secondary Heat Removal 4.3E-07 PDS 12 Small LOCA with Loss of HPSI Recirculation 0.OE+00 RSC 07-05 6

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Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table 2 (Continued)

SONGS Plant Darnage States Plant Damage Representative Sequence Frequency State

(/yr)

Small LOCA with Loss of I-PSI Recirculation. and Loss of Containment PDS 13 Spray Recirculation S.IE-06 Small LOCA with Loss of HPSI Recirculation. and Loss of Containment Heat PDS 14 Removal 0.OE+00 Small LOCA with Loss of HPSI Recirculation and Loss of Secondary Heat PDS 15 Removal 2.9E-07 Small LOCA with Loss of HPSI Recirculation. Loss of Secondary Heat PDS 16 Removal, and Loss of Containment Spray Recirculation 5.3E-06 Small LOCA with Loss of HPSI Recirculation, Loss of Secondary Heat PDS 17 Removal, and Loss of Containment Heat Removal 5.7E-08 PDS 18 Small LOCA with Loss of HPS1/LPSI Injection 1, IE-07 PDS 19 Large/Medium LOCA with Loss of Core Heat Removal 2-3F-07 Large/Medium LOCA with Loss of Core Heat Removal, and Loss of PDS 20 Containment Spray Recirculation 1.9E-10 PDS 21 Large/Medium LOCA with Loss of Core and Containment Heat Removal 3.4E-09 PDS 22 Large/Medium LOCA with Loss of HPSI recirculation 1.5E-07 Large/Medium LOCA with Loss of IHPSI recirculation, and Loss of PDS 23 Containment Spray Recirculation

3. IE-10 Large/Medium LOCA with Loss of HPSI recirculation, and Loss of PDS 24 Containment Heat Removal 8.9E-10 PDS 25 Large/Medium LOCA with Loss of HPSI/LPSI Injection 1.3E-10 PDS 26 Transient/LOCA with Loss of Containment Isolation and Heat Removal 7.OE-08 PDS 27 Interfacing System LOCA (ISLOCA) Initiating Event 4.8E-08 Steam Generator Tube Rupture (SGTR) Initiating Event, No Stuck Open PDS 28 Relief Valve (SORV) 6.3E-08 Steam Generator Tube Rupture (SGTR) Initiating Event, with Stuck Open PDS 29 Relief Valve (SORV)

5. IE-08 RSC 07-05 7

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Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table 2 (Continued)

SONGS Plant Damage States Plant Damage Representative Sequence Frequency State

(/yr)

TOTAL 3.004E-5 In order to develop a plant damage state person-remn dose it is necessary to associate each plant damage state with an associated release of radionuclides and from this information to calculate the associated dose.

The IP3 submittal (Reference 6) utilizes a multiplication factor to adjust the design basis leakage value (L,) that is based on generic information that relates dose to leak size. The CR3 submittal (Reference 8) utilized plant-specific dose estimates based on the predicted level 2 analysis results.

The SONGS PRA (References 16 and 17) contains the necessary information to convert the plant damage states to release categories. Using this information, the plant damage states are mapped to the six release categories: B, D, G, L, T, and W. In addition, the fraction of intact containment cases is determined using the split fraction information contained in References 16 and 17.

The SONGS PRA (References 16 and 17) release categories are defined by the reiease fraction of major radionuclides. The release fractions provided for SONGS are used to match the dose results from a surrogate NUREG-1150 plant (Surry) in order to identify the dose associated with each release category. This approach was utilized in the performance of a sensitivity study associated with accident sequence dose terms documented in Reference 4.

The intact containment dose is developed using intact containment leakage rate information for SONGS.

The development of the dose information is provided in Appendix B.

The SONGS release category dose information is presented in Table 3.

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Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table 3 Release Category Radionuclide Fraction Release Representative Frequency Category Sequence

(/yr)

Noble Gas' Iodine' Cesiurn' Telluiiurn' StrontiumI IC-1 (S) 2.45E-05 NA-NA NA NA NA B3 SGTR-20 9.67E-07 2.OE-3 4.OE-5 3.OE-5 3.OE-5 2.OE-6 iD4 SGTR2-66 5.07E-08 6.4E-I 3.4E-1 1.6E-I 2.7E-I 5.0E-3 G 5 SBO-17 3.226E-07 3-.IE-1

3. 2E-

-2E -2 5.0E-2 1.OE-3 L()

LOP-48 3.57E-06 2.9E-1 3.OE-3 3.OE-4 1.OE-4 2.OE-6 T7 VSEQ-2 4.79E-08 1.OE+0 9.OE-1 8.9E-1 S.6E-1 1.3E-2 WS MIO-20 5.82E-07 2.0E-3 2.OE-5 2.OE-5 2.OE-5 6.OE-6

1. Contributing fission oroduct gr,,ouDs are discussed in Reference 3.

2.

3.

4.

5.

6.

7.

S.

Release fractions not necessary for this calculation.

Release category B is defined by containment bypassed with less than 0.1% of volatiles released.

Release category D is defined by containment bypassed with up to 10% of volatiles released.

Release category G is defined by early, or isolation failure, containment failure prior to or at vessel failure with up to 10% of volatiles released.

Release category L is defined by late containment failure with tip to 1% of volatiles released.

Release category T is defined by containment bypassed with greater than 10% of volatiles released.

Release category W is defined by late containment failure with more than 10% of volatiles released.

Other inputs to this calculation include ILRT test data from NUREG-1493 (Reference 14) and the EPRL report (Reference 13) and are referenced in the body of the calculation.

3.0 ASSUMPTIONS

1. The maximum containment leakage for EPRI Class I (Reference 13) sequences is I La (Type A acceptable leakage) because a new Class 3 has been added to account for increased leakage due to Type A inspections.

2. The maximum containment leakage for Class 3a (Reference 2) sequences is 10 L. based on the NEI guidance and previously approved methodology (References 5, 6 and 8).

3. The maximum containment leakage for Class 3b sequences is 35 L, based on the NEI guidance (Reference 2) and previously approved methodology (References 5, 6 and 8).

4. Class 3b is conservatively categorized LERF based on the NEI guidance and previously approved methodology (References 5, 6 and 8).

5. Containment leakage due to EPRI Classes 4 and 5 are considered negligible based on the NEI guidance and the previously approved methodology (References 5, 6 and 8).

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Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach

6.

The containmenl releases are not impacted with time.

7. The containment releases for EPRI Classes 2, 6, 7 and 8 are not impacted by the ILRT Type A Test frequency. These classes already include containment failure with release consequences equal or greater than those impacted by Type A.

S. Because EPRI Class 8 sequences are containment bypass sequences, potential releases are directly to the environment. Therefore, the containment structure will not impact the release magnitude.

4.0 CALCULATIONS This calculation applies the SONGS PRA release category information in terms of frequency and person-remn estimates to estimate the changes in risk due to increasing the ILRT test interval.

The changes in risk are assessed consistent with the guidance provided in the NEI interim guidance document (Reference 2). This approach considers other similar analyses presented in EPRI TR-104285 (Reference 13) and NUREG-1493 (Reference 14).

The detailed calculations performed to support this report were of a level of mathematical significance necessary to calculate the results recorded. However, the tables and illustrational calculation steps presented may present rounded values to support readability. Exact values are presented enclosed in braces, 1}. where more precision is necessary to represent results.

4.1 CALCULATIONAL STEPS The analysis employs the steps provided in Reference 2 and uses risk metrics presented in Reference 15 to evaluate the impact of a proposed change on plant risk. These measures are the change in release frequency, the change in risk as defined by the change in person-rern, the change in LERF and the change in the conditional containment failure probability.

Reference 15 also lists the change in core damage frequency as a measure to be considered.

Since the testing addresses the ability of the containment to maintain its function, the proposed change has no measurable impact on core damage frequency. Therefore, this attribute remains constant and has no risk significance.

The overall analysis process is outlined below:

" Define and quantify the baseline plant damage classes and person-remn estimates.

Calculate baseline leakage rates and estimate probability to define the analysis baseline.

Develop baseline population dose (person-rem) and population dose rate (person-.rem/yr).

" Modify Type A leakage estimate to address extension of the Type A test frequency and calculate new population dose rates, LERF and conditional containment failure probability.

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Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach 0

Compare analysis metrics to estimate the impact and significance of the increase related to those metrics.

The first step in the analysis is to define the baseline plant damage classes and person-remn dose measures. Plant damage state information is developed using the SONGS PRA (References 17 and 18) results. The plant damage state information and the results of the containment analysis are used to define the representative sequences. The population person-remn dose estimates for the key plant damage classes are based on the application of the method presented in Appendix B and consistent with the approach used in Reference 4.

The product of the person-remn for the plant damage classes and the frequency of the plant damage state are used to estimate the annual person-rern for the plant damage state. Summing these estimates produces the annual person-rem dose based on the sequences defined in the PRA.

The PRA plant damage state definitions considered isolation failures due to Type B and Type C faults and examine containment challenges occurring after core damage and/or reactor vessel failure. These sequences are grouped into key plant damage classes. Using the plant damage state information, bypass, isolation failures and phenomena-related containment failures are identified.

Once identified, the sequence was then classified by release category definitions specified in Reference 13.

With this information developed, the PRA baseline inputs are completed.

The second step expands the baseline model to address Type A leakage.

The PRA did not directly address Type A (liner-related) faults and this contribution must be added to provide a complete baseline. In order to define leakage that can be linked directly to the Type A testing, it is important that only failures that would be identified by Type A testing exclusively be included.

Reference 2 provides the estimate for the probability of a leakage contribution that could only be identified by Type A testing based on industry experience. This probability is then used to adjust the intact containment category of the SONGS PRA to develop a baseline model including Type A faults.

The release, in terms of person-rem, is developed based on information contained in Appendix B and is estimated as a leakage increase relative to allowable dose (La) defined as part of the ILRT.

The predicted probability of Type A leakage is then modified to address the expanded time between testing. This is accomplished by a ratio of the existing testing interval and the proposed test interval. This assumes a constant failure rate and that the failures are randomly dispersed during the interval between the test.

The change due to the expanded interval is calculated and reported in terms of the change in release due to the expanded testing interval, the change in the population person-rem and the change in large early release frequency.

The change in the conditional containment failure probability is also developed.

From these comparisons, a conclusion is drawn as to the risk significance of the proposed change.

Using this process, the following were performed:

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Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach

1. Map the SONGS release categories into the 8 release classes defined by the EPRJ Report (Reference 13).

2. Calculate the Type A leakage estimate to define the analysis baseline.

3. Calculate the Type A leakage estimate to address the current inspection frequency.

4.

Modify the Type A leakage estimates to address extension of the Type A test interval.

5. Calculate increase in risk due to extending Type A inspection intervals.

6. Estimate the change in LERF due to the Type A testing.

7. Estimate the change in conditional containment failure probability due to the Type A testing.

4.2 SUPPORTING CALCULATIONS Siep I. Map tlhe Level 3 release categories into the 8 release classes defined by the EPRI Report EPRI Report TR-104285 (Reference 13) defines eight (8) release classes as presented in Table 4.

Table 4 Containment Failure Classifications (from Reference 13)

Failure Description Interpretation for Assigning SONGS Release Classification Category Containment remains intact with containment Intact containment bins initially isolated Dependent failure modes or common cause Isolation faults that are related to a loss of failures power or other isolation failure mode that is not a direct failure of an isolation component 3

Independent containment isolation failures due Isolation failures identified by Type A to Type A related failures testing 4

Independent containment isolation failures due Isolation failures identified by Type B testing to Type B related failures 5

Independent containment isolation failures due isolation failures identified by Type C testing to Type C related failures 6

Other penetration failures Other faults not previously identified 7

Induced by severe accident phenomena Early containment failure sequences as a result of hydrogen burn or other early phenomena 8

Bypass Bypass sequence or SGTR RSC 07-05 12 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table 5 presents the SONGS release category mapping for these eight accident classes. Person-rem per year is the product of the frequency and the person-reim.

Table 5 SONGS PRA Release Category Grouping to EPRI Classes (as described in Reference 13)

Person-Class Description Release Category Frequency Person-Rem Remlyr I

No containment failure IC-1 (S) 2.45E-5 1.70E+2 4.17E-3 Large containment None isolation failures Sinall isolation failures Not 3aNone 0.00E+0 (liner breach) addressed 3b Large isolation failures None Not 0.00E+0 (liner breach) addressed Small isolation failures -

4 None failure to seal (type B)

Small isolation failures -.

None.

5 None failure to seal (type C)

Containment isolation 6

failures (dependent failure, G

3.26E-07 2.06E+06 2 7.OOE-01 personnel errors)

Severe accident 7

phenomena induced failure L, W 4.15E-06 1.62E+06 2 6.73E+00 (early and late) 8 Containment bypass B, D, T 1.07E-06 2.14E+06 2 2.28E+00 Total 3.OOE-05 9.684

1. e represents a probabilistically insignificant value.

2.

The value presented represents an average of the Appendix B.

contributing release categories and is developed in Step 2: Calculate the Type A leakage estimate to define the analysis baseline (3 Year test interval)

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Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach As display"ed in Table 5 the SONGS PRA did not identify any release categories specifically associated with EPRI Classes 3. 4, or 5. Therefore each of these classes must be evaluated for applicability to this study.

Class 3:

Containment failures in this class are due to leaks such as liner breaches that could only be detected by performing a Type A ILRT. In order to determine the impact of the extended testing interval, the probability of Type A leakage must be calculated.

In order to better assess the range of possible leakage rates, the Class 3 calculation is divided into two classes. Class 3a is defined as a small liner breach and Class 3b is defined as a large liner breach.

This division is consistent with the NEI guidance (Reference 2) and the previously approved methodology (References 5, 6 and 8).

The calculation of Class 3a and Class 3b probabilities is presented below.

Calculation of Class 3a Probability The data presented in NUREG-1493 (Reference 14) is also used to calculate the probability that a liner leak will be small (Class 3a). The data found in NUREG-1493 states that 144 ILRTs were conducted. The data reported that 23 of 144 tests had allowable leak rates in excess of 1.OL.

However, of the 23 events that exceeded the test requirements, only 4 were found by an ILRT, the others were found by Type B and C testing or were identified as enrors in test alignments.

Data presented in Reference 2, taken since t/1/1995, increases this database to a total of 5 Type A leakage events in total of 182 events. Using the data a mean estimate for the probability of leakage is determined for Class 3a as shown in Equation 1.

5 PFI,,.,3a -

0.0275 (eq. 1) 182 This probability, however, is based on three tests over a 10-year period and not the one per fifteen-year frequency currently employed at SONGS (Reference 1). The probability (0.0275) must be adjusted to reflect this difference and is adjusted in step 3 of this calculation.

Multiplying the CDF times the probability of a Class 3a leak develops the Class 3a frequency contribution in accordance with guidance provided in Reference 2. This is conservative since part of the CDF already includes LERF sequences. The CDF for SONGS is {3.004E-5/yr} as presented in Table 2.

Therefore the frequency of a Class 3a failure is calculated as:

FREQci,*a 3a = PROB....,,.. x CDF = 0.0275 x 3.004E-5/yr = 8.254E-7/yr (eq. 2)

Calculation of Class 3b Probability RSC 07-05 14 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEl Approach To calculate the probability that a liner leak will be large (Class 3b) use was made of the data presented in the calculation of Class 3a. Of the events identified in NUREG-1493 (Reference 14), the largest reported leak rate from those 144 tests was 21 times the allowable leakage rate (L,). Since 21 L, does not constitute a large release. no large releases have occurred based on the 144 ILRTs reported in NUREG-1493.

The additional data point was also not considered to constitute a large release.

To estimate the failure probability given that no failures have occurred, the guidance provided in Reference 2 suggests the use of a non-informative prior. This approach essentially updates a uniform distribution (no bias) with the available evidence (data) to provide a better estimation of an event.

A beta distribution is typically used for the uniform prior with the parameters u=0.5 and 0=1.

This is then combined with the existing data (no Class 3b events, 182 tests) using Equation 3.

n + -a 0+0.5 0.5 P

=--

- 0.00273 (eq. 3)

N+f/

182+1 183 where: N is the number of tests, n is the number of events (faults) of interest, ca, 03 are the parameters of the non-informative plior distribution. From this solution, the frequency for Class 3b is generated using Equation 4 and is adjusted appropriately in step 3.

FREQcklssb = PROB..... x CDF = 0.00273 x 3.004E-5/yr = 8.209E-8/yr (eq. 4)

Class 4:

This group consists of all core damage accident accidents for which a failure-to-seal containment isolation failure of Type B test components occurs. By definition, these failures are dependent on Type B testing, and Type A testing will not impact the probability. Therefore this group is not evaluated any further, consistent with the approved methodology.

Class 5:

This group consists of all core damage accident accidents for which a failure-to-seal containment isolation failure of Type C test components occurs. By definition, these failures are dependent on Type C testing, and Type A testing will not impact the probability. Therefore this group is not evaluated any further, consistent with the approved methodology.

Class 6:

The Class 6 group is comprised of isolation faults that occur as a result of the accident sequence progression. The leakage rate is not considered large by the PRA definition and therefore it is placed into Class 6 to represent a small isolation failure and identified in Table 5 as Class 6.

FREQc.is 6 = 3.26E-7/yr (eq. 5)

Class 1:

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Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Although the frequency of this class is not directly impacted by Type A testing, the PRA did not model Class 3 failures, and the frequency for Class I should be reduced by the estimated frequencies in the new Class 3a and Class 3b in order to preserve the total CDF. The revised Class 1 frequency is therefore:

FREQci,,si = FREQciisi - (FREQc3aso3

+ FREQcass3b)

(eq. 6)

FREQISt = 2.45E-5/vr - (8.254E-7/yr + 8.209E-S/yr) = 2.359E-5/yr (eq. 7)

Class 2:

The SONGS PRA did not identify any contribution to this group above the quantification truncation.

Class 7:

The frequency of Class 7 is the sum of those release categories identified in Table 5 as Class 7.

FREQIas 7 = 4.15E-6/yr (eq. 8)

Class 8:

The frequency of Class 8 is the sum of those release categories identified in Table 5 as Class 8.

FREQla,8 = 1.07E-6/yr (eq. 9)

Table 6 summarizes the above information by the EPRI defined classes. This table also presents dose exposures calculated using the methodology described in Appendix B. For Class 1, 3a and 3b, the person-rem is developed based on the design basis assessment of the intact containment also developed in Appendix B.

The Class 3a and 3b doses are represented as OL, and 35L, respectively. Table 6 also presents the person-remn frequency data determined by multiplying the failure class frequency by the corresponding exposure.

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Evaluation of Risk Significance of ILRT Extension Based on the NE] Approach Table 6 Baseline Risk Profile Class Description Frequency Person-remn Person-rein Person-rem

(/yr)

(calculated)'

(from L;,

(/yr) factors)

No containment failure 2.36E-05 1.70E+2-4.01E-3 Large containment 0.OOE+00 isolation failures 3a Small isolation failures 8.25E-07 1.70E+34 1.40E-3 (liner breach) 3b Larue isolation failures 8.21E-08 5.95E+3-4.89E-4 (liner breach)

Small isolation failures -

O.OOE+..

failure to seal (type B)

Small isolation failures -

O.OOE+"O failure to seal (type C)

Containment isolation 6

failures (dependent failure, 3.26E-07 2.06E+66 7.OOE-1 personnel errors)

Severe accident 7

phenomena induced failure 4.15E-06 1.62E+6'6 6.73E+O (early and late.)

8 Containment bypass 1.07E-06 2.14E+6" 2.28E+O Total 3.OOE-05 9.685 I.

3.

4.

5.

6.

From the method presented in Appendix B.

I times L,, dose value calculated in Appendix B.

E represents a probabilistically insignificant value.

10 times L,,.

35 times L,,.

The value presented represents an average of the contributing release categories and is developed in Appendix B.

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Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach The percent risk contribution due to Type A testing is defined as follows:

%Risk*,,[ =[( Class3a.,,, + Class3bBASE)

Total,.S] x 100 (eq. 10)

Where:

Class3a'a,, = Class 3a person-rem/year =1.404E-3 person-rein/year Class3b.AS= Class 3b person-remi/year = 4.887E-4 person-rem/year Total...E = total person-remn year for baseline interval = 9.685 person-remi/year (Table 6)

%RiskBISI = [(1.404E-3 + 4.887E-4) / 9.685] x 100 = 0.020%

(eq. 11)

Step 3: Calculate the Type A leakage estimale to address the current i77spectiOn i07te1rval The current surveillance testing requirements as proposed in NEI 94-01 (Reference 12) for Type A testing and allowed by 10 CFR 50, Appendix J (Reference 10) is at least once per 10 years based on an acceptable performance history (defined as two consecutive periodic Type A tests at least 24 months apart in which the calculated performance leakage was less than LOLa).

According to References 2 and 14, extending the Type A ILRT interval from 3-in-10 years to I-in-10 years will increase the average time that a leak detectable only by an ILRT goes undetected from 18 to 60 months. Multiplying the testing interval by 0.5 and multiplying by 12 to convert from "years" to "months" calculates the average time for an undetected condition to exist.

The increase for a 10-yr ILRT interval is the ratio of the average time for a failure to detect for the increased ILRT test interval (from 18 months to 60 months) multiplied by the existing Class 3a probability as shown in Equation 12.

)0.0275x--j

=0.0916 (eq. 12)

A similar calculation is performed for the Class 3b probability as presented in Equation 13.

Pclass3b (10))

0.00273x

=0I 0.0091 (eq. 13) y 18, Risk Impact due to 10-year Test Interval Based on the previously approved methodology (References 5, 6 and 8) and the NEI guidance (Reference 2), the increased probability of not detecting excessive leakage due to Type A tests directly impacts the frequency of the Class 3 sequences.

Consistent with Reference 2 the risk contribution is determined by multiplying the Class 3 accident frequency by the increase in the probability of leakage.

Additionally the Class I frequency is adjusted to maintain the overall core damage frequency constant. The results of this calculation are presented in Table 7 below.

RSC 07-05 is Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table 7 Risk Profile for Once in Ten Year Testing Class Description Frequency (/yr)

Person-ren 2 Person-rem (/yr) 1 No Containment Failure'

2. 15E-5 1.70E+2 3.65E-3 Large Containment Isolation Failures Small Isolation Failures (Liner 2.75E-6 1.70E+3 4.68E-3 breach) 3b Large Isolation Failures (Liner 2.74E-7 5.95E+3 1.63E-3 breach)

Small isolation failures - failure to seal (type B)

Small isolation failures - failure to seal (type C)

Containment Isolation Failures 6

(dependent failure, personnel 3.26E-7 2.06E+6' 7.00E-I errors)

Severe Accident Phenomena 4.15E-6 1.62E+6' 6.73E+0 Induce Failure (Early and Late) 8 Containment Bypass 1.07E-6 2.14E+6 4 2.28E+O Total 3.OOE-5 9.689

1. The PRA frequency of Class I has been reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF.

2. From Table 6.

3. e represents a probabilistically insignificant value.

4. The value presented represents an average of the contributing release categories and is developed in Appendix B.

Using the same methods as for the baseline, and the data in Table 7 the percent risk contribution due to Type A testing is as follows:

%Riskio

[(Class3a,o + Class3bo) / Total,] x 100 (eq. 14)

Where:

Class3a,,

Class 3a person-rem/year = 4.68E-3 person-rem/year Class3b,o Class 3b person-remn/year = 1.63E-3 person-ren/year Total,, = total person-rem year for current 10-year interval = 9.689 person-rem/year (Table 7)

%Risk,o

[(4.68E-3 + 1.63E-3) / 9.689] x 100 0.065%

(eq. 15)

RSC 07-05 19 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach The percent risk increase (ARisko) due to a ten-year ILRT over the baseline case is as follows:

ARisk,, = [(Total,, - Total.ASE) I Total,.\\,,] x 100.0 (eq. 16)

Where:

Total,,,,, = total person-rein/year for baseline interval = 9.685 person-ren/year (Table 6)

Total. = total person-rem/year for 10-year interval = 9.689 person-rem/year (Table 7)

A%Risk,o= [(9.689 - 9.685) / 9.685] x 100.0 = {0.042%}

(eq. 17)

Step 3b: Calculate the Type A leakage estimate to address the current extension inspection i.t7terval The current inspection interval for SONGS Unit 3, is once per 15 years and supported by the analysis is documented in References 3 and 4.

In this case, the average time that a leak detectable only by an ILRT test goes undetected increases to 90 months (0.5 x 15 x 12). For a 15-vr-test interval, the result is the ratio (90/18) of the exposure times as was the case for the 10 year case.

Thus, increasing the ILRT test interval frorn 3 years to 15 years results in a proportional increase in the overall probability of leakage.

The approach for developing the risk contribution for a 15-year interval is the same as that for the 10-year interval. The increase for a 15-yr ILRT interval is the ratio of the average time for a failure to detect for the increased JLRT test interval (from IS months to 90 months) multiplied by the existing Class 3a probability as shown in Equation 18.

(15y) 0.0275>X\\1-1

- 0.1375 (eq. 18)

ý18)

A similar calculation is performed for the Class 3b probability as presented in Equation 19.

Pc,.,..,35 (15Y)= 0.00273 X 9-

= 0.0137 (eq. 19)

(18)

As stated for the 10-year case, the increased probability of not detecting excessive leakage due to Type A tests directly impacts the frequency of the Class 3 sequences.

The increased risk contribution is determined by multiplying the Class 3 accident frequency by the increase in the probability of leakage.

Additionally the Class I frequency is adjusted to maintain the overall core damage frequency constant.

The results of this calculation are presented in Table 8 below.

RSC 07-05 20 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table 8 Risk Profile for Once in Fifteen Year Testing Class Description Frequency (/yr)

Person-rem 2 Person-rem (Iyr)

No Containment Failure 21.00E-5 1.70E+2 3.40E-3 Large Containment Isolation 3

Failures 3a Small Isolation Failures (Liner 4.1 3E-6 1.70E+3 7.02E-3 breach) 3b Large isolation Failures (Liner 4.10E7 5.59E+3 2.44E-3 breach)

Small isolation failures - failure to seal (type B) tSmall isolation failures - failure to seal (type C)

Containment Isolation Failures 6

(dependent failure, personnel 3.26E-7 2.06E+64 7.OOE-I errors)

Severe Accident Phenomena 4.15E-6 1.62E+64 6.73E+0 Induce Failure (Early and Late.)

S Containment Bypass 1.07E-6 2.14E+6 4 2.28E+0 Total 3.OOE-5 9.692 I The PRA frequency of Class I has been reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF.

2. From Table 6.

3. E represents a probabilistically insignificant value.

4. The value presented represents an average of the contributing release categories and is developed in Appendix B.

Using the same methods as for the baseline, and the data in Table 10 the percent risk contribution due to Type A testing is as follows:

%Risk,, =[( Class3a,, + Class3b,,) / Totalj] x 100 (eq. 20)

Where:

Class3a,, = Class 3a person-ren/year = 7.02E-3 person-remn/year Class3b,, = Class 3b person-remn/year = 2.44E-3 person-rem/year Total,, = total person-remn year for 15-year interval = 9.692 person-ren/year (Table 8)

%Risk,, = [(7.0 2 E-3 + 2.44E-3) / 9.692] x 100 = {0.098%}

(eq. 21)

RSC 07-05 21 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach The percent risk increase (A%Risk,,) due to a fifteen-year ILRT over the baseline case is as follows:

A%Risk,, = [(Total,, - TotalB1SE) / TotalAS,,] x 100.0 (eq. 22)

Where:

TotalB SF = total person-remn/year for baseline interval = 9.687 person-renm/year (Table 6)

Total,, = total person-rern/year for 15-year interval = 9.692 person-rem/year (Table 8)

A%Risk,, = [(9.692 - 9.685) / 9.685] x 100.0 = {0.072%}

(eq. 23)

The percent risk increase (A%Risk,5 ) due to a fifteen-year ILRT over the 10-year case is as follows:

A%Risk,,.,, = [(Total,5 - Total,) / Total,,] x 100.0 (eq. 24)

Where:

Total,. = total person-remn/year for 10-year interval = 9.689 person-remn/year (Table7)

Total,, =total person-rem/year for 15-year interval = 9.692 person-rem/year (Table 8)

A%Risk,-,,, = [(9.692 - 9.689) / 9.689] x 100.0 = 10.030%}

(eq. 25)

Step 4: Calculate the Type A leakage estimate to address extended inspection inten'als If the test interval is extended to I per 16 years, the average time that a leak detectable only by an ILRT test goes undetected increases to 96 months (0.5 x 16 x 12). For a 16-yr-test interval, the result is the ratio (96/18) of the exposure times as was the case for the 10 year case. Thus, increasing the ILRT test interval from 3 years to 16 years results in a proportional increase in the overall probability of leakage.

The approach for developing the risk contribution for a 16-year interval is the same as that for the 10-year interval. The increase for a 16-yr ILRT interval is the ratio of the average time for a failure to detect for the increased ILRT test interval (from 18 months to 96 months) multiplied by the existing Class 3a probability as shown in Equation 26.

Pchm 3, (16 Y) = 0.0275 x (18)

= 0.1465 (eq. 26)

A similar calculation is performed for the Class 3b probability as presented in Equation 27.

PC,0.,,3b (16 Y) = 0.00273 x ý96/= 0.01457 (eq. 27)

\\18)

RSC 07-0-5 Printed: 07/ 25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach As stated for the earlier, the increased probability of not detecting excessive leakage due to Type A tests directly impacts the frequency of the Class 3 sequences.

The increased risk contribution is determined by mnultiplying the Class 3 accident frequency by the increase in the probability of leakage.

Additionally the Class I frequency is adjusted to maintain the overall core damage frequency constant. The results of this calculation are listed in Table 9 below.

Table 9 Risk Profile for Once in Sixteen Year Testing Class Description Frequency (/yr)

Person-remn 2 Person-rem (/yr)

I No Containment Failure' 1.97E-5 1.70E+2 3.34E-3 Large Containment Isolation 3

Failures 3a Small Isolation Failures (Liner 4.40E6 1.70E+3 7.49E-3 breach) 3b Large Isolation Failures (Liner 4.38E-7 5.95E+3 2.61E-3 breach)

Small isolation failures - failure to seal (type B)

Small isolation failures - failure to seal (type C)

Containment Isolation Failures 6

(dependent failure, personnel 3.26E-7 2.06E+6' 7.00E-I errors)

Severe Accident Phenomena Induce Failure (Early and Late) 8 Containment Bypass 1.07E-6 2.14E+64 2.2 8E+0 Total 3.OOE-5 9.693

1. The PRA frequency of Class 1 has been reduced b total CDF.

2. From Table 6.

3. e represents a probabilistically insignificant value.

y the frequency of Class 3a and Class 3b in order to preserve

4. The value presented represents an average of the contributing release categories and is developed in Appendix B.

Using the same methods as for the baseline, and the data in Table 9 the percent risk contribution due to Type A testing is as follows:

%Risk,, =[( Class3a,. + Class3b,,,) / Total 6] x 100 (eq. 28)

Where:

RSC 07-05

?3 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Class3a,o = Class 3a person-rem/year = 7.49E-3 person-rem/year Class3b,<, = Class 3b person-remi/year = 2.61E-3 person-remn/year Total,, = total person-rem year for 16-year interval = 9.693 person-rem/year (Table 9)

%Risk,,= [(7.49E-3 + 2.61E-3) / 9.693] x 100 = {0.104%}

(eq. 29)

The percent risk increase (A%Risk,,) due to a sixteen-year ILRT over the baseline case is as follows:

A%Risk,, = [(Total,, - Total,,,,) / TotalBA,,] x 100.0 (eq. 30)

Where:

Total,,B

= total person-rem/year for baseline interval = 9.685 person-rem/year (Table 6)

Total,, = total person-rem/year for 16-year interval = 9.693 person-rem/year (Table 9)

A%Rlsk,,, _,. = [(9.693 - 9.685) / 9.685] x 100.0 = j0.078%}

(eq. 31)

The percent risk increase (A%Risk,) due to a sixteen-year ILRT over the 10-year case is as follows:

A%Risk,, = [(Total,, - Total,,) / Total,,] x 100.0 (eq. 32)

Where:

Total,, = total person-rem/year for 10-year interval = 9.689 person-remi/year (Table 7)

Total,, = total person-rem/year for 16-year interval = 9.693 person-remi/year (Table 9)

A%Risk,.,, = [(9.693 - 9.689) / 9.689] x 100.0 = {0.036%}

(eq. 33)

Step 5: Calculate increase in r-isk dcue to extending Type A inspection intervals Based on the guidance in Reference 2, the percent increase in the total integrated plant risk for these accident sequences is computed as follows:

%Total,,-,, = [(Total,, - Total,,) / Total,,] x 100 (eq. 34)

Where:

Total,, = total person-remn/year for 15-year interval

= 9.692 person-rem/year (Table 8)

Total,, = total person-rem/year for 16-year interval = 9.693 person-rem/year (Table 9)

%Total,,.,, = [(9.693 - 9.692) / 9.692] x 100 = {0.006%}

(eq. 35)

RSC 07-05 24 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Step 6. Cclculate the change in Risk in ternts of Large EarhY Release Freqiency (LERF)

The risk impact associated with extending the ILRT interval involves the potential that a core damage event that normally would result in only a small radioactive release from containment could in fact result in a larger release due to failure to detect a pre-existing leak during the relaxation period.

From References 2, 5, 6 and 8, the Class 3a dose is assumed to be 10 times the allowable intact containment leakage, L:, (or 1,700 person-rem) and the Class 3b dose is assumed to be 35 times L, (or 5,950 person-remn).

The dose equivalent for allowable leakage (La) is developed in Appendix B.

This compares to a historical observed average of twice L,. Therefore, the estimate is somewhat conservative.

Based on the NEI guidance (Reference 2) and the previously approved methodology (References 5, 6 and 8), only Class 3 sequences have the potential to result in large releases if a pre-existing leak were present. Class I sequences are not considered as potential large release pathways because for these sequences the containment remains intact. Therefore, the containment leak rate is expected to be small (less than 2L,). A larger leak rate would imply an impaired containment, such as Classes 2. 3, 6 and 7.

Late releases are excluded regardless of the size of the leak because late releases are, by definition, not a LERF event. At the same time, sequences in the SONGS PRA (References 16 and 17) that result in large releases, are not impacted because a LERF will occur regardless of the presence of a pre-existing leak.

Therefore, the change in the frequency of Class 3b sequences is used as the increase in LERF for SONGS, and the change in LERF can be determined by the differences. Reference 2 identifies that Class 3b is considered to be the contributor to LERF. Table 10 summarizes the results of the LERF evaluation that Class 3b is indicative of a LERF sequence.

Table 10 Impact on LERF due to Extended Type A Testing Intervals ILRT Inspection 3 Years (baseline) 10 Years 15 Years 16 Years Interval Class 3b (Type A 8.2 IE-8/yr 2.74E-7/yr 4.1OE-7/yr 4.38E-7/yr LERF)

ALERF (3 year 1.92E-7/yr 3.28E-7/yr 3.56E-7/yr baseline)

ALERF (10 year) 1.37E-7/yr

!.64E-7/yr ALERF( 15 year 2.74E-8/yir current)

RSC 07-05 2-5 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEl Approach Reg. Guide 1.174 (Reference 12) provides guidance for determining the risk impact of plant-specific changes to the licensinE basis. The Reg. Guide 1.174 defines very small changes in risk as resulting in increases of core damage frequency (CDF) below IE-6/yr and increases in LERF below IE-7/yr. Since the ILRT does not impact CDF, the relevant metric is LERF. Calculating the increase in LERF requires determining the impact of the ILRT interval on the leakage probability.

Increasing the ILRT interval from the currently 15 years to a period of 16 years increases the LERF contribution by 2.74E-8/yr. This value meets the guidance in Reg. Guide 1.174 defining very small changes in LERF that are acceptable and not significant.

The LERF increase measured from the original 3-in-10-year interval to the 16-year interval is 3.56E-7/yr, which exceeds the criterion presented in Regulatory Guide 1.174.

Reference 18 indicates that plants with a CDF in excess of 1.OE-5/yr may have difficulty demonstrating a change in LERF less than 1.OE-7/yr.

It further states that the analysis as embodied in the NEI approach is conservative and provides additional guidance with respect to refining the initial analysis.

Since the target value is exceeded some refinement is necessary. The increase is explicitly tied to the Class 3b contribution which is generated by multiplying the total CDF by the defined split fraction (0.0027).

Using the entire CDF frequency is conservative since some sequence frequencies comprising the total CDF already account for other LERF sequences which may occur due to interfacing system LOCA events or steam generator tube ruptures.

The first refinement centers on this conservatism.

Sequences which result in LERF contributions are not influenced (change in outcome) by the potential for Type A leakage and can be excluded from the calculation of Class 3 leakage.

The second aspect defined in Reference 18 addresses the magnitude of the source term expected to be available for release during the accident sequence. If the debris escapes the reactor vessel but remains essentially covered with water (either due to large pools or continual containment sprays) the source term will be greatly reduced and a large source term would not be expected.

Therefore, if the accident sequence involves containment spray operation or coverage of the debris with large pools of water, the source term is not considered sufficient to support a LERF release and these contributions can be excluded.

The SONGS Level 2 containment event tree model was utilized to identify the characteristics necessary for determining the status of these aspects of the analysis. The existing containment event tree model provided clear branch points for LERF, debris flooded and containment spray status in recirculation such that a set of logical rules could be applied to the model to obtain the necessary results in terms of CDF. The analysis is summarized in Reference 19 and the results illustrated graphically in Figure 1.

RSC 07-05 26 Printed: 07/25/2007

Evaluation of Risk SiEnificance of ILRT Extension Based on the NEI Approach FR"M CE!

LERF FILOODED SPRAYS SPECIAL IS DEE.

0" JtINT.TIII FA leqeeEcI CASE THERE FL:ODEII SPRA'f FOR LERF AFTER It FLLISRDEE VESSEL RECIEC7 FAILI PRE 47A.1.FLOODED fhil R8l* letIr'I l Rule N

IOl.el-d Ritle dIel1hed Role Ilellled NOS LERS 2.946..005S F 10 'JEER 4.28 1*005 NOT1 FLOODRED 2.52 Iý.00E SPRAYS 2.175,-006 No SPRAY 2.10e 0.0ol SPRAYS

,..15e-R06 2.17se W.

U 3

3.002,00E5 111, SERRO 1.954e.-005 4

6.645e.O00.

5.586,-E07

. V.,

5.586e-007

,:,-Vwl,IEIRIRA3RIISCE-oSEIII.FHiL-L2E01.M CA.1-FLCSS[IETISTE)

L'e S ee-,1: Weai,,esI.'r,. Eehil.oe 16. 2005 SLi leloe IcecillS. Nt.Wve,4 28. 2800.

WinRIJICEE' 1.0 iere e iuoi alrgir Figure 1. Quantified Source Term Category Diagram for SONGS Case 1 (from Reference 19)

RSC 07/-05 2)7 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NET Approach The figure illustrates how the total CDF frequency is subdivided based on the three identified criterion. From the solution presented in the figure a table of results is obtained and reproduced in Table 11.

Table 11 Source Term Outcomes Source Term Outcome Frequency (/yr)

Description STC Refinement Outcome I Non-LERF sequence with the 2.175E-06 debris flooded and containment sprays functioning STC Refinement Outcome 2 Non-LERF sequence with the debris flooded and the containment sprays do not function after recirculation STC Refinement Outcome 3 Non-LERF sequence without water 6.645E-06 pools covering the debris and containment sprays functioning.

STC Refinement Outcome 4 Non-LERF sequence without water 1.854E-05 pools covering the debris and containment sprays do not function STC Refinement Outcome 5 5.586E-07 LERF sequence Only outcOme 4 contributes to the potential for a Type A LERF. This value is then utilized to calculate the LERF contribution from Class 3b fiequency by the following equation:

FREQIS..3 b = PROB,,_, x Adjusted CDF = 0.00273 x 1.85E-5/yr = 5.07E-8/yr (eq. 36)

This can then be extrapolated using the methods presented earlier to determine the 10-year, 15-year and 16-year contributions and to generate adjusted LERF values as presented in Table 12.

RSC 07-O5 28 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table 12 Class 3b Contributions Using Adjusted CDF Test Interval Frequency (/yr)

Delta Frequency from Prior Period

(/yr)

Baseline 5.07E-8 10-year 1.69E-7

1. 18E-7 15-year (current) 2.54E-7 8.45E-8 16-year 2.70E-7 1.69E-8 Summing the last column provides the total increase from the baseline (3 year) to the proposed (16 year) interval (2.20E-7/yr). This increase is still in excess of the guidance for a small change in risk.

Since only a single branch point contributes to the LERF estimation, an additional evaluation of the contributors associated with this LERF contribution was performed. From this effort, it was identified that the major initiating event contributing sequence was small LOCA. A significant portion of this contribution dealt with a failure of the sump suction due to debris clogging.

A somewhat conservative assumption is made for this sequence that if debris is released complete clogging will occur. In the case of a large LOCA where significant pressure and steam flows are present it may be somewhat reasonable to assume clogging if materials are found to be present in the containment that could be swept to the sumps due to forces present during the initial blowdown.

For a small LOCA, however, this is believed to be overly conservative since the effects would be more localized and if debris were to be dislodged it would be of lesser quantity and therefore less likely to plug the sump. A more realistic estimation of the probability of clogging given a small LOCA is provided in Reference 19. The investigation also identified a conservative error factor estimate associated with the human error probability (HIEP) to close the containment sump access hatch prior to closure of the containment for planned startup.

This information was used to refine the LERF estimate. The STC Outcome 4 contains the small LOCA contribution of interest. The frequency of the small LOCA contribution is 1.20E-5/yr and the frequency of the sequence of interest (i.e., SLE-I1) is 8.08E-6/yr.

The small LOCA contribution was examined and the identified conservatisms addressed. The adjustment to the small LOCA frequency is presented in Table 13.

RSC 07-05 2?9 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table 13 Adjustment to Small LOCA Contribution to STC Refinement Outcome 4 (from Reference 19)

Attribute Frequency (/yr)

Total Frequency Reduction (/yr)

Initial small LOCA contribution 1.20E-5 Initial contribution from small LOCA 8.08E-6 with sump clogging (SLE-11)

Adjusted Small LOCA Contribution 5.57E-6

.5 1IE-6 (SLE-I 1I)

The inclusion of these refinements reduces the small LOCA contribution for STC Outcome 4 fiom 1.20E-5/yr to 9.48E-6/yr. This is then combined with the unadjusted contribution to STC Outcome 4 (6.56E-6/yr) to arrive at the new LERF contribution of 1.60E-5/yr. Substitution into the LERF calculation equations defined earlier yields the final results presented in Table 14.

Table 14 Class 3b Contributions Using Adjusted CDF Test Interval Frequency (/yr)

Delta.Frequency from Prior Period.

(/yr)

Adjusted STC Outcome 4 1.60E-5 Baseline LERF 4.3S8E-8.

10-year 1.46E-7 1.02E-7 15-year (current) 2.19E-7 7.3 IE-8 16-year 2.34E-7 1.46E-8 Again, summing the last column provides the total increase from the baseline (3 year) to the proposed (16 year) interval (1.90E-7/yr).

This decrease is not sufficiently small to meet the guidance for a small change in risk.

The guidance is based on the desire to avoid a significant increase in LERF. In addition to frequency, LERF is defined by both release (large) and timing (early). It is this aspect that is examined to determine if the existing analysis can be refined further.

The characteristics release categories from the SONGS Level 2 assessment were reviewed to examine the accident sequence progression. In particular the time at which the core is uncovered was examined.*,Typically, early releases involve fairly rapid voiding of the core and results in a RSC 07-05 30 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach significant quantity of radionuclides being present in the containment and available for early release. Early release timing is typically associated with a time sufficiently short that there is an impaired ability to evacuate individuals near the plant such that a fatality could be possible.

Many studies indicate that the timing of such events should be on the order of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (Reference S21). For this assessment, the breakpoint between early and late is chosen as 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months in keeping with prior analyses.

The release category results based on characteristic accident sequences were reviewed and timing of the accident progress examined for those involving intact containments were examined and compiled as shown in Table 15 below.

Table 15 Intact Containment Release Category Timing Release Category Sequence Time to Uncovering the Time to Reactor Pressure Number Core (hrs)

Vessel Failure (hrs)

PCS-4 0.7 4.0 MLO-4 3.1 5.7 LLO-4 0.004 2.9 ATWS-6 0.02 1.3 PCS-35 12.7 16.0 MLO-20 0.8 3.2 Examining the data in the table identifies that the core uncovers for case PCS-35 substantially later than the other intact cases. The timing being greater than 12.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months provides considerable margin to the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months requirement and it is not considered to be capable of generating a LERF sequence.

The sequence is not representative and can be excluded from the assessment for LERF due to isolation faults. The frequency associated with PCS-35 is 6.62E-6/yr (Reference 21).

A portion of this release category includes the SLE-11 sequence involving small LOCA sequences with sump faults. This sequence has been already addressed and must be removed from the total. Based on information contained in Reference 21, the adjusted frequency is 5.06E-6/yr.

Another sequence with delayed release is MLO-4. However, the core uncovers prior to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and the reactor vessel fails at approximately 5.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months . This is considered to be close enough to the limit of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months that it cannot be removed due to timing alone.

RSC 07-05 31 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach In an earlier step tile potential for water scrubbing of debris was considered as a means to preclude a larOe source term in the containment which could be released to the environment.

This process was used to exclude some fraction of cases and is in keeping with tile guidance provided in Reference 18.

However, if the source term in the containment is on the same order as that involving covered debris it is reasonable to assume that any leakage would not be sufficient to be classified as LERF. If we add to that assertion that the timing is very near the separation point between early and late releases, this is sufficient to exclude the contribution from LERF.

A review of the intact containment cases examined the various release contributions for key accident sequences and this was compared to the MLO-4 case. For MLO-4 there is no water coverage (Reference 21, Table 2). In contrast the PCS-4 release category was fully flooded. A third case examined earlier, PCS-35, also has no water cover over the debris.

The fission product released to the containment for these three release categories are presented in Table 16 as predicted by MAAP (Reference 21, Table 1).

Table 16 Radionuclide Releases by Release Category Radionuclide PCS-4 MLO,4 PCS-35 Product Group (%)

Nobles 0.1 0.1 0.1 CsI 5.OE-5 7.OE-5 4.OE-3

TeO, 6.OE-5 4.0E-5 I.0E-3 SrO 5.0E-6 2.OE-5 1.OE-4 MoO-4.OE-5 6.OE-6 6.OE-5 CsOH 4.OE-5 5.0E-5 2.OE-3 BaO 1.OE-5 1.OE-5 9.OE-5 La20 3 E

1.0E-6 2.OE-6 CeO2 F

2.OE-5 7.OE-5 Sb 3.OE-3 Te2-5.OE-5 UO2 9.OE-8 1.OE-7 F - insufficient quantity to report.

RSC 07-05 32?

Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the N-EI Approach As the table indicates, the MLO-.4 case is similar to that of the flooded case PCS-4. The releases for the most part are very similar and within the same range. It is clearly different than for the other dry case (PCS-35) where the releases vary by factors of 10 and 100 greater than for MLO-

4. Therefore, it appears that the MLO-4 releases are more closely linked to a flooded debris state than for a state without debris coverage.

There is one notable difference in the MLO-4 and the PCS-4 cases. The MLO-4 case contains releases of cerium and lanthanum and could represent additional dose if it were released early in the event. The MAAP analysis output was reviewed and the two fission product releases plotted with time to determine their release distribution and is presented as Figure 2 below.

Release of La203 and Ce02 1000%

100 0D,.

90 0%

800%

70 0%

60 0%

50.0%

40 0%

30 0%

20 0%

10 0%

0.0%

00 0.8 1.6 24 3.3 4.1 4.9 5.8 66 7.5 83 9.1 100 108 11.7 12.5 13.3 14.2 Time (hours)

Figure 2. La203 and CeO2 Release Timing (derived from Reference 21)

The figure plots the incremental and cumulative release normalized to the total release predicted by MAAP. The plot shows that neither of the fission products appear in significant quantities prior to the time of reactor vessel failure.

Fifty percent of the released amount is not present inside the containment until over 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months into the event. From this information it appears that these fission products to not influence the early portion of the sequence and that the MLO-4 sequence behaves similar to those screened earlier. The release of these two radionuclide groups is not early in the sequence during the initial fuel melt and relocation. This delays any measurable quantities from being released until after reactor vessel failure and the subsequent core and concrete interaction. Contributions from this activity will not occur until after the four hour time period and contributions from this sequence can be excluded from evaluation with regard to LERF.

RSC 07-05 33 Printed: 07/25/2007

Evaluation of Risk Significance of FLRT Extension Based on the NEI Approach The MLO-4 sequence does not contain anl\\ portion of previously addressed intact containment contributions and the associated frequency can be removed without ad justment. The fiequency for MLO-4 is 4.78E-6/yr (from Reference 21). The total impact of this adjustment is provided in Table 17.

Table 17 Adjustment to Address Source Term Magnitude and Timing Release Category Sequence Number Prior to Adjustment (/yr)

After Adjustment (/yr)

Initial Intact Containment Frequency 6.2 1E-6 6.21 E-6 (not including PCS-35 or MLO-4)

PCS-35 Frequency 5.06E-6 MLO-4 Frequency 4.78E-6 Total Intact Containment Frequency 1.60E-5 6.21E-6 The additional refinement reduces the contribution as shown. Using the new intact containment contribution, an adjusted value for LERF is quantified. Substitution into the LERF calculation equations defined earlier yields the final results presented in Table IS.

Table IS Class 3b Contributions Using Adjusted CDF Test Interval Frequency (/yr)

Delta Frequency -from Prior Period

(/yr)

Adjusted Outcome 6.2 IE-6 Baseline LERF 1.70E-8 10-year 5.65E-8 3.96E-8 15-year (current) 8.48E-8 2.83E-8 16-year 9.05E-8 5.65E-9 Again, summing the last column provides the total increase from the baseline (3 year) to the proposed (16 year) interval (7.35E-8/yr).

The quantified value meets the definition for an acceptably small change in risk as defined by the gulidance in Reference 15.

RSC 07-05 34 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Step 7: Calculate the change in Conditional Containment Faiure Probabilit, (CCFP)

The conditional containment failure probability (CCFP) is defined as the probability of containment failure given the occurrence of an accident. This probability can be expressed using the following equation:

CCFP = 1 -

CDF I (eq. 37)

Where ftncf) is the frequency of those sequences which result in no containment failure. This frequency is determined by summing the Class I and Class 3a results, and CDF is the total frequency of all core damage sequences.

Therefore the change in CCFP for this analysis is the CCFP using the results for 16 years (CCFPI) minus the CCFP using the results for 15 years (CCFP,,). This can be expressed by the following:

_ACCFfl'; 1b _ý:CCFP16 - CCFP1 5

(eq. 38)

Using the data previously developed the change in CCFP from the current testing interval is calculated and presented in Table 19.

Table 19 Impact on Conditional Containment Failure Probability due to Extended Type A Testing Intervals I1LRT Inspection 3 Years (baseline) 10 Years 15 Years 16 Years Interval f(ncf) (/yr)

.442E-5 2..4.2) 3E-5 2.409E-5 2.406E-5 flncf)/CDF 0.813 0.806 0.802 0.801 CCFP 0.187 0.194 0.198 0.199 ACCFP (3 year 6.38E-3 1.09E-2 1.18E-2 baseline)

(0.64%)

(1.09%)

(1.18%)

ACCFP (10 year) 4.55E-3 5.46E-3 (0.46%)

(0.55%.)

ACCFP (15-year 9.11E-4.

current)

(0.09%)

4.3 REFERENCES

1.

Docket No. 50-362 San Onofre Nuclear Generating Station, Unit 3, Amendment to Facility Operating License, Amendment 189, License NO. NPF-15, August 24, 2005.

RSC 07-05 35 Printed: 07/25/2007

EvalUation of Risk Significance of ILRT Extension Based on the NEI Approach

2.

Haugh, J., et al, Interimn Guidance for Performing Risk impact Assessments in Support of One-Time Extensions for Containment Intearated Leakaae Rate Test Surveillance Intervals. Revision 4, Nuclear Energy Institute (NEI), November 2001.

3.

Miller. J., San Onofre Nuclear Generatin2 Station Probabilistic Risk Assessment Evaluation of Risk Significance of ILRT Extension, Revision 0, Ricky Surnmitt Consulting (RSC), Inc., RSC 04-02, March 2004.

4.

Surnmnitt, R., San Onofre Nuclear Generating Station Unit 3 Probabilistic Risk Assessment Evaluation of Risk Significance of ILRT Extension Based on the NFI Approach, Revision 0, RSC, Inc., RSC 05-03, April 2005.

5.

Summitt, R., Comanche Peak Steam Electric Station Probabilistic Safety Assessment.

Evaluation of Risk Significance of ILRT Extension, RSC, Inc., RSC 01-47/R&R-PN-110, November 2001.

6.

Indian Point 3 Nuclear Power Plant. "Supplemental Information Regarding Proposed Change to Section 6.14 of the Administrative Section of the Technical S pecification",

Entergy, IPN-01-007, January 18, 2001.

7.

Indian Point Nuclear Generating Unit No.3 - Issuance of Amendment Re: Frequency of Performance-Based Leakage Rate Testing (TAC NO. MBO178), United States Nuclear Regulatory Commission (USNRC), April 17, 2001.

8.

Evaluation of Risk Significance of ILRT Extension, Revision 2, Florida Power Corporation, F-01-0001, June 2001.

9.

Docket No. 50-389 St. Lucie Plant, Unit 2, Amendment to Facility Operating License, Amendment 140, License No. NPF-16, December 23, 005.

10, Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooling Power Reactors, U.S. Nuclear Regulatory Commission (USNRC), 10 CFR Part 50, Appendix J, January 2006.

11. San Onofre Nuclear Generating Station Unit 3, Containment Integrated Leak Rate Test, Rev. 4, Procedure S03-V-3.12.

12, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50.

Appendix J; Revision 0, Nuclear Energy Institute, NE1 94-01, July 26, 1995.

13, Gisclon, J. M., et al, Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals, Electric Power Research Institute, TRA104285, August 1994.

14, Performance-Based Containment Leak-Test Program, USNRC, NUREG-1493, July 1995.

RSC 07-05 3-6 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach

15. An Approach for Using Probabilistic Risk Assessment in Risk-Informed decisions on Plant-Specific Changes to the Licensing Basis. USNRC, Regulatory Guide 1.174, July 1998.

16. San Onofre Nuclear Generatin.

Station Living PRA. SONGS 2/3. PRA Level II Analysis Report, IPE-LEVEL2-000.

17. Sarmanian, L., Quantification of Level 2 Release Categories for SONGS Units 2/3 Internal and External Events, November 2006.

18. Letter from A. Pietrangelo, NEI, Titled: One-time extensions of containment integrated leak rate test interval - additional infformation, November 30, 2001.

19. Sarmanian, L., Investigation into the Contributors of Level 2 Source Term Categories for SONGS Unit 2/3, December 2006.

20. An Approach for Estimating the Frequencies of Various Containment Failure Modes and Bypass Events, Rev. 1, USNRC, NIJREG/CR-6595, September 2004.

21. Moieni, P., Update of Frequencies for Source Term Categories PCS-35 and MLO-4,.

SONGS, PRA-07-002, January 24, 2007.

RSC 07-05 37 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Appendix A:

External Events Sensitivity Study RSC 07-05 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach A.0 EXTERNAL EVENTS SENSITIVITY STUDY NEI guidanceI suggests the need to address external initiating events when estimating the impact of the proposed ILRT extension in cases where additional refinements are made to the analysis.

A sensitivity study using data for the plant damage state frequencies including seismic and fire contribution to release frequency is used to address this requirement.

A. I

SUMMARY

OF THE ANALYSIS This section is completed in the same manner as NEI baseline analysis.

Information and approaches defined in the NEI baseline analysis and presented in the main body of the report are utilized in this appendix. The methodology steps outlined in Section 4.1 of the main document are applied. The section only addresses areas of deviation from the earlier results and includes a summary of the results.

A.2

SUMMARY

OF RESULTS/CONCLUSIONS The specific results are summarized in Table A.1 below. The Type A contribution to LERF is defined as the contribution from Class 3b.

The detailed calculations performed to support this report were of a level of mathematical significance necessary to calculate the results recorded. However, the tables and illustrational calculation steps presented may present rounded values to support readability.

RSC 07-05 A. I Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table A. 1 Summary of Risk Impact on Extending Type A 1LRT Test Frequency Risk Impact for 3-years Risk Impact for 10-years Risk Impact tor 15-years Risk Impact for 16-years (baseline)

(current requirement)

Total integrated risk (person-rem/yr) 12.177 12.183 12.187 12.188 Type A testing risk (person-rem/yr) combination of 3a and 3b 2.80E-3 9.34E-3 1.40E-2 1.49E-2 7% total risk (Type A / total) 0.023%

0.077%

0.115%

0.123%

Type A LERF (Class 3b) (per year)

Taken from Table A. 16 2.01 E-8 6.68E-8 1.001E-7 1.07E-7 Conditional Containment Failure Probability (CCFP)

Taken from Table A. 17 1.57E-I 1.64E-I 1.68E-I 1.69E-I Changes due to extension from 15 years (cUrrent)

A Risk from current (Person-rem/yr) 8.58E-4

% Increase from current (A Risk / Total Risk) 0.007%

A LERF from current (per year) 6.68E-9 A CCFP from current 0.09%

RSC 07-05 A.22 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table A.I (continued)

Summary of Risk Impact on Extending Type A ILRT Test Frequency Changes due to extension from 10 years A Risk from 10-year (Person-rem/yr) 5.15E-3

% Increase from 10-year (A Risk / Total Risk) 0.042%

A LERF from 10-year (per year) 4.02E-8 A CCFP from 10-year 0.55%

Changes due to extension from 3 years (baseline)

A Risk from baseline

,Person-rem/yr)

1. 12E-2

% Increase from baseline (A Risk / Total Risk) 0.092%

A LERF from baseline (per year) 8.69E-8 A CCFP from baseline 1.18%

RSC 07-05 A.3 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Based on the analysis and available data the following is stated:

" The person-ren/year increase in risk contribution from extending the ILRT test frequency frorn the current once-per-fifteen-year interval to once-per-sixteen years is 0.000858 person-rem/year.

The risk increase in LERF from extending the ILRT test frequency from the current once-per-15-year interval to once-per-16 years is 6.68E-9/yr.

The change in conditional containment failure probability (CCFP) from the current once-per-15-year interval to once-per-16 years is 0.09%.

The change in Type A test frequency from once-per-fifteen-years to once-per-sixteen-years increases the risk impact on the total integrated plant risk by only 0.007%. Also, the change in Type A test frequency from the original three-per-ten-years to once-per-sixteen-years increases the risk only 0.092%. Therefore, the risk impact when compared to other severe accident risks is negligible.

Reg. Guide 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Reg. Guide 1.174 defines very small changes in risk as resulting in increases of CDF below 10 6/yr and increases in LERF below 107,yr. Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A ILRT test interval from a once-per-fifteen-years to a once per-sixteen-years is 6.68E-9/yr. Guidance in Reg. Guide 1.174 defines very small changes in LERF as below 10-7/yr, increasing the ILRT interval from 15 to 16 years is therefore considered non-risk significant. The value is below this guidance indicating that the change is not risk significant. In addition, the change in LERF resulting from a change in the Type A ILRT test interval from a three-per-ten-years to a once per-sixteen-years is 8.69E-8/yr, and is also below the guidance.

R.G. 1.174 also encourages the use of risk analysis techniques to help ensure and show that the proposed change is consistent with the defense-in-depth philosophy. Consistency with defense-in-depth philosophy is maintained by demonstrating that the balance is preserved among prevention of core damage, prevention of containment failure, and consequence mitigation. The change in conditional containment failure probability was estimated to be 0.09% for the proposed change and 1.18% for the cumulative change of going from a test interval of 3 in 10 years to I in 16 years. These changes are small and demonstrate that the defense-in-depth philosophy is maintained.

RSC 07-05 A.4 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach A.3 DESIGN INPUTS The inputs for this calculation are similar to the information in the baseline analysis. The only change is that the input information includes not only internal event initiators but fire and seismic. The inclusion of these initiating events comprises the SONGS Case 2 results. The plant damage states are summarized in Table A.2 (taken from References 2 and 3)).

Table A.2 SONGS Plant Damage States Including Fire and Seismic Plant Damage Representative Sequence Frequency State

(/yr)

PDS I Transient with loss of secondary heat removal 1.97E-05 Transient with loss of secondary heat removal, and loss of containment spray PDS 2 recirculation 1.07E-07 Transient with loss of secondary heat removal, and loss of containment heat PDS 3 removal 4.13E-07 PDS 4 Transient/SSL with loss of HPSI in recirculation 2.) IE-07 Transient/SSL with loss of HPSI in recirculation. and loss of containment heat PDS 5 removal 7.57E-14 Transient/SSL with loss of HPSI in recirculation. loss of secondary heat PSD 6 removal, and loss of containment spray recirculation

2. 17E-06 Transient with loss of H-PSILPSI injection and loss of containment heat PDS 7 removal 5.17E-06 Transient with loss of HPSI/LPSI injection, loss of secondary heat removal PDS 8 and loss of containment heat removal 1.55E-06 PDS 9 Small LOCA with loss of containment spray recirculation 6.40E-09 PDS 10 Small LOCA with loss of containment heat removal 1.16E-07 PDS 11 Small LOCA with loss of secondary heat removal 4.29E-07 PDS 12 Small LOCA with loss of HIPSI recirculation 0.OOE+00 RSC 07-05 A.5 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table A.2 (Continued)

SONGS Plant Damage States Including Fire and Seismic Plant Damage Representative Sequence Frequency State

(/yr)

Small LOCA with loss of HIPSI recirculation, and loss of containment spray PDS 13 recirculation 1.34E-05 Small LOCA with loss of IIPSI recirculation. and loss of containment heat PDS 14 removal 5.77E-08 Small LOCA with loss of HPSI recirculation and loss of secondary heat PDS 15 removal 2.91E-07 Small LOCA with loss of I-PSI recirculation. loss of secondary heat removal, PDS 16 and loss of containment spray recirculation 1.31E-10 Small LOCA with loss of I-PSI recirculation. loss of secondary heat removal,.

PDS 17 and loss of containment heat removal 9.93E-12 PDS 18 Small LOCA with loss of HPSIILPSI injection 1.69E-07 PDS 19 Large/medium LOCA with loss of core heat removal

).34E-07 Large/medium LOCA with loss of core heat removal, and loss of containment P.DS 20 spray recirculation 1.86E-10 PDS 21 Large/medium LOCA with loss of core and containment heat removal 3.41E-09 PDS 22 Large/medium LOCA with loss of -FPSI recirculation 1.49E-07 Large/medium LOCA with loss of 1-PSI recirculation, and loss of PDS 23 containment spray recirculation 3.09E-10 Large/medium LOCA with loss of I-PSI recirculation, and loss of PDS 24 containment heat removal 8.92E-10 PDS 25 Large/medium LOCA with loss of HPSI/LPSI injection 1.33E-10 PDS 26 TransientlLOCA with loss of containment isolation and heat removal 9.79E-08 PDS 27 Interfacing system LOCA (ISLOCA) initiating event 4.79E-08 Steam generator tube rupture (SGTR) initiating event, no stuck open relief PDS 28 valve (SORV) 6.32E-08 Steam generator tube rupture (SGTR) initiating event, with stuck open relief PDS 29 valve (SORV) 5.07E-08 RSC 07-05 A.6 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the N-EI Approach Table A-2 (Continued)

SONGS Plant Damage States Including Fire and Seismic Plant Damage Representative Sequence Frequency State

(/yr)

TOTAL 4.448E-05 The release category dose information is presented in Table A.3.

Table A.3 Release Category Radionuclide Fraction Release Representative Frequency Category Sequence

(/yr)

Noble Gas' Iodine' Cesium1 Telluriuri Strontium' IC-I (S) 3.76E-05 NA2 NA NA NA NA B3 SGTR-20 1.46E-06 2.OE-3 4.OE-5 3.OE-5 3.OE-5 2.OE-6 D 4 SGTR2-66 5.07E-08 6.4E-I 3.4E-1 1.6E-1 2.7E-1 5.0E-3 G5 SBO-17 4.90E-07 3.1E-1 3.2E-2 3.2E-2 5.0E-2 1.0E-3 L 6 LOP-48 4.OOE-06 2.9E-1 3.OE-3 3.OE-4 1.0E-4 2.0E-6 T 7 VSEQ-2 4.79E-08 1.0E+0 9.OE-1 8.9E-I 8.6E-I 1.3E-2 W8 MLO-20 8.28E-07 2.?E-3 2.OE-5 2.0E-5 2.0E-5 6.OE-6 L. Contributing fission product groups are discussed in Reference 4.

2. Release fractions not necessary for this calculation.

3.

4.

5.

6.

7.

8.

Release category B is defined by containment bypassed with less than 0.1% of volatiles released.

Release category D is defined by containment bypassed with up to 10% of volatiles released.

Release category G is defined by early or isolation failure, containment failure prior to or at vessel failure with up to 10% of volatiles released.

Release category L is defined by late containment failure with up to 1% of volatiles released.

Release category T is defined by containment bypassed with greater than 10% of volatiles released.

Release category W is defined by late containment failure with more than 10% of volatiles released.

RSC 07-05 A.7 Printed: 07/25/2007

Evahlation of Risk Significance of ILRT Extension Based on the NEI Approach A.4 CALCULATIONS Applying the methodology presented in Section 4.1, the following calculation steps were performed:

1. Map the SONGS release categolies into the 8 release classes defined by the EPRI Report5.

2. Calculate the Type A leakage estimate to define the analysis baseline.

3. Calculate the Type A leakage estimate to address the current inspection frequency.

4.

Modify the Type A leakage estimates to address extension of the Type A test interval.

5. Calculate increase in risk due to extending Type A inspection intervals.

6.

Estimate the change in LERF due to the Type A testing.

7.

Estimate the change in conditional containment failure probability due to the Type A testing.

A.5 SUPPORTING CALCULATIONS Step 1: Map the Level 3 release categories into the 8 release classes defined by the EPRI Report EPRI Report TR-104285 (Reference 5) defines eight (8) release classes as presented in Table A.4.

RSC 07-05 A.8 Printed: 07>`25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table A.4 Containment Failure Classifications (from Reference 5)

Failure Classification Description Interpretation for Assigning SONGS Release Category Containment remains intact with Intact containment bins containment initially isolated 1

Dependent failure modes or common Isolation faults that are related to a loss of cause failures power or other isolation failure mode that is not a direct failure of an isolation component 3

Independent containment isolation Isolation failures identified by Type A testing failures due to Type A related failures 4

Independent containment isolation Isolation failures identified by Type B testing failures due to Type B related failures 5

Independent containment isolation Isolation failures identified by Type C testing failures due to Type C related failures 6

Other penetration failures Other faults not previously identified 7

Induced by severe accident phenomena Early containment failure sequences as a result of hydrogen burn or other early phenomena 8

Bypass Bypass sequence or SGTR Table A.5 presents the SONGS release category mapping for these eight accident classes.

Person-remn per year is the product of the frequency and the person-rem.

RSC 07-05 A.9 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table A.5 Grouping to EPRI SONGS PRA Release Category Classes (as described in Reference 5)

Person-Class Description Release Category Frequency Person-Remn Rem!yr I

No containment failure IC-I (S) 3.76E-5 1.70E+2 6.40E-3 2

Large containment None E

isolation failures Salisolation failuresNo 3a Small isolation failures None Not 0.OOE+0 (liner breach) addressed 3b Large isolation failures None Not 0.OOE+0 (liner breach) addressed Small isolation failures -

None failure to seal (type B)

Small isolation failures -

None failure to seal (type C)

Containment isolation 6

failures (dependent failure, G

4.90E-7 2.06E+6 2 1.01E+0 personnel errors)

Severe accident 7

phenomena induced failure L, W 4.83E-6 1.62E+6 2 7.82E+O (early and late) 8 Containment bypass B, D, T 1.56E-6 2.14E+6 2 3.34E+0 Total 4.45E-5 12.174

1. E represents a probabilistically insignificant value.

2. The value presented represents an average of the contributing release categories and is developed in Appendix B.

Step 2: Calculate the Type A leakage estimate to deflne the analysis baseline (3 year test interval)

As displayed in Table A.5 the SONGS PRA did not identify any release categories specifically associated with EPRI Classes 3, 4, or 5. Therefore each of these classes must be evaluated for applicability to this study.

RSC 07-05 A. 10 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Class 3:

Containment failures in this class are due to leaks such as liner breaches that could only be detected by performing a Type A ILRT. In order to determine the impact of the extended testing interval, the probability of Type A leakage must be calculated.

In order to better assess the range of possible leakage rates, the Class 3 calculation is divided into two classes. Class 3a is defined as a small liner breach and Class 3b is defined as a large liner breach.

This division is consistent with the NEI guidance6 and the previously approved methodology (References 7, 8 and 9). The calculation of Class 3a and Class 3b probabilities is presented below.

Calculation of Class 3a Probability The data presented earlier from NUREG-1493' 0 and data presented in Reference 6 is used to calculate the probability that a liner leak will be small (Class 3a) as done earlier. Using the data a mean estimate for the probability of leakage is determined for Class 3a as shown in Equation 1.

5 PC,-

- 0.0275 (eq. 1) 182 This probability, however, is based on three tests over a 10-year period and not the one per fifteen-year frequency currently employed at SONGS (Reference 11). The probability (0.0275) must be adjusted to reflect this difference and is adjusted in step 3 of this calculation.

Multiplying the CDF times the probability of a Class 3a leak develops the Class 3a frequency contribution in accordance with guidance provided in Reference 6. This is conservative since part of the CDF already includes LERF sequences. The CDF for SONGS is 4.448E-5/yr as presented in Table A.2.

Therefore the frequency of a Class 3a failure is calculated as:

FREQciass3a = PROB,,,, x CDF = 0.0275 x 4.448E-5E-5/yr = { 1.222E-6/yr}

(eq. 2)

Calculation of Class 3b Probability To calculate the probability that a liner leak will be large (Class 3b) use was made of the data presented in the calculation of Class 3a. Of the events identified in NUZREG-1493 (Reference 10), the largest reported leak rate from those 144 tests was 21 times the allowable leakage rate (La). Since 21 La, does not constitute a large release, no large releases have occurred based on the 144 ILRTs reported in NUREG-1493.

The additional data point was also not considered to constitute a large release.

To estimate the failure probability given that no failures have occurred, the guidance provided in Reference 6 suggests the use of a non-informative prior. This approach essentially updates a uniform distribution (no bias) with the available evidence (data) to provide a better estimation of an event. A beta distribution is typically used for the uniform prior with the parameters oc=0.5 RSC 07-05 A. 11 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach and 3=1, This is then combined with the existing data (no Class 3b events, 182 tests) using Equation 3.

7 + a*

0 +0.5 0.5

PCl,

=-

-.. 0.00273 (eq. 3)

N+f6 182+1 183 where: N is the number of tests, n is the number of events (faults) of interest, a, 03 are the parameters of the non-informative prior distribution. From this solution, the frequency for Class 3b is generated using Equation 4 and is adjusted appropriately in step 3.

FREQCIa:,

3 b = PROB,,,-,, x CDF = 0.00273 x 4.448E-5/yr = { 1.215E-7/yr}

(eq. 4)

Class 4:

This group consists of all core damage accident accidents for which a failure-to-seal containment isolation failure of Type B test components occurs. By definition, these failures are dependent on Type B testing, and Type A testing will not impact the probability. Therefore this group is not evaluated any further, consistent with the approved methodology.

Class 5:

This group consists of all core damage accident accidents for which a failure-to-seal containment isolation failure of Type C test components occurs. By definition, these failures are dependent on Type C testing, and Type A testing will not impact the probability. Therefore this group is not evaluated any further, consistent with the approved methodology.

Class 6:

The Class 6 group is comprised of isolation faults that occur as a result of the accident sequence progression. The leakage rate is not considered large by the PRA definition and therefore it is placed into Class 6 to represent a small isolation failure and identified in Table A.5 as Class 6.

FREQcias 6 = 4.90E-7/yr (eq. 5)

Class 1:

Although the frequency of this class is not directly impacted by Type A testing, the PRA did not model Class 3 failures, and the frequency for Class I should be reduced by the estimated frequencies in the new Class 3a and Class 3b in order to preserve the total CDF. The revised Class 1 frequency is therefore:

FREQejassi = FREQciassi - (FREQclass3a + FREQclass3b)

(eq. 6)

FREQelassI = 3.76E-5/yr - (1.22E-6/yr + 1.215E-7/yr) = {3.626E-5/yr}

(eq. 7)

RSC 07-05 A. 12 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Class 2:

The SONGS PRA did not identify any contribution to this group above the quantification truncation.

Class 7:

The frequency of Class 7 is the sum of those release categories identified in Table A.5 as Class 7.

FREQcl.as 7 = 4.828E-6/yr-(eq. 8)

Class 8:

The frequency of Class 8 is the sum of those release categories identified in Table A.5 as Class 8.

FREQcta,,-ss= 1.559E-6/yr (eq. 9)

Table A.6 summarizes the above information by the EPRI defined classes.

This table also presents dose exposures calculated using the methodology described in Appendix B. For Class 1, 3a and 3b, the person-rem is developed based on the design basis assessment of the intact containment. The Class 3a and 3b doses are represented as 1OL, and 3 5 L, respectively.

Table A.6 also presents the person-rem frequency data determined by multiplying the failure class frequency by the corresponding exposure.

RSC 07-05 A. 13 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table A.6 Baseline Risk Profile Class Description Frequency Person-rem Person-rem Person-rem

(/yr)

(calculated)'

(from L,,

(/yr) factors)

No containment failure 3.63E-5 1.70E+22-6.17E-3 Large containment 3

isolation failures S alisolation failures 3a Small isolation failures 1.22E-6 1.70E+34 2.OSE-3 (liner breach) 3b Large isolation failures 1.22E-7 5.95E+3 7.23E-4 (liner breach)

Small isolation failures -

failure to seal (type B)

Small isolation failures -

failure to seal (type C)

Containment isolation 6

failures (dependent failure, 4.90E-7 2.06E+66 I.OIE+O personnel errors)

Severe accident 7

phenomena induced failure 4.83E-6 1.62E+6 6 7.82E+O (early and late) 8 Containment bypass 1.56E-6 2.14E+66 3.34E+O Total 4.45E-5 12.177 I.

2.

3.

4.

5.

6.

From Table A.5 using the method presented in Appendix B.

1 times L, dose value calculated in Appendix B.

E represents a probabilistically insignificant value.

10 times L,.

35 times La.

The value presented represents an average of the contributing release categories and is developed in Appendix B.

RSC 07-05 A. 14 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach The percent risk contribution due to Type A testing is as follows:

%Risk,,s, =[( Class3aB,,SE + Class3bBAs) / TotalBASE] X 100 (Eq. 10)

Where:

Class3a,,SE = Class 3a person-rem/year = 2.08E-3 person-rem/year Class3bBASE = Class 3b person-remi/year = 7.23E-4 person-remryear TotalBASE = total person-rem year for baseline interval = 12.177 person-rem/year (Table A.6)

%Risk,,,,, = [(2.0SE-3 + 7.23E-4) / 12.177] x 100 = {O.023%}

(Eq. 11)

Step 3. Calculate the Type A leakage estimate to address the current inspection interval The current surveillance testing requirements as proposed in NEI 94-01 (Reference 12) for Type A testing and allowed by 10 CFR 50, Appendix J is at least once per 10 years based on an acceptable performance history (defined as two consecutive periodic Type A tests at least 24 months apart in which the calculated performance leakage was less than 1.OL,).

According to References 6 and 10, extending the Type A ILRT interval from 3-in-10 years to I-in-10 years will increase the average time that a leak detectable only by an ILRT goes undetected from IS to 60 months. Multiplying the testing interval by 0.5 and multiplying by 12 to convert from "years" to "months" calculates the average time for an undetected condition to exist.

The increase for a 10-yr ILRT interval is the ratio of the average time for a failure to detect for the increased ILRT test interval (from 18 months to 60 months) multiplied by the existing Class 3a probability as shown in Equation 12.

pc,.,.,.

3 (10y) = 0.0275<(i-8) = 0.0916 (eq. 121)

A similar calculation is performed for the Class 3b probability as presented in Equation 13.

Pcmss,3b(101y) = 0.00273x

= 0.0091 (eq. 13)

Risk Impact due to 10-year test interval Based on the previously approved methodology (References 7, 8 and 9) and the NEI guidance (Reference 6), the increased probability of not detecting excessive leakage due to Type A tests directly impacts the frequency of the Class 3 sequences.

Consistent with Reference 6 the risk contribution is determined by multiplying the Class 3 accident frequency by the increase in the probability of leakage.

Additionally the Class I frequency is adjusted to maintain the overall core damage frequency constant. The results of this calculation are presented in Table A.7 below.

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Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table A.7 Risk Profile for Once in Ten Year Testing Class Description Frequency (/yr)

Person-rem Person-rem (/yr)

I No containment failurel 3.3 IE-5 1.70E+2 5.63E-3 2

Large containment isolation 3

failures 3a Small isolation failures (liner 4.07E-6 1.70E+3 6.93E-3 breach)

Large isolation failures (liner 4.05E-7 5.95E+3 2.41E-3 breach)

Small isolation failures - failure to seal (type B)

Small isolation failures - failure F

to seal (type C)

Containment isolation failures 4.90E-7

'2.06E+64 1.01E+0 6

(dependent failure, personnel errors)

Severe accident phenomena 4.83E-6 1.62E+6 4 7.82E+O induced failure (early and late) 8 Containment bypass 1.56E-6 2.14E+64 3.34E+O Total 4.45E-5

12. 183

1. The IPE frequency of Class I has been reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF.

2. From Table A.6.

3. e represents a probabilistically insignificant value.

4. The value presented represents an average of the contributing release categories and is developed in Appendix B.

Using the same methods as for the baseline, and the data in Table A.7 the percent risk contribution due to Type A testing is as follows:

%Riskio =[(Class3a,0 + Class3b,0) / Total,0 ] x 100 (eq. 14)

Where:

Class3ao = Class 3a person-rem/year = 6.93E-3 person-rern/year Class3bo = Class 3b person-rem/year = 2.41E-3 person-rem/year Total,, = total person-rem year for current 10-year interval = 12.183 person-rem/year (Table A.7)

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Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach

%Risko = [(6.93E-3 + 2.41E-3) / 12.183] x 100 = {0.077%}

(eq. 15)

The percent risk increase (A%Risk,,) due to a ten-year ILRT over the baseline case is as follows:

ARisk,0 = [(Total,, - Total.,AS) / TotalBAsE] x 100.0 (eq. 16)

Where:

TotaIBlS, = total person-rem/year for baseline interval = 12.177 person-rem/year (Table A.6)

Total,, = total person-rem/year for 10-year interval = 12.183 person-remlryear (Table A.7)

A%Risk,, = [(12.183 - 12.177) / 12.177] x 100.0 = {0.049%}

(eq. 17)

Step 3b. Calculate the Type A leakage estimate to address the current extension uinspection interval The inspection interval for SONGS Unit 3, is once per 15 years and is supported by the analysis documented in References 4 and 13. In this case, the average time that a leak detectable only by an ILRT test goes undetected increases to 90 months (0.5 x 15 x 12). For a 15-yr-test interval, the result is the ratio (90/18) of the exposure times as was the case for the 10 year case.

Increasing the ILRT test interval from 3 years to 15 years results in a proportional increase in the overall probability of leakage.

The approach for developing the risk contribution for a 15-year interval is the same as that for the 10-year interval. The increase for a 15-yr ILRT interval is the ratio of the average time for a failure to detect for the increased ILRT test interval (from 18 months to 90 months) multiplied by the existing Class 3a probability as shown in Equation 18.

PCI,..., (15y) = 0.0275x<

=0.1375 (eq. 18)

A similar calculation is performed for the Class 3b probability as presented in Equation 19.

Pcl,.sb(15 y) =0.00273X(--

= 0.0137 (eq. 19) y18)

As stated for the 10-year case, the increased probability of not detecting excessive leakage due to Type A tests directly impacts the frequency of the Class 3 sequences.

The increased risk contribution is determined by multiplying the Class 3 accident frequency by the increase in the probability of leakage.

Additionally the Class 1 frequency is adjusted to maintain the overall core damage frequency constant.

The results of this calculation are presented in Table A.8 below.

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Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table A.8 Risk Profile for Once in Fifteen Year Testing Class Description Frequency (/yr)

Person-remi2 Person-rem (/yr)

I No containment failurel 3.09E-5 1.70E+2 5.25E-3 Large containment isolation failures 3a Small isolation failures (liner 6.11E-6 1.70E+3 1.04E-2 breach) 3b Large isolation failures (liner 6.08E-7 5.95E+3 3.62E-3 breach)

Small isolation failures - failure to seal (type B)

Small isolation failures - failure to seal (type C)

Containment isolation failures 4.90E-7 2.06E+64 1.01E+0 6

(dependent failure, personnel errors)

Severe accident phenomena 4.83E-6 1.62E+64 7.82E+O induced failure (early and late) 8 Containment bypass 1.56E-6 2.14E+64 3.34E+O Total 4.45E-5 12.187

1. The IPE frequency of Class I has been reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF.

2. From Table A.6.

3. E represents a probabilistically insignificant value.

4. The value presented represents an average of the contributing release categories and is developed in Appendix B.

Using the same methods as was described earlier, and the data in Table A.8, the percent risk contribution due to Type A testing is as follows:

%Risk,, =[( Class3a,, + Class3b,5) / Total,] x 100 (eq. 20)

Where:

Class3a,, = Class 3a person-rem/year = 1.04E-2 person-rem/year Class3b,, = Class 3b person-rem/year = 3.62E-3 person-rem/year Total,, = total person-rem year for 15-year interval = 12.187 person-rem/year (Table A.8)

%Risk,, = [(1.04E-2 + 3.62E-3)/ 12.187] x 100 = {0.115%}

(eq. 21)

RSC 07-05 A. 18 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach The percent risk increase (A%Risk,5) due to a fifteen-year ILRT over the baseline case is as foillows:

A%Risk,5 = [(Total, - Total,,,,) / Totalling] x 100.0 (eq. 22)

Where:

Total,,,, = total person-rem/year for baseline (3 per 10 years) interval = 12.177 person-rern/year (Table A.6)

Total, = total person-rem/year for 15-year interval = 12.187 person-rerndyear (Table A.8)

A%Risk,, = [(12.187 177) / 122.177] x 100.0 = {0.085%}

(eq. 23)

The percent risk increase (A%Risk,5) due to a fifteen-year ILRT over the 10-year case is as follows:

A%Risk,, = [(Total,5 - Total,,) / Total,0 ] x 100.0 (eq. 24)

Where:

Total, = total person-remlyear for 10-year interval = 12.183 person-rem/year (Table A.7)

Total, = total person-remi/year for 15-year interval = 12.187 person-rem/year (Table A.8)

A%Risk,5 = [(12.187 - 12.183) / 12.183] x 100.0 = {0.035%}

(eq. 25)

Step 4: Calculate the Type A leakage estimate to address extended inspection intervals If the test interval is extended to I per 16 years, the average time that a leak detectable only by an ILRT test goes undetected increases to 96 months (0.5 x 16 x 12). For a 16-yr-test interval, the result is the ratio (96/18) of the exposure times as was the case for the 10 year case. Thus, increasing the ILRT test interval from 3 years to 16 years results in a proportional increase in the overall probability of leakage.

The approach for developing the risk contribution for a 16-year interval is the same as that for the 10-year interval. The increase for a 16-yr ILRT interval is the ratio of the average time for a failure to detect for the increased ILRT test interval (from 18 months to 96 months) multiplied by the existing Class 3a probability as shown in Equation 26.

Pc,s.,.3n(16 y) =0.0275x ('-6 0.1465 (eq. 26)

ý18)

A similar calculation is performed for the Class 3b probability as presented in Equation 27.

pc,.,3b,(16y) = 0.002731 X ) = 0.01457 (eq. 27)

RSC 07-05 A. 19 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach As stated for the 10-year case, the increased probability of not detecting excessive leakage due to Type A tests directly impacts the frequency of the Class 3 sequences.

The increased risk contribution is determined by multiplying the Class 3 accident frequency by the increase in the probability of leakage. Additionally the Class 1 frequency is adjusted to maintain the overall core damage frequency constant.

The results of this calculation are presented in Table A.9 below.

Table A.9 for Once in Sixteen Risk Profile Year Testing Class Description Frequency (/yr)

Person-remi Person-rem (/yr)

I No containment failurel 3.04E-5 1.70E+2 5.18E-3 Large containment isolation 3

failures 3a Small isolation failures (liner 6.52E-6 1.70E+3 1.1 IE-2 breach) 3b Large isolation failures (liner 6.48E-7 5.95E+3 3.86E-3 breach)

Small isolation failures - failure to seal (type B)

Small isolation failures - failure to seal (type C)

Containment isolation failures 4.90E-7 2.06E+64 1.01E+0 6

(dependent failure, personnel errors)

Severe accident phenomena 4.83E-6 1.62E+6 4 7.82E+O induced failure (early and late) 8 Containment bypass 1.56E-6 2.14E+6 4 3.34E+O Total 4.45E-5 12.188

1. The IPE frequency of Class I has been reduced by the frequency of Class 3a and total CDF.

2. From Table A.6.

3. c represents a probabilistically insignificant value.

4. The value presented represents an average of the contributing release categories Class 3b in order to preserve Using the same methods as was described earlier, and the data in Table A.9, the percent risk contribution due to Type A testing is as follows:

RSC 07-05 A.20 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach

%Riskb =[( Class3a,, + Class3b,o) / Total,] x 100 (eq. 28)

Where:

Class3a,, = Class 3a person-rem/year = 1.1 1E-2 person-rem/year Class3b,, = Class 3b person-rem/year = 3.86E-3 person-remlyear Total,, = total person-rem year for 16-year interval = 12.188 person-rein/year (Table A.9)

%Risk,, = [(I.1 IE-2 + 3.86E-3) / 12.188] x 100 = {0.123%}

(eq. 29)

The percent risk increase (A%Risk,,) due to a sixteen-year ILRT over the baseline case is as follows:

A%Risk,0 = [(Total,, - TotalBASE) / TotalBAS,] x 100.0 (eq. 30)

Where:

TotalBASE = total person-remi/year for baseline (3 per 10 years) interval = 12.177 person-rem/year (Table A.6)

Total,6 = total person-rem/year for 16-year interval = 12.188 person-rem/year (Table A.9)

A%Risk, 6 = [(12.188 - 12.177) / 12.177] x 100.0 = {0.092%}

(eq. 31)

The percent risk increase (A%Risk,,) due to a sixteen-year ILRT over the 10-year case is as follows:

A%Risk,, = [(Total,6 - Total,,) / Total,] x 100.0 (eq. 32)

Where:

Total,, = total person-rem/year for baseline 10-year interval = 12.183 person-rem/year (Table A.7)

Total, = total person-rem/year for 16-year interval = 12.188 person-rem/year (Table A.9)

A%Risk,, = [(12.188 - 12.183) / 12.183] x 100.0 = {0.042%}

(eq. 33)

Step 5: Calculate increase in risk due to extending Type A inspection intervals Based on the guidance in Reference 6, the percent increase in the total integrated plant risk for these accident sequences is computed as follows:

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Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach

%Total15-1J6 = [(Total., - Total,) / Total5] x 100 (eq. 34)

Where:

Total *5 = total person-remn/year for 15-year interval = 12.187 person-rem/year (Table A.8)

Total i, = total person-remi/year for 16-year interval = 12.188 person-remn/year (Table A.9)

% Total 15-16 = [(12.188-12.1S7)/ 12.187] x 100 = {0.007%}

(eq. 35)

Step 6: Calculate the change in risk in te17rs of large early release frequency (LERF)

The risk impact associated with extending the ILRT interval involves the potential that a core damage event that normally would result in only a small radioactive release from containment could in fact result in a larger release due to failure to detect a pre-existing leak during the relaxation period.

From References 4, 7, 8 and 9, the Class 3a dose is assumed to be 10 times the allowable intact containment leakage, La (or 1,700 person-remn) and the Class 3b dose is assumed to be 35 times L, (or 5,950 person-rem).

The dose equivalent for allowable leakage (L,) is developed in Appendix B.

This compares to a historical observed average of twice L,.

Therefore, the estimate is somewhat conservative.

Based on the NEI guidance (Reference 6) and the previously approved methodology (References 7, 8 and 9), only Class 3 sequences have the potential to result in large releases if a pre-existing leak were present.

Class I sequences are not considered as potential large release pathways because for these sequences the containment remains intact. Therefore, the containment leak rate is expected to be small (less than 2L,).

A larger leak rate would imply an impaired containment, such as Classes 2, 3. 6 and 7.

Late releases are excluded regardless of the size of the leak because late releases are, by definition, not a LERF event. At the same time, sequences in the SONGS PRA (References 2 and 3) that result in large early releases, are not impacted because a LERF will occur regardless of the presence of a pre-existing leak.

Therefore, the change in the frequency of Class 3b sequences is used as the increase in LERF for SONGS, and the change in LERF can be determined by the calculated differences. Reference 6 also identifies that Class 3b is considered to be the contributor to LERF. Table A.10 summarizes the results of the LERF evaluation assuming that Type 3b is indicative of a LERF sequence.

RSC 07-05 A.22 Printed: 07/25/ 2 007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table A.10 Impact on LERF due to Extended Type A Testing Intervals ILRT Inspection 3 Years (baseline) 10 Years 15 Years 16 Years Interval Class 3b (Type A 1.22E-7/yr 4.05E-7/yr 6.OSE-7/iyr 6.48E-7/yr LERF)

ALERF (3 year 2.84E-7/yr 4.86E-7/yr 5.27E-7/yr baseline)

ALERF (10 year 2.03E-7/yr 2.43E-7/yr baseline)

ALERF ( 15 year 4.05E-8/yr extension)

Reg. Guide 1.174 (Reference 14) provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Reg. Guide 1.174 defines very small changes in risk as resulting in increases of CDF below 1.OE-6/yr and increases in LERF below 1.OE-7/yr. Since the ILRT does not impact CDF, the relevant metric is LERF. Calculating the increase in LERF requires determining the impact of the ILRT interval on the leakage probability.

Since guidance in Reg. Guide 1.174 defines very small changes in LERF as below 1.OE-7/yr, increasing the ILRT interval to 16 years (4.05E-8/yr) from the current interval (15 years) meets this criterion. The LERF increase when measured from the original 3-in-10-year is 5.27E-7/yr, which is above the 1.OE-7/yr screening criterion in Reg. Guide 1.174.

Reference I indicates that plants with a CDF in excess of 1.OE-5/yr may have difficulty demonstrating a change in LERF less than 1.OE-7/yr.

It further states that the analysis as embodied in the NEI approach is conservative and provides additional guidance with respect to refining the initial analysis.

The change in LERF when extending the current ILRT period from 15 to 16 years is acceptable but the increase from 3 years to 16 years exceeds the limit. Therefore, the refinements defined in the main report are again required. Four refinement steps are implemented and are described in the main report. These are:

1. Remove existing LERF contributions from the estimate for Class 3b

2. Remove accident states where the debris would be scrubbed (covered by water) such that the source term in the containment would not be sufficient to support a LERF release.

3. Evaluate sequence timing to identify and exclude intact containment sequences with accident timing beyond that defined for the "early" class.

RSC 07-0-5 A.23 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach

4. Review a combination of containment source term and timing to exclude other sequences that could not meet the timing and source term requirements for LERF.

The first two refinements were conducted using the SONGS Level 2 containment event tree model to identify the characteristics necessary for determining the status of these aspects of the analysis. The existing containment event tree model provided clear branch points for LERF, debris flooded and containment spray status in recirculation such that a set of logical rules could be applied to the model to obtain the necessary results in terms of CDF. The analysis is summarized in Reference 15 and the results illustrated graphically in Figure A.1.

RSC 07-05 A.24 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach FROM CEI LERF FLcODED SPRAYS S

SPEIIAL iS DEB.

YONTAIEAMEIHS EASE THERE FL"L'DED SPRAY FT'R LERF AFTER IN FLC-CADE[,

VESSEL RECIRI(

FAILIUFRE

('A.2.FL-OODED IhW Bifell, elil1,l RIIII diM-11l Biull, d*.ill-d Bill, lefilned

SPRAYS, 5.286,00cES I

FLOODIED 7.3 A2V.RE TF PRA 12 I..

MY;AU N- -[EFE

-1.374-0-ER 5.201;,006 2.097l, 00f 1,5571,.005 2.078,605R 1.557,-005 NOT FLOODED 3..36,,-005

.1.443 e-005 HO SPRAY 2.078,-005 LFERF 5

1 6.959--007 6i.159e-007 C:-W.1.11 IPRA301.SCE-OSIANFRU.FL2.A?.M CA-24-FLOIDFEIS L::,eI S-, 1.:

U0IsSIV El,1101. 2005 Ls Upl~ipl[I.: Tuee$.1y. Ho-oile, 28. 2EAS MORI2IAP IA' Lceesed W: L-0Scl.....

M..,Ve~I Figure A. 1. Quantified Source Term Category Diagram for SONGS Case 2 (fiom Reference 15)

RSC 07-05 A.25 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach The figure illustrates how the total CDF frequency is subdivided based on the three identified criterion. From the solution presented in the figure a table of results is obtained and reproduced in Table A.I1.

Table A. I1 Source Term Outcomes Source Term Outcome Frequency (/yr)

Description STC Refinement Outcome I Non-LERF sequence with the 5.29E-6 debris flooded and containment sprays functioning STC Refinement Outcome 2 Non-LERF sequence with the

2. 1 E6 debris flooded and the containment sprays do not function after recirculation STC Refinement Outcome 3 Non-LERF sequence without water 1.56E-5 pools covering the debris and containment sprays functioning STC Refinement Outcome 4 Non-LERF sequence without water 2.08E-5 pools covering the debris and containment sprays do not function STC Refinement Outcome 5 6.96E-7 LERF sequence Only outcome 4 contributes to the potential for a Type A LERF. This value is then utilized to calculate the LERF contribution from Class 3b frequency by the following equation:

FREQclss 3b = PROB,,,,,, x Adjusted CDF = 0.00273 x 2.08E-5/yr = 5.69E-8/yr (eq. 36)

This can then be extrapolated using the methods presented earlier to determine the 10-year, 15-year and 16-year contributions and to generate adjusted LERF values as presented in Table A.12.

RSC 07-05 A.26 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table A. 12 Class 3b Contributions Using Adjusted CDF Test Interval Frequency (/yr)

Delta Frequency from Prior Period

(/yr)

Baseline 5.69E-8 10-year 1.90E-7 1.33E-7 15-year (current) 2.85E-7 9.48E-8 16-year 3.04E-7 1.90E-8 Summing the last column provides the total increase from the baseline (3 years) to the proposed (16 year) interval (2.47E-7/yr). This increase is above the guidance for a small change in risk and does not address the identified conservatism in the small LOCA contribution.

A more realistic estimation of the probability of clogging given a small LOCA is provided in Reference 15 and the resulting impact is found in Table A.13. The reduction in the small LOCA contribution reduces the STC Outcome 4 to 1.83E-5/yr from 2.08E-5/yr. Substitution into the LERF calculation equations defined earlier yields the final results.

Table A. 13 Class 3b Contributions Using Adjusted CDF Test Interval Frequency (Iyr)

Delta-Frequency from Prior Period:.

(/yr)

Adjusted STC Outcome 4 1.83E-5 Baseline LERF 5.OOE-8 10-year 1.67E-7 1.17E-7 15-year (current) 2.50E-7 8.34E-8 16-year 2.67E-7 1.70E-8 Again, summing the last column provides the total increase from the baseline (3 years) to the proposed (16 year) interval (2.17E-7/yr).

This increase is not sufficiently small to meet the guidance for a small change in risk with margin.

RSC 07-05 A.27 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Steps 3 and 4 involve looking at sequences in terms of the LERF definition. LERF is defined by both release (large) and timing (early).

It is this aspect that is examined to determine if the existing analysis can be refined further.

The release categories selected in the analysis documented in the main report also apply to this study since the external event sequences would be binned in a similar manner as the internal events to ensure consistency. The only difference will be the associated frequency.

The PCS-35 and MLO-4 release category frequencies are calculated including the external events and do not contain any portion of previously addressed intact containment contributions.

The associated frequency can be removed without adjustment.

The total impact of this adjustment is provided in Table A. 14.

Table A.14 Adjustment for PCS-35 and MILO-4 LERF Contribution Release Category Sequence Number Prior to Adjustment (/yr)

Adjustment (/yr)

PCS-35 Frequency (Internal, Small 6.62E-6 1.56E-6 LOCA)

PCS-35 Frequency (External) 1.29E-7 MLO-4 Frequency (Internal) 4.78E-6 MLO-4 Frequency (External) 1.0OIE-6 Intact Containment Frequency 1.25E-5 1.56E-6 Adjusting the total intact contribution by this assessment we arrive at the new total intact contribution to LERF as shown in Table A.15.

Table A. 15 Adjustment to Address Source Term Magnitude and Timing Release Category Sequence Number Prior to Adjustment:ý(Iyr)..:.::Afte rAdj ustment (/Yr)...

Initial Intact Containment Frequency 5.75E-6 5.75E-6 (not including PCS-35 or MLO-4)

PCS-35 and MLO-4 Frequency 1.25E-5 1.56E-6 Total Intact Containment Frequency 1.83E-5 7.34E-6 RSC 07-05 A.228 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach The additional refinement reduces the contribution as shown. Using the new intact containment contribution, an adjusted value for LERF is quantified. Substitution into the LERF calculation equations defined earlier yields the final results presented in Table A. 16.

Table A. 16 Class 3b (LERF) Contributions Using Adjusted CDF Test Interval Frequency (/yr)

Delta Frequency from Prior Period

(/yr)

Adjusted Outcome 7.34E-6 Baseline LERF 2.01E-8 10-year 6.68E-8 4.68E-8 15-year (current)

I.OOE-7 3.34E-8 16-year 1.07E-7 6.68E-9 Again, summing the last column provides the total increase from the baseline (3 year) to the proposed (16 year) interval (8.69E-8/yr).

The quantified value meets the definition for an acceptably small change in risk as defined by the guidance in Reference 14.

Step 7: Calculate the change in conditional containment failure probability (CCFP)

The CCFP is defined as the probability of containment failure given the occurrence of an accident. This probability can be expressed using the following equation:

CCFP = I

.f(

)

CDF (Eq. 37)

Where ffncf) is the frequency of those sequences which result in no containment failure. This frequency is determined by summing the Class 1 and Class 3a results, and CDF is the total frequency of all core damage sequences.

Therefore the change in CCFP for this analysis is the CCFP using the results for 16 years (CCFPI) minus the CCFP using the results for 15 years (CCFP,,). This can be expressed by the following:

ACCFF1 5-,, = CCFF] 6 - CCFP15 (Eq. 38)

RSC 07-05 A.29 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Using the data previously developed the change in CCFP from the current testing interval is calculated and presented in Table A.17.

Table A.17 Impact on Conditional Containment Failure Probability due to Extended Type A Testing Intervals ILRT inspection 3 Years (baseline) 10 Years 15 Years 16 Years Interval fincf) (lyr) 3.75E-5 3.72E-5 3.70E-5 3.70E-5 fincf)/CDF 0.843 0.836 0.832 0.831 CCFP 0.157 0.164 0.168 0.169 ACCFP (3 year 6.38E-3 1.09E-2 1.18E-2 baseline)

(0.64%)

(1.09%)

(1.18%)

ACCFP (10 year) 4.55E-3 5.46E-3 (0.46%)

(0.55%)

ACCFP (15 year 9.1IE-4 current)

(0.09%)

A.6 REFERENCES

1. Letter from A. Pietrangelo, NEI, Titled: One-time extensions of containment integrated leak rate test interval - additional infornation, November 30, 2001.

2.

San Onofre Nuclear Generating Station Living PRA, SONGS 2/3, Living PRA Main Report, IPE-MR-000.

3.

Sarmanian, L., Quantification of Level 2 Release Categories for SONGS Units 2/3 Internal and External Events, November 2006.

4.

Miller, J., San Onofre Nuclear Generating Station Probabilistic Risk Assessment Evaluation of Risk Significance of ILRT Extension, Revision 0, Ricky Summitt Consulting (RSC), Inc., RSC 04-02, March 2004.

5. Gisclon, J. M., et al, Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals, Electric Power Research Institute, TR-104285, August 1994.

6. Haugh, J., et al, Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals, Revision 4, Nuclear Energy Institute (NEI), November 2001.

RSC 07-05 A.30 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NE1 Approach

7.

Summitt, R., Comanche Peak Steam Electric Station Probabilistic Safety Assessment, Evaluation of Risk Significance of ILRT Extension, RSC, Inc., RSC 01-47/R&R-PN-110, November 2001.

8. Indian Point 3 Nuclear Power Plant. "Supplemental Information Regarding P1:oposed Change to Section 6.14 of the Administrative Section of the Technical Specification",

Entergy, IPN-01-007, January 18, 2001.

9.

Evaluation of Risk Significance of ILRT Extension, Revision 2, Florida Power Corporation, F-01-0001, June 2001.

10.

Performance-Based Containment Leak-Test Program, USNRC, NJUREG-1493, July 1995.

11.

Docket No. 50-362 San Onofre Nuclear Generating Station, Unit 3, Amendment to Facility Operating License, Amendment 189, License No. NPF-15, August 24, 2005.

12.

Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, Revision 0, Nuclear Energy Institute, NEI 94-01, July 26, 1995.

13.

Summitt, R., San Onofre Nuclear Generatin.g Station Unit 3 Probabilistic Risk Assessment Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach, Revision 0, RSC, Inc., RSC 05-03, April 2005.

14. An Approach for Using Probabilistic Risk Assessment in Risk-Informed decisions on Plant-Specific Changes to the Licensing Basis, U.S. Nuclear Regulatory Commission (USNRC), Regulatory Guide 1.174, July 1998.

15. Sarmanian, L., Investigations into the Contributors of Level 2 Source Term Categories for SONGS Unit 2/3, December 2006.

RSC 07-05 A.31 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Appendix B:

Population Dose Estimates RSC 07-05 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach B.0 POPULATION DOSE ESTIMATES This section provides the basis for the dose estimates utilized in the analysis. Dose estimates for both the intact case and for impaired containments are developed.

B.I ESTIMATION OF INTACT CONTAINMENT POPULATION DOSE The ILRT evaluation technique addresses the potential for a pre-existing leak in the containment such that a release could occur when successful containment performance would be expected.

To determine the dose for such a situation, it is necessary to develop an estimate for whole body population dose given that although the core has melted, the containment is intact with safeguards functioning.

This is the typical licensing basis assessment used for judging the acceptability of the containment and addresses the current San Onofre Nuclear Generating Station (SONGS) licensing basis assumptions.

B.1.1 Methodology The overall methodology utilizes a scaling factor to estimate the person-rem dose for the population within 50-miles of SONGS. The dose rates for the exclusion area boundary (EAB) and the low population zone (LPZ) are used to define a distance scaling factor. This scaling factor is then used to estimate the dose for distances beyond the LPZ up to the 50-mile radius.

An average person-rem dose is predicted assuming a uniform distribution of radionuclides that decreases with increased distance. A uniform distribution of the surrounding population is then combined to calculate the final total dose. The analysis depends on inputs from the licensing basis analysis' to arrive at the EAB dose rate, LPZ dose rate, LPZ total man-rem dose and population data.

B.1.2 Licensing Basis Information The information contained in References 1 and 2 provide the dose rates for the EAB and LPZ and the population data. The EAB is defined as the circular area within a radius of 576 meters

(-0.36 miles) from the containment (each containment has a similar sized area).

The LPZ extends the radius to 3,140 meters (-1.95 miles). The population is collected for a distance (radius) of 50 miles from the SONGS site.

Analysis presented in Reference 1 provides the dose rates listed in Table B.1, below, for the EAB and LPZ which are derived at a time of two hours following the initiation of the release and the 30-day LPZ dose.

RSC 07-05 B. I Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table B. 1 Predicted Dose Rates taken from Reference 1 Location Dose Rate (person-rem/h r)

EAB2hT 4.347E-1 LPZ2hr 1.234E-2 LPZ 30d 8.676E-2 Reference 2 provides the population data surrounding SONGS. This data indicates that there are no persons within the EAB and approximately 1,201 persons within the LPZ.

The LPZ is defined as an area within a 1.95 mile radius of the plant site. Reference 2 also indicates that the population within a 50-mile radius is approximately 8.32E+06 if the outer areas of San Diego and Los Angeles are included.

B.1.3 Dose Scalina Factor The calculation of the necessary scaling factor is based on a relationship of dose rate and distance.

This is consistent with the Indian Point 3 submittal (Reference 3).

The scaling equation is based on a ratio of the LPZ dose to the EAB dose. The equation is shown below:

1= X * (

)C (eq. 1)

Where:

Y = LPZ dose rate X = EAB dose rate dLpz = distance for LPZ dEAB = distance for EAB c = scaling constant This equation assumes that the dose rate is decreasing in a constant manner with distance and is consistent with the Comanche Peak submittal (Reference 4). Solving for the equation yields a value for the scaling constant (C). The input data is listed below in Table B.2.

RSC 07-05 B.2 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table B.2 Calculation Parameters taken from Reference I Parameter Value (units)

X 4.347E-1 (person-rem/hr)

Y 1.234E-2 (person-remlrhr) dEAB 576 (meters) dLpz 3140 (meters)

Solving yields a value of 2.1 for the constant, C. The next step involves extrapolation of. this information to the 50-mile radius.

B.1.4 Calculation of Population Dose To calculate the population dose for the 50-mile radius, Equation 1 is employed using somewhat different parameters. The LPZ total dose is extrapolated to calculate a population dose, it is important to estimate the dose out to 50 miles (although dose rates decrease significantly with distance, the population is much greater as distance increases) in order to account for the total exposed population. Consistent with Reference 4, the value at 25 miles can be used to represent an average dose for the 50-mile distance since it is the mid point between the plant and the 50-mile radius and the dose is assumed to decrease constantly with distance.

The values provided in Table B.3 are used to solve for the dose at 25 mile (Y).

Table B.3 Calculation Parameters taken from Reference I Parameter Value (units):.:-.

Y (LPZ30d) 8.676E-2 (person-rem)

C 2.1 dLpz 1.95 (miles) [3140 meters]

d215 25 (miles)

Solving Equation I yields a value for the person-rem dose of 4.09E-4 person-rem. The value represents an average individual dose.

RSC 07-05 B.3 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach The next step is to define the effected population.

The estimated population is 8.32E+06 persons. However, it is usually assumed that 95% of the population will be evacuated prior to release such that only 5% of the population would be involved.

Given a total population estimate of approximately 8.32E+06 persons, this equates to an exposed population of 4.16E+05 persons.

The population dose dpop is then developed using this information by employing the following equation:

d,,,,, = did 0 p (eq. 2) where:

dind is the dose calculated for a single individual (4.09E-04) p is the total effected population (4.16E+05)

Solving Equation 2 yields a total population whole body dose of 1.70E+2 person-rem.

B.1.5 Sensitivity Assessment of Evacuation Assumption on Intact Containment Dose Estimation The dose estimate can be sensitive to assumptions related to evacuation effectiveness.

A parametric sensitivity analysis is performed in this appendix to determine how a change in the evacuation assumption would impact the person-rem estimates used as a risk measure.

The use of evacuation is typical for most offsite dose analyses and a value of 95% evacuation or greater is typically assumed in many cases (Reference 6 for example assumed 99.5%). However, as the evacuation increases, the person-rem dose decreases. Most of the calculated parameters utilize the intact dose indirectly in the numerator. This is because the estimated doses for Class 1, Class 3a and Class 3b are the only dynamic estimates in the risk calculation and are the only values that can be used for the change in risk.

Expanding this discussion, since Class I represents an intact containment and Classes 3a and 3b are multiples of Class I the predicted values are linked. Also, since the results are typically utilized in the numerator, the smaller the value obtained, the lower the change in risk.

To reflect the most conservative case and maximize the numerator in the risk calculations the assumption of no evacuation can be made and the estimate for intact dose updated. The dose is proportional to the available persons and since no evacuation is assumed the value is increased by a factor of 20 (0.05/1). This increases the intact dose for Class I from a value of 170 person-rem to a value of 3,400 person-rem.

A sensitivity study is performed on the baseline analysis to determine how the increased Class I exposure would impact the conclusions of the analysis. The change in the baseline exposure has no impact on the LERF value or the conditional probability of containment failure since both are based on frequency values and they are not impacted by any exposure assumptions.

Only those metrics based on exposure are impacted. In those cases the increase in risk for the impacted classes (Classes 1, 3a and 3b) will tend to highlight the impact of any change by RSC 07-05 B.4 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach increasing the baseline value that is present in the numerator of the equation. It will also tend to increase proportionally the net change between the different time periods.

The total integrated risk and Type A risk are developed using the approach defined in the main report.

Specifically, the baseline intact dose (170 person-rem) is increased to the value associated with no evacuation (3,400 person-rem). Then the person-rem values for Classes 1, 3a and 3b are revised and the analysis quantified. Table B.4 presents a comparison of the Class 1, 3a and 3b contributions for the main report and the sensitivity case.

Table B.4 Comparison of Class 1, 3a and 3b Person-Rem for Baseline and No Evacuation Sensitivity Study Parameter Main Report Assessment No Evacuation Assessment 2 Testing 3 in 10 10 year 15 year 16 year 3 in 10 10 year 15 year 16 year Interval years years Class I 4.01E-3 3.65E-3 3.40E-3 3.34E-3 8.03E-2 7.3 1E-2 6.79E-2 6.69E-2 Class 3a 1.40E-3 4.68E-3 7.02E-3 7.49E-3

2. 81E-2 9.36E-2 1.40E-1 1.50E-I Class 3b 4.89E-4 1.63E-3 2.44E-3 2.6 IE-3 9.77E-3 3216E-2 4.89E-2 5.2 1E-2 Total for al Cas 9.685 9.689 9.692 9.693 9.799 9.885 9.946 9.958 All Classes Increase 1.18%

9.02%

2.62%

2.74%

1. Values taken from Tables 6, 7, 8 and 9 of the main report.

2. All values except for total represent an increase of a factor of 20 from the baseline.

Although the individual classes increase by a factor of 20, the overall increase is between 1.2 and 2.7 percent of the total. This is not a significant impact and the overall conclusion drawn from this sensitivity case is that the calculated change in risk is insensitive to the evacuation assumption.

B.2 ACCIDENT SEQUENCE DOSE ESTIMATION For existing accident sequences, such as interfacing system LOCA or early containment failure, the dose is essentially constant for this assessment since the dose is substantially higher than is predicted for the Type A leakage. Therefore, the contributions from Classes 6, 7 and 8 form a baseline risk value that is constant for the analysis. As the intact dose impacted the numerator, the doses predicted from these classes impact the denominator.

The larger the person-rem estimate for these classes, the smaller the increase in risk attributed to the Type A leakage because it reflects a relatively smaller increase in risk. Therefore a lower estimate for Class 6, 7 and 8 would tend to increase the estimates for risk increase due to the extension of interval for the ILRT.

RSC 07-05 B.5 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Reference 5 provides a SONGS-specific assessment using an accepted approach for developing source terms utilizing radionuclide release fractions. For this analysis, an approach is taken to use the results presented in the NUREG 11506 study as a surrogate estimate for person-rem for Classes 6, 7 and 8.

NUREG 1150 examined both pressurized water reactors (PWRs) and boiling water reactors (BWRs). The results presented for boiling water reactors (i.e, Peach Bottom, Grand Gulf) are not considered appropriate for this analysis since the core melt mechanics and design are substantially different between SONGS and the BWRs. Therefore, their results are excluded from consideration.

NUREG 1150 also analyzed Zion, Sequoyah and Surry PWR designs. Sequoyah utilizes an ice condenser design and the presence of ice and restricted flow paths can lead to sequences and conditions that are not found in a large dry design such as SONGS. Therefore, Sequoyah is not considered a good PWR design for comparison. Zion is a 4 loop Westinghouse design and may be somewhat closer to the SONGS design. However the 4 loop design and the potential for seal LOCA cases associated most often with Westinghouse designs could influence the analysis.

Therefore, it is not selected as a surrogate.

Surry is a Westinghouse 3 loop design and is considered the best surrogate after examination of the NUREG 1150 analyzed plants.

Reference 7 provides the Level 2 analysis and offsite consequence assessment for Surry. Table 4.3-1 of Reference 5 provides a summary of consequence results that includes population dose (exposure) within 50 miles for internal events. A range of outcomes exists for each source term group based on the consequence measures. A matrix is formed and values provided for figures of merit.

The exposure estimates for a range of 50 miles around the site are provided in Table B.5 for each reported source term group.

RSC 07-05 B.6 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table B.5 Reported Person Rem Estimates for Surry Source Term Groups (summarized from Reference 5)

Source Term Grouping Outcome I (Sv )

Outcome 2 (Sv)

Outcome 3 (Sv)

SUR-01 NA 2.33E+3 1.25E+3 SUR-02 5.33E+3 1.13E+4 5.82E+3 SUJR-03 1.15E+4 2.26E+4 1.13E+4 SUR-04 1.04E+4 1.45E+4 NA SUR-05 NA 5.15E+4 2.62E+4 SUR-06 NA 2.42E+4

.? 15E+4 S LJR-07 2.76E+4 3.43E+4 1.46E+4 SUR-08 1.68E+4 2.14E+4 1.61E+4 SUR-09 1.36E+4 1.74E+4 NA SLUR-10 4.73E+4 4.66E+4 3.34E+4 SUJR-11 4.56E+4 2.77E+4 2.78E+4 SUR-12 2.69E+4 3.01E+4 2.67E+4 SUR-13 2.15E+4 2.68E+4 NA SUR-14 1.88E+4 2.23E+4 NA SUR-15 4.28E-1 3.1OE+O NA SUR-16 4.28E+0 3.75E+1 NA S JR-17 2.66E+3 6.71E+3 NA SUR-18 0.OOE+0 NA NA

1. Values provided in Sieverts (Sv). Conversion factor I Sv = 100 rem.

In order to utilize this information it is necessary to convert it to the form needed in the ILRT analysis. This involves classification into one of the three EPRI classes and then determining the representative person-rem estimates.

Reference 5 provides some guidance with respect to the composition of the source term grouping. For example SUR-01 is dominated by bypass sequences. Using this information the SuITy results are grouped to the EPRI classes. The grouping is presented in Table B.6.

RSC 07-05 B.7 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table B.6 Assignment of Sul-y Source Term Groups to EPRI Classes EPRI Class Surry Source Term Groups Applied Class 6 SUR-14 Class 7 SUR-04, SUR-07, SUR-08, SUR-09, SUR-I 1, SUR-12, SUR-13, SUJR-15, SUR-16, SUR-17 Class 8 SUR-01, SUR-02, SUR-03, SUR-05, SUR-06, SUR-10

1. Group SUR-18 is not applied to an EPRI class since the listed outcomes in either 0.0 or NA.

Table B.5 are The source term exposure estimates for each source term group are first averaged to obtain a value for the source term group and then the individual groups are averaged to obtain a class estimate. An example calculation is provided below.

Source term group (STG) SUR-01 has two estimates for exposure (see Table B.5). These values are first averaged to obtain a STG average for SUR-1.

SVaVg = (2.33E+3 + 1.25E+3) Sv /2 = 1.79E+3 Sv (eq. 3)

Repeating this process arrives at the data provided in Table B.7.

It is noted that for Class 7 and Class 8 there are multiple source term groups included. In these cases the individual results using Equation 1 for each contributing Surry STG were summed and then averaged to obtain an estimate for the EPRI class.

RSC 07-05 B.8 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table B.7 Average Person-Rem for SUITy Source Term Groups Source Term Group Exposure (Sv)

SUR-01 1.79E+3 SUR-02 7.48E+3 SUR-03 1.51E+4 SUR-04 1.25E+4 SUR-05 3.89E+4 SUR-06 2.29E+4 SUR-07 2.55E+4 SUR-08 1.8IE+4 SUR-09 1.55E+4 SUJR-10 4.24E+4 SUP,- 11 3.37E+4 SUR-12 2.79E+4 SUR-13 2.42E+4 SUR-14 2'.06E+4 SUR-15 1.76E+O SUR-16 2.09E+1 SUR-17 4.69E+3 SUR-18 NA These results are then grouped into the EPRI Classes using Table B.7 and the average, minimum and maximum exposures are defined. The results are presented in Table B.8 in units of person-rem.

RSC 07-05 B.9 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table B.8 Average Person-Rem for EPRJ Classes Based on Sunry Source Term Groups EPRI Class Weighted Average Max Exposure in Class Min Exposure in Exposure (person-(person-rern)

Class (person-rem) rem)

Class 6 2.06E+6 NA NA Class 7 1.62E+6 3.37E+6 1.76E+2 Class 8 2.14E+6 4.24E+6 1.79E+5

1. Only one source term group applied.

B.3 REFERENCES

1.

Kimball, D. S., Calculation N-6060-002, LOCA Containment Leakage, CR & Offsite Doses - AST, Southern California Edison, Revision 1, March 2006.

2.

Southern California Edison, San Onofre Nuclear Generating Station 2&3 FSAR, Updated, May 2007.

3.

Indian Point 3 Nuclear Power Plant, "Supplemental Information Regarding Proposed Change to Section 6.14 of the Administrative Section of the Technical Specification",

Entergy, IPN-01-007, January 18, 2001.

4.

Summitt, R., Comanche Peak Steam Electric Station Probabilistic Safety Assessment, Evaluation of Risk Significance of ILRT Extension, RSC, Inc., RSC 01-47/R&R-PN-110, November 2001.

5.

Summitt, R., San Onofre Nuclear Generating Station Unit 3 Probabilistic Risk Assessment Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach, Revision 0, RSC, Inc., RSC 05-03, April 2005.

6.

Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, U.S.

Nuclear Regulatory Commission, NUREG 1150, Vols. 1-3, December 1990.

7.

Breeding, R. J., et al] Evaluation of Severe Accident Risks: Surry Unit 1, Main Report, Rev. 1, U.S. Nuclear Regulatory Commission, NUREG/CR-4551, Vol. 3, October 1990.

RSC 07-05 B. 10 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Appendix C:

Response to USNRC Request for Additional Information for Degradation of the Embedded Side of the Steel Drywell Structure RSC 07-05 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach C.0 RESPONSE TO USNRC REQUEST FOR ADDITIONAL INFORMATION FOR DEGRADATION OF THE EMBEDDED SIDE OF THE STEEL DRYWELL STRUCTURE This appendix provides the San Onofre Nuclear Generation Station (SONGS) response to previous United States Nuclear Regulatory Commission (USNRC) request of other licensee (Reference 1) for additional information regarding how potential degradation of the embedded side of the steel drywell containment structure under high pressure during core damage accidents could impact the risk assessment related to the extension of the integrated leak rate test.

C.I ANALYSIS APPROACH The analysis approach utilizes the Calvert Cliffs Nuclear Plant (CCNP) methodology (Reference 1). This methodology is an acceptable approach to incorporate the liner corrosion issue into the integrated leak rate test (ILRT) extension risk evaluation.

The results of the analysis, using CCNP methodology, were that increasing the ILRT frequency from three years to sixteen years did not significantly increase plant risk of a large early release.

C.2 ANALYSIS RESULTS Table C.1 summarizes the results obtained from the CCNP methodology (Reference 1) utilizing plant-specific data for SONGS.

Table C. I Risk Assessment Results Using CCNP Methodology SONGS Liner Corrosion Step Description Containment Cylinder and Containment Basemat Dome (85%)

(15%)

Historical liner flaw likelihood Events 2 Events: 0 Failure data: containment location specific (Brunswick 2 and North Anna 2)

Assume a half failure Success data: based on 70 2 /(70 x 5.5) = 5.19E-03 0.5 /(70 x 5.5) = 1.30E-03 steel-lined containments and 5.5 years since the 10CFR 50.55a requirements of periodic visual inspections of containment surfaces RSC 07-05 D. I Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table C.1 (Continued)

SONGS Liner Corrosion Risk Assessment Results Using CCNP Methodology Step Description Containment Cylinder and Containment Basernat Dome (85%)

(15%)

Aged adjusted liner flaw Year Failure rate Year Failure rate likelihood 1

2.05E-03 I

5.14E-04 During the 16-year interval, average 5-10 5.19E-03 average 5-10 1.30E-03 assume failure rate doubles 16 1.65E-02 16 4.13E-03 every five years (14.9%

increase per year). The average for the 5 'h to 10'h year 16 year average = 6.74E-03 16 year average = 1.75E-03 set to the historical failure rate.

3 Increase in flaw likelihood between 3 and 16 years Uses aged adjusted liner flaw 10.5%

2.5%

likelihood (Step 2), assuming failure rate doubles every five years.

4 Likelihood of breach in Pressure (psia)

Likelihood of Pressure (psia)

Likelihood of containment given liner flaw liner breach liner breach The upper end pressure is consistent with the current 76 0.10%

76 0.010%

SONGS probabilistic risk 98.6 1.54%

98.6 0.154%

assessment (PRA) level 2 137 16.14%

137 1.614%

analysis (Reference 2). 0.1%

182 57.54%

182 5.754%

is assumed for the lower end.

262 100%

262 10.00%

Intermediate failure likelihoods are determined through logarithmic interpolation as documented in Reference 3. The basemat is assumed to be 1/10 of the cylinder/dome analysis.

RSC 07-05 D. 2 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Table C.2 Changes Due to Extension from 15 Years (current) to 16 Years Method LERF Increase Person-rem/yr Percentage Increase in In crease Person-rem/yr NRC approved 5.65E-9 3.36E-5'. 2 3.36E-5 / 9.692' =

method (SONGS 0.00035%

submittal basis)

NRC approved 6.48E-9, 3.86E-51.2 3.86E-5 / 9.692' =

method with liner 0.00040%

corrosion

1. Person-rem and LERF increase taken from main report.

2.

Assumes all leaks associated with corrosion are large (which is conservative)

Person-tern = LERF x 5.95E+03 (from Table 8 of the main report:).

3. LERF increase = submittal LERF + 8.325E-10 (calculated corrosion LERF increase).

Table C.3 shows the changes with the corrosion-induced, non-detected containment leakage for the ILRT extension from the baseline test interval of 3 years to the proposed test interval of 16 years.

Table C.3 Changes Due to Extension from 3 Years (baseline) to 16 Years Method LERF Increase Person-remryr Percentage Increase in Increase Person-remryr NRC approved 7.35E-8 4.37E-41 4.37E-04 / 9.687 =

method (SONGS 0.00452%

submittal basis)

NRC approved 7.43E-8, 3 4.42E-4' 2 4.42E-04 / 9.687' =

method with liner 0.00456%

corrosion

1.

Person-rem and LERF increase taken from main report.

2.

Assumes all leaks associated with corrosion are large (which is conservative)

Person-rem = LERF x 5.95E+03.

3.

LERF increase = submittal LERF + 8.325E-10 (calculated corrosion LERF increase).

The results of the analysis, using CCNP methodology, indicate that increasing the ILRT frequency from the current fifteen years or the baseline three years to sixteen years did not significantly increase plant risk of a large early release. This supports the proposed extension of the ILRT interval.

RSC 07-05 D.4 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach C.3 REFERENCES

1. Letter to NRC from Calvert Cliffs Nuclear Power Plant Unit No. 1:

Docket No. 50-317, Response to Request for Additional Information Concerning the License Amendment Request for a One-Time Integrated Leakage Rate Test Extension, dated March 27, 2002.

2. San Onofre Nuclear Generatino Station Living PRA, SONGS 2/3, PRA Level I Analysis Report, IPE-LEVEL2-000.

3. Calculation RSC-CALKNX-2004-0301, Esti mation of Containment Overpressure Fragility Curve for SONGS, Revision 0, Ricky Summitt Consulting, Inc., March 2004.

RSC 07-05 D.5 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach Appendix D:

Evaluation of Relevant SONGS ILRT Experience RSC 07-05 Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach D.0 EVALUATION OF RELEVANT SONGS ILRT EXPERIENCE This appendix documents the relevant historical experience for the San Onofre Nuclear Generating Station integrated leak rate tests performed on both units 2 and 3. From this data, comparisons of the resultant leak size for each integrated leak rate test performed as compared to L,, are presented.

D.1 ANALYSIS RESULTS San Onofie Nuclear Generating Station (SONGS) has performed three integrated leak rate tests (ILRT) on unit 2 and unit 3. From the final report for each test, the overall mass point and total time analysis leakage rates (in terms of a percentage of the containment leaked per day) are noted and compared to the leakage rate equivalent to La.

Table D.1 presents the results of the comparison.

Table D. 1 SONGS ILRT Resultant Leak Rates Unit Test Date Mass Point Total Point As-Found La Comment Number Resultant. As-Resultant As-Leakage Rate Found Leakage Found Leakage Limit Rate Rate 21 02/1985 0.064 wt% I day 0.053 wt% / day 0.1 wt% / day Test result leakage is less than La 22 10/1991 0.0485 wt% day 0.0713 wt% day 0.1 wt% / day Test result leakage is less than La 23 03/1995 0.0427 wt% / day 0.0676 wt% / day 0.1 wt% / day Test result leakage is less than La 34 11/1985 0.054 wt% / day 0.054 wt% / day 0.1 wt% / day Test result leakage is less than La 35 03/1992 0.0536 wt% / day 0.0681 wt% / day 0.1 wt% / day Test result leakage is less than La 3 '

09/1995 0.0546 wt% / day 0.0732 wt% / day 0.1 wt% / day Test result leakage is less than La

1. Information for this test taken from Reference I.

RSC 07-05 E. I Printed: 07/25/2007

Evaluation of Risk Significance of ILRT Extension Based on the NEI Approach

2. Information for this test taken from Reference 2.

3. Information for this test taken fiom Reference 3.

4. Information for this test taken from Reference 4.

5. Information for this test taken from Reference 5.

6. inforrnation for this test taken from Reference 6.

From Table D.1, it can be seen that there have been no reported leakage rates equal to or greater than L,,. From this, the SONGS data is not atypical (or unique) from the industry data as a whole for ILRT results. This supports the proposed ILRT extension.

D.2 REFERENCES 1.

3.

4.

San Onofre Nuclear Generating Station, Unit 2, Reactor Containment Building Integrated Leak Rate Test, Final Report, February 1985.

San Onofre Nuclear Generating Station, Unit 2I Reactor Containment Building Integrated Leak Rate Test, Final Report, October 1991.

San Onofre Nuclear Generating Station, Unit 2. Reactor Containment Building Integrated Leak Rate Test, Final Report, March 1995.

San Onofre Nuclear Generating Station, Unit 3, Reactor Containment Building Integrated Leak Rate Test, Final Report, November 1985.

San Onofre Nuclear Generating Station, Unit 3, Reactor Containment Building Integrated 5.

Leak Rate Test, Final Report, March 1992.

6. San Onofre Nuclear Generating Station, Unit 3, Leak Rate Test, Final Report, September 1995.

Reactor Containment Building Integrated RSC 07-05 E.2 Printed: 07/25/2007

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ML072681094
Site:	San Onofre (NPF-015)
Issue date:	09/24/2007
From:	Katz B Southern California Edison Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RSC 07-05, Rev 0
Download: ML072681094 (126)
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ML072681094

Text

Subject:

Dear Sir or Madam:

Enclosures:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

Subject:

Subject:

Subject:

Subject:

Subject:

4.3 REFERENCES

SUMMARY

SUMMARY

4.3 REFERENCES

SUMMARY

SUMMARY

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