LR-N02-0393, Additional Information for License Change Request H02-013 Regarding One-Time Extension to Increase Interval of Integrated Leak Rate Test from Ten to Fifteen Years

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Additional Information for License Change Request H02-013 Regarding One-Time Extension to Increase Interval of Integrated Leak Rate Test from Ten to Fifteen Years
ML023370548
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 11/22/2002
From: Garchow D
Public Service Enterprise Group
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
LCR H02-013, LR-N02-0393
Download: ML023370548 (18)
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Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 NOV 22 2002 0 PSEG LR-N02-0393 Nuclear LLC LCR H02-013 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Gentlemen:

ADDITIONAL INFORMATION FOR LICENSE CHANGE REQUEST H02-013 REGARDING ONE-TIME EXTENSION TO INCREASE THE INTERVAL OF THE INTEGRATED LEAK RATE TEST FROM TEN TO FIFTEEN YEARS HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354

Reference:

Letter LR-N02-0319, Request for One-Time Extension to Increase the Interval of the IntegratedLeak Rate Test from Ten to Twenty Years, dated October 9, 2002 On October 9, 2002 PSEG Nuclear LLC (PSEG) submitted the referenced request for a revision to the Technical Specifications (TS) to extend the Type A Containment Integrated Leak Rate Test (ILRT) in Section 6.8.4.f from once per 10 years to once per 20 years for the Hope Creek Generating Station. Mr. George Wunder, NRC Hope Creek Project Manager, advised PSEG that the NRC could not support the review for extension to 20 years, but could support an extension to 15 years in the requested time frame. PSEG agreed to modify the request accordingly. Attachments 1 and 2 of our October 9, 2002 letter have been revised and are included with this letter. Revisions are denoted by marginal markings. Attachment 3 to our October 9, 2002 letter does not require revision to support the extension to 15 years.

If you have any questions or require additional information, please contact Mr. Michael Mosier at (856) 339-5434.

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

Executed on NOV 2 2 2002 Sincerely, D. F. G ah co w Vice President - Operations Attachments 95-2168 REV. 7/99

Document Control Desk NOV 2 2 2002 LR-N02-0393 C: Mr. H. Miller, Administrator- Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. George Wunder, Project Manager - Hope Creek U. S. Nuclear Regulatory Commission Mail Stop 08B3 Washington, DC 20555-0001 USNRC Senior Resident Inspector - Hope Creek (X24)

Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering PO Box 415 Trenton, New Jersey 08625 2

Document Control Desk LR-N02-0393 LCR H02-013 REQUEST FOR CHANGE TO TECHNICAL SPECIFICATIONS ONE-TIME EXTENSION TO INCREASE THE INTERVAL OF THE INTEGRATED LEAK RATE TEST FROM TEN TO FIFTEEN YEARS

1. DESCRIPTION ................................................................................................... 2
2. PROPOSED CHANGE ......................................................................................... 2
3. BACKGROUND ................................................................................................... 2
4. TECHNICAL ANALYSIS ...................................................................................... 4 4 .1 Methodo logy .............................................................................................. .. 4 4.2 Assumptions/Bases ..................................................................................... 5 4 .3 C a lculatio n ................................................................................................ .. 6 4 .4 R isk Im pact ................................................................................................ ..6
5. REGULATORY SAFETY ANALYSIS ..................................................................... 9 5.1 No Significant Hazards Determination ........................................................ 9 5.2 Applicable Regulatory Requirements/Criteria ........................................... 11
6. ENVIRONMENTAL IMPACT EVALUATION ........................................................ 11
7. REFERENCES .................................................................................................... 12

Document Control Desk LR-N02-0393 Attachment I LCR H02-013

1. DESCRIPTION This letter is a request to amend Facility Operating License NPF-57 for the Hope Creek Generating Station (HOGS). The proposed change would revise Technical Specification 6.8.4.f, "Primary Containment Leakage Rate Testing Program" to permit a one-time extension to the maximum ten-year interval to fifteen years to perform the Type A test. The proposed change will provide an economic benefit by eliminating the Integrated Leak Rate Test (ILRT) from Refueling Outage 11 (RF1 1), reducing the critical path by approximately 44 hours5.092593e-4 days <br />0.0122 hours <br />7.275132e-5 weeks <br />1.6742e-5 months <br /> with no significant impact on safety.

Approval of this proposed change is being requested by the end of February 2003 to support the scheduled implementation date of April 2003.

2. PROPOSED CHANGE The proposed changes to the Technical Specifications (TS) are included in Attachment 2 of this submittal. In summary, it is requested that:

Section 6.8.4.f, "Primary Containment Leakage Rate Testing Program," be amended to permit a one-time extension to the maximum ten-year frequency to be increased to fifteen years to perform the ILRT. The proposed TS change is based on past successful Type A, B, and C tests, and American Society of Mechanical Engineers (ASME)Section XI inspections (reference 7.12) at HCGS. The results for HOGS are shown in Table 1. Further justification is based on research documented in NUREG 1493 (reference 7.7) which generically shows that very few potential containment leakage paths fail to be identified by Type B and C tests. In fact, an analysis of 144 ILRT test results, including 23 failures, found that no failures were due to containment liner breach. The NUREG concluded that reducing the Type A (ILRT) testing frequency to once per twenty years would lead to an imperceptible increase in risk. A plant specific calculation provided in Attachment 3 demonstrates that the risk impact of the proposed change when compared to other severe accident risks is negligible. The purpose of this submittal is to request a one-time deferral of the Type A (ILRT) from April 12, 2004 to no later than April 12, 2009.

3. BACKGROUND ILRTs have been required of operating nuclear power plants to ensure the public health and safety in the case of an accident that would release radioactivity to the containment.

Conservative design and construction have led to very few ILRTs exceeding their required leakage. The NRC has extended the allowable ILRT test period from three times in ten years to once in ten years based on past successful tests. NUREG-1493 that supported the change to the ten-year interval also stated that test periods of up to twenty years would lead to an imperceptible increase in risk.

Section 3.8.2 of the Hope Creek Updated Final Safety Analysis Report (UFSAR) describes the primary containment. The steel containment is an ASME B&PV Code 2

-Document Control Desk LR-N02-0393 Attachment I LCR H02-013 Class MC vessel designed to house the Nuclear Steam Supply System (NSSS). The steel containment is a part of the Primary Containment System, which limits the postulated release of radioactivity from the NSSS. This section describes the structural design considerations for the primary containment and includes information that provides the bases for design, construction, and testing of the steel containment, except as modified by the plant unique analysis report, submitted to the NRC under separate cover (letter from R.L. Mittl to Albert Schwencer, dated February 10, 1984.).

The primary containment consists of a drywell, a pressure suppression chamber, and an interconnecting vent system. The drywell is a steel pressure vessel with a spherical lower portion 68 feet inside diameter, a cylindrical upper portion 40 feet 6 inches inside diameter, and a removable, flanged, hemi-ellipsoidal top head, 33 feet 2 inches inside diameter. Its overall height is 114 feet 9 inches. The bottom elevation of the spherical portion is 77 feet 10 inches. Inner and outer steel cylindrical skirts that are encased in concrete and anchored to a concrete pedestal support the drywell. The suppression chamber consists of 16-mitered cylindrical shell segments joined together to form a torus shaped pressure vessel located below and encircling the drywell. The suppression chamber has a major diameter of 112 feet 8 inches, a minor or chamber diameter of 30 feet 8 inches, and contains water to an approximate depth of 14 feet. Eight equally spaced vent pipes connect the drywell and the suppression chamber, each with an internal diameter of 6 feet 2 inches. These vent pipes are connected to a common mitered header within the suppression chamber with a major diameter of 112 feet 8 inches and a minor diameter of 4 feet 3 inches.

The satisfactory results from previous integrated leakage rate tests at HCGS, as well as continued satisfactory results of local leak rate tests, and containment inspections, support deferral of the RF1 1 test. The reactor containment will continue to be inspected under the requirements of ASME Section Xl Subsections IWE and IWL. The existing Type B and C containment penetration-testing program will continue to be performed in accordance with previous regulatory approvals.

PSEG has performed three operational ILRT tests. All tests passed the as-found acceptance criteria of 1.0 La, where La is the maximum allowable accident leakage rate.

The results are shown in Table 1.

Structural degradation of containment is a gradual process that occurs due to the effects of pressure, temperature, radiation, chemical, or other such effects. Such effects would be identified and corrected when the containment structure is periodically tested and inspected to verify structural integrity under ASME Section Xl Subsections IWE and IWL. The most recent 100% IWE inspection performed was during refueling outage RFO9, in Spring 2000. The next scheduled 100% IWE is RF1 1, Spring 2003 These surveillances provide a high degree of assurance that any degradation of the containment structure will be detected and corrected before it can produce a containment leakage path. The tests and inspections conducted to date have not identified degradation that threatens the integrity of the HCGS containment.

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-Document Control Desk LR-N02-0393 Attachment I LCR H02-013 The NRC has approved similar changes in Amendment No. 197 for Crystal River Unit 3.

Also, in Amendment No. 206 for Entergy Nuclear Operations, Inc.'s Indian Point Nuclear Generating Unit 3 the NRC approved a one-time increase 10 to 15 years for the ILRT interval. In addition the NRC approved a similar change in Amendment No. 234 for Salem Unit 2.

4. TECHNICAL ANALYSIS The purpose of this analysis is to demonstrate that extending the Type A Integrated Leak Rate Test (ILRT) interval from the current 10 years required by 10 CFR 50, Appendix J (reference 7.1) at Hope Creek Generating Station (HCGS) to 15 years has a negligible impact on risk. The risk in this analysis is defined in terms of population dose (person-rem) per reactor year, large early release frequency (LERF) and conditional containment failure probability (CCFP). Consequently, the impact of Type A extension is evaluated against the person-rem, LERF and CCFP.

This calculation evaluates the risk associated with various ILRT intervals as follows. The focus is the risk changes from the current 10 years to the proposed 15 years.

E 3 years - interval based on the original requirements of 3 tests per 10 years 0 10 years - current test interval required for HCGS 0 15 years - interval extension proposed for HCGS 4.1 Methodology The evaluation for HCGS follows the guidelines set forth in NEI 94-01 (reference 7.5),

the methodology used in EPRI TR-104285 (reference 7.6), NUREG-1493 (reference 7.7), EPRI Interim Guidance (reference 7.11), and the regulatory guidance on the use of Probabilistic Safety Assessment (PRA) findings in support of a licensee request to a plant's licensing basis, RG 1.174 (reference 7.8). The calculation applies the HCGS Individual Plant Examination (IPE) release categories, current core damage frequency (CDF), and the Level 3 PRA person-rem estimates to estimate the changes in risk due to increasing the ILRT test interval. This information is obtained from the HCGS IPE (reference 7.9), HCGS PRA, Revision 1.3 (reference 7.14), and a Level 3 PRA study (reference 7.10) performed by SCIENTECH for HCGS.

In addition to the references mentioned above, improvements suggested in references (7.11 and 7.13) are implemented in this evaluation. The previous methodology for LERF (Class 3b frequency) calculation involved conservatively multiplying the CDF by the failure probability for this class (3b) of accident. This was done for simplicity and to maintain conservatism. However, core damage sequences include individual sequences that either may already (independently) cause a LERF or could never cause a LERF, and are thus not associated with a postulated large Type A containment leakage path (LERF). These contributors should be removed from Class 3b release 4

  • Document Control Desk LR-N02-0393 Attachment I LCR H02-013 evaluation by multiplying the Class 3b probability by only that portion of CDF that may be impacted by type A leakage.

The analysis steps performed are listed below:

  • Calculate the Level 3 release category population doses.
  • Map the Level 3 release categories into the 8 release classes defined by the EPRI report.
  • Calculate the Type A leakage estimate to define the analysis baseline.
  • Calculate the Type A leakage to address the current inspection frequency.
  • Calculate the Type A leakage estimates to address extension of the Type A test interval.
  • Calculate the change in population dose due to extending Type A inspection intervals.
  • Calculate the change in LERF due to extending Type A inspection intervals.
  • Calculate the change in CCFP due to extending Type A inspection intervals.

4.2 AssumptionslBases

"* The maximum containment leakage for Class 1 sequences is estimated using the level 3 PRA results and is defined as 1 La unit for this analysis.

"* The maximum containment leakage for Class 3a sequences is 10 times the class 1 sequences based on the previously approved methodology.

(references 7.2, 7.3, and 7.11)

"* The maximum containment leakage for Class 3b sequences is 35 times the class 1 sequences based on the previously approved methodology.

(references 7.2, 7.3, and 7.11)

"* Containment leakage due to Classes 4, 5 and 6 are considered negligible based on references 7.2 and 7.3.

  • The containment releases are not impacted by time.
  • Because Class 8 sequences are containment bypass sequences, potential releases are directly to the environment. Therefore, the containment structure will not impact the release magnitude.
  • This calculation uses the CDF from the latest HCGS PRA (Revision 1.3) and the release categories and frequency distribution in the HCGS IPE. This approach is used for the following two reasons. First, the latest Level 1 PRA revision reflects the plant configuration more accurately, but the Level 2 PRA, except LERF, has not been updated. Second, the Level 2 PRA in the HCGS IPE is extensive and has enough information for distributing the latest CDF to various release categories. The CDF value used in this calculation is 8.89E 6/year, which is the CDF in the latest HCGS PRA.

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  • Document Control Desk LR-N02-0393 Attachment I LCR H02-013 4.3 Calculation The inputs for this calculation come from the information documented in the HCGS IPE (reference 7.9), HCGS PRA, Revision 1.3 (reference 7.14), and a Level 3 PRA study (reference 7.10) performed by SCIENTECH for HCGS. The Level 3 study used the MACCS2 computer code to develop person-rem dose results. The study also used site-specific inputs for meteorological and population data.

The current HCGS PRA is a non-safety-related tool and is intended to provide "best estimate" results that can be used as input when making risk-informed decisions.

The HCGS IPE (reference 7.9) is an earlier version of the PRA submitted to NRC in response to Generic Letter 88-20. Neither the PRA nor the IPE is considered as design basis information. Other inputs to this calculation include ILRT test data from NUREG-1493 (reference 7.7), EPRI Interim Guideline (reference 7.11) and the EPRI report (reference 7.6) are referenced in the body of the calculation.

4.4 Risk Impact The change in Type A test frequency from once every ten years to once every fifteen years increases the total integrated plant risk by only 0.13%. Also, the change in Type A test frequency from the original every three years to once every fifteen years increases the risk only 0.26%. 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 1 E-6/yr and increases in LERF below 1 E 7/yr. Since the ILRT does not impact CDF, the relevant criterion is the increase in LERF. The increase in LERF resulting from a change in the Type A ILRT test frequency from the current once every 10 years to once in every 15 years is 2.191E 8/yr. It meets the guidance in Reg. Guide 1.174 as a very small change in LERF; therefore, increasing the ILRT interval from 10 to 15 years is considered non-risk significant. The LERF increase for the cumulative change from a test frequency of three times in every ten years to once in every fifteen years is 5.248E-8/yr, which is still non-risk significant.

R.G. 1.174 also encourages the use of risk analysis techniques to ensure 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 is estimated to be 0.25% for the proposed change and 0.59% for the cumulative change of going from a test frequency of three times in every ten years to once in every fifteen years. These changes are small and demonstrate that the defense-in-depth philosophy is maintained.

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- Document Control Desk LR-N02-0393 Attachment I LCR H02-013 Table I Hope Creek Generating Station ILRT Results Total Time Method TS 3.3.3 Test Date Leakage Rate Test Duration (or)Ciei Acceptance

(%/day) (Hours) Criteria January 2, 1986 0.175*

0.180"* 24 0.75 La 0.084*

November 9, 1989 0.087"* 24 0.75 La April 12,1994 0.200* 11hours 10 0.75 La 0.217** minutes Measured leakage

    • Leakage with 95% upper confidence 7

Document Control Desk LR-N02-0393 Attachment I LCR H02-013 Table 2 Summary of Risk Impact on Extending Type A ILRT Test Frequency Risk Impact Risk Impact Risk Impact for 3-year for 10-year interval for 15-year interval interval (current (proposed)

(baseline) requirement)

Total Integrated Risk (Person-Rem/yr) 15.67 15.69 15.71 Type A Testing Risk (Person-Rem/yr) 0.010 0.034 0.051

% Total Risk (Type A / Total) 0.065% 0.216% 0.324%

Type A LERF (Class 3b)

(per year) 1.312E-08 4.369E-08 6.560E-08 Changes due to extension from 10 years (current)

A Risk from current (Person-rem/yr) 0.02

% Increase from current (A Risk / Total Risk) 0.13%

A LERF from current (per year) 2.191E-08 A CCFP from current 0.25%

Changes due to extension from 3 years (baseline)

A Risk from baseline (Person-rem/yr) 0.04

% Increase from baseline (A Risk / Total Risk) 0.26%

A LERF from baseline (per year) 5.248E-08 A CCFP from baseline 0.59%

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  • Document Control Desk LR-N02-0393 Attachment 1 LCR H02-013
5. REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration Determination PSEG Nuclear LLC (PSEG) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10CFR50.92, "Issuance of amendment," as discussed below:
1. Does the change involve a significant increase in the probability or consequences of an accident previously analyzed?

Response: No.

The proposed revision to Section 6.8.4.f adds a one-time extension to the current interval for containment integrated leak rate test (ILRT). The current test interval of 10 years, based upon past performance, would be extended on a one-time basis to 15 years from the last ILRT. The proposed extension to ILRT testing cannot increase the probability of an accident previously evaluated since the containment ILRT testing extension is not a modification to plant systems, nor a change to plant operation that could initiate an accident. The proposed extension to Type A testing does not involve a significant increase in the consequences of an accident since research documented in NUREG-1493, "Performance-Based Containment Leak-Test Program," found that very few potential containment leakage paths fail to be identified by Type B and C tests.

The NUREG concluded that reducing the ILRT testing frequency to once per twenty years would lead to an imperceptible increase in risk. Containment performance monitoring is performed in accordance with the Maintenance Rule (10CFR50.65) and inspections required by American Society of Mechanical Engineers (ASME) code are performed in order to identify indications of containment degradation that could affect leak tightness. Type B and C testing required by the technical specifications (TS) will identify any containment opening, such as valves, that would otherwise be detected by the ILRT. Reg.

Guide 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. It also recommends the use of risk analysis techniques to ensure and show that the proposed change is consistent with the defense-in-depth philosophy. The increase in large early release frequency (LERF) resulting from a change in the ILRT test frequency from the current once in every 10 years to once in every 15 years is less than 1 E-7 per year, thereby meeting Regulatory Guide 1.174 definition of a very small change in risk. The change in conditional containment failure probability (CCFP) is estimated to be 0.25% for the proposed change. These factors show that an ILRT test extension will not represent a significant increase in the consequences of an accident.

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  • Document Control Desk LR-N02-0393 Attachment I LCR H02-013
5. REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration Determination PSEG Nuclear LLC (PSEG) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10CFR50.92, "Issuance of amendment," as discussed below:
1. Does the change involve a significant increase in the probability or consequences of an accident previously analyzed?

Response: No.

The proposed revision to Section 6.8.4.f adds a one-time extension to the current interval for containment integrated leak rate test (ILRT). The current test interval of 10 years, based upon past performance, would be extended on a one-time basis to 15 years from the last ILRT. The proposed extension to ILRT testing cannot increase the probability of an accident previously evaluated since the containment ILRT testing extension is not a modification to plant systems, nor a change to plant operation that could initiate an accident. The proposed extension to Type A testing does not involve a significant increase in the consequences of an accident since research documented in NUREG-1493, "Performance-Based Containment Leak-Test Program," found that very few potential containment leakage paths fail to be identified by Type B and C tests.

The NUREG concluded that reducing the ILRT testing frequency to once per twenty years would lead to an imperceptible increase in risk. Containment performance monitoring is performed in accordance with the Maintenance Rule (1 OCFR50.65) and inspections required by American Society of Mechanical Engineers (ASME) code are performed in order to identify indications of containment degradation that could affect leak tightness. Type B and C testing required by the technical specifications (TS) will identify any containment opening, such as valves, that would otherwise be detected by the ILRT. Reg.

Guide 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. It also recommends the use of risk analysis techniques to ensure and show that the proposed change is consistent with the defense-in-depth philosophy. The increase in large early release frequency (LERF) resulting from a change in the ILRT test frequency from the current once in every 10 years to once in every 15 years is less than 1 E-7 per year, thereby meeting Regulatory Guide 1.174 definition of a very small change in risk. The change in conditional containment failure probability (CCFP) is estimated to be 0.25% for the proposed change. These factors show that an ILRT test extension will not represent a significant increase in the consequences of an accident.

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Document Control Desk LR-N02-0393 Attachment I LCR H02-013 Therefore, this proposed amendment does not involve a significant increase in the probability of occurrence or consequences of an accident previously analyzed.

2. Does the change create the possibility of a new or different kind of accident from any accident previously analyzed?

Response: No The proposed revision to Section 6.8.4.f adds a one-time exception to the current interval for the ILRT. The current test interval of 10 years, based upon past performance, would be extended on a one-time basis to 15 years from the last Type A test. Primary containment is designed to contain energy and fission products during and after an event. The Individual Plant Examination (IPE) identifies events that lead to containment failure. Revision to the ILRT test interval does not change this list of events. 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 a new or different kind of accident.

Therefore, this proposed amendment does not create the possibility of a new or different kind of accident from any previously analyzed.

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

Response: No The proposed revision to Section 6.8.4.f adds a one-time extension to the current interval for the ILRT. The current test interval of 10 years, based upon past performance, would be extended on a one-time basis to 15 years from the last ILRT. The proposed extension to ILRT testing interval will not significantly reduce the margin of safety. The NUREG-1493 generic study of the effects of extending containment leakage testing found that a 20-year exception in ILRT leakage testing resulted in an imperceptible increase in risk to the public.

NUREG-1493 found that the containment leakage rate contributes a very small amount to the individual risk, and that the decrease in Type A testing frequency would have a minimal affect on this risk since most potential leakage paths are detected by Type C testing. Type B and Type C testing will continue to be performed at a frequency currently required by the Technical Specifications (TS).

The containment inspections being performed in accordance with ASME, Section Xl, and Maintenance Rule (10CFR50.65) provide a high degree of assurance that the containment will not degrade in a manner that is only detectable by Type A testing.

Reg. Guide 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. It also recommends the use of risk analysis techniques to ensure and show that the proposed change is consistent 10

Document Control Desk LR-N02-0393 Attachment I LCR H02-013 with the defense-in-depth philosophy. The increase in large early release fraction (LERF) resulting from a change in the ILRT test frequency from the current once in every 10 years to once in every 15 years is less than 1E-7 per year, thereby meeting Regulatory Guide 1.174 definition of a very small change in risk. The change in conditional containment failure probability (CCFP) is estimated to be 0.25% for the proposed change.

Therefore, these changes do not involve a significant reduction in margin of safety.

Based on the above, PSEG 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 RequirementslCriteria 5.2.1 Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," September 1995 (RG 1.163).

5.2.2 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.

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

Final Report, September 1995 (NUREG-1493).

5.2.4 Title 10, Code of Federal Regulations, Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors".

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6. ENVIRONMENTAL IMPACT EVALUATION PSEG has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.

However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment 11

Document Control Desk LR-N02-0393 Attachment I LCR H02-013 meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7. REFERENCES 7.1. Title 10, Code of Federal Regulations, Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors".

7.2. Florida Power, 3F0601-06, "Crystal River- Unit 3 - License Amendment Request #267, Revision 2, Supplemental Risk-Informed Information in Support of License Amendment Request #267," June 20, 2001.

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

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

7.5. NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J", July 26, 1995, Revision 0 7.6. EPRI TR-1 04285, "Risk Assessment of Revised Containment Leak Rate Testing Intervals" August 1994.

7.7. NUREG-1493, "Performance-Based Containment Leak-Test Program", July 1995.

7.8. 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.

7.9. HCGS Probabilistic Risk Assessment Individual Plant Examination Submittal, Revision 0, March 1994.

7.10. SCIENTECH 17268-001, "Hope Creek MACCS2 Model," 9/2002.

7.11. EPRI Interim Guidance for Performing Risk Impact Assessments In Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals", November 2001.

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Document Control Desk LR-N02-0393 Attachment I LCR H02-013 7.12. 1998 Edition of Subsection IWE and IWL, "Requirements for Class MC and Metallic Liners of Class CC Components of Light-Water Cooled Power Plants,"

of Section XI, Division 1, of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code).

7.13. NEI Memo to the USNRC, 'One-time extensions of containment integrated leak rate test interval - additional information.' November 30, 2001 7.14. HCGS Probabilistic Safety Assessment, Revision 1.3, November 3, 2000.

7.15. HCGS Technical Specifications 13

Document Control Desk LR-N02-0393 LCR H02-013 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-54 DOCKET NO. 50-354 REVISIONS TO THE TECHNICAL SPECIFICATIONS (TS)

TECHNICAL SPECIFICATION PAGE WITH PROPOSED CHANGE The following Technical Specification for Facility Operating License NPF-57 are affected by this change request:

Technical Specification Page 6.8.4.f 6-16b 1

ADMINISTRATIVE CONTROLS 6.8.4.f Primary Containment Leakage Rate Testing Program A program shall be established, implemented, and maintained to comply with the leakage rate testing of the containment as required by 10CFR50.54(o) and 10CFR50, 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

ýntl ýent Leak-Te* Program", dated September 1995) co _Z The peak calculate containment internal pressure for the design basis loss of coolant accident, Pa, is 48.1 psig.

The maximum allowable primary containment leakage rate, La, at Pa, shall be 0.5% of primary containment air weight per day.

Leakage Rate Acceptance Criteria are:

a. Primary containment leakage rate acceptance criterion is less than or equal to 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are less than or equal to 0.6 La for Type B and Type C tests and less than or equal to 0.75 La for Type A tests;
b. Air lock testing acceptance criteria are:
1) Overall air lock leakage rate is less than or equal to 0.05 La when tested at greater than or equal to Pa,
2) Door seal leakage rate less than or equal to 5 scf per hour when the gap between the door seals is pressurized to greater than or equal to 10.0 psig.

The provisions of Specification 4.0.2 do not apply to the test frequencies specified in the Primary Containment Leakage Rate Testing Program.

The provisions of Specification 4.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.

6.8.4.g. Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBER(S) OF THE PUBLIC from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

HOPE CREEK 6-16b Amendment No.

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