ML082760752

Fourth 10-Year Interval Inservice Inspection Request Relief PT-02 for Proposed Alternative Inspection for Standby Gas Treatment System

ML082760752
Person / Time
Site:	Brunswick
Issue date:	11/19/2008
From:	Boyce T Plant Licensing Branch II
To:	Waldrep B Carolina Power & Light Co
Saba F
References
TAC MD8124, TAC MD8125
Download: ML082760752 (8)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 November 19, 2008 Mr. Benjamin Waldrep, Vice President Brunswick Steam Electric Plant Carolina Power & Light Company Post Office Box 10429 Southport, North Carolina 28461

SUBJECT:

BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 - RELIEF REQUEST PT-02 FOR THE FOURTH 10-YEAR INSERVICE INSPECTION INTERVAL OF THE STANDBY GAS TREATMENT SYSTEM (TAC NOS. MD8124 AND MD8125)

Dear Mr. Waldrep:

By letter dated February 6, 2008, Carolina Power & Light Company (the licensee) submitted Relief Request PT-02 (RR PT-02) for the fourth 1O-year interval of the Brunswick Steam Electric Plant (BSEP), Units 1 and 2 Inservice Inspection (lSI) Program pertaining to system pressure tests. The licensee requested relief from performing the American Society of Mechanical Engineers Code-required pressure test of the standby gas treatment system. Alternatively, the licensee proposed a visual examination during each refueling outage of the accessible pressure retaining boundary of the standby gas treatment system to assure its structural integrity. The fourth 10-year lSI interval began on May 11, 2008 and is scheduled to be completed by May 10, 2018.

Based on the information provided in RR PT-02, the Nuclear Regulatory Commission (NRC) staff concluded in the enclosed safety evaluation that the licensee's proposed alternative would provide an acceptable level of quality and safety. Therefore, pursuant to Title 10 of the Code of Federal Regulations, Section 50.55a(a)(3)(i), the NRC authorizes the lSI program alternative proposed in RR PT-02 for the fourth 10-year lSI interval of BSEP, Units 1 and 2.

The bases for the NRC staff's conclusion are contained in the enclosed Safety Evaluation. If you have any questions regarding this issue, please contact Farideh Saba at (301) 415-1447 or farideh.saba@nrc.gov.

Sincerely,

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Thomas . Boyce, hief Plant Licensing B anch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-325 and 50-324

Enclosure:

Safety Evaluation cc w/encl: See next page

Carolina Power & Light Company Brunswick Steam Electric Plant, Units 1 and 2 cc:

Sandra Spencer, Mayor City of Southport 201 East Moore Street Southport, North Carolina 28461 Additional Distribution via ListServ

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST PT-02 FOR THE FOURTH 10-YEAR INSERVICE INSPECTION INTERVAL BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 DOCKET NOS. 50-325 AND 50-324

1.0 INTRODUCTION

By letter dated February 6,2008, Carolina Power & Light Company (CP&L, the licensee) submitted Relief Request PT-02 (RR PT-02) for the fourth 1O-year interval of the Brunswick Steam Electric Plant (BSEP), Units 1 and 2 Inservice Inspection (lSI) Program pertaining to system pressure tests. In RR PT-02, the licensee requested relief from performing the American Society of Mechanical Engineers (ASME) Code-required pressure test of the standby gas treatment (SGT) system. Alternatively, the licensee proposed a visual examination during each refueling outage of the accessible pressure retaining boundary of the SGT system to assure its structural integrity.

The fourth 1O-year lSI interval began on May 11, 2008 and is scheduled to be completed by May 10, 2018.

2.0 REGULATORY EVALUATION

Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a(g) requires that lSI of ASME Code Class 1, 2, and 3 components be performed in accordance with Section XI of the ASME Code and applicable addenda, except where specific written relief has been granted by the U.S. Nuclear Regulatory Commission (NRC) pursuant to 10 CFR 50.55a(g)(6)(i). According to 10 CFR 50.55a(a)(3), alternatives to the requirements of paragraph 50.55a(g) may be used, when authorized by the NRC, if: (i) an applicant demonstrates that the proposed alternatives would provide an acceptable level of quality and safety, or (ii) the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Pursuant to 10 CFR 50.55a(g)(4), ASIVIE Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that lSI of components and system pressure tests conducted during the first 1O-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b),

12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The lSI code of record for the fourth 1O-year inspection interval of the BSEP, Units 1 and 2 is the 2001 Edition with 2003 Addenda.

Enclosure

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3.0 TECHNICAL EVALUATION

3.1 Relief Request PT -02 The licensee requests relief from the ASME Code requirement to perform a system leakage test and VT-2 visual examination that are required to be conducted at the system pressure obtained while the system is in service performing its normal operating function or at the system pressure developed during a test conducted to verify system operability.

3.1.1 Components for Which Relief is Requested SGT system including all pressure retaining components 3.1.2 Code Requirements The ASME Code,Section XI, 2001 Edition with 2003 Addenda, Article IWC-5000 "System Pressure Tests," requires that pressure retaining components be visually examined and tested at the frequency specified in Table IWC-2500-1, Examination Category C-H. The system pressure test and visual examination shall be conducted in accordance with Articles IWA-5000 and IWC-5000. The contained fluid in the system shall serve as the pressurizing medium.

3.1.3 Licensee's Proposed Alternative The licensee proposes to perform an alternative structural integrity visual examination of the accessible pressure retaining boundary defined in Subarticle IWC-5222 with a frequency of once every refueling outage in lieu of the ASME Code-required system leakage test with a frequency of once every inspection period (40-month interval).

3.1.4 Licensee's Basis for Requesting Relief PT -02 The SGT System and supporting system components consist of suction piping, two parallel 100 percent capacity filter trains and blowers, and a discharge vent. The system and the components perform several functions following a design basis loss-of-coolant accident (LOCA) and during other conditions when the reactor building ventilation system is isolated. Their safety-related functions include: (1) maintaining the secondary containment structure at a negative pressure of 0.25 inches of water by controlled venting of the building atmosphere at a rate of 100 percent of the structure volume per day, and (2) removing the halogens and other fission products from the atmosphere vented from the drywell and/or suppression chamber following a LOCA. These components also perform several functions that are not safety-related.

During normal plant operation, the SGT System is in a standby mode and aligned to take suction from the reactor building atmosphere in case an automatic start signal is received. In the unlikely event this system is needed for mitigation of a potential release, the system would be aligned to draw radioactive material from either primary or secondary containment through a series of high efficiency particular air/charcoal filters. Once drawn through the filters, the material would be exhausted to an elevated release point (i.e., 1OO-meter tall plant stack).

-3 The licensee, in Enclosure 2 of its letter dated February 06, 2008, stated that during normal operation, the affected components will experience minimal pressures. For the affected components, observing leakage during a system leakage test is unlikely since they will only experience minimal or negative pressure during a test. During a system leakage test, the SGT System will take suction from the reactor building atmosphere. Once a fan blower is started, the test medium (Le., reactor building atmosphere) is drawn through one of the filter banks and exhausted to the plant stack. Because the fan blower is creating suction to draw the atmosphere through the filters, the components upstream of the fan blower will experience little to no pressure during the system leakage test. In the unlikely event that a structural distress occurs, the process fluid would not escape since it would be drawn into the system. Since the fan blower is exhausting the filtered atmosphere to the stack, the components downstream of the blower could be slightly pressurized during the test. Again, it would be unlikely that leakage would be detected during the test since it is an open path to the plant stack and the test medium would take the path of least resistance.

The licensee, in RR PT-02, proposed an alternative to the ASME Code requirement to perform a structural integrity visual examination of the accessible pressure-retaining boundary, as defined in IWC-5222, during each refueling outage. The licensee determined that the proposed alternative will provide an acceptable level of quality and safety for the following reasons:

1. Since the system leakage test is not an affective method for identifying structural distress, the licensee will perform a structural integrity visual examination of accessible pressure retaining boundary each refueling outage. CP&L considers this structural integrity visual examination a superior and proven method for identifying potential degradation. Similar visual examinations are also performed on other safety significant components (e.g., components classified as Class MC). The licensee also stated that performing a structural integrity visual examination once each refueling outage would detect and correct potential degradation. The performance of this visual examination is also considered an acceptable alternative to the current test frequency specified in the ASME Code, which is once every inspection period (Le., approximately every other refueling outage).

2. Not performing the system leakage test or the visual (Le., VT-2) examination of the SGT System and supporting system components will not compromise quality or safety. The SGT System and supporting system components were designed and constructed and tested commensurate to ASIVIE Code,Section III, Class 2 and for seismic forces in accordance with seismic class I requirements. In addition to the non-destructive examinations performed on the components by their manufacturer, each butt weld associated with this piping had a surface and VOlumetric examination performed.

Following completion of the construction, the piping was also hydrostatically tested. As such, these components were constructed and tested to high quality standards.

The licensee stated that the operability of these safety-related components is assured by the performance of a series of surveillance requirements specified in BSEP, Units 1 and 2 Technical Specification Surveillance Requirements 3.6.4.3.1, 3.6.4.3.2, and 3.6.4.3.3. These surveillance requirements demonstrate acceptable operation of this system by verifying system flow, differential pressure, efficiency and performance.

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3. The SGT System is classified as a standby system under the Maintenance Rule (Le.,10 CFR 50.65) based on the classification as safety related and the design function of mitigating the consequences of design basis accidents and transients. In addition to assessing performance data, the structural condition of these systems is also periodically monitored as part of the Maintenance Rule Program.

The licensee, in RR-PT-02, concluded that "Since the system operability is verified by surveillance requirements and performance of the proposed structural integrity visual examinations each refueling outage, there is no safety significance associated with not performing the system leakage test or the visual (Le., VT-2) examination. Because of the SGT System operating characteristics, the performance of the system leakage test is not an effective method for assuring integrity of these components. As such, the described alternative will provide an acceptance level of quality and safety pursuant to 10 CFR 50.55a(a)(3)(i).

3.1.5 NRC Staff Evaluation of Relief Request PT-02 The code of record requires a system leakage test and a VT-2 visual examination conducted at the system pressure obtained while the system or portion of the system is in service performing its normal operating function. However, due to the operating characteristics of the SGT System, the system leakage test is not an effective method for assuring structural integrity of components in the system. The system operates at a slight negative pressure in the suction side of the blowers and at minimal pressure in the discharge past the blowers to the vent stacks.

Performing the ASME Code-required test and VT-2 visual examination will not achieve the purpose of the ASME Code requirement, which is to effectively detect and identify sources of pressure boundary leakage.

The proposed alternative includes a structural integrity walkdown once every outage (instead of every inspection period) to detect evidence of leakage, such as structural distress and evidence of degradation, including coating degradation. If a crack exists, it would likely cause paint irregularities and the appearance of surface rust staining, which would be observable since the piping is painted and uninsulated. The walkdown would also provide reasonable assurance of structural integrity of the piping system.

The subject system has low potential of experiencing degradation due to the following reasons:

(1) The safety-related components were constructed to standards commensurate to ASME Code,Section III, and Class 2 and included in the lSI program. As such, the SGT System was constructed and tested to high quality standards.

(2) Prior to installation, nondestructive examinations were performed on each pressure retaining component. Following the installation, each butt weld associated this piping had a surface and volumetric examination performed. In addition, the piping was hydrostatically tested.

-5 (3) The environment to which the systems are exposed is an ambient containment atmosphere. During normal operation, the SGT System is in a standby mode. When purging through the exhaust fan, pressure is controlled between -0.25 to 1 pound per square inch gauge (psig). The system will not be exposed to a severe environment or harsh conditions.

(4) The temperature and pressure design parameters for the SGT System are 150 degrees Fahrenheit and 5 psig, respectively, including corrosion allowance for the specified design life.

As discussed above, the systems were constructed and tested per ASME Code quality.

Previous nondestructive examinations have demonstrated their leakage integrity. The system is in standby mode during plant operation and will not be exposed to any harsh environment. The leakage integrity and structural integrity will not be challenged in a manner that would result in any significant degradation. The frequent structural walkdown will ensure that evidence of degradation would be detected in a timely manner. Furthermore, technical specification surveillance requirements also demonstrate acceptable operation of this system by verifying system flow, differential pressure across various filters including the heaters and moisture separators, mechanical efficiency of the filters, the ability of the heaters to maintain relative humidity, and the ability of the charcoal to remove appropriate amount of radionuclide. Since these requirements are verified with the system in operation, the test results are an indicator of actual system performance and operability.

Certified and properly trained personnel will perform the structural integrity visual examinations.

Personnel performing these examinations will be certified in accordance with the applicable requirements of the ASME Code,Section XI, IWA-2300. This level of certification will ensure that the capability and visual acuity of the personnel is sufficient to detect evidence of potential degradation.

The performance of the structural integrity visual examination once every refueling outage, in lieu of the current test frequency specified in the ASME Code, which is once every inspection period (Le., approximately every other refueling outage), is also considered to be more effective in detecting potential degradation early in operating life and taking timely corrective action.

Based on the above evaluation, the staff finds that performance of the alternative examination offers reasonable assurance of structural and leakage integrity of the SGT system and the supporting system components.

4.0 CONCLUSION

The NRC staff concludes that compliance with the ASME Code-required system leakage test with VT-2 visual examination would not effectively detect leakage or evidence of leakage. The licensee's proposed alternative of structural integrity visual examination with the test frequency of once every outage would assure structural and leakage integrity by detecting evidence of potential degradation and initiating timely corrective action. Therefore, the proposed alternative in RR PT-02 is authorized for the fourth 10-year lSI interval of BSEP, Units 1 and 2, pursuant to 10 CFR 50.55a(a)(3)(i), since it would provide an acceptable level of quality and safety. All other

-6 requirements of the ASME Code,Section XI for which relief has not been specifically requested remain applicable, including a third party review by the Authorized Nuclear Inservice Inspector.

Principal Contributors: Prakash Patnaik Date: November 19 , 2008

November 19, 2008 Mr. Benjamin Waldrep, Vice President Brunswick Steam Electric Plant Carolina Power & Light Company Post Office Box 10429 Southport, l\Jorth Carolina 28461 SUB.JECT: BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 - RELIEF REQUEST PT-02 FOR THE FOURTH 10-YEAR INSERVICE INSPECTION INTERVAL OF THE STANDBY GAS TREATMENT SYSTEM (TAC NOS. MD8124 AND MD8125)

Dear Mr. Waldrep:

By letter dated February 6, 2008, Carolina Power & Light Company (the licensee) submitted Relief Request PT-02 (RR PT-02) for the fourth 1O-year interval of the Brunswick Steam Electric Plant (BSEP), Units 1 and 2 Inservice Inspection (lSI) Program pertaining to system pressure tests. The licensee requested relief from performing the American Society of Mechanical Engineers Code-required pressure test of the standby gas treatment system. Alternatively, the licensee proposed a visual examination during each refueling outage of the accessible pressure retaining boundary of the standby gas treatment system to assure its structural integrity. The fourth 1O-year lSI interval began on May 11,2008 and is scheduled to be completed by May 10,2018.

Based on the information provided in RR PT-02, the Nuclear Regulatory Commission (NRC) staff concluded in the enclosed safety evaluation that the licensee's proposed alternative would provide an acceptable level of quality and safety. Therefore, pursuant to Title 10 of the Code of Federal Regulations, Section 50.55a(a)(3)(i), the NRC authorizes the lSI program alternative proposed in RR PT-02 for the fourth 10-year lSI interval of BSEP, Units 1 and 2 The bases for the NRC staff's conclusion are contained in the enclosed Safety Evaluation. If you have any questions regarding this issue, please contact Farideh Saba at (301) 415-1447 or farideh.saba@nrc.gov.

Sincerely, IRA!

Thomas H. Boyce, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-325 and 50-324

Enclosure:

Safety Evaluation cc w/encl: See next page Distribution:

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