NL-09-0359, Proposed Alternatives and Relief Request for the Fourth ISI Interval

Proposed Alternatives and Relief Request for the Fourth ISI Interval

ML092050741
Person / Time
Site:	Farley
Issue date:	07/24/2009
From:	Ajluni M Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-09-0359
Download: ML092050741 (26)

Text

Southern Nuclear Operating Company. Inc.

Pust Dtlice Box 1295 B,rllllrigharn, AlaiJama 35201-1295 If;) 7UJ ~8250[jU July 24, 2009 SOUTHERN'\\.

COMPANY Energy to Serve lour Worltl Docket Nos.: 50-348 NL-09-0359 50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant - Unit 1 and Unit 2 Proposed Alternatives and Relief Request for the Fourth lSI Interval Ladies and Gentlemen:

Pursuant to 10 CFR 50.55a(a)(3)(ii) and (a)(g)(5)(iii), Southern Nuclear Operating Company (SNC) hereby requests NRC approval of proposed alternatives to and relief from the specified ASME Boiler and Pressure Vessel Code Section XI requirements.

The details of these requests are contained in the enclosures. This is a re submittal of requests previously approved by the NRC for the Third lSI Interval.

Approval is requested by March 13, 2010 to support examinations to be performed during the Unit 2 Spring 2010 outage.

This letter contains no NRC commitments. If you have any questions, please advise.

Sincerely,

~9~

M. J. Ajluni Manager, Nuclear Licensing MJAlJLS/phr

U. S. Nuclear Regulatory Commission NL-09-0359 Page 2

Enclosures:

1) Proposed Alternative FNP-ISI-ALT-04 Version 1.0, in Accordance with 10 CFR 50.55a(a)(3)(ii)

2) Proposed Alternative FNP-ISI-ALT-05 Version 1.0, in Accordance with 10 CFR 50.55a(a)(3)(ii)

3) Proposed Relief Request FNP-ISI-ALT-06 Version 1.0, in Accordance with 10 CFR 50.55a(g)(5)(iii) cc:

Southern Nuclear Operating Company Mr. J. 1. Gasser, Executive Vice President Mr. J. R. Johnson, Vice President - Farley Ms. P. M. Marino, Vice President - Engineering RTYPE: CFA04.054 U. S. Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Mr. R. E. Martin, NRR Project Manager - Farley Mr. E. L. Crowe, Senior Resident Inspector - Farley

Joseph M. Farley Nuclear Plant - Units 1 and 2 Proposed Alternative for the Fourth lSI Interval Proposed Alternative FNP-ISI-ALT-04 Version 1.0, in Accordance with 10 CFR 50.55a(a)(3)(ii)

Plant Site Unit:

Interval Interval Dates:

Requested Date for Approval and Basis:

ASME Code Component(s)

Affected:

Applicable Code Edition and Addenda:

Applicable Code Requirement:

Reason for Request:

Proposed Alternative FNP-ISI-ALT-04 Version 1.0, in Accordance with 10 CFR 50.55a(a)(3)(ii)

Joseph M. Farley Nuclear Plant (FNP) Unit 1 and Unit 2.

4th lSI Interval extending from December 1, 2007 through November 30, 2017.

Approval is requested by March 13, 2010 to support 4th Interval examinations to be performed during the Spring 2010 Outage at FNP-2.

Thirty-nine Unit 1 and fifty-one Unit 2 small diameter (:=; 1-inch), Class 1, reactor coolant system (RCS) pressure boundary first isolation (inboard) vent and drain valves. A configuration drawing is provided in Figure 1.

Tables 1 and 2 provide the first isolation valve listing for Units 1 and 2, respectively.

ASME Section XI, 2001 Edition through the 2003 Addenda.

The 2001 Edition through 2003 Addenda of ASME Section XI, Table IWB 2500-1, Examination Category B-P, Item Number B15.1 0 requires examination of pressure retaining components in accordance with IWB 5222; specifically IWB-5222(b), which states, "the pressure retaining boundary during the system leakage test conducted at or near the end of each inspection interval shall extend to all Class 1 pressure retaining components within the system boundary."

Relief is requested to perform the Class 1 System 10-Year Test with the vent and drain valves in the closed position.

SNC believes there are potential personnel safety and ALARA issues associated with pressurizing these connections. These issues are as follows:

1. Pressure testing these connections to the outboard valve requires the inboard isolation valves to be opened and subjects the valves and piping to RCS nominal operating pressure. Opening the inboard valve at these conditions is contradictory to the requirement for double isolation of the RCS (10 CRF 50.55a(c)(2)(ii)), and thus creates the possibility for safety concerns for personnel performing visual examination of the E1 - 1 Proposed Alternative and Basis for Use:

Proposed Alternative FNP-ISI-ALT-04 Version 1.0, in Accordance with 10 CFR 50.55a(a)(3)(ii) connections.

2. Performing the test with the inboard valves open will increase the amount of dose personnel involved with the test will receive.

Several man-hours are required to position the valves for the test and to restore them after the test is complete. All of these valves are located in close proximity of the RCS main loop piping, thus requiring personnel entry into high radiation areas within the containment. Based on previous outage data, it is estimated that dose associated with valve alignment and realignment would be approximately 1.2 Rem per test (2.4 Rem every ten years for Unit 1 and Unit 2).

3. Since this test would be performed near the end of an outage, when all RCS work has been completed, the time required for opening and then closing these vent/drain valves could impact plant startup.

Imposition of the Code requirements creates the three issues described above, and does not provide a compensating increase in the level of quality or safety, as described below.

Proposed Alternative The RCS vent and drain connections will be visually examined with the isolation valves in the normally closed position each refueling outage for leakage and evidence of past leakage during the ASME XI Class 1 System Leakage Test (IWB-5220).

The RCS vent and drain connections will also be visually examined with the isolation valves in the normally closed position during the 10-year lSI pressure test (IWB-5222(b)).

Basis for Use These connections are equipped with manual valves which provide for double isolation of the reactor coolant system (RCS) pressure boundary.

These valves are generally maintained closed during all modes of operation and the piping outboard of the first isolation varve is, therefore, not normally pressurized.

The proposed alternative provides an acceptable level of safety and quality based on the following:

1. Requiring a leakage test and visual examination of these s; 1-inch diameter RCS vent/drain connections once each 1O-year interval is unwarranted considering that a repair weld on the same connections is exempted by the ASME XI Code. ASME Section XI E1 - 2 Duration of Proposed Alternative:

Precedents:

Proposed Alternative FNP-ISI-ALT-04 Version 1.0, in Accordance with 10 CFR 50.55a(a)(3)(ii)

Code, paragraph IWA-4540, provides the requirements for hydrostatic pressure testing of piping and components after repairs by welding to the pressure boundary. IWA-4540(b)(6) excludes component connections, piping, and associated valves that are 1 inch nominal pipe size and smaller from the hydrostatic pressure test requirement after welded repairs.

2. The non-isolable portion of the RCS vent and drain connections will be pressurized and visually examined as required. Only the isolable portion of the vent and drain connections is not pressurized.

3. A typical vent/drain connection includes two manual valves separated by a short pipe nipple which is connected to the RCS via another short pipe nipple and a half coupling. All connections are typically socket-welded and the welds receive a surface examination after installation. The pipil1g and valves are nominally heavy walls (Schedule 160 pipe able to withstand approximately 6000 psig and equivalent 6000# valve bodies). The design ratings are significantly greater than both RCS operating pressure (2235 psig) and design pressure (2485 psig).

4. The Technical Specifications (TS) require RCS leakage monitoring during normal operation. Should any of the TS limits be exceeded, then appropriate corrective actions, which may include shutting the plant down, are required to identify the source of the leakage and restore the RCS boundary integrity.

The contradictory conditions, added radiation exposure, and potential for outage impact associated with opening the valves, are not considered justifiable, since the proposed alternative visual examinations (in conjunction with the TS monitoring requirements for RCS leakage) provide reasonable assurance of structural integrity of the subject piping.

Examination in accordance with the Code requirements results in the potential for outage schedule impacts and radiation exposure without a compensating increase in the level of quality and safety; therefore, the proposed alternative should be granted pursuant to the requirements of 10CFR50.55a(a)(3)(ii).

The proposed alternative is applicable for the 4th Inservice Inspection Interval.

This alternative is a re-submittal of NRC approved 3'd Interval relief request RR-26 for both units. RR-26 was based on the 1989 Edition of Section XI, while this 4 th Interval request is based on the 2001 Edition through the 2003 Addenda. There have been no substantive changes to this alternative, to the Code requirements or to the basis for use which would alter the previous NRC Safety Evaluation conclusions.

E1 - 3

References:

Status:

Proposed Alternative FNP-ISI-ALT-04 Version 1.0, in Accordance with 10 CFR 50.55a(a)(3)(ii)

A similar 3rd Inservice Inspection Interval (for Unit 1) and 2nd Inservice Inspection Interval (for Unit 2) request was submitted and approved for Beaver Valley Power Station in 2007 (see TAC NOS. MD2936 and MD2937).

A similar 4th Inservice Inspection Interval request was submitted and approved for Hatch Nuclear Plant in 2005 (see TAC NOS. MC6526, MC6530, MC6531, MC6534, MC6535, MC6536, MC6537, MC6538, MC6539).

A similar 4 th Inservice Inspection Interval request was submitted and approved for H. B. Robinson Steam Electric Plant in 2002 (see RR-12 of TAC NO. MB2773).

SNC letter dated May 28, 1997 submitted RR-26 for both units.

RR-26 for both units was approved for 3rd Interval by NRC TAC numbers M98858 and M98859, dated January 12, 1999.

Awaiting NRC approval.

E1 - 4 Proposed Alternative FNP-ISI-ALT-04 Version 1.0, in Accordance with 10 CFR 50.55a(a)(3)(ii)

FIGURE 1 CONFIGURATION OF FNP CLASS 1 REACTOR COOLANT SYSTEM (RCS) PRESSURE BOUNDARY VENT AND DRAIN VALVES

.. -3".......

s 1" Piping

~-----I~I---------~~-

First Second Isolation Isolation Valve Valve Flow

• E1 - 5 Proposed Alternative FNP-ISI-ALT-04 Version 1.0, in Accordance with 10 CFR 50.55a(a)(3)(ii)

TABLE 1: FNP UNIT 1 RCS CLASS 1 FIRST ISOLATION (INBOARD) VALVES DRAWING TOTAL PLANT NUMBERING SYSTEM No.

LINE NUMBER VALVE DIAMETER Q1B13V093A CCA-15 Vent 3/4" Q1B13V094A CCA-22 Vent 3/4" Q1B13V095A CCA-7 Vent 3/4" Q1B13V096A CCA-24 Vent 3/4" Q1B13V097A CCA-27 Vent 3/4" Q1B13V072A CCA-27 Vent 3/4" 0175037 sheet 1 Q1B13V075A CCA-22 Vent 3/4" Q1B13V073A CCA-24 Vent 3/4" Q1B13V076A CCA-15 Vent 3/4" Q1B13V074A CCA-6 Vent 3/4" Q1B13V092A CCA-28 Vent 3/4" Q1B13V091A CCA-24 Vent 3/4" Q1B13V090A CCA-22 Vent 3/4" Q1B13V077A CCA-15 Drain 3/4" 0175037 sheet 2 Q1B13V089A CCA-15 Vent 3/4" Q1B13V078A CCA-15 Drain 3/4" Q1E21V561A CCA-26 Vent 3/4" Q1E21V415A CCA-21 Test 3/4" Connection Q1E11V054A CCA-22 Drain 3/4" Q1E11 V052A CCA-22 Drain 3/4" Q1E21V415B CCA-21 Test 3/4" Connection Q1E21V415C CCA-21 Test 3/4" Connection 0175038 Sheet 1 Q1E11V074A CCA-22 Drain 3/4" Q1E21V559A CCA-26 Vent 3/4" Q1E21V422A CCA-27 Drain & Test 3/4" Connection Q1E21 V412A CCA-30 Test 3/4" Connection Q1E21V411A CCA-29 Test 3/4" Connection Q1E21V422B CCA-27 Drain & Test 3/4" Connection E1 - 6 Proposed Alternative FNP-ISI-ALT-04 Version 1.0, in Accordance with 10 CFR 50.55a(a)(3)(ii)

TABLE 1: FNP UNIT 1 RCS CLASS 1 FIRST ISOLATION (INBOARD) VALVES I

DRAWING 0175038 Sheet 1 (continued)

TOTAL PLANT NUMBERING SYSYEMNo.

QIE21V412B QIE21 V411B QIE21V563A QIE21V412C QIE21 V411C LINE NUMBER CCA-30 CCA-29 CCA-28 CCA-30 CCA-28 VALVE Test Connection Test Connection Test Connection Drain Test Connection DIAMETER 3/4" 3/4" 3/4" 3/4" 3/4" 0175039 Sheet 1 0175041 Sheet 1 QIE21V535A QIE21V534A QIEI1V045B QIEIIV063A QIEI1V045A QIEIIV085A CCA-6 CCA-7 CCA-16 CCA-16 CCA-16 CCA-16 Drain Drain Drain & Test Connection Vent Drain & Test Connection Vent 3/4" 3/4" 1"

3/4" 1"

3/4" E1 - 7 Proposed Alternative FNP-ISI-ALT-04 Version 1.0, in Accordance with 10 CFR 50.55a(a)(3)(ii)

TABLE 2: FNP UNIT 2 RCS CLASS 1 FIRST ISOLATION (INBOARD) VALVES DRAWING D205037 sheet 1 D205037 sheet 2 TOTAL PLANT NUMBERING SYSTEM No.

Q2B13V093A Q2B13V094A Q2B13V095A Q2B13V096A Q2B13V097A Q2B13V072A Q2B13V075A Q2B13V073A Q2B13V076A Q2B13V074A Q2B13V092A Q2B13V091A Q2B13V090A Q2B13V077A Q2B13V081A Q2B13V078A LINE NUMBER CCA-15 CCA-22 CCA-7 CCA-24 CCA-27 CCA-27 CCA-22 CCA-I0 CCA-15 CCA-6 CCA-28 CCA-24 CCA-22 CCA-15 CCA-15 CCA-15 VALVE Vent Vent Vent Vent Vent Vent Vent Vent Vent Vent Vent Vent Vent Drain Vent Drain DIAMETER 3/4" 3/4" 3/4" 3/4" 3/4" 3/4" 3/4" 3/4" 3/4" 3/4" 3/4" 3/4" 3/4" 3/4" 3/4" 3/4" I

D205038 Sheet 1 Q2E21V415A Q2EIIV054A Q2Ell V083A Q2Ell V052A Q2EIIV080A Q2E21V589A Q2E21V415B I Q2Ell V082A Q2E21V582B Q2EIIV079A Q2E21 V415C Q2E21V582C Q2EIIV081A Q2E21V594A CCA-21 CCA-22 CCA-22 CCA-22 CCA-22 CCA-21 CCA-21 CCA-22 CCA-26 CCA-22 CCA-21 CCA-26 CCA-22 CCA-27 Drain & Test Connection Drain Vent Vent Drain Drain Drain & Test Connection Drain Vent Vent Drain & Test Connection Vent Drain Vent 3/4" 3/4" 3/4" 3/4" 3/4" 3/4" 3/4" 3/4" 3/4" 3/4" 3/4" 3/4" 3/4" 3/4" E1 - 8 Proposed Alternative FNP-ISI-ALT-04 Version 1.0, in Accordance with 10 CFR 50.55a(a)(3)(ii)

TABLE 2: FNP UNIT 2 RCS CLASS 1 FIRST ISOLATION (INBOARD) VALVES TOTAL PLANT LINE VALVE DIAMETER NUMBERING DRAWING 1

NUMBER SYSTEM No.

I Q2E21V412A CCA-30 Test 3/4" Connection CCA-29 Test 3/4" Connection Q2E21V594C 1 Q2E21V411A Vent 3/4" Q2E21V538A CCA-27 3/4" Q2E21V412B CCA-21 Drain 3/4" Connection Q2E21V41lB CCA-30 Test CCA-29 Test 3/4" 0205038 Sheet 1 Connection (continued) 3/4" Q2E21V587A Q2E21V586A CCA-28 Vent Drain 3/4" Q2E21V412C CCA-28 CCA-30 Test 314" Connection Q2E21V588A 3/4" Q2E21V411C CCA-28 Vent 3/4" Connection CCA-28 Test CCA-21 Q2E21V583A Vent 314" 1

Q2E21 V0539A CCA-27 Drain 3/4" 3/4" Q2E21V569A CCA-5 Drain 0205039 Sheet 1 Q2E2lV568A CCA-7 Drain 3/4" 0205041 Sheet 1 Q2E11V045B CCA-16 CCA-16 Q2E11 V063B Q2EllV101A CCA-16 Q2E11V045A CCA-16 Q2EI1V063A CCA-16 II Q2E11V046A CCA-16 3/4" Connection Vent 1 Drain & Test 3/4" Drain & Test 3/4" Connection Drain & Test 3/4" Connection 1

Vent 3/4" Drain & Test 3/4" Connection E1 - 9

Joseph M. Farley Nuclear Plant - Units 1 and 2 Proposed Alternative for the Fourth lSI Interval Proposed Alternative FNP-ISI-ALT-05 Version 1.0, in Accordance with 10 CFR 50.55a(a)(3)(ii)

Plant Site Unit:

Interval Interval Dates:

Requested Date for Approval and Basis:

ASME Code Component(s)

Affected:

Applicable Code Edition and Addenda:

Applicable Code Requirement:

Reason for Request:

Proposed Alternative FNP-ISI-ALT-05 Version 1.0, in Accordance with 10 CFR 50.55a(a)(3)(ii)

Joseph M. Farley Nuclear Plant (FNP) Unit 1 and Unit 2.

4th lSI Interval extending from December 1,2007 through November 30, 2017.

Approval is requested by March 13, 2010 to support 4th Interval examinations to be performed during the Spring 2010 Outage at FNP-2.

Portions of Spent Fuel Pool Cooling System piping immediately adjacent to the Spent Fuel Pit that are encased in concrete but are not provided with an annulus to facilitate visual examinations (

Reference:

Drawing 0-175043 and 0-205043, Sheet 1, line numbers 10-inch HCC-105, 8-inch HCC-107, and 1O-inch HCC-108 located in coordinates E-7 through E-9).

The two 10-inch lines, HCC-105 and HCC-108, are suction lines from the Spent Fuel Pit and the 8-inch line, HCC-107, is the common return line from the Spent Fuel Pool Heat Exchangers that discharges below the water level in the Spent Fuel Pool.

Configuration drawings are provided in Figures 1, 2, and 3.

ASME Section XI, 2001 Edition through the 2003 Addenda.

ASME Section XI, 200'1 Edition through 2003 Addenda, paragraph IWA 5244(b)(2) states, "The system pressure test for nonisolable buried components shall consist of a test to confirm that flow during operation is not impaired."

Relief is requested from performing the IWA-5244(b)(2) required test for the concrete encased portions of 10-inch HCC-105 and 10-inch HCC-108 and 8-inch HCC-107.

Portions of these pipe lines adjacent to the Spent Fuel Pool were encased in concrete during plant construction without an annulus to facilitate visual examinations. The plant's design also did not include any flow measuring instrumentation on the suction piping or any means of observing flow discharging from the return line. Therefore, a costly design modification would be required to meet the Code requirements. Imposition of the E2 - 1

Proposed Alternative and Basis for Use:

Duration of Proposed Alternative:

Precedents:

References:

Status:

Proposed Alternative FNP-ISI-ALT-OS Version 1.0, in Accordance with 10 CFR SO.SSa(a)(3)(ii) requirements is a financial hardship without a compensating increase in the level of quality and safety.

Proposed Alternative The piping sections immediately adjacent to the concrete encased sections will be examined each examination period to determine any evidence of material degradation or potential leakage during system inservice pressure tests.

Basis for Use All three piping segments normally experience low pressures <<100 psig based on water head pressure and on pump discharge pressure) and relatively low temperatures << 100 degrees Fahrenheit based on Spent Fuel Pool temperatures during hot months). Visual examinations of the piping segments adjacent to the concrete encased sections will provide reasonable assurance of the structural integrity of these low pressure, concrete encased, piping sections; therefore, approval should be granted pursuant to 10CFR50.55a(a)(3)(ii).

The proposed alternative is applicable for the 4th Inservice Inspection Interval.

This alternative is a re-submittal of NRC approved 3rd Interval relief request RR-29 for both units. RR-29 was based on the 1989 Edition of Section XI, while this 4th Interval request is based on the 2001 Edition through the 2003 Addenda. There have been no substantive changes to this alternative, to the Code requirements or to the basis for use which would alter the previous NRC Safety Evaluation conclusions.

SNC letter dated May 28, 1997 submitted RR-29 for each unit.

RR-29 for both units was approved for 3rd Interval by NRC TAC numbers M98858 and M98859, dated January 12, 1999.

Awaiting NRC approval.

E2 - 2 Proposed Alternative FNP-ISI-ALT-05 Version 1.0, in Accordance with 10 CFR 50.55a(a)(3)(ii)

FIGURE 1 CONFIGURATION OF 10" HCC-105 SEGMENT THAT IS BURIED IN CONCRETE (SIMILAR FOR BOTH UNITS)

E2 - 3 Proposed Alternative FNP-ISI-ALT-05 Version 1.0, in Accordance with 10 CFR 50.55a(a)(3)(ii)

FIGURE 2 CONFIGURATION OF 8" HCC-107 SEGMENT THAT IS BURIED IN CONCRETE (SIMILAR FOR BOTH UNITS)

E2 - 4 Proposed Alternative FNP-ISI-ALT-05 Version 1.0, in Accordance with 10 CFR 50.55a(a)(3)(ii)

FIGURE 3 CONFIGURATION OF 10" HCC-108 SEGMENT THAT IS BURIED IN CONCRETE (SIMILAR FOR BOTH UNITS)

E2 - 5

Joseph M. Farley Nuclear Plant - Units 1 and 2 Proposed Relief Request for the Fourth 151 Interval Proposed Relief Request FNP-151-ALT-06 Version 1.0, in Accordance with 10 CFR 50.55a(g)(5)(iii)

Plant Site Unit:

Interval Interval Dates:

Requested Date for Approval and Basis:

ASME Code Component(s)

Affected:

Applicable Code Edition and Addenda:

Applicable Code Requirement:

Impracticality of Compliance:

Proposed Relief Request FNP-ISI-ALT-06 Version 1.0, in Accordance with 10 CFR 50.55a(g)(5)(iii)

Joseph M. Farley Nuclear Plant (FNP) Unit 1 and Unit 2.

4th lSI Interval extending from December 1, 2007 through November 30, 2017.

Approval is requested by March 13,2010 to support 4th Interval examinations to be performed during the Spring 2010 Outage at FNP-2.

Class 2 piping segments in the Safety Injection System that cannot be isolated from Class 1 piping in order to perform pressure testing at the Class 2 required test pressure. See Table 1 for more details.

Configuration drawings are provided in Figures 1, 2, 3, and 4.

ASME Section XI, 2001 Edition through 2003 Addenda, Table IWC-2500-1, Category C-H, requires a system leakage test of Class 2 pressure retaining components in accordance with IWC-5220 once each inspection period.

ASME Section XI, 2001 Edition through 2003 Addenda, Table IWC-2500-1, Category C-H, requires pressure testing of Class 2 pressure retaining components in accordance with IWC-5220. At issue specifically is IWC 5221 which states, "The system leakage test shall be conducted at the system pressure obtained while the system, or portion of the system, is in service performing its normal operating function."

Compliance with the Code is impractical because:

1) Each of the subject piping segments includes an outboard motor operated valve (MOV) that allows for system isolation/injection to the Reactor Coolant System (RCS) and the inboard check valve(s) that provide the Class 1 to Class 2 change boundary. System flow is through the MOV and check valve(s) to the RCS. There are no other boundary valves downstream of the check valve(s) that can be used as a pressure test and/or class boundary.

2) The subject lines are all associated with either High Head Safety Injection (HHSI) or Low Head Safety Injection (LHSI). The configuration of the plant is such that these piping segments will not experience RCS pressure during normal operation. Consequently, no system surveillances achieving normal RCS system operating pressure exists for these piping segments. They do not exist because HHSI and LHSI operate only during emergency conditions.

E3 - 1 Burden Caused by Compliance:

Proposed Alternative and Basis for Use:

Proposed Relief Request FNP-ISI-ALT-06 Version 1.0, in Accordance with 10 CFR 50.55a(g)(5)(iii)

Since the function and design of these segments are for emergency use only, a major hardware modification would be required in order to test these segments at RCS pressure like the rest of the RCS Pressure Boundary piping.

Pressure testing of these piping segments would require pressure testing the entire RCS boundary and the temporary installation of pressure "jumpers" between vent/drain connections on each side of the subject check valves. These "jumpers" would provide pressure equalization between the Class 1 and Class 2 piping thus meeting the code requirement. However, the installation of these temporary "jumpers" adds extra time to the refueling outage. Extra time is added to the refueling outage due to: 1) the test would be performed in Mode 3 during startup concurrent with the Class 1 System Leakage Test (IWB-5220), 2) several temporary "jumpers" would be required, and 3) "jumpers" must be removed prior to progressing into Mode 2 (Startup).

Proposed Alternative:

The Class 2 piping segments will be included within the VT-2 visual examination boundary during the Class 1 system leakage test (IWB-5220) each refueling outage.

Additionally, once each inspection interval, the piping segments will be pressurized to approximately RCS nominal operating pressure in conjunction with the Class 1 10-year leakage test. With the Class 1 system at nominal operating pressure (approximately 2235 psig), these Class 2 piping segments will be pressurized, using a test pump or jumpers, to approximately RCS nominal operating pressure. All Class 1 system leakage tests are done during Mode 3 (Hot Standby) when RCS temperature is;::: 350°F.

These proposed alternative pressure tests and visual examinations provide adequate assurance of the pressure boundary integrity of the subject Class 2 piping segments.

Basis for Use:

The system design did not include provisions for isolating these Class 2 piping segments from the Class 1 system for Class 2 Code-required pressure testing. Since there is no practical method to isolate and pressure test these piping segments, they cannot be tested unless the associated Class 1 reactor coolant system (RCS) piping and components are pressurized as well.

The design rating of all the piping segments, except for the 10-inch line, is Schedule 160, designed to withstand at least 3500 psig. The 10-inch line is Schedule 140, designed to withstand approximately 2600 psig. As noted in Table 1, this is greater than the pressure these piping segments need to withstand during a safety injection.

E3 - 2 Duration of Proposed Alternative:

Precedents:

Status:

Proposed Relief Request FNP-ISI-ALT-06 Version 1.0, in Accordance with 10 CFR 50.55a(g)(5)(iii)

A low pressure flow is passed through the HHSI segments each refueling outage during a flow balance procedure designed to check the needle valves upstream. The presence of this flow during this test will manifest any leaks, cracks or imperfections within the line. A VT-2 examination will be on the segments during this test.

In conclusion, as an alternative to ASME Section XI, 2001 Edition through 2003 Addenda, Table IWC-2500-1, Category C-H, during every refueling outage these piping segments will be VT-2 examined only. During the 10 year test these piping segments will be pressurized and VT-2 examined.

The proposed relief request is applicable for the 4th Inservice Inspection Interval.

This relief request is a re-submittal of NRC approved 3rd Interval relief request RR-30 for both units. SNC letter dated May 28, 1997 submitted RR-30 for each unit. RR-30 for both units was approved for 3 rd Interval by NRC TAC numbers 1\\1198858 and M98859, dated January 12, 1999.

RR-30 was based on the 1989 Edition of Section XI, while this 4th Interval request is based on the 2001 Edition through the 2003 Addenda. There have been no substantive changes to this request, to the Code requirements or to the basis for use which would alter the previous NRC Safety Evaluation conclusions.

A similar 4th Inservice Inspection Interval request was submitted and approved for Surry Power Station in 2005 (see TAC NO. MC5586). Relief was asked for Class 1 piping that is similar to the aforementioned piping segments.

A similar 4 th Inservice Inspection Interval request was submitted and approved for H. B. Robinson in 2002 (see TAC NO. MB2773). Relief was asked for Class 1 piping that is similar to the aforementioned piping segments.

Awaiting NRC approval.

E3 - 3 Proposed Relief Request FNP-ISI-ALT-06 Version 1.0, in Accordance with 10 CFR 50.55a(g)(5)(iii)

Table 1: Affected Class 2 Components ORAWING LINE NUMBER OESCRIPTION SAFETY INJECTION SERVICE CONOITION psig of 0175038-1 (Unit 1) 0205038-1 (Unit 2) 2" CCB-30 3" CCB-30 Between motor operated valve 01(2)2E21V068 and check valves 01 (2)2E21V078A, B & C (Hot Leg Safety Injection) 2555 120 2555 120 0175038-1 (Unit 1) 0205038-1 (Unit 2) 2" CCB-31 3" CCB-31 Between motor operated valve 01 (2)E21V072 and check valves 01 (2)E21V079A, B & C (Hot Leg Safety Injection) 2555 120 0175038-1 (Unit 1) 0205038-1 (Unit 2) 2" CCB-22 3" CCB-22 Between motor operated valve 01 (2)E21V063 and check valves 01 (2)E21V066A, B & C (Cold Leg Safety Injection) 2555 120 0175038-1 (Unit 1) 0205038-1 (Unit 2) 2" CCB-21 3" CCB-21 Between motor operated valves 01(2)E21V016A & B and check valves 01 (2)E21V062A, B & C (Cold Leg Safety Injection) 2500 350 0175038-1& 2 (Unit 1) 0205038-1 & 2 (Unit 2) 6" CCB-29 10" CCB-29 Between motor operated valve 01(2)E11V044 and check valves 01 (2)E21V076A & B (RHR Hot Leg Injection)

E3 - 4 Proposed Relief Request FNP-ISI-ALT-06 Version 1.0, in Accordance with 10 CFR 50.55a(g)(5)(iii)

FIGURE 1 PIPING SEGMENT BETWEEN MOTOR OPERATED VALVE Q1(2)E21V068 AND CHECK VALVES Q1(2)E21V078A, B & C (HOT LEG SAFETY INJECTION; PIPING SEGMENT BETWEEN MOTOR OPERATED VALVE Q1(2)E21V072 AND CHECK VALVES Q1(2)E21V079A, B & C (HOT LEG SAFETY INJECTION)

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B-21 3/4" 0 Proposed Relief Request FNP-ISI-ALT-06 Version 1.0, in Accordance with 10 CFR 50.55a(g)(5)(iii)

FIGURE 3 PIPING SEGMENTS BETWEEN MOTOR OPERATED VALVES Q1(2)E21V016A &BAND CHECK VALVES Q1(2)E21V062A, B & C (COLD LEG SAFETY INJECTION)

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FIGURE 4 PIPING SEGMENT BETWEEN MOTOR OPERATED VALVE Q1(2)E11V044 AND CHECK VALVES Q1(2)E21V076A & B (RHR HOT LEG INJECTION)

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