ML081160094

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Transmittal of Revision 1 to Request for Relief No. 04-ON-012 Third Ten Year Inservice Inspection Interval
ML081160094
Person / Time
Site: Oconee  Duke Energy icon.png
Issue date: 04/21/2008
From: Baxter D
Duke Energy Carolinas, Duke Power Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RR 04-ON-012, R1
Download: ML081160094 (25)
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Text

I Duke DAVE BAXTER Vice President kEnergy. Oconee Nuclear Station Duke Energy Corporation ON01 VP17800 Rochester Highway Seneca, SC 29672 864-885-4460 864-885-4208 fax dabaxter@dukeenergy.com April 21,2008 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001

Subject:

Duke Power Company LLC d/b/a Duke Energy Carolinas, LLC (Duke)

Oconee Nuclear Station, Units 1, 2, 3 Docket Nos. 50-269,-270,-287 Third Ten Year Inservice Inspection Interval Request for Relief No. 04-ON-01 2, Revision 1 By letter dated December 21, 2004 (ADAMS Accession # ML043630367), Duke Power Company, now Duke Power Company LLC, d/b/a Duke Energy Carolinas, LLC, (Duke) submitted Request for Relief (RFR) 04-ON-01 2, seeking relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section Xl, Rules for Inservice Inspection of Nuclear' Power Plant Components, for Oconee Nuclear Station, Units 1, 2, and 3 for the duration of the Third Ten Year Inservice Inspection Interval. The third Inservice Inspection Interval for Oconee terminated January 1, 2004 for Unit 1; September 9, 2004, for Unit 2; and January 2, 2005 for Unit 3.

By letter dated July 6, 2006 (ADAMS Accession # ML061930413), Duke subsequently withdrew RFR 04-ON-012, with the statement that Duke intended to resubmit a revised version of the request at a later date. Duke is submitting RFR 04-ON-01 2 Revision 1, which replaces and supersedes the original request, for NRC review and approval in order to close out the third interval documentation.

The Oconee Reactor Coolant Pressure Boundary was originally designed and built to USAS B31.7 Nuclear Power Piping Code (February 1968) and Addenda (June 1968), which considered the Reactor Coolant Pressure Boundary to be through the first valve.

As a result, certain sections of pipe, specified in the attached request, were designed such that the system leakage test conditions specified for Class 1 cannot be attained without excessive hardship.

A4OY www. duke-energy.com

U. S. Nuclear Regulatory Commission April 21, 2008 Page 2 Request for Relief 04-ON-01 2, Revision 1, is to allow Duke to credit testing performed to date on these sections in lieu of the required system leakage tests.

Duke requests that the NRC authorize these alternatives under 10 CFR 50.55a(a)(3)(ii).

If there are any questions or further information is needed you may contact Corey Gray at (864) 886-6325.

Very truly yours, Dave axter Site Vice President Enclosure

U. S. Nuclear Regulatory Commission April 21, 2008 Page 3 xc w/att: Victor McCree Region II Administrator (Acting)

U.S. Nuclear Regulatory Commission Atlanta Federal Center 61 Forsyth St., SWW, Suite 23T85 Atlanta, GA 30303 L. N. Olshan, Project Manager, Section 1 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 xc(w/o attch):

G. A. Hutto Senior NRC Resident Inspector Oconee Nuclear Station Susan E. Jenkins, Section Manager Division of Waste Management Bureau of Land and Waste Management SC Dept. of Health & Environmental Control 2600 Bull St.

Columbia, SC 29201

U. S. Nuclear Regulatory Commission April 21, 2008 Page 4 bxc w/att: J. J. Mc Ardle III M. A. Pyne L. C. Keith J. M. Boughman P. A. Wells D. W. Peltola A. D. Best T. J. Coleman V. B. Dixon G. L. Brouette (ANII)

C. A. Gray J. E. Smith R. L. Gill, Jr.

ISI Relief Request File NRIA File/ELL EC050 Document Control

Duke Energy Corporation Oconee Nuclear Station Units 1, 2, & 3 Request for Relief No. 04-ON-012 Rev. 1 1.0 System / Component for Which Relief is Requested:

Relief is requested for sections of ASME Code Class 1 piping and components connected to the Reactor Coolant System (RCS) that are isolated from direct RCS pressure during normal plant operation. These sections of pipe are isolated from the RCS by their configuration because they are up stream of a check valve, between two or more check valves, or between two normally closed valves that remain closed when the unit is in Modes 3, 2, or 1.

At Oconee Nuclear Station, for each unit, there are five such sections of pipe that would result in hardship or unusual difficulty to perform a Class 1 hydrostatic test.

All five sections are totally contained within the reactor building. The five sections, of pipe are as follows:

1. Between LP-1 and LP-2
2. Between LP-103 and LP-104
3. Between LP-47, CF-I1, and CF-12
4. Between LP-48, CF-13, and CF714
5. Between LP-46 and LP-131 This represents 15 sections of pipe between the three units. The section of pipe between these valves can not be hydrostatically tested as specified in ASME Section XI 1989 edition, no addenda. Attachment 1 contains a table which gives the pipe material, header pipe size, and schedule as well as the length of the header pipe in each section.

For the three Oconee units the start and end dates for the third 10-year interval are as follows:

Third 10-Year Interval Unit Start Date End Date 1 7/15/1994 1/1/2004 2 12/16/1994 9/9/2004 3 12/16/1',994 1/02/2005 This document includes two attachments. Attachment 1 contains information from the UFSAR on piping classification and tables with information about the major components in each section of pipe. Attachment 2 contains portions of five drawings, one for each section of pipe addressed by this request for relief. These drawings were taken from the appropriate one line diagrams (Oconee Flow Diagrams). Since these piping sections are identical except for the differences described in other sections of this request a single drawing for each section was included.

Request for Relief 04-ON-012 Rev. 1 Page 2 of 16 Pipe Section 1: Configuration for Valves LP-l and LP-2:

LP-1 and LP-2 are part of the decay heat removal system. This section of pipe is placed in service after the reactor has been shutdown and depressurized to about 350 psig.

LP-1 and LP-2 are 12 inch gate valves separated by a piece of 12 inch schedule 140 stainless steel pipe. LP-1 is the first valve off the reactor coolant system (RCS) on the decay heat removal line.

Up Stream LP-I Between LP-1 Down Stream

__& LP-2 LP-2 Design Pressure 2500 2500 388 Design Temperature 650 300 300 Pipe size & Schedule 12 in, Sch 140 12 in, Sch 140 12 in, Sch 10s*

  • Unit 3 is schedule 40s See tables in Attachment # I for additional information.

Pipe Section 2: Configuration for Valves LP-103 and LP-104:

LP-103 and LP-104 are part of the boron dilution system. These valves would be used during a LOCA to prevent concentration of boron in the reactor vessel.

Down stream of LP-104 this pipe is open ended into the reactor building basement. When this section of pipe is placed in service, water from the RCS would discharge to the basement where it could be recycled through the LPI pumps.

LP-103 and LP-104 are 3 inch gate valves separated by a piece.of 3 inch schedule 160 stainless steel pipe. LP-103 is the first valve off the decay heat removal line upstream of valve LP-1.

Up Stream LP- Between LP- Down Stream 103 103 & LP-104 LP-104 Design Pressure 2500 2500 400 Design Temperature 650 425 425 Pipe size & Schedule 3 in, Sch 160 3 in, Sch 160 3 in, Sch 10s*

  • Unit 3 is 4 inch schedule 40 Valve 3LP-104 is a 4 inch valve See tables in Attachment # I for additional information.

Request for Relief 04-ON-012 Rev. 1 Page 3 of 16 Pipe Section 3: Configuration for Valves LP-47, CF-I 1, and CF-12 These three valves, LP-47, CF-11, and CF-12, exist in a section of pipe that directs water from the core flood tank or the LPI pump to the reactor vessel. As long as the RCS pressure remains above 600 psig these three valves remain closed.

LP-47 is a 10 inch check valve and CF-12 and CF-I I are 14 inch check valves.

This line is part of the emergency core cooling system. During a loss of coolant accident, water would be supplied to the core from the core flood tank through check valves CF-Il and CF-12. This would force closed check valve LP-47.

After the core flood tank empties, the RCS pressure would continue to decrease and eventually cooling water would be supplied to the core through check valves LP-47 and CF-12, and check valve CF-11 would be forced closed, Down Stream Between CF-11, Up Stream CF-1I CF-12 CF-12, & LP-47 & LP-47 Design Pressure 2500 2500 2500 Design Temperature 650 300 300 Pipe Size & Schedule 14 in, Sch 140 14 in, Sch 140 14 in, Sch 140 10 in, Sch 140 10 in, Sch 160 *

  • Unit 1 the 10 inch pipe is schedule 140 See tables in Attachment # 1 for additional information.

Pipe Section 4: Configuration for Valves LP-48, CF-13, and CF-14 These three valves, LP-48, CF-13, and CF-14 exist in a section of pipe that directs water from the core flood tank or the LPI pump to the reactor vessel. As long as the RCS pressure remains above 600 psig these three valves remain closed.

LP-48 is a 10 inch check valve and CF-14 and CF-13 are 14 inch check valves.

This line is part of the emergency core cooling system. During a loss of coolant accident, water would be supplied to the core from the core flood tank through ..

check valves CF-13 and CF-14. This would force closed check valve LP-48.

After the core flood tank empties, the RCS pressure would continue to decrease and eventually cooling water would be supplied to the core through LP-48 and CF-14, and check valve CF-13 would be forced closed.

Down Stream Between CF-13, Up Stream CF-13 CF-14 CF-14, & LP-48 & LP-48 Design Pressure 2500 2500 2500 Design Temperature 650 300 300 Pipe Size & Schedule 14 in, Sch 140 14 in, Sch 140 14 in, Sch 140 101in, Sch 140 10 in, Sch 160 *

  • Unit I the 10 inch pipe is schedule 140 See tables in Attachment # 1 for additional information.

Request for Relief 04=ON-0 12 Rev. 1 Page 4 of 16 Pipe Section 5: Configuration for Valves LP-46 and LP-131 LP-46 and LP- 131 are part of the alternate pressurizer spray supply line. This line is located entirely within the reactor building. During normal operations this line is full but isolated. During refueling outages this line is placed in service to cool down the pressurizer.

LP-46 and LP-131 are 1-1/2 inch check valves separated by a piece of 1-1/2 inch schedule 160 stainless steel pipe.

Down Stream Between LP'-46 Up Stream LP-131 LP-46 & LP-131 Design Pressure 2500 2500 2500 Design Temperature 670 300 300 Pipe Size and Schedule 1.5 in, Sch 160 1.5 in, Sch 160 1.5 in, Sch 160 See tables in Attachment # 1 for additional information.

When Oconee was constructed, the RCS boundary was defined as all connected piping, valve bodies, pump casings, heat exchangers, or vessels out to and including the first isolation valve. Therefore, all these sections of pipe were installed as B31.7 Class II. For Section XI implementation, these sections between the first and second isolation valves were identified as ASME Class I. (Reference Attachment 1.)

2.0 Code Requirement The 1989 edition with no addenda of ASMIE B&PV Code Section XI, Table IWB-'

2500-1 Examination Category B-P, Item B15.51, Pressure Retaining Boundary, System Hydrostatic Test. Note # 2 in this table states that the 'pressure retaining boundary during the system hydrostatic test shall include all Class. 1 components within the system boundary'. This test is to be conducted once during the 10 year interval, either at or near the end of the interval.

3.0 Code Requirement From Which Relief is Requested:

IWB-5222 System Hydrostatic Test (a) The system hydrostatic test may be conducted at any test pressure specified in Table IWB-5222-1 corresponding to the selected test temperature, provided the requirements of IWB-5230 are met for all ferritic steel components within

.the boundary of the system (or portion of system) subject to the test pressure.

(See IWA-5245.)

Request for Relief 04-ON-012 Rev. 1 Page 5 of 16 Specifically relief is requested from the requirement to extend the hydrostatic test boundary to include all ASME Class I piping and components.

"Specifically, relief is requested as follows:

1. For piping sections 1, 3, 4, and 5, relief is requested from the requirement to perform the System Hydrostatic Test at the pressure prescribed by IWB-5222.
2. For piping section 2, relief is requested from the requirement to perform a hydrostatic pressure test in accordance with IWB-2500, Table IWB-2500-1, Examination Category B-P, Item B115.51.

4.0 Basis for Relief:

In accordance with the requirements of 10 CFR 50.55a(a)(3)(ii), relief is requested from the requirements IWB-5222(a) and IWB-2500, Table IWB-2500-1 Examination Category B-P, Item B115.51.

The following discussion provides the basis for the relief request and approval of the proposed alternative testing in accordance with the provisions in 16CFR 50.55a(a)(3)(ii) due to the hardship that would be imposed by complying with the Code requirement.

Applying RCS hydrostatic pressure to Piping Sections 1, 2, 3, 4, and 5 of the Class 1 piping listed in Section 1.0 of this request would result in a hardship by exposing station personnel to:

  • Safety Hazards - ranging from physical exposure to temporary connections whose medium is pressurized to 2155 psig (and in some cases at 600 0 F) to their being "stationed" at open manual valves or near vent/drain valves serving as RCS single isolation pressure and temperature barriers in order to maintain the RCS boundary redundant valve protection requirement of 10CFR 50.55(a)(c)(2)(ii) during the test.

" Radiation Exposure - from activities in containment such as carrying, installing, performing test activities and removing hydro pump or temporary jumpers as needed. Further, activities to install and remove scaffold, as necessary, and removal and replacement of insulation and valve internals, as necessary, would result in additional exposure.

In addition, for piping section 2 performing any type pressure test would expose personnel to the hazards described above without a compensating increase in the level of safety and quality. The end results of a leak in this section of pipe is the same as if the second valve was open, the water ends up in the basement.

Request for Relief 04-ON-0 12 Rev. 1 Page 6 of 16 According to Technical Specification 3.4.3, RCS Pressure/Temperature Limitations the following are the minimum RCS temperatures for a RCS pressure of 2155 psig during the performance of a RCS hydrostatic test.

Unit Temperature Unit 1 260 Unit 2 235 Unit 3 235 Pipe Section 1: Between Valves LP-1 and LP-2 Due to the design parameters for the pipe down stream of LP-2, the only way to pressurize the section of pipe between LP-1 and LP-2 is to have valve LP-1 open during the pressure test. There is an interlock in the open circuit for LP-1 which prevents this valve from being opened when the RCS pressure is above 400 psig.

To pressurize the section of pipe between LP-1 & LP-2 from the RCS would require defeating this interlock. Defeating this interlock results in hardship and unusual difficulty since it would require a special and infrequently performed procedure (directions) and could result in additional cool downs and pressure cycles for the reactor coolant system to permit returning the interlock to normal and testing the interlock for proper operation.

Performing a hydrostatic pressure test is, in itself, an infrequent task. Adding to this task the task to bypass interlocks increases the chance for an error.

The RCS is limited to 360 thermal cycles of which about 100 have been used.

Pipe Section 2: Between Valves LP-103 and LP-104 LP-103 and LP-104 are 3 inch gate valves. Since this line is the primary boron dilution flow path, the pipe is open ended down stream of LP-104. Unlike the other sections of pipe described in this request for relief, this pipe does not directly feed another section of pipe. If a leak was to occur between LP-103 and LP-104 the water would end up in the reactor building basement which is where the discharge of LP-104 also terminates.

As described above, performing this hydrostatic pressure test would involve exposing personnel to a potentially hazardous condition without providing any increase in the level of quality and safety over the alternative of doing nothing. A leak between these two valves would result in the piping fluid being directed to the same location, the basement.

For Units I and 2, LP-104 is a 3 inch 1500 pound class valve manufactured by Walworth. The body is made of SA 351 CF8M with a design pressure rating of 2485 psig at 650 degrees F. For Unit 3 this valve is a 4 inch 1500 pound class

Request for Relief 04-ON-012 Rev. 1 Page 7 of 16 valve manufactured by Velan. The body is made of SA-182 Grade F316 with a design pressure rating of 3600 psig at 100 degrees F.

Pipe Section 3: Between Valves CF-i1, CF-12 and LP-47 CF-12 is a 14 inch pressure seal check valve. Hydrostatically testing this section of pipe to the required pressure would require the removal of the check valve bonnet, removal of the valve disc to provide a fill path, and re-assembly of the bonnet so the reactor coolant system (RCS) could be heated to increase pressure.

Once the hydrostatic test was completed, it would be necessary to depressurize the RCS so the check valve disc could be re-installed. This process results in an additional heatup to reach the test conditions followed by another cooldown to reassemble the valve. Performing the hydrostatic test creates hardship by inducing an unnecessary thermal cycle for the RCS. Removing and installing the disc would result in a hardship and unusual difficulty on personnel due to the location of these valves and the difficulty to access the valve bonnet.

An alternate fill path through a 1/2/2 inch tubing connection was considered. During this filling and pressurizing process the RCS would have to be at or near hydrostatic test pressure to ensure the check valve did not open. The fill water would have to be heated to prevent injecting cold water into a hot RCS. The existing equipment is not designed to handle water at elevated temperatures.

Working with water at elevated temperatures and the possibility of a reactivity addition represents hardships to performing the hydrostatic pressure test through this 1/2/2 inch line.

Pipe Section 4: Between Valves CF-13, CF-14, and LP-48 CF-14 is a 14 inch pressure seal check valve. Hydrostatically testing this section of pipe to the required pressure would require the removal of the check valve bonnet, removal of the valve disc to provide a fill path, and re-assembly of the bonnet so the reactor coolant system (RCS) could be heated to increase pressure.

Once the hydrostatic test was completed, it would be necessary to depressurize the RCS so the check valve disc could be re-installed. This process results in an additional heatup to reach the test conditions followed by another cooldown to reassemble the valve. Performing the hydrostatic test creates hardship by inducing an unnecessary thermal cycle for the RCS. Removing and installing the disc would result in a hardship and unusual difficulty on personnel due to the location of these valves and the difficulty to access the valve bonnet.

An alternate fill path through a 1/2 inch tubing connection was considered. During this filling and pressurizing process the RCS would have to be at or near hydrostatic test pressure to ensure the check valve did not open. The fill water would have to be heated to prevent injecting cold water into a hot RCS. The existing equipment is not designed to handle water at elevated temperatures.

Request for Relief 04-ON-012 Rev. 1 Page 8 of 16 Working with water at elevated temperatures and the possibility of a reactivity addition represents hardships to performing the hydrostatic pressure test through this 1/2/ inch line.

Pipe Section 5: Between Valves LP-46 and LP-131 LP-46 is a check valve in the alternate supply line to the Pressurizer Spray Nozzle in the top of the pressurizer. To prevent this valve from opening when performing a hydrostatic test, the reactor coolant system would have to be in a solid condition. Operation of the plant in a solid condition is a hardship on the operators and the plant. The concern with solid plant operations is a small increase in temperature will result in a' significant increase in pressure due to the positive coefficient of expansion for water.

5.0 Alternate Examinations or Tests:

Pipe Section 1: Between Valves LP-I and LP-2 Every cold shutdown the pipe between LP-1 through LP-2 is pressurized to 270 psig or greater.

During plant cool down to cold shutdown, the section of pipe down stream of LP-1 is placed in service and remains in service most of the outage. At least once a day an inspection is done of the reactor building looking for leaks and accumulation of boron. This pipe is in an accessible area so any leakage could be identified during one of these inspections and action taken to fix the leak.

Pipe Section 2: Between Valves LP-103 and LP-104 The pipe between LP-103 and LP-104 was not pressurized.

The consequence of not pressure testing this section of pipe is minimal because the purpose of this line is to direct water to the floor of the reactor building in the event of a LOCA. If this section of pipe, down stream of LP-103, were to fail, the worst to occur would be for this line to leak to the floor when placed in service. The reactor coolant would still end up on the floor of the reactor building and eventually get to the sump to be recycled for cooling water.

Pipe Section 3: Between Valves LP-47, CF-11, and CF-12 At the end of each refueling outage the piping between LP-47, CF-11, and CF-12 is tested to verify an open flow path. During the fuel cycle, this section of pipe is pressurized to approximately 600 psig. If leaks were to develop in this line it would be detected by having to fill the core flood tank or having to add nitrogen more

Request for Relief 04-ON-012 Rev. 1 Page 9 of 16 frequently during plant operation, by visual observation of boron during a reactor building inspection tour, or increased normal sump rate.

Pipe Section 4: Between Valves LP-48, CF-13, and CF-14 At the end of each refueling outage the piping between LP-48, CF-13, and CF-14 is tested to verify an open flow path. During the fuel cycle, this section of pipe is pressurized to approximately 600 psig. If leaks were to develop in this line it would be detected by having to fill the' core flood tank or having to add nitrogen more frequently during plant operation, by visual observation of boron during a reactor building inspection tour, or increased normal sump rate.

Pipe Section 5: Between Valves LP-46 and LP-131 At the beginning of each refueling outage, the section of pipe between LP-46 to LP-131 is place in service at 300 psig to cool down the pressurizer. During this same time there are several reactor building inspection tours looking for leaks. Any boron is cleaned and the area inspected for corrosion, and the source of boron determined.

6.0 Justification for the Granting of Relief Pursuant to the requirements of 10CFR50.55a(a)(3)(ii), compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety As described above, performing these tests at the required pressure present personnel safety concerns, and creates excess thermal cycles on the Class 1 piping and components.

As described in Section 5.0, a system leakage test was performed on piping sections 1, 3, 4, & 5 at a pressure less than that required for the hydrostatic pressure test. For piping section 2 no hydrostatic or leakage test was performed. A system leakage test at nominal operating pressure has already been established as an acceptable pressure test method based on the endorsement of Code Case N-498-1 in Regulatory Guide 1.147, Revision 12. Since the test pressures used for the piping in this request were lower than nominal operating pressure, N-498-1 does not directly apply.

For the piping sections that were pressure tested, the hydrostatic test does not provide an increase in quality and safety that compensates for the hardship or unusual difficulty. Assuming a crack or defect would behave like a fixed area orifice, it can be shown that the leakage varies proportional to the square root of the ratio of the differential pressure (ref. Crane Technical Paper # 410). For example, if a leakage of L was projected to be present at a pressure of 2155 psig that same leak would exist at a reduced pressure with a reduced magnitude of:

27:0 x L = 0.35L

" 2155

Request for Relief 04-ON-012 Rev. .1 Page 10 of 16 Inspections that reveal no leakage at 270 psig (where 35% of the leakage produced by 2155 psig would be present for detection during a VT-2 inspection) gives a reasonable assurance that no leakage would be present at 2155 psig. Through wall leakage that would occur at the higher pressure such as RCS pressure, would be expected to reveal itself at lower pressures. It may take longer for some leaks to propagate through the pipe wall at lower pressure and the amount of leakage from a lower pressure test would be reduced, but the leakage would still be visible for a VT-2 examination.

The reduced pressure testing performed as an alternate pressure test for Units 1, 2, &

3 piping covered by this request ranged from 270 psig to approximately 2155 psig except for the section of pipe between LP-103 & LP-104, which was not pressure tested. Pressure in that range is sufficient to detect any through wall leakage in the tested piping and components during the performance of the alternative tests.

Except for the section of pipe between LP-103 & LP-104, each alternate pressure test indicated in this request was performed using a reduced pressure for the specified hold time followed by visual examination (VT-2) to verify no through wall leaks exist in the tested pipe. The test boundaries included all piping described in Part 1.0 of this request.

There are also other plant measures that provide additional evidence that an appropriate level of quality and safety is being maintained:

First, welds within these sections of pipe are within the Class 1 ISI boundary and thus are candidates for both volumetric and surface examinations as required by the ASME Section XI Code.

Second, if a leak were to develop in any of the piping described in this request it could be detected by various means available to the operators, increased frequency for pumping the normal reactor building sump, increased frequency of filling the core flood tanks, and observing changes in the radiation levels in the reactor building are just some of the ways of monitoring for reactor building leakage. In addition, plant Technical Specification 3.4.13 requires that at least once every 72'hours when above MODE 5 a RCS operational leakage determination be performed. Technical Specifications limit the amount of unknown RCS leakage to 1 gpm. If this limit is exceeded then the source must be identified or the reactor must be in MODE 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and MODE 5 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

During reactor shutdown for outages, numerous inspections are made in the reactor building to look for indications of leakage. Any identified leak is evaluated and dispositioned.

Finally, all of the piping described in this request is inside the reactor building. The reactor building has been designed to contain any leakage and serves as a radiation barrier to protect the health and safety of the public.

Request for Relief 04-ON-012 Rev. I Page 1 Iof 16 Based on the hardships without a compensating increase in quality and safety discussed above, approval of this request is requested. The Nuclear Regulatory Commission has granted similar relief to Catawba Nuclear Station (04-CN-004),

Kewaunee Nuclear Power Station, McGuire Nuclear Station (01-03), and Sequoyah Nuclear Station in response to their request for relief from certain ASME Section XI Class 1 pressure testing.

7.0 Implementation Schedule:

Relief is requested for the third ten-year interval of the ISI schedule. The third interval ended as follows:

Unit 1 January 1, 2004 Unit 2 September 9, 2004 Unit 3 January 2, 2005 8.0 Other Information:

This relief request was discussed with the following individuals:

Aaron Best Core Flood System Engineer - ONS Basil Carney Repair and Replacement Engineer - ONS Charles Henson Weld Technical Support Supervisor - ONS Frank Eppler Low Pressure Injection System Engineer - ONS James Boughman Pressure Test Program Manager - GO Mark Ferlisi Containment ISI Program Engineer - GO Mark Pyne Section XI Inspection Program Manager - GO Scott Manning RCS System Engineer - ONS Vick Dixon Pressure Testing - Section XI - ONS

Request For Relief 04-ON-012 Rev. 1 Page 12 of 16 Attachment # 1 The following information was copied from the latest revision of the Oconee UJFSAR.

3.1.1 Criterion 1 - Quality Standards (Category A) 3.1.1.1 Oconee QA-1 Program To meet the requirements of IOCFR50 Appendix B, Oconee has defined its QA-1 program. The QA-1 program shall be applied to the "essential systems and components" listed above. The scope of these systems and components is provided in greater detail below. The QA-I program shall also be applied to the Reactor Protective System, and shall be applied to any systems and components committed to the NRC as being classified as1QA-1 per any correspondence subsequent to the orginal QA-I licensing basis.

Therefore, the general criteria used to determine if a SSC is QA-I is divided into two categories:

First category - provides general QA-1 criteria based on the original licensing basis of ONS, and Second Category - provides general criteria for SSCs that were added to the QA-1 licensing basis after issuance of the original operating licenses for ONS.

First Category, Original Oconee QA-1 Licensing Basis This first category includes the integrity of SSCs essential to prevention and mitigation of the Large Break LOCA coincident with loss of offsite power for the following five SSCs:

1) Reactor Coolant System, 2) Reactor Vessel Internals, 3) Reactor Building, 4)

Engineered Safeguards System, and 5) Emergency Electric Power Sources. In addition,

6) Reactor Protective System, another system not addressed in FSAR Section 3.1.1, was interpreted to be included in the QA-1 scope, even though not listed.

Clarification regarding the six SSCs identified above is provided below.

1. Reactor Coolant System From a quality assurance perspective, the Reactor Coolant System consists of all connecting piping, valve bodies, pump .casings, heat exchangers, or vessels out to and including the first isolation valve. The integrity of the pressure boundary of the connecting piping, valve bodies, pump casings, heat exchangers, or vessels is the function which determines applicability of the quality assurance program.

Request For Relief 04-ON-012 Rev.. I Page 13 of 16

!Attachment # 1 3.2.2.1 System Classifications Class I This class is limited to the Reactor Coolant System (RCS) and Reactor Coolant Branch lines, as described herein. The Reactor Coolant Branch lines include connecting piping out to and including the first isolation valve. This section of piping is Class I in material, fabrication, erection, and supports and restraints. A Class I analysis of the piping to the first isolation valve has been completed for the following systems:

1. High Pressure Injection (Emergency Injection)
2. High Pressure Injection (Normal Injection)
3. High Pressure Injection (Letdown)
4. Low Pressure Injection (Decay Heat Removal Drop-line)
5. Low Pressure Injection (Core Flood)
6. Reactor Coolant Drain Lines
7. Pressurizer Spray
8. Pressurizer Relief Valve Nozzles Modifications that affect the Reactor Coolant System and the Class I portion of the branch lines must demonstrate that the impact on the Class I piping is acceptable. The impact may be assessed by performing a Class I analysis or by other conservative techniques to assure Class I allowable limits are not exceeded. Isolation valves can be either stop, relief, or check valves. Piping 1 inch and less is excluded from Class I.

Request For Relief 04-ON-012 Rev. 1 Page 14 of 16 Attachment # 1 PIPE MATERIAL, SIZE AND SCHEDULE Segment Pipe Header Pipe Diameter & Segment Length Segment Length Segment Length Material Schedule Unit 1 Unit 2 Unit 3 LP LP-2 SA 376 12 in Sch. 140 7 ft 6 in. 5 ft 6 in 5 ft 6 in TP 304 LP-103 - LP-104 SA 376 3 in. Sch. 160 6 ft 3 ft 3 ft 6 in TP 304 4 in. Sch. 160 1 ft CF-12 - LP-47 SA 376 10 in. Sch. 140 162 ft 158 ft 160 ft TP 304 14 in. Sch. 140 CF-14 - LP-48 SA 376 10 in. Sch. 140 110 ft 108 ft 108 ft TP 304 14 in. Sch. 140 LP-46 - LP-131 SA 376 1.5 in. Sch. 160 100 ft 94 ft 98 ft TP 304 This table does not include the length of the small diameter vents and drains that are attached to the header pipe.

OPERATING CONDITIONS Normal Design Accident Segment Press. Temp. Press. Temp. Press. Temp.

LP-1 - LP-2 0* Ambient 2500 300 0 400 LP-103 - LP-104 0

  • Ambient 2500 4250 **

CF-11, CF-12, LP-47 600 Ambient 2500 3000 600 3000 CF-13, CF-14, LP-48 600 Ambient 2500 3000 600 300 0 LP LP-131 0* Ambient 2500 3000 **

  • Normally isolated
    • RCS pressure at time of activation

Request For Relief 04-ON-012 Rev. 1 Page 15 of 16 Attachment # 1 NUMBER OF WELDS NOT BEING PRESSURE TESTED TO RCS PRESSURE Unit 1 Unit 2 Unit 3 Pipe Size # of Welds Pipe Size # of Welds Pipe Size # of Welds LP-1 and LP-2 1/2" 6 '/2" 11 1/2" 5 45 3"/4 5 3/4 5 12" 2 12" 2 12" 2 LP 103 and LP-104 1/2" 6 1/2 4

/4" 2 3/4 2 3" 2 3" 3 3" 3 4" 2 CF-11, CF-12, & LP-47 14" 12 14" 12 14" 11 10" 15 10" 15 10" 15 19 6 1" 6 1" 7 CF-13, CF-14, & LP-48 14" 6 14" 6 14" 6 10" 14 10" 14 10" 13 1" 4 1" 5 1" 5 LP-46 and LP-131 1.5" 28 1.5" 19 1.5" 23 1" 3 1" 2 1" 2 I

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Request For Relief 04-ON-012 Rev. I Page 16 of 16 Attachment # 2 Portions of Applicable Oconee Flow Diagrams Five Sheets- each showing one of the sections.

PIPE PIPE 1St PIPT ISl LINE vai DESIGN DESIGN NO. CLAS PNEUUE tDC.. MTERIA. VC NO. SOL "M. CLASS NOTE a 4 C 3ISI 2WF a36s 153.3 WOESa B NOTES S 5 S 20 PM 3wF 36 151.2 NOTEa a NOTE3 2 S S 3 479 3110/595 3110 36VF/25VP SS 351.2 WOE32 a NOTE7 A 12 - A 25M PS16 £WF SS Mo9.31 WE 7 A 23-E - KDO.- 3WF-SS- 151.4-NOWE 6- -

29 6 3 PM1 369V 3 393.2 504. 26 B NOTE a -

153.4 WOE 0 - NO1TE l -

27 Of AT". 3WF 83 3WF 95 361.2 WCE to 9 NOTE is 34 Kc 36SPItO p ss 151.3 NOTE S 0 NOTE a B 35 eK am-. 36PF -NOTE to -

II 36 1-S 305 PS - 3WF SS3 361.2-NOTE 10- 0 37 s 3" PM1 31WF 36 391.2 NOTEt3 S SMx. MU 0 43 3 2500 P310 OWF 33 1551.2 SC4. 366 a NOTE 7 A 47 I 2599 P31 31F 98 353.2 WYTE 7, a NOTE 0 c 49 c 3 PSIC 2WF 93 051.3 NOTEa c -NOTE a -

59---E-3 PS - 2W9F- So- 953.4-NOE 0- - NOTE 13 a 51 oC 3308M 39SF 93 301.3 NOTE13 S amH. 211 B 52 S amO. 3WF sI 151.2 504. 20 9 NOTES S 50 C 3 510 ISwF s3 353.3 NOTESa NOYE -

I9 ATPL IWF 4515.4 NOE S -

-SD. 435 -

-S0 300 3110 3WF - SS-361.2-SCH. 4416-s8 3M310I 3WF 55 M6.3 5SM.59 9 SCM. of 9 99 Kc 77 0 365 PM2 3WF as 363.2 904. Is 0

- NOTE7 A 57-I 3 3510 425&F---SS- 13.2-- NOTE7 - A sm.o a A 9S 8 250 P333 425'F $5 2701.1 S0I. 90 A 304.459 sm.'r me 357 1 3310 M WF 53 361.4 NoTE Is 1E 471 PIO/M P31 3WIZWF5V 93 363.4 SM. 4K sm*. Mol 113 -

304.44 -

133 8 4733110/US PMO 3MP/25rF IS 301.4 UK4.406 -

314. H 49 P9RO 425*F 5 381.4 SM4. 46 -

NOTET I' 17 g3~1 CO 3124 C ýV_40_ tc3/!D 4 ,

3LP- I

(

LINE DAM DSIGN CESIGN pWE PIPE Is

  1. a LAS IPIES52M TEW. -MMA 9CC. 04 SOL. NCL M.A 4 C-- NPIM 2-F 99 152.3 ,m e I o

5 8 m 3W1F U 253.2 OuWE o 8 a 4?g esW.M= PSIG 3WF12SF S1 3"1.2 " M2 t2 32 A 2111 . 6FF .ss Mi. I $o0TE7I 23- E- Um.L 31F - S1 151.4- NDTE0 -

I 26 27 S4 s S

H 3M59MW 3651 ATI.

3FF 3WF 21F ss i11 311.2 251.4 391.3

$CH.2f laE8a wmOT is S

35 w @I= 3FF ss 151.3 OT a 36-3 - 365 9 3- 3FF- 93-361.2-MOTE tf-- B 3*a I PM591 3FF 08 313.2 MM10 423 250 P= 1FF SS 1M1.2 S10. 169 B 47 6 m =193 3FIF 96 We11. 2 MOTE 7 A 49 C 3--M 21FF s$ 352.3 WYEE c 50- E- 3 M1Q2-- -FF-S- 151.4- NOTE0 -

51 eK 315913 3FF is 232.3 MOTE 13 a 52 Sa Ow. 3FF 96 351.2 SM0. 21 B so C 2963 15FF is 151.3 NOTE 9 S w f A*,14 25FF 66 151.4 DOTES -

go- -*3* 0 9M - 31FF t-'361.2- U. 4$6 -

610 4K 30w 3FOF $5 301.3 50M. as 9 a7 316910 31FF is 361.2 104. 2t a cT~D- Ca2

-~6 2511 P13 35FF__ 91 2792. I 2- NOTEI3~

U- a__ A E. 31659620 WF 61$ 371.4 10.1 A 11 9 470 P135/l PBX0 3WFri3F ss 391.4 1CH. 495S 113 a 47P10mumPraw 3WFj25FF $1 331.4 SM4.499 13414 411 "11 42F It 321.4 304. 49 H¸

-SC '-

F - ------(g )

F;,

TAW* FLOOD CC ZA RC CO)P.

2 DRAIN HEAOR PIBF30

-5 , , -ý-, 0 -1, -P --

f fi 94- /3 4-- / ftýesl 't TO REACTOR VESS p oDESIGN PARAM*TERS LINE UXKE DESIGN DESI G PIPE PIPe 1St NO. CLASS PRESSURE TE9. MTERIAL. SPEC. ND. SOL NO. CLASS 9 B 70 PSIG 3000F SS 601.2 NOTE S 8 Is A 2586 PSIG 38eF SS 1581. 1 NOTE 2 A 11 B 2509 PSIG 3@FF SS 1501.2 NOTE 2 B Is E 156 PS31G 30-F SS 151.4 NOTE 3 -

24- F 10 PSIG - 2'F-1- CS - 158.4 - SCH. 168- B 29 AC 2500 PSIG 36FF SS 1581.3 NOTE 2 A 36 BC 25M PSIG 3W*F Ss 1581.3 NOTE 2 a 31 BC 710 PSI6 300'F SS 6:81.3 NOTE S B 42 E ISO PSIG 3WF CS 158.4 NOTE 8 45- E - 7M6 PSIG - 3 WF - SS - 601.4 - NOTE 6 --

55 BC 786 PSIS 3WF SS 601.3 SCH. m8S B 57 BC 700 PSIG 3WF SS 681.3 SC). 48 a 61 E 700 PSIG 30FF SS 69.4 SCH.8 S -

88 BC 786 PSIG 666F SS 601.3 SCH. BeS B NOTE 13 NOTE 13 II0- 8 - 256 PSIG-- 3OrF - SS-2701. la - 8 1e8 E 2500 PSIG 366F SS 1581.4 NOTE 2 -

9 E 25M PSIG 300' F SS c(2781.2..

1561.3 SCH,. - MIS

=W c

FC CP. TO

  • AIN DRAIN HEADE'R VASTE TAN

,J0e-

ý14 TO REACTORX VESSEL 4 DESIGN PARAMETERS LINE BxE DESIGN DESIGN PIPE PIPE ISI NO. C A';s PRESSURlE TIEP. MATERIN_ SPEC. NO. SOi. NO. CLASS 9 B 789 PSIG 30'F SS 681.2 NOTE 6 B 10 A 250 PSIG 38FF SS 1501.1 NOTE 2 A 11 a 2569 PSIG 300F SS 1581.2 NOTE 2 a 16 E 150 PSIG 38BF SS 151.4 NOTE 3 -

24- F 2ee PS5G - 2M*F CS 150.4 -- SCH. 168--B 29 AC 2600 PSIG 30*F SS 30 1581.3 NOTE 2 A BC 258 PSIG 301F SS 1501.3 NOTE 2 a 31 BC 789 PSIG 3WF SS 601.3 NOTE 6 B 42 E 1ee PSlG -- 380F CS 159.4 -- NOTE 8 -

30&'F - SS - 6,01.4 NOTE 6-- -

45- E 799 PSIG 55 SC 789 PSIG 3Oe F 5S 601.3 SCH. 8SS B 57 BC 7M PSIG 389F SS 601.3 SCl. 48 B 6a E 789 PSlG 3O9 F 53 68.4 SCH. 803 -

a8 8C 78 PSIG 65rF SS 601.3 SCH. SOS B NOTE 13 NOTE 13 Ile- B 259 PSIG -309F-- 2781. le . B E SS--

lee 250M PSIG 38FF ss 1501.4 NOTE 2 -

109 E 2W09 PSIG 3WF SS 1501.3 e {2701.2a So'. 8S1

2LP- 131 (1" RRC-205 1 /-

2 P LS/GN PIPE PIPE ISI 6UKE DESIGN 1T)4. FIIRTRIAL SPEC. NO. sC1H. NO. CLASPS LIPE NO. ClASS PRESSURE 670]F SEmW BQ N NOTE 10 A

! A 250 PS1G A A 2.500 PSIG 670'S SE&W 5k&W SC54140 2 A 25OB PS10 650"F BE 8k B&W S561 160 3 A A A 2500 PSIO 6I0"F SO 4504. & SCH l60 5 t1O -- A S AC--2500 PSIO--670"F -- S- 1501. 3----SCH 670'F Es 1501. 3 SON 160 8 7 BC 2500 PS10 55O"F SS 1501.3 SCH 160 A 10 AC 2500 P5I0 3800" SS 1501. 2 NOTE 3 A 17 5 2500 PSIG 300°F Es 1501.3 NOTE 3 A 19 SC 2500 PSlG C---700 P51G -500'S -- SS -- 601. 3 - NOtE 2-.----C 21 --

2701. 21 2 N,OTE 150 B 27 9 2500 PS2C 300'S -CHSS 150.

300'P SE 2701. t NOTE 2 A 40 6 2500 P61G

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