Text

ASME Section XI Inservice Inspection Program, End of Intervals System Pressure Testing
	ML050180495
Person / Time
Site:	Surry, North Anna
Issue date:	01/10/2005
From:	Hartz L; Virginia Electric & Power Co (VEPCO)
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	04-766
	Download: ML050180495 (39)
	v • d • e

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 January 10, 2005 United States Nuclear Regulatory Commission Serial No.: 04-766 Attention: Document Control Desk NL&OS/ETS: Rev. 1 Washington, D. C. 20555 Docket Nos.: 50-338/339 50-280/281 License Nos.: NPF-4/7 DPR-32/37 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

NORTH ANNA POWER STATION UNITS 1 & 2 SURRY POWER STATION UNITS 1 & 2 ASME SECTION Xi INSERVICE INSPECTION PROGRAM END OF INTERVALS SYSTEM PRESSURE TESTING North Anna Units 1 and 2 are currently in the second period of the third Inservice Inspection (ISI) interval. Surry Units 1 and 2 are currently in the first period of the fourth ISI interval. The attached requests for relief are necessary for certain pressure tests and associated visual examinations required at or near the end of the ISI interval on ASME Class 1 components for each of the units. The relief requests address components and piping considered part of the extended Class 1 boundary and only require pressurization at or near the end of the ISI interval. Pursuant to 10 CFR 50.55a(a)(3)(i) or (ii), Dominion requests relief from the applicable ASME Code pressure testing requirements for North Anna and Surry. To support the Staff's review of the relief requests, a CD with the associated system drawings is also attached.

If approved, the relief requests for the system pressure test program that are provided, will be appropriately added to the North Anna and Surry ISI Plans in lieu of creating separate plans for system pressure tests. The ISI Plan for each unit addresses all Class 1, 2, and 3 components and their supports for examination and pressure testing.

If you have any questions or require additional information, please contact Mr. Thomas Shaub at (804) 273-2763.

Very truly yours, Leslie N. Hartz Vice President - Nuclear Engineering Attachments Commitments made in this letter: None < 1'

cc: U.S. Nuclear Regulatory Commission (without CD)

Region II Sam Nunn Federal Atlanta Center 61 Forsyth Street, SW Suite 23T85 Atlanta, Georgia 30303 Mr. M. S King (without CD)

NRC Senior Resident Inspector North Anna Power Station Mr. N. P. Garrett (without CD)

NRC Senior Resident Inspector Surry Power Station Mr. S. R. Monarque NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 8-H12 Rockville, MD 20852 Mr. J. E. Reasor, Jr. (w/out attachments)

Old Dominion Electric Cooperative Innsbrook Corporate Center Suite 300 4201 Dominion Blvd.

Glen Allen, Virginia 23060 Mr. M. Grace (w/out CD)

Authorized Nuclear Inspector North Anna Power Station Mr. R. A. Smith (w/out CD)

Authorized Nuclear Inspector Surry Power Station

Attachment 1 Letter Serial No.04-766 Surry Power Station Units Relief Requests 1 and 2 Virginia Electric and Power Company (Dominion)

SURRY UNIT 1 SPT 005,006 and 007

SPT-005 IDENTIFICATION OF COMPONENTS System: Residual Heat Removal (RHR) and Safety Injection (SI)

Components:Class 1 components and piping between:

1-SI-109, 1-SI-107 and 1-SI-HCV-1850B on drawing 11448-CBM-089B-4, sheet 1 of 4 1-SI-130, 1-SI-128 and 1-SI-HCV-1850D on drawing 11448-CBM-089B-4, sheet 2 of 4 1-RH-MOV-1720A on drawing 11448-CBM-087A-4, sheet 2 of 2 1-SI-147, 1-SI-145 and 1-SI-HCV-1850F on drawing 11448-CBM-089B-4, sheet 3 of 4 1-RH-MOV-1720B on drawing 11448-CBM-087A-4, sheet 2 of 2 ISI Class: 1 CODE REQUIREMENT IWB-5222(b) (1998 Edition through the 2000 Addenda of ASME Section XI) requires the system leakage test shall be conducted at or near the end of the inspection interval to extend to all Class 1 pressure retaining components within the system boundary. IWB-5221(a) states the system leakage test shall be conducted at a pressure not less than the pressure corresponding to 100% rated reactor power.

Ill. BASIS OF REQUEST FOR RELIEF Surry Unit 1 is in the fourth ISI interval utilizing the 1998 Edition through the 2000 Addenda of ASME Section Xl. The unit is currently in the first ISI period.

Normal reactor coolant pressure at 100% rated power is approximately 2235 psig. The piping in question is separated from this reactor coolant pressure by a single check valve, and as such does not normally see this pressure. Part of the area in question is pressurized during normal operation to approximately 660 psig from the passive safety injection accumulators. The accumulator pressure is monitored from the control room throughout the operating cycle. An external pressurization source would be necessary to meet requirement of normal reactor coolant pressure. Since the check valve would be part of the test boundary, a pressure differential would be required between the reactor coolant system and the area in question to maintain check valve closure. Maintaining the differential

pressure and ensuring no test fluid intrusion into the reactor coolant system (reactivity control issue) is considered unusually difficult to meet with no compensating increase in quality or level of safety when considering the alternative below.

IV. ALTERNATE PROVISIONS The areas in question are examined (VT-2) each refueling as part of the normal Class 1 system leakage test (normal valve line-up) for evidence of leakage.

Except for the areas between,1 -SI-108 and 1-SI-HCV-1850B,1 -SI-129 and 1-SI-HCV-1850D, and 1-SI-146 and 1-SI-HCV-1850F, the areas will be examined at that time at the safety injection accumulator nominal operating pressure.

Additionally, except for the areas between, 1-SI-108 and 1-SI-HCV-1850B, 1-SI-129 and 1-SI-HCV-1850D, and 1-SI-146 and 1-SI-HCV-1850F, the components are under safety injection accumulator pressure during normal operation.

Through-wall leakage for these areas would be identified by the control room through their monitoring of the pressure on the safety injection accumulators.

The areas between 1-SI-108 and 1-SI-HCV-1850B, 1-SI-129 and 1-SI-HCV-1850D, and 1-SI-146 and 1-SI-HCV-1850F would also be examined at or near the end of the interval by using an external pressurization source, or by opening the isolation valves separating the lines from the safety injection accumulator pressure. The test pressure would again correspond to the safety injection accumulator nominal operating pressure.

In accordance with 10 CFR 50.55a(a)(3)(ii) Dominion requests that the tests described above be alternatively performed for the fourth inspection interval.

SPT-006 I. IDENTIFICATION OF COMPONENTS System: Residual Heat Removal (RHR)

Components:Class 1 components and piping between:

1-RH-MOV-1700 and 1-RH-MOV-1701 (RHR suction) on drawing 11 448-CBM-087A-4, sheet 1 of 2 ISI Class: 1 II. CODE REQUIREMENT IWB-5222(b) (1998 Edition through the 2000 Addenda of ASME Section Xl) requires the system leakage test shall be conducted at or near the end of the inspection interval to extend to all Class 1 pressure retaining components within the system boundary. IWB-5221(a) states the system leakage test shall be conducted at a pressure not less than the pressure corresponding to 100% rated reactor power.

III. BASIS OF REQUEST FOR RELIEF Surry Unit 1 is in the fourth ISI interval utilizing the 1998 Edition through the 2000 Addenda of ASME Section XI. The unit is currently in the first ISI period.

Normal reactor coolant pressure at 100% rated power is approximately 2235 psig. The piping in question is separated from this reactor coolant pressure by a single closed valve, and as such does not normally see this pressure. Opening valve 1-RH-MOV-1700 is prevented by a pressure interlock, which prevents opening, when pressure in the reactor coolant system is above 465 psig. The interlock protects the low pressure RHR system from being over-pressurized by the higher pressure reactor coolant system. There is no other valve that would allow pressurization of the area from an external source. The system design prevents Code compliance in this area and it is therefore considered a hardship, since a plant modification would be needed, to meet code requirements.

IV. ALTERNATE PROVISIONS The area in question is examined (VT-2) each refueling as part of the normal Class 1 system leakage test (normal valve line-up) for evidence of leakage.

Additionally, it is proposed that the area be examined as part of the Class 2 system leakage test pressure boundary using the Class 2 test requirements associated with the adjoining Class 2 piping. This would result in an additional

pressure test each period at the Class 2 RHR system nominal operating pressure with the associated visual VT-2 examination.

It is requested per 10 CFR 50.55a(a)(3)(ii) that the tests described above for the fourth inspection interval be alternatively performed.

SPT-007 IDENTIFICATION OF COMPONENTS System: All Class 1 systems Components:Class 1 components and piping located in the extended portion (beyond the first normally closed valve) of the Class 1 boundary

11. CODE REQUIREMENT Table IWB-2500-1 Category B-P, note 2, 'The system leakage test (IWB-5220) shall be conducted prior to plant startup following reactor refueling outage," and IWB-5221 (a) states the system leakage test shall be conducted at a pressure not less than the pressure corresponding to 100% rated reactor power.

IWB-5222(b) requires the system leakage test shall be conducted at or near the end of the inspection interval to extend to all Class 1 pressure retaining components within the system boundary.

Code: Above references based on ASME Section Xl, 1998 Edition through the 2000 Addenda III. BASIS OF REQUEST FOR RELIEF Surry Unit 1 is in the fourth ISI interval utilizing the 1998 Edition through the 2000 Addenda of ASME Section Xl. The unit is currently in the first ISI period.

Normal reactor coolant pressure at 100% rated power is approximately 2235 psig. At or near the end of the interval the system leakage test at 2235 psig is to be extended to all Class 1 components. Note 2 in Table 2500-1, Category B-P restricts performance of the required system leakage test to prior to plant startup following a reactor refueling outage. The note 2 requirement for the normal system leakage tests (non 10-year, every refueling) can be met, but it is considered an unnecessary burden for the end of interval (10-year) test requirement on the extended Class 1 boundaries.

A significant number of extended Class 1 boundary areas are isolated from RCS pressure by check valves. Pressurizing the extended Class 1 boundaries could lift these check valves off the seats. Current practice is to test many of these check valves for positive closure following any potential opening to verify reactor coolant system pressure boundary integrity. These check valve tests are normally run with the plant shutdown and with a lower reactor coolant pressure.

Requiring the ASME Section Xl 10 year test at the note's prescribed time would

put the plant in a situation where it would possibly be required to reduce pressure again following the ASME Section Xl test to repeat check valve closure testing.

Secondly, the test in question is performed only once in 10 years. At all other times the system leakage test is performed with normal valve line-up examining the extended boundary at a pressure less than 2235 psig. The "when" at the end of the interval to verify piping integrity of the extended Class 1 boundary is not technically relevant. Tests performed anytime at or near the end of the interval prior to startup at test pressure would verify piping integrity. This is exemplified by Code Case N-498 (-1,-2,-3,-4), which was used with earlier ASME Section Xl codes as an alternative for the Class 1 end of interval (10 year) test (hydrostatic).

The code case does not have the same restriction, stipulating testing following a refueling outage. Inquiry IN04-002 from ASME indicates the N-498 (-1,-2,-3,-4) test may be conducted anytime prior to startup without the "following refueling outage" stipulation. Considering the information above, it is our position performing these extended boundary tests at test pressure at or near the end of the interval at anytime prior to startup would be equivalent in quality and safety.

IV. ALTERNATE PROVISIONS The system leakage test at or near the end of the interval of the extended Class 1 boundary will be performed prior to startup without the "following refueling outage" stipulation.

It is requested per 10 CFR 50.55a(a)(3)(i) that the test described above for the fourth inspection interval be alternatively performed.

SURRY UNIT 2 SPT 004, 005 and 006

SPT-004 IDENTIFICATION OF COMPONENTS System: Residual Heat Removal (RHR) and Safety Injection (SI)

Components:Class 1 components and piping between:

2-SI-109, 2-SI-107 and 2-SI-HCV-2850B on drawing 11548-CBM-089B-3 sheet 1 of 4 2-SI-130, 2-SI-128 and 2-SI-HCV-2850D on drawing 11548-CBM-089B-3 sheet 2 of 4 2-RH-47 on drawing 11548-CBM-087A-3 sheet 2 of 2 2-SI-147, 2-SI-145 and 2-SI-HCV-2850F on drawing 11548-CBM-089B-3 sheet 3 of 4 MOV-2720B on drawing 11548-CBM-087A-3 sheet 2 of 2 ISI Class: 1 II. CODE REQUIREMENT IWB-5222(b) (1998 Edition through the 2000 Addenda of ASME Section Xl) requires the system leakage test shall be conducted at or near the end of the inspection interval to extend to all Class 1 pressure retaining components within the system boundary. IWB-5221(a) states the system leakage test shall be conducted at a pressure not less than the pressure corresponding to 100% rated reactor power.

Ill. BASIS OF REQUEST FOR RELIEF Surry Unit 2 is in the fourth ISI interval utilizing the 1998 Edition through the 2000 Addenda of ASME Section Xl. The unit is currently in the first ISI period.

Normal reactor coolant pressure at 100% rated power is approximately 2235 psig. The piping in question is separated from this reactor coolant pressure by a single check valve, and as such does not normally see this pressure. Part of the area in question is pressurized during normal operation to approximately 660 psig from the passive safety injection accumulators. The accumulator pressure is monitored from the control room throughout the operating cycle. An external pressurization source would be necessary to meet requirement of normal reactor coolant pressure. Since the check valve would be part of the test boundary, a pressure differential would be required between the reactor coolant system and the area in question to maintain check valve closure. Maintaining the differential

pressure and ensuring no test fluid intrusion into the reactor coolant system (reactivity control issue) is considered unusually difficult to meet with no compensating increase in quality or level of safety when considering the alternative below.

IV. ALTERNATE PROVISIONS The areas in question are examined (VT-2) each refueling as part of the normal Class 1 system leakage test (normal valve line-up) for evidence of leakage.

Except for the areas between, 2-SI-1 08 and 2-SI-HCV-2850B, 2-SI-1 29 and 2-SI-HCV-2850D, and 2-SI-146 and 2-SI-HCV-2850F, the areas will be examined at that time at the safety injection accumulator nominal operating pressure.

Additionally, except for the areas between, 2-SI-1 08 and 2-SI-HCV-2850B, 2-SI-129 and 2-SI-HCV-2850D, and 2-SI-146 and 2-SI-HCV-2850F, the components are under safety injection accumulator pressure during normal operation.

Through-wall leakage for these areas would be identified by the control room through their monitoring of the pressure on the safety injection accumulators.

The areas between 2-SI-108 and 2-SI-HCV-2850B, 2-SI-129 and 2-SI-HCV-2850D, and 2-SI-146 and 2-SI-HCV-2850F would also be examined at or near the end of the interval by using an external pressurization source, or by opening the isolation valves separating the lines from the safety injection accumulator pressure. The test pressure would again correspond to the safety injection accumulator nominal operating pressure.

In accordance with 10 CFR 50.55a(a)(3)(ii) Dominion requests that the tests described above be alternatively performed for the fourth inspection interval.

SPT-005 I. IDENTIFICATION OF COMPONENTS System: Residual Heat Removal (RHR)

Components:Class 1 components and piping between:

MOV-2700 and MOV-2701 (RHR suction) on drawing 11 548-CBM-087A-3 sheet 1 of 2 ISI Class: 1 II. CODE REQUIREMENT IWB-5222(b) (1998 Edition through the 2000 Addenda of ASME Section XI) requires the system leakage test shall be conducted at or near the end of the inspection interval to extend to all Class 1 pressure retaining components within the system boundary. IWB-5221 (a) states the system leakage test shall be conducted at a pressure not less than the pressure corresponding to 100% rated reactor power.

Ill. BASIS OF REQUEST FOR RELIEF Surry Unit 2 is in the fourth ISI interval utilizing the 1998 Edition through the 2000 Addenda of ASME Section Xl. The unit is currently in the first ISI period.

Normal reactor coolant pressure at 100% rated power is approximately 2235 psig. The piping in question is separated from this reactor coolant pressure by a single closed valve, and as such does not normally see this pressure. Opening valve MOV-2700 is prevented by a pressure interlock, which prevents opening, when pressure in the reactor coolant system is above 465 psig. The interlock protects the low pressure RHR system from being over-pressurized by the higher pressure reactor coolant system. There is no other valve that would allow pressurization of the area from an external source. The system design prevents Code compliance in this area and it is therefore considered a hardship, since a plant modification would be needed, to meet code requirements.

IV. ALTERNATE PROVISIONS The area in question is examined (VT-2) each refueling as part of the normal Class 1 system leakage test (normal valve line-up) for evidence of leakage.

Additionally, it is proposed that the area be examined as part of the Class 2 system leakage test pressure boundary using the Class 2 test requirements associated with the adjoining Class 2 piping. This would result in an additional

pressure test each period at the Class 2 RHR system nominal operating pressure with the associated visual VT-2 examination.

It is requested per 10 CFR 50.55a(a)(3)(ii) that the tests described above for the fourth inspection interval be alternatively performed.

SPT-006 IDENTIFICATION OF COMPONENTS System: All Class 1 systems Components:Class 1 components and piping located in the extended portion (beyond the first normally closed valve) of the Class 1 boundary

11. CODE REQUIREMENT Table IWB-2500-1 Category B-P, note 2, 'The system leakage test (IWB-5220) shall be conducted prior to plant startup following reactor refueling outage," and IWB-5221 (a)states the system leakage test shall be conducted at a pressure not less than the pressure corresponding to 100% rated reactor power.

IWB-5222(b) requires the system leakage test shall be conducted at or near the end of the inspection interval to extend to all Class 1 pressure retaining components within the system boundary.

Code: Above references based on ASME Section Xl, 1998 Edition through the 2000 Addenda Ill. BASIS OF REQUEST FOR RELIEF Surry Unit 2 is in the fourth ISI interval utilizing the 1998 Edition through the 2000 Addenda of ASME Section Xl. The unit is currently in the first ISI period.