ML20114E767

Provides Suppl Info Supporting Proposed Relief Request Applicable to Unit 2 First 10-yr Interval ISI Program Submitted to NRC on 960408

ML20114E767
Person / Time
Site:	Beaver Valley
Issue date:	06/18/1996
From:	Jain S DUQUESNE LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9606270106
Download: ML20114E767 (10)

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Duquesne Lic)t Company ;gg: v a- < sia'ioa e

Shippingport. PA 15077 0004

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. S. Nuclear Regulatory Commission r

lune 18,1996 Attention: Document Control Desk b- [E Washington, DC 20555-0001 c

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Subject:

Beaver Valley Power Station, Unit No. 2 Docket No,50-412, Lipense No. NPF-73 ISI (Inservice Inspection) Program: Relief Request-Supplement 7

i The purpose of this submittal is to provide supplemental information supporting a proposed relief request applicable to the Unit No. 2 First Ten-Year Interval ISI Program which was submitted to the NRC on April 8,1996. The current 10-year interval ends on November 17,1997. The ISI Progrt.m is based on the 1983 Edition through the Summer 1983 Addenda of ASME Boiler and Pressure Vessel Code,Section XI.

On May 9,1996, a conference call between the NRC reviewers and Beaver Valley staff members resulted in an agreement to submit supplemental information in support of the proposed relief request.

Information regarding the potential use of Code Case N-408-2, leak before break, and service history sf the regenerative heat exchanger and similar components was requested. As a result, the proposed relief request has been revised as shown in the attached Relief Request BV2-C.130, Rev.1. Please note that the photographs referenced in the relief request are not attached since they were submitted on April 8,1996.

It is proposed to substitute visual exammations in accordance with the attached relief request in order to reduce occupational radiation exposure. A proposed alternative to the requirements, with supporting basis is included in the attached request for relief.

10 CFR 50.55a(a)(3)(ii) provides for NRC approval of proposed alternatives to the ASME Code requirements when it can be demonstrated that compliance would result in hardship or ynusual difficulty without a compensating increase in the level of quality and safety. The attached relief request satisfies this acceptance criteria.

.sp'f l DEllVERING QUALITY ENERGY 9606270106 960618 PDR ADOCK 05000412 G

PDR

i Beaver Valley Power Station, Unit No. 2 j

ISI (Inservice Inspection) Program: ReliefRequest-Supplement Page 2 4

1 It is requested that this review be completed before July 8,1996, in order to allow sufficient time to plan and prepare for any follow-up examinations that may be required during the sixth refueling outage scheduled to begin August 30,1996.

If you have any questions regarding this issue, please contact Mr. Roy K. Brosi at (412) 393-5210.

i Sincerely, Sushil C. Jain 1

Attachment 1

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Mr. L. W. Rossbach, Sr. Resident Inspector Mr. T. T. Martin, NRC Region I Administrator Mr. D. S. Bnnkman, Sr. Project Manager i

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I DUQUESNE LIGHT COMPANY Beaver Valley Power Station Unit No. 2 RELIEF REOUEST NO. BV2-C1.30. Rev. I 1

COMPONENT Regenerative Heat Exchanger (2CHS*E23) - Tubesheet to Shell Welds 2, 3, 6, 7,10, and 11 DRAWING NO.

l ISI-E-2C SECTION XI REOUIREMENT (83S83)

Item No. C1.30 (IWC-2500-1, Category C-A) requires volumetric examination.

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BASIS OF RELTEF l

The general radiation levels in the area of this heat exchanger are between 500 mR/hr and 3000 mR/hr and up to 5000 mR/hr on contact (based on surveys taken in the second and fifth refueling outages at BV-2). The manrem estimate to prepare, inspect, and reinsulate these six welds is 9800 mrem. This estimate is based on working space dose rates and estimated work duration as noted on the attached manrem estimate.

I The design and function of this component give rise to areas where " hot spots" can readily occur due to corrosion and wear products buildup. Flushing provides little benefit toward reducing radiation fields in this area. The location of the welds, between two branch connections with a welded support between the two welds, makes shielding of these areas impractical. Reference attached drawing and preservice photographs.

1 A visual examination for leakage, in this situation, would provide an adequate method for monitoring the integrity of the component. The required ultrasonic (UT) examinations are limited by component geometric restrictions. Approximately 1/3 of the required volume of each weld could not be examined in the preservice inspection. The adjacent branch connections and the welded pad, plates and lugs of the support located between the welds limit the UT exam,(Note: Relief Request BV2-CI.10-2 identified these limitations and was approved in the SER dated August 22,1991.) Preservice UT examinations performed on welds 2,3,6,7, and 10 had no reportable indications. ExanJaation of weld 11 found a manufacturing defect that was subsequently repaired and reexamined satisfactorily.

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Page 2 of 4 i

Relief Reauest No. BV2-C1.30. Rev.1 i

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Code Case N-408 has been implemented at Unit 2 for Class 2 components. The regenerative

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heat exchanger was initially evaluated for exclusion from the ISI Plan based on N-408 and l

subsequently evaluated against N-408-2. In both cases, Duquesne Light Company has adopted a conservative position and included the regenert,tive heat exchanger and its associated piping i

l in the ISI Plan for examination. It is recognized that N-408-2 would permit the regenerative heat exchanger to be excluded from the ISI Plan. The regenerative heat exchanger is not part of the Emergency Core Cooling System (ECCS) flow path and could be excluded from examination; however, the regenerative heat exchanger may remain part of the Reactor i

Coolant System (RCS) boundary even though there is no flow through the heat exchanger.

l This conservative position was adopted because the two valves, (2CHS*LCV-460A and B) that isolate the shell side of this heat exchanger from the RCS, do not automatically close upon a safety injection actuation, but do close upon a Pressurizer low level signal. If a leak occurred at one of the these welds, in excess of ECCS make-up capability, resulting in a low level in the i

Pressurizer, the regenerative heat exchanger would then be automatically isolated from the RCS by valves 2CHS*LCV-460A and B.

a The regenerative heat exchanger shell material is SA-351 CF 8, which is cast stainless steel material made under the same specification (different grade) as the reactor coolant loop piping.

l Both grades have identical chemical requirements, with similar tensile and yield strengths. BV-l 2 has been licensed to " leak before break" for large bore stainless steel piping. The j

regenerative heat exchanger was not included in this evaluation. But, many of the factors j

included in the large bore piping " leak before break" evaluation are applicable to the l

regenerative heat exchanger. As noted, the material for the regenerative heat exchanger shell i

and the loop piping is covered by the same material specification (SA 351) which is considered a high strength, ductile material. Also, the fluid contained on both the tube and shell sides of the regenerative heat exchanger is the same as the RCS loop piping. Strict chemistry standards are maintained to ensure a non-corrosive environment. Oxygen, chloride, fluoride and other i

contaminant concentrations are maintained below the thresholds known to be conducive to j

stress corrosion cracking. Further, there is a low potential for water hammer at the 4

j regenerative heat exchanger, since the design and operation preclude voids in the system. The j

regenerative heat exchanger is designed for normal, upset, emergency and faulted condition i

transients.

l To our knowledge, no plant has experienced a through wall leak at their regenerative heat exchanger. An informal survey ofeight utilities, representing fourteen Westinghouse PWR's, j

found that none of these plants have had leaks at the regenerative heat exchanger. Likewise, BV-1 and BV-2 regenerative heat exchangers, having a combined service history of i

approximately 30 years, have not experienced any leaks and have proven to be very reliable.

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The feed and return piping to the regenerative heat exchanger are included in the ISI Plan. The examinations performed to date have found no unacceptable indications.

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Page 3 of4 Relief Request No. BV2-C1.30. Rev.1 i

Several methods are available to detect leakage from these welds if sufficient weld degradation

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occurred to cause a through-wallleak. Listed below are some examples:

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a The control room operators perfonn Operation Surveillance Test (OST) 2.6.2A

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" Reactor Coolant System Water Inventory Balance" every three days when the plant is operating at steady conditions. Leakage through the subject welds would j

be discovered by the conduct of this OST.

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b. Containment cirborne radiation monitors continuously sample the containment i

atmosphere and alarm in the control room. This method is sensitive enough to detect a 1 gpm leak in less than I hour.

2 Two leakage monitoring systems are available in the containment sump. One c.

system uses the flow indication of the containment sump pumps to determine leak rates. A programmable controller monitors the flow and pump operating i

times so that a 1 gpm leak could be detected by this method in I hour. A second system uses changes in the water level of the sump to determine a leak rate. The

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containment sump levelis continuously monitored by instruments that alarm in i

the control room. This system is capable of detecting a 5 gpm leak in I hour.

j The Regenerative Heat Exchanger is readily isolable should a leak occur. Double valve

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isolation from the reactor coolant system is provided by valves 2CHS*LCV-460A and B.

CONCLUSION Obtaining the limited UT data on the identified welds would result in hardship and unusual difficulty without a compensating increase in the level ofquality and safety. A visual examination would provide an alternate inservice examination that would provide an acceptable level of quality and safety. This conclusion is based on the following:

i the reliable service history of the regenerative heat exchangers at both BVPS and throughout the nuclear industry I

the code required UT examinations would result in less than the required exam coverage if they were performed l

the fracture toughness of SA 351 material i

the detection methods and isolating capabilities should a through-wall leak occur the radiation dose associated with the preparation for and examination of these welds

i Page 4 of 4 1

Relief Recuest No. BV2-C1.30. Rev.1 i

1 ALTERNATIVE EXAMINATION l

A visual examination of these welds for leakage is performed in conjunction with the boric acid walkdown, every shutdown while the RCS and associated piping remain at operating pressure and temperature. Also, the regenerative heat exchanger is included in the Mode 3 walkdown i

of the RCS boundary, performed during each startup following refueling outages. Both these activities are performed by qualified VT-2 examiners. These examinations are augmented by the leakage detection methods noted above.

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2CHS-E23 EXAMINATIO..s t

WELDS 2, 3,6,7,10 and 11 MANREM ESTIMATE t

Relief Reauest BV1-C1.30. Rev.1 1

2CHS-E23 Regenerative Heat Exchanger General area radiation level: 500 mR/hr to 3000 mR/hr Contact radiation level: 1000 to 5000 mR/hr TIME IN EXPOSURE ESTIMATED l

WORK

1. OF WORKERS RAD FIELD RATE EXPOSURE TASK AND JOB CLASS (hrs)

(mR/hr)

(mrem)

CONSTRUCT SCAFFOLDING (2)

CARPENTERS x

1 x

600 1200

=

REMOVE INSULATION (2)

INSULATORS X

1 x

600 1200 i

=

WELD L

PREPARATION (2)

FITTERS X

1 x

800 1600

=

i WELD l

INSPECTION (2)

EXAMINERS x

2.5 x

800 4000

=

t REINSTALL INSTALLATION (2)

INSULATORS X

1 x

600 1200

=

REMOVE SCAFFOLDING (2)

CARPENTERS x

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600

=

600 TOTAL FOR SIX WELDS:

9800 mrem

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