ML20212F767
| ML20212F767 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 09/24/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20212F752 | List: |
| References | |
| NUDOCS 9909280368 | |
| Download: ML20212F767 (6) | |
Text
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% UNITED STATES j
- Ij . NUCLEAR REGULATORY COMMISSION WA5HINGTON, D.C. '8aaaa aani 4.....
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION EVALUATION OF RELIEF REQUESTS COMANCHE PEAK STEAM ELECTRIC STATION. UNIT 1
- RELIEF REQUESTS A-3. B-15. B-16 AND B-17 COMANCHE PEAK STEAM ELECTRIC STATION. UNIT 2 RELIEF REQUEST C-4 TXU ELECTRIC COMPANY DOCKET NOS. 50-445 AND 50-446
1.0 INTRODUCTION
The inservice inspection of the American Society of Mechanical Engineers (ASME) Code Class 1,2, and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code (the Code) and applicable addenda as required by Section 50.55a(g) of Title 10 of the Code of FederalRegulations (10 CFR), except where
- specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).
Section 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (i) the proposed alternatives would provide an acceptable level '
of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Pursuant to 10 CFR 50.55e.(g)(4), ASME Code Class 1,2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code.,Section XI, " Rules for inservice Inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) on the date 12 months prior to the start of the 120-month ,
interval, subject to the limitations and modifications listed therein. The applicable edition of the ASME Code,Section XI, for Comanche Peak Steam Electric Station (CPSES), Units 1and 2, during the first 10-year inservice inspection (ISI) interval, is the 1986 Edition. The components (including supports) may meet the requirements set forth in subsequent editions and addenda of the ASME Code incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein and subject to Commission approval.
ENCLOSURE 9909290368 990924 PDR ADOCK 05000445 p PDR
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Pursuant to 10 CFR 50.55a(g)(5), if the licensee determines that conformance with an
' examination requirement of Section XI of the ASME Code is not practical for its facility, :
- j. information shall be submitted to the Commission in support of that determination and a request made for relief from the ASME Code requirements. After evaluation of the determination, pursuant to 10 CFR 50.55a(g)(6)(i), the Commission may grant relief and may impose i
attemative requirements that are determined to be authorized by law, will not endanger life, property, or the common defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed.
By letter dated February 17,1999, as supplemented by letter dated August 16,1999, TXU Electric (the licensee), requested relief from the examination requirements of the 1986 Edition l ASME Code,Section XI, for the containment spray heat exchanger and the reactor vessel during the first 10-year ISI interval for the CPSES, Unit 1. The requests for relief were identified in the February 17,1999, submittal as A-3, B-15, B-16, and B-17. In addition, by letter dated July 16,'1999, the licensee requested relief from the examination requirements of the 1986 Edition of ASME Code,Section XI, for the reactor vessel during the first 10-year ISI interval for CPSES, Unit 2. The request for relief was identified in the July 16,1999, submittal as C-4.
2.0 DISCUSSION AND EVALUATION 2.1 Relief Reauest A-3. Rev.1 Systems / Components for Which Relief is Reauested:
Containment Spray Heat Exchanger CP1-CTAHCS-02, ASME Code Class 2 Examination Reauirements from Which Relief is Requested The ASME Code Section XI,1986 Edition, paragraph IWA 5250(a)(2) requires that "(a) The source of leakages detected during the conduct of a system pressure test shall be located and evaluated by the Owner for corrective measures as follows...(2) if leakage occurs at a bolted connection, the bolting shall be removed, VT-3 visually examined for corrosion, and evaluated in accordance with IWA-3100." <
2.1.1 Licensee's Basis For Relief The heat exchanger is pressurized once per calendar quarter during ASME Section XI inservice test of the containment spray pump; This test duration is less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and the subsequent leakage is less than 10 drops per minute. As the duration and leakage is small, no appreciable materialloss is expected to occur.
The effort and radiation exposure required to replace the gasket is not warranted based on the l unlikely occurrence of botting degradation. In order to replace the gasket, all 441%-inch bolts must be removed, the missile shield removed, engineered rigging installed, and the heat exchanger decoupled. This effort would require a large amount of manpower as well as radiation exposure. Additionally, this system would be out of service for a duration in excess of the current Technical Specification outage time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
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The leakage occurs at two bolts of the connection. Assuming the worst case with the loss of i L
' . two bolts, the, subject flange joint has been evaluated per the ASME Code requirements with i only 42 with out of 44 bolts present and determined to remain operable and in compliance with the ASME Code.
2.1.2 Licensee's Alternate Examinations 4
Following each ASME Section XI containment spray pump test, the insulation will be removed !
from the leaking flange area and the area will be cleaned. Also, following the pump test, an :
ultrasonic examination of the bolting in the leaking area will be performed to identify gross . I material loss. - If gross material loss is detected, then suspect bolting will be replaced. Typically,- l a 0-degree angle ultrasonic examination of a bolt will identify material loss starting at
! 15-20 percent through depth. - The bolts in the area of leakage are not accessible for removal and subsequent visual examination due to existing supporting structure obstacles.
l 2.1.3 Evaluation (Relief Reauest A-3. Rev.1) 4 The licensee states that during the inservice test of the containment spray pump once every calendar. quarter, the containment spray heat exchanger that is pressurized, develops leakage of 10 drops per minute at the bolted connection of the flanged joint on the heat exchanger.
This leakage is observed in the vicinity of two bolts of the connection having a total of 44 bolts.
The licensee is unable to comply with the requirement of para 0raph IWA 5250(a)(2) of the ASME Code,'Section XI,1986 Edition, which requires that the botting be removed for a VT-3 visual examination for corrosion followed by an evaluation in accordance with IWA-3100. The bolts in the area of leakage are not accessible for removal and subsequent visual examination
[ ' due to obstructions from supporting structures. In order to stop the leakage, the gasket needs to be replaced, which is associated with significant radiation exposure. Alternatively, the licensee proposes to perform an ultrasonic examination of both bolts, in place, following each inservice test of the containment spray pump. The staff believes that the ultrasonic examination of the bolts will detect any materialloss greater than 15 percent through-wall with reasonable confidence and, therefore, will assure structural integrity of the bolts. Nevertheless, l
. the duration of the pump test being short and the leakage being very small, no appreciable material loss is expected to occur. However, in an unlikely event of not timely detecting gross .
. material loss by ultrasonics, it is estimated that the remaining 42 oct of 44 bolts are also capable of maintaining the integrity of the flanged joint. The licensee further states that should there be any indication of a gross material loss, the bolts in question would be replaced. The
' NRC staff has, therefore, determined that the licensee's proposed alternative provides a reasonable assurance of operational readiness.
2.2 Relief Reauests B-15. B-16. and B-17
- Items For Which Relief is Reauested
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- a. TBX-1-1100-4 Reactor Vessel Shell to Bottom Head Circumferential Weld
- b. TBX-1 1100 5 Reactor Vessel Lower Head Circumferential Weld
- c. TBX-1 1100A-19,22,23, and 26 Reactor Vessel Outlet Nozzle to Shell Welds
- d. .TBX-1 1100A-20,21,24, and 25 Reactor Vessel inlet Nozzle to Shell Welds o
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p L 4 Examination Reauirement From Which Relief is Reauested:
l Volumetric examination of essentially 100 percent of weld volume 2.2.1 Licensee's Basis For Relief For welds TBX-1-1100-4 (Reactor Vessel Shell to Bottom Head Circumferential Weld) and l TBX-1-1100-5 (Reactor Vessel Lower Head Circumferential Weld), physical interference from l core support lugs and bottom mounted instrument tubes allows examination of approximately 61 to 75 percent of weld volume, respectively. For welds TBX-1-1100A-19,22,23, and 26 l
(Reactor Vessel Outlet Nozzle to Shell Welds) approximately 42 percent of weld volume can be examined and for welds TBX-1-1100A-20,21,24, and 25 (Reactor Vessel inlet Nozzle to Shell 1 Welds) approximately 79 percent of the weld volume can be examined. The limitation to examination of the above nozzle to shell welds is due to the geometric configuration of the nozzle welds.
l 2.2.2 Licensee's Afternate Examination (as stated)
L ' " Cameras on the examination tool will be used to visually access some of the areas that are not accessible to the ultrasonic transducer scanning sled."
2.2.3 Evaluation (Relief Reauests B-15. B-16. and B-17)
The staff has determined that the reactor vessel shell to bottom head and the lower head circumferential welds designated as TBX-1-1100-4 and TBX-1-1100 5, respectively, are
! inaccessible to ultrasonic scanning due to interference from the core support lugs and the bottom mounted instrument tubes. The licensee's best-effort examination would result in volumetric coverages of approximately 61 to 75 percent, respectively for the subject welds.
i Further, the welds joining the four outlet and the four inlet nozzles to the vessel shell designated as TBX-1-1100A-19,22,23, and 26 and TBX-1-1100A-20,21,24, and 25, respectively, can be
, i volumetrically examined up to approximately 42 to 79 percent, respectively, due to weld -
configuration. Therefore, it is impractical to obtain a volumetric coverage of greater than 90 percent required by the Code for each of the subject welds unless the vessel is redesigned to improve access to the welds, which certainly would impose a significant burden on the licensee. However, the staff believes that the examination coverage expected to be obtained for each weld with the existing physical constraint or the geometric configuration should l discover the existence of any service-related degradation with reasonable confidence.
The welds are located outside of the vessel beltline. Therefore, the welds are not subject to embrittlement due to neutron irradiation. The NRC staff notes that the licensee has proposed
- to supplement the limited volumetric examination of the welds with a remote visual examination i of the subject welds. Therefore, the proposed alternative provides reasonable assurance of structuralintegrity of the subject welds.
2.3 Relief Reauest C-4 I
.ltem for Which Relief is Reauested:
Containment Spray Heat Exchanger Shell to Flange Weld, TCX-2-1180-1-2 I
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Examination Reauirement from Which Relief is Reauested:
! Volumetric examination of essentially 100 percent of weld volume as stated in the ASME Code,Section XI,1986 Edition, Table IWC-2500-1, Examination Category C-A, item No. C1.10.
2.3.1 Licensee's Basis for Relief L
For weld TCX-2-1180-1-2, interferences from the heat exchanger welded support pads preclude the complete ultrasonic examination of the volume required by Figure IWC-2500-1.
Approximately 86 percent of the weld volume was examined with no recordable indication
' found. Completion of the Code-required volumetric examination would require extensive modifications of the heat exchanger, which would impose a significant burden on the licensee.
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2.3.2 Licensee's Proposed Alternate Examination None 2.3.3 Evaluation (Relief Reauest C-4)
The staff finds that due to interferences from the welded support pads of the containment spray heat exchanger, essentially 100 percent of the shell to flange weld could not be examined. The licensee's best-effort examination resulted in 86 percent volumetric coverage. In order to obtain the Code-required examination coverage, the heat exchanger must be extensively modified, which would cause a significant burden to be placed on the licensee. Nevertheless, if there were service-related degradation in the subject weld. 86 percent volumetric coverage of the weld should have detected it with reasonable confidence. Therefore, the licensee's best-effort examination provides reasonable assurance of structuralintegrity of the weld.
3.0 CONCLUSION
3.1 Relief Reauest A-3. Rev.1
, The staff has determined that the licensee's proposed alternative of monitoring material loss by ultrasonics for two bolts in the vicinity of leakage at the containment spray heat exchanger flanged joint, in lieu of the Code-required removal of bolting and VT-3 visual examination
. provides reasonable assurance of structuralintegrity of the bolts and operational readiness.
l The staff concludes that compliance with the Code requirements would result in hardship
, .without a compensating increase in the level of quality and safety and, therefore, relief is authorized pursuant to 10 CFR 50.55a(a)(3)(ii) for CPSES, Unit 1, during the first 10-year ISI interval.
3.2 Relief Recuests B-15. B-16. and B-17 l ' The staff concludes that the reactor vessel shell to bottom head circumferential weld, the lower I head circumferential weld, the vessel shell welds to the outlet nozzles, and vessel shell welu to the inlet nozzles are inaccessible for essentially 100 percent volumetric coverage either due to physical interference from other components or due to the configuration of the welds. The vessel would have to be redesigned in order for the licensee to comply with the Code-required inspection, which would impose a significant burden on the licensee. The Code requirements l
are, therefore, impractical. The NRC staff concludes that the licensee's best-effort examination supplemented by a remote visual examination of the subject welds would assure structural integrity and, therefore, pursuant to 10 CFR 50.55a(g)(6)(i) relief is granted for the subject welds identified in Relief Requests B-15, B-16, and B-17 for CPSES, Unit 1, during the first 10-year ISI interval. The relief granted is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
3.3 Relief Reauest C-f The staff concludes that the Code-required examination is impractical for weld TCX-2-1180-1-2 and the licensee has provided reasonable assurance of structuralintegrity of the weld with the best-effort examination. Based on the burden on the licensee if the Code requirements were imposed, relief is granted pursuant to 10 CFR 50.55a(g)(6)(i) for the subject weld identified in Relief Request C-4 for CPSES, Unit 2, during the first 10-year ISI interval. The relief granted is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest giving due consideration to the burden upon the 1icensee that could result if the requirements were imposed on the facilityJ Principal Contributor: P. Patnaik Date: September 24, 1999 o