ML18026A479
| ML18026A479 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 04/28/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17158C142 | List: |
| References | |
| NUDOCS 9705090014 | |
| Download: ML18026A479 (13) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 205554001 SAFETY EVALUATION BY THE OFFIC OF NUCLEAR REACTOR REGULATION OF THE SECOND 10-YEAR INTERVAL INSERVICE INSPECTION PLAN I
RE VEST FOR REVISION TO RELIEF NO.
RRPT-8 PENNSYLVANIA POWER 8( LIGHT COMPANY SUS UEHANNA STEAM ELECT IC STATION UNITS 1
AND 2 DOCK T NOS. 50-387 AND 50-388
1.0 INTRODUCTION
The Technical Specifications (TSs) for Susquehanna Steam Electric Station, Units 1 and 2 states that the inservice inspection of the American Society of Mechanical Engineers (ASHE)
Code Class 1,
2 and 3 components shall be performed in accordance with Section XI of the ASHE Boiler and Pressure Vessel Code (ASHE Code) and applicable addenda as required by 10 CFR 50.55a(g),
except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).
Pursuant to 10 CFR 50.55a(a)(3),
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 difficulties without a compensating increase in the level of quality and safety.
Pursuant to 10 CFR 50.55a(g)(4),
ASHE Code Class 1,
2 and 3 components (including supports) shall meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASHE
- 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) twelve months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein.
The applicable edition of Section XI of the ASHE Code for the Susquehanna Steam Electric Station, Units 1 and 2, second 10-year inservice inspection (ISI) interval is the 1989 Edition.
'ursuant to 10 CFR 50.55a(g)(5), if the licensee determines that conformance with an examination requirement of Section XI of the ASHE Code is not practical for its facility, information shall be submitted to the Commission in support of that determination and a request made for relief from the ASHE Code requirement.
After evaluation of the determination, pursuant to 10 CFR 50.55a(g)(6)(i),
the Commission may grant relief and may impose alternative requirements that are determined to be authorized by law, will not ENCLOSURE
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.
In a letter dated January 29,
- 1997, pennsylvania Power and Light Company (the licensee),
submitted to the NRC its Second 10-Year Inservice Inspection Interval Program Plan Request for Relief RRPT-8, regarding Byron Jackson Reactor Recirculation Pump Nos.
1P401 A and B and 2P401 A and B for Susquehanna Steam Electric Station, Units 1 and 2.
The licensee also provided additional information in its letter dated March 19, 1997.
- 2. 0 EVALUATION The NRC staff, with technical assistance from its contractor, 'the Idaho National Engineering and Environmental Laboratory (INEEL), has evaluated the information provided by the licensee in support of its Second 10-Year Inservice Inspection Interval Program Plan Request for Relief RRPT-8 for Susquehanna Steam Electric Station, Units 1 and 2.
Based on the information submitted, the staff adopts the contractor's conclusions and recommendations presented in the Technical Letter Report (TLR) attached.
Request for Relief RRPT-08, the 1989 Edition of the Code, requires that when leakage occurs at bolted connections, all bolting is required to be removed for VT-3 visual examination for Reactor Recirculation Pump Nos.
1P401 A and B
and 2P401 A and B.
For the above reactor recirculation pump cases and cover bolts, the licensee proposed in lieu of removal of all bolting to perform a
VT-3 visual examination, to evaluate the bolted connection.
The evaluation will consider the potential for bolting degradation as well as the cause of the leakage.
If the evaluation indicates the need for a more detailed analysis, all of the bolts will be volumetrically examined and evaluated in accordance with IWB-3515.,
The Code has recognized that the requirement to remove all bolting at. leaking bolted connections to perform a VT-3 visual examination results in an undue burden on licensees.
Later editions of the Code have relaxed the requirement for bolt removal to only one bolt closest to the source of the leak.
Experience by the licensee and the pump manufacturer has found that seepage may occur at the bolted connection during startup, and will diminish at rated operating pressure and temperature.
However, the Code of record would require corrective action for this leakage.
To perform the corrective action required by Code, the design of the subject pumps would require the removal of the pump motor to allow the removal of bolting.
Regardless of the removal of one or all of the bolts, a major effort, requiring many manhours from skilled maintenance personnel, causing excessive radiation exposure, and raising personnel safety concerns, would result.
In addition, requiring the licensee to remove stud(s) could result in
thread damage or damage to the pumps.
The staff determined that requiring the licensee to remove the reactor recirculation pump case-to-cover joint bolting when leakage occurs results in a hardship without a compensating increase in quality and safety.
The licensee has proposed to evaluate the leaking bolted connection.
If it is determined that further evaluation is required, 100X volumetric examinations will be performed on the studs in place.
The staff determined that the licensee's alternative to bolt removal when leakage occurs at the reactor recirculation pump case-to-cover is based on sound 'engineering judgment.
Therefore, the staff concludes that the licensee's proposed alternative provides reasonable assurance of operational readiness.
- 3. 0 CONCLUSION The staff has reviewed the Pennsylvania Power and Light Company's Request for Relief RRPT-8 regarding Byron Jackson Reactor Recirculation Pump Nos.
1P401 A
and B and 2P401 A and B for Susquehanna Steam Electric Station, Units 1
and 2.
Based on this evaluation, the staff concluded that for the reactor recirculation pump case-to-cover bolting, requiring the licensee to remove
.bolting as part of the evaluation when leakage occurs will result in a burden without a compensating increase in quality and safety.
The staff concluded that the licensee's proposed alternative, to evaluate the subject reactor recirculation pump case-to-cover bolting when leakage
- occurs, augmented with a volumetric examination if a more detailed analysis is needed, provides reasonable assurance of continued operational readiness of Reactor Recirculation Pump Nos.
1P401 A and B and 2P401 A and B.
Therefore, the staff concludes that the licensee's pi oposed alternative for Reactor Recirculation Pump Nos.
1P401
'A and B and 2P401 A and B is authorized pursuant to 10 CFR 50.55a(a)(3)(ii).
h
Attachment:
Technical Letter Report Principal Contributor:
T. McLellan
'ate:
August 28, 1997
TECHNICAL LETTER REPORT ON THE SECOND 10-YEAR INSERVICE INSPECTION INTERVAL RE UEST FOR RELIEF RRPT-8 FOR SUS UEHANNA STEAM ELECTRIC STATION UNITS 1
AND 2 PENNSYLVANIA POWER
& LIGHT COMPANY OCKET NOS.
50-387 AND 50-388
- 1. 0 INTRODUCTION By letter dated January 29,
- 1997, Pennsylvania Power
& Light Company (PP&L) submitted Relief Request RRPT-8 for the Susquehanna Steam Electric Station, Units 1 and 2.
By letter dated March 19, 1997, the licensee provided additional clarification on the submittal.
The Idaho National Engineering and Environmental Laboratory (INEEL) staff has evaluated the subject request in the following section.
2.0 EVALUATION The Code of record for the Susquehanna Steam Electric Station, Units 1 and 2, second 10-year inservice inspection (ISI) intervals, which began June 1,
- 1994, is the 1989 Edition of the American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code,Section XI.
The information provided by the licensee in support of Request for. Relief RRPT-8 has been evaluated and the basis for disposition is documented below.
Relief Re uest RRPT-8 IWA-5250 a 2
Corrective Action Resultin from Leaka e at Bolted Connections Code Re uirement:
IWA-5220(a)(2) requires that the source of leakage detected during a system pressure test shall be located and evaluated by the Owner for corrective action.
When the leakage is at a bolted connection, the bolting shall be removed, VT-3 visually examined for corrosion, and evaluated in accordance with IWA-3100.
ATTACHMENT
Licensee's Pro osed Alternative:
In accordance with 10 CFR 50.55a(a)(3)(ii),
the licensee proposed an alternative to the ASHE Section XI requirements for removal and VT-3 visual examination of the reactor recirculation pump case and cover bolts (Byron Jackson Pumps IP401 A 8 B, and 2P401 A & B) when leakage is detected.
The licensee stated:
"The source of such leakage detected during a system pressure test shall be evaluated to determine bolting corrosion and potential failure.
This evaluation shall consider, as a minimum:
I) 'ocation and source of leakage 2) history of leakage 3) fastener materials 4) evidence of corrosion with the connection assembled 5) corrosiveness of the process fluid 6) history and studies of similar fastener material in a similar environment 7) other components in the vicinity that may become degraded due to the leakage "If the evaluation indicates the need for further analysis, then all of the studs shall be volumetrically examined, and evaluated in accordance with IWB-3515.
"If the evaluation determines that the leaking condition has not degraded the fasteners, then no further action is required.
Should significant leakage from this bolted connection persist, it would be detected in the control room via the leakage collection system (drywell sump) serving this equipment, be investigated, and be corrected, in accordance with plant Technical Specification 3.4.3.2."
Licensee's Basis for the Pro osed Alternative (as stated):
"Pursuant to CFR 50.55a(a)(3), relief is requested from the requirement of ASME Code Section XI paragraph IWA-5250(a)(2).
The requirements described above create a hardship for this plant, potentially compromise radiation safety, and are less effective than the alternative actions, which Susquehanna proposes.
"The bolting at the reactor recirculation pump case-to-cover bolted connection, being Code Examination Category B-G-l, Item Number B6. 180, is required by Table IWB-2500-1 to receive a volumetric examination, to the Acceptance Standard of paragraph IWB-3515, once each 10-year Inspection Interval.
In both Susquehanna units, all reactor recirculation pump case-to-cover studs have been volumetrically examined by ultrasonic examination once during the first 10-year Inspection Interval.
No unacceptable indications were found by this examination.
The past examination results confirm the soundness of these studs at
- present, and the periodic repetition will confirm their continued soundness in the future.
r "BWR industry experience shows no evidence of service-related corrosion damage for the recirculation pump case-to-cover studs Because. the BWR pump environment contains only condensate-quality water with no boron, there is an absence of any chemical agents that can aggressively corrode these studs.
Throughout the industry, there have been no reports of any deterioration of BWR recirculation pump case-to-cover connection
- studs, despite several plants having disassembled these bolted connections in
'the past to reduce observed leakage.
An additional factor inhibiting corrosion of the bolting at these connections is that they are located within the drywell, which is inerted with a nitrogen atmosphere throughout the operating cycle.
The elimination of atmospheric oxygen diminishes the potential for corrosion in this environment.
"The geometry of this particular bolted connection mitigates the potential for corrosion due to leakage.
The sealing
- area, which includes a flexitallic gasket, is inside the stud circle.
The cover is of compl'ex shape, having a smaller diameter (within the stud circle) on the bottom, next to the pump; and a large'r diameter (enclosing the stud circle) above.
The small diameter of the lower part of the cover, inside the stud circle, permits direct visual inspection of about 3
inches of the lower portion of the shank of each stud, after the thermal insulation is removed.
This includes the location where the stud enters the threads of the pump case.
This configuration elevates the nut well above the area where any water from minor leakage could impinge on the nut.
An annular space between the pump case and the overhang of the cover permits direct visual examination of the lower shank an portions of the threads of each stud.
"The flexitallic-type gasket of this particular bolted connection tends to seal poorly if recrushed following relaxation during any gross disturbance to the pattern of bolt tension.
"The ASNE Code requirement cited above creates a hardship for this plant and potentially compromises minimizing radiation exposure.
Removal of insulation and all bolting from this bolted connection is not required to assure absence of significant corrosion and other forms of deterioration.
- Further, removal of this bolting is not advisable.
This connection would require the installation of a new gasket of special materials and be more.difficult to seal after disassembly.
The complete disassembly of this connection, which is a High Radiation area, would result in significant personnel radiation exposure, in contradiction of the ALARA principle.
The ASNE has recognized the potential for such
conditions and has subsequently modified this requirement to allow the removal and evaluation of one bolt closest to the leak in a bolted connection reported as leaking.
"Although ASME Code editions subsequent to 1989 have improved the approach to corrective action for leakage at bolted connection, the improvement does not well address the situation of the reactor recirculation pump case-to-cover bolted connection.
In the 1990 Addenda and later editions of ASME Code Section XI, paragraph IWA-5250(a)(2) has been enhanced to state, 'f leakage occurs at a bolted connection, one of the bolts-shall be removed, VT-3 examined, and evaluated in accordance with IWA-3100.
The bolt selected shall be the one closest to the source of leakage,.
When the removed bolt has evidence of degradation, all remaining bolting in the connection shall be removed, VT-3 examined, and evaluated in accordance with IWA-3100.'his approach is not appropriate for the reactor recirculation pump case-to-cover bolted connection for the following reasons:
I)
Experience at Susquehanna and the pump manufacturer both suggest that any seepage past the gasket of this connection will disappear or greatly diminish at rated temperature and pressure'onditions, as the pump case and cover reach thermal equilibrium with each other.
Susquehanna did not observe any leakage in the pump area during the drywell entry following a 50X rod pattern scram at rated temperature and pressure.
Further evidence is provided by monitoring the leakage.collection system (drywell sump) that serves this equipment over the operating cycle.
No sizable leakage was observed.
2)
The thermal insulation covering the reactor recirculation pump makes it impractical to determine the precise location of any seepage from this connection.
Leakage of interest might occur only at ASME Class I System Leakage Test pressure.
'During the ASME Class I
System Leakage Test, plant Technical Specification 3. 10.6 limits coolant temperature to less than or "equal to 212'F.
While at test pressure and increasing coolant temperature, insulation removal will be carefully controlled because the removal could introduce unusual temperature distribution patterns.
This could have possible adverse effects on dimensional clearances of rotating components of the pump and on achievement of thermal equilibrium between the pump case and
- cover, and create or exacerbate leakage pa'st the gasket.
3)
The reactor recirculation pump case and cover are designed to support removal of studs from the pump case only after removal of the cover (and of the motor above it).
Lack of access to the full length of each stud when the cover is in place impedes the stud removal operation and precludes removal of any stud that galls.
4)
The stainless steel material of the pump case is vulnerable to galling during stud removal.
If galling occurs, the old stud must be drilled and machined to extract it.
A repair of the threaded'ole in the pump case, prior to installation of a new stud would also be necessary.
"Thus, removal of any stud for visual examination would result in the hardship or drastic increase in wor k scope and personnel radiation exposure without any compensating increase in safety."
Evaluation:
In accordance with the 1989 Edition of the Code, when leakage occurs at bolted connections, all bolting is required to be removed for VT-3 visual examination.
In lieu of removal of all bolting to perform a VT-3 visual examination, the licensee has proposed to evaluate the bolted connection.
The evaluation will consider the potential for bolting degradation as well as the cause of the leakage.
If the evaluation indicates the need for a more detailed analysis, all of the bolts will be volumetrically examined and evaluated in accordance with IWB-3515.
The Code has recognized that the requirement to remove all bolting at leaking bolted connections to perform a VT-3 visual examination results in an undue burden on licensees.
Later editions of the Code have relaxed the requirement for bolt removal to only one bolt closest to the source of the leak.
Experience by the licensee and the pump manufacturer has found that seepage may occur at the bolted connection during startup, and will diminish at rated operating pressure and temperature.
However, the Code of record would require corrective action for this leakage.
'o perform the corrective action required by Code, the design of the subject pumps would require the removal of the pump motor to allow the removal of bolting.
Regardless of the removal of one or all of the bolts, a major effort, requiring many manhours from skilled maintenance personnel, causing excessive radiation exposure, and raising personnel safety concerns, would result.
In addition, requiring the licensee to remove stud(s) could result in thread damage.
Because of the burden on the licensee, the INEEL staff
- 6'-
believes. that requiring the licensee to remove the reactor recirculation pump case-to-cover joint bolting when leakage occurs results in a hardship without a compensating increase in quality and safety.
The licensee has proposed to evaluate the leaking bolted connection.
If it is determined that further evaluation is required, 100X volumetric examinations will be performed on the studs in place.
The INEEL staff believes that the licensee's alternative to'olt removal when leakage occurs at the reactor recirculation pump case-to-cover is based on sound engineering judgment.
Therefore, it is concluded that the licensee's proposed alternative provides reasonable assurance of operational readiness.
3.0 CONCLUSION
The INEEL staff has 'reviewed the Pennsylvania Power and Light Company's Request for Relief RRPT-8 for Susquehanna Steam Electric Station, Units 1 and 2.
Based on this evaluation, it concluded that for the reactor recirculation pump case-to-cover bolting (Byron Jackson Pumps 1P401 A
8 B,
and 2P401 A & B),
requiring the licensee to remove bolting as part of the evaluation when leakage occurs will result in a burden without a compensating increase in quality and safety.
The licensee's proposed alternative, to evaluate the subject reactor recirculation pump case-to-cover bolting when leakage
- occurs, augmen'ted with a volumetric examination if a more detailed analysis is needed, will provide reasonable assurance of continued operational readiness.
Therefore, it is recommended that the proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(ii).