ML20029D002
| ML20029D002 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/25/1994 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20029C999 | List: |
| References | |
| NUDOCS 9405030217 | |
| Download: ML20029D002 (10) | |
Text
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' UNITED STATES 3
"i NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D.C. 20555-0001
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ENCLOSURE 1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTQ_R REGULATION OF THE THIRD TEN-YEAR INTERVAL INSERVICE INSPECTION PROGRAM RE00ESTS FOR RELIEF NORTHEAST NUCLEAR ENERGY COMPANY MILLSTONE NUCLEAR POWER STATION. UNIT 1 QOfKET NO. 50-245
1.0 INTRODUCTION
The Technical Specifications for Millstone Nuclear Power Station, Unit 1, state that the inservice inspection and testing 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 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)(1).
10 CFR 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 50.55a(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 inspection interval,. subject to the limitations and modifications listed therein. The applicable edition et Section XI of the ASME Code for the Millstone Nuclear Power-Station, Unit 1, third 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.
9405030217 940425 PDR ADOCK 05000245-P PDR
4 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, information shall be submitted to the Commission in support of that determination and a request made for relief from the ASME 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 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 October 29, 1993, the licensee, Northeast Nuclear Energy Company (NNECO), proposed an alternative examination to and requested relief from the requirements of the ASME Boiler and Pressure Code Section XI that it determined to be impractical to perform at Millstone Nuclear Power Station, Unit 1.
2,0 EVALUATION AND CONCLUSIONS The staff, with technical assistance from its contractor, the Idaho National Engineering Laboratory (INEL), has evaluated the information provided by NNECO in support of its Requests for Relief B-G-1-1, IWA-2, and IWC-1.
Based on the information submitted, the staff adopts the contractor's conclusions and recommendations presented in the Technical Evaluation Summary attached.
The alternatives contained in Request for Relief B-G-1-1 and IWA-2 are authorized pursuant to 10 CFR 50.55a(a)(3)(1) provided the proposed alternative actions described by NNEC0 are followed.
Request for Relief IWC-1 has been denied since an acceptable level of quality and safety has not been demonstrated.
The NRC staff has determined that certain Section XI required inservice inspections cannot be performed to the full extent required by Section XI.
The staff has determined that the requested relief (B-G-1-1 and IWA-2),
provides an acceptable level of quality and safety and may be granted pursuant to 10 CFR 50a(a)(3)(i).
Principal Contributor:
K. Battige Date:
p ENCLOSURE 2 TECHNICAL EVALUATION
SUMMARY
0F THE THIRD TEN-YEAR INTERVAL INSERVICE INSPECTION RE0 VESTS FOR RELIEF B-G-1-1 (10/93). IWA-2. AND IWC-1 (10/93)
FOR NORTHEAST NUCLEAR ENERGY COMPANY MILLSTONE NUCLEAR POWER STATION. UNIT 1 DOCKET NUMBER:
50-245
1.0 INTRODUCTION
'I By letter dated October 29, 1993, the licensee, Northeast Nuclear Energy i
Company, requested relief from the requirements of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI.
Two of these requests for relief [B-G-1-1 (12/90) and IWC-1 (12/90)] had been previously submitted and subsequently denied in the NRC Safety Evaluation dated January 5, 1993.* The Idaho National Engineering Laboratory (INEL) staff has evaluated the new information provided by the licensee in support of these requests in the following section.
2.0 EVALUATION The information provided by the licensee in support of the requests for relief has been evaluated and is documented below. Millstone Nuclear Power Station, Unit l's third 10-year inservice inspection (ISI) interval began in May 1991.
Based on this date, the applicable edition of Section XI of the ASME Code for the third 10-year ISI interval is the 1986 Edition, a
A.
Reauest for Relief No. B-G-1-1 (10/93). Examination Cateaory B-G-1.
Item B6.10. Reactor Pressure Vessel (RPV) Closure Head Nuts Code Reauirement:
Section 'XI, Table IWB-2500-1, Examination Category B-G-1, " Pressure Retaining Bolting, Greater than 2 in.
Diameter," Item B6.10, " Closure Head Nuts," requires a 100% surface examination once each 10-year interval.
j Licensee's Code Relief Reauest:
The licensee requested' relief from performing the Code-required surface examination of the RPV closure head nuts.
Licensee's Basis for Reauestina Relief (As stated):
" Extensive surface preparation and cleaning of the reactor vessel closure head nuts is required prior to performing an acceptable surface examination. The cleaning material refuse results in ' mixed waste,'
i.e., contaminated and hazardous material that cannot be disposed of.
"J. F. Stolz letter to J. F. Opeka, " Safety Evaluation of the Third Ten-Year Interval Inservice Inspection Program Plan for Millstone Nuclear Power Station, Unit 1," dated January 5, 1993.
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This extensive cleaning has resulted in additional costs and inefficient use of available manpower resources.
The service history of closure head fasteners at Millstone Unit No. I has shown that there have been no problems with closure head nuts.
Potential defects would develop in the threads of the bolt or stud prior to developing in the nut's threads because the stresses in the male threads are higher than that of the nut threads.
When the reactor vessel head fasteners are tightened for closure or loosened for opening, the studs are tensioned and the nuts are run on the threads with no load by utilizing a tensioning device.
The service history reinforces this position.
"NNECO believes that continuing to perform surface examinations on the closure head nuts will not provide any potential increases in plant safety margins, and the additional costs of these examinations are no longer warranted based upon changes now published in the 1989 Addenda of Section XI of the ASME Code.
The 1989 Addenda to the 1989 Edition of the ASME Code (although not presently approved by reference in 10CFR50.55(a) changed the examination requirement from a surface examination to a visual (VT-1) examination.
" Additionally, by letter dated February 19, 1992,1 the Connecticut Yankee Atomic Power Company was granted relief to perform a visual (VT-1) examination in lieu of a surface examination at the Haddam Neck Plant."
Licensee's Proposed Alternative Examination (As stated):
"A visual examination (VT-1) will be performed on all 56 reactor vessel closure head nuts once during the third ten-year inspection interval."
Evaluation: The 1986 Edition of the Code requires that reactor vessel closure head nuts receive a surface examination once each inspection interval.
The licensee has proposed to perform a VT-1 visual examination in lieu of the Code-required surface examination.
The licensee points out that this change in the examination requirement for RPV closure head nuts has been incorporated in the 1989 Addenda of the Code.
If a service-induced defect were to develop, it would occur in the threads of the bolt or stud before developing in the threads of the nut due to the higher stresses in the male threads. Additionally, when the reactor vessel head fasteners are tightened for closure or loosened for opening, the studs are tensioned and the nuts are run on the threads l
with no load by utilizing a tensioning device.
The service history of j
closure head fasteners shows no problems with closure nuts.
It is
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concluded that the licensee's proposed alternative VT-1 visual l
examination will provide an acceptable level of quality and safety.
Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), it is recommended that
' John F. Stolz letter to John F. Opeka, " Requests for Relief From Section XI Code Requirements for the Haddam Neck Plant," dated February 19, 1992.
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t the proposed VT-1 visual examination of the closure head nuts be authorized.
B.
Reauest for Relief No. IWA-2. Article IWA-5250. Corrective Measures for J
Leakina Bolted Connections Code Reouirement:
Section XI, Article IWA-5250(a)(2) states that 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.
Licensee's Code Relief Reaues1: The licensee' requested relief from removing all of the bolts, as required by IWA-5250(a)(2), if leakage occurs at a bolted connection.
Licensee's Basis for Reauestina Relief (As stated):
"NNECO believes that the overall benefit of removing all bolting for examination if leakage occurs at a bolted connection is not commensurate with the expense of manpower resources and, in certain cases, additional radiation exposure.
Since Millstone Unit No. 1 is a Boiling Water Reactor, corrosion resulting from boric acid leakage is not a concern.
The concept of the 1990 Addenda to examine a sample of bolts, with provisions for sample expansion in the event degradation is observed, is believed to be a more appropriate alternative.
" Additionally, by letter dated March 1,1993, the NRC staff granted the Limerick Generating Station approval to utilize the requirements of-Article IWA-5250(a)(2) of the 1990 Addenda of Section XI of the ASME Boiler and Pressure Vessel Code."
Licensee's Proposed Alternative Examination (As stated):
"If 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 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.
Reference:
1990 Addenda of Section XI of the ASME Boiler and Pressure Vessel Code, paragraph IWA-5250(a)(2)."
Evaluation:
The 1986 Edition of the Code requires that the source of leakages detected during the conduct of a system pressure test be located and evaluated for corrective measures.
If leakage occurs at a-bolted connection, the bolting is to be removed and VT-3 visually examined. The licensee proposed to use the provisions of the 1990 Addenda, which states that one bolt, closest to the leakage, shall be removed, VT-3 examined, and evaluated.
When the removed bolt has evidence of degradation, all remaining bolting in the bolted connection shall be removed, VT-3 examined, and evaluated.
Because a Boiling Water Reactor does not contain borated water, the 3
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corrosion resulting from boric acid is not a concern.
The removal of the bolt closest to any leakage should provide a sufficient indication of corrosion if it has occurred.
For this reason, following paragraph IWA-5250(a)(2) from the 1990 Addenda should disclose a significant safety problem, and will provide an acceptable level of quality and sa fety.
Therefore. pursuant to 10 CFR 50.55a(a)(3)(i), it is recommended that this alternative examination be authorized.
C.
Reauest for Relief No. IWC-1 (10/93). Paraaraoh IWC-1220. Exemption Criteria for Feedwater Coolant In.iection (FWCI) System in Examination Cateaories C-A. C-B. C-C. and C-F-2.
Code Reauirement:
Section XI requires the examination of Category C-A,
" Pressure Retaining Welds in Pressure Vessels", Category C-B, " Pressure Retaining Nozzle Welds in Vessels", Category C-C, " Integral Attachments for Vessels, Piping Pumps and Valves", and Category C-F-2, " Pressure Retaining Welds in Carbon or Low Alloy Steel Piping," components in accordance with Table IWC-2500-1.
Section XI, Article IWC-1221(a), " Components within RHR, ECC, and CHR Systems (or portions of Systems)" exempts vessels, piping, valves, and other components within the FWCI system that are NPS 4 inch and smaller.
Licensee's Code Relief Reouest:
The licensee requested authorization to exempt the condensate portion of the FWCI system by the exemption criteria of Article IWC-1222(c) of the 1986 Code that applies to components within systems or portions of systems other than RHR, ECC, and CHR systems. Article IWC-1222(c) exempts vessels, piping, pumps, valves, other components, and component connections of any size that operate (when system function is required) at a pressure equal to or less than 275 psig and at a temperature equal to or less than 200*F.
Licensee's Basis for Reauestina Relief (As stated):
"During the first and second Millstone Unit No. I ten-year inspection intervals, 10CFR50.55a required licensees to determine the extent of examination of Class 2 components by the requirements of paragraph IWC-1220, " Exempted Components," Table IWC-2520, " Examination Categories C-F and C-G," and paragraph IWC-2411, " Nondestructive Examinations," in the 1974 Edition and Addenda through the summer 1975 Addenda of the Code.
Paragraph IWC-1220(c) exempted feedwater coolant injection system components from examination (i.e., components which perform an emergency-core cooling function, provided the control of the chemistry is verified by periodic sampling and test).
"The January 1, 1990, revision to 10CFR50.55a that is the basis document used in preparation of the third ten-year program, does not allow licensees to utilize the exemption criteria within paragraph IWC-1220 of.
the 1974 Edition of the Code for Class 2 systems.
" Millstone Unit No. 1 is the only BWR in the country that utilizes the normally operating condensate, condensate demineralizer, condensate 4
booster, and feedwater system equipment as its high pressure coolant injection emergency core cooling system (ECCS).
"The FWCI system supplies make-up water from the main condenser hotwell through the condensate and feedwater system to the reactor vessel in the event of a loss-of-coolant accident (LOCA) due to a small break in the primary coolant system.
The Class 2 portion of the FWCI system consists of piping from the condenser hotwell to the outside containment isolation feedwater check valve and includes three condensate pumps, two steam jet air ejectors, one steam packing exhauster, three condensate booster pumps, seven condensate demineralizers, three reactor feed pumps, and ten feedwater heaters.
Piping and components from the outside containment isolation feedwater check valve to the reactor vessel is Class 1 and examined in accordance with the Code. The FWCI system operates continuously during all modes of power _ operation. The components normally in operation are the selected motor-driven condensate pump, selected motor-driven condensate booster pump, selected motor-driven feedwater pump, and associated piping and valves.
In the event a LOCA occurs and off-site power is maintained, the system will continue to operate in this mode.
If off-site power is lost in conjunction with a LOCA, the on-site gas turbine generator unit provides the motive power for the selected FWCI pumps restart and operation (i.e., one condensate pump, one condensate booster pump, and one feedwater pump). Additionally, redundant ECCS (i.e., low pressure coolant injection, core spray, automatic pressure relief, and the isolation condenser) are available to mitigate any size LOCA should the FWCI system fail.
"The FWCI system piping and components are carbon steel and not susceptible to intergranular stress corrosion cracking. The condensate and feedwater purity is stringently controlled via seven parallel-arranged mixed resin bed,' full-flow condensate demineralizers. During power operation, water conductivity and oxygen levels are monitored continuously. Daily water samples are taken and analyzed for conductivity, oxygen, chlorides, sulfates, and organic compounds.
"An erosion / corrosion pipe inspection program in areas susceptible to turbulent flow is ongoing in part because of IE Bulletin 87-01,
' Thinning of Piping Walls in Nuclear Power Plants,' and subsequent Notices on the same subject.
This testing program provides a thorough inspection since it inspects the entire pipe section (although it does not specifically inspect the weld).
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" Ultrasonic thickness test inspections since the Surry Nuclear Power
.l Plant incident has identified no significant degradation on the system
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which has been in operation for 20 years.
Remote visual examinations using TV cameras inside feedwater piping were also performed on selected areas with no observed erosion or corrosion degradation. As such, these inspection programs serve to provide assurance that the piping within i
the feedwater/FWCI system maintains its structural integrity and is, therefore, capable of performing its safety function.
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" Eddy current examinations are performed routinely on feedwater heaters as part of the ' Balance of Plant Reliability Program.'
" Visual examination of FWCI component supports to Article IWF requirements and surface examinations of integral attachments to Category C-C requirements have been performed in accordance with the Code during the second ten-year interval.
" Operability testing of the FWCI system's safety-related pumps and valves are performed in accordance with Generic Letter 89-04, ' Guidance on Developing Acceptable Inservice Test Programs.'
"The inclusion of the condensate portion of the FWCI system into the Class 2 Inservice Inspection Program is not warranted because the system is in operation continuously which provides constant verification that the system is operable. Millstone Unit No. 1 is the only nuclear power plant which utilizes the condensate and feedwater systems for routine operation and also as a ECCS for accident scenarios. The maintenance history for the feedwater system and components illustrates that the entire system is reliable and does not require frequent maintenance for repairs. Additionally, it should be noted that the FWCI system is not credited for any high energy line break in the turbine building.
Safe shutdown can be achieved by utilizing the isolation condenser or the low pressure coolant injection systems.
These methods provide redundant means for achieving safe shutdown following a postulated seismic event."
"It has been determined that the additional costs associated with performing inspections of the entire FWCI system in accordance with 10CFR50.55a is not commensurate with its increased margin of safety.
The present ongoing examinations and testing are believed to provide adequate assurance that the system's structural integrity is maintained and that the system will meet its intended function for those events for which it is credited."
Licensee's Proposed Alternative Examination (As stated):
... the following examinations and tests will be performed on the feedwater coolant injection system during the third ten-year inservice inspection interval at Millstone Nuclear Power Station Unit No. 1.
1.
"All of the FWCI system's component supports (that includes portions of the system that operates less than 275 psig and less than 200 degrees F) will continue to be visually (VT-3) examined in accordance with the requirements of Subsection IWF as amended byproposedreliefpequestIWF-1,whichwasapprovedbytheStaff on January 5, 1993.
2J. F. Stolz letter to J. F. Opeka, " Safety Evaluation of the Third Ten-Year Interval Inservice Inspection Program Plan for Millstone Nuclear Power Station.
Unit 1 (TAC No. M79663)," dated January 5, 1993.
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l 2.
"All of the FWCI system's pressure retaining piping and components (that includes portions of the system that operates less than 275 psig and less than 200 degrees F) will be visually examined for leakage every operating cycle as amended in relief req C-H-1,whichwasapprovedbytheStaffonJanuary5,1993,yest relief request C-H-1.
3.
"The non-exempt FWCI system components (i.e., components that are greater than 4" NPS and that operate at pressures greater than 275 psig and at a temperature greater than 200 degrees F) will be examined in accordance with Section XI of the ASME Code 1986 Edition Table IWC-2500-1 Categories C-A, p',B, C-C, and C-F-2 as approved by the Staff on January 5, 1993.
4.
" Eddy current examinations of the feedwater heaters, which are
,l downstream of the condensate booster pumps, will continue to be performed as part of the Balance of Plant Reliability Program.
5.
" Pipe ultrasonic inspections will continue to be performed on components located downstream of the condensate booster pumps as part of the Northeast Utilities Erosion / Corrosion Program."
Evaluation:
10 CFR 50.55a(b)(2)(iv) states that Class 2 pipe welds in the emergency core cooling system (ECCS) shall be examined; Article IWC-1221(a) of the Code allows the exemption of ECCS components that are NPS 4 inch and smaller.
The licensee is requesting authorization to exempt the condensate portion of the FWCI system by use of Article IWC-1222(c), which is for systems other than ECCS, based on operating pressure s 275 psig, and operating temperature 5 200*F.
The licensee has determined that the number of FWCI non-exempt piping and component welds that are eligible for examination [using the appropriate exemption criteria of IWC-1221(a)] is approximately 1200.
The examination sample specified for Code Examination Category C-F-2 welds is a 7.5% statistical sampling. The ASME Code committees and the NRC have determined that 7.5% is a representative sample for the safety systems cited in the regulation.
Reasonable assurance of the continued inservice structural integrity of the ECCS can be provided if this 7.5%
statistical sampling is maintained at the different nuclear plants.
Based on the importance of the ECCS, it is determined that the licensee's proposal will not provide an acceptable level of quality and safety. A 7.5% representative sample of the welds in the ECCS should be examined each inspection interval.
Therefore, it is recommended that relief be denied.
'The licensee's referenced approval of this alternative examination in the January 5,1993, Safety Evaluation is not applicable to non-exempt FWCI system components.
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3.0 CONCLUSION
The INEL staff has reviewed the licensee's submittal and concludes that for Request for Relief No. B-G-1-1 (10/93) and Request for Relief No. IWA-2, the proposed alternatives will ' provide an acceptable level of quality and safety, and, therefore, should be authorized pursuant to 10 CFR 50,55a(a)(3)(i),
For Request for Relief No. IWC-1 (10/93),.it is recommended that relief be denied.
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