ML19345G962
| ML19345G962 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/10/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19345G961 | List: |
| References | |
| NUDOCS 8104220911 | |
| Download: ML19345G962 (12) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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E WASHINGTON. O. C. 20555 SAFETY EVALUATION SY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING RELIEF FROM REOUIREMENTS OF SECTION XI ASME CODE NORTHEAST NUCLEAR ENERGY COMPANY MILLSTONE NUCLEAR PCWER STATION UNIT NO.1 C0CKET NO. 50-215
1.0 INTRODUCTION
3/ letter dated Sectember 13, 1980, the Nortneast Nuclear Energy Company (the licensee) submitted :he Millstone Unit 1 Inservice Ins;:ection and Testing Pro-gram for the second tr..'-year inspection interval. 3ased on the date of comercial oceration, tne second ten-year inscaction interval will begin on Cecemcer; 23, 1980, and end on December 23, 1990. The Inservice Ins::ection and Testing Program is based on the current inscection program, wnich was accreved in the Safety Enluation Report issued Se;: ember 19, 1979, in support of Amend-ment No. 54 to Provisional Operating License No. JPR-21, but u:: graded to the requirement of Section XI of the ASME Boiler and Pressure Vessel Code,1977 Edition, including Summer 1973 Addendum, i
Paragrapn (g)S(iii) of 10 CFR Part 50.55a requires that the licensee notify the Nuclear Regulatory Commission when it is determined that conformance with
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certain Section X1 ASME Code requirements are impractical and that information be submitted in support of this determination. After review and evaluation, relief may be granted and alternative requirements may be imposed pursuant to paragraph (g)6(i) of 10 CFR Part 50.55a by the Nuclear Regulatory Commission.
We have completed our review and evaluation of the request for relief from certain Section XI ASME Code inservice inspection requirements for the Millstone Unit 1 facility. The specific requests for relief and our evaluation follow.
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. 2.0 EVALUATION OF RE00ESTS FOR RELIEF FRCM SECTION XI CODE REQUIREMENTS 2.i CLASS 1 COMPONENTS A.
Reactor Vessal 1.
Relief from the volumetric examination of one longitudinal and one circumferential shell weld in the beltline region of the vessel.
(Item No. 31.10, Examination Category 3-A, Table lWB-2500-1)
Code Recuirement The Code requires tnat one circumferential and one longitudinal beltline region weld be volumetrically axamined to essentially 100% of the lengtn during the second inscection interval. The welos should be located at a structural discontinuity, if any. The examinations may be performed at or near the end of the inspection interval.
Basis for the Recuested Relief The reactor vessel is insulated with permanent reflective insulation and surroundeo by the concrete biological shield. The annular scace between the inside diameter of tae biological snield and tne outside diameter of the insulation is a ncminal 5-1/2 inches. Thus, access for removal of the insulaticn panels is extremely limited, and this inaccessibility precluces direct examination of tne beltline region welds ficm the outside surface.
The interior surface of the reactor vessel is stainless steel clad, and the vessel 's internals, shroud, and jet pumps would make-an internal volumetric examination of the beltline region welds imoractical.
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The Northeast Nuclear Energy Company agreed to perform the augmented program approved by Amendment No. 64 to Provisional Operating License No. CPR-21 when examining these welds. The augmented program consists of:
(a) Examine volumetrically at least 100% of accessible length of each longitudinal weld and at least 100% of the accessible length of each circumferential weld, from either inside or outside the vessel.
l (b) Visually inspect to the extent practical, and fran the vessel inside 3urface, the areas of the welds required to be examined.
i (c) In the event that Code unacceptacle flaw is detected, ICC% volumetric examination snail be performed on the welds to meet Section XI ASME Coce requirement. This examination will be cerformed from tr.e inside surface of the reac or vessel.
3 Evaluation and Conclusion Imposition of the Section X1 ASME Code requirement would suoject the licensee to extreme hardship in necessitating removal of the concrete biological shield and the permanently installed insulation to perform the required examination of the welds fran the vcssel outside surface.
The boiling water reactor vessel is calculated to receive a relatively low fluence. Therefore, the amount of irradiation damage in the belt-line region should not ce significant curing this inspection interval.
The reactor vessel is monitored for radiation damage in the beltline region. The surveillance program is in ccmpliance with ASTM E185-66 and has been evaluated with resoect to the requirements of ASTM E185-73.
We have determined tnat the program meets the essential requirements of ASTM E185-73 and conforms to the intent of Appendix H of 10 CFR Part 50.
The program will provide sufficient data to monitor radiation damage to the vessel beltline materials tnrougnout the service life.
In addition, the vessel was designed and fabricated in accordance witn the rules of Section III of the 1965 Edition of the ASME Boiler and Pressure Vessel Code. We have evaluated the fracture toughness procerties and find that they meet the principle requirements set forth in Appendix G of 10 CFR Part 50. Utilizing the results of the surveillance program to monitor material damage from neutron irradiation and the guidelines in Regulatory Guide 1.g9 to establish operating limitations will insure that the reactor vessel will be operated in compliance witn the requirements of Appendix G of 10 CFR Part 50. This will provide an acceptable margin
~ of safety to prevent brittle fracture of the vessel during any conditions of normal operation, including antic! pated operational occurrences and system hyorostatic tests, to which it may be subjected during the remain-ing service life.
We conclude that the vessel design, on going surveillance program of the reactor vessel materials from the beltline region, and the augmented examination requirements are adequate to provide an acceptable level of safety and assurance that the vessel structural integrity will not be compromisec curing the inspection period.
. 3.
Pumos 2.
Relief frem the scheduling requirement for the visual examination of internal surfaces of at least one pumo in each group of pumps performing similar functions in the system.
(Item No. 312.20, Examination Category S-L-2, Table IWB-2500-1)
Code Reouirement A visual examination (VT-1) of the internal pressure boundary surfaces of one pump in each of the group of pumas performing similar functions in the system shall be examined during each inspection interval. The examination may be performed at or near the end of the inspection interval.
Basis for the Recuested Relief The Northeast Nuclear Energy Comoany orcocses to costoone the visual examination of the internal surfaces of the recirculation pumos until the need arises to perform major maintenance On the pumps.
In the absence of required recirculation puma maintenance, the Section XI ASME Code require-ment to disassemble and perform a visual examination of the internal sur-face of the pumo casing is an impractical requirement. Disassemcly and inspection of a recirculation pump would result in personnel exposures of about 100 man-rems and require approximately 5200,000 in spare parts and other costs. The turden imposed by the Section XI Code examination require-ment in the absence of maintenance warrants neither the personnel exposure nor the spare parts and other costs.
The Millstone Unit I recirculation pumps have coerated satisfactorily during the first inspection interval. Visual examination of the pump casing exteriors during operational leak tests at refueling outages has shown no sign of degradation. Based upon SWR service experience, cast stainless steel pump and valve bodies have performed acceptably with no degradation noted. The integrity of the Millstone Unit i recirculation pumps will be assured by performing monthly vibration signature analyses.
The analysis is sufficiently sensitive to detect enange: occurring as a result of either wear, internal surface erosien, or cavitation tamage.
-5 Accelerometers are mounted at right angles on each recirculation pump in areas most sensitive to detect change in internal configuration and coolant flow. Although monthly vibration signature analysis will be made, accelerometers will continuously monitor the pumps operational and structural characteristics.
In addition, the integrity of the recirculation pumps will be assured by a sensitive leak detection system. The Millstone Unit No.1 Tecnnical Scecifications limit unidentified leakage inside containment to 2.5 spm.
In the event a through-wall crack develops in eitner pump casing, the leak detection system is adequate to provide warning of pump degradation at an early stage of leak occurrence. Northeast Nuclear Energy Comoany will continue the augmented visual inspections of coth ::uma casing exteriors during periodic leak tests or hydrostatic tests during each refueling outage.
Evaluation and Conclusion,
We have reviewed and evaluated the basis of the request for relief frem the scheduling requirement of Examination Category 3-L-2, pump casings, of Section XI ASME Code. We conclude that the Section XI ASME Ccde requirement is impractical to conduct, and that relief is necessary. We have reviewed the alternate methods of inspection proposed in lieu of the impractical scheduling requirement, and conclude that they will ensure an adequate margin of component integrity. Pursuant to 10 CR Section j
50.55a(g)6(i), we have granted relief from the specific requirement that
- the licensee identified to be impractical for the Millstone Unit No. I facility, giving due consideration to the burden placed upon the licensee l
if the Section XI ASME Code requirement was imposed, and wnica we have determined that by granting such' relief will not endanger life or property or comrc-
%nse and security of the public.
[(Also see NRC safety Evaluation issued by letter (D. Crutchfield to W. Counsil) dated November 19, 1980.)]
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. C.
Valves 3.
Relief from visual examinatton of the internal surfaces of valves listed below in the Feedwater and Recirculation Water Systems (Item No.12.40, Examination Category B-M-2, Table IWB-25CO-1) 1-FW-11A 1-FW-ilB 1-RR-1A 1-RR-2A 1-RR-1B l-RR-23 1-RR-JA 1-RR 48 Code Recuirement Examination of the internal surfaces of one valve within each group of valves that are of tne same constructional design, e.g. globe, gate, or check valve, manufacturing T.etacd and that are performing similar functions in the system, e.g. containment isolation,-system' overpressure protection. The examination may be performed at the end of the inscection interval.
Basis for the Recuested Relief The valves are located in piping whicn penetrate the reactor vessel and cannot be isolated for disassembly in a practical manner for examination of the interna: surfaces. Relief is requested frcm the Code requirement on certain valves wnich either require drainir.] of the reactor vessel or installing plugs in the piping prior to disassembly of the valve for examination.
l The valves are subjected to the system leakage and hydrostatic tests required by IWB-5221 and IWB-5222 of Section XI of the ASME Code. There are valves of the same constructional design in other piping systems whien penetrate the reactor vessel. These valves will experience similar thermal-hydraulic conditions as the listed valves. Northeast Nuclear Energy Comcany will conduct the required Section XI examination on tnese 7
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. valves during this inspection interval.
In the event unacceptable conditions are observed, the listed valves will be examined in ccm-pliance with Section XI ASME Code requirements; or, in the event the reactor vessel is drained for scme other purpose, the internal surfaces of the listad valves will be examined.
Evaluatien and Conclusion We conclude from our review that impositien of the Section XI ASME Code requirement for the examination of the internal surfaces of the listed valves would subject the licensee to extreme hardship. Disassembly of the listed valves would require either draining the reactor vessel or installing plugs in the piping leading to the valves. The examination of valves of the same constructional design and experience similar thermal-hycraulic conditions as the listed valves will crovide assurance of acceptable structural integrity. The Section XI ASME Coce recuirement is impractical for the listec valves, relief is required and granted for tnis examination requirement.
2.2 CLASS 2 CCMPONENTS A.
Pressur: ?assels 4.
Request to substitute surface examination for volumetric examination of the circumferential shell welds of the LPCl heat exchanger. The surface examination sh.a1 cover at least 20% of the weld in proximity to the nozzles. (Item No. C1.30, Examination Category C-A, Table IWC-2500-1)
Code Recuiremen_t Volumetric examination of essentially 100% of the length of the tubesheet-to-shell weld. The examination is to be conducted during each inspection interval.
.a.
Basis for the Recuested Relief The gecmetry of the circumferential tubesheet-to-shell weld on the LPCl heat exchanger is suca that volumetric examination of tnis weld, using either ultrasonic or radiographic techniques, cannot be perfomed to obtain meaningful results. Northeast Nuclear Energy Company proposes to perfom a surface examination of at least 20% of the weld in the no::le arcas.
Evaluation and Conclusien Volumetric examination of the circumferential tubesneet-to-shell weld in :ne LPCl heat excnanger is an impractical requirement for the Millstone Unit i facility because the cesign of the neat exchanger pronibits meaning-ful examination of these welds by eitner ultrasonic or radiogracnic tecnniques. The licensee has proposed to use surface examination in lieu of volumetric examination. We require surface examination of at least 20% of eaca circumferent.ial tubesheet-to-snell weld be perfor:ec during d
tnis ins;;ection interval and that the examination be per omec on :ne part of the welds in the vicinity of the no::les. Due to the design of tne heat exchanger and the weld configuration, we conclude that the surface examination of the welds in these areas is adequate to assure the structural integrity of the iieat exchanger.
t Request to substitute surface examination for surface and volumetric examination of no::le welds to the Shutdown Heat Exchanger.
(Item No. C2.20, Examination Category C-8, Table IWC-2500-1)
Code Reauirement l
l Volumetric and surface examinations fcrpressure retaining no::le welds (7-1/2 inch nominal thickness) during each inspection interval (Reference l
Fig. IWC-2520-4,Section XI)
-g-Basis for Requesting Relief The snutdown heat exchanger nozzle to vessel weld is similar to Fig. I'AB-2500-10. The design of the nozzle prevents volumetric examination of the entire length of the weld. 'lolumetric examination of the weld, using either ultrasonic or radiographic techniques, cannot be perfor ned to obtain meaningful results. Northeast Nuclear Energy Comoany will perform surface examination of tnis weld.
Evaluation and Conclusion 3ased on drawings and the description of the shutdown heat excnanger nozzle to vessel weld, we conclude that this weld cannot be volumetrically examined over its entire length. 'lolumetric examination may be practical over part of the lengtn of tr.e weld. We require that volumetric examina-tien be performed to the extent practical, and that surface examination be perfomed on the nozzle to vessel weld during this inspection interval.
We conclude that volumetric examination to the extent practical,comoinea with surface examinat1on, will assure the integrity of the snutdown heat exchanger nozzle to vessel weld.
0.3 SYSTDI PRESSURE TESTS A.
Class 1 - Class 2 System Scundary 6.
Request relief from testing portions of the systems listed at the Code required test pressure:
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System Class 1 Soundarf Class 2 Soundar/
Feedwater 1-FW-llA/B HP Heater Discharge Isolation '/alves Standby Liquid Control 1-SL-8 1-SL-6 Core Spray 1-CS-6A/B l-CS-5A/B LPCI l-LP-11A/B l-LP-10 A/B i
Reactor Cleanup Return 1 -C'J-29 CR0 Return Penetration Cap
. Code Recuirement The pressure retaining components shall be subjected to a hydrostatic test at 1.10 times the system operating pressure at least once toward the end of each inspection interval.
Basis for the Recuested Relief Prr. cautions must be taken in view of the differences that exist in Class 1/ Class 2 system test oressures to prevent overpressucitation of the Class 1 components. The location of eneck valves in several systems preclude the Class 1 pressure test boundary from extending outward beyond the first of such valves, usually located inside contain-ment even though the class change boundary is outside containment.
In tnese cases, the Class i leakage and pressure test boundary would be the inside eneck valve. Conversely, pressure tests of Class 2 systems would have to ::e bounced at a stop valve wnich may or may not be the Class 1/ Class 2 boundary.
Evaluation and Conclusion The design of the listed Class 1 and Class 2 systems does not pernit system isolation at the boundary of the systems. In order to prevent overpressurization of the Class 1 portions, the portions of the systems which cannot be tested to ASME Section XI required cressure must be visually inspected to tne extent practicable during operation of the facility. Ic addition, portions of the system piping will be volu-i metrically examined under Category 3-J.
We conclude that the examinations and alternate testing procedure are acceptable to assure the systems integrity for safe operation of the facility curing this inspection interval.
l 3.
Class 3 Systems l
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Request relief from pressure testing the systems listed at pressures i
less than tne code required test pressure.
. Class 3 Systems System Pressure Test Pressure Service Water 55 55 Atnospheric Control 0
43 Code Requirement The pressure retaining components shall be subjected to a pressure test at 1.10 times the system design pressure at least once toward the end of each inspection interval.
Basis for the Recuested Relief These systems are required for continuous cooling of vital system components and cannot be removec frem service for the period of time required for the hydrostatic test. Due to the essential cooling functions of this system, the licensee will test these systems by utilizing the pump head ard throttled ficw for continuous cooling of the vital system ccmcc-nents. Referring to the. atmospheric control,'the system will be pneumatically tested in ccmpliance to the requirements of the integrated leak test progran of Appendix J,10 CFR Part 50.
Evaluation and ;onclusion l
The licensee has proposed testing the Service Water System for leakage by utilizing the pumo head and throttled flow for continuous cooling of the vital system components and pneumatically testing the Atmossneric Control System in compliance to Appendix J of 10 CFR Part 50. We conclude that the pressures at which the systems will be tested are greater than normal operating pressure and are considered adequate to provide assurance of the structural integrity of the systems.
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. for the
SUMMARY
Sased on our review, we have found that the Inservice Inspection Program 3.0 23, 1980 to December 23, 1990 is in compliance ten-year interval from December ty and materials to the extent practical, within the limitations of design, geome r,
E Boiler and Pressare of construction, to the requirements of Section XI of the ASM
'Ae, therefore, con-Vessel Code,1977 Edition, including Summer 1978 Addendum.
l from clude that tne Inservice Inspection Program for the ten-year interva in for Millstone Station Unit No.1 is December 28, 1980 to December 28, 1990 compliance with the requirements of 10 CFR Section 50.55a(g).
the certain Pursuant to 10 CFR Section 50.55a(g)(5)(i) we hereby grant relief from ll stone Unit i ten-year requirements, as discussed in this Safety Evaluation, for Mi 23, 1930 to December 23, 1990, and have determined inspection interval from Cecember d nger life or property that granting such relief is authori:ed by law and will not en a t consider-or the common defense and security and is otherwise in the public interes the relief is not grantad.
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ing tne burden imcosed on the licansee CONCLUS!CM ht (1)
'de have concluded, based on the considerations discus;ed above, t a :
4.0 bil i ty because the relief does not involve a significant incerase in the proba l
of consequences of accidents previously considered and does not invo ve a i
i f-significant decrease in a safety margin, the relief does not involve a s gn health icant hazards conr,ideration, (2) there is reasonable assurance that the d
and safety of the public will not be endangered by operation in the propose h
manner, and (3) such activities will be conducted in compliance with t e i l Commission's regulations and the issuance of this relief will not be inim ca d safety of the public.
to the common defense and security or to the health an April 10,1981 Date:
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