ML20028B471
| ML20028B471 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 11/19/1982 |
| From: | Withers B PORTLAND GENERAL ELECTRIC CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR TAC-07716, TAC-7716, NUDOCS 8211300411 | |
| Download: ML20028B471 (11) | |
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November 19, 1982 Trojan Nuclear Plant Docket 50-344 License NPF-1 Mr. Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington DC 20555
Dear Mr. Eisenhut:
Control of Heavy Loads, NUREG-0612 Attached is a supplemental response to our letters of September 22, 1981 and May 17, 1982 in order to close out action in accordance with Section 2.1 of your December 22, 1980 letter.
Sincerely, e
.-1 1.. n_
Bart D. Withers Vice President
! OM Nuclear Attachment c:
Mr. Lynn Frank, Director State of Oregon Department of Energy 8211300411 821119 PDR ADOCK 05000344 P
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i Trojen Nuciscr Pl:nt Darrall G. Eistnhut Dockst 50-344 Novatber 19, 1982 License NPF-1 Attachment Page 1 of 10 SUPPLEMENTAL RESPONSE TO PGE LETTERS OF SEPTEMBER 22, 1981 AND MAY 17, 1982 1.
In accordance with Section 2.1.3.d of the NRC letter of December 22, 1980, the following information is provided concerning compliance of special lif ting devices (SLD) at the Trojan Nuclear Plant with ANSI N14.6-1978 requirements as referenced in NUREG-0612. -This response supersedes previous responses concerning SLDs.
NRC Request:
2.1.3 With respect to the design and operation of heavy-load-hand-ling systems in the containment and the spent fuel pool area and those load-handling systems identified in 2.1-1, above, provide your evaluation concerning compliance with the guide-lines of NUREG-0612, Section 5.1.1.
The following specific information should be included in your reply:
d.
Verification that lif ting devices identified in 2.1.3-c, above, comply with the requirements of ANSI N14.6-1978 or ANSI B30.9-1971, as appropriate.
For lif ting devices where these standards, as supplemented by NUREG-0612, Section 5.1.1(4) or 5.1.1(5), are not met, describe any proposed alternatives and demonstrate their equivalency in terms of load-handling reliability.
PGE Response:
The SLDs governed by ANSI N14.6-1978 at the Trojan Nuclear Plant are the reactor vessel head lifting device, reactor vessel internals lifting device, and reactor coolant pump motor sling. SLDs used with the inservice inspection tool and the spent fuel shipping cask are provided by outside contractors. Plant procedures require that con-tractors' compliance with ANSI N14.6-1978 requirements be verified during procurement of their services.
An evaluation of the reactor vessel head lifting device, reactor vessel internals lifting device, reactor coolant pump motor sling, load cell, and load cell linkage was performed to demonstrate the acceptability of these devices to meet the requirements of NUREG-0612.
l This evaluation consists of a comparison of the ANSI N14.6-1978 requirements and the requirements used in the design and manufacture of these devices (since ANSI N14.6-1978 had not been issued at the time these devices were f abricated) and a stress analysis in accor-dance with the design criteria of ANSI N14.6-1978.
In areas where i
this evaluation determined noncompliance with ANSI N14.6-1978 require-ments, actual design or fabrication methods are provided or suitable alternatives are identified to demonstrate compliance with the intent of NUREG-0612 and/or ANSI N14.6-1978.
Trojan Nuclocr Plcnt Darra11 G. Eistnhut Dockat 50-344 Nov2aber 19, 1982 License NPF-1 Attachment Page 2 of 10 The following conclusions were reached as the result of this evaluation:
1.
The ANSI N14.6-1978 requirements for design, fabrication, and quality assurance are generally in agreement with those used for these SLDs.
2.
The SLDs adequately meet the ANSI N14.6-1978 criteria for tensile and shear stresses and meet other appropriate criteria for buckling and for loading conditions that result in bearing or combined shear and tensile stress.
As stated in our letter of May 17, 1982, the dynamic load effects _during a lift of the reactor vessel head were determined to be approximately one percent of the static load
[as discussed in NUREG-0612, Section 5.1.1(4)] and are, therefore, considered to be adequately accounted for in the stress design factors of 3 and 5 for yield and ultimate strength, respectively.
The comparison of ANSI N14.6-1978 requirements and the requirements used in the design and manufacture of these SLDs has shown that these devices are not in strict compliance with all of the ANSI N14.6-1978 requirements. Listed below, for those areas of noncompliance which are considered most important in demonstrating continued load-handling reliability, are the requirements under which these devices were designed and fabricated and/or methods for demonstrating continuing compliance with the intent of NUREG-0612 or ANSI N14.6-1978:
ANSI N14.6-1978 Section Number Requirement / Response 3.1.4 The designer shall indicate what repair proce-dures are permissible and set criteria for acceptable repair procedures and testing.
PGE Response:
Any repair to these SLDs is considered to be in the form of welding. Pins, bolts or other fasteners shall be replaced, in lieu of repair, in accordance with the original or equivalent requirements for material and nondestructive testing. Weld repairs and examinations shall be performed in accordance with current Plant procedures.
l 3.2.1.1 Some materials have yield strengths very close and to their ultimate strength. When materials 3.2.6 that have yield strengths above 80 percent of their ultimate strength are used, each case requires special consideration and the fore-going stress design factors do not apply.
Design shall be on the basis of the material's fracture toughness, and the designer shall establish the criteria.
Trojtn Nuclear Plsnt Darrell G. Eistnhut Dockat 50-344 N vssbar 19, 1982 o
License NPF-1 Attachment Page 3 of 10 ANSI N14.6-1978 Section Number Requirement / Response Unless exempted by the provisions of Paragraph AM218 of the ASME Boiler and Pressure Vessel Code,Section VIII, Division 2, materials for 4
4 load-bearing members shall be subjected to a drop weight test in accordance with ASTM E208 or a Charpy impact test in accordance with ASTM A370 at a temperature 40*F (22'C) below the anticipated service temperature and shall meet the requirements of the design specification.
PGE Response:
i High strength materials are used in these SLDs.
Although fracture toughness was not tested, materials used ( ASTM A304, AISI 4340) which have yield strengths above 80 percent of their ultimate strengths do have acceptable fracture toughness _haracteristics.
The stress design factors of 3 and 5 of Paragraph 3.2.1, together with the original material selection, are con-sidered sufficient to provide reasonable assurance that brittle performance of the SLDs is not a concern.
L 3.3.6 An actuating mechanism shall be used, if needed, to securely engage or to disengage a SLD and a container. A position indicator shall be used in conjunction with an actuating j
mechanism when it is difficult to see the connection between the lifting device and the container.
PGE Response:
The reactor vessel internals lifting device employs a long-handled tool to engage the SLD and the internals. The tool depresses a i
spring-loaded tube and turns the engaging screw into the internals.
No specific position indi-cation is provided.
Scribe marks on the tool and the visual difference in the top of the spring-loaded tube are considered sufficient indication that the internals are engaged. For the reactor vessel head and RCP motor lifting devices, an actuating mechanism is not utilized.
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Trojcn Nuclser Pltnt Darrall G. Eicinhut Dockat 50-344 Nav:nber 19, 1982 License NPF-1 Attachment Page 4 of 10 ANSI N14.6-1978 Section Number Requirement / Response 4.1.6 Provision of a quality assurance program neces-and sary to conform to the applicable requirements 4.1.7 of Code of Federal Regulations, Title 10, Part 50, Appendix B, which requires the fabri-cator to organize, plan, establish, document, implement, and maintain systems using written procedure.s. The extent to which 10 CFR 50, Appendix B, applies to the SLD in question is defined in the design specification - critical items list.
Provision for identification and certification of materials, as required by the design speci-fication and the critical items list.
PGE Response:
A formal quality assurance program for the manufacturer was not required for all items.
The manufacturer's welding procedures and non-destructive testing procedures were reviewed by Westinghouse prior to use.
Most of the criti-cal load carrying members required lettet. of compliance for material requirements.
Westing-house performed certain checks and inspections during various steps of manufacturing. Final Westinghouse review included visual, dimen-sional, procedural and cleanliness inspections, a review of personnel qualifications and, in most cases, issuance of a quality release to ensure conformance with drawing requirements.
No information that a quality release was issued for the reactor coolant pump motor sling has been found, although Westinghouse performed the final inspection.
5.1.1 Verification by records furnished by the and designer and the fabricator that the perfor-5.1.2 mance criteria have been met by the design specification and that the design specifica-tion has been met by the fabricator.
Verifying by acceptance and functional testing performed or observed by himself or his repre-sentative that the performance criteria have been met.
Trojen Nuclesr Plant Darrall G. Eisenhut Dockat 50-344 Novenber 19, 1982 License NPF-1 Attachment Page 5 of 10 ANSI N14.6-19/8 Section Number Requirement / Response PGE Response:
Letters of compliance for materials and speci-fications were required for verification with I
original specifications. The Westinghouse quality release is considered an acceptable alternative to verify that the criteria for letters of compliance for materials and speci-fications required by the Westinghouse drawings and purchasing document were satisfied. Load testing was performed on the reactor vessel 1
I head and internals lifting devices at field I
assembly. These were 100 percent load tested, and nondestructive testing was conducted on critical welds following the load tests. The l
reactor coolant pump motor sling legs were l
proof tested to 2.7 times the rated load value.
5.1.3 Verifying by scheduled periodic testing that the SLD continues to meet its performance cri-teria and continues to be capable of reliable and safe performance of its functions, and providing a system that indicates the date of expiration of the validity of the test.
PGE Response:
The PGE testing and inspection program is i
discussed under Section 5.3 of the Standard.
i 5.1.4 Providing an operating procedure for the use of the SLD outlining proper use and maintenance and noting any limitation to its use.
PGE Response:
Present Plant operating procedures for the reactor vessel head and internals SLDs address the use and limitations of these devices. The use and limitations of the reactor coolant pump motor sling will be included in the reactor coolant pump motor removal procedure. Main-j tenance, testing and inspection requirements, addressed under Sections 5.3 and 5.4 of the Standard, will be included in the appropriate Plant maintenance procedures.
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Trojcn Nuclser Plsnt Darr211 G. Eiccnhut D:ckst 50-344 Nov:xber 19, 1982 License NPF-1 Attachment Page 6 of 10 ANSI N14.6-1978 Section Number Requirement / Response 5.1.5 Providing each SLD with identification that will serve to relate to its intended use and that may be used to record its history.
PGE Response:
It is obvious, from their designs, that these SLDs can only be used for their intended purpose and that their parts are not inter-changeable. Therefore, labeling each SLD for its intended purpose is not necessary.
5.1.6 Maintaining a record of the history of the SLD or component, including documentation of required testing, all uses of the device, any incidents in which the device or any parts may have been loaded beyond the loads for which it was qualified, damage, distortion, replacement, repair, alterations, and inspections.
PGE Response:
A record of the history of the SLDs is being maintained in accordance with current Plant procedures.
5.1.7 Removing from service any SLD or component for which the period of test validity has expired, which has experienced any incident causing doubt as to its continued compliance, or which has been damaged.
PGE Response:
Any SLD which has experienced any incident causing doubt as to its continuing compliance, or which has been damaged will be removed from service until it can be shown to be in compli-ance or until repaired.
SLDs shall not be used until the required visual inspections have been performed in accordance with PCE's rasponse to Section 5.3 below.
5.1.8 Since the SLD may be employed by users other than the owner, the owner may have to delegate some of his responsibilities to a user.
In such cases, the owner shall verify that the user will conform to his practices of use and
Trojsn Nucicar Plcnt Darrsll G. Eiz:nhut D ckat 50-344 NavrEber 19, 1982 License NPF-1 Attachment Page 7 of 10 ANSI N14.6-1978 Section Number Requirement / Response recording of use, incidents, or damage, and will remove from service any device about which there is some doubt.
PGE Response:
SLDs at the Trojan Nuclear Plant will only be used by the owner.
5.2.1 Prior to its initial use, each device shall be subjected to a load test equal to 150 percent of the maximum load to which the device is to be subjected. Af ter sustaining the load for a period of not less than 10 minutes, critical areas, including all load-bearing welds, shall be subjected to nondestructive testing in accordance with 5.5 of this Standard.
PGE Response:
The head and internals lif ting devices were load-tested at assembly to approximately 100 percent of the rated load followed by non-destructive testing of critical welds. The reactor coolant pump motor sling legs were proof tested to 2.7 times the rated load value.
Load testing to 100 percent is considered adequate in view of the safety factors designed into these SLDs.
5.2.2 Except where load-bearing welds are involved, replacement parts fabricated to the same design, from the same heat of material, and processed in the same lot at the same time as parts that have successfully passed the load test described in 5.2.1 may be qualified by the testing of the initial sample.
Any part with load-bearing welds shall be individually tested and inspected in accordance with 5.5 of this Standard.
PGE Response:
Replacement parts, should they be required, shall be made of identical or equivalent material and inspected as originally required.
Only pins, bolts, and nuts are considered replacement parts for the reactor vessel head and internals lifting devices.
e Trojnn Nuclscr Plant Darrall G. Eic2nhut D:ckst 50-344 Novrnb:r 19, 1982 License NPF-1 Attachment Page 8 of 10 ANSI N14.6-1978 Section Number Requirement / Response J
I 5.3 Testing to verify continuing compliance.
PCE Response:
The requirements for testing of SLDs imposed by ANSI N14.6-1978, Section 5.3.1, are not justi-fled for Trojan. The 150 percent load test option is impractical because of the limited capacity of the Containment polar crane.
l Removal of the devices from Containment for j
load testing is also impractical. The dimen-sional testing and inspection option is more practical but cannot be justified on an annual basis since the frequency of use and, there-4 fore, the wear on the SLDs is much lower than for shipping containers for which the Standard was written.
In addition, the SLD for the reactor vessel internals is painted and highly contaminated from being submerged in the refueling cavity and, therefore, the testing and inspection on an annual basis is not justi-fled from an ALARA standpoint as it would result in an exposure of 5 to 10 man-rem.
Therefore, PGE is implementing the following program for continued testing and maintenance at the Trojan Nuclear Plant which we feel adequately meets the intent of NUREG-0612 and ANSI N14.6-1978. This program will require an annual comprehensive visual examination by qualified personnel prior to first use of the i
lifting devices. This inspection will check all critical load-bearing welds and components for evidence of degradation or cracking.
Qualified personnel will also conduct a visual I
inspection for obvious deformation or cracking prior to each successive use of the SLDs during j
j that year. Additionally, a nondestructive examination of major load-carrying welds and critical areas will be conducted once every 10 years, which is the periodicity for repaint-l ing the reactor vessel internals lifting device.
This testing interval is justified because of the low usage (20 to 30 times) the SLDs receive l
over a 10 year period.
Since these SLDs are I
very large, dimensional testing would be of l
little if any benefit and therefore will not be conducted.
i
Trojan Nucicar Plant D2rrall G. Eissnhut Docket 50-344 Noverbar 19, 1982 License NPF-1 Attachment Page 9 of 10 ANSI N14.6-1978 Section Number Requirement / Response Should major maintenance or repair be required, 4
or if the SLDs are subjected to an incident causing permanent distortion of any load-bearing component, a 100 percent load test shall be performed following repair to verify operability.
Should out-of-Plant repairs be j
required, a load test in excess of 100 percent will be considered.
j 5.4 Maintenance and repair.
PGE Response:
Maintenance and repair shall be conducted in accordance with current Plant maintenance procedures.
i 5.5 Nondestructive testing procedures, personnel qualifications, and acceptance criteria.
PGE Response:
Nondestructive testing, personnel qualifica-tions, and acceptance criteria shall be in accordance with ASNT-SNT-TC-1A-1968 and Plant operating procedures.
6 SLDs for critical loads.
PGE Response:
]
Through load-drop analyses and procedural con-trols, compliance with NUREG-0612, Section 5.1 has been demonstrated, and therefore this section is not applicable to the use of SLDs.
2.
A further review of the handling of heavy loads over an open reactor vessel has demonstrated the need for more flexibility in the load-handling restrictions previously implemented as a result of our September 22, 1981 submittal. There are a number of miscellaneous loads (eg, inservice inspection tool, vacuum cleaner, etc) that at times are required to be lif ted over an open reactor vessel for main-l tenance, inspection, etc.
In lieu of completing a detailed analysis l
for each of these loads, yet in the interest of meeting the intent of NUREG-0612, it has been determined that in such cases a safety factor of 10:1 for the lift would be a sufficient alternative. This requires that the sling used to lif t the load be selected such that a safety factor of 10:1 is maintained and, for the Containment polar crane i
Trojan Nuclear Plant Darrall G. Eicanhut Dockst 50-344 Nov:2btr 19, 1982 License NPF-1 Attachment Page 10 of 10 auxiliary hoist (which has a 25-ton capacity with a 5:1 safety factor),
that the load not exceed 12.5 tons. Provisions to allow such lif ts will be incorporated into Plant operating procedures, however, previously implemented restrictions will not be totally relaxed.
As discussed in out letter of May 17, 1982, these procedure changes will be approved by the Plant Review Board prior to implementation.
Unnecessary lifts over an open reactor vessel will not be allowed simply because a 10:1 safety factor exists. These provisions will only be for necessary lif ts over the vessel and must be planned for and approved by supervisory maintenance or operations personnel. All other precautions and limitations of heavy load handling procedures will apply to these lifts.
SAB/4b1m936
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