ML20106C319

Forwards Assessment of Special Lifting Devices for Compliance W/Guideline 4 of DG Eisenhut 801222 Request for Review of NUREG-0612 (Control of Heavy Loads).Facility in Full Compliance w/NUREG-0612 Section 5.1.1

ML20106C319
Person / Time
Site:	McGuire, Mcguire
Issue date:	01/31/1985
From:	Tucker H DUKE POWER CO.
To:	Adensam E, Harold Denton Office of Nuclear Reactor Regulation
References
REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR GL-81-07, GL-81-7, TAC-47127, NUDOCS 8502120222
Download: ML20106C319 (16)

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{ January 31, 1985 Mr. Harold 'R. Denton,' Director (Office of Nuclear Reactor Regulation

'U.

S_. Nuclear Regulatory Commission 4

Washington, D. C. 20555

Attention: 'Ms. E. G.IAdensam, Chief

Licensing Branch No. 4 l

Subjects

-McGuire Nuclear Station Docket Nos. 50-369 and 50-370 NUREG-0612. " Control of Heavy Loads at Nuclear Power Plants" Dear Mr..Denton't I

0n December 22.-1980 Mr. D. G. Eisenhut (NRC/0NRR) issued la letter. requesting that Duke Power Company review its controls for the handling'.of heavy loads to

' determine the extent to which the guidelines of NUREG-0612 were satisfied at McGuire Nuclear Station, identify the changes and modifications that would be required in order to fully satisfy those guidelines, and provide information

-documenting the results of our review and implementation.of-the required' changed

'and modifications. NRC Generic Letter 81-07.." Control of~ Heavy Loads", was issued on February 3, 1981 correcting several minor' errors in the December 22, 1980 letter.

4 Duke Power Company;has. submitted several responses to the.. letter, the most recent

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.being my letter dated' August 17, 1984..As indicated in.that response, the following are the.lif ting devices at McGuire which.can be categorized as special-lifting' devices and which must be evaluated'for compliance with the requiremenis of ANSI N14.6-1978-(as' supplemented by NUREG-0612l Section 5.1.1(4))'in accordance 1

with Guideline No. 4. "Special Lifting. Devices", of the December 22,'1980 letters

- a.; ' Reactor vessel' head ~1ifting rig an'd load cell' r

b.. ~ Reactor internals litting rig a>.

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Reactor l Coolant Pump motor lifting' rig J

d.

Control Rod Drive Mechanism (CRDM) missile sh'ield lifting rig-

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The NRC's' August 31,~1982 Draft TER supplied a list of the specific sections of-ANSI N14.6-1978 which must be addressed to determine compliance with Guideline'-

No. 4.

The August 17, 1984 submittal.provided an asses'ement of each of these.'

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sections demonstrating compliance for-itsas e and d above (which were designed and constructed by Duke Power Company),-and indicated that Duke was pursuing I

additional information on' items a and b from Westinghouse-(which manufactured g

i these twof tems) necessary for their evaluation, with an assessment to be made F'-

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)Mr. Harold R'.= Denton, Director i anuary 31,i1985 J

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and submitted later.- Attachment No. 1 provides this assessment for items a and b.

In-general, ANSI'N14.6 contains detailed requirements for the design, fabrication,

. testing, maintenance and quality. assurance of special'11fting devices. The o

evaluation l performed on. items a and b to determine the acceptability of these r

idevices to meet the above requirements included a detailed comparison'of the

-information contained-in ANSI N14.6~-1978 with the information that was used to

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7 design, manufacture,1 inspect and test-these special lifting devices. This y

comparicon shows thacL these devices meet the intent of the ANSI document for

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< design, fabrication and quality. control, and will meet the' requirements for.

1 periodic maintenance, proof and functional testing upon completion of certain actions (as indicated in Attachment No. 1).

Thel August ~17, 1984 submittal also indicated (Attachment No.1. Section 5.2.1)

_.that load tests for the McGuire Reactor Coolant Pump Motor Lif ting, Rig and. the CRDM' Missile. Shield Lifting Rigs were conducted, followed by non-destructive testing of.each rig for which no indications were present except for a surface

-lamination on one of the Unit 1 CRDM Missile Shield Lifting Rig's bolts used to

' attach the Missile Shield to the lif ting rig (this lamination did not occur

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as a result of the load test, but rather appeared to have been made at the time..

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of> manufacture).- Even though the bolt'successfully carried.the test load,' Duke

. Power stated that it would be replaced and an NDE performed' on the new bolt prior "to placing'it into service..The bolt heving the' surface lamination has been

< replaced,'with the new bolt having received an NDE. LIn addition, when the.RCP caotor. lifting rig was tested, the turnbuckles used with the rig-were mistakenly

.left off. 'ItLwas stated that the turnbuckles would be<1oad tested separately

' and then ' inspected, with' the results of: these tests to be forwarded upon completion.

These; turnbuckles were tension tested by Superior Rope & Sling (Atlanta).- Each turnbuckle was tested to 33,125 pounds and held for 10 minutes. This test load

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'm was equivalent to 150% of. the individual turnbuckle. load at' the rated capacity of the rig. An NDE was performed at'McGuire and.the-turnbuckles have been placed

'into service. This completes load testing ~of these special. lifting rigs for' j

NUREG-0612 at McGuire.

L With the submittal of.this information full compliance with Guideline No. 4 has been' demonstrated.. Consequently, McGuire has now demonstrated full compliance with all guidelines of NUREG-0612 Section 5.1.1.

Since information regarding

the-guidelines.of Sections 5.1.2 through 5.1.6 of NUREG-0612 as well as other.

aspects of the December 22, 1980 letter has been previously' submitted, this completes Duke Power Company's response to this letter for the McGuire Nuclear-

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Station..and.it is' requested that this issue be closed. Accordingly, Duke Power Company also considern that Facility Operating License NPF-17'(McGuire Nuclear..

Station Unit 2)-Condition 2.C.(8), " Heavy Loads", is satisfied..Please. advise if Lthere are.any questions'regarding this matter.

Very truly yours, f),b>

jg Hal B. Tucker

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Attachment

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tMr.' Harold R. Denton, Director January 31. 1985 L_ Page' -_

cc: IMr.2J. P. O'Reilly, Regional Administrator U.fS. Nuclear Regulatory Commission' Region II-101 Marietta Street, NW, Suite 2900 Atlanta.: Georgia 30323

'Mr.fDarl Hood Division of Project Management Office of Nuclear Reactor Regulation

'U. S. Nuclear Regulatory Commission Washington, D. C. 20555

'Mr. W. T. Orders Senior Resident Inspector McGuire Nuclear Station Mr. Amarjit - Singh Auxiliary Systems Branch

Division ~of Systems Integration Office of Nuclear Reactor Regulation

-U. S. Nuclear Regulatory Commission-

-Washington, D. C. 20555 s

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ATTACHMENT 1 ASSESSMENT OF SPECIAL LIFTING DEVICES FOR COMPLIANCE WITH GUIDELINE NO.4 ThE following.is an~ assessment of the Reactor Vessel. Head Lifting Rig and g;

Load Cell.and Reactor Internals Lifting Rig for Compliance with the require-

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"ments of ANSI.N14.6-1978 (as. supplemented by NUREG-0612, Section 5.1.1(4)).

Strict. interpretation of compliance of existing special lifting device

' design with'the criteria of.. ANSI N14.6-1978 cannot be made. Accordingly,

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_ 'only those-sections directly related to load-handling reliability of the s

_ lifting devices need be addressed.

Several sections of ANSI N14.6-1978

-dolnoicontainrequirementsconcerningload-handlingreliability: Scope

.(Section 1)' Definitions (2), Design Considerations to Minimize Decontami-

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'l nation Efforts (3.4), Coatings (3.5), Lubrication (3.6), Inspector's

. Responsibilities (4.2), and Fabrication Considerations. (4.3).

Evaluation u.

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of compliance with Section 6 (Special Lif ting Devices for Critical Loads)

need not be-included since no load has been determined to be a " critical

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i fload." The specific sections of-ANSI N14.6-1978 referenced below are'those fatate'd;in'theAugust 31,1982' TER for which compliance or equivalence must -

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, ibe demonstrated in order to determine compliance with guideline no.-4.

-(Note-that the TER' referenced the applicable ANSI sections both by. number

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1and brief description.- 'However, several ofethe" numbers did not correspond 3#

stolthefsectionsindicatedbythe.' description.- For'these cases Duke. Power _-

. assumed the descriptions referenced the section for which compliance orc equivalence with was intended.)

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'Section'3.1.1:. No l design s.pecification; was-written concerning these ' specific {

requirements.::Theireactor vessel head lift rig,'the reactor vessel intiernals clifting rig and load cell'and' load ce11' linkage, were designed and built for e

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.the McGuire Units 1 and 2 Nuclear Power Plants,'circaL1972-74. These devices'

. ere ' designed"to the requirement that the resulting stress in theLload' carry-w ing members when: subjected to-the total combined lifting weight should not

- i' exceed the ' allowable stresses specified in the AISC code (Manual of Steel

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-SConstruction, Seventh Edition, American Institute of Steel. Construction).,

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' 1Also'a(125 percent _ loa'd test was required on both devices, followed'by appropriate nondestructive tetting. These items were not classified as nuclear safety components and require: tents for formal documentation of design. requirements and stress reports'were not applicable. Thus, stress-reports and design' specifications-vere not formally documented.

. Westinghouse' defined the design, fabrication and quality assurance

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~ ' requirements on detailed manufacturing' drawings and purchase order

'. documents. Westinghouse also issued field assembly and operating instruc-tions, where applicable. Although no specific design specification was-written, the assembly and detailed manufacturing drawings and purchase

order documents contain equivalent requirements, including:

Material specification for all=the critical load path items e

to ASTM, ASME specifications or special listed requirements.

e All welding, weld procedures and welds to be in accordance

- with ASME Boiler and Pressure Code - Section IX.

.Special nondestructive testing for specific critical load e

. path items to be performed to written and approved pro-cedures in accordance with ASTM or specified requirements.

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e'.All: coatings to be performed to strict compliance:with specified requirements.

Letters of compliance for materials and specifications e

.were required for verfication with original specifications.

.No' limitations were identifed as to the use of these devices under adverse environments. Markings and nameplate information were not addressed.

These devices. for the'most part, were manufactured under Westinghouse surveillance with identified hol'd points, procedure review and personnel qualification which ' adequately meet' these ;related ANSI requirements.

Section 3.1.2:? A critical items' list has been prepared per Section 3.1.2 for the reactor vessel head lift; rig, the reactor vessel internals lift

- rig and the load' cell and load cell linkage. Load carrying members and

-welds of these special lifting devices are considered to be the critical

' items.- The list includes the materia 1' identification, and the applicable nondestructive volumetric and surface inspections that were performed in 1.

.the fabrication of these special lifting devices.

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nondestructive testing was not specified since the material selection and strength result in very low tensile stresses and thus, nondestructive test-ing was not justified. The material selection for all critical load path

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items was made'to~ ASTM, ASME or special material requirements. The material zrequirements were supplemented by Westinghouse imposed nondestructive test-ing, and/or'special-heat treating requirements for almost all of the critical 11tems.; Westinghouse required all welding, welders, and weld procedures to be in accordance with ASMF Boiler and Pressure Vessel Code Section IX for all welds.1 Westinghouse required a certificate, or letter of compliance y

that the materials and processes used by the manufacturer were in accordance s

_ with~the purchase order and drawing requirements. Westinghouse also performed 4

-final' inspections on these devices and issued quality releases for the in-ternals'and head lifting rigs.

3ection 3.1'.3:

Although a stress report was not originally required (as indicated in;Section 3.1.1), Section 3.1.3 of ANSI N14.6 and NUREG 0612 b

'Section 5.1.1(4) require a stress report to be prepared. Special loads 5-h,

'and allowable. stress criteria <are specified for this analysis.

Stress

-analyses were performed on the McGuire Units 1 and 2 reactor vessel-head

. lift rigi reactor vessel internals = lift' rig and the load cell and' load

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[f cellalinkage to determine the acceptability of these devices to meet the design requirements.of ANSI N14~.6. L As.part of the invoking of the ANSI

'N14.6 ' document,; the NitC requested 'utilitiies to demonstrate their compliance Y

with the stress criteria with some qualifying conditions. The stress-

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report for these devices wasiperformed in ace'ordance'with the criteria'of.

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' IANSI N14.6', including the'NRC qualifying. conditions of NUREG 0612;(Sectio'n

'521.1 (4))'. s /A11lof1the" tensile :and sh' ear Lstresses with' the except' ion of the '

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k rod housing and the guide sleeve meet the design criteria of Section.

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3.2.1.1 of ANSI N14.6,irequiring. application of stress design factors.of 4

threeland five:with the accompanying. allowable stress limits of yield and

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ultimate! strengths resp'ectively.1 In addition,L all of the~ tensile 1and

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7 cshear stresses meet-the; requirements of the AISC code..It should be noted.

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y that:the' design weight-used.for.;the reactor vessel internals life rig'in

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the stress calculations is based on the weightfof the lower internals.-

3 The lower internals are only removed when;a periodic inservice inspection C

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of the vessel is required (once/10 years). Prior to removal of the lower internals, all fuel is' removed. Thus the concern for handling over fuel

?is non-existent in this particular case. Normal use of this rig is for moving the upper internals which weighs approximately one-half of the lower internals. The design weight is based on lifting the lower inter-nals..Thus all the stresses could be reduced by approximately 50 percent and considered well within the ANSI N14.6 criteria'for stress design

. factors.

~Section 3.1.4: Although repair procedures were not originally identified by Westinghouse', Westinghouse states that-any repair to these special lifting; devices is considered to be in the form of welding. Should pins.

-bolts.or other fasteners need repair, they should be replaced, in lieu Lof repair, in accordance with the original or equivalent requirements for b

_ material and' nondestructive testing. Weld repairs should be performed in

~ accordance with the requirements identified in NF-4000 and NF-5000

,(Fabrication and Examination) of the ASME Boiler and Pressure Vessel Code,?Section III,. Division 1 Subsection NF.

All special lifting' devices Lare unique to the item they were designed to lift. -Any repairs made would.

,also:be_ unique to:each lifting device; therefore, all repairs-would be handledLon an individual basis and would depend.on the severity of'the

problem, TAny major repair;would~be subject to 125 percent load test.

Section 3.2.1:. NUREG 0612, 5.1;1(4) states:

'"In. addition, the st'ress-

' design factor stated in Section 3.2'.1.1 of ANSI N14.6 should be' based'

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<on the combined maximum-static and dynamic loads that couldibe imparted N

'on-the handling-device ~ based ~on characteristics of.-the crane'which'will-

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3 be used.- This is in lieu of the guideline in Section_3.2.1.1 of-ANSI s

N14'.6'which1 bases the stress design factor on only the. weight (static

load)-of-thefload and of?the intervening components of the special:

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handling device."' It: can be inferred from-this paragraph that the ' stress' a --7 W

design factors specified:in Section 3.2.1.1 of-ANSI N14.6"(3 and 5) are

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Enot all inclusive. -The application'of the ANSI N14.6 criteria for stress 0 : de' sign factor of 3 and 5 are'only:for shear and. tensile ' loading conditions.

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/Other!1oading.conditionsaretobeanalyzedtoother:appropriatecriteria.

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LAlso. it can beJinferred that the static load should be increased by an

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-amount' based on the crane dynamics characteristics. The dynamic charac-

teristics of the crane would be based on the main hook and associated

.wireiropes holding the hook. Most main containment cranes use sixteen

, (16);or more-wire ropes to handle the load.

Should the crane hook t

suddenly stop during the lifting or lowering of a load, a shock load

.could be transmitted.tolthe connected device. Because of the elasticity of.the sixteen or more wire ropes, the dynamic factor for a typical con-tainment crane.is considered to be not much larger than 1.0.

Both the

. reactor.vesse tl head and internals lift rig were originally designed to

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the requirement that all resulting stresses in the load carrying members, when' subjected to the' total combined lifting weight, should not exceed

the allowable stresses spec'ified'in the AISC code. The design criteria ofiSection:3.2.1.1 of7 ANSI'N14.6, requiring application of stress design factors of three and five with.the accompanying allowable stresses, are

to'be used for evaluating load' bearing members of a special lifting 4

Ldevicelwhen. subjected to loading conditions resulting in shear or tensile

. stresses. Application o t ese design load factors to other loading f h

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zconditions"is not addressedtin ANSI N14.6. However, these two stress

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designifactors have been used to determine tlie stresses-of the load

' ; carrying members when subject to other loading conditions',' viz, bending,

. bearing. This is.an extremely conservative. approach 1 and in several instances the resulting= stresses exceed the accompanying ^ allowable-stress y

Llimit; ;The,ps ts of~the internals-lift rig that'do not meet the ANSI' fN14 5 chiteria'when analyzed for'b~ earing. stresses are the. lower sling

<1eg clevis..the.. lower clevis pin,1the spreader-leg, the leg lug, and' tthe* guide' sleeve. However, siince bearing stresses are localized stresses, theylcanTb'e considered under Section3.2.'1.2[shichstatesthatthestress

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idesign' factors ofl3.2.1.1 areinot intended to apply to situations 1where-

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M d11'=high'lo'callstressesarerelievedbyslightyielding. None of the. bearings dtressesreachtheyieldstress,~nd,infact,allof'thebearingstresses a

Jaeet:the design crit'eria'of the AISC code.' The-combined tensile stress-

. from bending and tension, in the. leg lug of'the: internals lift: rig' exceeds Lthe'Section13.2.1.'1 criteria.-1 Bending, however, sis not a uniform stress,-

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I but a local fiber stress, and.is at maximum at the outermost fiber.

Even if the fiber stress reached anywhere near the yield stress, the rest of the cross section could assume the additional load. As indicated above,

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bending too can be considered under Section 3.2.1.2.

Bending contributes to'the major. portion of the stress, and, as a result, the tensile stress without bending is extremely low and well within the - Section 3.2.1.1 criteria. The combined stress also meet the 1AISC code criteria. The fillet weld connecting the leg lug to the leg channel and the mounting block welds on the_ internals lift rig meets the

' ASME criteria.for weld stresses based on base material properties. How-

'ever,.when applying the ANSI N14.6 3W and SW criteria to the nominal stress:value,'the ASME allow'ble stress value is exceeded.

But, since a

the ASME allowable stress is satisfied for this weld, it is considered acceptable. The rod housing and the guide sleeve do not meet the ANSI

=N14.6 criterion'in tension. However, they do meet the AISC allowables for' tension and thus the design of these is considered acceptable. High strength materials are used in some of these devices (mostly for pins, load cell). Although the. fracture toughness was not determined, the

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material was selected based on its fracture toughness characteristics.

However, in lieu of a different stress design factor, the stress design factors. listed in ANSI ~N14.6 Section 3.2.1 of 3 and 5 were used in the

. analysis and the resulting stresses are considered acceptable.- In sum-mary, a stress report has been generated which addresses the capability

-of these rigs.to meet the ANSI design stress. factors. The stresses on each of the parts which make.up the reactor. vessel head, load cell and

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. load cell linkage'and'the internals lift' rig were: determined. -All.of

~the tensile and, shear stresses, with the exception of the' rod housing and the guide' sleeve meet the design criteria of Section 3.2.1.1 of-

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' ANSI N14.6~,-requiring application of stress design factors of three and five with ' accompanying allowable stress limits of -yield and ultimate strength, respectively.

In' addition, all of the tensile and shear-stresses: meet the requirement of not exceeding the allowables.of the-u

,AISC code.

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1 Section -3.2.4 : Where necessary. the weight of pins was considered for handling.

A Section 3.2.5:

All slings at McGuire. Nuclear Station are inspected and

~ tagged with'a color coded I.D. tag annually. The slings comply with ANSI B30.9 (1971).

Section 3.2.6:

Fracture toughness requirements were not identified for.all the' material used in these special lift 1ng devices. However, the material

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selection was.-based on its fracture toughness characteristics.

- Section 3.3.1:

The lifting rigs are protected from the environment and

from galling..Lamellar tearing is not a problem.

Section 3.3.4:

The rig's design assures even distribution of load.

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. Section '3.3.5:

Locking plates, pins, etc.,' are used throughout.these.

. special lifting devices.

Section 3.3.6:

Remote actuation is only used when engaging the internals lift; rig'with the internals and position ~ indication is provided from the operating platform.

Section 4.1.3:

All critical-load. carrying members require certificates of coupliance for material requirements.. Westinghouse-performed.certain checks.

and' inspections during various steps of manufacturing. Final' Westinghouse review 11ncludes visual,. dimensional -procedural, cleanliness, personnel qualification, etc., and issuance of.a quality release to ensure conformance with drawing requirements.

Section 4;1.4: All the manufacturer' ~ welding procedures and nondestructive-s

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' testing procedures were_ reviewed by Westinghouse prior to c.a.

Westinghouse performed'certain checks and. inspections during various steps of manufactur-

'ing..' Liquid penetrant,' magnetic-particle, ultrasonic and radiograph 1inspec-P i tions wereL performed 1Ln accordance with ASTM specifications, 'ASME Code,

' Westinghouse process specifications or as-noted on detailed drawings and

- provide cimilar results to the requirement of the ASME' Code referenced in

- ANSI.N14.6 Section 5.5.. Westinghouse ' Quality Assurance' personnel performed 3

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some in-process and final inspections similar to those identified in ANSI

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[N14.6 Section 4.2, and 11ssued a Quality Release. Westinghouse also required

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? nondestructive tests and inspections on critical load path parts and welds both'as raw material and'as finished items. Westinghouse's objective was fs 'tio provide a quality product! and this product was designed, fabricated, iassembled and inspected in accordance with internal Westinghouse require-

. General good manufacturing processes were followed in the manufac-ments.

ture of these devices. Final Westinghouse review includes visual, dimen-sional, procedural, cleanliness, personnel qualification, etc., and issuance of a quality release to ensure conformance with drawing requirements.

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Sect ion - 4.' 1. 5 : All the manufacturers welding procedures and nondestructive

' testing procedures were reviewed by Westinghouse prior to use.

Final

Westinghouse review-includes visual, dimensional, procedural, cleanliness,

. personnel' qualification, etc., and issuance of'a quality release to ensure conformance vith drawing requirements.

Section 4.1.6: Westinghouse performed certain checks and inspections during various steps of manufacturing.- Final Westingbouse review includes. visual,.

dimensional, procedural, cleanliness, personnel qualification, etc., and

issuance of 's quality release to: ensure conformance -with drawing require -

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,Section 4^.1.7: JAll; critical load' carrying members require certificates of-compliance for material requirements. Westinghouse performed certain checks

.and inspections during1various-steps of manufacturing. Final Westinghouse 1 review"inclu' des visual,Idimensional,-procedural, cleanliness, personnelL f-

, qualification,'-etc.,'and issuance of a quality ~~ release tol ensure conformance

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.with. drawing' requirements.

iSection 4.1.9: ?All the manufacturers welding procedures and nondestructive C '

te'stiing' procedtires were' reviewed'by Westinghouse prior to use. _All critical.

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load carrying members require certificates of compliance for material require-b '

1 ments.' Westinghouse performed.certain checks and inspections d'uring various

-steps of, manufacturing. ' Final Westinghouse. review includes visual, dimen-

.sional, procedural, cleanliness,' personnel qualification, etc., and. issuance,

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4-of a-quality. release to ensure conformance with drawing requirments.

Although there wasn't any design; specification for these rigs, the t

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. Westinghouse Quality Release is considered toLbe an acceptable alter-

~nate to verify that the criteria for the letters of compliance for

-materials and specifications required by Westinghouse drawings and purchasing document were satisfied, i

-Section 5.1.3:

All special lifting devices are covered under our preven-tative maintenance (PM) inspe'ction program.- They are inspected prior to

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use at the beginning of an outage. A visual inspection is performed via

.a PM Work Request. Ths PM lists step by step instructions of which welds require visual inspection.

Section 5;1.4: - Operating instructions for the reactor vessel internals l'ifting rig' were furnished to Duke Power Company and operating procedures avere prepared and are used. Procedures for the removal or movement of equipment associated with special lif ting devices were developed or

. incorporated in.the procedures for work on that equipment. The operating procedureL does not outline. maintenance of the devices, however, = there is a yearly preventative maintenance which covers inspection. Each lifting

-device is. designed to lift only one piece of equipment and therefore are limited to those applications

' procedures specify use.of-appropriate rigs.'

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Section 5.1.5:

It is obvious, from their designs, that these' rigs are--

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. specific -lifting devices and can only. be used for their intended purpose and parts are not interchangeable.. Specific identification--of the rig

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willfbe made by marking with steucils, the rig name and rated capacity (probably on the spreader assembly).

s Section 5.1.6: - An equipment history is'kept for all PM Work Requests.

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Any. discrepancies found during the.PM inspections are repaired on that

'PM Work Request.

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e Section 5.1.7: ' All=special lifting' devices are inspected on an as 7

required basis. -If any indications.are found in the inspection areas, Lthe-device will be tagged and removed from service until the problem-Lis resolved..

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Section 5.2.1:.The reactor vessel head and internals lifting rigs were designed.and built for McGuire Units 1 and 2, circa 1972-74. A 125 per-cent load test was required on both devices, followed by appropriate nondestructive testing. Westinghouse also required nondestructive tests and inspections ~on critical-load path parts and welds both as raw material and as' finished items. Both the reactor vessel head and internals lifting rigs and-load cell were proof tested upon completion with a load of approxi-mately 1.25 times the design weight. Upon completion of the test, all parts, particularly welds, were visually inspected for cracks or obvious deformation and critical welds were magnetic particle inspected.

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addition the Westinghouse Quality Release verified that the criteria for 1etters of compliance for materials and specifications rcquired by the Westinghouse drawings and purchasing documents were satisfied. Although

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these rigs were not subjected to an initial acceptance load test prior to use' equal to 150 percent of the maximum load, the 125 percent of maximum load test that was performed should be acceptable in lieu of the 150 per-

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cent load test. The NRC has recognized that the specification of a 150 percent overload test is somewhat arbitrary and has provided for excep-tions to verbatum compliance with NUREG-0612 guidelines via the " Synopsis of issues associated with-NUREG-0612" (dated May 4,1983) which was trans-mitted with the January 12, 1983 TER. NUREG-0612 Section 5.1.1(4) also-indicates that certain load tests may be' accepted in lieu of certain material r equirements in the ANSI standard. The rigs were tested to-F 125% overload which has been' standard industrial pract' ice for some time.

s At-pres'nt, no-spare parts are' stocked for special lifting Section 5.2.2:

e devices. Appropriate measures (i.e., compliance with Section 5.2.2);will be taken should any spare parts"be stocked'in the future.

Replacement parts, should they be required, would be mEde of identical (or equivalent) material and inspections as originally required. Only pins, bolt and nuts are-considered replacement parts for the reactor vessel head and internal lift rigs.

'Section 5.3.1:.These special lifting devices are used during plant refuel '

ing which is,approximately.once per year. During plant operation these i

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- special-lifting devices are inaccessible.since they are permanently A

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-installed and/or remain in the containment. They cannot be removed from the containment unless they are (-eassembled and no known purposes exist

-for. disassembly., Load testing t o 150 percent of the total weight before

- each use would require special fixtures and is impractical to perform.

Crane = capacity could also be limiting.- Since the 150 percent load test is:v'ry impractical, annual load testing per Section 5.3.1 Part (1) for e

McGuire's lif ting' devices are omitted in accordance 'with Section 5.3.1 Part: (2) 'of ANSI N14.6-1978. Consequently, a minimum'of nondestructive-

. testing'of major load-carrying welds and critical areas is performe.1 as 1seraitted by.Section 5.3.1 Part (2). Prior to McGuire Unit l's second refueling the head and internal removal procedures will include the following: :" Prior. to use an'd after reassembly of the spreader assembly, l'ifting lug, and upper lifting legs'to the. upper portion of the reactor 4

~ vessel head lift rig, visually check all welds. Raise the vessel head siightly above its support'and hold for 10 minutes. During this time, visually inspect the sling, block lugs to the lif ting. block welds, and

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spreader lug to spreader' arm weld. :If no problems are apparent, continue to lif t,' monitoring the load cell readout at all' times. ~ Prior to 'use 'of theLreactor vessel internals lift rig,_ visually inspect the rig components and welds while on the storage stand for signs of cracks or deformation.

Check all bolted joints tofensure that they are tight and secure. After connection to the upper or lower internals, raise the' assembly slightly off lts s'upport and hold'for 10 minutes.' During this time, visually-tinspect the. sling block lugs to the lifting block welds. If no problems

-are apparent,' continue to lift, monitoring the load cell-readout'at all times.", [The above actions-do' not include a no'ndestructive test ~ of - these

- welds'because:

a)' access to thelwelds for. surface examinationLis diffi '

cult.,These rigs. are in containment and some contamination' is present;.

a b)L-a11' tensile and shear stresses in the welds are within the~ allowable

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stress;-c);the items that are welded remain assembled and cannot be.

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h misused for anyLother:lif,t.other than their intended function; d) to f:

l perform nondestructive tests would require:.1) removal of paint around'.

the area..to'be" examined which is contaminated; 2) performance of=either 7

s magnetic' particle inspecion.or: liquid penetrant inspection and 3) re-paintingjafter testing'is completed; 4) cleanup of contaminated items. -

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Performing nondestructive tests on these welds every refueling would increase the critical path refueling time. Dimensional checking is not included since these structures are large (about.14 feet diameter by 44-feet high) and the results of dimensional checking would always be questionable. :0ther checks on critical load path parts such as pins, are also not included since an examination of"these items would require disassembly of the special lift' devices.]'In summary, it is impractical to perform the 150 percent load test prior to each use.

It is considered

. that a 100 percent load test, performed on each device, at each refueling, followed by a visual check of critical welds is sufficient to demonstrate compliance. Upon completion of'the field assembly of the reactor vessel

- head. lifting rig, the assembly procedure calls for a 100 percent load test

- (lifting of the assembled head), with a visual inspection for any signs of distortion. A check (visual) of critical welds and parts will be conducted

-at initial lift prior to moving to full lift and movement for these devices.

Further note that with the use of the load cell for the head and internals lift rig, all lifting and lowering is monitored at all times.

Section 5.3.2:

McGuire intends to load test all special-lifting devices

- after major modifications or repairs in accordance with Section 5.3.2 of ANSI N14.6-1978.

Section'5.3.3: McGuire's special lifting devices are designed to specific pieces of equipment and should never be subjected to stress substantially greater than they were designed;for.

If they were subjected to an'over-stressed condition, appropriate measures would be taken.

Sectica 5.3.6:

All.special lifting devices are visually inspected by personnel using the device prior to each use-as specified'in the Lift

- Supervisor's Handbook.

Section 5.3.7:

Maintenance personnel inspect each lifting rig in accord-ance with PM requirements prior to outages (unless rig was inspected and used within the last 30 days). These rigs are inspected prior to outages (as opposed to every three months) in view of ALARA considerations, and since the rigs are only used during outages there is no reason to inspect them at three month intervals during power operation in which they wouldn't s_

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be used. In addition,'a periodic nondestructive surface examination of

. critical welds and/or parts will be performed once every ten years as part of an inservice inspection outage. - A PM will be written to cover this inspection.

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