ML20127K655

Rept to ACRS in Matter of Util Application for CP for Monticello,Mn,Nuclear Unit 1

ML20127K655
Person / Time
Site:	Monticello
Issue date:	01/24/1967
From:	US ATOMIC ENERGY COMMISSION (AEC)
To:
Shared Package
ML20127K646	List: ML20127K642 ML20127K655
References
NUDOCS 9211200463
Download: ML20127K655 (31)

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/' q l Yk MN l' 4 7367 U, S. ATOMIC ENEROY COF"GSS10N DIVISION OF REACTOR I.ICENSINO REPORT TO ADVISORY COMMITTEE ON REACTOR S AFECUARDS IN THE MATTER OF NORTHERN STATES POWER CO"PANY gPLICATION FOR CONSTRUCTION PERMIT FOR THE MONTICELLO, MINNES(YTA. NUCLEAR UNIT NO, 1 DOCKET NO, 50-263 Note by the Director of the Division of Reactor licensine I

The attached report has been prepared by the Division of Reactor Licensing for consideration by the ACRS at its February 1967 ineeting.

w OFMC4AL "d3E GJAY 9211200463 67012.4 PDR ADOCK 05000263 0 PDR

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4 1.0 Int roduction l 1.1_ Purpose This report is confined to a review and evaluation of the design, fabrication,

, and erection of a reactor vessel which Northern States Power Company proposes to-i field-assemble, at the company's Monticello, Minnesota, site, from segments small enough to ship by railroad. Since this is the first time such a proposal has been made (as we indicated to you in our September-2,1966 report), and since it is i likely that additional proposals will follow, we believe.that a detailed evalua-1 tion of this proposal should be made. The following questions have set the goals a

for this report:

. (1) How is the integrity of the vessel affected by field-assembly methods instead of the generally accepted shop-assembly methods?

(2) What can be done beyond the ASME code requirements to improve the reactor vessel integrity?

1.2 ,S,i,ta it The Northern States Power Company, . Minneapolis, Minnesota, has submitted an L

3 application dated August 1,1966, for a Section 104b (Utilization Facility) i License and Construction Permit, for a single cycle. boiling water reactor of -

1469 FMc (494 FMt) to be located on a 1325 acre site on the west bank of the j -- Mississippi : River 'in Wright County, ~ Minnesota, approximately 3 miles northwest of f Monticello, Minnesota. General Electric will- furnish the complete nuclear power 1 -

plant which is similar except for size, to Dresden Units 2' and 3, and Guad-Cities a

No. 1 and 2.

Tha Monticello, Minnesota, site is inaccessible for a barge of the capacity.

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- to transport a completely. fabricated reactor vessel (the accepted practice for d

nuclear power applications in the United States up to the present time) and 2

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therefore it is a practical necessity that pre-assembled reactor vessel senments small enough to be shipped by railroad be field erected.

It should be noted that site erection of heavy-wall,'non-nuclear vessels, l comparable to the NSP reactor vessel, has been preferred over shop fabrication in at least one situation that we are familiar with and will discuss in more detail in part 1.3.1 of this report.

j The Chicago Bridge and-Iron Company will design (with some transient analysf s e

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assistance f rom General Electric), fabricate, and simultaneously field assemble

the reactor vessel and containment structure.

l.3 Review Method Emploved by DR1, The requirements for the design.of the Monticello reactor vessel as contained 4

j in the ASME Code Section III and the General Electric APED Specifications are identical to those of other reactor vessels- currently beinn fabricated for 3 Ceneral Electric. Our evaluation of the technical feasibility of erectinn large-i reactor vessels in the field concentrates essentially, therefore, on those features associated with field erection which are different from the presently j accepted shop practices.

l A comparison of the reactor. vessel field. f abrication methods to the methods employed by the two leading shop f abricators has been made and is . presented in i'

Appendix-A of this' report. In making the comparison- between shop- and field-4 f abricated reactor vessels, considerations important to improved reactor vessel

! ' integrity have been identified: and logically become a part ' of our evaluation as .

I presented in part 3. We have been aided in this effort by pressure vessel consultants , Parameter, Inc. , of Elm Grove , Wisconsin.

Much of the information presented in the comparison of shop- and . field-fabrication methods, Appendix A, was obtained orally during; field trip meetings, i

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1.3.1 Meetines and Field Trins 4

Bethesda Meetine During a meeting on October 11, 1966, in Bethesda, . Maryland, of Northern States Power Company, General Electric Company, and Chicago Bridge & 1ron

Company representatives, we and our consultants heard an oral presentation which, in general, described the methods to be employed in field erection of the NSP pressure vessel and in some areas made comparisons to shop pressure-vessel assembly practices.

Atlantic Re finerv. Philadelphia. Pennsv1vania A 575-ton stainless steel clad vessel, approximately .79 f t. in heicht and 13 ft. inside diameter with 6-7/8 in thick walls, was being assembled at the time.of our visit. When placed in service, this vessel will operate at approxi-i mately 1800 pai and 800*F.

4 Conditions for transporting a shop-assembled pressure vessel to this site by barge could be considered very favorable, and a second slightly smaller shop-fabricated high pressure vessel is to be shipped to this same location by water.

It appears, therefore, that in a choice between field and shop erection of large i

pressure vessels, other factors besides shipping considerations influence the l

l outcome.

! Our attention during the tour was drawn to the methods and procedures for assuring that correct weld electrcde materials are used, to the techniques employed .fer hct ultrasonic tests of field weld areas, to the provisions for preheating during welding, to the gas burners and temperature sensors and recorders for stress relieving operations, to the shelter provided for protection of the vessel and workers from inclement weather during- assembly, and, finally, l

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to the small trailer located near the vessel where all field-weld gamma graphs were stored. In contrast to the proposed Monticello vessel, there was no removable head, nor any necessity for vertical field welds since rings arrived intact from the shop.

Mount Ve rn eo . Indiana We toured the new Babcock & Wilcox facilities for fabricating large pressure (including reactor) vessels.

Chattancoca. Tennessee We toured the Combustion Engineering reactor vessel shop f abric& tion facilities.

Birmincham. Alabama We toured the Chicago Bridge & Iron Company shop facilities where the steel plates from the Lukens Steel Mill will be press-formedi stress-relieved, clad with stainless steel, and then as reactor vessel segments will be prepared for shiprent 1

to the Monticello site.

1,3.2 Paramete r. Inc. , Reoorts i Our reactor vessel consultant, Parameter, Inc. , has prepared the following reports :

AEC - DCL - 1 October 17, 1966 Report of Fketing on Site Assembly of Pressure Vessel for Monticello Nuclear l Generating Plant i AEC - DCL - 3 Novembe r 21, 1966 Report of Visits to Babcock & Wilcox, Combustion Engineering, and Chicago Bridge

& Iron Co. to Review Nuclear Pressure Vessel Construction Practices OFFHCHAL USE ONLY

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- 5-l AEC - DCL - 4 November 28, 1966 Supplementary Comments and Recommendations

! on Fabrication of Pressure Vessel for

- Monticello Nuclear Generating Plant AEC - DCL - 5 January 6,1967 Comments and Recommendations, Review of i Northern States Power Company Report,

" Design, Fabrication, and Erection of the Reactor Vessel" 1.3.3 Applicant Submittals Related to Reactor Vessel Evaluation Original August 1, 1966 . Facility Description and Safety -

License Application Analysis Report l

Amendment 1 September 8,1966 - Site Assembly of the Pressure Vessel Amendment 2 November 21, 1966 Design, Fabrication, and Erection of the -Reactor Vessel.

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e 4 2.0 Description of the Monticelle peactor Vessel l The internal pressure of the primary plant which includes the reactor vessel will be controlled during normal operation at approximately 1000 psie. The design ,

i pressure for the reactor vessel is 1250 psig at 575'F with a maximum allowable primary safety valve pressure setting of 1313 psig (design pressure plus 5% in

accordance with code).

l The reactor vessel is a vertical cylindrical pressure vessel with an inside length of approxit.ately 63 feet and an inside diameter of 17 feet 1 inch. The-cylindrical wall thickness is 5-1/16 inches minimum plus a 1/8 inch minimum stainless steel internal cladding. The bottom head dollar plate through which

control rods and instruments pass is 5-7/8 inches minimum thickness plus 1/8 inch j minimum SS cladding. The top head thickness is 2-1/16 inch minimum plus 1/8 inch minimum SS cladding.

Water in the annulus between the- core shroud and the vessel will significantly 1

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reduce radiation levels at the vessel wall so that estimated exposure of t F.e vessel

at the reactor belt line will be approximately 8.2 x 10 17 nyt for neutron eneraies greater than 1 Mev at the end of 40 years.

. Steam exits through four 18-inch nozzles in the reactor vessel body, thus

! eliminating the need to break flanged joints in the steam lines when removing the 1

I vessel heads for- refueling. Relief valves and safety valves are mounted on

nozzles in the main steam lines.

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) Reactor vessel penetrations are listed below:

Quantity Size Function 2 36" x 28" Recirculation outlet 10 12" Recirculation inlet 4 18" Steam outlet 4 10" Feedwater

2 8" Core spray and flooding l 2 8" Instrumentation 1 4" Vent 2 4" Jet pump instrumentation 1 3" Control rod drive hyd. system return 1 2" Core diff. press, and liquid control 2 2" Instrumentation 2 2" Instrumentation J

j 1 1" High pressure seal leak det.

1 1 1" Low pressure seal leak det.

1 2" Drai- 4 121 6" Control rod drive 40 2" Flux monitor-The reactor vessel is supported by a steel skirt, the top of which is welded

)- to the bottom head of the vessel. The base of the ' skirt is welded to a support-l ing steel cylinder rising' from the secondary containment vessel which is ths i

! support for the vessel during erection.

The individual fuel assemblies of the nuclear reactor rest' en small' fuel support pieces mounted on top of the control' rod guide tubes. Each cuide tube OFFECHAL USE ONLY f

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i JFFHCHAL USE ONLi with its fuel support piece bears the weight of four fuel assemblies and rests on the control rod drive housing which is welded to the reactor vessel bottom head. An upper and lower grid and plate provides lateral support and aliennent 1 of the fuel assemblies.

A stainless steel cylindrical core shroud surrounding the core to separate

' core coolant from recirculated coolant is joined to a shroud support. The shroud support and legs are welded to the bottom head knuckle and essentially sustain all of the vertical weight of the reactor vessel internals with the exception of

the core.

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[. . JRCHAL USE ONLi 30 Evaluation l 31 Design Analysis As stated earlier in this report, the Monticello Nuclear Power Plant reactor vessel will be designed in accordance with the ASPE Boiler and Pressure Vessel l

Code,Section III,1965 Edition. The applicant's Amendment No. 2 Section III, i states the principal design requirements of the Code. Further, in the same section, the additional design requirements of General Electric are specified.

4 One of these is that all of the reactor vessel desien drawings and stress analysis calculations must be submitted by Chicago Bridge & Iron to General Electric for l

! an independent review and approval by an engineer experienced in pressure vessel design. However, the Atomic Power Equipment Department of General Electric j Company is providing a calculational service to Chicano Bridge & Iron Company in f the steady-state and transient thermal stress analysis of the reactor vessel i

because CB&I is not staffed sufficiently for a timely completion of the analysis.

l The net effect is that a General Electric desien engineer will review and approve i

calculations that , in part, are made by General Electric. It is recoenized that i

! this arrangenent does not provide for a complete independent stress analysis; but since there are no significant differences

• between the Monticello vessel design

• The only dif ference between the Monticello reactor vessel desinn and.that of

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other plants is in the supportina structure. The base of the reactor vessel I

support skirt will be welded to a steel supportine cylinder rather than to a flange to support the vessel durina field erection. The steel support cylinder will be designed to withstand the imposed loads resultina from the weight of the j vessel- test water and other construction and test loads. Shear rines or other i means will be provided to transfer the additional vessel loads resulting from the weight of fuel, water, piping, etc. - to the concrete pedestal and :to make the vessel skirt and pedestal act as a composite structure.

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JFFHCHAL USE ONL and the vessels of other CE plants, with the possible exception of the 36" x 28" diameter recirculation line nozzles which have been repositioned to permit shop-installation, this arrangement is acceptable. It is understood, however, that if during the design sta :e any new or unusual design considerations develop, the need for an independent review of those design analyses will be re-evaluated.

, It is our belief that the responsible stress analyst for the Monticello reactor vessel should be selected now to insure a coordinated effort between General Electric and Chicago Bridge & Iron. Further, we believe that the desien analyst should keep informed on the fabrication progress and review all specifica-tion deviation notices. The final stress report for the vessel should include all significant calculations required to accept any deviation and include a checklist of those parts requiring detailed analysis, as well as those which do not.

" State of the art" stress analysis methods for the "as-built" vessel, as well as the ideal Monticello vessel, should be used to improve the knowledge of vessel performance capability and reliability.

We expect to discuss this subjecc more fully with the applicant before the February ACRS meeting.

3,2 Materials According to. the applicant's Amendment No. 2, -Section 3.3, the materials selected for the Monticello reactor vessel are the same materials used in shop-f abricated vessels for boiling water reactors; namely, low alloy steel plate made in accordance with ASME-SA 533, Grade B, Class 1, per /.SME Code Case 1339-2 Heavy rolled plates commonly concain defects which of ten require that material be removed and replaced with a weld material of suitable properties. The defects are usually detected by ultrasonic test methods; and after further additional OFFECHAL USE ONLY s a -- - ,- , , , y

JFFHCHAL USE ONLi tests and evaluation at the fabricator's shop, repairs are made to meet General Electric specifications. We understand that at the present tire there is only-one supplier of plate for reactor vessels and'that the quality of plate -is on somewhat of a plateau with little near future prospect of reducing the size of inherent laminations so that repairs by velding will be unnecessary. This is not a problem unique to field erection, however, and the acceptable repair methods i developed in the past will be used as necessary on the Monticello reactor vessel plates.

All plate material will be detailed to the maximum. length and width dimensions that can be delivered from the mill and handled at the CB&I f abrica-tion shop in Birmingham, Alabama. The shell plates will be purchased in the quenched- and tempered condition, flame-cut to approximate size, and cold-formed by.a 6000-ton hydraulic press now being installed in the Birmingham shop. Cold-j pressing of rec.ctor vessel places to the desired shape is new in reactor vessel f abrication, although the ASME Code permits forming in any manner, provided specified materials properties can be obtained in the final condition. Appendix F of Amendment 2 describes the test to qualify cold-forming. The final acceptance of the vessel plate forming method, therefore, cannot be ascertained until the tests have been performed and the results analyzed.

The method of plate forming is determined largely by the shop facilities-and preferences of the fabricator and-is not a requirement due to field-erection.

Cold-forming because it is new to us, will be followed closely as the test results become available. -

3,3 Field Weldine Figure IV-4 ' of Amendment 2 shows the proposed shop-assembled pieces which must be welded together at the Monticello site to complete the reactor vessel.

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The site work can be divided into two parts.

Part 1 of Field Reactor Vessel Assenbly.

The basic assembly yard fabrication process will proceed as follows on head and shell components : (a) join shell halves into rings, bottom head to skirt I extension, and top head to flange on level work tables; (b) preheat to 300*

to 400*F and weld sections together; i.e., four shell rings, one bottom head with i

skirt and one top head with flange, and, finally, vessel closure flange to the j number 4 shell ring; (c) magnetic particle check weld periodically during i

deposition of metal as preliminary inspection step and replace any unsound material a

found therein; (d) hot ultrasonic test of welds before postweld heat treatment; (e) postweld heat t reat at 1150*F; (f) cool and radiograph welds; (g) cold ultrasonic test of welds; (h) manual overlay welds; (i) postweld heat treat; i

(j) cool and ultrasonic overlay; and (k) dye check overlay, In the. manner described, 8 vertical seam welds and 3 girth welds will be comnieted on level l ' ~

work tables. All of the foregoing will be performed within temporarv enclosures as protection against weather.

I Part II of Field Reactor Vessel Assembiv l

3 The bottom head and stub support skirt asseebly will be set in place on a 17-foot diameter cylindrical support skirt furnished in the drywell base of the containment structure.. The bottom head skirt assembly will be leveled using the devices on the stub skirt and intermediate skirt extension provided.

3 The leveling and plumbine procedure will be repeated af ter each. of the four

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shell rings is placed. The No.1 shell ring, complete with the -36 by 28 -inch l

diameter recirculation c.d other nozzles will be placed in position atop -the bottom head.. The girth seam will be fit, preheated, and the- hand welding will be i

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The No. 2 ring will then be placed, fit, and welded before any pestweld 4

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treatment is done in place on either ring. Temporary weather protection will be 9

provided by the multipurpose postweld heat treatment furnace structure._ The preheat will be maintained on the bottom head to No.1 ring girth seam until the No.1 to No. 2 girth seam is ready for postweld heat treatment. The_ shroud

! support skirt will be installed. At this time, the two rings will be stress relieved simultaneously in this temporary furnace.

Af ter the No. I and No. 2 girth seams have been postweld heat treated, the i temporary furnace will be converted into an air conditioned and ventilated work room around the bottom head and No.1 shell ring. A temporary cover will be j installed above this work area so that the balance of the vessel can be erected in the manner described above without interfering with the bottom head ' work i

(other than for in-place radiographic work on the closure girth seams of subsequent

rings). Four girth seam welds will thus complete the in-place field erection of i

the Monticello reactor vessel.

I The main joints to be welded in the field will utilize joints of Category A and/or B (paragraph N-462i Section III, ASME Code,- 1965 Edition). These joints are double-beveled and utilize a square bar as a spacer and temocrary backup bar .for

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welding. Weld edge preparation will be accomplished in the shop by flame-cutting i the desired bevels and land upon the place edges followed by grinding and ma;metic i-particle inspection in the field prior to welding.- Approximately 1-1/2 inches I of weld metal will be deposited on the inside before removing the spacer bar completely by arc gousing f rom the outside surf ace. of the vessel. The are gouged groove, as well as the weld surf ace inside, will be magnetic particle inspected I before welding from both sides progresses simultaneously.

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JFRCHAL USE ONLi It is anticipated that in order to deposit the first 1-1/2 inches of weld metal, a total of 30 passes will be required. These passes will be made with 5/32-inch and 3/16-inch diameter electrodes. Upon completi' s of the arc-gouging operation, the joint will be velded using 5/32, 3/16, and 7/32-inch diameter electrodes as the groove size permits until approximately 220 passes have been made to completely fill the groove.

The plate edges will be held in place by fit-up lugs welded securely in position. These lugs will be fabricated from material similar to the base metal and removed when no longer required.

The shielded metal are process was selected for all of the field welding of the major joints in the pressure vessel._ A special low alloy steel coated electrode developed to match closely the physical and chemical properties of the base metal will be utilized for the welding. The low hydronen and the composition 1

of the electrode result in a deposited weld metal that has notch touchness advantage from added nickel and an NDT temperature lower than the base metal.

A review of the typical results indicates a close matching of the base metal tensile properties and lower NDT temperatures. - Multiple pass techniques with i

l this electrode have the ef fect of producing good, sound welds without reducing any of the physical properties of the base metal below the standards required by Section III, ASME Code.

All electrodes will be removed from the hermatically sealed tins and baked at 800'F for one hour before using. Af ter the 800*F baking cycle, the electrodes will be stored in holding ovens maintained at 250 to 350*F until used. Welders will be permitted to remove only the number of electrodes that can be used in a 2-hour period.

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Preheating for welding shall be accocplished using electric strip heaters, insulation, and metal foil to provide a uniform temperature within the preheat ran ge. This preheat will be applied before any welding begins and will continue until the parts being welded have been given a postweld heat treatment.

Welding will be accomplished by the manual shielded metal are process (i.e. , hand welding using stick electrodes) and will be continuous as dictated by job site conditions. Continuous visual inspection of the welding will be accomplished by assigning one welding supervisor for every 6 to 8 welders The welding supervisor will have the responsibility to assure that all weldine procedures are carefully followed and that any irregulatity that occurs in a weld bead deposit is removed. There will be one project veldina supervisor who is responsible for all of the welding performed under the direction of the welding supe rvis ors . The welding supervisors will be under the direct control of the proje et welding supervisor.

All welding supervision, electrode care and preheating, tocether with constant visual inspection will be instrumental in reaching the ultimate. objective of no defects when the weldments are subjected to radiographic examination.

As a further check to assure a "no defect" condition after the base metal weldments are completed and prior to postweld heat treatment, these welds will be subjected to ultrasonic testing (hot U.T.) which is over and above the requirements of the ASME Code or the specifications that have been issued by Ceneral Electric.

Past C3&I experience has indicated that this inspection method (hot U.T. ) is an excellent quality control measure. IrregularPies that are dAclosed by this method permit corrections to be performed before the postweld heat treacment and radiographic inspection.

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JFFHCHAL USE ONLi After ultrasonic (hot U.T.) inspection has been performed and the base metal veldments p oven satisfactory, the joints will be given an interstage postweld heat treatment at ' 150* F (+25'F; -50* F) for approximately one hour before alleu-ing th2 weldments to cool to ambient temperatures. Ultrasonic and radiographic examinations of the veld joints will be performed upon the base metal prior to overlay welding of the joints with the shielded metal are process. Preheat will again be provided while overlay welding is being performed.

Field overlay welding will be performed using the stringer bead technique with a minimum of two layers to accomplish the full weld overlay thickness. The

first layer will be made with 1/8, 5/32, 3/16-inch diameter or a combination of sizes of E309 electrodes, and the remainder of the overlay will be accomplished with 1/8, 5/32, 3/16-inch diameter or a combination of sizes of E308 electrodes.

Close visual inspection will be performed by the welding supervisors to assure quality. Instruction and training of the welders will have been accomplished in the dob site school. The school will also train and qualify welders for the overlay welding of the forgings for the top and bottom flances that will be shipped to the job site direct from the forging manufacturer. Ultrasonic and dye penetrant examinations will be performed on the overlay in addition to close l

visual inspection during the actual welding operation.

1 Other Inconel (INCO 182 or INCO 82) overlays that may be - required as the design is finalized will be performed with either the shielded metal arc, metal inert gas arc or _ submerged arc processes or a combination of these processes.

Inconel will be deposited directly to the base metal and not on -top of a stainless steel overlay. This sequence eliminates the objectionable considerations that develop if an interlayer of stainless steel is between the bas; metal and the l Inconel, 1

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e The shroud support overlay buildup as well as the overlay buildups for the T

control rod penetration openings will be made with Inconel. The control rod penetration sleeves will be made from Inconel.

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1 Temporary furnaces will be utilized at the site to accomplish postwald heat J

1 treatment on the ground. The heat treatment will be monitored using thermocouples l

strategically placed upon the assembly and throughout the entire furnace.

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- Multiple point strip chart recording potentiometers will be used to record the actual heating progress of the assembly during the entire postwald heat treating i

Wela.

Assemblies completed in place wf11 be enclosed and postwald heat treated in i

a similar f ashion to those on the ground. Insulation, thermocouples, heatinn

sources, recording instruments, etc. , will '.,e geared to assure complete compliance with the Code and General Electric specifications. The postweld heat treatment

of subassemblies ca the ground or completed assemblies in place in the field will 1

be performed with the same limitations and controls as those in the shop.

The field welders will come frcm a Boilermaker Union which requires 100 percent local velders; however, an agreement has been made with the union to brina-50 percent outside welders from other CB&I job locations in to do the containment i

vessel. Som of these welders will be interchanced to work on the reactor pressure vessel. Thus, a large portion of the welders used in the reactor vessel

! will be experienced C3&I welders.

A welder training school will be set up and conducted by the experienced and qualified welding supervisors assigned to the Monticello project. Every welder will receive training in the job-site school using the materials of construction until a proficiency has been attained that will permit the welder to qualify fn i OFFHCHAL USE ONLY

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accordwice with Section IX, ASME Code and the General Electric specifications.

The training program will be conducted on the same material, including thickness, as the reactor pressure vessel. The material vill be obtained during the f abrication of the vessel. This training will be on test plates at the shop or field, not on the actual vessel. The welding program will indoctrinate the welders into what is expected in the form of " built-in quality" following approved welding procedures and of all the necessary examinations to assure

" built-in quality." The welding supervisors and welders will be trained and indoctrinated to the fact that veldments can be made to a quality that will not require time-consuming repairs after radiographic and ultrasonic inspections are made.

High quality and reliability of welds could be further assured, according to our welding consultant, Mr. J. Chyle of parameter, Inc., if additional information over and above the requirements of the ASME Boiler Code,Section III, were supplied for technical evaluation. Our combined recommendations are t

1. The vessel test plate metallurgical evaluation should include macro-etch cross sections showing the full thickness of the base plate and the veld inetal, and should include a thorough exploration of the heat affected zones by hardness surveys. These macro specimens will reveal plate cleanliness and any unusual structures. In addition, metallo-graphic examinations should be made of base metal, weld tnetal, and heat af fected zones for correlation with mechanical properties. Information and interpretation from the above procedures should be reported by an experienced metallurgist.

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$ i 2 yrom each vessel test plate, suf ficient specimens should be obtained so that the complete Charpy temperature transition curve can be

• 2 established for the base metal, heat af fected zone, and weld natal. The ,

i radius of the Charpy specimen notch should be checked using a maeni-a l fying radius comparator. The other sensitive dimensions should also be i 3

measured. The root of the notch should be located as close to the coarse grain region as possible (as determined by etchina) and should be wholly i

within the heat ~affected zone, with the notch oriented parallel to the  !

i i plate thickness.

j j 3. A complete table showing the chemical analysis of each plate, forginn, i and vessel test plate veld should be submitted. This analysis should ,

l include the trace elements, such as soluble and total A1, Ti, Sb, Sn, i

Pb, Cu, V, Zr, etc.

i j 34 Mondes t ructive Testine >

CB&I, in Amendment 2, reports that several vessels have been fabricated in f the field that were too large to be shop-built, and these vessels have been

! radiographed in the field, using Cobalt-60 sources. The thickness of these i

l vessels ranged from 2-3/4 inches up to 1/4 inches, and were built to the l requirements of the ASME Code. No dif ficulties were encountered in obtaining i

l the quality level required by this' code, i

With respect to radiographic examination of completed field welds on the .

I ' Monticello reactor vessel, it is reported that" a radioactive source such as-l cobalt-60 or Iridium 192 will be employed.- The type of source used will be determined by the thicknesses ranging from 1/2 inch up to 3 inches, whereas i

Cobalt-60 will be used on steel thicknesses ranging from 1-1/2 up to 8 inches.

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!  ! 4 ll 1 The film to be used for radiography will be fine grain similar to Eastean Kodak 1

j Type AA, or equal.

The ASME Code requires, and CB&I will t.se, a .060-inch thick penetrameter ,

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for 4-1/2 inch thick steel having three holes, the smallest of these beine l

2T (0.12 inches) in diameter.

It is contended by CB&I that the radioactive source is more adaptable to j work in the field than a linear accelerator or betatron and that the quality level of inspection will be the same for bott. plant and field regardless of the type of radiation equipeent used. CB&I claims that the nondestructive testina pro-

, cedures in the field will be essentially the same as those performed in the shop.

j It is planned that radiography will be performed on the finished stress-relieved 1

weld utilizing a 100 to 150 Curie camma source during third shif ts or weekends so that work on the vessel will not be delayed.

Weld surf aces will be magnetic particle inspected at each 1-inch layer of j weld deposit. As e further check to assure a "no defect" conditen af ter the base metal weldments are completed and prior to postweld heat treatment, these welds l will be subjected to ultrasonic testing (hot U.T.) over and above the require-ments of the ASPI Code or specifications that have been issued by General Electric.

Irregularities disclosed by this method can be corrected before the postweld heng 4 treatment and radiographic inspection. When the base metal weldments have been proven satisfactory by the hot ultrasonic tests, the joints will be niven an i

interstage postweld heat treatment at 1150*F for approximately one hour before allowing the weldments to cool to ambient temperatures. Additional ultrasonic tests of the veld joints will be made, followed by radiographic examiaation as described above, prior to overlay welding the insida of the joints. Ultrasonic 4

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l and dye penetrant examinations will be performed on the overlay af ter completion.

Mr. Al F. Cota of Parameter, Inc. , our nondestructive-testing consultant,

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has reviewed tne nondestructive-testing procedures in Amendrent 2 lie advises i l

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1. The water washable, visible dye penetrant to be used by CB&I is rated as the least sensitive of these penetrants.

I 2. Permanent record penetrant systeem are available for all types of j penetranta.

j 3. Abrasive blasting, polishing, or buffing tend to close or lap the discontinuities that might be revealed by penetrant testing. If this surface preparation is necessary, etching should follow in order to allow the penetrant to enter any discontinuities that may be present.

4 The marnetic particle "Co-no go" type of report is undesirable. Tape transfer method of recording questionable or borderline areas will 1

overcome this- deficiency.

5. A specially trained ultrasonic test operator is necessary to obtain valid ultrasonic test results.

6. The ultrasonic sensitivity must be established with the same couplant applied to the standard reference area when calibrating the equipnent i

and evaluating sensitivity.

4

7. During radiographic examinations, the resolution of the penetrameter establishes an arbitrary radiographic quality level but does not assess radiographic ability to detect linear or crack-like discontinuities.

The gamma ray system proposed (C060, AA film and .020-inch _ lead filter) fails to resolve the fine crackline discontinuities. Extremely high t

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JFRCllAL USE ONL - 22-I j resolution radiography will detect cracks .002 inch wide. The crack i

sensitivity of a radiographic system is dependent upon the aliennent of I the center line of radiation with respect to an axis of the crack. Even under ideal radionraphic conditions, cracks, which present an axis of l more than 7' of f a parallel to the line of radiation, will not be

. detected.

i Two source positions (essentially stereo-radionraphy) will improvri l

! the crack detecting ability of the radiographic system. Three source positions et, eld be arranged to cover 40' of the material being inspected.

Source-spot size and source-film distance control the ceometric unsharpness of the radiographic imane. The smallest possible source size is indicated to keep this unsharpness at a minimum.

4 Other unsharpness factors are those caused by the inherent unsharpness of the source, the Compton scattering, and the film graininess.

The compton scattering reduces to a minimum value at between i

6 and 12 1EV.

I  ?!alti-energy sources produce varying degrees of scattering within l

the film emulsion, each energy registering the imane in a slichtly i dif ferent location, i

l Film graininess is considered to be a reflection of the grain clumping characteristics of the emulsion so this unsharpness can be controlled only by film selection and/or film processing. Film selection, as well as source energy level, influence the density and contrast of the radiographic image.

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JFFHCHAL USE DE Linear accelerators and betatrons both operate in the optimum I energy level range necessary to keep unsharpness at a minimum value.

j Both are capable of providing monochromatic energies in the 6 to 12 WN range.

While the linear accelerators have extremely small focal spots (0.1 mm) and extremely high intensities (1500 to 12,000 R per minute at i a meter), portable linear accelerators are not available, so the use of this equipment could not be entertained for field radiography.

A portable betatron with an 0.2 nm focal spot and an output of 150 R per minute at a meter, is available. Power requirements are 4-150 kva, 480 volts, 3 phase, 60 cycles. Use of the betatron would provide the ability to resolve the imace of many of the possible cracks, and certainly, supplemented by the Peneprint system, would improve the quality assurance level for the field fabricated reactor vessel.

The superior radiographic resolution of the betatron can be approached with the best gamma ray system, but such a system would i

require long source film distances, type T or M film (or equivalent) and extremely precise source location.

We feel that the best attainable nondestructive-testine meth:3s should be employed as an audit of vessel integrity which is independent of administrative procedures, controls , trainine, welding techniques , individual performance , etc. ,

and, therefore we fully endorse Mr. Coca's recommendations. The items listed above have been discussed with the applicant, and it is expected that the

applicant's response will be known by the February ACRS meetine, i

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3.5 Field Hnehininn ,

{ concurrent with erection of the vessel shell, the vessel top head will be

) positioned for drilling the 5-1/4-inch diameter hold-down bolt holes. With the i 1

i cover in the same position, the grooves for the two 1/2-inch diameter stainless t

"0" ring gaskets will be machined with portable CB&I equipment shown in rigure l IV-13 of Amendment 2.

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! The vessel closure flance after being welded to the No. 4 shell ring in the l

assembly yard will be drilled and tapped in place to receive the bushines and .

hold-down bolts as shown in Figure IV of Amendment 2 i

j The holes and sleeves- for the 1216-inch diameter control rod drive thimbles i

and the 40 2-inch. diameter holes for the in-core flux sensors will be provided, ,

4 utilizing precision-bored guide templates, optically aligned in a temperature

] controlled housing to guide- a vertical boring bar and cutter head as shown in Figure IV-14 of Amendment 2

i. r i At this time, not enough information has been provided to make a complete evaluation of field machining techniques to be used relative to the detailed dimensional specifications of the lionticello reactor. While there appears to be i

l no reason to believe that tolerance envelopes needed to meet over-all functional i requirements cannot be met by a combination of in-place machininn and built-in adjustments, it is felt that the detailed machining procedures should not be I overlooked as - construction procaeds. The chances to desien 'of the reactor 1

i internals, if any, (relative to shop constructed vessels) necessitated by field

erection should, we- feel, be identified and evaluated with respect to function, I structural integrity, and reliability.

4 a 1

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JFMCHAL-?$- USE ONL i 3.6 Fit-uD ,

t 6 f The plate edges for field joints will be held in place by fit-up lugs f

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welded securely in position, tie note that the shroud ring will be machined q

complete at CB&I's Creenville works for subsequent assembly into the lower head knuckle section at the site. As close fit-up for location and welding of the i

l shroud to knuckle circumferential seam is desirable, the shroud weld prep and knuckle veld prep should be match-machined. Since installation of the shroud ring precedes the first stress relieve, and post-heat will have to be maintainec i after welding of the knuckle to the dollar plate, installation of the shroud i

j ring will necessarily be a " hot" operation. Our consultant is concerned that i

distortion of the veld preparation for the shroud attachment during this weldint

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operation would affect fit-up and location of the shroud. In view of these f complications, consideration should be given, he believes, to shipping the ,

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shroud-knuckle as one piece. In addition, the erection sequence calls for  !

i leveling of lower head knuckle assembly with respect to the top of the knuckle.

If the shrinkage of the weld between the knuckle and dollar section is not I

i exactly uniform, leveling this top surface would not be a true indicator of the i orientation of. the centerline of the head. This centerline is defined by the ,

center penetration and the center of the knuckle top diameter. (The surface on i- which this definition center is based is not well defined.) Assuming that the above centerline controls, it is possible that the top weld preparation of the t l

knuckle will not be exactly level for attachment of the .first shell course, Our consultant feels that some. provision should be made for shimming to obtain ,

optimum verticality of the shell, that the need for this shim or adjustment should be taken into account in the welding procedure, and that the- need .for leveling

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JFFECHAL USE ONLi adjustment would also apply to the subsequent attachment of each shell course.

We must obtain the ox11 N '= input on these matters to properly resolve any concern for safety, althouch we presently feel that vessel interrity is not involved.

One area that must be clarified is the exact maximum allowable offsets for shell ring longitudinal joints and shell ring circumferential joints that exist prior to welding these joints. ASME Code, Case Interpretation 1365, extends the maximum allowable of facts of Table N-525 of Section III of the Code (for thick-nesses over 2 inches) from 1/8 inch to 3/8 inch for loncitudinal joints and 1/4 inch to 3/4 inch for cirettmferential joints. However,1365 may 1 e interpreted to mean that of fsets greater than the maximums allowed by 1365 are acceptable, provided that the requitements of Section N-414 of the Code, which pertains to basic stress intersity limits, are met. Normally, if offsets before weldine are larger than those permitted by 1365, calculations would be made af ter the vessel is completed. We recommend that the maximum allowable of fset be deternined before welding of the joints is allowed.

3.7 Oualttv control Quality control for the Monticello nuclear reactor is directed by a <:uality control manager assisted by two quality control coordinators. The primary objective of this group is to coordinate CB&I's many quality connected functions into a system that will assure that the reactor vessel produced will meet the quality requirements and to document the f act that these quality requirements have been met. Authority lines for project manacement and project quality control are separated by virtue of having both of these managers report directly to the regional manneer who, in turn, reports to the Vice President and Manneer OFFHCHAL USE ONLY

JFRCllAL-27-USE ONL of Operations. The quality control coordinator for manufacturing is concerned with nuclear reactors which will be scheduled so that vessels are spaced at least three months apart. The quality control coordinator for engineering, purchasing, and construction is concerned, however, with only one vessel from inception of contract through to completion.

Quality Control is responsible for proper procedures, calline upon experienced engineers or technicians to write each procedure (plant personnel for plant processes, construction personnel for site processes, and inspection personnel for inspection). Full use is made of operator experience. Each process procedure is reviewed by an "in-house" review board consisting of top company officers and managers prior to being submitted to the customer for approval.

Each item or piece of material received at the shop or site is to be covered by a work order and traveler cara which lists in sequence all of the operations and inspections which that particular item must underno. CB&I lists in Section IV (p.18) the various quality control checks which are above those

required by the ASitE code and because of the space requirement they_are not i

i repeated in this report. Written nondestructive-test reports are to be t

prepared for each radiographic, ultrasonic, magnetic particle, and liquid

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penetrant inspection. Walders' performance qualification certificates and test results will be available for review.

On the basis of quality control measures described by CB&I in Amendment 2, .

and considering our remarks elsewhere in Sections 3.3 and 3.4 of this report , we-feel that the CB&1 responsibilities in this important area relating to the i

vessel integrity are reasonable and well documented, i

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3. 8 livdrostatic Test Upon completion of the in-place machining work on the vessel flange and control rod sleeves, the reactor cover will be attached and the vessel prepared for test. A hydrostatic overload pressure test will first be perfornad at 125%

design (1560 psig) in accordance with AS!E Code paragraph N-714.1 following which the cover will be removed and the service gaskets installed, the cover replaced, and a leakace rate test will be performed at the design pressure 1250 psig. (Shop-fabricated vessels are shop tested at 125% of desien pressure.)

Af ter the over-pressure test is completed, all weld surfaces, including those of welds used to repair material, must be subjected to a magnetic particle examination or to a liquid penetrant examination to satisfy ASME Code paracraph N-618.2. We have listed the Code requirements as backnround for our- recommenda-tions which are to instrunent the reactor vessel so that stress measurements can be made during the hydrostatic tests and to examine the entire outer surface of the reactor vessel by magnetic particle and ultrasonic tests after the hydro-static tests are completed.

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j 40 Summary and conclusions A significant portion of the Monticello reactor vessel f abrication will be l performed in the shop and, in this respect, vill be comparable to similar vessels made by other vessel fabricators. The balance of the f abrication and assembly at the Monticello, Minnesota, site will be performed according to CB&I utilizing proven techniques and equipment without compromise in the design quality or functional requirements of the vessel. Our evaluation of the information pre-sented-in Amendment 2 and Appendix A of this report leads to the conclusion that s-the integrity of the field-assembled reactor vessel will be at least equivalent to that of a shop-fabricated reactor vessel.

l Our understanding that all nozzles on the vessel, except one small nozzle 1

in the bottom head, were to be installed and nondestructively tested in the i

l Birmingham, Alabama, shop of CD&I helped to shape this conclusion. There is 4

I some uncertainty, however, in whether or not recirculation nozzles will be i

installed in the field, and therefore some concern exists that the problems j related to the installation of the 36 x 28-inch recirculation nozzles are 4

clearly understood.

i- We feel that the following tests and procedures beyond the minimum ASME code Rad CB&I and CE requirements would significantly increase the confidence

{ in the quality level of the Monticello vessel, and expect to have the applicant's l reaction by the February 1967 ACRS meeting.

(1) Use of betatron radiography at the site as well as the CB&I: Birmingham

) shop to assure an extra margin in obtaining the required radionraphic quality.

level and to provide a greater probability of finding crack-like defects.

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(2) Use of more sensitive and permanent-record type of surf ace inspection '

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] methods .

(3) Feedback of all dimensional and materials condition data to the design l

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analyst for evaluation and disposition. Application of analytical methods to i

l assessment of the as-built vessel with respect to the theoretical vessel.

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(4) Stress analysis determination of waximum allowable of fsets for shell l ring longitudinal and circumferential joints ortor t,o,o weldine.

(5) Attention to the effect of dimensional and mechanical changes from i the design of previous vessel internals and attachments to accommodate field i

l assenbly and machining on functional reliability.

i j (6) Performance of tests and recording of detailed properties of materials i of construction to evaluate effects on the welding processes and over-all i structural integrity of the vessel.

i l (7) Magnsflux inspection and ultrasonic inspection (shear wave and longi-tudinal wave) of all vess,el surfaces, except veld ' deposit cladding, following completion of the 125% design pressure hydrostatic acceptance test.

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