LD-83-017, Forwards Draft Mods to Sys 80 Fuel Design for CESSAR-F. Mods Include Completion of Seismic & LOCA Loading Analyses. Changes Will Be Formally Incorporated Into CESSAR-F in Next Amend

Forwards Draft Mods to Sys 80 Fuel Design for CESSAR-F. Mods Include Completion of Seismic & LOCA Loading Analyses. Changes Will Be Formally Incorporated Into CESSAR-F in Next Amend

ML20079N921
Person / Time
Site:	05000470
Issue date:	02/28/1983
From:	Scherer A ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
LD-83-017, LD-83-17, NUDOCS 8303040467
Download: ML20079N921 (6)

Text

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' ' " C-E Power Systems Tet. 203/688-1911 Combustico Engineenng. Inc. Telex 99297 1000 Prospect Hill Road Windsor, Connecticut 06095 M POWER SYSTEMS Docket No.: STN 50-470F February 28, 1983 LD-83-017 Mr. Darrell G. Eisenhut, Director Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

Fuel Design Evaluation for CESSAR-F i

Dear Mr. Eisenhut:

The NRC'S evaluation of the Combustion Engineering (C-E) System 80" fuel design, as documented in the Safety Evaluation Report (SER), notes that some of the results of the fuel design analyses are not provided in CESSAR-F. Since the issuance of the SER in November 1981, C-E has performed the stress, strain and strain fatigue analyses for the System 80" fuel assembly, fuel rod, burnable poison rod and CEA, except for the seismic and LOCA analyses which are addressed on a plant-specific basis. The assembly lift-off analysis and the CEA axial growth analysis have also been performed and show acceptable results. These two analyses do not involve seismic and LOCA loadings.

l Attached are rough drafts of the modifications to be made tc CESSAR-F to reflect the completion of these analyses. A rough draft has been provided to permit the continuation of the Staff review of CESSAR-F. Formal incorporation of these changes into CESSAR-F will be accommodated in the next amendment.

If you have any questions on the attached, please contact me or Mr. G. A. Davis of my staff at (203) 688-1911, extension 2803.

Very truly yours, COMBUSTION ENGINEERING, INC.

A. E. Scherer Director Nuclear Licensing AES:las cc: Gary Meyer (Project Manager /VSNRC) 0 0 bbb 0 A PDR

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excess of Sg indicates an unusual shipping occurrence in which case the j

4y fuel assembly is inspected for damage prior to releasing it for use. -

l The axial shipping load path is through either end fitting to the guide j [ e

! tubes. A Sg axial load produces a compressive stress level in the guide tubes less than the two-thirds yield stress limit that is allowed h>y'-

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for normal condition events. The fuel assembly is prevented from buckling by being clamped at grid locations. For lateral or vertical $ N t/1 I

shipping loads, the grid spring tabs have an initial preload which yQe exceeds five times the fuel rod weight. Therefore, the spring tabs see s e-no additional deflection as a result of Sg lateral or vertical accelera- t N tion of the shipping container. In addition, the side load on the grid M3Jo%c faces produced by a Sg lateral or vertical acceleration is less than the measured impact strength of the grids. 2 l 5 @ y, The fuel assembly shall be capable of sustaining a 5000 pound axial B.

load applied at the upper end fitting by the refueling grapple (and --( jgy resisted by an equal load at the lower end fitting) without sustaining stress levels in excess of those allowed for normal operation. The q4

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%;; 5C 5000 pound load was chosen in order to provide adequate lift capability should an assembly become lodged. This load criterion is greater than h* J -J' S any lift load that has been encountered in service. j%

L C. The fuel assembly shall be capable of withstanding a 0.125-inen deflection 7.->@A 0 E in any direction whenever the fuel assembly is raised or lowered from a 2 5g$

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horizontal position without sustaining a permanent deformation beyond 41

% ,,,j E -5 the fuel assembly inspection envelope. ,

7 %>Ej Fuel handling procedures required the use of a strongback to limit the 2 k *S q

fuel assembly deflection to a maximum of 0.125-inch in any direction whenever the fuel assembly is raised or lowered to a horizontal position. - $ $ ~g This limits the stress and strain imposed upon the fuel assembly to e :7 ?

values well below the limits set for normal operating conditions.

adequacy of the 0.125-inch criterion is based on the inclusion of this The q

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limitation in specifications and procedures for fuel handling equicment, D}w%, NN # t, which is thereby constrained to provided support such that lateral deflection is limited to 0.125 inches. _ _I r . _ _ - . . - - .-.

4.2.3.2 Fuel Rod Desion Evaluation

, The evaluations discussed in tnis section are based on assumed fuel rod operation within certain linear heat rate limits related to avoiding excessive fuel and clad temperatures. Information concerning the bases for these limits is contained in Section 4.4 4.2.3.2.1 Results of Vibration Analyses l

Three sources of ::eriodic excitation are recognized in evaluation the fuel

! rod susceptibility to vibration damage. These sources are as described in l

Section 4.2.3.1.1.

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EEMIT@@Mf These sources of periodic motion are not expected to have an adverse effect on the performance of the fuel rod. Section 4.2.3.2.4 includes additional information on fuel rod response to the sources.

4.2.3.2.2 Fuel Rod Internal Pressure and Stress Analysis A fuel rod cladding stress analysis is conducted to determine the circumferen-tial stress and strain resulting from normal, upset, and emergency conditions.

The analysis includes the calculation of cladding temperatures and rod internal pressures during each of the occurrences listed in Section 4.2.1.1.

The design criteria to be used to evaluate the analytical results are specified in Section 4.2.1.2.1. Fuel rod stresses resulting from seismic events are calculated, using the methodology described in Reference 50.

4.2.3.2.3 Potential for Chemical Reaction A. Corrosion Zircaloy-4 fuel rod tubing has been visually examined in the spent fuel pool after three reactor cycles at Ft. Calhoun, two reactor cycles at .

Clavert Cliffs, and one reactor cycle at Millstone, St. Lucie-1, and Maine I Yankee. In addition oxide thicknesses were measured in the hot cell after one cycle at Maine Yankee. In all instances the oxide appearance and oxide thickness measured similar to autoclave behavior for that time and tempera-ture.

Coolant chemistry parameters have been specified that minimize corrosion product release rates and their mobility in the primary system. Specifically, the precare hot functional environment is controlled (pH and oxygen) to provide a thin, tenacious, adherent, protective oxide film. This approach minimi:es corrosion product release and associated inventory on initial startup and sucsequent operation. During operation, tne recommended litnium concentration range (0.2-1.0 ppm) effects a chemical potential gradient or driving force between hot and cooler surfaces (fuel cladding and steam generator tubing, respectively) such that soluble iron and .1ickel species will preferentially deposit on the steam generator surfaces. The associated ,

pH alsa minimizes general corrosion product release rates from primary  !

system surfaces. Moreover, the specified hydrogen concentrations range (10- l l 50cm'/kgSTPhinsuresreducingconditionsinthecore,therebyavoidinglow l solubility Fe . Additionally, dissolved hydrogen promotes rapid recombination of oxidizing species. (Recall, oxidizing spe:ies and a fast neutron flux are synergistic prerequisites to accelerated Zircaloy-a corrosion).

During operation lithium, dissolved oxygen, and dissolved hydrogen will be monitored at a frequency consistent with maintaining these parameters within their specifications.

Post-operational examinations of fuel cladding that has operated within these specifications, has shown no significant chemical or corrosive attack of the Zircaloy cladding.

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5fBMT @@{PV local power and burnup in the rod being examined. This procedure will yield conservatively high stored energy in the fuel rod under consideration.

The maximum power density, including the local peaking as affected by antici-pated operational occurrences, is discussed in Sections 4.3, 4.4, and Chapter 15.

4.2.3.3 Burnable Poison Rod 4.2.3.3.1 Burnable Poison Rod Internal Pressure and Cladding Stress A poison shim cladding anaiysis will be performed to determine the stress and strain resulting from the various normal, upset, and emergency conditions discussed in Section 4.2.1.1. Specific accounting will be made for differen-tial pressure, differential thermal expansion, cladding creep, and irradiation induced swelling of the Al 20,8 4C burnable poison material. Owing to the very low linear heat generatTon rates in these rods (maximum local is less than 1.5 kW/ft), the stress analysis can be accomplished using conventional strength of materials formulae, except for determining clad collapse resistance which will be done using the CEPAN computer model, (Reference 22).

The desi;;n criterie used to evaluate the analytical results are specified in Cu t i a- '.2.'.3.'.

4.2.3.3.2 Potential for Chemical Reaction A discussion of possible chemical reaction between the poison material and x the coolcnt was presented in Section 4.2.1.3.3.3, along with information on chemical compatibility between poison material and cladding. Since the cladding material is identical to that of the fuel rod (Section 4.2.1.3.2),

the description of potential chemical reactions between cladding and coolant in Section 4.2.3.2.3 is applicable to both fuel and poison rods.

The potential for waterlogging rupture in poison rods is much lower than that in fuel rods because of the smaller thermal and dimensional changes that occur in a poison rod during reactor power increases. Refer to Section 4.2.3.2.10 for a discussion of the potential for waterlogging rupture in fuel rods.

4.2.3.4 Control Element Assembly The CEAs are designed for a 10 year lifetime based on estimates of neutron absorber burnup, allowable plastic strain of the Inconel 625 cladding and the resultant dimensional clearances of the elements within the fuel assembly ,

guide tubes. J A. Internal Pressure The valve of internal pressure in the control elements is dependent on the following parameters:

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1. Initial fill gas pressure

2. Gas temperature

3. Helium generated and released

4. Available volume including B4C porosity Of t.'.e absorber materials utilized in the CEA design, only the B,C contributes to the total quantity of gas which must be accomodated withjn thdgontrol7 cl emep' t. The helium is produced by the nuclear reaction n +g8 T 3Li H and the fraction of the quantity generated which is actaally released

+to2th ]e, plenum is temperature dependent and is predicted by the empirical cquation discussed in Section 4.2.1.4.4 listing A.3. Temperatures used for release fraction calculations are the maximum p edicted to occur during norm:1 operation. l B. Thermal Stability of Absorber Materials None of the materials selected for the control elements are susceptible to thermally induced phase changes at reactor operating conditions. Linear thermal expansion, thermal conductivity, and melting points are given in Szction 4.2.1.4. .

I C. Irradiation Stability of Absorber Materials Irradiated properties of the absorber materials are discussed in Section

4. 2.1. 4 Irradiation induced chemical transmutations are produced in 8 40.

N3utron bombardment of B-10 atoms results in the production of lithium and helium. The percent of helium released is given by the expression in Section 4.2.1.4 Irradiation enchanced swelling characteristics of the absorber materials are given in Section 4.2.1.4. Accomodations for swelling of the absorbers nave bsen incorporated in the design of the control elements and include the following measures:

1. All B3 C pellets have chamfered edges to promote sliding of the pellets in the cladding due to di,fferential thermal expansion and irradiation enhanced swelli.;g.

2. Dimensionally stable Type 304 stainless stee: spacers are locatec at l the bottom of all absorber stacks adjacent to the nose cap to minimize strain at the weld joint.

3. A felt metal sleeve containing reduced diameter B,C pellets is located in the bottom length of the absorber stacks in fuT1 lengtn control elements. The felt metal sleeve laterally ; ositions the reduced diameter B C pellets ut.iformly witn respect to the clad and in addition absores 4

the differential thermal expansion and irradiation induced swelling of the B C pellets thereby limiting the amount of induced strain in the 4

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...- DDDBMF @@py D. Potential for and Consequences of CEA Functional Failure The probability for a functional failure of the CEA is considered to be very small. This conclusion is based on the conservatism used in the design, the quality control procedures used during manufacturing and on testing of similar full-size CEA/CEDM combinations under eimulated reactor conditions for lengths of travel and numbers of trips greater than that expected to occur during the design life. The consequences fo CEA/CEDM functional failure are discussed in Chapter 15.

A postulated CEA failure mode is cladding failure. In the event that an element is assumed to partially fill with water under low or zero power conditions, the possibility exists that upon returning to power, the path of the water to the outside could be blocked. The expansion of the entraoped water could cause the element to swell. In tests, specimens of CEA cladding were filled with a spacer representing the poison material. All but 9% of the remaining volume was filled with water. The sealed assembly was then subjected to a temperature of 650F and an external pressure of 2250 lb/in.2 followed by a rapid removal of the external pressure. The resulting diametral increases of the cladding were on the order of 15 to 25 mils and were not sufficient to impair axial motion of the CEA, which has a 0.084 diametral clearance with the fuel assembly guide tubes. This test result, coupled with the low prooability of a cladding failure leading to a waterlogged rod, comonstrates that the probability for a CEA functional failure from this cause is low.

Another possible consequence of failed cladding is the release of small quantities of CEA filler materials, and helium and lithium (from the neutron-boren reactions). However, the amounts whicn would be releasec are too ffhk, h f$$ '

4.2.4 ' TESTING AND INSPECTION PLAN Fuel bundle assembly and control element assembly quality assurance is attained by adherence to the procedures described in Chapter 17.

Vendor product certifications, process surveillance, inspections, tests, and material check analyses are performed to ensure conformity of all fuel assembly and control element assembly components to the design requirements from material procurement through receiving inspection at the plant site.

The following are basic quality assurance measures wnich are performed:

4.2.4.1 Fuel Assembly A comprehensive quality cu. trol plan is established to ensure that dimensional requirements of the drawings are met. In those cases where a large numcer of measurements are required and 100% inspection is impractical, these plans provide a high statistical confidence that these dimensions are within  ;

tolerance. Sensitivity and accuracy of all measuring devices are within j

'j 10% of the dimensioned tolerance. Sensitivity and accuracy of all measuring '

l devices are within +10% of the dimerisioned tolerance. Tne basic quality assurance measures which are performed in addition to dimensional inspections and material verifications are described in the following sections.

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