ML17252B199

Supplement F to Second Reload License Submittal

ML17252B199
Person / Time
Site:	Dresden
Issue date:	01/09/1974
From:	Commonwealth Edison Co
To:	US Atomic Energy Commission (AEC)
References
Download: ML17252B199 (46)

Text

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DRESDEN UNIT 3 - SUPPLEMENT F to SECOND RELOAD LICENSE SUBMITTAL

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,. QUESTIONS AND RESPONSES REF~RRING_ TO ATTACHM~NT A Dresden 3 Nuclear Power Station Second Reload License Submittal, September 1973 CONTENTS 6.2 6.3 6.4 6.5 6.6 6.7 6.8 6.9 6.10*

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QUESTION 6.2:

For the steam line break inside containment, provide curves of power flow, quality, MCHFR and pressure differentials in the highest powered 8 x 8 fuel assembly as a function of time.

RESPONSE

The highest powered bundle has been analyzed to determine the thermal-hydraulic response to a steam line break inside the containment.;

The models employed are those described in Appendices A and B of NED0-10329.

Nonnalized core power yersus time is presented on Page D-6 of NED0-10625.

Nonnalized core inlet flow, core inlet and exit qualities and MCHFR versus time are plotted on figures 6.2.1, 6.2.2 and 6.2.3 respectively. As shown on figure 6.2.3 the bundle will not pass into transition boiling and therefore will not experience heatup.

The total pressure drop through the core versus time is shown on figure 6.2.4.

The pressure drop across the channel wall_ using conservative pressure loading assumptions has been presented in the response to Question 3.69 .

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' I QUESTION 6. 3:

Provide curves c:if flow, quality at the channel inlet and outlet convective heat transfer coefficient and the clad temperature of several representative rods in the highest powered 7 x 7 fuel assembly as functions of time following the design basis loss-of-coolant accident.

RESPONSE

The requested information can be found in the Interim Acceptance Criteria submittal Amendment 27 (Item 3) and 28 of Quad Cities.

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QUESTION 6.4:

Provide curves similar to figure C-19 of NED0-10329 of flow and MCHFR in the hottest 8 x 8 fuel assembly versus time for both the cases of 1) no flow redistribution occurs and 2) flow redistribution is considered.

RESPONSE

The results of these analyses are given on the attached Figure 1. The assumptions used in the analysis are identical to those used in the analyses presented in NED0-10329. As can be seen from the figure, the hot channel flow is conservatively predicted for the majority of the transient when no flow redistribution is considered. Also, consideration of the effect of flow redistribution has a relatively insignificant effect on the minimum calculated MCHFR following the postulated DBA. Therefore, it can be con-cluded that flow redistribution effects are not an important consideration during a DBA.

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QUESTION 6.5:

Provide curves similar to figure C-21 of NED0-10329 which show the sensitivity of the MCHFR in an 8 x 8 assembly following the design basis loss-of-coolant accident to variations in initial flow.

RESPONSE

The results of the sensitivity study are given on the accompanying Figure l. The solid li.ne represents the history of the MCHFR during the transient assuming no initial flow distribution effects and is identical to that presented in Figure 6.14 of the second reload license submittal. The two dotted curves represent the history of the MCHFR assuming an initial 10% and 15% flow reduction in the hot channel flow.

It can be seen from the results that even with a 15% flow reduction the calculated MCHFR during the transient did not go below 1.0 until the flow window. It can therefore be concluded that the MCHFR calculated for an 8 x 8 assembly is not sensitive to variations in initial flow.

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QUESTION 6.6:

Provide the local rod-to-rod peaking factors for each rod, the axial and radial peaking factor and the peak LHGR in the highest powered 8 x 8 fuel in the ECCS analysis at 5,000, 15,000 and 30,000 MWD/T exposure.

RESPONSE

The peaking factors applicable to the Dresden 3 8 x.8 LOCA Reload analysis are as fo 11 ows:

1. Local rod-to-rod peaking factors - See attached Figures.

2. Gross peaking factor i.e. (axial peaking factor X radial peaking factor) = 2. 38

3. Peak LHGR = 13.4 KW/ft at a design local peaking of 1.22. (See

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• QUESTION 6. 7;*--

Al though the lo,wer stresses in a rod in an 8 x8 assembly during a LOCA may result in fewer perforations than in a 7 x 7 assembly, the extent of*

swelling and blockage may be greater. Provide an estimate of the degree of blockage in a 8 x 8 assembly following the design basis LOCA.

RESPONSE

In order to evaluate the effects of fuel rod swelling and perforations on the effectiveness of the emergency core cooling heat transfer, full scale bundle tests have been performed. It is General Electric's philosophy that the effects of fuel rod perforations on ECCS perfonnance are best evaluated by full scale tests rather than analytical investigations where several simplifying assumptions are necessary to couple the heat transfer and geometry effects. Such a test was performed for 7 x 7 fuel and the results are re-ported in GEAP 139112. In this test no affect of fuel rod swe 11 i ng were observed on ECCS effectiveness.

These conclusions also apply to an 8 x 8 lattice. Recently an. 8 ~ 8 full scale internally pressurized bundle was tested under simulated LOCA conditions typical of Dresden 3. The results of this test, which are reported in a topical report to be released in January 1974, sho\'1ed that no change in the

. heat transfer or the ability of the core heatup model to predict cladding temperatures was observed subsequent to ci adding swe 11 i ng and perforations.

Therefore, it can be concluded that the degree of blockage that would occur in the 8 x 8 assembly for Dresden 3 will not impair the effectiveness of the ECCS and hence will not alter the previously reported peak cladding temperature.

[)cmori~;Lrute that the 1*10Ler rod (nod No. 37) vJill not be du111<1gcd or distorted duP to depressurization during tile 1*1or'.>t liquid or stea111 line brcuk LOCI\.

RESPONSE

The pri111ary stresses in the 1*1ater tube are much less than the material yield strength for the worst liquid or steam line break LOCA, thus no significant damage 6r di~tortion would be expected .

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QUESTION 6.9:

-~ Pro vi de an analysis of the long-term cooling capability of the LPCI system which is applicable to a core with 8 x 8 fuel assemblies. Provide curves of swollen level and clad temperature versus bundle power for the 8 x 8 assemblies.

RESPONSE

Because of the higher heat fluxes associated with 7 x 7 fuel, the long-term cooling analysis presented in Quad Cities Amendment 26 is considered to be conservatively applicable to 8 x 8 fuel, i.e. the long-term cladding tempera-tures for an 8 x 8 core will not exceed those presented in Amendment 26 .

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• --......._ QUEST! ON fr. 10:

Provide the range of parameters (e.g. inlet flow and enthalpy, bundle

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power, peaking factors and flow coastdown and pressure decay rates) in the full scale 8 x 8 bundle transient or steady state CHF tests that can be used to "verfiy the applicability of the existing model" (Section 6.1.2.3).

Provide the schedule for submittal of the data and evaluation of these tests.

RESPONSE

The Hench-Levy Critical Heat Flux correlation was introduced in 1966, and its basis and formulation was presented to the AEC in Reference l~ The range of significant parameters used in the collection of data for the correlation are presented in Table 1. The test data, coupled with a multi-l~ channel hydraulic model to predict the local coolant behavior in more complex geometries than those tested, formed the critical heat flux design basis.

Since its initial introduction in 1966, the Hench-Levy Critical Heat Flux correlation has been accepted by the AEC as a licensing basis for the evaluation of thermal margin for a variety of boiling water reactor fuel geometries. A tabulation of principal geometric parameters for these lattice configurations appears in Table 2. Also shown in Table 2 are comparable parameters for the 8 x 8 assembly.

Use of the Hench-Levy correlation was continued for recent submittals covering 8 x 8 reload fuel, because the geometrical, thermal, and hydraulic parameters are similar to those of the fuel in plants previously licensed under this

thermal design basis, and well within the range of previous applications.

Reactor operating conditions for the 8 x 8 fuel in terms of pressure, mass flux, and quality do not differ from those previously experienced.

Furthermore, th.e use of the transient Critical Heat Flux Correlation given in Appendix C of NED0-10329 and AEC approved for use in evaluating the thermal hydraulic performance of"7 x 7 BWR fuel bundles following a postu-lated LOCA is similarily equally appropriate for use in evaluating 8 x 8 BWR fuel bundles. Since the geometric configuration of the 8 x 8 assembly has not changed significantly (See Table 2) applicability of the transient Critical Heat Flux model is considere.d appropriate for determining the consequences of the LOCA.

TABLE l HENCH-LEVY CORRELATION TEST PARAMETERS l Multirod Geqmetry Four- and Nine-Rod in Square Array Heated Length 36, 45, 48, 60 inches Hydraulic Diameter 0.324-0~485 inches Rod-to~Rod Spacing 0.060~0~187 inches Rod~to~Channel Spacing 0.060-0.135 inches Pressure 600-1450 psia i

0.2-l.6xlo6 lb/hr-ft 2

~j Flow Rate Steam Qua 1ity 0-0.6 Heating Distribution Uni form and Increased Heating in Corner Rod or Central Rod

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TABLE 2

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) GE-BWR FUEL ASSEMBLY PARAMETERS Geometry 6x6 7x7 9x9 llxll 8x8 Heated Length, inches 77.5-109 79-144 70 70 144 Hydraulic Diameter, inches .453-.567 .494-.576 .499 .497 *516 Rod-to-Rod Spacing inches . 134-. 177 .145-.175 . 145 . 128 . 147 Rod-to-Channel Spacing, inches .087-. 140 .135-.144 . 162 . 140 . 153

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REFERENCE

./ 1. Healzer, J. M., Hench, J. E., Janssen, E. and Levy, S. ,- "Design Basis for Critical Heat Flux Condi ti on in Bailing Water Reactors ; 11 July, 1966. APED-5286

QUESTIONS

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AND_RESPONSES REFERRING TO_ATIACHMENT B NED0-20103 General Design Infonnation for General Electric Boiling Water Rea-tor Reload Fuel Corrrnencing in Spring 1974 CONTENTS 3.1 I

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3.1 QUESTION Provide an assembly drawing of the fuel assembly and, if necessary for completeness and clarity, detail dra\'lings of components. The drawing should be similar to figure 6.2-2 in GESSAR, but the following additional information should be provided:

a) Dimensional tolerances.

b) Fuel pellet dimensions, including pellet length, edge chamfer and end dishing.

c) Filler gas pressure and composition including water vapor and other impurity content.

d) A description of the getter, including the volume, ~eight, surface area and alloy constituents.

• e) Water *rod dimensions, including diameter, wall thickness and number, location and size of vent holes .

RESPONSE

3.l(a) Attached is an assembly drawing (GE drawing number 814E954) providing fit-up dimensions and tolerances.

Proprietary information contained in separate submittal.

3. l(b) The fuel pellet dimen_sions and end chamfer dimen.sions are as follows:

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3.l{e) The water rod dimensions requested are noted below.

Water Rod O.D., inch 0.493 Water Rod I.D., inch 0.425 Water Rod Inlet and Outlet Flow Holes Inlet Number

• 3 LOCl.tion 1 inch from base of tube spaced 120° Diameter 0.089 inch

Question 3.1 (Continued) Outlet Number 8 Location 4 groups of two holes per group spaced 90° apart, at 2.00 inches, 1.66 inches, 1.33 inches and 1.00 inch from the top of the tube Diameter O. 188 inch

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° QUESTIONS AND ANSWERS REFERRING TQ ATTACHMENT C GESSAR QUESTIONS APPLIED TO DRESDEN 3 RELOAD 2

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QUESTION 3.69:

Provide detailed information concerning the effects of a steam lfne break on reactor internals. Include the effects on final assemblies and the possible interference to control rod insertions.

RESPONSE

~worst-case reactor internal pressure difference analyses has been per-fonned using the model described in Appendix A of NED0-10329. The conditions that the steam line* breaks are analyzed"are the turbine rated steam flow (Case I) and at 20% rated power (Case II). The 20% power case represents the minimum power attainable at maximum reci rcul ati on flow. This condition gives the maxi mum internal loadings because the difference between the energy l:

removal rate through the break and energy addition rate to the reactor vessel inventory increases with decreasing power level. Therefore, the depressurization rate following a steam line break increases with decreasing initial power level.

As the initial recirculation flow decreases *at a given power level, the initial steady state pressure differentials decrease as well as their maximum values during blowdown.

The maximum differential pressures across the reactor internals are presented in Table 3.69. l. The transient pressure loading histories during the blowdown are shown on figures 3.69.l and 3.69.2 for cases I and II respectively.

As discussed on Page 3.6-10 of the Dresden SAR, a pressure difference across the channel of as much as 25 psi could be applied to an 80 mil

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Therefore, no interference between the channel wall and the control

• blades will occur.

As discussed on Page 3.6-11 of the Dresden SAR, the control rods are capable of withstanding an external pressure up to 54 psi without col-lapsing. Therefore, the co~trol rod guide tubes ~ill not.coll~pse.

Loading the Dresden core with 8 x 8 bundles wi 11 not impose excessive stresses on core internals in the event of a worst case main steam line break.

TABLE 3.69.l MAXIMUM PRESSURE DIFFERENCES OCCURRING DURING THE ACCIDENT PRESSURE DIFFERENCES Reactor Component Initial Steady Maximum Occurring State Values (~si) During a Steam Case l Case 2 Line Break (Psi)

Case 1 Case 2 Core Plate and Guide Tube 17.7 15.5 17.7 18.0 Baffle Plate and Lower Shroud 22. 5' 22.0 25.5 36. l Upper Shroud and Shroud Head 5.0; 4. l 10.9 21. 9 Channel Box 7.4 6.5 7.7 8.4

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4. 16 QUESTION For the cyclic loadings in Section 4.2.l .l.2.8, specify the time durations of each cyclic condition. Explain how the cyclic loading would be reflected in typical technical specifications .

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. RESPONSE The answer to this question was recently provided in response to the same question on the GESSAR docket. The answer has been incorporated into the revised Section 4.2.1.3.2~3 of GESSAR, and is provided in the following paragraph.

The expected time duration for each of the subject cyclic loadings is not specified, and for the startup and reduced power cycles can vary according to the reactor status and power demand. The cyclic condi-tions relating to abnormal operational transients would result from a single operator error or equip-ment malfunction, and would therefore be expected

• to be of short duration (less than eight hours) .

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In addition, the time/pressure/power relationship for abnormal operational transients is identified in Section 15 of the FSAR .

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• 4.21

• QUESTION Is there any stiffness limitation on the spacer grid assembly and individual grid spring?

RESPONSE:.

For the portion of the question relating the spacer grid assembly stiffness, please refer to the answer to Question 4.34.

Stiffness of the spacer springs is based upon experience and past testing which have demonstrated that a positive spring force is necessary to assu~e that fuel rod vibration does not result in excessive clad fretting. *1 Proprietary information contained in separate submittal.

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These spring design cnaractenstics have been employed since the mid 1960's with very satisfactory results.

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4. 211 QQEST ION

\*lith respect fo fuel rod and iY,scmbly hchav"ior, discuss the dcsi'..)n lirnits for the accidents such ils LOCI\ <rncl seisrnic design.

RESPOi*!SE:

To tile extent that the Interim Acceptance Critcri a cun he construed as "clcs*ign limits," they apply to the fuel under LOCA conditions. Othernise, there arc no fuel mechanical design limits which apply specifically to LOCA conditions.

The most limiting fuel assembly comp6nent under seismic loading is the fuel channel. Sinte the channel design and the bundle v:eight are the same as for the initial core, the fuel seismic performance is unchange'd. The allovrnbl<::-

and predicted channel seismic loads are as shown below: .*

ALLOHAf3LE PREDICTED 13encling Moment (in-lb).

l/2 SSE 27,600 11'600 SSE . 42, 400 23,200 Shear (lb) l /2 SSE 2,890 203 SSE 4,660 406

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• 33 QUESTION Discuss how the different loading categories are combined to

• satisfy the design limit for each component of the fuel assembly.

RESPONSE

The fuel rod applied stresses are combined using the ASME Boiler and Pressure Vessel Code, Section Ill as a guide as discussed in Section 3.2.3 of 11 Dresden 3 Nuclear Power Station, Second Reload License Submittal, 11 S~ptember 1973, previously provided to the AEC. For the other fuel assembly components, the*stresses are combined using the theory of constant elastic strain energy of distortion, or where component loading tests are performed, loading and/or deformation limits are derived directly from the test results .

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...,/ ' *. Evalutitc the behavior of tile Spc1e<~r urid in hiuh tc111pcrature und oscilli1tin~ fluid environments. Shol'/ that dur1ng fuel rod axitil expansion und contruction no b*inding is c_xpr.ctcd due to l>cndin9 of fuel rod an*d usynirnetry of support in un 8x8 design. Ev0luate the stresses in the spticer grid lock.

Show that the spacer is adequtitc to support ltiteral loading during LOCA und seismic latercJl loadings.

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RESPONSE

Mechanical design of the Bx8 spticer is essentially identical to that of the 7x7 spacer with the exception that it contains

• additional internal members. Based upon the excellent performance of the 7x7 spacers in the Dresden 2, Dresden 3, and other operating .BHR 1 s, it is expected that the 8x8 spacer will also perform satisfactorily in th~ BWR environment. "7 As noted in the response to (Juestion 4.21, the GE spacer design uses relatively soft springs to position the fuel rods, 1vhich assures tl1at the fuel rods arc free to expand and contract axially. Therefore, no significant binc.ling of the fuel rods is expected.

As noted in the response to Question 11.2'1, the mos . limiting fuel assembly component under seismic louding conditions is the fuel channel. The functional requirements of the fuel spticer under seismic lotidings arc that it must transmit the acceleration lotidin9s fro111 the contained fuel rods to the fuel cl1annel v1hi le maintaining tl1c llppropriotr. furl rod positioning. As discussl'd Jliove, the 8x f spJccr i~. cs'.;entii!lly idcntic,11 to the 7x7 spJcc*r 1*1it.l1 the c>:cPptio11 that it hJs been str<'n~tl1'~11cd by ;1d<li1.io11:1l i11t1.:r11t1l '.;Lr11ctu1-.1l 111<:11ilH:rs and 1*JL~lds. Tc~,ti11Q uf lite 1:<7 sp~ic:cr lli1'.. VL'rificd tl1<1t it'.'.

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RESPONSE TO 4.34 Continued seismic loading capability exceeds that of the fuel channel ..

The seismic capability of the stronger 8x8 spacer should also .

exceed that of the fuel channel. ,,

In the full-scale internally pressurized 8x8 Zircaloy ECCS tests of October 1973, significant bo\'ling of fuel rods at elevated temperature was expected and observed. This bowing is the significant lateral loading on the spacers which will occur as a result of a LOCA. No significant deformation of spacers was observed, and no mechanical failures of spacers occurred.

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. 4. 36 QUESTION Evaluate the axial and radial thermal cycling problem of a fuel channel. Discuss the pre-operation and normal operating loads on the channel, including stresses due to control rod interference and thermal differential expansion stre~s between stainless steel tie plate and Zr-4 channel.

RESPONSE

As discussed in the response to Question 4.26, the fuel channel is rigidly attached to the fuel bundle only at the upper end where it is bolted to the upper tie plate~ Axial differential thermal expansion between the bundle and the channel is accommo-dated by slippage at the fuel rod spacers and lower tie plate.

Under situations of adverse tolerance stack up, differential thermal expansion between the stainless steel tie plates and the Zircaloy channel can result in an interference fit; however, the resultant secondary membrane stress in the channel does not exceed the material yield strength. The loads and resultant stresses imposed on the fuel channel in the event of control rod interference are quite low in that the channel is relatively flexible and will deflect to accommodate the interference.

Approximately one pound force is applied to the fuel channel for each mil of control rod interference.

The mechanical fitup between the channel, fuel bundle, and control rod for this design is basically the same as that which has been used for essentially all GE fuel designs dating back to 1960.

Operating experience with this design has been completely satis-factory in all cases.

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4.46 QUESTION Quality assurance and inspection procedures used in the manufacture and loading of gadolinia-poisoned rods should be.

presented. Specifically, methods used to verify the gadolinia concentration in the uo 2-Gd 2o3 blend, the sintered pellet uo 2-Gd 2o3 .

solid solution homogeneity, the gadolinia-urania pellet identification, and the gadolinia-urania fuel rod identification should be presented.

RESPONSE

The same rigid quality control requirements observed for standard uo2 fuel are employed in the manufacture of gadolinia-urania fuel and pro~id~ verification of the powder, pellet and fuel rod characteristics.

The fuel fabrication process parameters and procedure~ are and have been made available for audit by customers and the* AEC.

The amount of Gd 2o3 in fuel pellets is verified by t~sting pellet samples using X-ray spectru111etry. The presence or absence of burnable poison in the assembled fuel rods is also verified by gamma scanning.

Thr..: capability also exists to detect Gd o variations in the assembled 23 fuel rods using a neutron irradiated gamma scan procedure.

Each separate pellet group is characterized by a single stamp. The gadolinia pellets are specially stamped to differentiate them from uranium pellets. Fuel rods are individually numbered prior to fuel loading: (l) to identify which pellet group is to be loaded in each fuel rod, (2) to identify which position in the fuel assembly each fuel rod is td be loaded, and (3) to facilitate total fuel material accountability for a given project. Further identification of individual fuel rod gadolinia concentrations and uranium enrichments is accomplished by geometric djfferences in the upper end plug shank for each differing rod. Each upper end plug is ensured proper place-ment within a fuel rod by reference to the lower end plug identification i1ililiw:li'll*t'illi*...__ _ _ _ _ ._._....Kl.- - - ,.,-- -- .* -

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4.4b Question *(Continued) number. Each fuel rod is ensured of proper plJcement within a fuel bundle by reference to visual geometric differences in the ~pper

~nd plugs.

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