ML19316A084

Forwards Request for Addl Info Re Fsar.Some Questions Concern BAW-10008,Part 2

ML19316A084
Person / Time
Site:	Oconee
Issue date:	03/03/1970
From:	Morris P US ATOMIC ENERGY COMMISSION (AEC)
To:	Thies A DUKE POWER CO.
References
700308, NUDOCS 7911210809
Download: ML19316A084 (10)

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March 3, 1970 DR Reading i

j DRL Reading RPB-3 Reading C. K. Beck M. M. Mann i

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ADDITIONAL INF0PM TION REQUIRED 3.8 Reactor Internals (The following questions apply to B&W Report BAW-10008, Part 1) 3.8.1 Briefly describe the manner by which Figure 10 of the report

" Shear Force on Core for 36-Inch and 28-Inch Rupture," is derived f rom the pressure differential transients.

3.8.2 The report states that all components will be designed to ensure

~ against structural instabilities, regardless of stress level. We note that the core support shield and core barrel shells were analyzed for un-stable collapse due to external pressure. Were the control rod guide tubes analyzed for column buckling ef fects due to combined LOCA and seismic loadings?

Identify any other components of the reactor internals for which buckling is a possible mode of f ailure under any of the design loading combinations. Provide the bases for using static loads in lieu of the dy-namic response loads.

i 3.8.3 Provide the bases for the dynamic analyses and the associated dynamic load f actors which are used in the stress and deflection analysis for horizontal and vertical excitation input, _ including bell mode responses.

Give typical examples of such factors and their effect on the results.

3.8.4 The report states that seismic loads were determined from the response spectra for the design basis and maximum hypothetical earthquakes specified for the Rancho Seco Station site.

Discuss how the seismic loads were determined from the response spectra. Give sufficient detail to show the development of the seismic loadings from the ground motion 1-ats for the containment structure to the final input used for the an.mysis of the internals structural members.

In addition, describe in detail all dynamic analysis methods used in deter =ining stresses and deflections for reactor internals under seismic loadings.

Include in the discussion the following:

(a) A detailed description of all =athematical models of the system including a discussion of the degrees of freedom and methods of lumping masses, determining section properties, etc.

(b)

A discussion of the analytical methods used including, where applicable, the metheds of computing periods, mode shapes, and modal p.articipation factors.

(c) A listing of and the bases for any damping values that were used.

(d)

A list of points at which there are changes in stress analysis methods, e.g., dynamic to static, and the bases for such changes.

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Indicate the =odal responses that were codbined, e.g.,

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deflection, acceleration, or stresses, and the procedure for combining these responses.

3.S.5 The discussion of the =ultimass model, Figure 22, refers to a

ore detailed multimass model.

Describe the more detailed multi = ass model and discuss the basis upon which the results frc= this =cdel deter-

=ined the adequacy of the model used in Figure 22.

3.8.6 It is stated that the plenum cylinder and reinforce =ent plate i

were treated as a flat plate with a unifor= pressure load in the calcula-tion of stress and deflection. Describe the configuration and similitude of the =odel and the plenum cha=ber and reinforcement plate, including the boundary conditions assumed, e.g.,

edge fixity.

3.8.7 As discussed in the January =eeting, the combined stress, Pb + ?= for the control rod guide tube,Teported"in Section 3.2.2.3 of the report, should be clarified.

3.3.8 In referance to the stress su==ary of Table 1 of the~reporh providc the following information:

- x (a) Examples of how LOCA and seismic stresses were co=bined to give conservative results for these concurrent loading conditions.

l (b) A separate su==ary of stress intensities due to the naxi-

=um hypothetical earthquakc and the applicable allowable stress intensities.

3.8.9 For loading combination case IV in Appendix A, provide a com-parison on an elastic basis between the stated stress limits and a medbrane unifor= strain for the materials associated with this loading combina: ion.

3.8.10 Equations (5) ~an'd (7) of ~ Appendix A_should be corrected as 4

discussed in the January =eeting.

3.8.11 Appendix C indicates that the case IV loading combination stress 11=1 utilizes ulti= ate strength curves published by U.S. Steel which are nor=alized at roc = te=perature to minimum ultimate strength values given by Table N-421 of Section III.

These U.S. Steel ultimate strength curves cannot be considered as conservative unless the lower bound value of the ulti= ate strength of each =aterial at an appropriate design te=perature has been established.

Indicate how this concern will be resolved.

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-E 3.8.12 A=plify the discussion of Appendix D of the report concern #ng the stress limits and S values chosen for load combination caser II, III, and IV.

This disc 5ssion should consider:

'N (a)

The bases upon which S, values and stress limits were,~

l selcceed, since code limits are not specified.

(b)

The ef fect upcn bolts of preload, pressure, and differen-tial thermal expansion on the stress limits specified, for cases II, III, and IV.

3.9 Fuel Assembly Structural Design (The following questions apply to BAW-10008, Part 2) 3.9.1 Section 2.4 of the report does not sufficiently define the stress and strain limits for the design basis earthquake (DBE) and simultaneous maximum hypothetical earthquake (MHE) and loss-of-coolant accident (LOCA) nor the manner and extent to which the cited limits provide an assured margin against failure for these loadings. Our specific concerns are:

3. 9.1.1 DBE Criteria (a)

Confirm that the type of stresses referred to in paragraph 1 are in the primary

- as defined in Article 4 of ASME Code,'Sec-tion III.

Describe oasis for establishing 75% of the stress rupture l

life of the material as a numerical limit and whether that limit is con-structed upon the average stress or the minimum stress to produce rupture at the end of 105 hours0.00122 days <br />0.0292 hours <br />1.736111e-4 weeks <br />3.99525e-5 months <br />.

(b)

Clarify whether stresses of the type referred to in paragraph 2 are in the secondary category in the same context as above.

i Where stresses exceed yield, are they calculated on an equivalent elastic basis, i.e., pseudo-elas tic basis as in Section III?

Identify the source of the f atigue curves used for each material of concern (e.g., Article 4,Section III). Where fatigue data are e= ployed which are not included in any codes or standards, specify whether a basic data curve is used or a design curve which incorporates design / correction f actors and correction for maximum effect of mean stress.

Provide the bases for the statement that strain limits will be set using no more than 90% of the material's fatigue life.

Specify the number and type of cycles that have been estab-lished for design purposes and indicate the margin of safety that exists over,the-expected number _and type-of op~ ration cycles to be experienced.

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For the combination of stresses in (a) and (b) above, l

specit'y the stress limits that apply (e.g., 3 S or S ).

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3.9.1.2 Combined LOCA and MHE

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' (a)

Clarify whether 'the applied stress referred to in paragraphs

-I and 2 is a primary stress, exclusively.

Provide the basis for establish-ing 85% of ulti= ate strength of the material as a numerical stress limit.

l is the ultimate strength normalized to the minimum tensile strength of the material as specified-in the appropriate ASME or ASTM material specification?

Is this stress calculated on an elastic basis?

Provide the elastic stresses corresponding to this limit for each of the materials of concern. Furnish the corresponding strain limits of each material.

(b)

Identify the components referred to in paragraph 2 that contribute to the stability of the control rod guide tubes.

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(c)

Provide the basis for the allowance of 85% of the critical buckling load as a limit.

Identify the theoretical column formulce used (i.e., Euler or o ther).

I 3.9.2 Relate quantitatively Figure 3 of this part of the report to the j

figures of Part 1.

I 3.9.3 3riefly describe the analytical techniques that the FLASH com-puter code utilizes and its capabilities in relation to its employment on this problem.

3.9.~

The model used to describe the dyna =ic behavior of the re actor i

vessel and internals is not described in sufficiant detail to perm t an i

assessment of the accuracy by which the vessel and internals have ceen analytically described.

Provide:

(a)

Engineering drawings and/or sketches of the structural I

features of importance.

(b) JL precise description of the location of and basis for couputation of masses and section properties / boundary conditions.

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(c)

Details on the manner'in which flexibility coeffielents 4

have been co=puted and the results achieved.

3.9.5 The design loadings and their =anner of application to the structure require more precise description. Provide:

(a)

The complete digitalized acceleration record that was used in the analysis.

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3.9.1G In reference to the spacer grid compression tests described in Section 5.3, provide a sketch showing the test spec 1=en, its orientation in the loading fixture, and the direction of loading.

Explain how correc-

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tions were made for te=perature ef fects. Provide elaboration on the load cycling pheno =enon noted in paragraph 2 and show graphically how this occurs.

1 3.9.11 Horizontal contact analysis results are given in Section 6.1 in terms of =argins of safety calculated on the basis of allowable and Provide the maximum stresses that were calculated from applied loads.

the applied loads for -the applicable-components in both Sections 6.1 and

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Specify how LOCA and seismic stresses are combined.

3.9.12 Section 6.2.1.2 shows the margin of safety for guide tube g q

buckling under LOCA loadings only.

Indicate the margin of safety.for i

co=bined LOCA and seis=le loads.

Confirm that seismic loads are ir}cluded in the reported results of Section 6.2, vertical contact analysis.

3.9.13 Provide a detailed explanation for the conclusion in Section 6.2.2.1 that loads due to LOCA and/or earthquake are not additive to those due to no=al operation because the maximum loads are limited by the i

available friction loads between the end grids and the fuel rods.

3.10 Control Rod Drive System

(

3.10.1 Identify in the FSAR or in B&W Report BAW-10007 the design codes j

which are applicable and applied-to-the_. rod _ drive system. For non-code ite=s indicate the stress, defor=ation and faii~gue limits used. Discuss f

the analytical approaches taken in a format which will include the above items and which will demonstrate the =argins of safety provided under l

nor=al operating conditions and hypothetical accident conditions.

i 3.10.2 Provide descriptive infomation and a discussion of the function i

of the springs which release the roller nuts.

Include infomation on spring =aterial and =aterial specification, fabrication techniques, and design stresses.

3.10.3 We understand that, in addition to the motor torque tests referenced in BAW-10007, tests have been performed to assess the ability of the control rod drive mechanis= to drive-in a stuck rod.

Describe these tests. and provide the results.

l 3.10.4 All tests reported in BAW-10007 have been perfor=ed on a 2

p ro to type unit.

Indicate any significant differences in design, materials, tolerances, and f abrication techniques between the prototype units and the

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production units, and discuss their i=portance in determining the need to repeat the basic tests with production units. Discuss the tes't program conte = plated for the production units and the acceptance criteria to be I

applied.

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i (b) Discuss the stress limits applicable to the simultaneous f

LOCA and seismic loads and the basis therefor.

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(c) A general description of the manner of digital-to-analog i

conversions of data, an estimate of the accuracy of the process and a l

description by which the acceleration was inserted into the electronic differential analyzer.

(d) A complete acceleration response spectrum comparison at i

1 and 10 percent critical damping.

(e)

The manner in which the vertical seismic component has been factored into the analysis and the importance of the stresses and deflections therefrom with respect to the horizontal seismic and LOCA loadings.

f 3.9.6 The manner in which analog computations have been perfor=ed is not presented. Provide a detailed description of the manner in which these computations have been performed.

In addition, provide strip chart recorder output results for severa~ typical runs and a tabulation of significant stress, strain and deflection results at critical locations for these same runs.

3.9.7 Provide a sketch of the second model segment (as discussed in l

Section 4.1.4 of the report) and discuss its interaction with the first

=odel segment.

i 3.9.8 In reference to Figures 7 and 8 of the report which show the l

mathematical model for the vertical contact analysis and its load-

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deflection curve, specify the spring constant variation for the. fuel assembly in relation to its location within the core for that part of q

the load-deflection curve which occurs af ter the gap is closed.

i 3.9.9 Section 5.1 of the report discusses the frequency and da= ping i

tests performed for full-size and subsized specimens.

Further detailed information is required to complete our review.

Provide discussion of the following:

j (a)

The basis for test amplitudes and frequencies used.

1 (b)

A description of and bases for tht type of loadings used, i

including test fixtures employed.

(c) A detailed description of the full-size and subsized specimens used including the. identification of specimen materials.

s (d). Description of test data obtained.

(e)

Interpretation and analysis of results, i

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3.10.5 Discuss the tests and/or analyses that have been employed to assess the damage which would result f rom operator errors or minor mal-j functions, such as over-driving a limit switch.

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3.10.6 Provide a lis t of the metals, lub ricants, insulation materials,

etc, which were tested in the prototype unit and discuss their long-term j

reliability in the reactor enviro...nent.

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11.8 We understand that you intend to rely on the RIA-36 reactor coolant letdown radiation monitors for detection of prompt fuel f ailures.

Describe the sensitivity and response time of these monitors.

Indicate the smallest number of failed fuci elements that the monitors can detect as well as the highest activity they can withstand without loss of func-tion.

Discuss the effects of crud buildup and provisions for decontamina-tion of the section of letdown line being monitored.

14.3 Steam-Line-Rupture Accident 14.3.1 We understand that the main turbine stop valves serve to isolate the unaf fected steam generator in the event of a steam-line-rupture acci-dent.

Describe the design, operation, and inspection of the main turbine stcp valves.

Discuss the capability of a turbine stop valve to close against reversed critical flow.

14.3.2 Describe the extent that the system which trips the curbine stop valves by a reactor trip signal meets IEEE-279.

14.3.3 We understand that in your analysis of the steam-line-rupture accident you have assumed that portions of the Integrated Control System (ICS) function (e.g. closing the main turbine stop valves, and the feed-water valves).

For those portions of the ICS which you have assumed to function properly, either provide an evaluation for our review to show that the system design conforms to IEEE-279 Criteria or analyze the steam-line-rupture accident at 100% power with an end-of-life moderator coefficient, minimum shutdewn margin and a stuck rod condition, assuming that the ICS and the operator fail to function or function in an adverse manner.

14.3.4 Describe the hybrid analog-digital computer program used for analysis of the steam-line rupture including physical models, equations, assumptions, numerical approximations, and input parameters.

i 14.3.5 For the analysis of the worst case steam-line-rupture accident provide the following (a) All input quantities including fluid inventories, time delays and constants, instrumentation time delays, negative reactivity insertions, flow rates, and heat transfer coefficients.

Justify each and cxplain why each is a conservative assu=ption.

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systen that detects and alarms c hig,h cnd low o.rassurizar levcis and meets the criteria of IEEE-279.

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