ML20079G006

Responds to NRC 810226,0323 & 0409 Requests for Addl Info. Responses to Questions CS.210.1,3 & 7; 250.4,5,6,7,8,& 9; 491.5; 410.18 & 230.3 Will Be Incorporated Into PSAR Amend 69 Scheduled for Submittal Later in June

ML20079G006
Person / Time
Site:	Clinch River
Issue date:	06/02/1982
From:	Longenecker J ENERGY, DEPT. OF
To:	Check P Office of Nuclear Reactor Regulation
References
HQ:5:82:040, HQ:5:82:40, NUDOCS 8206080292
Download: ML20079G006 (18)

Text

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Department of Energy Washington, D.C. 20545 Docket No. 50-537 HQ:S:82:040 JUN 0 21982 Mr. Paul S. Check, Director CRBRP Program Office Office of Nuclear Reactor Regulation

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

Dear Mr. Check:

RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION

Reference:

Letter, P. S. Check to J. R. Longenecker, "CRBRP Request for Additional Information," dated February 26, March 23, and April 9, 1981 This letter formally responds to your request for additional information contained in the reference letter.

Enclosed are responses to Questions CS 210.1, 3, and 7; 250.4, 5,-6, 7, 8, and 9; 491.5; 410.18; and 230.3; which will also be incorporated into the PSAR Amendment 69; scheduled for sut.nittal'later in June.

Sincerely, Jo n R. Longenec ,r,. Manager Li ensing & Envi 'nmental Coordination Office of Nuclear Energy Enclosures cc: Service List Standard Distribution Licensing Distribution Q0 N

8206080292 820601

{DRADOCK 05000537 PDR

Page - 1 82-0287 [8,22]

Ouestion CS410.18 (9.7.3) 1 Seismic Category I piping.he normal chilled water systs is a non-safety relate with some and equignent designed which to Seismic Category is located I criteria. in cells containing ng is sodium or NaK Resoonse With the exception of the SGB loop cells, all normal e

chill d equipnent are located outside the cells containing water piping and sodium or NaK pi i p ng.

to Seismic Category III.For the SGB loop cells the normal chilled water piping - e i evaluated to determine if Seismic Category IIIp cells adequate. pipingis in the SGB loo amendment. We results of this evaluation will be provided in a future l

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QCS410.18-1 Amend. 69 -

May 1982

82-UJ43_.

Pag 2 - 8 [82-0343] #70 i

~0uestion CS210.1 In piping systems at elevated teroperatures, local deformation may occur at creas of gemetric discontinuity, such as at fittings. Provide methods and procedures for the following:

A. Define elastic follow-up.

B. . Evaluate creep rupture and fatigue damage.

C. ' Justify the use of simplified creep ratcheting bounding techniques used in cmputer codes.

Eggponse This question was answered orally at the May 6-7, 1982 High Tenperature Design Meeting. The documentation requested will be provided in Reference QCS210.1 by June'17, 1982. .

References QCS 210.1-1 ES-LPD-82-009, Engineering Study Report on CRBRP Special Stress and Criteria Considerations, June 1982.

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j- 7 QCS210.1-1 Amend. 69

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cz-ua43

. Page - 9 [82-0343] #70 Question N10.3 h ermal expansion and creep rate generally vary among different materials.

l Describe methods and procedures used to evaluate local stressen and strains at places where Bi-metallic and Tri-metallic transition welds are applied.

Response

his question was answered orally at the May 6-7, 1982 High Tenperature Design Meeting. We documentation requested will be provided in Reference CS210.3-1 by June 17. 1982.

Reference CS210.3-1 CRBRP Engineering Study Report on CRBRP Transition Joints, ES-LPD-82-007, June 1982.

QCS210.3-1 Amend. 69 May 1982

82-0345

. Page - 10 [82-0343] #70 l

l ouestion cr>10.7 Due to the constant evolution of rules in Code Case 1592 (C-47) daring the period when the PSAR was prepared. Identify any areas where the rules delineated in current Appendix T of Code Case N-47 have not been satisfied.

Provide the basis to show that such deviations, if any, are acceptable in.

terms of design margins to ensure safety.

Response

his question was answered orally at the May 6-7, 1982 High Tenperature Design Meeting. %e documentation requested will be provided in Reference QCS210.1-1 by June 17, 1982.

QCs210.7-1 Amend. 69 May 1982

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Pag 2 - 2 [82-0343] #70 Ouestion rR?50.4 Provide the method for structural evaluation of weldments and associated materials for service at elevated temperature. %e following should be addressed:

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( (i) room tarperature and elevated temperature material acceptance criteria:

1 (ii) creep rupture damage resulting from residual forming and welding stresses:

(iii) mass transfer effects:

(iv) metallurgical notch effects:

(v) fracture tcughness criteria:

(vi) thermal aging effects: and (vii) irradiation effects.

Response

his questicn was answered orally at the May 6-7, 1982 High Tenperature Design meeting. We documentation requested will be provided in Reference QCS250.9-1 by June 17, 1982.

QCS250.4-1 Amend. 69 May 1982 i

. t!z-0343 P;ga - 3 [82-0343] #70 Ouestion N 50.5 Provide the method and data base for the structural evaluation and acceptance of Bi-metallic and Tri-metallic transition welds for service in the primary and intermediate heat transport systems.

f Response I eJ h is question was answered orally at the May 6-7, 1982 High Tenperature Design meeting. %e documentation requested will be provided in Reference CS210.3-1 cutznitted in response to Question CS210.3 by June 17, 1982.

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QCS250.5-1 Amend. 69 May 1982

U2-U343 Page - 4 [82-0343] #70 Ouestion CS250.6 Provide justification for the use of the simplified creep ratcheting bounding methods in Code Case 1592 at structural discontinuities.

Response

%ere is no generic justification for the use of the simplified creep rctchetting bounding methods in Code case 1592 (N-47) at structural discontinuities for final design assessment. % e Code Case does not allow the use of such methods at structural discontinuities. If the bounding method is used, it must be justified for the specific application.

OCS250.6-1 Amend. 69 May 1982

t!2-0343 Paga - 5 [82-0343] #70 Ouestion rR950.7 How do you account for the elastic follow-up in elevated temperature ccuponent and piping (elbows) system analyses?

Response

his question was answered orally at the May 6-7, 1982 High Tenperature Design meeting. Se dcv,=ntation requested will be provided in Reference QCS210.1-1 by June 17, 1982.

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t QCS250.7-1 Amend. 69

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. Page - 6 [82-0343] #70 Ouestion CS250.8 Provide the design criteria for the elevated ternperature core support structure, including the welds in the forging and the reactor vessel.

Response

'Ihe response to this question is provided in Amended Sections 4.2.2.1.1.1, 4.2.2.3.1.1 and 4.2.2.4.1.1. Additional information will be provided in Reference QCS210.1-1 by June 17, 1982.

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QCS250.8-1 Amend. 69

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Paga 1 (82-0344) [8,22] 471 Numerous conbinations exist for the implementation of these final three fallback positions. In addition, the extent to which they are needed for verification purposes cannot be accurately assessed until final design and the ongoing and planned testing is completed. Plans to utilize these fallbacks for verification or to increase the ultimate operational limits for the reactor will be prepared on a schedule consistent with that for preparation of the FSAR.

4.2.2 Paactor vaamel Internala 4.2.2.1 Desian Races 4.2.2.1.1 Functional Requirements Wo major assablies form the reactor intiernals support structure. %ese are the multiple component lower internals structure (LIS) and the multiple caponent upper internals structure (UIS) . An additional major system is the core restraint system (CRS), which consists of the core formers and the removable radial shielding.

W e lower internals structure consists of the core support structure, lower inlet modules, bypass flow modules, fixed radial shielding, fuel transfer and storage assably, and horizontal baffle. Support for the core formers of the core restraint system is provided by the lwer internals structure.

l  % e following are primary functional requirements for these components. Also j included are the maintainability and surveillance requirements.

4.2 4.1.1.1 Core *rmrt Struchwe he core support structure consists of the core plate, the core barrel, and the module liners. %e core p ate and support cone form the upper surface of the inlet plenum, provide the structure for the upper surface of the inlet plenum and provide the structure for the support of the core, %e core barrel encloses the core and provides a base for the core former rings. S e core cupport structure is welded to the support cone which is an integral part of the reactor vessel. %e Core Support Structure to Core Support Cone and Core Support Cone to Reactor Vessel Welds are both in an ares where the temperature is below 8000F and therefore the high temperature ASME Code Criteria is not cpplicable. W e welds made to the Core Support Cone were analyzed utilizing the ASME Ccde Criteria and Material Allowables. As a permanent structure, it must be designed for the " stretch" conditions

• and have a design life of 30 years. Se core support structure has the following principal functional requirements:

a) Provide support for the weight of the core assemblies, lower inlet modules, bypass flow modules, fuel transfer and storage assenbly, horizontal baffle, fixed radial shielding, rencvable radial shielding, spacers and core formers.

• See Section 1.1 4.2-118 Amend. 69 May 1982

U2-0.144 Paga 2 (82-0344) [8,22] #71 4.2.2.3.1 Lower Internain Structures (LTR)

The LIS c mponents and Core Former Structure (CFS) are evaluated as nuclear cmponents in accordance with the rules of: h e ASME Boiler and Pressure Vessel Code,Section III.

Where these rules cannot be applied, the following rules are invoked:

a. Code Case 1592, Class I Cmponents in Elevated Teperature Service,Section III.

b. RDP F 9-4 Components at Elevated T eperature (Supplement to ASME Code Cases 1592,1593,1594,1595, and 1596) .

c. RDF F 9-5, Guidelines and Procedures for Design of Nuclear Syctems Cmponents at Elevated 'Asaperatures (Non-mandatory) .

d.  % e special purpose strain controlled high-cycle fatigue criterion discussed in Section 4.2.2.3.2.3 may be applied to 304 and 316 austenitic stainless steel at temperatures up to 11000F.

Material properties not given in the Code are taken from the Nuclear Systems Materials Hancbook, TID-26666, and section 4.2.2.3.3.1 below.

4.2.2.3.1.1 Core Sucoort Structure (CSS) he CSS ' ace nn:1y cd using the following additional rules:

a. We 1974 Edition of the Code, Subsection NB and selected portions of Subsection NG with Addenda through Sumer 1975.

b. RDT Standard E 15-2NB, November 1974, (Supp1 ment to ASME Code Section III, Subsections NA and NB) .

c. Modification to the high t m perature design rules for Austenitic Stainless Steel creep fatigue evaluation per paragraph 4.2.2.3.2.3.

d. In the use of Code Case 1592, the effect of the sodium and radiation awavmua. on material properties was evaluated as defined in 4.2.2.3.3.2.

e. Because of radiographic examination limitations for the 20 inch thick weld between the core plate forging and the ring forging, progressive liquid penetrant examination was performed per paragraph NG-5231 as specified for Type I welds in Table NG-3352-1.

4.2.2.3.1.2 Lower Inlet Modale (LIM) , Bvnana Flow Madnle (BPFM) , and Core Former Structure (CFS) he 1974 Edition of the Code with Addenda through Winter 1976 were used for the LIM, BPEN, and CFS analyses.

4.2-176 Amend. 69 May 1982

t$2-UJ44 Pagm 3 (82-0344) [8,22] #71 4.2.2.3.1.3 Horizontal Baffle (HB), Fuel Transfer & Storage Asssbly (FT&SA) .

and Fixed Radial Shield (FRS)

'Ihe HB, ET&SA, and FRS are internals structures and are not covered by mandatory Code rules, but the owner's designee has required that the rules stated in 4.2.2.3.1 be applied to the design and analysis of these 1

4.2-176a Amend. 69 May 1982

82-0344 F gs 4 (82-0344) [8,22] 071 4.2.2.4.1.1 Analysis of Core Saccort Structure Umi Structural Evaluation Five sections in the CSS axisymetric model shown in Figure 4.2-51 and 10 sections of the core support plate sector model shown in Figure 4.2-52 were selected for the structural evaluations. nese sections represent the high stress areas in the CSS structure, and their selection was based on a thorough review of the finite element stresses for pressure, dead weight, seismic, and the thermal transients.

%e primary plus secondary stress intensity limit of 1.5 Sm was reduced to 1.35 Sm in the perforated region of the core support structure. his reduction was made to account for the actual bending shape factor for the geometry of the ligament. %e 1.5 shape factor is only applicable for. a rectangular cross section.

% e primary, primary plus secondary, and fatigue evaluation results are sumarized in Table 4.2-23 hese data are limited to the maximum of each stress intensity category for the locations identified in the table.

Simplified inelastic analysis techniques were utilized to show that the areas with primary plus secondary stress intensity values exceeding the 3 Sm allowable limit are acceptable. A fatigue factor of 0.9 was used in evaluating the fatigue damage in the area of the full penetration weld, Section C-C, Figure 4.2-51. his factor is imposed to meet the requirements in paragraph 4.2.2.3.1.1 for the method of examination enployed.

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4 4.2-197 Amend. 69 May 1982

82-0343 Page - 7 (82-0343] #70 Ouestion CS250.9 Provide the elevated temperature data base for the mechanical properties of 1 the following materials:

(i) Type 304 stainless steel; (ii) Type 316 stainless steel; (iii) 21/4 Cr - 1 Mo alloy; (iv) Alloy 718.

Resoonse h e elevated temperature data base for mechanical properties of types 304 and 316 stainless steel, alloy 718 and 2 1/4 Cr-1 Mo alloy are given in:

(i) Type 304 Stainless Steel (a) ASME Code,Section III and Code Case N47.

(b) NSM Handbook, Volume I, Group I, Section 2.0.

(c) Reference QCS250.9-1 (available June 17,1982)

(ii) Type 316 Stainless Steel (a) ASME Code,Section III and Code Case N47.

(b) NSM Handbook, Volume I, Group I, Section 4.0 (annealed) and Section 5.0 (20% cold worked) .

(c) Reference QCS250.9-1 (available June 17, 1982)

(iii) 21/4 Cr-1 Mo Alloy (a) ASME Code,Section III and Code Case N47.

(b) NSM Han& ook, Volume I, Group 2, Section 2.0.

(c) Reference QCS250.9-1 (available June 17,1982)

(iv) Alloy 718 (a) Code Case N47 (for bolting only).

(b) NSM Handbook, Volume I, Group 4, Section 5.0.

'c} Reference QCS250.9-1 (available June 17,1982) he ASME Code and N47 Code Case are the primary sources of materials properties data with the NSM Handbook and Reference QCS250.9-1 providing the data that are not covered in the primary documents.

W e data correlations presented in NSM Handbook Volume I are documented in NSM Han & ook Volume 2.

References QCS250.9-1 ES-LPD-82-008, Engineering Study Report on CRBRP Materials De a Base, June 1982.

QCS250.9-2 Amend. 69 May 1982

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Pag 2 4 (82-0315) [8,22] #57 Ouestion CS491.5 c) In one case you claim one dollar of reactivity is inserted within .3 seconds after rod motion begins. Why doesn't this sgree with the presentation in Figure 4.2-122?

b) Why aren't the treatments of FCRS and SCRS scram parallel?

Response

a) h applicant has noted the inconsistency and will provide an updated discussion in the PSAR by August 1982.

W e scram insertion speed requirements of the SCRS are shown in Figure 4.2-94 and discussed in Section 4.2.3.1 of the PSAR.

Test results to date have indicated that scram reactivity insertion requirements are met.

b) he two shutdown systems are diverse. Design and scram performance requirements have been established for each syst s. Given the distinct performance requirements for each system, discussion of the scram requirement and predicated scram performance can be made for each syst s.

QCS491.5-1 Amend. 69 May 1982

G2-0342

-- Feg3 - 2 [82-0342] #68 Ouestion CS230.3 In paragraph 2.5.2.10 of the PSAR, the Safe Shutdown Earthquake is represented es being a peak horizontal ground acceleration of 0.25g used in conjunction with a Regulatory Guide 1.60 response spectrum. Were is no indication for what foundation conditions this SSE is established. Since some category I structures are to be found on rock and others are to be found on soil / fill, ,

you should specify SSE peak acceleration and spectra for both of these '

foundation conditions.

W e same conditions should be considered for establishing the T E.

Response

h e Regulatory Guide 1.60 response spectra are based on recorded data of rock and soil sites and they are, therefore, applicable to both foundation conditions. .

h e peak ground acceleration of 0.259 selected for the SSE is based on the Newman intensity / acceleration relationship. A similar value is obtained using the Trifunac-Brady correlation which was developed from recorded data on the ground surface, primarily soil. mis value is applicable to both soil and rock foundation conditions although conservative for structures founded on rock. % e value of 0.25g was derived by assuming an associated earthquake intensity of VIII MM (the May 31, 1897 Giles County event) in conformance with NRC direction, notwithstanding the conclusions reached by other investigators that the appropriate intensity rating is VII-VIII as indicated in the PSAR.

l Similarily, for the CBE, the peak ground acceleration and the design response spectra are the same for both foundation conditions.

QCS230.3-1 Amend. 69 May 1982

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