ML20063C230

Forwards Response to 820305 Request for Addl Info 210 Re Mechanical Engineering.Specific Effects of Strain Rate & Strain Hardening Not Included in Plastic Sys Analysis

ML20063C230
Person / Time
Site:	Seabrook
Issue date:	04/30/1982
From:	Devincentis J PUBLIC SERVICE CO. OF NEW HAMPSHIRE, YANKEE ATOMIC ELECTRIC CO.
To:	Miraglia F Office of Nuclear Reactor Regulation
References
SBN-265, NUDOCS 8205040613
Download: ML20063C230 (35)

Text

o sua swa m ornee-IPUBLIC Companyof NewSERVICE Hampshir e 1671 Worcester Road Framinoham. Massachusetts 01701 (617) - 872 - 8100 to April 30, 1982 2  %

SBN- 265 T.F. B 7.1.2 2- pp EUED' l

MAyg L au 1982w ~

United States Nuclear Regulatory Conmission \ 7  %

Washington, D. C. 20555

% 0 Attention: Mr. Frank J. Miraglia, Chief O 6 Licensing Branch #3 Division of Licensing

References:

(a) Construction Permits CPPR-135 and CPPR-136, Docket Nos. 50-443 and 50-444 (b) USNRC Letter, dated March 1, 1982, " Request for Additional Information," F. J. Miraglia to W. C. Tallman (c ) PSNH Letter, dated April 21, 1982, " Response to 210 Series RAIs; (Mechanical Engineering Branch)," J. DeVincentis to F. J. Miraglia (d) USNRC Letter, dated March 5, 1982, " Request for Additional Information," F. J. Miraglia to W. C. Tallman Su bjec t : Response to RAI 210 Series RAIs; (Mechanical Engineering Branch)

Dear Sir:

We have enclosed responses to the following RAIs which you forwarded in Re f e rence (d):

210.4, 210.8, 210.18, 210.19, 210.21, 210.22, 210.23, 210.24, 210.38, 210.39, 210.42, 210.43, 210.44, 210.4 5, 210.4 6, 210. '+ 7, 210.48, 210.4 9, 210.51, 210.52, 210.57, 210.58, 210.59, 210.60, 210.61, 210.62, 210.63, 210.66, 210.67, 210.68, 210.69 The outstanding 210 series RAIs are the following:

210.15, 210.30, 210.50, 210.56, 210.64, 210.65 We plan to submit responses to the outstanding 210 series RAIs at the May 11-13, 1982, MEB Review Meeting.

Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY un b {

8205040613 820430 . /J. DeVincentis '

PDR ADOCK 05000443 A

Project Manager pop

210.4 Section 3.2.1. Table 3.2-2, Sheet 1 O

Explain note 9 as it applies to the reactor coolant pump flywheel.

RESPONSE: Note 9 does not apply directly to the RCP flywheel. The flywheel is designed to maintain structural integrity during an SSE and under overspeed conditions that could result from a LOCA. A detailed discussion of the RCP flywheel is contained in FSAR Chapter 5. Since Note 9 to Table 3.2-2 is not applicable to the RCP flywheel it will be deleted and Table 3.2-2 revised to reflect the appropriate information for the Seabrook FSAR. The revised inf ormatioa will be provided at the MEB review meeting.

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210.8 Section 3.2.2.2, Table 3.2-2 Sheet 1 O

Explain your rationale for classifying the shell side of the reactor coolant pump thermal barrier heat exchanger as ASME Code Class 3 although the tube side is Code Class 1.

RESPONSE: Definition of the RCP thermal barrier relative to a tube side and shell side is not totally accurate. The tubes are located inside the pump casing and reactor coolant flows around the tubing.

There is no shell side in the strict sense of a heat exchanger.

Table 3.2-2 will be revised to correctly describe the thermal barrier and its classification.

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, 210.18 (3.6(B).2.3, Page 3.6(B)-15)

Provide a reference or further justification for the use of a maximum fiber strain of 50% of ultimate strain as an adequacy requirement for the load carrying capacity of piping.

RESPONSE: We refer to ANSI ANS 58.2 (draf t) November,1978, Section 6 3.2.a.1, for the use of this value as defining an upper-bound design limit.

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.s 210.19 (3.6(B).2.3, Page 3.6(5)-17)

What strain-rate and strain-hardening effects have you included in plastic system analysis?

RESPONSE: Specific effects of strain-rate and strain-hardening have not been included in plastic system analysis. The effect of strain-rate and strain-hardening is to increase the energy absorption capability of the material. For conservatism, we have treated the material as behaving in a purely plastic manner for the design of pipe rupture restraints.

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210.21 Section 3.6(N).2.1, Page 3.6(N)-1 In the primary loop, what size breaks are postulated for the design of pipe whip restraints? What size breaks are pontulated in the primary loop for determination of compartment pressurization and asymmetric loads? If breaks for either case are less than size, provide justification.

RESPONSE: In the design of reactor coolant system pipe whip restraints Westinghouse assumes that the restraints must limit the worst postulated pipe breaks (e.g., a full circumferential break or one full area slot break). It should be recognized that these restraints are designed to limit pipe displacements resulting from such postulated pipe breaks.

The break areas .sssumed in various analyses performed by Westinghouse at the break locations identified in Section 3.6N vary in size and do include limited break areas. These break opening areas are limited by the physical constraint of pipe whip restraints. A further description of the specific break opening areas assumed in analyses performed for the Seabrook plant and further justification for these limited areas will be provided at the MEB review meetings. It should be noted that justification for limited break opening areas has previously been presented to the HEB on other dockets and found acceptable. The information previously presented to the MEB is also applicable to the Seabrook plant.

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.s 210.22 Section 3.6(N).2.3, Page 3.6(N)-7 Provide a copy of test results of pipe-to-pipe impact. Also provide test results that show whipping or bending of a stainless steel pipe does not cause the section to become a missile.

RESPONSE: Westinghouse has performed tests demonstrating the characteristics of whipping stainless steel piping including its missile characteristics. The test data developed by Westinghouse is proprietary and has not been published. It should be further recognized that the criteria used by Westinghouse for whipping pipe has been confirmed through other non-Westinghouse tests and has been incorporated by the NRC in Standard Review Plan 3.6.

Further discussion of this item will be provided at the MEB review meeting.

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.s 210.23 Section 3.6(N).2.5, Page 3.6(N)-11 Review of this section shows that you have used a cumulative usage factor 0.2 for postulated pipe rupture criteria. Branch Technical Position MEB 3-1 specifies a cumulative usage factor of less than 0.1. Provide a commitment to meet this criteria.

RESPONSE: Based on the number of break locations considered in the reactor coolant system a cumulative usage factor of .2 has been justified. Additionally, the NRC approved the .2 criteria in WCAP-8082 (Reference 1 to Section 3.6N) which is applicable to the Seabrook plant. It should be noted that this item was discussed with the MEB during previous review meetings on other dockets and it was agreed that the FSAR would be revised to delete the .2 cumulative usage factor and replace it with a reference to WCAP-8082.

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210.24 Section 3.6(N).2.5, Figure 3.6(N)-2 In addition to showing postulated break locations, they must be identified as either circumferential or longitudinal. Structural barriers, if any, restrain location and constrained directions must also be included in order to complete our review.

RESPONSE: Table 3.6N-1 identifies postulated reactor coolant system pipe breaks as circumferential (guillotine) or longitudinal (slot break).

A description of the RCS pipe whip restraints and a description of their location is provided in Section 5 of the Seabrook FSAR.

Additional information on the exact location of these restraints will be provided at the upcoming MEB review meeting.

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210.38 (3.9(B).3.4, Page 3 9(B)-26)

Provide an example of the analysis performed on ASME Code Class 1, 2 and 3 valve supports.

RESPONSE: To date, analyses of actuator supports for Class 1, 2 and 3 valves have been delayed because of a lack of valve dimensional information from valve vendors. This information is currently being generated at the jobsite in the form of as-built data. A typical analysis will be complete and submitted by July 1, 1982.

However, in all cases, supports for actuators of Class 1, 2 and 3 valves will be analyzed and designed in accordance with Subsection 3.9(B).3.4a, ASME Code Class 1, 2 and 3 Piping Supports.

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210 39 (3.9(B).3.4, P3.9(B)-26)

The design criteria used for mechanical equipment supports needs clarification. Subsection NF, ASME Code,Section III is applicable to these supports. Justify the use of AISC allowable stresses and demonstra Re that your design criteria satisfy the requirements of Subsection NF.

RESPONSE: The supports of certain mechanical equipment purchased circa 1974 were designed in accordance with the requirements and allowable stresses defined in the AISC Manual for the Normal and Upset Conditions. The allowables are the same as those specified by ASME III, Subsection NF where Level A and B Service Limits apply.

For the faulted condition, tensils and bending stresses were limited to 90% of the material yield strength and shear stresses were limited to 60% of the material yield strength which compares favorably with the limits defined by ASME III, Subsection NF, for Level D Service Conditions. Buckling evaluations were performed in accordance with the AISC criteria using faulted condition loads.

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4 210.42 Section 3.9(N).1.2, Page 3.9(N)-20 Provide references 1 and 2 for our review.

RESPONSE: Reference 1 (WCAP-8252) hss been reviewed and approved by the NRC. Reference 2 (WCAP-8929) has been submitted to the NRC and is currently being reviewed by the NRC and Oak Ridge National Labs.

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4 210.43 Section 3.9(N).1.4, Page 3.9(N)-33 How is the critical buckling strength for supports determined?

RESPONSE: Westinghouse performs buckling analysis in accordance with the requirements of the ASME Code,Section III, Appendix F. Further details of the buckling analysis will be presented at the upcoming MEB review meeting.

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210.44 3.9(N).2.5, Pages 3.9(N)-38 to 44 Previous analysis for other nuclear plants have shown that certain reactor system components and their supports may be subjected to previously under-estimated asymmetric loads under the conditions that result from the postulation of ruptures of the reactor coolant piping at various locations.

The applicant has described the design of the reactor internals for blowdown loads only. The applicant should also provide information on asymmetric loads. It is, therefore, necessary to reassess the capability of these reactor system components to assure that the calculated dynamic asymmetric loads resulting from these postulated pipe ruptures will be within the bounds necessary to provide high assurance that the reactor can be brought safely to a cold shutdown condition. The reactor system components that require reassessment shall include:

a. Reactor pressure vessel.

b. Core supports and other reactor internals.

c. Control rod drives.

d. ECCS piping that is attached to the primary coolant piping.

e. Primary coolant piping.

f. Reactor vessel supports.

The following information should be included in the FSAR about the effects of postulated asymmetric LOCA loads on the above mentioned reactor system components and the various cavity structures.

1. Provide arrangement drawings of the reactor vessel support systems in sufficient detail to show the geometry of all principal elements and materials of construction.

2. If a plant-specific analysis will not be submitted for your plant, provide supporting information to demonstrate that the generic plant analysis under consideration adequately bounds the postulated accidents at your facility. Include a comparison of the geometric, structural, mechanical, and thermal-hydraulic similarities between your facility and the case analyzed. Discuss the effects of any differences.

3. Consider all postulated breaks in the reactor coolant piping system, including the following locations:

a. Steam line nozzles to piping terminal ends.

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b. Feedwater nozzle to piping terminal ends.

, c. Recirculation inlet and outlet nozzles to recirculation piping terminal ends.

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4. Provide an assessment of the effects of asymmetric pressure

, differentials

• on the systems and components listed above in combination with all external loadings including safe shutdown earthquake loads and other faulted condition loads for the postulated breaks described above. This assessment may utilize the following mechanistic effects as applicable:

a. Limited displacement break areas.

b. Fluid-structure interaction.

c. Actual time-dependent forcing function.

d. Reactor support stiffness.

e. Break opening times.

5. If the results of the assessment on item 3 above indicate loads leading to inelastic action of these systems or displacement exceeding previous design limits, provide an evaluation of the inelastic behavior (including strain hardening) of the material used in the system design and the effect of the load transmitted to the backup structures to which these systems are attached.

6. For all analyses performed, include the method of analysis, the structural and hydraulic computer codes employed, drawings of the models employed and comparisons of the calculated to allowable stresses and strains or deflections with a basis for the allowable values.

7. Demonstrate that safety-related components will retain their structural integrity when subjected to the combined loads resulting from the loss-of-coolant accident and the safe shutdown earthquake.

8. Demonstrate the functional capability of any essential piping l . when subjected to the combined loads resulting from the i loss-of-coolant accident and the safe shutdown earthquake.

RESPONSE: Westinghouse included asymmetric loads in the design of the l Seabrook plant. The analyses methods used are consistent with NUREG-0609. The Seabrook FSAR addresses asymmetric LOCA load analysis in Section 3.9N. A further discussion of this item will be provided at the upcoming MEB review meeting and further information provided in the FSAR if necessary.

• Blowdown jet forces at the location of rupture (reaction forces), transient differential pressures in the annular region between the component and the wall, and transient differential pressures across the core barrel within s the reactor vessel.

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210.45 Section 3.9(N).2.5, Page 3.9(N)-41 Your statement that the loading imposed by the SSE is generally small compared to blowdown loads implies that in certain cases you have neglected loads due to an SSE. If this is true, provide analysis details justifying your doing so.

RESPONSE: This section of the FSAR refers to the dynamic analysis of the reactor internals. For the reactor internals both LOCA and SSE locals are evaluated. It should also be pointed out that all Seismic Category I components are seismically analyzed regardless of the relative magnitudes of LOCA (if appropriate) and SSE loads.

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210.46 Section 3.9(N).4.3, Page 3.9(N)-60 The statement "The stress limits are established not only to assure that peak stresses will not reach unacceptable values, but also limit the amplitude of the oscillatory stress component in consideration of fatigue characteristics of the materials" needs clarification. What are these stress limits and from what source were they obtained?

RESPONSE: For the CRDM pressure housing the stress limits and fatigue criteria defined in the ASME Code for Class 1 components are applicable.

Information on the seismic analysis, including scram capability, of CRDMs will be added to Section 3.9N.4 of the FSAR. Evaluation of non pressurized components was addressed at a previous MEB review meeting based on review of the design specification, discussion of testing performed by Westinghouse and Westinghouse licensees, and the contents of Section 3.9N.4 of the FSAR.

This information will again be discussed at the upcoming HEB review meeting. In addition, sinor changes to the FSAR may be required to resolve this iten.

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210.47 Section 3.9(N).4.3, Page 3.9(N)-61 Provide assurance that deformation limits are sufficient to guarantee control rod drive system integrity and functioning after a dynamic event such as an OBE.

RESPONSE: Following a dynamic event such as a LOCA or SE Westinghouse and NRC (SRP 4.2, Appendix A) criteria require that control rod

. Insertability be demonstrated. Since the control rods are passively designed to be inserted following a reactor trip and depend only on gravity, it is necessary to limit deflections and deformation in the reactor internals and fuel grids to guarantee this function. For the Seabrook plant the reactor internals and fuel grids are designed to minimize deflections to ensure control rod insertability. These analyses are further described in FSAR Section 3.9(N).2 and 4.2. Details of this item will be further discussed at the upcoming HEB review aceting.

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210.48 Section 3.9(N).5.2, Page 3.9(N)-69 The statement "The stress limits are established not only to assure that peak stresses not reach unacceptable values, but also limit the amplitude of the oscillatory stress component in consideration of fatigue characteristics of the material" needs clarification. What are these stress limits and from what source were they obtained?

RESPONSE: The referenced paragraph will be clarified to define the criteria used in the design of reactor internals. This item will be further discussed at the upcoming MEB review meeting. Also, see the response to item 49 for further clarification.

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210.49 Section 3.9(N).5.2, Page 3.9(N)-68 to 71 Subsection NG, ASME Code,Section III should be referenced as the design criteria for all design analyses, not just for the design basis accident.

RESPONSE: As discussed with the NRC on other plant dockets, the Seabrook

- reactor internals were procured prior to implementation of

,, Subsection NG of the ASME Code. However, the Seabrook reactor internals were designed and fabricated in accordance with Subsection NG although there is no specific Code Stress Report or Stamp. The FSAR will be revised to address Subsection NG of the ASME Code.

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210.51 Section 3.9(N).5.4, Page 3.9(N)-71 What are the stresses associated with the maximum deflections in Table 3.9(N)-177 State the basis for these deflections limits.

Justify .ce lack of safety margin for radial outward deflection.

RESPONSE: The stresses and deflections on reactor internals have been discussed on other dockets and determined to be acceptable.

_, Although the Seabrook reactor internals were procured prior to implementation of Subsection NG of the Code, stresses are generally consistent with Code limits. Deflections are below the limits identified in Table 3.9(N)-17. This item will be discussed further at the upcoming MEB review meeting.

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210.52 Section 3.9(N).2.3, Table 3.9(N)-1 Provide information on how the number of occurrences of steady-state fluctuations were determined. How was the effect of transients listed in this table considered for BOP equipment?

RESPONSE: The requirements for steady-state fluctuations are based on the overall NSSS design and operating experience. This condition is

, incorporated in the design of NSSS systems and components. The NSSS design transients are provided to the BOP designer as interface criteria such that they can te incorporated in BOP equipment as appropriate. This ites will be discussed further at the upcoming HER review meeting.

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210.57 (3.6(B).2.1, Page 3.6(B)-7)

Explain in more detail how intermediate break points are determined in Class 1 BOP piping when stresses and usage factors are below 2.4 Se on 0.1, respectively.

RESPONSE: The criteria of Regulatory Guide 1.46 were followed for the intermediate break point determinations. PSAR Paregraph 3.6(B).2.la(1)(d) is being deleted in Amendment 45.

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. s 210.58 (3.6(B).2.1, Page 3.6(B)-7)

It is the staff's position that if a structure separates a high energy line f rom an essential component, that separating structure should be designed to withstand the consequences of the pipe break in the high-energy line which produces the greatest effect at the structure irrespective of the fact that the pipe break criteria might not require such a break location to be postulated. Provide assurance that the Seabrook plant meets the above requirement.

RESPONSE: In evaluating the effects of high energy line breaks on essential components, it was found that protection is most of ten provided by

separation distance due to the high energy lines being located f ar enough away from essential components that the danger of impact or jet impingement did not exist. In the case of the electrical trays in the control building, which were separated by the building wall from the adjacent main steam and feedwater lines, a guard pipe was provided to prevent jet impingement from the main steam line, and an energy-absorbing bumper was provided to prevent impact from main steam or feedwater lines that could damage the wall.

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210.59 (Appendix 3C, Page 4)

For circumferential breaks that are axially restrained, provide justification that the axial and lateral movement of the pipe will result in the fan jet impingement that is assumed. The staff's position is that a fan jet can occur when the broken piping is physically restrained f rom significant separation (axial pipe movement equal to or less than 1/2 pipe diameter and lateral pipe movement less than pipe wall thickness).

RESPONSE: Seabrook BOP piping is not arially restrained so as to restrict pipe action in such a f ashion as to form a f an jet. Fan jet calculations were not used in the analysis.

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210.60 What is the maximum allowable tip deflection of a restrained whipping pipe? Provide assurance that the deflections of the restrained whipping pipe will not affect the function of any safety-related components.

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RESPONSE: For BOP piping, pipe whip restraints are provided to maintain the motion of the ruptured spipe end within controlled limits. The limit of motion is the area within which no essential component can be af fected by inpact ur jet impingement.

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-> 21 0 3.1 6 - Provide a description .on examples of the di f f ereat types of pipe

< whip restraints and jet impingement barriers that are used in the

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RESPONSE: For BOP piping, pipe whip restraints are either bumpers (with or

+ 4. 't without energy-at.; orbing crushable pads), guides, . U-bolt.

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restraints, or a combination of these. Guides are structural-

, J. . steel frames surrounding the pipe. . U-bolts are only used in

,r tension. . Jet impingement barriers consist of sleeves or guard y

- ! t pipes. Drawings will be available at the MEB Review Meeting for P 3 T ",

, .< review.

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210.62 Section 3.9(N).2.3, Page 3.6(N)-9 When calculating the dynamics effects of jet impingement, what values are assumed for Ko for the various initial fluid conditions?

I RESPONSE: Westinghouse is currently reviewing their jet impingement analysis for the RCS and will provide the values of Ko used in this

, analysis at the upcoming MEB meeting. The values used are consistent with those used for other Westinghouse plants and j discussed at other MEB review meetings.

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210.63 Table 3 . 6 ( ;: )-2 Why a re only seven break locations listed? There are eleven design breaks postulated. Provide the CUF and moments for the other f our designer break locations. In addition, provide any

.other modes where the CUF exceeds 0.1. Provide the CUF values specific for Seabrook at each design break location and at any locations where the CUF exceeds 0.1.

RESPONSE: Westinghouse is reviewing the Table 3.6N-2 and will discuss this item further at the upcoming NEB review meeting. The cumulative usage factor was discussed relative to Item 23 and it was noted that the criteria in WCAP-8082 are acceptable relative to a cumulative usage factor of .2.

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Provide a more detailed description or t. h e l o.n t s , load combinations, and stress limits that are used in the design of ASME Code Class 1, 2 and 3 components and component supports.

RESPONSE: The plant loading conditions and load combinations are summarized in Table 3.9(B)-2 for ASME components and their supports. Stress limits applicable to the various components (excluding supports) for each loading condition are summarized in Tables 3.9(B)-3 through 3.9(B)-5. Note (1) on each of these tables instructs the reader to refer to Table 3.9(B)-2 for definitions of the plant loading conditions.

Tables 3.9(B)-2 and 3.9(B)-3 are being revised in Amendment 45.

Supports are addressed in FSAR Subsection 3.9(B).l.4(a). All component support designs, except those listed in the response to RAI 210.29, satisfy the design requirements and stress allowables of ASME III, Subsection NF. Those listed in the RAI 210.29 response are designed according to AISC criteria.

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TABLE 3.9(B)-2 DESIGN LOADING COMBINATIONS FOR ASME CODE CLASS 1, 2 AND 3 COMPONENTS Plant Loading Conditions Design Loading Combinations Normal P+D+L+M+NOL+T Upset P + D + L + M + N0L + OBE + H + Ft+T Emergency P + D + L + M + NEL + H + F t Faulted P + D + L + M + NFL + SSE + H + F t Where:

P = Pressure corresponding to the loading condition D = Dead weight L = Live weight of fluid handled; for liquid content of vessel or tank M = Snow or wind load for outdoor storage tank NOL = Nozzle load for Normal / Upset condition NEL = Nozzle load for Emergency condition NFL = Nozzle load for Faulted condition OBE = Operating Basis Earthquake (inertia load)

SSE = Safe Shutdown Earthquake (inertia load)

H = Dynamic fluid head effects (where applicable)

Ft = Valve thrust loads (where applicable)

T = Thermal load (where applicable)

e TABLE 3.9(B)-3 STRESS LIMITS FOR NON-ACTIVE CATECORY I, ASME CODE CLASS 2 AND 3 PUMPS Plant Loading

. Conditions (l) Stress Limits (2)

Normal ASME III, NC-3400 or ND-3400 Upset P, 1.lS (P, or P1) + Pb 1.65S Emergency P, 1.5S (P, or P1) + Pb 1.8S Faulted .

P, 2.0S (P, or Pg) + Pb 2.4S 1 Where:

S = Material allowable stress at maximum temperature from Appendix I of

'- ASME Section III.

P, = Primary general membrane stress, the average primary stress across the solid section under consideration. Excludes effects of discontinuities and concentrations. Produced by pressure and mechanical loads.

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= Primary bending stress. This stress is produced by pressure and mechanical loads including inertia earthquake ef fects but excluding

. effects of discontinuities and concentrations.

Py w Primary local membrane stress, the average stress across any solid f

section under consideration. Same as P, except that' discontinuities l are considered.

j NOTES i (1) Plant loading conditions are defined in Table 3.9(B)-2.

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210.67 l' ro vi de the stress limits used for !>olts.

U.S l'o:iS E 1. Anchor Botts for Eeluipment and finilding Columns Allowable Tensile Allowable Shea r Bolt Ma te ria l S t res s Stress ASTM A193 Grade B7 Under 2-1/2"O Fy = 105 ksi Ft = 0.6 Fy 0.5 Fu Fv = 0.4 Fy Fu = 125 ksi = 62.5 ksi = 42 kai ASTM AS40 Crade B23 Class 4 Up to 3"O Fy = 120 ksi Ft = 0.6 Fy 0.5 Fu Fv = 0.4 Fy Fu = 135 ksi = 67.5 ksi = 48 ksi ASTM A354 Grade BD For 1/4" to 2-1/2"O Fy = 130 ksi Ft = 0.6 Fy 0.5 Fu Fv = 0.4 Fy Fu = 150 ksi = 75 ksi = 52 ksi

2. liigh strength bolts for equipment on structural steel and for steel-to-steel connections.

ASTM A325 1/2" to 1"O l-1/8" to 1-1/2"O Fy = 92 ksi Fy = 81 ksi Fu = 120 ksi Fu = 105 ksi ASTM A490 1/2" to 1-1/2"O Fy = 130 ksi Fu = 150 ksi .

All allowable tension shear and bearing values are in accordance with pages 4-3 through 4-9 of Manual of Steel Construction - ASIC.

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?!U.68 Describe to what ex ten t . h t ;;h- :: r. ng t h bo l t s a re used.

RESPONS!:: ANCllOR 110LTS

1. The following three types of II.S. anchor bolts are used for equipment on concrete elements:

a. ASTM A193, Grade'B7

b. ASTM AS40, Grade B23, Class 4'

c. ASTM A354, Grade BD This represents approximately 10% of the total anchor bolts used for this purpose.

2. The following two types of anchor bolts are used for equipment on structural steel members:

a. ASTM A325

b. ASTM A490

3. Anchor bolte used for building columns (waste process building exterior columns):

a. ASTM A193, Grade B7.

CONNECTION BOLTS

1. The following two types of bolts are used for structural steel connections:

d a. ASTM A325 - 96% of total (approximately)

b. ASTM A490 - 4% of total (approximately) l i

4 e *

.O 210.69 ' The seabrook plant incorporate:4 the Westinghouse !!odel F steam generator. We will require that the results of the analysis to-determine the tube plugging eri teria' he presented to the Staff when_they become available.

RESPONSE: The plugging criteria used by Westinghouse f or steam generator tubes is defined in 'Section 1.8 relative to 'the Westinghouse position on Regulatory Guide 1.121.

The analysis to determine the tube plugging criteria is being performed and will be submitted to the Staff prior to commercial operation.

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