ML20196F633

Provides Revised Response to Integrated Design Insp Repts 50-327/87-48 & 50-328/87-48,Item U3.5-1 to Reflect Correct ASME Code Ref.List of TVA Commitments Attached

ML20196F633
Person / Time
Site:	Sequoyah
Issue date:	03/02/1988
From:	Michael Ray TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
NUDOCS 8803040135
Download: ML20196F633 (13)

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TENNESSEE VALLEY AUTHORITY CHATTANOOGA. TENNESSEE 37401 5N 105B Lookout Place MAR 2-1988 U.S. Nuclear Regulatory Connission ATTN: Document Control Desk Washington, D.C. 20555-Gentlemen:

In the Matter of ) Docket Nos. 50-327 Tennessee Valley Authority ) 50-328 SEQUOYAH NUCLEAR PLANT (SQN) - NRC INSPECTION REPORT 50-327/87-48 AND 50-328/87 INTEGRATED DESIGN INSPECTION (IDI)

Enclosure 1 provides TVA's revised response for IDI item U3.5-1 to reflect the correct ASME code reference, and a supplemental response to IDI. item DS.2-10, provided previously on December 29, 1987. Enclosure 2 provides a list of cotnitments being made by TVA in this submittal. It is our understanding that the revised responses provided herein complete TVA's actions for SQN unit 2 restart on these items.

If you have any questions, please telephone D. L. Williams at (615) 632-7170.

Very truly yours, TENNESSEE VALLEY AUTHORITY W

M. J.

I ay, Deputy Dire tor Nuclear Licensing and Regulatory Affairs Enclosures cc: See page 2 8803040135 880302 PDR. ADOCK 05000327 \

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An Equal Opportunity Employer

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• 9 .ew U.S. Nuclear Regulatory Commission

'21988 hAAR Rnclosures cc (Rnclosures):

Mr. K. P. Barr, Acting Assistant Director for-Inspection Programs TVA Projects Division U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, coorgia 30323 Mr. C. C. Zech, Assistant Director.

for Projects-TVA Projects-Division U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852 Sequoyah Resident Inspector.

Sequoyah' Nuclear Plant 2600 Igou Ferry Road Soddy Daisy, Tennessee 37379 1

ENCLOSURE 1 SEQUJYAH NUCLEAR PLANT (SQN)

ITEM NO: U3.5-1 TITLE: Piping Code of Record

SUMMARY

OF ITEM:

The piping code of record for design as stated in the Final Safety Analysis Report (FSAR) is ANSI B31.1-1967 Edition. Because ANSI B31.1-1967 did not

, define combinations for the normal, upset, and faulted conditions, TVA used l the stress-allowable equations from ASME Section III, Subsection NC-3000, Winter 1972 Addendum, for these plant conditions. In addition to the stress equations. TVA used the stress-allowable limits specified in the ASME Code.

This is documented in CEB's Rigorous Piping Analysis Handbook for Piping.

CEB's use of the ASME Code stress-allowable limits is not' consistent with FSAR Table 3.9.2-3, which commits to the use of ANSI B31.1-1967 stress-allowabla limits, l CLASSIFICATION: Documentation REVISED RESPONSE:

ANSI B31.1-1967 did not define stress allowables or equations for the loading combinations required for SQN; therefore, the stress allowables and equations from ASME Section III, Subsection NC, 1971 Edition through the Winter 1972 Addenda were used. This is consistent with industry practice (seo attachment) l and is considered appropriate and conservativo. This is also consistent with l ANSI B31, code Case 115, which permits the piping designed and constructed in l accordance with ASME Section III to be accepted as complying with ANSI l . B31.7-1969.

l Differences exist between the codo stress equations in ANSI B31.1-1967 and i ASME Section III-NC-W72 on both the right-hand side (RHS) and left-hand side (LHS) of the equations. For the RHS (the stress-allowable side), ASME Section l III-NC-W72 providos steess allowables for primary stress combinations classified as upset, emergency, and faulted. ANSI B31.1-1967 provides no such l allowables. For the LHS, ASME Section III-NC-W72 provides for the stress l intensification factor (SIF) to be applied to the resultant of the three i

moments (square root of the sum of the squares [SRSS) of the two bending

( moments and the torsional moment). ANSI B31.1-1967 applies the SIF to the  ;

i bending moments but not to the torsional moment. It also defines an in-plano  ;

! and out-of-plano SIF for reduced outlet tees. The LHS values determined based

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upon ASME Section III-NC-W72 generally bound allowable stress values determined from ANSI B31.1-1967.

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The stress in the RHS is the S value at temperature (S h) for both codes.

For many materials, these values are essentially identical. This is not generally true for austenitic steels (such as SA-312 304) that are widely used at SQN. The stress allowables (Sh values) in ASME Section III-NC-W72 for austenitic steels exceed the corresponding values in ANSI B31.1-1967. The basis for both values is consistent. The ANSI B31.1-1967 stress allowables (in accordance with note 6 in Appendix A Table A-1, ANSI B31.1-1967) are limited by 90 percent of yield at temperature. The same is true for ASME .

Section III-NC-W72 (in accordance with Table 1-7.2, note 3). Because ANSI A31.1-1967 does not have an Sy (yield strength) table, it is not possible to g eompare Sy values between ANSI B31.1-1967 and ASME Section III-WC-W12, but a comparison of ASME Section III-NC-W72 with ANSI B31.7-1969 shows that the ANSI B31.7 Sy values for austenttic steels at higher temperatures are lower.

The basis of this is not readily apparent; but because later (1973) B31.1 codes revise S and Sy values for these austenitic steels consistent with the values in ASME Section III-NC-W72, TVA has concluded that the Sh values given in ASME Section III-NC-W72 are appropriate for use at SQN. The allowable stresses promulgated in ASME Section III and later made consistent in B31.1 did not result in additional materials, fabrication, or inspection requirements in B31.1. Further, material regulrements for these materials in ANSI B31.1-1967 ANSI B31.7-1969, and ASME Section III-NC-W72 are similar.

Therefore, the use of ASME Section III-NC-W72 is appropriate for use and consistent with industry practices for B31.1 plants as well with later code editions wherein ANSI B31.1 and ASME Section III are simllar.

Equivalence of the allowable stresses between ANSI B31.1-1967 and ASME Section III, Class 2 and 3 (1971 Edition, Winter 1972 Addendum), is eneured because:

1. The stress criteria are identical; and

2. The allowable stresses were developed by the same committee for both documents.

ANSI B31.1-1967 allowable stresses (S e , S h) were based on the lower of 0.25 Su and 0.625 S y, where su is the ultimate tensile strength and Sy is the yield strength. These are the same criteria that were used in Sections I and VIII of the ASME Boller and Pressure lessel Code in 1967. When ASME Section III Class 2 and 3 rules were published, they adopted the stress criteria for Sections I and VIII and were, therefore, identical to ANSI B31.1-1967 in this aspect. ANSI B31.1 was revised in 1973 to eliminate the citation of specific criteria and to replace them with reference to the criteria of ASME Sections I and VIII. Thus, the stress criteria were, and are, identical for ANSI B31.1 and ASME Section III, Class 2 and 3.

The allowable stresses fo'c ANSI B31.1 and ASME Section III are both developed by the Subcommittee on Properties of Metals and its predecessors of the ASME Boiler and Pressure Vessel Committee, using the same database for both sets of allowable stresses. Even though the stress criteria are identical, minor differences may exist from time to time between' ANSI B31.1 and ASME Section '

III, Class 2 and 3, because of a las on the part of one document or the other in the adoption of changes to the allowable stresses resulting from the addition of later test results to the database.

The FSAR will be revised in the next annual update to change the source of allowable piping stresses from ANSI B31.1.0-1967 to ASME Section III, 1971 Edition through Winter 1972 Addenda. Attached to this response are the proposed revisions to the FSAR.

REFERENCES:

None ,

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ATTACHMENT TO ENCLOSURE 1 P

ANSI B31.1-1967 Code was establisheo for nonnuclear power piping design. The [

code equations specified in this code are to calculate longitudinal pressure  !

stresses (S ep) and expansion stresses (S E ) as defined in equation 8 of the Code.

No clear definitions were given as to how to calculate the stressen because of seismic and other dynamic loads.

Most of all, ANSI B31.1-1967 Code, as is, is not suf ficient for use in the design of nuclear power plants. Specifically, the Code does not provide i sufficient detailed code rules to comply with NRC regulator requirements, such as design limits and loading combinations that were specified in Regulatory cuides 1.48 and 1.57, etc.

ANSI B31.7 Code was established in 1969 for nuclear power piping. However, the design rules of nuclear classes 2 and 3 piping are referred back to ANSI B31.1-1967. Thereforo, ANSI B31.7-1969 Code also cannot provide sufficient -

design rules for nuclear power piping to meet regulatory requirements; 1 further, ANSI-B31.7-1969 Code was replaced by ASME III-1971 Code. However, the ASME IIINC, ND Code equations, equations 8, 9,10, and 11 (or ANSI B31.1 code equations 11, 12, 13, and 14), were not issued until ASME Section III-1972 Addenda. ANSI B31.1 Code was also revised in Summer 1973 Addenda to incorporate these code equations without change of material requirements. In  ;

December 1973, ANSI B31.1 Code case 115 was issued to accept ASME Section III

• Code rules as complying with the requirements of ANSI B31.7-1969. Therefore, it is the nuclear industry practice to use ASME Section III Code equations to design nuclear classas 2 and 3 piping for those nuclear power projects

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committed to ANSI B31.1 Code.

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FSAR' REVISION.  ;

-(Revised dates on ASME III code reference from 1974 through Winter'1976 - .

Addenda to 1971 through Winter 1972 Addenda. This is the only change made '- .

from the TVA December 29, 1987 IDI response.). ,

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Valve Set Accunulation Blowdown Mk. No. Press.  % Press.  % Press.

47W400 101 1064 10.0 1170 10 958 47W400 102 . 1077 8.6 1170 10 969 47W400-103 1090 7.3 1170 10 981 4 s 47W400 104 ,'

1103 6.1 1170 10 993 ,

4 7W400- 105 1117 4.7 1170 10 1005 All valves are connected.to a rigidly supported common header that is in turn connected to the main steam piping through a branch pipe equal in size to the main steam piping. The header and valves are located immediately outside containment in the rain steam valve building. ~

The safety valves are mounted on the header such that they' produce torsion, bending, and thrust loads in the header during valve operation. The header has been designed to accomodate both dynamic and static loading ef fects of all-valves blowing down -

simultaneously. The stresses produced by the following loading ef fects assumed to act concurrently are within the .".201. 221.1.^.

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ASME Se.c. tion III l'l71 Edition therup Winter M72 Addenda..

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1. de'adweight ef fects,

2. thermal loads and movements,

3. seismic loads and movements,

4. safety valve thrust, moments, torsional loading, and

5. internal pressure. .

, The nozstes connecting each valve to the header are analyzed to assure that for both dynamic and static loading situations, the stresses produced in the nozzle wall are within ANSI B31.1.0, 1967 code allowable for the same loading consideration as the header,

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The safety valves and power-operated atmospheric relief valves are Seismic Category I components. They have been seismically quallfled by analyses per criteria presented in 3.7.3 and Table 3.9.2-3.

Pressure relief valves in auxiliary safety-related systems have been installed considering loads carried in the support members produced by:

1. deadweight of valve and appurtenances, s.

/. t hemal ef fect s, 3.9-24 s

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8. Residual. heat removal system

9. Component cooling system

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10. Essential raw cooling water

11. Auxiliary boiler piping

12. Upper head injection piping

13. Parts of other systems which require rigorous analysis. .

3.9.2.5.2 Analytical Methods "

Loadinz Conditions and Stress Limits' The design loading combinations and the allowable stress limits '

considered in the design of TVA piping systems'within the scope of Subparagraph 3.9.2.5.1 are shown in Table 3.9.2-5. Design loading combinations are categorized with respect to normal, upset, and faulted conditions. "lpin; :::.;:n:nt; h: : 5::n d::!;n:d S11 i: 211: ti: :tr::: int:::ity 1: :1; giv;n by th '

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1. Stress evaluations due to loadings such as deadweight, thermal expansion, and anchor movements are performed using static analysis techniques, while stress evaluations due to earthquake loadings are performed using dynamic analysis

' techniques. The computer programming for appilcation of both techniques is described in subparagraph 3.9.2.5.3.

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2. Loads on equipment nozzles are combined and evaluated against allowables as follows:  !

. i Fat + Fgy + F1/2SSE i Allowable

3. Seismic valve accelerations are generaly maintained below i 2 s vertical, and 3 g horizontal. Cases exist such that valve l

accelerations can exceed these standard limits. Such cases are 4 evaluated and approved individually; this process is

• I controlled by the Rlrorous Piping Analysis Itandbook_ {

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INSERT A Piping components have been designed to the ANSI B31.1 1967 power piping code '

utilizing the equations and rules from ASME Section III, Winter 1972 Addenda for loading combinations not defined in ANSI B31.1, and the allowable stress and material property values form Appendix I of the ASME'Section III, 1971 Edition through Winter 1972 Addenda.

While the ANSI B31.1 1967 code did not define allowable stress limits for some of the loading combinations considered in Tabic 3.9.2-5, the allowable stress levels are-in basic agreement with Appendix I of the ASME Section III, 1971 Edition through Winter 1972 Addenda.

The rules and criteria of ANSI B31.1 1967 are considered to be equivalent to those of section NC3000 of the ASME section III. Winter 1972 Addenda with the appropriate additional consideration of the equation 9 requirements of the' ASME code.

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Table 3.9.2-3 SAFETY CLASS B, C, AND D 00MPONDff IDADING CONDITIONS AfD STPISS LI.V ITS CDMIONCIT Flant Atmospheric Contain:xnt Icading Vessels and Storage Storage Penetrations Conditior. Piping Tanki (0-15 psig) Tanks Ptucps Valves (Nozzles)

Primary Stresses' Primary Stresses StressesiSe ASE III Draft ASME III, ASME,Section III, TSh U f r Purgs and 1971 Edition, m 1968/ ANSI Valves. Per- B16.5 Subsection NE formance Ratings Normal Expansion Stress

• Primary + 4 c d Ua See ndary-3 Sm with standards '

of the Hydraulic Institute -

Procedures .

Primary Stresses Primary Stresses Stressesbe Structural integrity is d 1.2 Sh ue by the integrity ASME,Section III, ism 1971 Edition, Up' set of the connecting piping. Subsection NE Pu::ps and valves,sre -

gansionStress Primary + supported to assure each

-S a e- Secondaryk S , component is not seismically loaded in excess of the "g" loading specified in the ASME,Section III, ~

design specification.

1971 Edition, Duergency NA NA NA Pumps and valves have Subsection NE *

' been demonstrated to be rigid (r o225 n,). .lligher accelerations and lower Primary Stresses Pr Stresses Stresses natural frequencies 12A Sh 1. 3 -1.25 S e may be approved on a 5 Taulted Erpension stresses Secormiary stresses Case-by-Case basis. NA need not be need not be i

evaluated evaluated S 31.1."

h= "C " Code allowable stress Sa =5^' ^31.1.0 - 13',7 Code allowable -

at design tensperature.. expansion stress. f 5, = 13MC Section III,1968 Edition Code allowable NA = No loam nr condition assigned.

stress at design temperature

,52 = AWA Standard D100 Code allowable stress.

ASME S e,t,Non III~,1971 Ediben necqh IN 4 '" ISTA Ndd* '

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ITEM No.: DS.2-10 TITLE: Adequacy of ERCW Instrumentation Provided for Detection of Break in Non-Selsmic ERCW Piping l

SUMMARY

OF ITEM:

In the event of a pipe break in the~non-seismic portion of the ERCW piping, operator action is taken based on a high flow alarm and status light in the control room that monitors each ERCW header.

CLASSIFICATION: Design Deficiency ,

SUPPLEMENTAL RESPONSE:

In the December 29, 1987 response to this item. TVA committed to presue implementation of an automatic isolation' scheme to isolate the Class H ERCW piping going to the turbine building. By this letter, TVA commits to completing the implementation of this modification before SQ4 unit 2 restart. -

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ENCLOSURE 2 l LIST OF COMMITMENTS )

1. U3.5-1 i The FSAR will be revised in the next annual update to change the source of allowable piping stresses from ANSI B31.1.0-1967 to ASME Section III, 1971 Edition through Winter 1972 Addenda.

2. D5.2-10 Before unit 2 restart TVA will complete implementation of an automatic isolation scheme to isolate the class H ERCW piping going to the turbine l building.

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