ML20114E761
| ML20114E761 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 02/28/1979 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML19241B424 | List: |
| References | |
| NUDOCS 7907170131 | |
| Download: ML20114E761 (14) | |
Text
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. l ENCLOSURE 1 PRELIMINARY EVALUATION OF SEQUOYAH UNIT 1 FLAW INDICATION
[b: SUPPLEMENTAL REPORT SwRI Project 17-5339 February 1979
' Prepared for Dccht #5g327 ' ~""3 Tennessee Valley Authority Cc::!rt D3!eMl#J[pf 79d7/90/,q,g' O 505 Edney Building Chattanooga, Tennessee 37401 MEULU0aycccyay?C22d-g g-Prepared by: Approved by:
i.
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&** u YY s. . _
- / Clarenc ed,-1
- autzenheiser Vice President e' I Quality Assurance Systecis
-Reviewed by: and Engineering Division
, Project Manager W / f t - 7907170/3 d ' d
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A aE FRELIMINARY EVALUATION OF SEQUOYAH UNIT l' FLAW INDICATION l' l
? f i 1.0 Introduction i s / .
, The preservide ultrasonic examination of Sequoyah Unit I reactor vessel p
closure head weld WO9-10 revealed a flaw indication exceeding the ac'cep-tance standards of the American Society of Mechanical Engineers (ASME) j i Boile; a'nd Pressure Vessel Code Section XI, . Division 1 IWB-3500. Re-(- _ , evaluation _of .the weld joint radiograph identified the indication as sub-
/ i' [ ject-to exemption from IWB-3500 standards in accordance with Code Case
/ N-2091 .
f CodeCaseN-209r$quiresevaluationoftheflawindicationbytheIWB-
' 3600 analytical procedures. This report describes a preliminary evalu-ation to determine the allowability of the indication and the potential
'd', effects of this flaw on. operating procedures. The evaluation presented herein is based, on the pressure vessel manufacturer's stress report 2, l 4/' A more detailed analy is and evaluation may be required to justify leav-t ing this flaw indication in place. The final disposition of this indi-cation and the justifi, cation for this disposition is ':o be established by others. - - i 2.0 General Discussion of Eotential Primary Coolant Boundary Failure at the Flaw Indication Lo;ation and the Consequences of Such Failure The flaw locatio'n Oith respect to any structural discontinuities (such as ar flange, nozzle, or lug) is sudh that the stresses at this section are about equal to'the general primary membrane pressure stresses and the local thermal bending stresses-(discussed as " skin effect stress" in the
.- ' referenced stress ' report). The local surface area is not subjected to
' coolant injection flow cooling during postulated incidents.(such as loss
?of coolant i accidents) so that the increase in thermal stress at the flaw 11ocation,r in response to emergency and faulted incident loadings, is
. expected to be less than the decrease in pressure stress at ~ specific times after incident initiation. Neutron irradiation effects on mate-L . rial propirties are negligible at this location.
I The growth o& a flaw through the coolant boundary at this location would result only in a steam leak (if unstable crack propagation is inter-dicted). This 'would not diminish the system capability for core submer-gence and cooling or safe shutdown in response to an emergency or faulted - incident. System instrumentation would detect a steam leak. 1
- 1 Wil'liam McGaughey and Sam Wenk, Sequoyah Unit 1 Draft Field Notes, Circumfer-ential Weld WO9-10, Southwest Research Institute, January 31, 1979.
N 2" Analysis of the Main Closure Including Core Support Ledge," Rotterdam Stress
-Report,' Document No. 30616-1105 (The Rotterdam Dockyard Company, September 22, 1975T
/
1
. j 3.0 Fracture Mechanics Analysis (n
\s_,/
ASME Section XI, Division 1, Appendix A, A-1100, summarizes the analyt-ical procedure to be used. This procedure was followed step-by-step, as described below. (1) Ultrasonic and radiographic examinations evaluated by McGaughey and Wenk determined the actual flaw indication configuration. Figures 1 and 2 depict these results. (2) The flaw indication was resolved into elliptical shape as depicted by Figure 3 Flaw Indication Characterization. (3) The stresses at the flaw indication location were obtained from the ( manufacturer's stress report and posted in Table 1, Flaw Growth Prediction, for the various design conditions listed. (4) The stress intensity factors for each condition were calculated and posted in Table 1 (see Figure 4, Sample Calculation). l l (5) The necessary material properties were taken from the stress j report. (6) The analytical procedures described in A-5000, as appropriate, were t used to determine the critical flaw parameters (see Table 1 and
- Figure 5, IWB-3600 Analysis).
.I \ (7) The flaw evaluation criteria of IWB-3600 were used to determine I
(s_,/ that the observed flaw indication is acceptable for continued oper-ation provided that metal temperature exceeds the reference nil-
, ductility temperature (RTNDT) by 120*F during the primary side hydrostatic test.
I 4.0 Conclusions e The fracture mechanics analyses performed using the conservative method l of Section XI, Appendix A, the manufacturer's stress report, and the cri-l teria of IWB-3600 indicate that: (1) If any flaw growth occurs in response to the specified operating ,. loadings, it would be negligible. I (2) . Unstable crack propagation initiation from the defect is not cred-ible if the metal temperature exceeds RT NDT by 120*F during the primary side hydrostatic test. This is the highest stress / lowest metal temperature worst case condition postulated in the stress report. The minimum critical fisw size (at) for emergency and faulted conditions has not been determined because of lack of information. There are two conditions to consider:
\
( _,/ (1) Large steam line break. (2) LOCA-ECC thermal shock. 2 __. g .- ,
/
Both of these conditions have been evaluated by Westinghouse on a generic [m basis and for specific reactor vessels. For the preliminary evaluation, it can be noted that: (1) The large steam line break minimu:n fluid temperature is 212*F. This is - >200*F above RTNDT for the material at this location. The fracture toughness at this temperature is sufficient to permit yield strength loading of a through wall crack without unstable pro-pagation (see Figure 4). (2) The flaw is located above any cold water injection location for LOCA-ECC event, and the consequent thermal shock would induce no higher stress than the large steam line break. . (3) Figure 5 is an example of the result of an overly conservative analysis which would be more severe than either condition. As can be seen, neither condition would result in unstable crack propaga-tion. G 3
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$ ] SAMPLE CALCULATION FLAW GR0hTH INSERVICE Loading No. 1 - Primary Side Hydrostatic Test @ 3,105 psig for 5 cycles 0 60'F. From Stress Report o ys-
= 50.07 kai o, = 20.22 ksi'
. -1.98 ksi ;
ob From Section XI, Appendix A (1977 Edition) ! Fig. A-3300-1, Q = 1.9 l Fig. A-3300 Point 1 Point 2 ,
- M,= 1.035 1.025 i Fig.-A-3300-4 j
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t l f - Fig. A-4300-1 (Summer 1977 Addenda) i , da/dN = 2 micro-inches / cycle
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1 i i. IWB-3600 ANALYSIS FOR HYDROTEST Ag =. 0.625 inches (see. flaw growth inservice) ! Assume cA* = 0.7 inches > 4
'o m = 20.22 + 8 =.28.22 ksi (1) o og = 1.98 kai 4
- Then K*la = 28.22 x 1.035 6 /0.7/1.87 - 1.98 x 0.47 6 /0.7/1.87 4
{ = 30.67 ksi dn~ I K Ia
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.?. '. '
<r pyi !: . i* ' y' "j Memorandum TENNESSEE VALLEY AUTHORITY rx CC'
- 7. :/
c. ( ? : H. S. Fox, Director of Power Production, 716 EB-f7 (2) v . . - _ . FRoM
.Roy H. Dunham, Director of Engineering Design, Wil A9 C-K -
DATE : g !, _. . _. ,
SUBJECT:
SEQUOYAH NUCLEAR PLANT UNIT 1 - ASME SECTION XI FRACTURE MECHANICS' EVALUATION OF A FLAW INDICATION IN THE REACTOR PRESSURE VESSEL CLOSURE T.' . T ? ~.~ HEAD 7 sM
? I' kiL~.:....
.7._._,...
Reference conversations between E. A. Merrick (EN DES-MEB) and E. ,F. E ~ ',T[ ', Harwell (P PROD). t-
, p .f.'( '....m....
On February 1,1979, we received data from E. F. Harwell on an ind'ication in the unit 1 reactor pressure vessel (RPV) closure head dollar plate to 82 ?. flange transition forging weld. The indication had been judged unacceptable IC to the requirements of ASME Section XI, and a decision was made to justify it to the requirements of ASME. Code Case N-209. E. F. Harwell requested pr] wo perform an analytical evaluation of the flaw. >. I nH.__I EN DES-CEB has performed that evaluation by application of procedures.i93 h._.t from Appendix A to ASME Code Section XI. Compliance with current Sect 101M .
~
p XI flaw acceptance criteria has been demonstrated. e -{ { Our evaluation predicts that the existing subsurface planar flaw will grow to an elliptical crack with 2.69- and 2.07-inch major and minor diameters [3
~ after 40 years of service under specified operating conditions. Applied stress intensification factors are calculated for the final crack dimen-sions and compared to allowable stress intensification factors from para-graph IWB-3612 of Section XI, thereby demonstrating compliance with the flaw acceptance criteria.
The complete evaluation will be released as report CEB-CQS-79-1 on approximately February 23, 1979, after completion of our internal review. EN DES has the necessary expertise and component information inhouse to perform work of this nature in a timely responsive manner. This would be especially important for an operating plant. Consequently, we strongly recommend that you consider using EN DES to perform future evaluations rather than having the work performed outside of TVA. We shall forward you copies of the report upon completion of the review cycle. EN DES engineers will be available to provide support for your presentation of the information on the NRC upon request. f% $ //./g.'M w v) i j Ro~y H. Dunham - - -
. DRP:EAM:MGR 1' cc: R. G. Domer, W9D224 C-K R. M. Pierce, 204 GB-K (2)
MEDS, E4837 C-K G. G. Stack, Sequoyah CONST (4) D. R. Patterson, W10Cl26 C-K E. F. Thomas, 550 CST 2-C H. H. Mull, E7824 C-K 167127/4 Buy U.S. Savings Bonds Regularly on the Payroll Savings Plan
... ENCLOSURE 3/
7 . . . . ..,4 ,_ , T N"'.'.'D STATES GO\*ERNMENT , '3 ;, # 67M0Tdn d um TENNESSEE VALLEY AUTIIORITY p
)0 : H. S. ' Fox, Director of Power Production, 716 EB-C (2) 0 01 790410 MS FROM '
' Roy H. Dunham, Director of Engineering Design, Wil A9 C-K h .iw S/tm3 DATE D g, p.
SUBJECT:
SEQUOYAH NUCLEAR PLANT UNIT 1 - ASME SECTION XI FRACTURE MECHANICS EVALUATION OF A FLAW INDICATION IN THE REACTOR PRESSURE VESSEL CLOSURE HEAD yl190330904 Reference my memorandum to you dated / February 13, 1979 (MEB 790214 353). Attachment 1 is the fracture mechanics evaluation report no. CEB-CQS-79-1 promised you in the referenced memorandum. Attachment 2 (R. G. Domer's memorandum to D. R. Patterson dated faarch 19, 1979, CEB 790319'011) contains comments developed by our Civil Engineering Branch on Southwest Research Institute's evaluation of the indication in the reactor vessel head. It is our position that report no. CEB-CQS-79-1 provides adequate assurance that the flaw indication is acceptable for the full service life of the unit 1 reactor vessel and that the Southwest Research Institute report gm basically substantiates this conclusion. It is our position that no repair ( ) is necessary for this indication. Finally, we do not recommend further C/ - faulted condition evaluation by Southwest Research Institute.
,, EN DES has the necessary expertise and component design information inhouse to perform work of this nature in a timely and responsive manner. This would be especially important for an operating plant. Consequently, we strongly recommend that you consider using EN DES to perform future evalua-tions rather than having the work performed outside of TVA.
EN DES engineers will be available upon request to provide support for your presentation of the information to the NRC.
..- a
/ Roy H. Dunham DRP:EAM:MGR Attachments N cc: R. G. Domer, W90224 C-K N" d "*"
MEDS, E4B37 C-K ON N WD' H. H. Mull, E7824 C-K / D. R. Patterson, W10Cl26 C-K - R. M. Pierce, 204 GB-K (2) G. G. Stack, Sequoyah CONST (4) N- !' 'y
-(N E. F. Thomas, 550 CST 2-C g' 600 / .l .( } M.L
- e &f)1.53
-cfC-
,(C z.,
i r N y
- j[:i (l '2 l $13
- 97026/1 Buy U.S. Savings Bonds Regularly on the Payroll Savings Plan
. . . . . _ -}}