ML20040F702
| ML20040F702 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Limerick |
| Issue date: | 06/29/1971 |
| From: | Case E US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Morris P US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML20040F238 | List: |
| References | |
| FOIA-81-385 NUDOCS 8202100151 | |
| Download: ML20040F702 (14) | |
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i Peter A. Morris, Director, Division of Reactor Licensing 1
LIMERICK CEN STATION UNITS 1 AND 2 DOCKET NOS 50-352 AND 50-353
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The interin evaluation for the subject plant which was prepared by the DRS Mechanical Engineering Branch, dated March 27, 1971, has been revised i
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to reflect significant changes submitted in Amendments through Ro. 7 and proposed by the applicant in recent discussions. New report sections l
are enclosed as direct replacements for those sections of the March 27, 1971 l
evaluation which have undergane major revisions. Instructions for making minor revisions are on: separate pages under the appropriate section heading.
I Origial 5gned By l
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df.T # 50 - &Q E.G.Can s Edson G. Case, Director Division of Reactor Standards
Enclosure:
Final Evaluation - Mechanical for Limerick 1 and 2
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S. Hanauer, DR i
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R. Boyd, DEL j
R. DeYoung, DEL D. Skovholt, DRL I
'j R. Maccary, DRS
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D. Knuth, DRS t
D. Lange, DRS C. Lear, DR'.
R. Kirkwood, DRS i
K. Wichman, DRS i
J. Knight, DRS i
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DRS:MEB MEB omer>
Id" J. Knig D. Lange a
E. G. Case m,
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Form AEC Ste (Rev.9-53) AECM 0240 e u a sovre wee me wre s onct sovo-.aov 7ss
("202100151 811019 F
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Final Evaluation-Mechanical LIMIRlCK GENERATING STATION UNI'TS 1 AND 2 DOCKET NOS. 50-352 AND 50-353
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4n System Ouality Group Classifications IFP-
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The applicant has applied a system of code classification groups to.
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.a those pressure containing components which are part of thei reactor
, :s coolant pressure boundary and other fluid systems important to safety.
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These classification groups generally correspond to the tentative code classification groups A, B, C and D developed by the regulatory staff.
The codes and standards applicable to the compenents in each of the classification Groups are identified in Table CS-1.
i k'e and the applicant are in general agreement on the application of the code classification groups for the reactor coolant pressure boundary i.
ETSi n rr and other fluid systems important to safety.
For those systems, portions W
i of systems, or components where the applicant's classification' grouping differs from ours, the applicant has upgraded his classifications by proviciens for a quality level subst antially equivalent to that' normally required by.the applicable staff classification code groups. Clarification 1
~ of these upgraded classifications is documented in amendment 6, supplement 5 to the PSAR.
For pressure-retaining cast parts of pumps and valves in systems which are classified as Groups A and B ve believe that the non-destructive examination requirements in the ASME Pump and Valve Code for lines over.
2 inch up to and including 4 inch are inadequate. To be acceptable, e e
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D will require volueetric. examination (radiography or ultrasonic testing) of these pressure-retaining cast parts in lines over 2 inch up to an including 4 inch. Where size or configuration does not permit effective.
volumetric examination, surface exa=ination (magnetic particle, or v:;;
liquid penetrant testing) may be substituted.
Examination procedures and acceptance standards should conform with those specified in the ASME p<.
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Code for Pumps and Valves. We recommend this requirement be imposed IV on the applicant as the basis for the acceptance of the system
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l classification in Groups A and B.
i T.ie app icant has agreed to supp y piping and instrumentation diagrams l
l of the reactor coolant pressure boundary and those other fluid cystems important to safety.
In the interim, Group Classification Diagrams, Figures A.2.1 and A.2.2 are acceptable.
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Wefind$that the system quality group classification an specified by the applicant and supplemented by provisions for upgrading quality levels, and additiensi nendestructive examination requirement discussed above are acceptabic for this facility.
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..p TABLE CS-1 Summary of Codec and Standards for Components of-Water-Cooled Nuclear Power Units 2/12/71,,
I Code Classifications I
3 Component Group A Group B Group C Group D I
Prassure ASME Boiler and Pressure-ASME Boiler and Pressure ASME Boiler and Pressure ASME Boiler and Pressure I
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Vessels Vessel Code,Section III, Vessel Code,Section III, Vessel Code,Section VIII, Vessel Code, Scction VIII, Class A Class C Div.ision 1 Division 1orEquivaley 4
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i 0-15 P'sig API-620 with2the NDT API-620 with the NDT API-620 or Equivalent i
Storage Examination Eequiremchts Examination Requirement's Tenks in Tabic NST.-1, Class 2 in Table NST-1, Class 3 Atmospheric Applicable Storage Tank Applicable Storage Tank API-650, AWAD100 or Codes such as API-650, codes such as API-650 ANSI B 96.1 or Equivalent Storage Tenks AWAD100 or ANSI B 96.1 AWAD100 or ANSI B 96.1 with the NDT E>. amination with the NDT Examination Requirements in Table Requirements in Table NST-1, Class 2 NST-1, Class 3 Piping ANSI B 31.7, Class 1 ANSI B 31.7, Class II ANS1 B 31.7, Class III ANSI B 31.1.0 or Equiva-Ient Pumps and Draft !+/6 Code for Pumps Draft ASME Code for Pumps Draft ASME Code for Pumps Valves - ANSI B 31.1.0 or Velvec and V n nos ' lass I.
See and Valves Class II.
See and Valves Class III Equivalent p
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Foott i a 'i Footnote (a)
Pumps' - Draf t ASME Ci, f
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$[1 pressure-retaining cast parts shall be radiographed (or ultrasonically testdd to equivalent standards).
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FOOTNOTE:
(a)
Rhere size or configuration does not permit effective volumetric examination, magnetic particle or liquid penetrant examination may be substituted.
Examination proceduces and acceptance standards shall be at 1 cast equitalent to those specified in the applicabic class in the code.
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Reactor Coolant Pressure Boundary See Interim Evaluation-Mechanical dated March 27, 1970.
The text of I;
f this Section remains unchanged.
Remove parentheses around tentative l,
conclusions and remove the last paragraph, in brackets, completely, f
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N-o Reactor Internals - Desien The reactor internal structures are classified as Clas's I (Seismic)
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components and will be designed to withstand normal design loads, anticipated transients and the Operational
- Basis Earthquake within the stress limit criteria of Article 4,Section III of the ASME Boiler and c)Ul lcQ Pressure Vessel Code.
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Under the loads calculated to result from the Design Basis Accident, sh:s the Design Basis Earthquake and the combination of these postulated V ,
events Limerick reactor internal structures will be designed to the requirements of the G. E. Nuclear Steam System (NSS) Loading Criteria, Appendix C of the PSAR.
The NSS loading criteria contain a number of design stress, deformation, f at'igue and buckling limits some of which
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may exceed applicable limits as specified in the nuclear component codes.
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Ilowever, during recent discussions representatives of GE have stated that only those limits of Appendix C which are consistent with the limits specified in the nuclear component codes wil.' be employed; on this basis we find the design criteria for the Limerick reactor internals acceptable.
Documentation of verbal agreements is expected in the near future.
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Other Class I (Seismic) Mechanical Fluid Systems-All Class I systems, components, and equipment outside' of the reactor i
coolant pressure 1.undary will be designed to sustain normal loads, anticipated transients and the Operational Basis Earthquake within the I
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appropriate code allowable stress limits and the Design Basis Earthquake l.
I within stress limits which are comparable to those associated with the emergency operating condition category (within the yield strength of the material for membrane stresses).
k'e consider that these stress criteria provide an adequate margin of safety for Class I systems and components which may be subjected to scismic loadings.
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s Seismic Input See Interin: Evaluation Mechanical dated March 27, 1970, ne text of this Section remains unchanged with the exception of the last paragraph, n,
.t in parentheses, which should be removed.
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Seismic System D> manic Analyses See Interi:s Evaluation-Mechanical dated March 27, 1970. The text of this Section remains unchanged with the exception that references to natural frequencies greater than 20 Hz should be changed to 30 Hz.
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3 SEISMIC OUALITY ASSURANCE
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The quality assurance requirements for Class I (seismic) structures, fi F I:-
cystems, and components are specifically stated in Supplement No. I to
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the application for. license. We believe that these quality assurance
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t Class I for design, comply with the requirements of Appendix B,
" Quality Assurance Criteria for Nuclear Power Plants" of 10 CFR 50.
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Reactor Internals - Dynamic System Scisnic, Oncrat'inc and LOCA Analysis Seismic loading on the reactor internals vill be' deter =ined by means of a normal mode-time history analysis. We find this procedure acceptable.
rN Dynamic loading due to normal and upset operating conditions vill be computed by means of quasi-dynceic methods based on the measured vibration response of similar reactor designs.
We are presently reviewing a summary of BWR vibration test histories cnd correlations of vibration test data with design predictions submitted as Amendment 19 (Proprietary) to the Quad Cities docket and referenced in Supplement 5 to Limerick.
A request for the additional information required for us to complete our review of Quad Cities Amendnent 19 has been sent to the General Electric Company; their reply is expected socn.
Final acceptance of i
the procedures proposed for determining the dynamic loading due to
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normal and upset operating conditions for the Limerick reactor internals
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will be withheld until we c.mplete our review of Quad Cities Amendment 19.
We expect to complete this review early in the post-construction permit period for Limerick.
If necessary, confirmation of the procedures 3
employed to determine the normal and upset condition design dynamic loadings for the Limerick reactor internals structures can be obtained during the pre-operational vibration test program for this plant by the use of additional instrumentation and analyses beyond that currently contemplated.
(See Vibration Control Section.)
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Design loadings,for the postulated Loss-of-Coolant Accident (LOCA) will be determined by computing the response of each structural member to the calculated peak pressure differential applied as an equivalent static f*,. ;:
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In response to our concerns regarding the validity of this static
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analysis the applicant has stated (Supplement 5) that the natural frequency of the Bk'R internal structures is more than ten times the calculated frequency of the LOCA loads thus assuring no significant dynamic ampli-I j "'. ' -
On the basis of the information submitted by the applicant fication.
we find this analytical method acceptable.
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As part of previous design efforts by General Electric, extensive vibration T
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test' programs have been conducted at several large BWR plants. The nuclear y,",.2 wk steam supply system supplier, General Electric, has stated that other plants g[f."
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with similar internals supports will be vibration tested prior to the com-pletion of Limerick and that these tests will identify the potential prob 1cm k
l%. 'M areas, if any, for the Limerick plant.
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We believe that the following should be included in the required preoperational Y
test program for the Limerick plant to confirm that the vibrational character-istics are not unlike the prototype plants:
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The reactor internals important to safety should be subjected during u
the preoperational functional testing program to all significant flow
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Af ter the reactor internals have been subjected to the significant flow modes of normal reactor operation, the reactor internals should be
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removed and visual or surface examinations of reactor internals should
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be conducted to detect any evidence of excessive vibrations, and the
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examinations should be conducted at all major load-bearing structural
.M 2-elements whose f ailure could adversely affect structural integrity of the reactor internals, and at all areas of lateral, vertical and
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In lieu of tihe visual or surface examinationa of the reactor internals
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preoperational functional testing program by using sufficient and appro-Wpi Xfn-priate vibration-measuring instrumentation to detect the predominant gf~
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vibratory responses observed in the prototype reactor.
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4 A summary of the inspection of 2. above, or the results from the vi-
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bration test program of 3. above, should be the subject of a report, M
submitted to the Commission within 3 months af ter completion of the f(-
inspection and/or tests.
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