ML20009A001
| ML20009A001 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 06/30/1981 |
| From: | Tramm T COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8107070352 | |
| Download: ML20009A001 (45) | |
Text
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1 / N Commonwealth Edison
/ ) One First N;tional Plazi, Chicigo, lilinois 7 Address Reply to: Post Office Box 767
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\ ,/ Chicago, Illinois 60690
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June 30, 1981 p7
- JUL 0 61981 ** g u.h ***w*is"o"*n
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Mr. Harold R. Denton, 9trector U Office of Nuclear Reactor Regulation 4 g, U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Byron Station Units 1 and 2 Braidwood Station Units 1 and 2 FSAR Review Meeting NRC Docket Nos. 50-454/455/456/457
Dear Mr. Denton:
Enclosed are minutes of the meeting held on May 11, 12, and 13, 1981, in Bethesda to review portions of the draft Byron /Braidwood SER prepared by the Mechanical Engineering Branch.
Resolution of open items from the draft SER involved discussion with the staff, presentation of technical information, and/or agreement to provide FSAR revisions. A summary of the discussion, presentations, and resolution of each open item is included as Attachment A. In addition, a list of FSAR revisions and/cr commitments to provide additional information is included as Attachment B. A copy of the meeting agenda defining the subject matter of the open items is included as Attachment C.
As a result of this meeting, all but nine of the open items in the SER draft have been resolved. The nine open items are described below:
N4 -
Westinghouse is to provide a comparison to the NRC for i their review.
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! B1 -
This item involves seismic re-evaluation of the l station, which has not yet be_n completed.
N17 - This item is a disagreement between Westinghouse and the NRC.
N15, B16, B18, N23, N25 - All of these items are related to asymetric loading. The analysis is scheduled for completion in fall, 1981.
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=The programs for preservice testing of pumps and valves and for_ inservice testing of pumps and valves will be submitted to the NRC in early 1982 and September 1982,-respectively.
Please address corrections to these minutes and further questions regarding this matter to my office.
Very truly yours, k FAN = =
T. R. Tramm Nuclear Licensing Administrator Pressurized Water Reactors Enclosure 2236N
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O Attachment A Minutes of NRC MEB SER Meeting Byron /Braidwood May 11-13, 1981 e
w s Item N1-(SER Page 1, Section"3.6.2, Paragraph 3)
A comparison of the Byron /Braidwood design transients and the WCAP-8082 transients was presented. It was pointed out that the Byron /Braidwood transients include all of the WCAP-8082 transients plus a number of additional transients.
Use of the reference analysis was also explained, indicating that the reference analysis utilized the same methods and criteria as WCAP-8082 except that the Byron /Braidwood tran-sients were used rather than the WCAP-8082 transients. In this way, the actual plant thermal and OBE moments calculated for Byron /Braidwood have been shown to be less than the moments of the reference analysis. Thus, the number of breaks and their locations determined from the reference analysis are the same as those given in WCAre8082, and the stresses and the usage factors calculated in the reference analysis are applicable to Byron /Braidwood. It was concluded that the results of the reference analysis are consistant with WCAP-8082, i.e, no addi-tional breaks need be postulated as a result of the Byron /
Braidwood transients.
FSAR-Section 3.6.5 will be modified to indicate that the refer-ence a'nalysis specifically nsiders the same transients as Byron /Braidwood and to confirm that the break locations are the same as indicated in WCAP-8082. With these changes and the meeting discussion, this item is resolved.
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Item N2 (SER Page 2, Section 3.6.2, Paragraph 4)
It was explained that a detailed fatigue analysis was performed for the loop isolation valve-to-pipe welds using the Byron /
Braidwood transients. It was also indicated that the Byron /
Braidwood reactor coolant loop (RCL) model included the loop isolation valves in the fatigue analysis to determine the actual plant moments. The fatigue results showed that the Equation (10) stresses were less than 2.4 S,a'nd. that the usage factors were less than 0.01. Therefore, it was not necessary to postulate new breaks in addition to those defined by WCAP-8082.
Based on the above discussion, this item was resolved.
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h Item'N3 (SER Page'2, Section 3.6.2, Paragraph-5)
The responsibilities of S&L and Westinghouse concerning the interface between.the design of the primary equipment supports and the RCL stress analysis were discussed. Also, the design flow process between S&L and Westinghouse, assuring that RCL analysis ~was consistant with the final support design, was explained. It was noted that actual calculated support stiff-nesses'were included in the RCL model and that the resulting loop loads and displacements were included in the final support design.
Comparisons between the S&L - designed supports for Byron /
Braidwood and typical Westinghouse - designed supports were presented showing the similarity between the two designs. It was further indicated that the Byron /Braidwood-loop piping, loop layout, and primary _ equipment are essentially the same as used in'the WCAP-8082 and the reference analysis.
FSAR 3.6.5 will be revised to reference FSAR 3.9.3.4.4.1 which
" discusses the interface between S&L and Westinghouse. This change and the above discussion constitute resolution of this item.
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D Item N4 (SER Page 2, Section 3.6.2, Paragraph 6)
At the time the Byron /Braidwood loop piping analyses were per-formed there were no NRC approved computer codes. SATAN IV was in use at that time. Since then the NRC has approved the MULTIFLEX Code for piping and reactor internals analyses. For l Byron /Braidwood, the MULTIFLEX Code is used for the internals.
l For the piping analyses, Westinghouse has done comparisons j between MULTIFLEX and SATAN IV and has shown that comparable l I results-are obtained. Westinghouse will provide these E comparisons and the SATAN IV modeling scheme to the NRC for their review.
I The comparison.is scheduled to be completed by the end of July, 1981. Submittal of this comparison will constitute a f
7 resolution for this item.
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Item B1 (SER page 5, Section 3.7.3, paragraph 6)
A meeting between the applicant and the NRC was held on May 13, 1981-to-discuss the adequacy of seismic margins for the Byron /
Braidwood plants. Commonwealth Edison will complete its seismic margin reassessment as quickly as possible and submit it as the l . response to Question 130.06. The concerns expressed by the MEB will be covered in that response.
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b Item B2/N5-(SER Page-5, Section 3.7.3, Paragraph 7)
Sargent and Lundy indicated that they require the valve vendor to seismically qualify the valves to certain "g" values and that they check the piping analysis to verify that these values are conservative.
Sargent and Lundy was asked-if they have reviewed and rejected
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any seismic analyses for valves. They answered affirmatively and provided an example for NRC review.
Westinghouse discussed the seismic analysis methods it used to qualify equipment and piping for Byron /Braidwood. Equipment with more than one mode below 33 Hz and all piping systems are qualified using the response spectrum analysis technique. All L other equipment is qualified using static analysis method.
Westinghouse agreed to provide changes to the FSAR that would sunmarize the discussions of analysis technique.
j A question was asked concerning the modeling of valves in 4 piping systemn. Westinghouse stated that rigid valves are modeled as a mass on an extended structure. Non-rigid valves i are modeled as flexible in the piping analysis, Westinghouse requirements for rigid valves and its review of valve vendor y reports were discussed. Westinghouse will provide a statement b of its valve modeling techniques in the FSAR.
Based on these discussions and the changes to be made to the I FSAR these items are resolved.
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4 Item-B3/N6 (SER Page 5, Section 3.7.3, Paragraph 8)
Sargent and Lundy stated that all seismic restraints in piping systems within Sargent and Lundy's scope of work are considered to be infinitely rigid for analytical purposes.
Westinghouse stated that for all piping systems they are evaluating for Byron, the calculated support stiffness is included in the analysis. Westinghouse will provide a state-ment to this effect'in the FSAR to resolve this issue.
For Byron, Westinghouse is responsible for the analysis of all piping inside containment, and for the safety injection system, residual heat removal system, chemical and volume control sys-tem, and containment spray system in the auxiliary building.
Based on this discussion and the FSAR changes which will be made, this item is closed.
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Item B4 (SER page 6, Section 3.7.3, paragraph 4)
A discussion of the extent of use of the static load method took place. It became apparent that Section 3.7.3.5 of the
'fSAR is unclear. . We propose to revise it to read:
No static load method is utilized in the seismic analysis of piping systems.
However, in the seismic analysis of equip-ment,_the equivalent static load method is used if the equip.: lent is not rigid and a dynamic analysis is not performed.
If the fundamental natural period (FNP)...
This will resolve the item.
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Item N7-(SER-Section 3.7.3, Page'6, Paragraph 9)
Westinghouse stated that all equipment with more than one mode
~below 33 Hz and all piping systems that'they are responsible for (see Item B3/N6) are analyzed with response spectrum tech-niques. All other equipment is analyzed using static analysis methods.. Westinghouse indentified the equipment with more than one mode below 33 liz as the steam generator, reactor coolant pump, pressurizer, control rod drive icechanisme,. reactor inter-nals and fuel. Westinghouse will incluca this information in the FSAR~to resolve this item.
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a-Item N8-(SER'Page 6,-Section 3.7.3,-Paragraph 10)
The method.used by Westinghouse for the combination of closely spaced modes has been accepted previously by the NRC (RESAR 41,
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RESAR 414, numerous plant dockets) as an acceptable alternative to Regulatory Guide 1.92. The NRC Mechanical Engineering Branch will notify the Structural Branch that on this basis this item is considered resolved.
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Item B5 (SER page 6, Section 3.7.3, paragraph 11)
Sargent & Lundy described buried cat I structures and
) the analysis of them, and_ agreed to provide a write-up to be added to the existing FSAR Section 3.7.3.12 de-linating stress limits used for buried piping analysis, as follows:
"Since all buried essential service water piping falls under subsection i NC of ASME B&PV code, section l((
the following. stress limits are met:
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Stresses due to sustained loads < l.0Sh
) Stresses due to occasional loads (OBE) < l.2Sh t~ Stresses due to occasional loads (SSE) < l.8Sh Stresses due to bending moments g caused by soil settlement and/or < 3.0Sc overburden pressure L For all buried concreto electrical duct runs associated with the essential service l- water. system, the design is in accordance i with ACI-318-71 requirement.
i i Based on_the-discussion, and the proposed FSAR changes, this item is' resolved.
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Item N9-(SER Section 3.7.3, Page 6, Paragraph 12)
Westinghouse explained that the damping values in Regulatory
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. Guide 1.61, i.e. two (2). percent for the OBE and four (4) per-cent for the SSE-are utilized. Based on the above explaina-tion and the addition of:this information to the PSAR, this item is resolved.
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i Item-B6/N10-(SER Page 7, Section-3.9.1, Paragraph 2)
Sargent and Lundy stated that the PIPSYS (integrated piping analysis system) computer code was bench marked against two public domain computer programs, DYNAL and NASTRAN, and was found to be acceptable. This is documented in FSAR Appendix D.8.
Westinghouse discussed the method utilized in piping analysis for lumping masses and referred to Dr. Lim's paper, "How to Lump the Masses" as the basis for this method. Dr. Lim's paper is referenced in Section 3.7 of the FSAR.
For. flexible equipment, Westinghouse utilizes multiple degree of freedom dynamic analysis models (e.g. over 200 degrees of freedom for the steam generator) to assure that a sufficient number or modes are calculated. This method is consistant with SRP 3.7.2.
Westinghouse also provided test results for review that support the validity of their modeling techniques. These results were for piping systems in two plants and compared the analysis results with test data obtained at the plant site.
Based on these discussions this item was resolved.
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Item Nil (SER Page 7,:Section 3.9.1, Paragraph 3)
-No response to'this.. item.is required for MEB review, 1
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Item N12-(SER Page8,-Section 3.9.1, Paragraphs 4 and 5)
-Westinghouse-described the test load method of qualification of i
I the reactor vessel-support pad and shoe and the elastic system c-1 analysis with inelastic component analysia used for qual-ifica-Ltion~of the reactor coolant pump support foot. Both of these methods are-used~in lieu of the Appendix F limits of the ASME Code,-Section III.
'For the test load method, Westinghouse will provided a revised section-for the FSAR that will'contain.more information on the techniques. A revised section for the FSAR will also be pro-
.vided thatl justifies'the use of the elastic system with inelas-
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tic component analysis.
Based on.this-discussion and the FSAR changes which will be made,-this item is resolved.
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Item N13-(SER'Page 8, Section 3.9.1, Paragraph 6) j: Westinghouse uses 4% of critical damping for the SSE seismic 1.-
1 analysis fo.the reactor coolant system and supports. The justification of this damping value is provided in WCAP 7921-AR, which has been reviewed and approved by the NRC. This WCAP'is' referenced in the Byron /Braidwood FSAR. Based on this discussion this~ item is resolved.
b Item B7 (SER page 10, Section 3.9.2.1) h The preservice inspection program will include the visual b examination of all hydraulic and mechanical snubbers in-stalled on safety related systems. The inspection will
) verify that the snubbers are installed correctly, and are
[ undamaged.
Al list of all snubbers (both hydraulic and mechanical) on safety related systems will be developed. Documentation will be provided to record the inspections conducted on each snubber. The documented inspection will be conducted no longer than 6 months prior to the preservice testing program in order to satisfy preservice testing requirements.
-For hydraulic snubbers the fluid will be verified to be at the recommended level and not leaking.
During hot functional testing, snubber thermal movements for systems whose operating temperature exceeds 2500F will be verified. The thermal monitoring program will be included .
in the test program. The thermal monitoring program consists of visual verification of snubber movements, as indicated on the snubber, from room temperature to maximum operating tem-perature. If maximum operating temperature is not attained
- during testing, the amount of movement expected will be cal-culated by multiplying the movement indicated on the snubber by the ratio of the temperature rise to the test temperature (ATm /ATt ). .If snubber movement differs from the expected movement by more than 1/8 inch, as assessment will be made to verify that the snubber will satisfy its design function for the design load, i Based on this discussion and commitment, this-item is resolved.
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Item-B8/N14'(SEE-Page'11,-Section 3.9.2.2)
-Review of.the dynami c quali fi cation of mechanical equipment will be covered _by the Equipment Qualification Branch. No k' response is required for the MEB review.
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Items N15 and N25 (SER' Section 3. 9. :. 3, - Page 12, Paragraph-2,
- f. and-SER'Section-3.9.5,-Page-20; Paragraph 4)
Westinghouse stated that for reactor internals, the LOCA evaluation has always considered the asymmetric loading "inside" the reactor vessel. (See FSAR Section 3.9.7, Reference #7: 'G. J. Bohm and J. P. LaFaille "R'eactor Internals Response Under a Blowdown Accident" First Int'l Conference on
. Structural Mechanics in Reactor Technology, Berlin, September 20-24, 1971.)
FSAR Section 3.9.2.5, " Dynamic System Analysis of the Reactor l
Internals for Faulted Conditions", describes in detail the l
l reactor internals blowdown analysis and also refers to
" Reference #7" noted above. Question 110.62 concerns the asymmetric loads "outside" the reactor vessel. These loads due to cavity pressure along with loop loads were shown to be insignificant with respect or to the reactor internals.
'Any additional information required on the internals concerning asymmetric loadings will be submitted with the response to Question 110.62. (Also see the response to item N23.)
, The abov.e discussion along with the response to Question 110.62 will resolve this item.
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Item B10 - (SER page 13, Section 3.9.3.1, paragraph 1)
The loading combinations for ASME Class I component supports were presented from FSAR Table 3.9-2. The analytical method I
used to combine the loads is by algebraic summation. The i signs of OBE and SSE are chosen to maximize the magnitude of the total load.
} A summary of the faulted loads which control the design and
$ the resulting largest stress as a percentage of allowable per FSAR Section 3.9.3.4.5 was presented for each component sup-port. In all cases, the actual stresses are less than or I
j equal to the allowable stresses. The staff requested that i 'he t highest stressed member for each component support be l
identified on the FSAR figures with a corresponding qualita-tive discussion of stress state and nominal margin to failure for each critien1 member. Commonwealth Edison Company agreed to include this requested information in a future FSAR amendment.-
The Staff questioned which allowables were used for bolt mat-1' f erials. Sargent & Lundy stated the faulted allowables were l
- obtained by using factors calculated in accordance with Appen-
- dix F of ASME Section III, Article F1370 times the normal Ii allowables. The Staff requested that the allowables be pro-vided in tabular form to ensure they are below 0.90 of yield strength.
The ratio of the faulted allowables to the yield strength was presented for the high strength bolts used in the Class I component supports. The factors calculated in accordance with ASME Section'III, Appendix F, result in faulted condition al-lowables which are less than 90% of yield strength (Sy) in tension and/or shear. Commonwealth Edison Company agreed to includd this information in a future FSAR amendment.
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- . i The above discussion concerns Class 1 component supports. -In addition,. tables'in FSAR section 3.9 will be revised to reflect the load combination methodology used for Class 2 and 3 piping, equipment, piping supports and equipment supports.
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Based on the discussions and submittal ~of the FSAR revisions,-this item is resolved. i I
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l Item N16-(SER'Section 3.9.3.1, Page 13, Paragraph 1)
The methodology of load combinations and applicable stress limits for Class 1, 2, and 3 equipment was discussed. Several necessary modifications to the FSAR were identified:
- Table 3.9-3 should be clarified indicating what stress limits. apply for each of.the operating condition classifi-cations.
- The FSAR should include the load combination methodology applicable to the-loads identified in Tables 3.9-2 (Class
- 1) and 3.9-5 (Class 2 and 3).
- The source of the stress limits for Class 2 and 3 equipment should be identified.
Class 2 and 3 equipment supports stress limits should be provided.
2 Additional information on load combination for reactor inter-rials is presented in items N18 and N26.
Based On these FSAR changes and the meeting diccussion, this item is' resolved.
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4 h1 4 -. It$m Bll '(SER page 13, Section 3.9.3.1, paragraph 1)
Sargent &.Lundy described their requirements for functional-capability for essential piping. All piping systems that
'are designated essential and are within Sargent & Lundy scope of work, are evaluated using-the functional capability
. criteria ~ outlined in GE's Topical Report #NEDO-21985, September, 1978, which was evaluated and approved for use
. by the Mechanical Engineering Branch of NRC.
All essential Byron /Braidwood piping will fall within the
} following range 50<Do/ t<100, or D
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0 For Sargent & Lundy scope piping, this item is closed.
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Item N17 (SER Page 13, Section 3.9.3.2, Paragraph 1) f-t For Class 2 and 3 austenitic steel bends and elbows, Westinghouse
- and the NRC 'could not reach a mutually acceptable resolution of? stress criteria for' functional capability. This item will t.
- remain open.
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Westinghouse has formed a task force to resolve this issue.
The resolution is expected within approximately two months and
[ will bersubmitted to the NEC for their review.
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t Item N18 and N26 (SER Section 3.9.3.1, Page 14, Paragraph 2, and SER Section 3.9.5, Page 20, Paragraph 5)
A discussion of rnactor internals stresses and deformations was presented. It was indicated that although the Byron /Braidwood
. internals are not contractually required to meet ASME Code requirements, essentially the design and fabrication require-ments of Section NG of the ASME Code have been satisfied.
Exceptions to code requirements discussed at the meeting were no code stamp'and no plant-specific stress report.
l-Additionally, all stresses and deformation are below j allowable limits. Westinghouse agreed to provide a statement I in the FSAR indicathig the differences between the Westinghouse criteria used for the Byron /Braidwood internals and the ASME 1 .
I Code requirements. Additionally, Westinghouse agreed to l
provide a statement in the FSAR relative to the acceptability
.L of stresses and deformations for the Byron /Braidwood internals.
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l Although new specific issues have been identified relative to i Regulatory Guide 1.20, Westinghouse stated that Byron /Braidwood references Indian Point Unit 2 and Trojan as the prototype plants for internals vibration monitoring. It should also be noted that the Indian Point tests were conducted both with and without fuel assemblies in the core at the time of the L
. vibration monitoring. The vibration levels under actual operating conditions (i.e. with fuel in the core) are typically lower than those obse'rved without the fue'l in place.
Westinghouse further stated that when appropriate, e.g. simple l
. beam analysis, LOCA and SSE loads are combined on a reactor I
y internals structural component basis per the SRSS method, and the resultant stress intensities calculated. For more complex j structural geometries (e.g. core barrel shell) the stress
) components due to LOCA and SSE are combined either by absolute 4
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- sum or by SRSS, preserving the appropriate signs. These stress components are used to determine the stress intensity for the
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I structural component. For the LOCA, the maximum stresses from the time history response are used. Since the seismic stresses are : calculated using response spectrum techniques, the responses are unsigned; therefore, when the LOCA and SSE stresses are f combined, the most unfavorable sign convention for the SSE is assumed.
6 Based'upon the above, this item will be resolved upon revision of the FSAR.
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o Item-N19'(SER Section-3.9.3.1, Page 14, Paragraph 2)
Regulatory Guide 1.121 " Bases for Plugging Degraded Steam Gen-l erator. Tubes" Q1 -
NRC requires that a margin of 3 against tube burst as l
outlined in this R.G. must be maintained.
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Westinghouse uses a margin of 2 against tube failure.
The definition of tube failure is plastic deformation
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of a crack to the extent that the crack opens to a l.
non-parallel elliptical configuration. NRC defines tube failure as tube burst. Since Westinghouse uses a different definition of tube failure we use a smaller margin. The position on R.G. 1.121 will be expanded to include this information.
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NRC would like the position to reference the tech spec requirement to plug all degraded tubes that have been reduced in wall thickness by 40% of the nominal tube wall thickness.
A2 - The 40%'T.S. limit is a reference limit for Westing-
. house steam gener tors.- R.G. 1.121 analyses have not been completed for model D4 and D5 steam generators used in Byron /Braidwood. These analyses will be com-pleted prior to first refueling and at that time the T.S. limits will be re-evaluated and this information I
can then be included in the R.G. 1.121 position, if necessary.
Based on the above, this item is resolved.
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Item B12/N20 (SER Section'3.9.3.2,~Page 15, Paragraph 2)
Review of pump and valve operability will be covered by the
' Equipment-Qualification Branch.- No response is reqLired for the MEB' review.
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-Items N21 and N22 (SER Section 3.9.3.3, Page 15, Paragraphs 2.
and 4)
Westinghouse discussed the analysis methods used to evaluate l
the: pressurizer safety and relief valve discharge piping.
Additional FSAR information will be provided which describes i
the hydraulic and structural analysis methods, loading combina-tions, inclusion of the effects of water. slugs from the loop
-seals, and valve opening sequence. Also to be provided are the makes and types of valves used and their mounting arrangement.
[ With the submittal of this FSAR write-up, this item will be resolved.
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j '*' Item B14 (SER page 16, Section 3.9.3.3, paragraph 3)
Sargent & Lundy discussed the use of. design load factors k
(DLP) in'the design-of-relief valves. This discussion revealed that the main steam relief valves utilize a DLF less than 2.0. The design basis for utilit_.3 a DLF'less than 2.'0 was a parametric study based on a dynamic analysis as allowed for in Code Case 1569. A parametric study done for Zion was shown to the NRC to illustrate the basis for
. utilizing the lower DLF. All other relief valves utilize a DLP of 2.0.
FSAR Appendix A, Al.67 will be revised to reflect this
\' ' discussion. This discussion and the resulting FSAR change will close this item.
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Item B16 (SER page 17, Section 3.9.3.4, paragraph 2)
An assessment of the NSSS component supports for the faulted
- condition against 67% of critical buckling was presented. The 1
t stresses were calculated on the basis of an SRSS combination of LOCA and SSE for the faulted condition. The stresses cal-1 L culated on this basis are less than 0.67 times the critical I buckling stress in all cases except one. The steam generator h lower lateral support.has one member which is stressed to 0.73 of critical as calculated from ASME Code Appendix XVII
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interaction equation.
A figure of the steam generator lower lateral support was presented and the member stressed to 0.73 of critical was identified. The stress in this member of the steam generator lower lateral support is primarily due to a jet impingement load resulting in weak axis bending. The ultimate capacity of this wide flange type member which has acceptable width thick-ness ratios for its flanges would be governed by the plastic capacity of the section and, therefore, the recommendation of L
the Regulatory Guide 1.124 and 1.130 on critical buckling would not directly apply. The staff agreed with this justification.
Commonwealth Edison Company acknowledged that the response to Question'110.50 and Appendix A will be revised to point out this exception to'the regulatory guides.
t The effects of asymmetric pressurization loads will be assessed I
in response to FSAR Questions 110.14 and 110.62. This item will remain open:pending-submittal of the FSAR revisions and responses to Questions 110.14 and 110.62.
Item-B17 (SER page 17, Section 3.9.3.4, paragraph 3)
Sargent & Lundy presented a table of operating temperatures
-for each Class I component support. In linear clastic analy-i sis,.the effect of_ temperature is accounted for by a reduc-l t tion in yield stress -(Sy) and ultimate tensile stress (Su)
I as specified in Subsection NP-3229, Appendix XVII Article 1121, and Appendix F Section 1370 (a) of the ASME Code, Sec-tion III, Division I, Summer 1975 Edition.
The Staff requested the source of these reduced values for yeild stress and ultimate tensile stress. The reduction in yield and ultimate stress are in accordance with ASME Section III, Appendix I or Code Case 1644 Commonwealth Edison Company agreed to include this information in a future FSAR amendment.
Based on the discussion and the FSAR changes to be made, this item is resolved.
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< Item B18 (SER page 17, Section 3.9.3.4, paragraph 4)
Sargent & Lundy presented a discussion of their analysis of NSSS. component supports.
- Sargent & Lundy has assessed NSSS component supports for asymmetric pressure loads provided by Westinghouse for breaks in the hot, cold and cross-over legs. We find J r- that these supports are within the limits described in FSAR paragraph 3.9.3.4.5.
l l We are in the' process of completing the assessment for
[ asymmetric pressure loads and will transmit our results
) in response to FSAR Questions 110.14 and.110.62.
j This. item will remain open pending completion of the ana-i lysis described above, and documentation of the results in the FSAR.
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Item N23 (SER Section:3.9.3.4, Page 17, Paragraph 4)
- - A general discussion describing how asymmetric loads (Question r 110.62) are included in the analysis of reactor coolant system piping and components was presented. The FSAR changes neces-sary to respond to Question 110.62 will be provided. Upon I
l submittal of this information, this item is resolved. (Also
- t. .see items N15 and N25 for. reactor internals.)
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Item-N24-(SER Section 3.9.4,'Page 18, Paragraph 2)
Evaluation of the loading combination and stress limits for pressure bcundary components of the CRDMs, and the design criteria and loading combinations used for non-pressure coundary components of the CRDMs were satisfactorily addressed based upon review of the Westinghouse design specification, discussion of testing performed by West'nghouse and Westing-house licensees, and the contents of the,FSAR.
It was agreed that the information provided in the PSAR is sufficient provided a statement is added to the FSAR which (1) statcs that FSAR TC,la 3.9.2 is applicable to Class 1 compo-nents of the CRDM and (2) discusses the operational testing and experience gained by Westinghouse & Westinghouse Licensees.
Upon revision to the FSAR, this item will be resolved.
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e Item N27 (SER-Section 3.9.5, Page 20, Patagraph 6) i
(- Test results were presented by Westinghouse indi ating that for strain, ranges below 1.8%, the irradiated material has higher
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j fatigue allowable (allowable stress at given number of cycles) l f than the unirradiated material. The .ests performed included both. laboratory-irradiated material and irradiated material
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). from operating plants. It was noted that the strain levels in
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the internals structures are less than 1.8%.
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Based upon the information presented, this item was resolved, r
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-Itam C1 (SER page 22, Sectio'n 3.9.6)
The preservice testing of pumps and valves is performed during i system pre-operational testing. The prograui plan will be sub-mitted with the other portions of the preservice inspection i plan. The complete plan is expected to be available in early h 1982.
Commonwealth Edison will submit its program for inservice testing of pumps and valves as requested by Question 110,64.
This program will include valves between the reactor coolant j system boundary and low pressure systems that penetrate the h containment.
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[ It is anticipated that the program will be submitted for Staff s review in September, 1982.
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., Item C2 f During preoperational testing normal operating modes will be observed for vibration. Engineers familar with the subject
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) piping will visually inspect the lines to determine the accept-ability of the steady state vibrations. Abnormal vibration will I
i be noted. If piping systsm vibration is judged excessive, l
t corrective action will be either:
) 1. The cause of the excessive vibration will be eliminated.
) 2. The support system will be modified to reduce the l vibration to acceptable limits.
l 3. The piping will be monitored by instrumentation at
(- locations which appear to be excessive to demonstrate j that the measured pipe deflections when converted to r stress will not exceed 50% of the material endurance E
l . limit selected from the value at 10' cycles from the
} curves of Appendix I-9.0 of Section III of the ASME l Code.
Based on the discussion and the commitment stated above, this item is closed.
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FSAR Revisions Required Item : Info, Needed FSAR Section Impacted j
N1 3.6.5 .
N3 3.6.5 l N4 SATAN IV vx. MULTIFLEX N5 3.7.3 N6- 3.7.3 B4 3.7.3 N7 3.7.3 B5 3.7.3 N9 3.7.3 N12 3.9.1 N15, N25 Answer 0110.62
,B10 3.9.3 i N16 3.9.3 L Bil/N17 Open Item on Pipe bends and elbows l N18, N26 3.9.3, 3.9.5 N19 Appendix A Reg. Guide 1.121 response and analysis needed N21, N22 3.9.3 B14 Appendix A, Reg. Guide 1.67
- B16 Answer Q110.14, 3.9.3.4.5, Appendix A Q110.62' B17 3.9.3.4 B18 Answer Q110.14, Q110.62 N23 Answer 0110.62 H
N24 3.9.4 l:
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MEB SER Review Meeting Agenda It' ens
. Monday Afternoon SER Reference Item Page Paragraph -
Description Response N1 1 3.6.2 (3) Justify break locations from Bob Kelly WCAP-8082A Brad Maurer N2 2 3.6.2 (4) Justify use of WCAP-8082A with Bob Kelly loop isolation valves Brad Maurer N3~ 2 3.6.2 (5) Describe Westinghouse / BOP Bob Kelly interface for component supports Brad Maurer N4 2 3.6.2 (6) Justify use of SATAN-IV Bob Kelly Brad Maurer B1 5 3.7.3 (6) Demonstrate adequacy of seismic J. T. Westermier margins (Deconvolution)
B2/N5 5 3.7.3 (7) Provide natural frequencies for B2-R. J. Netzel.
Seismic Cat. I structures NS-Bob Kelly Brad Maurer
- B3/N6 5 3.7.3 (8) Provide discussion of modeling B3-A. A. Deguermendjian for pipe supports and snubbers N6-Bob Kelly.
Brad Maurer B4/N7 6 3.7.3 (9) Identify equipment for which B4-K. L. Adlon the natural period is not known N7.-Bob Kelly Brad Maurer N8 6 3.7.3 (10) Response to Q110.33 part (4) is Bob Kelly not satisfactory Brad Maurer B5 6 3.7.3 (11) Identify buried structures and A. A. Deguermendjian describe seismic analysis a
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MEB SER Review Meeting Agenda Items Tuesday Morning SER Reference Itnm Page Paragraph Description Response B6/N10 7 3.9.1 (2) Provide basis for selection of B6-A. A. Deguermendjian number of masses in seismic N10-Bob Kelly model of piping and components Brad Maurer Nll 7 3.9.1 (3) Specify use of simplified elas- Bob Kelly tic-plastic methods / provide Brad Maurer tabular summary of stress ranges N12 8 3.9.1 (4&5) Justify use of ASME Section TT7 Bob Kelly Appendix F for general compo Brad Maurer nents/Specify deformation / dis-placement limits N13 8 3.9.1 (6) Justify 4% critical damping for Bob Kelly reactor coolant loop and Brad Maurer supports B7 10 3.9.2.1 Document preservice examination L. A. Bowen and pre-operational testing program for all snubbers B8/N14' 11 3.9.2.2 Dynamic Qualification of B8-K. L. Adlon mechanical equipment N14-Bob Kelly Brad Maurer Bll/N17 13 3.9.3.1 (1) Address functional capability Bll-A. A. Deguermendjian of equipment when service B N17-Bob Kelly limits are exceeded (Question Brad Maurer 110.40)
N19 14 3.9.3.1 (4) Revise response to Q110.61 Mae Wright regarding tube plugging limit B12/N20 15 3.9.3.2 Pump and valve operability B12-K. L. Adlon assurance N20-Bob Kelly Brad Maurer
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MEB SER Review Meeting Agenda Items Tuesday Afternoon SER Reference Item Page Paragraph Description Response N9 6 3.7.3 (12) Resolve concerns regarding Mark Beaumont damping valves used for reactor internals seismic analysis B9/N15 12 3.9.2.3 (2) Impact of revised seismic input B9-R. J. Netzel on safe shutdown evaluation N15-Mark Beaumont B10/N16 13 3.9.3.1 (1) Provide discussion of load com- B10-R. J. Netzel bination methods and max loads, N16-Mark Beaumont stresses, deformations N18 14 3.9.3.1 (2) Address reactor internals as Mark Beaumont requested in Q110.15 and 0110.41 N24 18 3.9.4 (2) List load combinations and Mark Beaumont stress limits for pressure boun-dary items /Specify design crit-eria and load combinations for non-pressure boundary items N25 20 3.9.5 (4) Address asymetric loading on Mark Beaumont reactor vessel intervals N26 20 3.9.5 (5) Clarify manner of compliance Mark Beaumont with ASME code N27 20 3.9.5 (6) Clarify use of unirradiated Mark Beaumont material properties for fatigue evaluation of reactor intervale m-
t O O MEB SER Review Meeting Agenda Items o -
Wednesday Morning SER Reference Item Page Paragraph -
Description Response.
B13/N21 15 3.9.3.3 (2) For safety and relief valves Bob Kelly discharging to closed system, Brad Maurer include effects of water slugs B14 16 3.9.3.3 (3) For safety valves that discharge A. A. Deguermendjian to an open system, justify use of DLF less than 2.0 B15/N22 15 3.9.3.3 (4) Identify safety and relief valves B15-A. A. Deguermend'jian locations, mounting arrangements, N22-Bob Kelly opening sequence, and load com- Brad Maurer bination and stress limits B16 17 3.9.3.4 (2) Justify use of stress limits R. J. Netzel 50% greater than normal allowables 2
Bl7 17 3.9.3.4 (3) Identify component supports R. J. Netzel which are subjected to tempera-
- tures greater than ambient B18/N23 17 3.9.3.4 (4) Respond to Q110.14 and 110.62: B18-R. J. Netzel asymetric loads on component N23-Bob Kelly supports Brad Maurer C1 22 3.9.6 (4) Submit program for preservice L. A. Bowen and inservice testing of pumps and valves
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