ML20093D088
| ML20093D088 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 10/01/1984 |
| From: | Tramm T COMMONWEALTH EDISON CO. |
| To: | Deyoung R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| References | |
| 9262N, NUDOCS 8410110079 | |
| Download: ML20093D088 (34) | |
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. Commonwealth Edison
[ One Firsi National Plur. Chicago. Illinois i , ! Address R'. ply to: Post Offics Box 767 j Chicago. Illinois 60690 October 1, 1984 R. C. DeYoung, Director Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Sub ject: Byron Generating Station Units 1 and 2 Independent Design Inspection NRC Inspection Report No. 50-454/83-32 References (a): August 16, 1984 letter from D. L. Farrar to J. G. Keppler.
(b): August 16, 1984 letter from Cordell Reed to R. C. DeYoung.
Dear Mr. DeYoung:
This letter provides additional information to address NRC questions . raised during the review of our response to the NRC's report on their Integrated Design Inspection (IDI) and to the report of the Bechtel Independent Design Review (IDR). Submittal of this information was requested in a meeting in Glen Ellyn on September 14, 1984 and in a conference call on September 21, 1984.
Attachment A to this letter contains nearly all of the information requested of Commonwealth Edison to resolve the issues related to the IDI. The item numbers were arbitrarily assigned and do not correspond to any numbering scheme previously used. The revised FSAR pages included in Attachment B will be incorporated into the FSAR in the next amendment.
There are only three items which remain to be provided to resolve IDI/IDR concerns. FSAR changes necessary to close IDR Observation 8.47 will be provided later this week. Additional information on auxiliary building ficoding will be provided to address IDI Finding 2-19 late 1 this week. A description of the methodology used to address pipe whip in the jet impingement study provided in reference (a) will also be provided later this week, m8mm8m 5 G 4
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(_ a LR. C..DeYoung October 1, 1984
.. Please address further questions regarding this. matter to uthis. o f fice. .
.One signed original and.. fifteen copies of this letter and gthe' enclosures:are provided for NRC review.
Very truly yours,.
f k<f/&W T. R. Tramm Nuclecr Licensing Administrator im.
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'.cc: J. G. Keppler - Region III
.J. Streeter J. Milhoan i.
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BECHTEL RESPONSES TO NRC FROM MEETING OF 9/14/84 I
Item 2 The statement, "there is no reason to expect this to be a concern !
elsewhere" was used frequently in close-out of observation reports. Bechtel should document the basis of why the use of this statement was appropriate for each observation report. l Bechtel Response:
Each Observation Report (OR), was analyzed and a determination was made of whether or not the OR condi^' n was limited and not expected to be a concern elsewhere. Aiso, a determination was made of whether or not a safety-significant condition existed in accordance with the Program, Plan.
When it was concluded that the condition was not expected to be a concern elsewhere, the above quoted statement was made. The basis for these statements are summarized in Table-1, which give specific reasons for making that; judgment on each such OR.
It should be noted that the purpose of Table-1 is only to explain the bases of non-concern elsewhere. It does not deal with resolution of the concern for the specific design work covered by the OR, which is covered by the Final Report.
, In making these determinations, each OR was considered from the following standpoints: (a) can it significantly impact design performance, (b) is the condition likely to be transferred, and (c) is it relevant to other safety related designs. Also, in considering impact on design performance, the criterion was consistently applied of being able to achieve safe-shutdown.
Using these standards, the IDR Team thus concluded that in the case of each OR "there is no reason to expect this situation is cause for a significant concern elsewhere.
Item 5 We agreed to discuss if any component could not perform its
. function.
Bechtel Response:
There were no cases where, to the knowledge of the IDR team, any reviewed safety-related component was found which could not perform its intended safety function.
BECHTEL RESPONSES TO NRC l
FROM MEETING OF 9/14/84 Item 5 There was an instance, documented by OR 8.24, of potential damage (Cont'd) to portions of the CCW or ESW systems piping, from postulated HELB associated jet forces determined to exist. However, in each case identified in .that 0.R., the IDR team concluded the affected portions of the systems had no safety function relative to achieving safe plant shutdown for the specific postulated breaks associated with each case. -
Another Observation Report, OR 8.38, merits discussion relative i to this item. An unanticipated consequence of the issuance of OR 8.38 was the conservative decision by Westinghouse to make a 10CFR21 report to the NRC regarding a potential overpressure condition in the CCW system caused by postulated primary coolant in-leakage to that system. Subsequent Westinghouse clarification was that the decision to make the report was based on generic system design information and not as a result of Byron specific analysis. It was the judgement'of the IDR team that, for Byron, such an overpressure condition occurrence would not be expected to cause loss of system function such that loss of capability to achieve safe shutdcwn would occur.
, Item 14 Bechtel was requested to document their present review of the S&L High Energy Line Break Report and provide a description or final
- - statement of how Observation Report 8.47 could be closed out.
Bechtel Response Regarding the HELB/MELB Confirmatory Report on jet impingement, the IDR Team has reviewed it for responsiveness to OR 8.47 and
- concludes it meets the resolution comitment. That is, the Report covers the appropriate scope, uses necessary criteria, L
clearly presents results, and makes an organized, controlled review of design for jet impingement. The IDR team did not review the Confirmatory Report for technical adequacy. However the Report does satisfy the concern for design process identified in OR 8.47. The results reported (no design changes required) evidences that an adequate design process had existed to achieve such results.
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TABLE-1 OBSERVATION REPORT
SUMMARY
OF LIMITS TO CONCERNS
- OR File #- Subject Reasons for No Concern Elsewhere 8.1- SRV Discharge Path This, as with other minor discrepancies in the FSAR, was a random occurrence. The observation was i issued as a result of IDR need to treat each FSAR I state-ment as a licensing commitment. No reason !
was identified by the IDR team for expecting '
any similar FSAR problems to represent concern for the adequacy of other systems or to have any adverse impact on the plant's ability to achieve a safe shutdown condition. ;
8.2 Column Baseplate The issue was one of insufficient documentation Thickness of engineering judgment and not one of adequacy.
The IDR* concluded there was no real cause for concern elsewhere, because a similar application of judgment would have produced a similar result.
8.3- Alarms for ESW Same as for OR 8.1 Makeup Pumps 8.4 Burial Depth of Same as for OR 8.1 ESW Pipes ,
8.5 . Seismic Analysis Same as for OR 8.2 for Screenhouse 8.6 . Valve Disc Require- Same as for OR 8.1 ments 8.9- Relay Protection A review of the S&L drawings has identified no in 125 V-dc system other instance where non-Class lE instruments fed from Class 1E power supplies are connected up-stream of the second isolation breaker without fuses. Also, it was concluded the application of these de instruments does not degrade the Class lE de bus below an acceptable level, even without the additional fuses.
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TABLE-1 OBSERVATION REPORT
SUMMARY
OF LIMITS TO CONCEP.NS OR File-# Subject Reasons for No Concern Elsewhere 8.10 Battery Capacity This condition is not likely to be a problem with the ac system because conservative estimates of the Class 1E ac loads are required by RG 1.9 Further, the SER indicates that the electrical design had previously been reviewed for compliance with RG 1.9 and had been found acceptable.
Conservative assumptions of electrical loads have been found in all other cases reviewed by the IDR Team.
, 8.14 .ESW Makeup Pumps This appeared to be a random discrepancy since Seismic Qualifi- other items such as the structure and piping were cation reviewed for the new spectra. Also, only the river screenhouse spectra were revised, at that
, time, and not those for the other Seismic Category 1 buildings.
8.16 Component Support The issue was that an S&L document addressing Weld Sizes weld design did not require weld size in strict conformance with the applicable portion of the ASME B&PV Code. The IDR team judged that U
design was adequate since S&L analysis had established that such welds met stress limits and further qualification of the welds had been performed. While the particular situation exists throughout the design, the IDR team .
concluded that the other welds would likewise be adequate. While S&L had already applied for
, a Code case (to allow the situation) prior to L the IDR, CECO decided to review all affected i welds on all systems to bring them into strict
!' code conformance.
J l 8.17' Structural Steel The issue was similar to OR 8.16 as it relates i Weld Size to conformance to the AISC Code for structural l steel weld sizes. The IDR team conclusion was similar to that of OR 8.16 The welds reviewed by the IDR, and those for other safety-related l
structures, were done to a qualified weld procedure, and the welds had been qualified for strength requirements.
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TABLE-1 OBSERVATION REPORT
SUMMARY
- OF LIMITS TO CONCERNS i
OR File #: Subject Reasons ~for'No Concern Elsewhere 8.19 NPS Pipe Support The IDR team, upon receipt of clarifying Calculation Review information, concluded no discrepancy existed.
-8.21 Interchangeable The IDR team concluded the situation was unique Components for Corner & Lada pipe sunport components and was satisfied with the existing situation, once clar-ification was received from S&L regarding field commodity control procedures.
8.22 ESW Piping Design The issue was one of compliance with the Code Pressures and not one of adequacy. Although the higher
' pressure conditions were not code required, the piping'was capable of withstanding these
- improbable higher pressures. It was shown that ic' there was actually Code compliance.
'8.23 ESW Valve The issue was one of inconsistency between the Testing FSAR and procurement specifications and not one of adequacy. The supplier did, in fact, test the valves. If valves are not tested in the shop they are tested during preoperational testing.
8.25 Stress Calc. The issue was one of clarity 'in defining 1SX-17 changes in pipe support locations and not one of adequacy. The final piping stress report including addenda does match.the actual piping support configuration.
( 8.27- Pump & Valve The issue was a minor inconsistency between the-L Testing FSAR and procurement specifications and not one l of adequacy. Testing requirements have been met j or will be met during preoperational testing.
l 8.28 CCW Electrical The issue was one of readily locating documents.
Penetrations Upon receipt of clarification by S&L, the IDR l concluded that no discrepancy existed which would adversely affect the intended safety function of the components. This was supported by a review of a significant number of additional packages.
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TABLE-1 ;
OBSERVATION REPORT SUM 4ARY I 0F LIMITS TO CONCERNS OR File # Subject Reasons for No Concern Elsewhere 8.29 Non-pressure Boundary Stress The issue was one of documentation which raised a concern of review adequacy. However, upon Criteria clarification by S&L of its standard practice, the IDR concluded that S&L had an adequate review process and that it functioned. This was supported by a significant sample of valve stress analyses.
~ 8.31 CCW Partial- This issue was one of AWSDl.1 code Pressure Welds compliance and there was no concern that the weld in question was adequate to perform the intended safety function. An extensive S&L review 'of other welds established that this discrepancy was a unique occurrence.
-8.32 Aux. steel support This observation related to a convenient and overstress technically justifiable design practice which used terminology ("overstress factor") which appeared to lack compliance with the AISC Code.
It was established no discrepancy existed.
8.34 Welded Connec- The issue was one of the lack of adequate tions documentation of weld design review. The IDR
- concluded the weld was adequate, based on analysis, and, therefore, the application of judgment was effective. The IDR further concluded that such similar application of judgment for other safety-related systems would '
have produced an adequate design.
8.35 Piping Support The issue was one of documentation of design Calculations change review judgments. The IDR concluded the situations reviewed were adequate and that the judgment application was substantiated. The IDR further concluded similar applications of judgments for other safety systems would have produced an adeouate design.
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' TABLE-l' OBSERVATION REPORT'SUW1ARY OF LIMITS TO CONCERNS 0R File i Subject Reasons' for No Concern Elsewhere r 18.36 Expansion plates The observation dealt with a _ question of the adequacy of the design margin to meet the NRC IE Bulletin 79-02 requirement provided by a S&L design standard. S&L.provided a calculation of an appropriate limiting condition and the IDR team accepted the calculation as demonstrating the-standard's adequacy. The standard was used throughout the plant, _ and since it was judged adequate, no concern exists for its application elsewhere.
. 8.37 Support Swing The Observation dealt with a question of the Angle Limit adequacy of the design procedure to ensure proper application of component supports.
Clarification by S&L of the design process, and also of the checks of the conditions in question by walkdowns during hot functional . testing satisfied the IDR team that an adequate, controlled process existed. Therefore, the process was judged adequa b.
8.38 CCW Design Pressure The Observation originally dealt with the adequacy of the selection of the ASME B&PV Code design pressure _for the CCW system. The S&L l response on this point was judged adequate by the IDR team.
The Observation resulted, for _other reasons, in the designer (Westinghouse) notifying the NRC of
" a 10CFR21 situation as a result of an identified-potential overpressure condition. The IDR team judged that the situation was such that the plant's capability to achieve safe plant shutdown was not adversely affected. Also, the IDR team judged that tho'CCW design, and the effects on it which might lead to the postulated overpressure condition, was unique compared to other safety-related systems, and no concern existed that the situation would be replicated for other plant systems.
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TABLE-1 l OBSERVATION REPORT SUM 4ARY OF LIMITS TO CONCERNS OR File # -Subject Reasons for No Concern Elsewhere 8.39 Power Cables in The resolution of the observation pointed out that Cable Spreading the uncovered power cables actually was not Room included within the defined area of 'the cable spreading. areas. Therefore, this was not a deficiency.
.- 8. 40 Cable Separation Other manhole drawings were checked and did not contain any conflicting lines or any lines at all.
Therefore, this appeared to be isolated to the -
subject drawings. Field inspection showed that the cables are installed correctly.
8.41 Motor Operated To resolve this observation S&L performed a calcu-Valve Operators lation to verify that MOVs required to function upon a safety signal will perform their safety function. In this calculation S&L included MOVs of.all safety related systems in the plant. S&L expanded the scope of this observation to assure that the design of power supplies to 460V motors
- and MOVs of other systems are adequate with regard
-to this concern. -
8.42 Cable Saddles in Since the design of the cable saddle was proven as Manholes adequate, use of these saddles elsewhere would
- e. also be acceptable.
8.44 CCW Nozzle Loads The Observation dealt with a question of whether the designer's judgment that the effects of
- thermal growth produced insignificant stress levels and nozzle loads was justified. After extensive review within S&L and by Bechtel, L
the IDR team judged the configuration in question to be unique, and concluded that there was no reason for concern with similar judgments else-where in the plant.
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TABLE-1 OBSERVATION REPORT
SUMMARY
OF LIMITS TO CONCERNS OR File # Subject Reasons for No Concern Elsewhere
.8.47 HELB Jet The design process for HELB jet impingement
, Impingement effects is considered adequate for the entire plant based on the process identified by the IDR and supplemented by the results reported in
" Confirmation of Design Adequacy for Jet Impingement Effects" which examined postulated breaks plant-wide for jet effects and reported that no plant modifications were required.
8.49 ESW and CCW The Observation dealt with a question of strict Piping Flanges compliance with the ASME BP&V Code, as interpreted by the IDR team. In this case, there was a difference of opinion on code interpretation.
S&L calculations for limiting conditions demonstrated design adequacy. There was a conclusion on the part of the IDR team that a technically adequate situation existed throughout the plant, and the interpretation of the ASME Code did not in any way affect any safety-related system's capability. to perform its intended safety function. Despite extensive reviews for code compliance, no significant deficiencies were found elsewhere.
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10-01-84 Page 1 .
SARGENT & LUN6Y RESPONSE TO- .
NRC FROM MEETING OF 09-14-84 1
- Item 1 Edisonshas agreed to do all items committed.in the Bechtel-
. Report. Sargent.& Lundy has developed a tracking mechanism for Byron I and will make pqriodic submittals.of.the close-out status to the affected-project distribution. In addition, the Bechtel Report should be reviewed for actions to be' taken ,
oniByron'II and Briadwood I and II. A similar tracking much-anism will be (?eveloped and . distributed.
Sargent ^ & Lundy Response A tracking mechanism has been developed for the Byron Unit I-IDR. The only remaining open item is the required FSAR update resulting from OR 8.47 dealing with HELB. The. applicability of any_ Byron I IDR Commitments will be tracked and implemented j as appropriate for Byron II and Braidwood I and II.
. Iten 3 _, ' ,
A s$hedule for updating the FSAR for those items committed in the Bechtel Report should be'provided. -
Sargent &,Lundy~ Response- .
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IA11' items areLattached except the changes associated with OR'8.47
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3 which will be submitted the. week of October 1,19R4 Item 4
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i, In discussing the battery cross-tie, we agreed to document the
- operating limitations. We will. prepare a discussion with in-put from CECO Operating Station personnel.
! - Sargent & Lundy Response The de cross-tie consists of a manually operated breaker at each end of the cross-tie (i.e., one manually operated' breaker .
in the' Unit'2 de dictribution center, and one in the Unit 1 dc
- distribution center). All cross-tie breakers are normally
- padlocked in the open position with administrative controls on release of keys.- Use of the de cross-tie is presently limited by. Station Technical Specifications such that:at least O
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Sargent-6 Lundy Response to 10-01-84 NRC From Maeting of 09-14-84 Pago 2 Item 4 - Sargent & Lundy Response (Cont'd) i one of the two units must be in either a cold shutdown or refueling mode of _ operation (Modes 5 or 6) . The purpose of the cross-tie is to supply de power to some of, the loads in the bus of the "down" unit when, and if, it is desirable to
-isolate the battery of the "down" unit for maintenance or testing.
With one unit shutdown (Mode 5 or 6), the operating proceduceu for closing. the dc cross-tie ACBs, including the limitation on ;
the allowablu crouu-tie load, will include the following: !
- 1. Specific circuit breakers on the distribution panol will -
be opened to ensure that the cross-tie load will be pro-perly limited.
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- 2. The cross-tie breakers at Bus 111 and 211 vill then be
. unlocked and closed.- ! Note that a " cross-tie ACB closed" alarra at- the MCB annunciator will alert the Control Room '
operator when the ACBs are closed.) ,
- 3. The battery breaker at Bus 111 would then be opened 1(note
-that a " battery 111 ACB open" alarm on the MCB annunciator ,
will alert the Control Room operator when the-breaker is opened). -
With this' procedure, closing the de cross-tie is an admini-stratively controlled procedure in which the load circuit breakers are opened in a deliberate and-preplanned order,-prior to closing of the cross-tie breakers and disconnecting the ,
ba'ttery .
The reconnection of Battery 111, the opening of the cross-tie breakers, and the closing of the load circuit breakers for re-turn to normal operation, will be carried out in the reverse order, . again using documented procedures / checklist and fadmini- y strative controls. ,
Item 7 (Observation Report 8.21) ,
Provide a schedule for cla'rified and revised drawings for e OR 8.21. '
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_ Sargent & Lundy Response b 1-The drawings are currently being revised to clarify the inter-
,_ changeability of safety and non-safety related hanger. parts.
The drawings are scheduled to be revised, reviewed, and appro-ved by about October 5,,1984.
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. ' .' Sargent & Lundy Response to 10-01-84 NRC From Meetinf of 09-14-84 Page 3
- Item 8 (Observation Report 8.29)
Provide a commitment and a sc edu'le to change the Sargent &
Lundy design procedures to document when active allowable stresses were used rather than passive values.,
Sargent & Lundy Response No change to Sargent & Lundy procedures are required because a change to the procedure has already been made which addresses the documentation of the allowable stress values used, Sargent & Lundy is currently using, and has used since November 1982, a revised checklist which requires the reviewer to list the total stresses and the allowable stress values at criti-cal locations. This allows an auditor to determine whether active or passive allowables were used by the reviewer and satisfies documentation requirements.
Item 9 (Observation Report 8.32)
Provide an expanded basis why a 10% over-stress is not a problem. i The answer should address both the past and future.
Sargent & Lundy Response
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~~F 6F"th5 assessment of as-built small bore pipe supports, a driterion _-
was established such.that up to a 10% calculated over' stress was
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considered acceptable before additional calculations were required to establish code compliance. This was due to the fact that the
~~ hanger analysis was ' known to be~9ery conservative, and that refine-ments to this analysis _Would_ demonstrate _that_the hanger met all applicable design requirements. .
The 'known conservatism ~ in~smal1~ bbre~ pipe support design ~ include
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conservative loadings and conservative analysis techniques.
Loadings The design loads used for small bore pipe cupports are conser-vative because each support is designed for the peak plant seismic excitation.' The actual excitation of any wall or slab in the plant can be'much smaller than the peak excitation. This is a simplifying loading assumption which is reasonable con-sidering the small amounts of material required for small bore pipe supports. Also, the loads used are not based on the actual gravity load on a given support but rather the upper bound load. This is because small bore pipe supports are chosen by the contractor from a table based on allowable loads.
The support load always falls between two table capacities.
For example, if support Detail 1 on)the table is designed for 50 pounds and Detail 2 is designed for 100 pounds, a contractor with a 60 pound load must choose Detail 2. The result.s of the use of design tables and the use of peak plant acceleration values is a very conservative design load on any given small bore pipe support. -
10-01-84 Sargent & Lundy Response to
-NRC From Meeting of 09-14-84 Paga 4 Item 9 - Sargent & Lundy Response (Cont'd)
Analysia The analysis technique used'for small bore piping analysis in-volves a simplified method of piping analysis which gives con-servative piping loads at the supports'. This method basically considers one support at a time.. This is a very conse'rvative analytical procedure. A detailed dynamic computer analysis of the piping including all supports will always give smaller cal-
.culated pipe support reactions.
Conclusion
'The criteriiri t'o allow an apparent 70%71ncrease aliove cesign -
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allowables when simplified design methods are used is justi-fled because these conservative engineering methods of deter-mining loads and performing analysis for small bore pipe support design would not result in an actual over-stress if specific calculations were made.
Item 10 (Observation Reports 8.34 and 8.35)
Describe the basis for the engineering. judgement that was used on these two items. Discuss the relationship of the depart-mental standards with respect to these items also. .
Sargent & Lundy Response ,
The calculation for 1CC01009R indicates that the connection design was performed by utilizing the Review Manual with addi-tional hand' calculations. This'" Review Manual" contains design guidelines and assumptions. These design guidelines and as-sumptions apply to standard hanger configurations with member sizes and weld requirements and contain associated load tables.
The load tables have, among other things, built in consider-ations of the effects of installation tolerances and member deflections.
The original hand calculation performed verified the adequacy of the plate. Comparison of the weld capacities in the " Review Manual"=provided a basis for weld adequacy. Documentation for weld adequacy has been provided in a revision to the origianl calculation.
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- '. 'Sargnnt & Lundy Response to 10-01-84 g,, . NRC From Meeting of 09-14-84 Page 5 Item 10 - Sargent & Lundy Response (Cont'd)
OR 8.34 (4.2) Pipe Support 1CC01047, (4.3) Pipe Support 1CC01042 and (4.J) Pipe Support 1CC01034 The weld evaluation of the specified flare-bevel weld on the support drawings ICC01034, ICC01042 and 1CC01047.was based on engineering-judgement. The judgement was made by comparing the actual load to the maximum load carrying capability of the strut (all three supports are.Elcen Size 2 Struts).
- Maximum load carrying capability of the Elcen Size 2 Strut is:
Strut Design Load Strut Emergency Load 2870 lbs. 3710 lbs.
Piping loads on the support drawing are:
Actual Design Load ' Actual Emergency Load 1CC01034 831 lbs. 1597 lbs.
ICC01042 421 lbs. 1018 lbs.
lCC01047 647 lbs. -
1344 lbs.
The piping loads are-less than 50% of the load as tabulated above. The flare-bevel weld (the effective throat of the flare-bevel weld is 0.156" ccmpared to 0.176" for the fillet
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weld) was judged to be adequate for the actual design and i emergency loads.
Sargent & Lundy has performed calculations to verify the
- l. engineering judgement.- The calculation demonstrated that the l- design as specified is acceptable.
L OR B.35, Item 4.1, Pipe Support 1CC01010X
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L The original weld configuration - outside and inside weld at l both flanges - was based on an " Emergency" load of 696.7 lbs.
! Through subsequent minor revisions, this weld configuration remained the same even though the actual " Emergency" load was l' ,
..,._ reduced by almost one-half _to 3639 lbs g. '
The weld configuration was subsequently changed by omitting the weld at the inside of both flanges. The weld configuration i
prior to this change had a design margin of approximately 5 to 1.
l The judgement to reduce the weld'section was based on the actual
{ loading for the support. Calculations have been performed verifying this judgement. The design margin for the weld as l revised was in excess of 2 to 1.
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.: Sargent & Lundy Response to 10-01-84 NRC From Macting of-09-14-84
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Pago 6 Item 10 - Sargent & Lundy Response (Cont'd)
OR 8.35 (4.2) Piping Support 1CC01051X Sargent & Lundy has developed standard concrete expansion anchor. tables and charts for given anchor bolt assemblies.
.These tables and. charts allow a graphical selection of expan-
.sion anchor sizes. For Support 1CC01051X, the support design was changed from a 4-bolt assembly to an 8-bolt assembly.. New
. calculations on the 8-bolt assembly were not gencrated since the strength of the two assemblies can be determined by com-paring two charts in the standard. As a result of this obser-vation, calculations have been generated verifying that the
' determination that was made by comparing the two charts was accurate.. .
i IOR 8.35, Item 4.3, Pipe Support 1CC01012R The calculation accounts for the location tolerance and the
-proper load for. Support No. M-1CC01012R and M-1CC14009R utili-zing the " Review Manual" which was referenced in the calculation.
No engineering judgement was used.
Documentation of Engineering Judgements
-In the future, rngineering judgements similar to those de-scribed above will be'4ocumented as required by the following Sargent & Lundy standards that are in places Electrical Standard ESI-253 Structural Standard SAS-22
- Mechanica1' Standard MAS-22 Item '13 (Observation Reports 8.23 and 8.27) l-We agreed to revise the specifications or the FSAR as neces- .
p -sary to clarify the testing requirements to aid future pur-chases.. A schedule for~these revisions should be provided.
- Sargent~& Lundy Response
~T he_ specification and the FSAR have been reviewed relative to
- i. the in-shop testing requirements for pumps and valves. The
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PSAR'is being revised to clarify the testing. requirements.
The specifications contain all of the necessary testing require-
- ments and'do not require-revision.
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Sargent & Lundy Response to 10-01-84 l
. NRC From Meeting of 09-14-84 Page 7 Item 15 .-
Provide a summary of corrective actions taken as a result of the trends shown on Page 72. Discuss that no corrective action was needed on code items.
l Sargent & Lundy Response l l
The following actions have been taken by'Sargent & Lundy rela-tive to the trends identified on Page 72 of Volume I of-the Bechtel IDR.
. Use of Undocumented Judgements Standards have been issued by Sargent & Lundy in the Electrical, Structural, and Mechanical areas via Standards ESI-253, SAS-22,
. and MAS-22, these standards require documenting engineering
.judgements.
Insufficient Control of the FSAR
- The FSAR is being updated for all Observation Reports requiring FSAR update.. Other minor updates will be made in future amend-ments as appropriate.
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!- . Insufficient Review of Changes
- Sargent & Lundy Quality Assurance Procedure GO-3'07, Sargent & .
Lundy Drawings, requires that the reviewer of the drawing re-l view the-drawing for technical adequacy in accordance with 7
departmental standards. Other Quality Assurance Procedures cover design activities other than Sargent & Lundy drawings.
These procedures also require that revisions be prepared, re-l- viewed, and approved, in accordance with the same procedures L as the original activity.
Bechtel concluded "The review of the S&L design. process indi- -
cated that:each of these processes was controlled, but IDR~
Observations were made.for each area related to reviewing ~
changes and coordinating them within S&L. This indicated that
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certain minor deficiencies may exist in the S&L process but
- does not lead the IDR to conclude that.the process is generally inadequate."
i.
Sargent & Lundy.has, however, made the IDR Report'available to
' tha Design Directors in the Mechanical, Electrical, and Struc-tural disciplines and has requested that the Design Directors emphasize.the requirements for the review of' design changes l- to design personnel.
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Sargant & Lundy Response to 10-01-84 l
, . NRC From Meeting of 09-14-84 Paga 8 Item 15 - Sargent & Lundy Response (Cont'd)
Noncompliance with Code Requirements Sargent'& Lundy recognizes that code compliance is required and has addressed and resolved the Observation Reports that ;
deal with OR 8.16, 8.31 and 8.49. j Furthermore, Sargent & Lundy does not consider this to be a
-trend. The code circumstance identified in OR 8.16 was recog- l nized by Sargent & Lundy prior to the IDR and corrective action was being pursued. The partial penetration weld of OR 8.31 is considered to be an isolated case and OR.8.49 is a dif-ference of opinion on an interpretation of what the code requires.
Sargent & Lundy performed flange analysis in response to the OR, which demonstrates that the moment requirements of ASME
.Section III'have been met. None of the OR's have resulted in a question of design adequacy including OR 8.49.
In addition, with respect to the code interpretation identified in OR 8.49, Sargent & Lundy is developing a generic procedure for flange analysis. This procedure will require flange analy-sis for future ASME Section III piping analysis. In the interim, piping analysis personnel have been instructed to perform the flange analysis for Section III piping containing flanges.
^
Item 16 Change the appropriate page in the FSAR to state that the
- valve performs an isolation function not a throttling function.
Sargent & Lundy Response The required FSAR change; is attached.
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Sargent & Lundy Response to 10-01-84 NRC From Meeting of 09-14-84 Page 9 HIGH ENERGY LINE BREAK REPORT Item 1 A phone call will be held with.the NRC the week of September 21, 1984, to discuss hinge points, whipping pipe, secondary hinges, shape of breaks, zone of influence, etc.
Sargent & Lundy Response The phone call was held and Sargent & Lundy will provide the
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one additional item requested as a result of the phone call fo'r submittal' to the NRC~ the' week of OctoDer 17 1984.
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Item 2 Provide a schedule to revise the FSAR to make it consistent with.the High Energy Line Break Report. We should make sure that we reflect the existence of the existing jet impingement shields and the various longitudinal break locations.
Sargent'& Lundy Response The FSAR update will be submitted to CECO the week of October 1, 1984.
Item 3 Provide a copy of the Westinghouse letter which agreed with the confirmatory High _ Energy Line Report.
Sargent & Lundy Response Copy attached.
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CAW-7732 CBW-4754 I
"'" 0" "5 C *
- Westinghouse Water Recctor .
Electric Corpcration Divisions acms PinsDq Perrai.asa 15210
{
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August 1, 1984 ,,
14r. D. L. Leone, Project Director -
Ref: SL'4C-3121, Sargent and Lundy Engineers 7/26/34 55 East !!onroe Street Chicago, Illinois 50533 Attention: K. J. Green C0:4:13?iNEALTH EDISO:1 CD:4PA'iY BYR0:1 A iD 3RAIDN000 STATIO:iS - UtiITS 1 A!iD 2 SARGEftT AfiD LU:i3Y JET I:4?Ill3EttEllT P.EPORT - WESTIfi3 HOUSE RE'.'IEW
Dear ifr. Leone:
Per your request, Westinghouse has reviewed the subject draft report and has Do CoK1ents.
Our staff had reviewed a previous draft and our coments have been incorporated.
Very truly yours, WESTIrlGHOUSE'ELEC j uC C03PORATIO4 L .
.. /d
!.E.Kortier,:4ansg.r'p.
l Comonwealth Edison Projects l ,,
JLT/b:ns/',545D l D. L. Leone, 30L-cc: J. D. Deress, 2L C W. Fruehe, 2L X.-J. Green, IL W. C. Cleff, IL r.
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COMMONWEALTH ^ EDISON COMPANY FIELD VERIFICATIONS IN RESPONSE TO NRC MEETING OF 9-14-84
- Item 6 (Observation Report 8.16)
Provide status of NF weld size review and a schedule for
, ' completion.
F-Commonwealth Edison Co. Response The program regarding the NF weld size matter has'been completed and the component supports have Q.C. inspections verifying that subsection NF minimum fillet weld size requirements have been met. You will recall that resolution on this item was in progress prior to the Bechtel IDR.
' Item 11 (oFJ 9,30 Provide a status and a schedule for the completion of this design change including its implementation in the field.
-Commonwealth Edison Co. Response The design change for the revision to the CCW system has
-been issued. Field completion should-occur-by about~
10-22-84.
Item 12 / Observation Report 8.9)
Provide confirmation if the fuse'has been added in the field.
P
< Commonwealth Edison Co. Response [
-New fuse blocks'are currently being purchased and are anticipated to-be. installed by about October 12, 1984.
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ATTACIPENT B 10/1/84 Bechtel
' observation' FSAR Pages Changed Report Number:
- 8 .1 -. 9.2-17 8 '.' 3 -
9.2-31
'S . 4 Q10.8-1 8.6 3.9-96 8.14 3.9-94
'8.23 3.9-50, 51 8.27 3.9-47 8.38 9.2-16, 9.2-17 In addition, FSAR page 9.2-19.has been revised >
- .per discussion on-page A.2-34 in Volume II of the Bechtel Final Report.
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. B/B-FSAR
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^ 3.9.3.2.1 Pumps Balance of Plant All active pumps as listed in Table 3.9-15 are qualified for operability by first being subjected to rigid tests both prior to installation in the plant and after installation in the plant.
The in-shop tests include (1) hydrostatic tests of pressure-retaining parts; and (2) performance tests, while the pump is l operated with flow, to determine total developed head, minimum and maximum head, net positive suction head (NPSH) requirements, and other pump / motor parameters. After the pump is installed in the plant, . it undergoes the cold hydro tests, functional tests, and the required periodic inservice inspection and operation. These tests demonstrate reliability of the pump for the design life of the plant.
~
All active pumps, listed in Table 3. 9-15 are qualified for operability by first being subjected to rigid tests both prior to installation in the plant and after installation in the plant.
The in-shop tests include (l) hydrostatic tests of pressure-retaining parts to 150% of the design pressure times the ratio of material allowable stress at room temperatur a to the allowable stress value at the design temperature, and (2) performance l tests to determine total developed head, minimum and maximum head, net positive suction head (NPSH) requirements, and other pump parameters. Also monitored during these operating tests are bearing temperatures and vibration levels. Bearing temperature limits are determined by the manufacturer based on the bearing material, clearances, oil type, and rotational speed. These limits are approved by Westinghouse. After the pump is installed in the plant, it undergoes the cold hydro tests, hot functional tests, and the required periodic inservice inspection and operation. These tests demonstrate that the pump will function as required during all normal operating conditions for the design life of the plant.
In addition to these tests, the safety-related active pumps are qualified for operability by assuring that the pump will start up, continue operating, and not be' damaged during the f aulted condition.
The pump manufacturer is required to show by analysis correlated by tests, prototype tests, or existing documented data that the pump will perform its safety function when subjected to loads imposed by the maximum seismic accelerations and the maximum U
3.9-47
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'B/B-FSAR- ,
In-case the: natural. frequency ir found-to be below'33 Hz, a dynamic or pseudo dynamic analysis is' performed to determine-the amplified input accelerations necessary to perform the stress analysis.
- b. Additional loads considered ..n the stress analysis of the pumps and their supports are the nozzle loads for T he applicable plant condition from .
interconnecting piping. systems.
- c. In addition to the stress analysis, a static shaft deflection analysis of tne rotor is per-formed. The deflection determined from the static shaft analysis is compared to the allowable rotor clearances.
- d. To complete the seismic qualification procedures, the pump motor and all appurtenances vital to the operation of.the pump are independently qualified
'for' operation during the' maximum seismic event in accordance with IEEE Standard 344-1975 (see Section 3.10). 'In the analysis interaction between the pump and motor is considered.
- e. Alternatively, the entire pump assembly with appurtenances may be qualified by testingInin accordance with IEEE Standard 344-1975.
performing the seismic testing the nozzle loads .
for the applicable plant condition must be applied.
From this, it is concluded that the safety-related pump / motor assemblies will'not be damaged, will continue operating under SSE loadings and will perfcrm their. intended functions. These requirements take into account the complex characteristics of the pump and are sufficient to demonstrate and assure the seismic operability of the' active pumps.
.3.9.3.2.2' Valves Balance of-Plant Safety-related active valves as listed in Table 3.9-16 must
-perform their' mechanical motion in times of an accident.
Assurance must be supplied that these valves will operate
~during.a seismic event. Qualification tests and/or analyses have been; conducted for all active valves to assure valve opera-
.bility under. seismic and/or environmental conditions.
The valves are subjected to testing prior to service (in-shop and preoperational-field) and in situ (during plant life) as required by_ specific service and' functional requirements.
In-shop tests include the following: a) ASME Code - required hydrostatic tests to assure pressure boundary integrity:
3.9-50
(-
-B/B-FSAR- .
J .b) Specified conforImance to Manufacturers' Standard Practice cois requirements regarding hydrostatic tests and main seat leakag'e;
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,' c)._ Specified timed operational tests (valve stroking) when addi-tional verification of design requirements is necessary.
Cold hydro qualification' tests, hot functional. qualification tests, and periodic inservice-operation are performed in situ to verify and ensure the functional ability of the valve.
These. tests and appropriate maintenance ensure operability of the-valve for the-design life of the' plant. The valves are designed using either the standard or the alternate design rules of ASME III.
On all active valves, an analysis of the extended structure is also performed for static equivalent seismic loads applied at the center of gravity of the extended structure. The maximum stresses and deflection allowed in these analyses demonstrate operability and structural integrity.
Valves which are safety-related but can be classified as not having an overhanging structure, such as check valves and safety-relief valves, are considered separately.
Due.to the particular simple characteristics of the check valves, they will be qualified by a combination of the following tests and analysis:
- a. stress analysis including the seismic loads where applicable,
- b. in-shop hydrostatic tests,
- c. in-shop seat leakage tests, and
- d. . periodic.in situ valve exercising and inspection to ' ensure the functional capability of the valve.
The safety / relief valves a're qualified by the following procedures. These valves are also subjected to tests and analysis similar to check valves; stress analyses including
.the seismic loads, in-shop hydrostatic seat leakage and performance tests. In addition to these tests, periodic in situ valve inspection, as applicable, and periodic valve removal, refurbishment, performance testing, and reinstallation are performed to ensure the functional capability of the valve.
Using the methods described, all the safety-related active valves in the systems are qualified for operability during a seismic event. These methode, proposed conservatively, simulate the seismic event and' ensure that the active valves will perform their satsty-related function when necessary.
D 3.9-51
B/B-FSAR TABLE 3.9-8 DESIGN CRITERIA FOR ACTIVE PUMPS AND PUMP SUPPORTS COEDITION DESIGN CRITERIA
- Design and Normal ASME Section III
. Subsection NC-3400 and ND-3400 Upset- o ,5 1 0 S o,+o b $ 1 .5 S Emergency o,3 1 2S o,+o b 1 1 65 S Faulted o,5 1 2S o, + o b1 1 8S t-
- The stress limits specified for active pumps are
- more restrictive than the ASME--III limits. For the Faulted Condition (membrane plus bending), stresses may exceed 1.8.S but must remain below the material yield stress. In such cases, a deflection analysis
'is performed to assure that the maximum displacements are within the deflection limits which will not impair '
the operability of the equipment. t 3.9-94 L' -
B-B/FSAR-TABLE 3.9-9 (Cont'd) e
'4. Design requirements listed in this table are not applicable to' valve discs, ~ stems, seat. rings, or other parts of valves which are contained within the confines of the body and bonnet. l ,
- 5. The maximum pressure resulting.from upset, emergency, or faulted conditions shall not exceed the tabulated factors listed under P times the design pressure or the rated
-pressure at thE*Epplicable operating condition temperature.
If the pressure rating limits are met at the operating conditions, the stress limits in this table are considered to be satisfied.
- 6. Stress limits are taken-from ASME III, Subsections NC and ND, or, for valves procured prior to the-incorporation of these limits into ASME III, from Code Case 1635.
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3.9-96
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'*t; 3~ B/B-FSAR 3
a 4- "9.'2.2.4.1 system Availability and Reliability
' i Either unit may be aligned with two completely independent, I : parallel trains, each consisting of one pump and one component 1 ' cooling-heat exchanger. Either train provides cufficient cooling M .to accommodate the heatzloads experienced by that unit during a 4 ' loss-of-coolant accident. Hence, any single active or passive
- failure in the system does not prevent it from performing its
- j. design 7 function.
1 -Insid' e the containment, - mcst of the piping, valves and instru-1 mentation are located outside the shield wall at a location above the calculated water level in the bottom of the containment at 7] postaccident conditions. In this location, the portions of the ih system within-the containment are: protected against missiles and
] against flooding during postaccident operations. This location
- also provides radiation shielding which permits maintenance and
[ -inspection to be performed during normal power operation.
3 d The component : cooling pumps, heat exchangers, surge tanks and associated valves, piping and instrumentation are. located outside the containment and are, therefore, available for maintenance and
- ) ; inspection during power. operation. Replacement of a pump or heat exchanger- may be performed in accordance with technical
, specification limitations while the other units are in service.
l 1 Sufficient cooling capacity is provided to fulfill all system requirements' under nora&l and accident conditions. Adequate a .. safety margins are included ln the size and number of components
'1 .to preclude the possibility of a component malfunction adversely affecting , operation of safety features equipment. The relief l valves on the component cooling water lines downstream from each reactor cooling ~ pump are designed with a capacity equal-to the
. maximum rate at which reactor coolant can enter the component-cooling system for a severance-type break of the ' reactor coolant pump thermal barrier cooling coil. The valve set pressure equals
.the design pressure of the component cooling piping.
"'t The relief valves on the cooling -water lines downstream from the sample, excess letdown, letdown, seal water, spent fuel pit, and
. residual heat ' exchangers are sized to relieve the volumetric expansion occurring if the exchanger shell side is isolated and high-temperature coolant flows through ~ the tube side. The set
. pressure equals the design pressure of t.he shell side of the heat
-exchangers.
9.2-16 I
- B/B-FSAR 9.2.2.4.2 Leakage Provisions and Activity ' Release Welded construction is used where possible throughout the Component Cooling system piping, valves and equipment to minimize the possibility of leakage. The component cooling water could become contaminated with radioactive water due to a leak in any
_ heat exchanger tube in the chemical and volume control, the sampling, the residual heat removal or the spent f uel pit cooling systems or due to a leak in the cooling coil for the reactor coolant pump thermal barrier.
-Leakage froa or to the component cooling system can be detected by a' change of level in the ccmponent cooling surge tank. The rate ~ of water-level change and the area of the water surface in the tank permit determination of the leakage rate. In-leakag e' i s detected -anytime. by radiation monitors located on the main return headers. To assure accurate determinations, the operator must check that temperatures are stable.
AL cooling water temperature increase of about 2500 E in one of w_ .the units would be required to overfill its component cooling surge. tank. However, should a large leak develop in a residual heat exhanger, letdown heat exhanger, or due to a ruptured reactor coolant pump thermal-barrier, the water level in the component cooling surge tank of that unit would rise, and the: operator would be alerted by a high-water. level alarm.
The vent on the surge tank is automatically closed in the event of high radiation level detected at the component cooling heat exchanger discharge header. If the leaking component
-is not isolated _from the loop before the inflow fills the surge tank, the overflow line with a. loop seal will prevent component cooling system overpressurization. The overflow is routed to the chromated drains system.
Three heat exchangers are provided to serve the two units.
L During all conditions of plant operation, this provides for one backup exchanger. If a11'three exchangers are available, however, the backup exchanger may be employed on the unit undergoing a LOCA or shutdown (RHR heat exchanger in operation) .
Design cooldown rates are determined on this basis f(two exchangers operating on the unit recovering from a LOCA or
- shutdowns , but 'the consequence of the loss of one heat exchanger during tnis time only slows down the cooldown rate from the design valce and does not, aff ect the safe operation of the plant.
' Five pumps are provided to serve the two units. Under the limiting case, four pumps are required for the two units leaving one pump. as bi kup pump for either - unit.
9.2.2.4.3 Incident Control i_ Containment isolation valves are automatically closed on a saf ety
. features. actuation "T" signal. The cooling water supply header to the reactor coolant pumps contains a check valve inside and
, remotely operated valves outside the containment wall. The 9.2 ,
B/B-FSAR -
The instrumentation in,-the CCWS is provided primarily for initial system flow balancing and for monitoring purposes during normal operation. Thus failure of any of this instrumentation has no effect on system performance. Exceptions to this are:
- a. letdown heat exchanger CCWS flow controllers, e
Eb . : reactor coolant pump thermal barrier outlet flow controller, and
- c. component cooling surge tank radiation control valve.
The letdown heat exchanger tube side outlet temperature controls a butterfly valve which regulates the CCWS flow to the shell side of this heat ~ exchanger. Should the controller fail in a way to shut off CCWS flow to the circuit, a high temperature alarm will sound in~the control room allowing the operator to take
. corrective action.
Safety-related indication of component cooling water flow from the reactor coolant pump motor oil coolers is provided The reactor coolant pump and alarmed in the main control board.
-(FCP)-
causes thermal barrier outlet a motor-operated header valve has ainflow to close this controller line in thewhich l event of high flow (an indication of a broken RCP thermal barrier). Should the controller not operate properly, an in-creasing level is noted in the CCWSA surge tank, resulting in a second motor-operated valve high level alarm, if not isolated.
in series with the flow control v.alve is available for manual isolation of the line if required. Addi,tionally, two level instruments are provided on each surge tank, both of which will give a high level alarm in the control room.
Each component cooling surge tank vent has an air operated valve which will close on a high radiation signal from the radiation i monitors in the discharge Leaders from the CCWS heat exchangers.
This high radiation alarn normally indicates a primary to CCNS leak. Three radiation monitors are provided. The monitor on the common heat exchanger will alarm and close the vent valve on both surge tanks. The radiation monitors on each unit's heat exchanger will alarm and close its respective surge tank vent valve.
9.2.2.4.5. Electrical Power Supply A
The normal p. er supply to the system is from the ESF buses.
full description of the power supply is given in Subsection 8.3.1.1.
9.2.2.5 Tests and Inspections During-the life o'f' the Station, the CcLponent Cooling System is
.in continuous operation and performance tests are not required.
Standby pumps are rotated in service on a scheduled basis to obtain even wear. Preoperational tests are performed on the system. The equipment manufacturer's recommendations and station practices are considered in determining' required maintenance.
9.2-19
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I BYRON-FSAR j The worst case heat transfer to atmosphere condition of 820 F wet l'
bulb for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> on July 30, 1961 would result in a cold water 9- outlet temperature of 94. 60 F at a heat rejection rate of 580 x
' 106 Btu /hr based upon predicted tower performance curves.
The cooling ' tower is, therefore, adequate for all worst case meteorological conditions concurrent with a loss-of-cooling i accident in one unit while the other unit is being safely shut 4 d own.
3 The essential service water makeup pumps may be started manually from the control room, locally at the river screen house, Or automatically on level controls of the cooling tower basins.
Once. started automatically, they continue to operate until the i
2000 gallon fuel supply to each engine drive (approximate fuel i consumption is 10 gallons per hour) is exhausted or until the i engines are manually stopped from the control room or locally.
j The engines and pumps are capable of meeting makeup requirements
] for the actual post-LOCA heat rejection rates under worst case
] evaporative loss conditions.
1 1 9.2.5.4 Tests and Inspections 4
Since complete redundance is provided in the system, both towers
]9 are normally operated, with one tower providing cooling for one unit and the other tower providing cooling for the other unit.
( The normal operating heat load of one unit (142 x 106 Btu / hr) or 4 the refueling and maintenance outage heat load (13 x 106 Btu / hr)
] are more than adequate to prevent freezing of the basin and fill j under winter design ambient conditions. Tower makeup may be i switched from the Rock River source to the onsite wells. In thic
[ manner, continuous surveillance of all equipment availability and operability is maintained.
9.2.5.5 Instrumentation Requirements category I level switches are provided in each essential l service water cooling tower basin. In the event of low level in j a cooling tower basin, the corresponding essential service water makeup pump is automatically started. It continues operating until it is manually stopped, or exhausts the supply of diesel 1 fuel oil in its 2000-gallon storage tank.
i j Local alarms and shut down equipment f or the diesel engine makec; i pump drives are provided for high cooling water temperature in
[
the closed cycle cooling system, low lubricating oil pressure,
' engine overspeed, and engine overcrank. Annunciation is trans- l mitted to the main control room indicating " Engine Trouble ,"
auto-start, and auto trip for each engine.
I 9.2-31
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B/B-FSAR.
,, :s QUESTION 010.8
" Provide piping arrangement drawings (plan and elevations) ,
for~the: essential service water supply.and return lines from the ultimate heat sink to the essential service water pumps. Verify that the essential service water piping has not been routed through areas such that a seismic event will not prevent the system'from performing
'its safety function." ,
RESPONSE
The essential service water supply and return lines from the ultimate heat sink to the essential service water pumps has not been' routed-through areas.such that a seismic event will not prevent the system from performing its safety function.
At Byron, these lines are buried minimum 25 feet below grade level'and the soil is such that through a seismic event, it will retain its supporting and restraining capability and limit the seismic movements of the buried essential service water pipe to an acceptable level.
At Braidwood, the top soil-has a potential for liquefaction.
Therefore, the essential service waterlinec have been buried
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.below the top soil level and rest within the undisturbed till, which will retain its supporting and restrai.ning func-L tion-through a seismic event and limit the seismic movements '
of buried essential service waterline to an acceptable level.
Note: This response has been superseded by the response ;
'to' Question 010.21. !
O O
'010.8-1
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