ML20083J707
| ML20083J707 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 12/30/1983 |
| From: | Reed C COMMONWEALTH EDISON CO. |
| To: | Deyoung R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| References | |
| 7668N, NUDOCS 8401100275 | |
| Download: ML20083J707 (147) | |
Text
/
\\ Commonwe lth Edison j
) One First National Plaza. Chicago Hhnoes
_(
C Address Reply to. Post Office Box 767
(
Chicago. tilinois 60690
~
December 30, 1983 Mr. R. C. DeYoung, Director Of fice of Inspection and En forcement U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Byron Generating Station Units 1 and 2 Integrated Design Inspection Inspection Report 50-454/83-32 NRC Docket Nos. 50-454 and 50-455 References (a):
September 30, 1983 letter from R. C. DeYoung to Cordell Reed (b):
October 28, 1983 letter from R. C. DeYoung to Cordell Reed (c):
October 12, 1983 letter from D. G.
Eisenhut to Cordell Reed
)
Dear Mr. DeYoung:
References (a & b) provided the results of the integrated design inspection conducted in May end June, 1983 regarding Byron Generating
' Station.
The report of that inspection identified a number of findings, observations, and unresolved items.
This letter contains Commonwealth Edison's response to those items and addresses the general NRC conclusions presented in reference (a).
4 In Reference (c) the NRC also requested that we address the generic implications of the Byron IDI issues with respect to Byron's duplicate units at Braidwood Generating Station.
Most of the IDI issues pertain to duplicate design aspects and the information presunted here adequately addresses both Byron and Braidwood stations.
There are certain issues, however, that, involve field construction and design changes which need to be~ addressed separately for Braidwood Station.
These site-specific issues which will be addressed early next year for Braidwood are Findings 2-1, 3-2, 3-8, 3-10, 3-13, 3-14, 3-15, 4-8, 4-10, 6-4, 6-6, and Unresolved Items 2-3, 3-5, 5-1, and 6-1.
At the outset, we would like to thank the members of the IDI Team for-the professional'and courteous manner in which they conducted the inspection.
This inspection occurred during a period of intense activity on our part and the IDI Team made concerted efforts to avoid disrupting our routine work.
Nonetheless, it was a thorough and pain-staking inspection and required a considerable manpower commitment by B401100275 831230 19h PDR ADOCK 05000454 I lII G
PDR j
. Commonwealth Edison, Sargent and Lundy, NPS, and Westinghouse.
It is estimated that approximately 8000 manhours will have been devoted to the inspection itself.
Naturally, in a review of such scope and depth, communication difficulties are bound to arise.
In spite of such difficulties, the IDI 7eam made a commendable effort and accomplished a great deal in a relatively short time.
Attachment A to this letter contains Commonwealth Edison's response to the 96 specific findings, unresolved items and observations chich were identified by the IDI team.
In some cases, new information is presented to support our conclusions regarding specific issues.
Reference (a) also requested that Commonwealth Edison address three general areas of IDI findings and the possible need for additional design audits.
These issues were reviewed with NRC management personnel on December 13, 1983.
Substantial agreement has already been achieved regarding our responses on these general issues, as outlined below.
To fully appreciate our responses on these general areas it is necessary to review the responses to the relevant specific findings, observations and unresolved items.
General discussions are also provided in the areas of electrical separation and instrument setpoints.
Line Break and Flooding Analyses With regard to the analyses of postulated cracks and breaks in high and moderate energy piping and internal flooding, the IDI Team was unable to conclude that the design effort was complete, adequate, and controlled.
This matter has been reviewed in considerable dctail and we do not concur in that overall assessment.
We find that this aspect of plant design work is generally adequate and is being performed in a manner consistent with the FSAR commitments.
Responses to specific deficiencies are provided in Attachment A but these relatively minor problems are not indicative of systemic weaknesses.
None were found to be significant from a sa fety standpoint.
Our review confirms that separation of redundant and diverse ungineered safeguards equipment has been a fundamental element of the L'yron/Braidwood design.
This approach provides inherent protection in the evant of a crack or break in a high or moderate energy pipe.
It also minimizes complex engineering analyses of pipe break consequences and makes the plant less sensitive to minor changes in pipe routing and potential break locations.
The Standard Review Plan clearly advocates separation as the primary made of protection and the Byron SER confirms the acceptability of this approach.
The information necessary to determine the adequacy of the plant design with respect to the effects of high and moderate energy line breaks exists in studies completed prior to the IDI.
A planned report chich provides a road map to this data and explains its application was not completed at the time of the IDI.
This report,
" Jet Impingement and Water Spray Documentation Summary" will be completed in January 1984.
.~.
. For jet impingement, the Byron /Braidwood FSAR describes the to deal with postulated pipe breaks. analytical methods to be applied when sepa The response to Finding 2-16 delineates the work done by Sargent and Lundy to assess the potential impact of jet impingement.
It also delineates the extent of the documentation which was available at the time of the IDI.
clearly show the adequacy of the Byron design, additional documentation To more is now being prepared.
This documentation will be in the form of an impingement on each safe shutdown component and examines t results of jet im active failures. pingement damage in conjunction with postulated single Jet Impingement Ef fects on Safe Shutdown Equipment," is expected to b completed by the end of January 1984.
l' review has confirmed that the approach taken satisfies the FSARWith r commitments.
The response to Finding 2-17 contains a detailed description of the approach taken to evaluate moderate energy line failure and, in particular vulnerable to water spray.,. explains in detail why the plant is not Flooding, which is also the subject of Findings 2-18 and 2-19, l
- However, was evaluated and documented prior to the IDI.
better document the engineering judgements origintily made.the auxiliar Our review has indicated that a total of six findings and one unresolved item may have been pertinent to the recommendation for the review, audit and corrective action program made by the IDI Team.
The findings were 2-14, 2-15, 2-15, 2-16, 2-17, 2-18, and 2-19 and the unresolved item was 2-3.
On the basis of the foregoing discussion and the results of additional reviews described in the individual responses, ce conclude that further audits in this area are unwarranted.
l Machanical Systems Design Calculations After reviewing some of the mechanical systems design calcula-l tions, the IDI Team was unable to conclude that valid, updated analyses I
are generally available.
After further review, we agree that mechanical systems designs are generally adequate although documentation of certain calculations should be reviewed.
We found that hydraulic ef fects of dosign changes have always been adequately considered.
dasign changes have been made without documentation of superficialIn some cases, l
ef fects upon the hydraulic aspects of system design.
Normal conservatisms and design margins have been shown to adequately encompass the effects of such enanges.
. To address the IDI concerns relative to aechanical systems calculations, all safety-related calculations (approx. 100) in the Project Management Division (PMD) calculation books (Books C-1A, C-18 and 0-lC dated prior to October 1983) are being reviewed to verify that if they are technically adequate to support the current Byron /Braidwood design and to determine if the format conforms to tne applicable version of GQ 3.08 in ef fect at the time the calculations were performed.
These reviews are being done in accordance with the requirements of an approved project instruction.
To date, approximately 80% of the calculations have been reviewed and no hardware changes have resulted from the reviews.
All calculations are expected to be completed by January 20, 1984.
No hardware changes are expected to result from the remaining reviews.
Additionally, a review of all safety-related systems is being made with respect to maximum pressure as noted in the response to Unresolved Item 2-2.
No piping changes are expected to result from this review.
Our review indicated that eight findings and one unresolved item may have been pertinent to the IDI Team's recommendation for a systematic review and update of the mechanical systems design calculations.
The findings were 2-5, 2-6, 2-7, 2-8, 2-9, 2-10, 2-11 and 2-20 and the unresolved item is 2-2.
As discussed above and in the attached responses, specific reviews have been undertaken in response to some of the findings.
Because no safety-significant deficiencies have been identified, we conclude that existing design control measures are generally adequate without further audits.
Reviews presently underway are considered confirmatory and not essential to plant licensing.
We have every reason to believe that no plant changes will be required as a result o f these reviews.
Instrument Setpoints The IDI report noted the lack of documented bases for instrument setpoints, reset values, accuracy requirements and margins.
The actual design was found to be generally sound but improvements in the documentation were suggested.
To address this concern, S & L will make an assessment of the safety related instruments in their scope, to identify the instruments which are complex in application and scope.
A documented calculation will be provided for those instruments identified, if one does not adready exist, to verify the adequacy of the setpoints (Finding 6-3) and will include a verification of the setpoint accuracy (Finding 6-7),
and the reset value (Finding 6-8), if applicable.
. Electrical Separation We agree with the IDI Team's determination that some of the engineering analyses utilized to justify exceptions to the electrical separation criteria did not individually document all of the potential means by which Non-Class lE cables might degrade Class lE cables.
However, this does not mean that all of these potential means were not considered in the analysis.
Certain Commonwealth Edison Company generic design practices are very conservative, well-known to Sargent & Lundy,)and were not included in eacn individual analysis (Interface Review Report because they are documented elsewhere.
Examples of such conservative practicer relating to fault current and voltage transients in Non-Class lE circuits (i.e., the use of Class lE qualified caules for Non-Class 1E applications, the use of 600 volt cable for 120 volt applications, and the use of " oversized" conductors for control applications) are described in the response to Finding 5-1.
Westinghouse balance of Plant Piping Design The IDI Team was unable to review enough of the Westinghouse piping design work to make any conclusions.
The Team raised some questions and recommended that additional review work be done to look for indications of systematic weaknesses.
We have conducted additional reviews as documented in the attached responses.
It is apparent from our review of the IDI Team's findings, and from the additional reviews, that there are no systemic deficiencies in the control of tnis work.
There are a number of isolated discrepancies between licensing commitments and detailed plant implementation.
These isolated iristances are all minor in significance and do not contradict the overwhelming evidence of good engineering practice employed in the conduct of this work.
A review of the individual safety implications of each item has had been conducted with the conclusion that the a g cgate of these items,f the they not been noted, would have had negligible impact on the safety o plant.
Three categories are identified es describing the nature of the unresolved item or findings made by the IDI Team:
Obvious Conservatism -
In these cases, the analytical work was performed in such a manner that certain ef fects of weights, loads and other parameters potentially affecting the analytical result were not explicitly addressed because they were insignificant.
Evidence to support these actions has been included in our responses.
Our review indicates that one unresolved item (3-4) and three findings (3-13, 3-14, 3-17) are included in this category.
. In-Process Work - Several unresolved items and findings addressed issues which would automatically be rectified or corrected once the entire design process is completed.
In short, the IDI Team inferred I
inconsistencies and errors while viewing the design process as yet uncompleted.
Our review indicates that one unresolved item (3-6) and three findings (3-12, 3-15, 3-16) are included in this category.
Documentation Inconsistencies - In a few instances, the IDI Team found l
inconsistencies in detail between design criteria contained in licensing documents and those of the procedures being followed in the analytical design process.
In these cases, however, good engineering practice was followed and design margins were maintained.
Our review indicates that one unresolved item (3-5) and one finding (3-11) are included in this category.
Future actions will be taken by Commonwealth Edison to amend the Final Safety Analysis Report with design criteria that are consistent with the procedures being used in conducting the work.
These cases are minor in number and do not present serious concerns with respect to degradation of design margins.
No apparent evidence of systemic deficiencies or significant departure from good engineering practice was found in our reviews.
Since our review has indicated that the unresolved items and findings of the IDI report do not constitute a significant safety deficiency or any serious lapse of quality control, we conclude that good engineering practice has been maintained in the balance of plant design uork being performed by our contactor.
Audits of Other Areas In the IDI inspection report, the NRC requested that we address the necessity for conducting audits of design implementation in areas other than those audited by our inspection so as to assure ourselves that deficiencies of similar importance either do not exist or are corrected.
As indicated in the responses to specific IDI findings, unresolved items, and observations, in many cases we have exam,ined other areas to assure ourselves that significant deficiencies do not exist.
Since the IDI, considerable time has clso been spent by our Q.A. Department in reviewing the adaquacy of the responses for IDI finding-unresolved items' and observations.
The lessons learned from this $n,spection will be incorporated into their future audit plans, Here are some examples where our responses include examination of creas other than those audited by the IDI Team:
1.
Findings 2-2, 2-6, 2-7, and 2-10 deal with flow calculations for pumps in the auxiliary feedwater and containment spray systems.
We have reviewed and upgraded where necessary these calculations to verify that they are consistent with the current design.
We have also reviewed and upgraded calculations for the essential service water system and the suction portion for the residual heat removal and containment spray system.
, 2.
Finding 2-4 dealt with pipe whip calculations.
We have performed a review of all calculations for other unrestrained pipe ruptures in the plant and determined that pipe impacts onto concrete walls have been addressed in the calculations.
3.
In response to Finding 2-22, an S & L Project Management Division file index has been generated that indicates that files exist and are located, for the most part, in the individual engineer's offices.
4.
In response to Finding 3-2, all safety-related subsystems requiring a functionality check have been identified.
5.
In response to Finding 3-8, a field procedure has been revised to provide a check to assure that piping supports that have been deleted are removed.
6.
In response to Finding 3-11, all 521 subsystems in the Westinghouse scope were audited by Westinghouse with respect to damping values.
7.
In response to Finding 5-1, all Interface Review Reports have been reviewed and revised where necessary to address all potential means by which a Non-Class lE cable might degrade Class lE cables.
8.
In response to Finding 6-10, the C&I logic diagrams were revised as required to include sa fety-related stamping.
Commonwealth Edison Company Quality Assurance Department has also conducted an in-depth audit of the 66 findings,19
- u. resolved items and 11 observations identified in the above report.
The audit was conducted concurrent with the accumulation and evaluation of information by CECO Engineering, S & L, and Westinghouse which was required in order to respond to the items.
This' approach resulted in Q.A. directly auditing the input used in the preparation of the response to the NRC IDI items.
The Edison audit, which took about 400 man-hours, was directed at assuring that all the facts of the NRC IDI report were properly addressed, that there was adequate documentation to support the Edison response, that the approach to the response for each item was reasonable and appeared technically acceptable, and finally, to determine if there cere any significant programmatic problems associated with the implementation of design activities.
. As part of the audit process, it became clear that many of the NRC findings resulted because of changing practices over the years in documenting assumptions, calculations, approach to calculations and updating design information to reflect actual conditions.
For the most part, the key problem identified by the IDI Team was the manner in which some supporting design information cas documented.
In the early stages of design, documentation was not always as complete as we now require it to be.
Documentation inadequacies for design calculations similar to those identified by NRC in the IDI report were also found during our past audits.
In the specific cases challenged in our audits, re-calculations were required to be made during the audits to demonstrate the adequacy of the design.
Although the initial calculations were not fully documented, it was proven to our satisfaction by such recalculations that the designs were adequate and met the design requirements.
In the upcoming Edison audits of the Sargent & Lundy corporate office, chich are generally done about three times a year, continued attention to documentation of design activities will be maintained in our coverage of the total scope of design.
As to our auditing of engineering design, Commonwealth Edison has an extensive record of regular technical design analysis audits conducted of the Sargent & Lundy design in all major engineering disciplines by Quality Assurance Engineers and independent consultants with design experience.
These audits are directed at a broad range of project design activities, including those areas examined in the IDI.
The scope of Edison's Q.A.
coverage is described in detail in the report titled " Commonwealth Edison Company Quality Assurance Statement Regarding Verification of Adequacy of Design and Construction of Byron Nuclear Power Station j
Unit #1," dated April 19, 1983.
(See Pages 13-24; 29-34; Exhibits A,
M, N, P).
Specifically, Commonwealth Edison Corporate Quality Assurance Mangement has given special attention to architect /
I engineer and NSSS vendor design activities.
Audits have covered all areas of Quality Assurance Program and procedure implementation and in many cases went into considerable depth to examine specific designs including evaluation of design bases, computer code validation, and review of calculations.
Problems identified in those audits were pursued to determine root causes and to seek out generic deficiencies.
These comprehensive audits have served Edison tell as a basis for demonstrating that the Byron plant is properly designed in accordance with the FSAR, applicable codes, regulatory guides and standards.
Edison will continue this auditing approach until completion of design activities.
.. Summary The IDI was a significant effort in that it provided an independent assessment of the ef fectiveness of a particular set of design control measures on the Byron /Braidwood project.
Ei inspectors went over a single plant system in great detail.ghteen In numerous cases, the IDI Team looked at design and construction cctivities in systems other than the auxilary feedwater system.
A large number of findings, unresolved items and observations were documented in the IDI report, partially because not all of these issues could be satisfactorily resolved during the inspection period due to time constraints.
The additional information provided in the attached responses responds to all of the issues.
None of the issues raised in the IDI report are significant in the context of design adequacy of the plant.
They deal largely t:lth documentation and analytical techniques and resolution requires no physical changes to the plant.
We view the IDI as an independent confirmation of design adequacy and verification that licensing commitments have been fulfilled.
We believe that this response contains enough information for the NRC to conclude that all IDI icsues have been adequately addressed.
Please contact this office if additional discussion of our specific responses is needed.
Very truly yours, Co wR( R2m-Q Cordell Reed Vice President i
1m i
7668N
O Attachment A Byron IDI Report Response
.i Table of Contents O
Section I
Detailed Responses to Findings II Detailed Responses to Unresolved Items III Detailed Responses to Observations l
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FINDING 2-1:
DIESEL ENGINE AIR INTAKE This finding states that there was no analysis or justifi-cation for the seismic vulnerability of the air intake line for the auxiliary feedwater pump diesel.
The finding indicates ' that-this conflicts with the Auxiliary Feedwater i
Design Criteria which states that a single active failure plus a design basis seismic event shall not prevent the auxiliary feedwater system frcm meeting its functional requirements.
RESPONSE
We do not agree with the conclusions of this finding.
The auxiliary feedwater pump diesel combustion air intake line
-was intentionally routed as close as possible to the common turbine / auxiliary building seismically designed wall ("L" wall) to an area in the turbine building that was judged by 1
the Sargent & Lundy design team to be free of non-safety-related components and equipment that would jeopardize or impair the
(
function of the line during a seismic event.
Although the
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air intake line is routed inside the turbine building, which fr N is not classified as a seismic Category I structure, it was V
recognized by the angineers.that the design of the turbine building is such that it will not fail during a seismic event (re ference FSAR Subsection 3.7.2.8 and the response to Question 13 0. 30 ) ^.
We do agree that at the time of the IDI there was no documented evidence that the seismic vulnerability of a f
.c carbon dioxide tank located approximately 10 feet from the air intake line was considered.
Sargent & Lundy has recently verified'by calculation that this' carbon dioxide tank will not fail _during a seismic event.
Further support can be found p
in the' Zion Probabilistic Safety Study where the fragilities for nor.-seismic tanks f ar exceeded the design basis earthquake.
In a6dition, a documented walkdown of this area has concluded that' there are no additional non-safety-related components IM.the vicinity that will impair the function of the air in-1 take,line.
The stress report for the air intake line, reviewed fZ j
by 'the IDI Team but not referenced in the IDI Report, demon-
.strates that this line was seismically supported and was seismically
<f4 ja"* Y 'd -
L t;We! elieve that the routing of the air intake line b
(proximity g/o do "L"% wall), the calculation verifying seismic capability
_ F of the carbon dioxide tank (per formed as a result of the IDI),
c.
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turbine' building will not fail during a seismic event, and l
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tha; existence of a seismic calculation for the line indicates that the' seismic vulnerability of the air intake line was i
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O We therefore conclude that ~ the design of the air intake for the diesel-driven auxiliary feedwater pump does not violate the Design Criteria for single active failure during a design basis seismic event and it will not prevent the auxiliary feedwater system from meeting its functional requirements.
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B/B sl FINDING 2-2:
FLOW CALCULATION PROCEDURAL ITEMS This finding states that Sargent & Lundy calculation AFJK-1 (auxiliary feedwater flow to the steam generators under accident conditions) had some documentation deficiencies that added to the difficulty of reconstructing the calcula-tions which is contrary to Sargent E Lundy Procedure GQ-3.08.
For example, the bases are not stated for the constant used for pipe surface roughness, the resistance coefficient constants and the static head for the f aul'ted loop.
The revision number is not stated for the drawings used to calculate friction losses.
The calculation does not list the assumptions or input data that must be verified as the design proceeds in accordance with GQ-3.08.
Despite the above documentation deficiencies, the IDI Team con-sidered the calculation to be technically adequate.
RESPONSE
We do not agree with the conclusions of this finding for the following reasons.
Sargent & Lundy Procedure GQ-3.08, Revision 3,
in effect at the time AFJK-1 was prepared, did not require a listing of assumptions or input data that must be verified rs as the design proceeds.
This was added in Revision 4 of Procedure
(
)
GQ-3.08 dated 3-5-79.
Therefore, AFJK-1 did not violate Revision 3 of GQ-3.08.
Calculation AFJK-1 was based on input data and assumptions furnished by Westinghouse to Sargent & Lundy.
The revision numbers of the drawings used for the calculation were not listed; however, a review of the latest drawing revisions indicated that there have been no major piping changes that would affect the validity of the calculation.
The static head used in the calculation is the elevation difference between the auxiliary feedwater pump discharge and the steam generator nozzle.
Pipe friction losses were based on the well-known and widely accepted Williams & Hazens formula.
The friction factor (C) chosen is an engineering judgment.
An engineer often provides margin in the design by using a conservative "C"
factor.
For conservativeness, the calculation used the l
friction factor, C=100 (an input to the Williams & Hazens formula), used for water flow in "old pipes."
Calculation AFJK-1 also reviewed the effect on the results for friction factors C=120 and C=140 (used for new pipe).
The change in-friction factors produced negligible differences in the calculated flowrates.
Although the auditor was apparently not familiar with the Williams and Hazens formula, fluid dynamics literature was pro-vided to him during the IDI to verify the wide acceptance and use of this formula.
Procedure GQ-3.08 does not require documenting Od F2.2-1
B/B well-known eng ineering formulas.
Calculation AFJK-1 is a difficult and complex calculation that we feel can be recon-structed based on the above discussion.
Sargent & Lundy approved the test results for the flow orifice pressure drops in 1981.
At the time (1981), it was judged that the test results had a negligible effect on the calculated flow rates, and therefore, a revision to AFJK-1 was not necessary.
Sargent & Lundy performed a calculation during the IDI to confirm to the IDI Team that the actual orifice data did not affect the flow results.
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F2.2-2
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FINDING 2-3:
BASIS FOR TIME DELAY This finding states that the technical basis was deficient for approval of the addition of a time delay to the trip circuitry of the motor-driven auxiliary feedwater pump.
RESPONSE
We agree that there was no test or analysis performed to deter-mine if a time delay was actually needed for the motor-driven euxiliary feedwater pump.
Tha decision to add the time delay was based on a judgement by the responsible design engineer that the time delay was necessary due to the similarity between the trip circuits of the motor-driven auxiliary feedwater pump and the essential service water pumps (which were tested and found to require a time delay).
The time delay on the trip circuit for the motor-driven auxiliary feedwater pump has been removed because it has been determined that it is not necessary.
Because the time delay has been removed, we have not addreased in detail the questions the IDI Team felt should be addressed relative to whether the f'
15 second time delay would be harmful.
We, however, conclude that pump damage would not have occurred because of the amount
%s of water in the condensate storage tank available to provide adequate suction pressure.
i-F2.3-1 1-i
____.__m__._
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, p FINDING 2-4:
ADDITION OF A TIME DELAY ON LOGIC DIAGRAM The finding describes a field change request (FCR) which, i-although not specific to the auxiliary feedwater system, resulted in a schematic diagram revision to the motor-i driven auxiliary feedwater pump trip circuit on starting of the pump but did not indicate the corresponding change on the pump's logic diagram.
The finding also states that the responsible Project Management Division engineer had not been informed of the change.
RECPONSE We agree with conclusion of the IDI Team that a logic diagram had not been updated.
The design process early in the project requires that logic diagrams be prepared first, with electrical schematic diagram preparation following based on the logics.
4 Later in the project life, as construction proceeds, design changes to modify the design are requested by the field on FCR's.
FCR's are written referencing drawings issued for i
construction, or in this case, electrical schematic diagrams.
Logic diagrams are not issued for construction.
FCR's received by Sargent & Lundy are reviewed with the responsible engineers s
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to determine the acceptability of the design change.
If the FCR is approved, the drawings referenced therein are revised to close the FCR.
Subsequently, other affected documents such as-logic diagrams are revised.
As such, at any point in time, logic diagrams revisions may lag the schematic diagrams revision.
Contrary to the finding, the details of the design change relative to FCR-21265 were discussed with the Project Management Division at the time the FCR was received.
Also contrary to the finding, since the FCR was only written against the motor-driven essential service water pumps, design drawings for the auxiliary feedwater pump would not (and should not) have been referenced therein.
A design change to the trip circuit of the motor-driven auxiliary feedwater pump controls was made for the reasons described'in the response to Finding 2-3 and not explicitly due to the FCR.
Commonwealth Edison Company, with Sargent & Lundy concurrence, has-decided to remove the time delays as it has been deter-mined that they are not required.
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F2.4-1
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FINDING 2-5:
RECIRCULATION ORIFICE CALCULATION The finding describes a safety-related, preliminary orifice plate calculation which was never completed although signed off.
RESPONSE
We agree with the conclusions of the IDI Team that an orifice plate calculation was incomplete although signed off, and that this is contrary to procedure GQ-4.3 (current procedure is renumbered as GQ-3.08).
Upon further investigation, it was explained to the IDI Team that the calculation in question was incorrectly classified as safety-related.
The orifice plate referenced in the calculation is not and was not safety-related.
Subsequent to the IDI the applicable non-safety-related calculation for the recirculating orifice plate was found in the non-safety-related calculation book.
The IDI Team found the error after reviewing the Control and Instrumentation Division, Byron Project, safety-related calcu-lation book which contained twenty calculations.
No other problems were cited by the IDI Team, tF' As a follow-up to the IDI Team review, Sargent & Lundy has performed an additional review of the twenty calculations and confirmed that no other incomplete calculations exist.
The incomplete orifice calculation represents an isolated error and has been removed from the safety-related calculation book and nullified.
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l FINDING 2-6:
NET POSITIVE SUCTION HEAD FOR AFW This finding states that Sargent & Lundy Calculation AFJD-1 is deficient as a documented basis for determining that adequate NPSH is available to the auxiliary feedwater pumps and violates Procedure GQ-3.08.
i l
RESPONSE
l We do not agree with the conclusions of this finding.
While we agree that Calculation AFJD-1 is not as well documented as desired by the IDI Team, we disagree that this calculation was technically deficient for determining that adequate NiSH is available to the auxiliary feedwater pumps.
AFJD-1 calculated the minimum and maximum NPSH available to the
~
auxiliary feedwater pumps based on a proposed revision to the suction piping from the condensate storage tank.
The results of this calculation provided sufficient information to the Sargent & Lundy designers, relative to NPSH, to provide a design change to the suction piping.
Calculation AFJD-1 established that 21 feet of NPSH is available based on an i
empty condensate storage tank.
The Sargent & Lundy designers recognized that adequate NPSH would be available because 2 n' '.
feet of water was all that was required in the 45-foot tall condensate tank to provide the required 23 feet of NPSH.
Normal makeup to the condensate storage tank is initiated at a tank water level of 26 feet which corresponds to approxi-mately 283,000 gallons of water available above the auxiliary feedwater pumps required NPSH.
Note that an additional design feature is the automatic switchover to essential service water on low suction pressure to insure that water is available 1
for the auxiliary feedwater system should adequate NPSH from the condensate storage tank not be available.
Calculation AFJD-1 has been superceded and replaced with Calcu-lation AFTH-01 (Reference 2.98) which was reviewed by the IDI Team.
Calculation AFTH-01 is based on the current piping arrangement and includes all pipe fittings, branch tees, and pipe entrance losses.
Based on the current piping arrangement, it was determined that calculation AFJD-1 did not neglect the friction loss from six branching tees.
The documentation defi-ciencies noted for AFJD-1 did not affect the validity of the calculation and did not result in a design deficiency.
4 a
F2.6-1
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N#
FINDING 2-7:
NET POSITIVE SUCTION HEAD FOR CONTAINMENT SPRAY Several alleged deficiencies, given below, were noted in Calculation CS-5 which determined the available NPSH for the containment spray pumps and residual heat removal pumps:
1.
a.
Assumptions are not listed.
Input data and/or assump'lons that must be verified b.
c as the design proceeds are not identifed.
2.
The assumed piping and valve arrangement does not represent the current piping configuration.
3.
The calculation does not account for a partially blocked screen per Regulatory Guide 1.82.
RESPONSE
We do not agree with the conclusions of this finding for the following reasons.
Although the input for Calculation CS-5 did differ in some respects from the current configuration
(~')
of the containment spray system, the calculation is adequate
[m/
to determine that the system will operate properly.
The system changes, subsequent to the calculation, all served to increase the NPSH margin.
Since no potential existed for a system design deficiency, a revision to the calculation serves only to quantify the NPSH margin, not to determine adequacy.
The alleged deficiencies can be addressed individually:
1.
a.
We agree that the assumptions are not listed separately.
However, the input to the calculation is clearly shown and is not ambiguous.
The calculation was reviewed and approved in accordance with Sargent & Lundy Quality Assurance requirements.
In addition, Calculation CS-2, dated 8-25-83, has been performed to verify that adequate NPSH has been provided for the RHR and CS pump 1B.
This calculation (CS-2) has a listing of design informatica and pipe flow parameters.
b.
This requirement was implemented in Revision 4 (3-5-79) to Sargent & Lundy QA Procedure GQ-3.08 and therefore, was clearly not applicable to Calculation CS-5 (7-31-75).
In addition, the responsible engineers judge whether or not a system requires a review to determina if reverification of the adequacy of the design is required.
This judgment is made based on the type of change (whether the change increases or decreases the margin),
s) and the severity of the change (whether a large decrease i
in the margin will occur).
F2.7-1
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()
2.
Revision to the design of the containment spray and RHR pump suction piping systems since 1975 have served to increase the available NPSH.
Only two potentially signifi-cant differences exist.
The suction line from the contain-ment sump has been changed and the centerline of the pump suction connection has been changed.
Calculation CS-5 included 20-inch sump suction piping and associated fittings and valves.
The current design utilizes a 24-inch suction line.
A revised calculation (Calculation CS-2) shows that the flow losses in the piping are in fact reduced, as a result of these changes, by over 1.5 feet in the containment spray system and in the RHR system.
The final pump design and installation resulted in a pump suction centerline approximately 2 feet lower than utilized in Calculation CS-5.
As a result, the NPSH is about 3.5 feet greater than calculated in CS-5.
3.
At the time Calculation CS-S was made (July 1975), the screen arrangement in the plant did not conform to Regulatory Guide 1.82 (June 1974).
An additional screen was added in 1982.
Regulatory Guide 1.82 states "The effect of partially blocked screen should be considered in the evalu-ation of the overall NPSH."
However, the preceeding sentence in Regulatory Guide 1.82 states "For the recommended design
(}
velocity at the fine inner screens considered in this
(/
guide, a negligible pressure drop is anticipated across the screens."
It is not clear why omission of a negligible quantity from a calculation constitutes a deficiency.
Nonetheless, revised calculation CS-2 does include losses for a partially blocked screen.
The screen loss has in-creased from 0.017 feet in Calculation CS-5 to 0.031 feet in Calculation CS-2 (additional pressure drop of 0.014 feet [0.0061 psi]).
In summary, we do not agree that Calculation CS-5 is deficient as a basis for determining that adequate NPSH is available to the containment spray and RHR pumps.. CS-5 did not totally reflect the existing system but it did conservatively represent it.
Calculation CS-2 has been completed to quantify the NPSH margin available.
O
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F2.7-2
B/B O(Q FINDING 2-8:
MISSING CALCULATION FOR CONTAINMENT SPRAY FSAR Section 6.5.2 contains a detailed discussion of the NPSH available for the B train containment spray pump.
No calculation was available to support the January 1979 revision of this section.
The lack of an available calcu-lation was contrary to Procedure GQ-3.08.
RESPONSE
We do not agree with the conclusions of this finding.
the
" missing" calculation in question is not a design calculation because the purpose of GQ-3.08 is to " describe the quality assurance requirements for the preparation, review and approval of calculations that support the design of safety-related structures, systems, and components."
The information in FSAR Section 6.5.2 did not provide the basis for, or in any way support the design of the containment spray system.
As explained in response to Finding 2-7, Calculation CS-5 provided an adequate basis for the system design.
During the IDI, Calculation CS-2 was completed which quantifies the NPSH nargin based on as-built data.
This calculation shows that the final design provides more margin than indicated by the PSAR description.
O The FSAR is being revised to eliminate outdated information.
b-/
The FSAR description is of a preliminary system layout which was slightly modified to provide additional NPSH.
Inclusion of this description or calculation in the FSAR is not required by Standard Review Plan Section 6.5.2 and was not required to meet any licensing commitment.
The information wss not the basis for any safety-related design and was not part of a licensing commitment.
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FINDING 2-9:
FSAR DESCRIPTION OF CONTAINMENT SPRAY The discussion of the containment spray system in FSAR Section 6.5.2 is not consistant with the current design.
Although the specific items did not constitute deficiencies of technical significance, the FSAR description and the design should be consistent.
RESPONSE
We agree with the conclusion of the IDI Team that the FSAR description and the design should be consistent.
We also agree that the noted discrepancies are not of technical significance.
As explained in the response to Finding 2-8, the FSAR description was not used as design input and was not required to support licensing commitments.
FSAR Section 6.5.2 will be revised to eliminate the outdated information.
The FSAR will be updated to reflect the change.
l O
F2.9-1
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~ ~ ~ ' ' ' " ~ ~
~ ~ ~ ~ " ~ ~
l S/B FINDING 2-10:
CONTAINMENT SUMP SCREEN VELOCITY A calculation to support the design of the containment sump screens could not be found during the inspection.
Calculation SI-03 was performed during the inspection and indicated a worst case velocity of 0.24 fps in contrast to the recommended velocity of 0.20 (Regulatory Guide 1.82).
RESPONSE
We agree with the IDI Team that the calculation could not be found during the inspection.
The missing calculation (CS-012 dated 4-6-83) was unavailable during the inspection because it was sent to the Sargent & Lundy Quality Assurance Division for microfilming as required by procedure.
Calculation CS-012 was a design calculation performed to determine the required screen area.
Examination of the calculation revealed the reason for the difference in flow velocities (0.24 fps in contrast to 0.20 fps).
Calculation SI-03, which was completed very quickly during the IDI to demonstrate to the IDI Team the adequacy of the design was overly conservative in that the area occupied by stiffeners and braces was not considered to be flow area.
Calculation CS-012 included the area of b
these small but numerous supports under the assumption that the approach velocity, a short distance from the screen, is more indicative of the likelihood of debris to settle out than the velocity in the screen itself.
Calculation SI-03 has been revised such that the assumptions and bases are con-sistent with Calculation CS-012.
The velocity with a 50%
screen blockage is actually calculated to be 0.16 fps.
This demonstrates the design to be adequate.
4 O
F2.10-1
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FINDING 2-11:
MAXIMUM PIPING PRESSURE This finding stated that Sargent & Lundy had not performed a calculation to determine the maximum pressure of the auxiliary feedwater system to assure that the piping is acceptable.
The finding states this is contrary to ASME Code,Section III, Subsection HD-3612.4 which states that pump discharge piping shall be designed for the maximum pressure exerted by the pump.
RESPONSE
We do not agree with the conclusions of this finding.
The values for auxiliary feed pump discharge maximum pressure are as follows:
Design Pressure Maximum Operating (psig)
Pressure (psig)
Design Criteria 2080 Mechanical Department Piping Line List-1750 2080 I \\
S&L Wall Thickness h/
Calculation 1750 Design Specification DS-AF-01BB*
1750
- Test Loads - Section 404 of DS-AF-OlBB indicates "The auxiliary feedwater pumps will be tested at 100 gpm and 2080 psig."
The ASME Section III Code (1974 Code, Summer 1975 addenda, is the applicable code for Byron) defines tne design pressure as follows:
ND-3112.1 Design Pressure
" Components shall be designed for at least the most severe condition of coincident pressure and temperature expected in normal operation."
The normal operation of the auxiliary feedwater system is when water is being supplied to the steam generators.
Under this operating condition, the most severe condition of pressure is 1750 psig.
O F2.ll-1 l
i
)
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I)
The test load represents the maximum operating pressure (2080 psig) at which the auxiliary feedwater system would be normally operated, other than the design condition.
i l
In April of 1982 the Byron project decided to consolidate and reverify the minimum wall requirements for the pipe.
' Confirmatory calculations were performed at that time based on design pressures.
The calculations indicate the maximum required aall thickness (t per the Code for the design pressure and the nominal wall thickRe)ss actually provide for the piping.
For the auxiliary feedwater system piping, these calculations indicate:
Code Calculated Minimum Wall (t Wall Thickness Required for a,)
Provided Design Pressure of 1750 psig Nominal Wall Minimum Wall (t,)
4" Sch 120 pipe 0.251 inches 0.438 inches 0.383 inch 6" Sch 120 pipe 0.369 inches 0.562 inches 0.492 inch If these wall thicknesses are converted to maximum allowable design pressure (this is readily done for nominal wall using Sargent & Lundy Standard MES 2.5 [IDI Ref. 2.64]), the infor-mation would be as follows:
MINIMUM WALL-4" PIPE 0.251 inch 0.383 inch Code Calculated Wall Thickness Min. Wall Provided (t )
Required (t )
m Maximum Design Pressure @l5,000 1750 psig 2742 psig psi allowable
_ stress MINIMUM WALL-6" PIPE 0.369 inch Code 0.492 inch Calculated Wall Thickness l'
Min. Wall Thickness Required (t,)
Provided (t,)
Maximum Design Pressure @l3,000 1750 psig 2367 psig psi allowable stress O
We do not agree that non-conservative values were used for -
wall thickness calculations.
We believe that design conditions is-F2.ll-2 4
-.-------,.--r--
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,..a.- - - -
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were identified as required by the applicable Code and that substantial margin was provided for the piping.
The IDI Team indicated that they "would consider it more appro-priate to use, concurrently, a condition where suction is taken from the essential service water system, which would result in a pressure of 2165 psig."
This condition would be extremely unusual and if it did occur, would occur less than 14 of the time.
Nonetheless, it is only 85 psig greater than what has been identified as the test load (2080 psig).
It can be seen by inspection that the 2080 psig test load and the 2165 psig postulated by the IDI Team are well within the capability of the piping system. This was indicated by the IDI Team in the IDI Report as follows:
"However, the installed piping is adequate since it is rated for 2367 psig based on the allowable working pressure listed in Sargent & Lundy (S&L) Standard MES-2.5 for 6-inch schedule 120 pipe (Reference 2.64)."
It should be noted that S&L engir r ars have access to and utilize the above-mentioned S&L standard This S&L Standard is a document where code maximum allouable working pressures are delineated for various pipe sizes, schedules, and materials.
The IDI Team indicated that the calculations performed were contrary (relative to the IDI Team's postulated "more appro-priate pressure") to the ASME Code,Section III, Subsection HD-3612.3, which states (as indicated in the IDI Report) that the pump piping shall be designed for the maximum pressure exerted by the pump.
The complete code statement is:
" Pump discharge piping shall be designed for the maximum pressure exerted by the pump at any load and for the highest corresponding temperature."
1 We conclude that we have satisfied this requirement.
The pump discharge piping provided (nominal wall pipe) has a code design maximum pressure capability in excess of the postulated pressure.
The Code does not define this postulated pressure I
as the " design pressure" (as indicated earlier, the design pressure definition in accordance with ND-3112.1 is 1750 psig)r nor does it require that this pressure be used in determining the minimum wall.
In addition, the following Code section states:
"ND-3612.3 Allcwance for Variations From Design Conditions O'
(a)
It is recognized that variations in pressure and temperature inevitably occur and therefore the piping system shall be considered safe for occasional operation F2.ll-3
B/B O
for shot periods at higher than the design pressure or temperature as limited in (b).
(b)
Either pressure or temperature, or both, may exceed the design values if the stress in the pipe wall calcu-lated by the formulas using the maximum expected pressure during the variation does not exceed the S-value allowable for the maximum expected temperature during the variation by more than the following allowances for the periods of duration indicated:
(1)
Up to 15% increase above the S-value during 10% of the operating period; (2)
Up to 20% increase above the S-value during 1% of the opening period."
Note:
There is a typographical error in the Code, 14 of the opening should be 1% of the operating period.
The postulated 2165 psig that the IDI Team considers more appropriate would fall under ND-3612.3 which allows a 20%
increase above the S-value (allowable stress) if.the duration is expected 1% of the operating period.
We have demonstrated O
that this condition can be met with no code permissible in-crease in allowable stress.
This is clearly conservative.
Information was forwarded to IDI Team after the IDI by letter dated July 5, 1983 transmitting additional information for calculating minimum wall thickness.
This information included a table provided by the Piping Fabricator.
We indicated that it was a practice (non-required, non-safety-related function) of the Fabricator at the time (1975) to do a minimum wall calculation.
The Fabricator used the larger of the maximum operating pressure or design pressure in his calculation.
At the time, the maximum of the two pressures was 1830 psig.
The Piping Fabricator did include, as part of his tabulation, the maximum calculated pressure for the piping.
His tabulation indicates that the piping is capable of 2369 psig for 6-inch Sch 120 pipe.
The purpose of sending the information to IDI Team was to indicate that the piping is checked in several non-documented ways (in addition to documented ways) and to show that it is not uncommon to establish maximum calculated pressures for piping.
The maximum calculated pressure indi-cates for the designer, and in this case the Piping Fabricator, the margin that exists in the design that can be used as a comparison for unusual or postulated operating conditions.
The fact that the Piping Fabricator had calculated the maximum pressure was not mentioned in the IDI Report.
O F2.ll-4
B/B k
Sargent & Lundy Standard MES-2.5 is a tabulation of maximum calculated pressure for nominal wall pipe.
This tabulation for the piping under discussion was also forwarded to the IDI Team by the above-mentioned letter and was referenced in the Report.
The tabulation showed, as noted in the IDI Report, a design pressure for the 6-inch Sch 120 pipe of 2367 psig.
The concern of the IDI Team was probably caused because the minimum wall calculation was based on the design pressure i
(1750 psig) which is lower than the maximum operating pressure (2080 psig).
Although we conclude that this is consistant with the Code and that adequate margin exists, we have reviewed the entire Pipe Line list to verify that the piping is capable of with-standing the greater of the design and operating pressures.
There are only 12 safety-related lines where the operating pressure is identified to be in excess of the design pressure as noted below:
Max.
Line Design Operating Pressure
("')
Size Pressure Pressure Pipe Capability
\\s,/
Quantity (in)
System (psig)
(psig).
.ccha 0. t 3 (psig) l 4
4 FW 1750 1855 160 3600 t
4 6
FW 1750 1855 120 2367 4
1 SI 2485 2735 160 6199 l
In addition, as indicated in response to Unresolved Item 2-2, the maximum pressure will be identified for all cafety related piping and the ability to satisfy code pressure considerations will be addressed and documented for the maximum pressures.
l l
)
s-F2.ll-5 l
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. - - - -, - - - ~ - - -
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FINDING 2-12:
DESIGN REVIEW CLOSEOUT This finding indicated that Sargent & Lundy procedures and standards do not adequately describe the method for completing system design reviews.
Procedure GQ-3.10 states that the department standards shall describe the method for verifying and documenting the resolution of open items and discrepancies.
Mechanical Department Standard MAS-4 does not describe tha method for verifying and documenting that correction of discrepancies has.been completed.
RESPONSE
We agree with the conclusion of the IDI Team that a revision to MAS-4 would clarify the required method for verifying and documenting that correction of discrepancies has been completed.
Mechanical Department Standard MAS-4 states that the Mechanical Project Engineer shall maintain a file of all design review records for the project as well as ? list of all unresolved items.
The Mechanical Project Engineer assigns the unresolved 1
items to a Responsible Engineer and documents when each item is satisfactorily resolved.
The System Design Review Status Report is prepared and issued by the Mechanical Project Engineer.
l' A system review designated as " completed" documents that all open items have been resolved.
A commitment to update a design' drawing or document was considered by the Mechanical Project Engineer to be a satisfactory response and a resolved item.
EIt was not required to maintain the system review open until the drawing or document was actually released.
Sargent & Lundy is currently reviewing the Byron /Braidwood system design reviews and the Responsible Engineers are advising the Project Manager that all open items have been, in fact, resolved and that any commitments to update drawings or documents have been completed.
In addition, Sargent & Lundy Mechanical Department Standard MAS-4 is being revised to indicate that a system design review shall not be designated as completed on the status report unless all required documents are revised or a follow-up close out system is provided.
1 i
- y F2.12-1 i
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FINDING 2-13:
DESIGN CRITERIA UPDATING This finding indicated discrepancies between the auxiliary feedwater system Design Criteria DC-AF-01-BB and the actual installation.
These discrepancies were summarized as follows:
1.
Design criteria states that the minimum auxiliary feedwater flow rate supplied to one or more unfaulted steam generators is 470 gpm.
Sargent & Lundy calcu-lation AFJK-1 calculated 459 gpm which was used in the Westinghouse accident analysis and documented in Chapter 15 of the FSAR.
2.
The design criteria references a pressure drop across the flow restricting orifices of 155 psi based on a flow of 160 gpm.
Calculation AFJK-1 used a pressure drop of 230 psi at 160 gpm based on data points from the Daniel flow orifice calculator.
Data from the orifice supplier indicated the pressure drop to be 197.7 psi for 160 gpm.
3.
The design criteria states that the developed head of the auxiliary feedwater pump under minimum flow conditions is 4800 feet while the actual pump perfor-
[
mance curve indicates a developed head of 4700 feet I
under minimum flow conditions.
This finding also requested that the general practice of not updating design criteria be addressed.
RESPONSE
We do not agree with the conclusions of this finding.
The Byron /Braidwood design criteria were prepared and issued as the basis and starting point for the initial system designs.
Design criteria are not intended to be up-to-date system des-criptions, nor are they intended to reflect the final design.
The design drawings and documents themselves do this function.
The following responses address the individual IDI Team's concerns regarding the auxiliary feedwater system design criteria:
1.
The auxiliary feedwater design criteria was based on the recommended Westinghouse flow rate of 470 gpm to the intact steam generators.
Although AFJK-1 calculated 459 gpm (which was accepted by Westinghouse), the design criteria remains as 470 gpm.
The deviation from the design criteria was documented in the FSAR (prior to the IDI) and no change to the design criteria is necessary.
F2.13-1
B/B D
2.
The pressure drops listed in the design criteria for the flow orifices and pipe friction were estimated values based on the preliminary piping layout to demonstrate-the derivation of the auxiliary feedwater pump discharge i
head requirements.
Calculation AFJK-1, based on the actual piping configuration, determined there was less pipe friction pressure loss and therefore more pressure drop across the orifice.
The adequacy of the auxiliary feedwater pump was verified based on the actual differential pressure of the system.
Purchase of the flow orifices was based on a specified beta ratio, not on the differential pressure listed in the design criteria.
As stated in the response to Finding 2-2, the actual flow orifice test data had no impact on the calculated flow rates.
3.
The pump developed head of 4800 feet under minimum flow conditione (100 gpm) specified in the auxiliary feedwater design criteria was based on preliminary information received from the pump manufacturer.
This number is not a design point but was included in the design criteria for infor-mation to assure that the piping is properly designed.
The actual test performance curves provided by the pump manufacturer indicate a developed head of 4700 feet at g{s/}
minimum flow.
The pump head / flow condition has been found through existing calculations to satisfy system design requir ements.
We recognize that some design criteria have been updated to reflect modifications made to the design while others contain either outdated or obsolete information.
To eliminate potential confusion, the Design Criteria Status Report will be revised to include a status for each design criteria.
The classifi-L cation of the design criteria will address the IDI Team's concern about. future use of these documents or attempted use by someone not familiar with the actual status of a particular design criteria.
In general, modifications to a system at this time in the project are based on the latest design docu-ments not on design criteria.
i O
I F2.13-2 L.
B/B FINDING 2-14:
PIPE WHIP CALCULATIONS The feedwater piping inside the main steam tunnel has 27 postulated breaks.
The design calculations address 22 of these breaks.
Three of the remaining five breaks did not require any calculations.
For the last two breaks, C22 and C29, there were no calculations and it did not appear obvious that the impacted concrete tunnel walls would be capable of withstanding the pipe impact forces.
RESPONSE
To the original design engineer, it was obvious that the 5 foot 6 inch thick concrete walls being impacted by pipe breaks C22 and C29 identified in EMD File 6535 (Reference 2.34) are structurally adequate based on previous calculations done for more critical combinations of break force and wall thickness.
For this reason, no design calculations were performed.
During the inspection, calculations were performed to confirm this conclusion and submitted to the NRC IDI Team (Reference 2.154).
These calculations proved the structural adequacy of the impacted walls and substantiated the original design engineer's engineering judgment.
()
A review of all calculations for other unrestrained pipe ruptures in the plant has shown that all pipe impacts onto concrete walls have been addressed in the calculations.
Therefore we consider this finding to be resolved.
I f
F2.14-1
I B/B D
FINDING 2-15:
ENERGY DISSIPATION ASSUMPTION The IDI Team reviewed the basis for Sargent & Lundy's 50%
assumption, a report by Chelapati and Kennedy on " Prob-abilistic Assessment of Aircraft Hazard for Nuclear Power Plants" (Reference 2.90).
The report indicates that, for an aircraft striking a nuclear structure, the energy is diminished by the process of deforming the aircraf t.
This is conceptually similar to the situation for a whipping pipe, but we found no direct connection or specific basis for the 50% assumption that had been used in the calculations.
Procedure GQ 3.06 states that the calculation should list the data and assumptions which establish the basis of the design.
In this case the assumption was listed, but an adequate technical basis for the assumption was not established by the listed reference.
RESPONSE
The technical basis for the assumption that 50% of the kinetic energy of a whipping pipe is absorbed by plastically deforming the pipe is a report by Chilapaki and Kennedy on "Probabilistic Assessment of Aircraft Hazard for Nuclear Power Plants" (Reference f],
2.90).
Figure 9 in this report indicates that for plastic F-impact, the maximum absorption factor for the wall is 0.5.
This report was used because the missile effects of the aircraft and the pipe are similar in as much as both of them are relatively soft when compared to the rigid concrete structures they are impacting.
Therefore, the pipe will plastically deform as the aircraft does and the original design assumption is valid.
However, in order to further substantiate the adequacy of the design, subsequent calculations have been made using an alternative approach and have been submitted to the NRC IDI Team (Reference 2.151).
These calculations employ the conser-vation of momentum principle to show that the wall can adequately abs. orb the impact energy without considering the deformation of the pipe.
Since both techniques confirm design adequacy, we consider this finding to be resolved.
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FINDING 2-16:
JET IMPINGEMENT ANALYSIS The Team found that, although FSAR Section 3.6.2.2.2.1.4 indicates, for postulated breaks, how the jet impingement force will be calculated, no jet impingement analysis had not been performed.
The Team found that, although Sargent & Lundy personnel indicated they had intended to do the work, no documented program or procedure was in place to provide assurance that the work would be completed
RESPONSE
We agree that the documentation of the jet impingement analysis was not complete and available.
However we do not agree that a jet impingement analysis had not been done for the following reasons.
FSAR section 3.6.2.2.2.1.4 describes the methodology for calculating jet forces if jet impingement shields are found to be necessary.
It does not contain the procedure for assessing the potential for jet impingement damage.
This is most effectively done by assessing the vulnerability of safe shutdown systems rather than calculating the properties of the jets.
Although no specific analysis of jet effects exists, various studies have been completed which demonstrate that jet impingement effects will not prevent safe shutdown.
OI The " work" that had not been completed was not an additional analysis but an accumulation of information (i.e., a summary
" road map" report) that recreates the engineering considera-tions and assessments that were made and which demonstrates the applicability of the existing studies to the question of jet impingement.
This summary report, " Jet Impingement and Water Spray Documentation Summary," will be completed in January 1984.
These existing studies and documentation can be categorized in three areas:
those which identify and locate high energy line breaks, those which demonstrate the separation and protec-tion of safe shutdown systems, and those which describe the Byron approach to protection against jet impingement.
The following is a list of the more significant studies and documentation (asterisk indicates information made available to the IDI Team) :
1.
Location of High Energy Lines and High Energy Line Breaks a.
High Energy Line Location 1)
FSAR - 1978*
2)
Final Review - 1981 (S&L Calculation)*
b.
High Energy Line Location (Based on Stress) l' 1) 1980 - 1982 (S&L/W Calculations)
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)
c.
HELB Located on Composites 1) 1982 (Informal)
- 2.
Separation of Safe Shutdown Systems l-a.
Color Coded Mechanical Systems Composite - 1975 (Informal) b.
Fire Protection Report - 1977 (Dock eted)
- c.
Safe Shutdown Report - 1981 (Docketed)*
d.
Color Coded Mechanical / Electrical Composites - 1981 (Informal)*
e.
Environmental Qualification Report - 1982 (Docketed)*
3.
Jet Impingement Approach a.
FSAR Sections 3. 6.1.1. 2, 3.6.2.3.2.7, 3.6.2.3.1.2*
i b.
NRC Question Response (10.40) - 1982*
c.
NRC Byron SER Supplement No. 2 - 1983*
In order to more clearly show the adequacy of the Byron plant, O
additional documentation is being prepared which will clearly show the potential for and effects of jet impingement on safe shutdown equipment and components.
This additional report will be structured as an independent document rather than relying heavily on other documents such as the Fire Protection Report.
The input to this additional jet impingement report will be the list of Safe Shutdown Equip-ment, the identified high energy lines, high energy line break-locations, and the plant design drawings showing the system configuration and plant arrangement.
Each physical area in i
the plant which contains safe shutdown equipment will be identi-fled.
Areas with no high energy lines in the proximity of the equipment will be eliminated.
In the remaining areas, an assumption will be made that all equip-ment is disabled j
by a jet, and, in addition, a limiting single active failure occurs.
The capability of the plant to achieve safe shutdown will then be evaluated.
In the event that a safe shutdown path cannot be found, a more detailed analysis of the area will be performed, including a review and evaluation of alter-nate shutdown paths.
The final conclusion is expected to be identical to the summary " road map" report; namely, that
' jet impingement will not jeopardize the ability to safely shut down the plant.
This report is scheduled to be complete by January 31, 1984.
The remainder of this response describes the approach which has been used to address jet impingement and the documentation which was available at the time of the inspection.
F2.16-2
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FSAR Subsection 3.6.2.2.2.1.4 is a description of a mathemati-cal method for calculating the force of a fluid jet from a postulated pipe rupture.
This section contains no commitment l
to perform this calculation ~for all high energy pipe ruptures.
The relevant FSAR commitment is in FSAR Section 3.6.1.1.2:
"The criteria used for protection against pipe whip and the Commission's letter from Mr. Giambuso, dated December 1972, have been met.for designs inside and outside the containment respectively.
By virtue of the Construction Permit date for this plant, the above is the required minimum.
" Subsequent criteria, including that in the Commission's letter from Mr. O' Leary, dated July 1973, and Branch Technical positions APCSB 3-1 and MEC 3-1, have been employed to the extent possible and practical, given the stage of l
design / construction.
Essentially all of the above criteria have been met, with the exception of maximizing optimi-zation of plant layout to provide remote location of potential sources of pipe failure with respect to equipment essential to protect against such failures.
In these few cases, physical separation or whip restraint / impingement barriers have been employed."
Clearly, no jet force calculation is required if protection has been provided by the preferred means of remote location l
of the high energy lines and/or the safe shutdown equipment i
or by physical separation.
In the event a barrier is required to protect from jet impingement, the design approach is described in FSAR Subsection 3.6.2.3.2.7 for piping other than reactor coolant system (RCS) piping:
" Jet impingement shields are provided as required to protect safety-related equipment and components.
"To account for these effects in the design, a combination of component restraints, barriers, and layout is utilized to ensure that for a loss-of-coolant, steam or feedwater line break, propagation of damage from the original event is limited, and the components as needed, are protected and available.
Design loads are obtained from Subsection 3.6.2.2.2.1.4.
Allowable stresses are the same as in Subsection 3.6.2.3.2.4."
No jet deflectors were required on non-RCS piping and there-fore, the equations in FSAR Subsection 3.6.2.2.2.1.4 were not used.
The approach to jet impingement analysis for RCS breaks is explained in FSAR Subsection 3.6.2.3.3.3 and is consistent with the commitment in FSAR Subsection 3.6.1.1.2.
}
F2.16-3 e---,-
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Jet deflectors were required at four locations on the reactor coolant piping.
Design of these deflectors is described in FSAR Subsection 3.6.2.3.1.2.
A summary of the jet impingement design approach was provided 1
in the response to NRC Question 10.40 (FSAR page Q10.40-9) :
i "The approach to jet impingement is described in FSAR Subsection 3.6.2.
The break locations defined for the l
pipe whip investigation were also examined for jet impinge-ment effects.
The majority of locations had no effect or. equipment required for safe shutdown.
This was a result of the criteria used in $2 sign to maintain separation i
of redundant systems and the use of compartments to isolate high energy line effects.
Equipment which could be affected by jet impingement was analyzed and moved or protected if protection was required."
The Byron design approach of using separation of redundant safe shutdown systems and diversity of safe shutdown paths is consistent with the NRC guidance in the Standard Review Plan (SRP), NUREG-75/087.
Branch Technical Position APCSB 3-1 (included in Section 3.6.1 of the SRP) states:
I)
"Although various measures for the protection of safety-h-
related systems and components are ou'tlined in this position, I
the preferred method of protection is based upon separation and isolation by plant arrangement."
The Byron approach was reviewed by the NRC staff in preparation of the Safety Evaluation Report (SER), NUREG-0876, and found to be adequate.
Supplement No. 2 to the Byron SER (January 1983) addresses this issue in Section 3.6.1 which states, in part:
"The plant design accommodated the effects of postulated pipe breaks and cracks, including pipe whip, jet impinge-ment, and environmental effects.
The means used to protect essential (safety-related) systems and components include physical separation, enclosure within suitably designed structures, pipe whip restraints, and equipment shields.
To be consistent with BTP ASB 3-1, the applicant used separation as the primary means of protection, where separa-tion was not feasible, one of the other acceptable methods of protection was'used."
The physical separation of safe shutdown systems has been verified by a variety of efforts throughout the design of the plant.
In late 1975, the safe shutdown systems piping and equipment were located and color coded by division on
!~ hm-Containment and Auxiliary Building Piping plan and section drawings.
The Fire Protection Report (FPR) was submitted F2.16-4
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B/B O
in November of 1977.
The FPR analyzes in detail the ability to safely shut down the plant following an initiating event which affects various unrelated systems in an area.
The FPR contains color coded drawings showing safe shutdown equipment, components and cables 6.hroughout the plant.
The FPR was recog-nized to be applicable to the jet impingement issue with two qualifications.
The FPR did not rigorously consider single failure and did not consider LOCA or Main Steam Line Breaks as initiating events.
Ability to withstand single failure is inherent because the FPR did not take credit for diversity in shutdown paths.
Each safe shutdown system is backed up by another safe shutdown system or combination of systems that can perform an equivalent function.
Since these backup systems are,also checked for separation of redundant trains in the FPR, safe shutdown capability is assured.
LOCA and MSLB inside containment can be mitigated with two systems not~ included in the safe shutdown report, Safety Injection and Containment Spray.
Both of these systems have well separated redundant trains and do not contain active components inside containment.
As a result, jet impingement will not impair safe shutdown capability.
Location of safe shutdown equipment remote from high energy lines and the enclosure of high energy lines in protective s
enclosures (subcompartments and tunnels) also contributes to the safety of the plant.
The initial submittal of the FSAR I
(November 1978) contained piping schematic. diagrams (Figures 3.6-1 through 3.6-12), which located all high energy lines.
The corresponding P&ID's (Piping Instrumentation Drawings) have high energy lines marked with asterisks to alert designers to the need for special consideration of high Energy Line Break (HELB) effects.
At a later stage in the project, it was felt that design was finalized to the extent that high energy line locations should be final.
As a result, Sargent & Lundy calculation 3C8-ll81-001 (December 1981) was completed.
This calculation located all auxiliary building high energy lines and was used to update FSAR Figures 3.6-1 through.3.6-12.
The purpose of this calcu-lation was to ensure that all HELB effects had been considered in the appropriate areas of the auxiliary building.
At approx-imately the same time, a set of piping composite drawings was marked to show postulated HELB locations in the contain-ment and auxiliary building.
l Also, in late 1981 and early 1982, sets of mechanical and j
electrical composite drawings were marked up to show the location and routing of safe shutdown components, piping, and cables.
4 This was dor.3 to provide an even more detailed identification g
of safe shutdown systems and to show the separation of the redundant trains.
2 F2.16-5
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Sh IbI Team was shown all the above-mentioned F5AR sections, the set, ~ drawings showing tha high energy line break locations and" safe shutdown components, the FPR, and Sargent & Lundy Calculation 3C8-ll81-001.
The only additional effort planned in this area was a summary report which documented the appli-cability of_these studies and surveys to the jet impingement issue.. The location of breaks was done not for the purpose of defining jet properties but to determine the general areas in' which jets could occur.
The Byron approach was to identify equipment required to safely shutdown the plant and to verify that redundant trains of this equipment were not susceptible to common damage by a single high energy line break rather than examining the effects of all postulated high energy line breaks.
This approach eliminates most uncertainty associated with location and direction of jets and other HELB effects.
As noted, the safe shutdown capability of the plant following damage-to safe shutdown components is being documented in detail.
These reports will be prepared in accordance with Sargent
& Lundy QA requirements and will be retained in the calculation files.
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F2.16-6
B/B
()
FINDING 2-17 :
MODERATE ENERGY PIPE CRACK ANALYSES FSAR Section 3.6.2.1.2.2 indicates that through-wall pipe leakage crack locations are postulated based on stress levels to maximize effects f rom fluid spraying and flooding.
The Team determined that this had not been accomplished and it did not appear that the work was programmed to be done.
The design cannot be considered adequate in this regard until this work has been done to locate those instances where leakage might damage essential equipment and to protect equipment as appropriate in accordance i
with the licensing commitment.
RESPONSE
We do not agree with the conclusions of this finding because they misinterpret the FSAR commitment and disregard the methodology actually used to evaluate the ef fects of. moderate energy pipe cracks.
This methodology (briefly discussed below), which is very conservative and exceeds the FSAR commitments, was explained in detail to the IDI Team but is not mentioned in i
Section.2.4.4 (Moderate Energy Pipe Cracks) of the IDI Report.
FSAR Subsection 3.6.2.1.2.2. contains a commitment to evaluate moderate energy-line cracks in accordance with Standard Review Plan guidelines:
1 i
"Through-wall leakage cracks are postulated in Seismic Category I moderate-energy ASME Section III, Class 2 and 3 and seismically qualified ANSI B31.1 piping located both inside and outside containment except where the maximum etress range is less than 0.4 (1.2 S + S ).
In unanalyzed w
moderate-energy ASME Section III Clams 2,and 3 and ANSI B31.1 piping, this exception based on stress is not taken.
The cracks are postulated individually at locations that result in.the maximum effects from fluid spraying and flooding, with the consequent hazards or environmental l
conditions developed."
l l
Use of the stress criteria in the FSAR will result in cracks being postulated at only a small portion of the potential crack locations.
In practice, however, the need for protection must be evaluated early in the design process, before the final stress values are available.
As a result, a more conser-vative approach of designing the plant to accommodate a crack at any moderate energy location was adopted.
This effectively bounds the FSAR commitment.
O F2.17 -1
. - - - -, - ~ _
i B/B To evaluate possible flooding effects, moderate energy line j
cracks were postulated at locations which would result in the most severe flooding regardless of stress level.
The limiting locations for poatulated cracks are documented in Sargent & Lundy Calculation 3C8-1281-001, "ALxiliary Building Flood Level Calculations (IDI Report Reference 2.15) and in Sargent & Lundy Calculation RAS-FL-1, " Flood Level Inside Containment."
Calculation 3C8-1281-001 utilized a calcula-tional procedure to determine the crack or break flow rate based on the pipe size, wall thickness and internal pressure.
Each area in the auxiliary building was surveyed and the highest bresk flow rate determined.
The location chosen and the flow rate calculation for each area is documented in an appendix to Calculation 3C8-1281-001.
Section 2.4.5, Flooding Analysis of the IDI Report states "Sargent & Lundy's Nuclear Safeguards and Licensing Division performed an analysis of flooding in the auxiliary building from postulated high-energy and moderate energy line f ailures for the lines that would produce the worst flooding conditions in each area."
This statement and Finding 2-17 are contradictory.
Evaluation of water spray from moderate energy line cracks is not required in areas where effects are bounded by effects
(
of high energy line breaks (SRP Section 3.6.2).
This would eliminate essentially all of the containment.
However, FSAR i
g I
r Subsection 3.6.2.1.2.2 does commit to evaluate effects of l
water spray inside containment.
The capability to withstand water spray is established in the Byron Environmental Qualifi-cation of Electrical Equipment report (June 1982).
All electri-cal equipment inside containment required for safe shutdown is qualified for spray.
Water spray in the auxiliary building will not damage equip-ment such that safe shutdown capability is jeopardized.
This l
is a result of the basic design which separates redundant I
and diverse safety systems.
Water spray is postulated to affect only electrical equipment and not cables or mechanical equipment.
Water spray is also assumed to not simultaneously affect two systems separated by 20 feat or more.
The Byron l
Fire Protection Report (FPR) documents this separation with color-coded drawings showing safe shutdown components as well as tables and descriptions of the locations of these components.
The FPR does not address systems required following a design I
basis event (LOCA and Main Steam Line Break) but does cover all systems necessary for safe shutdown following damage or failure in a system in the auxiliary building.
The FPR does not rigorously postulate single f ailure but it also does not take credit for diversity of safe shutdown paths.
By examining the requirements for safe shutdown and the systems
-s i
F2.17-2 l
B/B utilized to achieve safe shutdown, 'it can easily be seen that O
water spray esulting from failure of a moderate energy line will not result in the inability to safely shut down the plant.
Safe shutdown requires three functions - (1) prevention of release of excessive offsite radioactivity; (2) negative reactivity; and (3) removal of decay heat.
Goal (1) is met for a moderate energy line crack outside of containment because the initiating accident does not breach the primary system and containment isolation is not required.
Goal (2) requires only that the reactor protection system function for hot shut-down.
This will not be affected by water spray in the auxiliary building.
Cold shutdown may require boron addition, depending upon plant conditions.
As discussed in the FPR, damage to the boron transfer system can be repaired within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
The charging system is used to inject boron.
In the event both centrifugal charging pumps are lost through a combination of water spray and single failure, the plant could be main-tained in a hot shutdown condition until the damage is repaired or the non-safety-related positive displacement charging pump is made available.
Goal (3), removal of decay heat, is of most interest following a moierate energy line crack.
FSAR Table 3.G-3 lists six systems required for safe shutdown in this case.
These six systems, used to remove decay heat, are:
a.
Residual Heat Removal System (RH);
b.
Chemical and Volume Control System (CV);
c.
Auxiliary Feedwater System (AF);
d.
Component Cooling System (CC);
e.
Essential Service Water System (SX); and f.
Essential Service Water Makeup System (Byron Only)
(SXM)
The function and redundancy of these systems can best be seen by examining a diagram of the various paths for safe shutdown decay heat removal as shown in Figure 1.
This diagram shows four paths.
Two of these reach cold shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> while the other two (without RHR) may taken longer than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
These paths assume offsite power has been lost.
Section B.3.6.3 of Branch Technical Position ASB 3-1 states that if the initiating failure is in a normally operating moderate energy system, postulation of a single active failur?
is not required in the redundant train of that system.
The RN, CC, SX, and SXM systems fall into this category.
F2.17-3
B/B i
The SX and SXM systems are required for all shutdown paths.
l The redundant SX trains are separated and contain relatively few active components.
A single failure analysis of the SX system is included in the FSAR in Table 9.2-2.
The SX pumps are the only vulnernble components of interest.
In their location in the auxiliary building basement, the pumps could be affected only by a crack in the same SX train and therefore j
no single failure is postulated in the redundant train.
Even in the event both trains of the SX system for one unit were out of service, the SX valves could be aligned to supply the CC and RH systems from the other unit.
The only source of spray in the vicinity of the CC pumps is the CC piping.
Due to the arrangement of the pumps, spray from a single source would not affect more than one other pump in addition to the failed train.
The plant can be safely shut down with any two of the five pumps operating.
Since single failure is not required and no other components are vulnerable to spray, safe shutdown can be achieved.
As shown in Figure 1, the RH sysc~em is required only to reach cold shutdown in a short time.
The only RH component outside containment which could be affected by spray is the RHR pump motor.
This motor is located in a room adjacent to the contain-ment which contains only one train of the RH system.
As a O
result, any crack which causes failure of one RHR train will not require postulation of a single failure in the redudnant 7
train.
I The essential portion of the CV system consists only of the charging path into the containment.
The valves in this path are normally open and fail in that position.
The pumps are in cubicles separated from the redundant train and other systems.
In the event one centrifugal charging pump fails and the other l
is disabled by a single active failure, the CV system is backed l
up by the Safety Injection System.
The SI system becomes l
available when the primary system pressure is reduced to 1700 psig and is not subject to common failure with the CV system.
Secondary cooling (per Figure 1) is used to reduce the primary pressure.
The AF system hec no components which could be damaged by spray, except for the pump drives.
These are in rooms which contain only one AF train, Tnerefore, water spray will affect only one train of the AF syctem.
The reliability of the AF system has been documented in the response to NRC Question 10.53 and accepted by the NRC.
In the event spray disables one AF train and single failure disables the other, safe shut-down can be achieved per Figure 1 by feed and bleed of the
-primary system with or without RHR.
()
Capability to safely shut down the plant following a moderate energy line crack and the resulting spray in conjunction with i
a single failure as required by Branch Technical Position F2.17-4 l
l r
B/B
- O ASB 3-1 will be documented in a report titled " Jet Impingement and Water Spray Documentation Summary".
Section IV of the FSAR response to NRC Question 10.40 (Pages
'Q10.40-1 through Q10.40-5, attached) summarizes the approach to moderate energy line break analysis.
This approach was reviewed.by the NRC and discussed on page 3-2 (Section 3.6.1) 1, of the Byron Safety Evaluation Report, NUREG-0876, Supplement No. 2, January 1983:
"The applicant has also analyzed the effects of moderate energy line breaks outside containment an nafety-related systems.
The moderate eneegy systems are also designed to meet the criteria set forth in BTP ASB 3-1.
The staff has determined that the applicant has adequately demon-i strated that a postulated crack in a moderate energy line will not cause loss of function of any safety-related system because of environmental effects such as flooding, high temperatures, or water spray."
i The response to FSAR Question 10.40 provides a detailed example of the effects of a. moderate energy line crack in an essential service water line in the auxiliary building bacement demon-strating that shutdown of the plant is unaffected.
In response l
to a specific request from the Auxiliary Systems Branch, this question response also includes a detailed examination of failures in the component cooling system.
Again, the ability L
to safely shut down the plant is maintained.
During the IDI, the IDI Team requested that the area around the Auxiliary Feedwater (AF) regulating valve station be evaluated.
This was done and it was found that even though only two locations in the piping examined (the two largest subystems in the AF system) exceeded the criteria for postulation of through wall cracks, a' crack could occur at any location on the subsystems without affecting safe shutdown capability.
This antlysis was transmitted to the IDI Team (IDI Report Reference 2.153) but was not mentioned in the IDI Report.
This finding is based on a misinterpretation of the licensing commitments in FSAR Subsection 3.6.2.1.2.2 and does not accurately reflect the actual methodology used to meet regulatory require-ments and the information supplied to the IDI Team.
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ATTACHMENT TO FINDING 2-17 RESPONSE B/B-FSAR AMENDMENT 39 SEPTEMBER 1982
(::)
1 QUESTION 010.40
" Provide a response to Question 010.17 and include the following in your response.
Provide the results of analyses of the effects on safety-related systems of failures in any high or moderate energy piping system in accordance with the J. F. O' Leary letter of July 12, 1973, as defined in Branch Technical Position ASB 3-1, Appendix C.
Provide a table which identifies the method of protection provided all. safety-related systems listed in FSAR Table 3.6.1 from failures of any high or moderate energy systems listed in FSAR Table 3.6-2.
Include figures depicting the locations of failures relative to the systems of FSAR Table 3.6-1 giving dimensions, locations and protective method for each postulated break or crack in a high or moderate energy system.
Include the assumptions used in your analysis such as flowrates through postulated cracks, pump room areas, sump capacities, and floor drainage system capaci-ties."
RESPONSE
I.
INTRODUCTION O
To ensure safe and reliable operation of the Byron and Braidwood Nuclear Power Stations, the possibility of high or moderate energy line breaks have been considered in the design.
This response documents a confirmatory study of the potential high and moderate energy line breaks which demonstrates that all design features necessary to mitigate the results of line breaks have been incorporated.
Standard Review Plans (SRP) 3.6.1 and 3.6.2 were used as the basis for this study.
SRP 3.6.1 includes Branch Technical Position (BTP) APCSB 3-1.
Appendix B of the BTP, the attachment to letters sent to applicants and licensees by A. Giambusso in December 1972, and Appendix C to the BTP, the July 12, 1973 letter to applicants, reactor vendors and architect-engineers from J. F. O' Leary, provide the basis for identification of high energy line breaks and evaluation of their consequences.
Piping drawings which identify the high energy lines are included in the FSAR (Figures 3.6-1 through 3.6-12).
Breaks have been postulated at the locations required by Branch Technical Position i
APCSB 3-1 for the purpose of assessing pipe whip and jet impinge-ment effects.
Pressure and temperatures in areas were calculated assuming the break occurs in the limiting location in the area.
Locations of mitigating features such as pipe restraints and cO Drawings showing the location of high energy lines have been impingement shields are shown in Section 3.6 of the ESAR.
provided to the NRC ASB reviewer.
These drawings also indicate location of subcompartment walls and pipe tunnels.
Q10.40 1
ATTACHMENT TO FINDING 2-17 RESPONSE (Cont'd)
B/B-FSAR AMENDMENT 39 SEPTEMBER 1982 i
II.
SCOPE The effects of high and moderate energy line breaks inside I
containment have been assessed as described in FSAR Sections 3.6 and 6.2.
Additionally, an investigation into the effects l
of high and moderate energy line breaks outside containment has been made and is described in this response.
Non-safety related areas, such as the turbine building, were not investi-gated because damage to or 2ailure of equipment in these areas will not affect plant safety.
The possible effects considered are structural loads due to pressurization, increases in pressure and temperature which could affect environmental qualification of equipment, and damage due to pipe whip and jet impingement.
Flooding is a potential effect but is not addressed in this response.
The response to Question 010.47 demonstrates that high and moderate
{
energy line breaks will not cause flooding which would adversely affect the plant safety.
Because of variations in requirements, techniques, and failure effects, high and moderate energy lines are addressed separately.
Similarly, the pipe whip, subcompartment pressurization, and environmental analysis all have somewhat different approaches.
[)
The following sections are divided to reflect these distiactions.
III.
HIGH ENERGY LINE ANALYSIS Standard Review Plans 3.6.1 and 3.6.2 were followed in defining and identifying high energy lines.
High energy lines are those larger than 1 inch diameter for which either:
The service temperature is greater than 200 F;
a.
or b.
The design pressure is greater than 275 psig.
Galy a limited number of systems in the auxiliary building meet either of these criteria.
The following systems have been identified as containing high energy lines in the auxiliary building:
Chemical and Volume Control (CV)
Auxiliary Steam (AS) i Steam Generator Blowdown.
(SD)
Radioactive Waste Processing (WX)
Boric Acid (AB)
Main Steam (MS)
(FW)
Safety Injection (SI) 0010.40-2
ATTACHMENT TO FINDING 2-17_ RESPONSE (Cont'd)
B/B-FSAR AMENDMENT 39 SEPTEMBER 1982 Systems which. are normally not used or at reduced temperature and pressure are not necessarily required to be considered as high energy lines.
A guideline has been established (Branch Technical Position MEB 3-1) that if the system is at high energy conditions less then 2% of the time, it may be considered a moderate energy line and its normal conditions applied to the line break analysis.
On this basis, the last three systems (AF, RH, SI) are not considered as high energy systems.
The Byron /Braidwood AF system is not used for normal startup as at some other plants.
The only high energy line in the boric acid system is a steam supply line to the boric acid batching tank.
This line is essen.tially a part of the nuxiliary steam system and, as a result, was not identified in FSAR Table 3.6-2.
Subcompartment pressurization is investigated for all lines with temperatures above 200' F.
Lower temperature lines do not have the potential for flashing to steam and thus will not increase the pressure of a subcompartment in the event of a break.
Pressurization is of concern only in small subcom-partments with relatively large high energy lines or subc*ompart-ments with limited pressure relief venting.
High energy lines below 200* F have only minor effects on the environmental conditions.
The absence of steam and the ability to drain warm liquid from the break area limits the temperature J
rise from these breaks.
The auxiliary building HVAC has suffi-cient capacity to accommodate these lower temperature breaks.
Breaks of other high energy lines may influence the expected maximum temperature in some areas of the auxiliary building
'even if high pressures do not result.
The auxiliary building contains several large areas with high energy lines that are not subject to pressurization but are investigated for environ-mental effects.
Certain postulated break locations in high energy piping systems are used to investigate the potential for damage due to pipe whip and jet impingement.
The guidelines in Standard Review Plan 3.6.2 are used to determine the number and locations of the pipe breaks.
Pipe restraints are added as required to prevent damage to structures and saf ety-related equipment.
I V.
MODERATE ENERGY LINE BREAKS Moderate energy lines are lines which operate at temperatures below 200' F and pressures below 275 psig.
A break in a moderate energy line will not result in ficshing of the liquid to steam and, as a result, has no potential for pressurization of areas.
The relatively low temperature and reduced heat transfer effects of the liquid blowdown precludes significant temperature increases s],~3 in the area of the break.
The reduced break area applicable to i
)
these breaks and the absence of steam allows the auxiliary building HVAC to maintain temperatures within those specified Q10.40-3
ATTACHMENT TO FINDING 2-17 RESPONSE (Cont'd)
B/B-FSAR AMEND::En? 39 SEPTEM3ER 1982
)
s in the environmental qualification program.
The results of moderate energy line breaks are, therefore, confined to the physical effects of liquid discharge into the plant.
?lant safety is affected only if equipment required to mitigate the break or to safely shut down the plant can be damaged by resultant flooding or water spray.
Flooding is discussed in the response to Question 010.47.
Water spray was not found to affect plant safety because of the separation of redundant safe shutdown systems and components.
Moderate energy line breaks do not result in pipe whip.
As an example, the auxiliary building basement at elevati6n 330 feet is subject to severe flooding following a moderate energy line failure in the essential service water system.
The limiting failure in this area is a crack in a 36-inch essentj/sec.
al service water line.
The predicted leak rate is' 2.8 ft The basement is divided into two completely independent sections.
These sections are separated by a wall which has been designed to withstand the ficoding.
Each section contains redundant essential service water purps which can supply both units.
Therefore, flooding or spray from a break cannot affect the equipment in the other section of the basement and essential service water will be supplied to both units.
_) h This separation is well documented in the Fire Protection Report.
This report lists and locates equipment required for safe shutdown.
When redundant safe shutdown systems are separated by fire walls or by more than 20 feet, spray fron j
a crack in a moderate energy line would not impair the safe shutdown capability of the plant.
j A moderate energy line break in the component cooling system was given special consideration because the component cooling system is not supplied with a Category I source of makeup water.
j A leak in this system could theoretically drain the surge tanks and result in damage to the component cooling pumps.
A significant leakage in the component cooling system is not expected.
The system is a moderate energy, low pressure system and is not subject to severe loading.
In the event the system is inoperable, the plant may be safely maintained in a not shutdown condition until the component cooling system is restored.
If a crack is postulated in one of the large lines in the system, I
the level in the surge tank of the affected unit will drop.
~
When the level reaches the low setpoint level, alarms will sound and the affected units component cooling pumps will be i
automatically tripped to prevent damage to the pumps.
I If reactor water or derineralized water makeup is available, the component cooling pumps may be restarted and the unit oper-ated normally while the leak is located and isolated.
Otherwise, 010.40-4
ATTACHMENT TO FINDING 2-17 RESPONSE (Cont'd)
B/B-FSAR AMENDMENT 39 SEPTE"BER 1982 s
6 Iw' '
the reactor will be tripped because of the interruption of component cooling to the reactor coolant pumps and the unit will be placed in a bot shutdown condition.
Component cooling is not required to safely maintain the unit in hot shutdown mode.
The component cooling system can be operated af ter a failure of the piping by closing the appropriate system valves to isolate the break location and maintain component cooling flow.
V.
SUBCOMPARTMENT PRESSURIZATION For the purpose of protecting subcompartments from overpressuri-zation, the CV, AS, SD, WX, MS, and FW systems were traced through the auxiliary building and all subcompartments containing high energy lines were identified.
The most severe break in the subcompartment was analyzed.
The main steam (MS) and feedwater (FW) systems are routed entirely in an enclosed tunnel in the auxiliary building.
The limiting break in this tunnel is a main steam line rupture.
Section C3.6 fully describes an analysis of a break in this tunnel.
The remainder of the auxiliary building was surveyed level by level to identify all subcompartments which could be pressur-f ized by high energy line breaks.
Figures 010.40-1 through Q10.40-5 identify all areas containing high energy lines.
The identification of the limiting line in each zone is also included.
The zone numbers do not correspond to environmental
. qualification zones (Section 3.11).
Figure Q10.40-1 represents elevation 346 feet 0 inch.
Zone 1, the recycle waste evaporator room, has been analyzed and the results are reported in Section A3.6 of the FSAR.
Zones 2 and 3, letdown reheat heat exchanger rooms and valve areas, have been analyzed and the results are reported in Section i
A3.6.
The assessment in A3.6 addressed Zone 3, the more limiting Ione.
Figure Q10.40-2 represents elevation 364 feet 0 inch.
Assess-ment of Zones SA and SB, the positive displacement charging
~
pump areas, predict a peak pressure of 2.42 psid and a peak temperature of 190
- F.
These results are being added to Section A3.6 of the FSAR.
Zones 6A, 6B, 7A, and 7B, the centrifugal charging pump rooms, contain high pressure, low temperature lines.
Failure of these lines (normal temperature of 1155 F) will not cause pressurization or increase temperatures.
Pipe whip and impingement are considered.
Zones 9A and 9B contain portions of the steam generator blowdown system.
Control valves upstream of these lines limit the blowdown flow and prevent the postulated breaks from impacting plant design.
Zones 8A and 8B, blowdown condenser rooms, have been analyzed and the results are included in FSAR Section A3.6.
Zones 11 and 12, blowdown condenser rooms, have been analyzed and the results are reported 1 Section A3.6.
010.40-5
B/B f
FINDING 2-18: FLOODING ANALYSIS The IDI Team noted that the Project Management Division
)
made an undocumented engineering judgment that high predicted flood levels in the plant.would cause doors to open prior to floor damage.
The. Team felt that this was inconsistent with the FSAR response to Question 10.47 and that other factors such as (1) the lock might be stronger than the minimum rating or (2) the door might open inward and the door frame might be stronger than the lock, should be considered.
RESPONSE
We do not agree with the conclusions of this finding.
The prisary purpose of the flooding calculation was to evaluate the ivoact of flooding on safety-related equipment.
Structural concernt.were addressed at a much earlier stage in design.
The response to FSAR Question 10.47 states " Doors are considered to be open or closed to maximize the flood levels."
FSAR Section D3.6.2.1 includes the explanation " Doors were assumed to remain closed to calculate the maximum flood in the area containing the break and assumed to open to check on potential flooding of adjacent creas."
This is a calculational technique O
and is notemployed to ensure that all affected equipment is identified inconsistent with the engineering judgment made about the relative door and floor capabilities.
These assumptions nilow the analyst to identify all flooding effects on equipment without determining the specific structural capabilities.
However, for completeness the auxiliary building flooding calculation is being revised to address the adequacy of both structural and mechanical / electrical aspects of the design for those areas with high flood levels.
This will serve to document the engineering judgment used.
All areas which could be flooded to levels which would be of structural concern are in the lower elevations of the aux-iliary building.
The floors, walls and supporting structures are ggnerally adequate for loads much greater than the 200 lb/ft live load considered in design.
The finding discusses the live load rating as if it were the structural capability 2) when, in fact, it is a design minimum.
This value (200 lb/f t also includes a design load factor of 1.7, which is not required under accident conditions such as flooding.
The finding mentions two examples of factors which should be considered.
The first was the possibility that the lock might be stronger than the minimum rating.
For the areas in question, the real concern is whether the weakest link O
ments, or if itin the rooms is the d3or latch, door frame or the hinge attach-is the structural floor and walls.
In the F2.18-1
B/B Byron auxiliary building, the areas with predicted high levels have been analyzed and it has been verified that even if the doors do not open, no structural damage will result.
The second concern about doors opening inward is not valid because the only areas which could have high flood levels are'subcompartmepts and all subcompartment doors open outward.
The six areas noted in the report were reexamined to confirm our assessment of the structural impact as noted below:
A.
Two of the areas, the RHR pump cubicles, are located on Elevation 343, 3 feet below adjacent areas.
The floor slab and lower 3 feet of the walls (with the exception of one block wall) are mats backed by compacted soil and easily capable of withstanding high flood levels.
The reinforced concrete walls above Elevation 346 would easily handle the high flood levels and, in any event, be stronger than the block wall and door.
Failure of the block wall would not have an adverse safety effect.
B.
The other four areas are the RHR heat exchanger cubicles.
These areas, although nominally on the 364 feet 0 inch level, are actually recessed and have a floor O's at the 357 feet 0 inch level.
These cubicles have no outflow path for water other than floor drains below 364 feet 0 inch.
In effect, they are designed to be flooded and to prevent RHR leakage from traveling to other parts of the auxiliary building.
As a result, the walls and floor of these cubicles are 40 inches thick.
Based on this information, we believe the judgment on the structural implications of the flooding analysis were justified given the location of the high flood levels and the design of the Byron auxiliary building.
The reexamination of the structural design demonstrates there is no safety significance of this finding.
The revision of the calculation will document the engineering judgments made.
This revision will be completed by January 31, 1984.
F2.18-2
B/B O
FINDING 2-19:
FLOOD LEVEL Project Management Division personnel sent a memorandum to the Electrical Department requesting that equipment be located to be protected from a predicted 13 inch flood in the Unit 2 auxiliary feedwater diesel drive pump room.
This level was not shown on the General Arrangement drawings.
The Team interviewed cognizant Sargent & Lundy personnel and alleged that no record or personnel recollection of the specific PMD memorandum or the resultant activity existed.
The Team was also concerned because they did not find a systematic tracking system to assure that actions needed have been taken.
RESPONSE
We agree that the design drawings were not updated in a timely manner.
However, we believe the finding is inaccurate in the allegation that cognizant personnel were unaware of the design change.
The design drawings have now been updated to show the flood levels which could affect safe shutdown capability.
The IDI concern about the need for a systematic tracking system is addressed by the normal design procedure of updating design documents to reflect changes.
This procedure, in conjunction with project communications (memos, documented project meetings, etc.) insure that required reviews are made. -
This finding is based on an 8-4-82 memorandum transmitting the revised flcod level information to the Electrical Department The finding is inaccurate in that the memorandum was in the Electrical Project Engineering files and shown to the IDI Team and the cognizant Electrical Project Engineer was aware of the flood requirements.
A copy of the memorandum was also available in the files of the Electrical Design and Drafting Division at the site.
These files were not reviewed by the IDI Team.
The memorandum was written as a result of a discussion between the mechanical and electrical design personnel involved with this area.
It was determined that transmittal of the information by memo prior to inclusion on design drawings would insure an early review of the affected area.
This review was, in fact, made as a result of the meno but not documented at that time.
The review has since been documented and the design drawings have been updated.
Although the drawings should have been revised more quickly, we believe initial transmittal of the data by memo was effective in initiating a design review in this instance where the design was almost complete.
i( )
The flood levels are contained in Sargent & Lundy Calculation 3C8-1281-001, " Auxiliary Building Flood Level Calculations" (December 1981) and Calculation RAS-FL-1, " Flood Level Inside F2.19-1
3/B O
g,/
Containment" (Revision 1, July 1982).
These are approved design calculations in accordance with Sargent & Lundy procedure GQ3.08.
At the time the information in question (13 inch flood level) was developed, the design of the Unit 2 auxiliary feedwater diesel driven pump room was complete, and construction on this room was in progress.
As a result, the most effective way to insure that the postulated flooding could be accommodated, was to make a review of the affected design.
Appropriate design checks were made prior to the IDI to insure that electrical equipment would not be damaged by the predicted flood levels.
No design changes were required.
Revised flood levels in the containment and the auxiliary feedwater diesel-driven pump rooms have been added to piping composite drawings.
0 f
O F2.19-2
i B/B I
FINDING 2-20:
DIESEL ENGINE HEAT LOAD This finding states that calculation VA-100 used a non-conservative number for calculating the total heat released to the ambient from the auxiliary feedwater pump diesel due to a revised insulation package and exhaust piping.
RESPONSE
l We do not agree with the conclusions of this finding.
Calculation VA-100 was performed prior to the award of the insulation contract for the auxiliary feedwater pump diesel exhaust manifold and associated piping.
The diesel heat load value used in Calculation VA-100 was supplied by the engine manufacturer cased on his proposed insulation package.
Although the actual insulation installed is different from that proposed by the diesel engine manufacturer, the difference in heat loads is not significant in comparison with the design margin of the room cooling system.
Similarly, the short length of piping did not add significantly to the heat load.
These facts were obvious to the designers and no calculational refine-ment was deemed necessary.
However, to confirm this point, Calculation VA-100 has been revised to reflect a 2.6% increase O
in the heat load based on the actual installed insulation l
on the exhaust manifold and piping.
Based on the revised calculation, there is still an additional 11.4% margin in the capability of the HVAC system for this area.
The calcu-lated room temperature is 108' F while the design temperature for this area is 122' F.
I Due to the adequate margin in the design of the HVAC system, i
there were no design errors resulting from the revision to the exhuast insulation.
We have demonstrated that this finding is dealing with minor effects, which the designer correctly judged to be insignificant.
1 i
O F2.20-1 l>. - -. - - - - -
B/B
/~'T FINDING 2-21:
EMERGENCY SERVICE WATER TEMPERATURE id The Westinghouse feedline break analysis assumed the aux-iliary feedwater temperature to be a maximum of 95 F.
This temperature is inconsistent with FSAR Section 9.2.5.3 which indicates that essential service water temperature may be as high as 98 F, Design Criteria DC-AF-01-BB which indicates a water temperature of 100 F,
and specification F/L-2758C for the auxiliary feedwater pumps which indicates a temperature range of 40-100 F.
RESPONSE
We agree with the IDI Team's assessment that these differences are not significant.
The minor temperature discrepancies listed in this finding have no impact on the Westinghouse accident analysis or the auxiliary feedwater system design.
The 100 F temperature listed in the design criteria is based on ambient conditions that thw v_tdoor condensate storage tank will experience.
This temperature is used as a reference for the purchase of equipment such as the auxiliary feedwater pumps and valves.
The design outlet water temperature of the essential service water cooling towers is 98 F bcsed on maximum heat load and worst case environmental conditions.
[( )
The difference in water density between 95 F and 100 F is less than 0.1%.
Since this has no impact on the accident analysis, no document changes are necessary or required.
hs-F2.21-1
B/B D
FINDING 2-22:
PROJECT MANAGEMENT DIVISION FILING SYSTEM This finding states that the Project Management Division File Index was incomplete and did not meet the requirements of Sargent & Lundy Procedure GQ-6.01.
RESPONSE
i We agree with the conclusions of this finding that there should have been an updated file index.
The majority of the project information in the Project Management Division (PMD) is maintained by the Responsible Engineers in the division.
At the time of the audit, information on hand with the Responsible Engineers was not identified by the File Index as required by GQ-6.01.
The Project Management File Index has been updated and supplaments the file index used for the main file '," Records Center").
If a responsible engineer leaves or is transferred, his files (those identified by the PMD File Index) will remain part of the project files.
We do not consider this to be a problem, systematic or otherwise, since the information was being retained even though the File Index was not being employed.
We believe the corrective action taken resolves this finding.
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F2.22-1
_ ~ - _ _.,,. -,
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\\
FINDING 3-1:, DIGITI 7ED RESPONSE SPECTRUM This findir:9 addresses a discrepancy between the OBE Response Spectrum 101-OB-EW and the Response Spectra Design Criteria.
RESPONSE
We agree with the conclusions of the finding.
We also agree with the IDI Team's conclusion that this did not indicate any systematic weakness and did not cause any design deficiency.
The Digitized Response Spectra (101-OB-EW), which is one of the 18 Response Spectra used in the analysis, had a differential acceleration value of 0.ii29 (0.26g instead of 0.289) at 18Hz.
Furthermore, the envelope of the 3 E-W Response Spectra, including
~
101-OB-EW used in the analysis, resulted in an acceleration value of 0.8g at 18 Hz.
This one discrepancy represents 0.3%
of all digitized points comprising the 18 Response Spectra identified in the analysis.
D l
f F3.1-1 I
l
t B/B f)
FINDING 3-2:
FUNCTIONALITY CRITERIA In summary, the memorandum defining systems to be checked for functionality was incomplete.
Some of the analysis packages that had been defined as needing the check did not state that the criteria had been met and, for some of the packages that did state the criteria was met, it was not clear that valid analyses had been performed to make this determination.
These items indicated systematic weaknesses in the Program for Meeting Functionality Criteria so that the licensing commitment was not being met.
RESPONSE
We agree that the memorandum for defining the sytems to be checked for functionality had not been updated and that some of the analysis packages that had been defined as needing the check (inclusion of a Class 1 stress calculation or have documentation that the piping met the Class 2 and 3 function-ality criteria) did not state that the criteria had been met.
We also agree that this item is not expected to cause any design deficiencies because the stresses are generally low.
Sargent & Lundy has updated all the system design specifica-tions outlining Functional Capability Requirements for ASME O
Section III Piping in the Sargent & Lundy scope.
This provides a documented basis for systems requiring a functionality check.
i S&L has also revised EMD TP-2 to clarify qualification methods available to satisfy functional capability requirements (EMD TP-2, Rev. 4, EMD-046032).
The review of fif teen (15) auxiliary feedwater systems for functional capability resulted in the following:
a.
Four (4) system stress packages did not contain a statement concerning functional capability (AF-08, AF-09, AF-10, and AP-15).
Two of these stress packages, AF-09 and AF-10, did contain a Class 1 Analysis Qualification (Run ID 650 UGS and 327 UGS, respectively).
Stress Peckage AF-08 required the-Stress Scanning Method for qualification, and Stress Package AF-15 required a calculation which was not included in the stress package at the time of the r.udit, b.
The remaining eleven (11) system stress packages did have a statement concerning functional capability.
Five (5) of these packages used the Class 1 Analysis Method for Qualification (AF-01, AF-06, AF-07, AF-14, and AF-16).
Four (4) of these packages used the Stress Scanning Method for qualification (AF-03, AF-04, AF-ll, and AF-12).
The remaining two (2) stress packages (AF-02 and AF-05) required D'.
calculations which were not included in the stress packages at the time of the audit.
F3.2-1 b u m' un's
'"ei-
B/B 4
- O Based on the above discussion, all the auxiliary feedwater systems satisfied the functionality requirements.
There were three (3) stress packages (AF-15, AF-02, and AP-05) that required a calculation method to satisfy the functional capability criterion per NEDO-21985.
These calculations were provided to the IDI Team (EMD-044004, June 10, 1983).
The IDI Team also questioned the applicable code version used for NEDO-21985 and the PIPSYS program.
In particular, they indicated that there had been "many stress intensification factor changes" for the Class 2 and 3 piping analysis require-ments versus the functional capability requirements.
The NEDO report references ASME Section III 1977 issue through summer of 1978 addenda.
The PIPSYS computer program has been updated for the 1977 issue through the winter of 1978 addenda for Class 2 and 3 piping.
Only minor differences occurred between these issues of the code, none of which impacted the stress intensification factors used for analysis.
As a result of this, it is possible for an analyst to screen the Clasa 2 and 3 stress results and compare them visually against the functional capability criteria since the design rules are the same in both cases.
It should be noted that the NEDO report also references the i f) 1977 issue through the summer of 1978 addenda for Class 1 analysis.
In this case, PIPSYS has been updated to the winter of 1979 addenda.
It should be noted that in this case, the PIPSYS version being used is either equal to or more conservative than the criteria recommended in the NEDO report.
It should also be noted that the Section III code revisions for Class 1 analysis were based on the same " seed" report by E. C. Rodabaugh and S. E. Moore (Reference 1 of NEDO-21985) which was prepared for the NRC to evaluate this very subject.
Consequently, Sargent & Lundy believes that the practice being followed is either equivalent to or more conservative than the NEDO report criteria.
The IDI Report states "it is difficult to assure that the l
General Electric Report Class 1 criteria are met when only ASME Code Class 2 stresses are calculated".
This is not Sargent &
Lundy's practice, nor has it ever been.
The IDI team member's observation is invalid since the functional capability requirement for ASME Section III Class 2 & 3 piping, which Sargent & Lundy l
follows, is delineated in Section 2.2.e of NEDO-21985 (IDI Reference 3.93) which is referred to as the General Electric Report by the IDI Team, and states as follows:
O F3.2-2
B/B O
" Piping constructed in accordance with the code rules for class 2 or 3 may be evaluated for functional capability using the criteria given herein for Class 1 piping."
All the safety-related subsystems in the S&L scope requiring a functionality check have been identified and will have a doc-umented functionality check performed.
O O
F3.2-3
I B/B O
FINDING 3-3:
NUMERICAL INTEGRATION ALGORITHM The numerical integration scheme outlined in FSAR Subsection 3.7.3.1.1.5 is not identical to the algorithm actually employed in the PIPSYS Computer Code.
RESPONSE
We do not agree with the conclusions of this finding.
As stated in the IDI Report, the Numerical Integration Method actually employed is technically adequate and the differences between the two methods are insignificant.
All piping analyses for Byron performed by Sargent & Lundy utilized the PIPSYS Computer Code which is referenced in Appendix D of the FSAR.
The Numerical Integration Method outlined in FSAR Subsection 3.7.3.1.1.5 is noted as "one available method for the numerical integration," and as such, does not constitute a commitment to use this specific method.
O O
F3.3-1
I B/B O
FINDING 3-4:
SUBSYSTEM NUMBER This itera indicates that, on Drawing M-37 (Reference 1.57),
the reference to the subsystem boundary showing the line from the Condensate Storage tank to the suction sides of the auxiliary feedwater pump should be labeled ICD 06 rather than lAF06.
This finding states that it is an isolated drafting error.
RESPONSE
We agree with the conclusions of this finding.
However, the subsystem identification numbec on piping and instrumentation diagram is for reference only as stated on Drawing M-34, Sheet 2.
It is not a design input for piping analysis.
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B/B O
FINDING 3-5:
FLANGE WEIGHTS This is a discrepancy of flange weights in Stress Report 1AF03.
RESPONSE
We agree with the IDI Team that a discrepancy ot 19 lbs between the weight shown on the analytical drawing (62 lbs)
)
of the 6-inch flanges and the weights used (43 lbs) in the piping analysis will not constitute any design deficiency and does not require any further consideration.
This has been documented in an addenda to the stress report.
l Y l
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F3.5-1 4
, - - - - -,. -,. =,, - - - -
B/B FINDING 3-6:
PIPE SUPPORT ADDED MASS The input data for two out of six hangers were not properly prepared to account for the additional weight or mass.
RESPONSE
We do not agree with the IDI Team's conclusion that this is a systematic weakness for the following two reasons:
1.
The addition of supports, weight and mass into the piping analysis is governed by the criteria outlined in Paragrcples d.4.5, Table 6.4.5-1, and Figure 6.4.5-1 of Engineering Mechanics Division's " Lesson Plan for Training Personnel it. Tiping Analysis," EMD TP-1, Revision 3 (Reference 3.50).
2.
According to the criteria, the weight of the support assembly lAF14003R (45 lbs) doet not have to be con-sidered in the analysis since it does not exceed the weight of 1 foot of pipe (47 lbs).
The weight of O.
the support acaembly 1AF03014X (53 lbs), while it exceeds the limitation of the criteria for the exclusion of the weight by one pound, represents 2% of the weight of a 1-foot section of pipe (52 lbs) and is considered not significant in the context of the low stresses in the piping system and the conservativeness inherent in the analysis techniques used.
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B/B m
FINDING 5-3:
HEAT DISSIPATION CALCULATION This finding alleges three minor procedural errors associated with Heat Dissipation Calculation 19AI-18.
The first error was that pages were not numbered 1 of 80, 2 of 80, etc., nor was the final page identified with a ' final page' designation as required by Sargent & Lundy's Procedure GQ-3.08.
The second error was that the review method used was not documented in accordance with Sargent & Lundy's Procedure GQ-3.08.
The third error was the use of an assumed motor efficiency of 95 percent when the actual vendor data on the motor efficiency was available at the time the calculation was performed.
RESPONSE
We do not agree with two of the alleged errors cited in this finding.
We do agree that the review method should be docu-mented.
The following response to this finding addresses each of the alleged errors identified in the finding.
1.
There was no error in page numbering.
Calculation 19AI-18 is an 81 page calculation and clearly states " final" (s
on page 81 of the calculation.
No correction is required.
a' 2.
Sargent & Lundy agrees that the review method should have been documented in accordance with GQ-3.08.
The review method used for Calculation 19AI-18 has now been documented.
3.
There is no error in motor efficiency.
The calculation clearly states that assumed motor efficiencies of 95 percent are used for all 4000- and 6600-volt motors.
The use of assumed motor efficiencies for all motors simplifies the calculation.
This assumption is valid due to the conser-vative factors and approximate nature of the calculation.
As the IDI Report states, the use of the actual motor efficiency (94%) has no effect on the design.
Furthermore, as acknowledged in t'he Report, there is a Departmental Standard (ESI-253) that requires that the input data and assumptions used in safety-related calculations be sub-l jected to a final check.
This (final check) program is in process; the subject calulation is on the list of calculations to be checked, and the final check will be complete prior to fuel load.
4 A draft revision (never formally issued) of Sargent &
Lundy Electrical Engineering Reference ESC-525 (Reference 5-25) is mentioned in the Report as requiring that efficiency figures for the motor given by the manufacturer should
,A be used for the calculation when available.
This reference
}
F5.3-1 l
.. _ _ =
I 3/B f]
does not govern the preparation of the calculation and is identified in the calculation references as only appli-cable for additional heat dissipation information not available in the governing Sargent & Lundy Standard.
No correction is required.
P O
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FS.3-2 l
B/B O
FINDING 5-4:
ALARM LIGHT LOGIC DIAGRAM This finding addresses a logic diagram draf ting error wherein a flow setpoint was shown as 100 GPM when it should have read 160 GPM.
RESPONSE
We agree that the 100 gpm logic diagram flow setpoint was a drafting error.
We also agree with the IDI Team that the error is an isolated case.
We have further verified that other documents which specify instrument requirements and have been released for construction and purchase, specifically the flow switch and indicating light instrument data sheets, have the correct setpoint.
The logic diagram has subsequently been corrected to indicate the correct flow data.
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F5.4-1
B/B FINDING 5-5:
ELECTRICAL DRAWING This finding identifies a minor transcription error on a design drawing in which a conduit number was incorrectly listed.
RESPONSE
We agree that a transcription error on a design drawing was i
made.
However, the IDI Report also contains a transcription error i.e.,
the correct Conduit Number is "57" (rather than "51",
as listed in the report).
This drawing error has been corrected to show Conduit No.
"57".
j We also agree with the IDI Team that the error was minor and "in general" the methods, procedures, and documents associated with manual routing of cables in the field (Cable 2AR190) are in good order.
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F5.5-1 I
B/B ge s FINDING 6-1:
OPENING OF FLOW CONTROL VALVES q
This finding state that the auxiliary feedwater air-operated flow control valves should, if closed, open on receipt of a safety injection signal as described in the auxiliary feedwater system design criteria (DC-AF-01-BB).
This has not been implemented in the system design.
RESPONSE
We agree with the conclusions of the finding.
However, the intent of the design criteria statement is to preclude leaving any of the eight flow control valves (nornally open) closed after testing or maintenance.
As an alternate design to supplying a safety injection signal to the flow control valves to assure that they are open, the design is such that closure of a flow control valve produces a signal to the Equipment Status Display indicating an inoperable status of an auxiliary feedwater train.
The operator would reopen the valve (s) to return the auxiliary feedwater system to a normal standby and ready condition.
The current design more than meets the intent of the design t
criteria to assure that failure to reopen the flow control r~
valves after testing or maintenance does not jeopardize the T,3 function of the auxiliary feedwater system.
Please refer
/
m to the response to Finding 2-13 for a discussion concerning j
update of the design criteria.
t l,(
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F6.1-1
B/B
(
FINDING 6-2:
RELAY DESIGNATION NUMBER This finding concerns an inconsistency in the bus undervoltage relay designation between an AF System schematic diagram and an EF System schematic diagram.
RESPONSE
We agree that there was an inconsistency on the drawing.
We also agree with the IDI Team that this is an isolated (random) error.
The EF System schematic diagram was revised to eliminate the inconsistency.
,O F6.2-1
_ _ _. - -.. _ - _, _ _ - _ _ - -. _ - - -. - - - ~ - - - - - - - - - - - - - - -
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B/B n
FINDING 6-3:
BASES FOR SETPOINTS This finding states that no calculations exist for the auxiliary feedwater pumps suction pressure switches' setpoints.
The lack of a calculation or other documented design basis for the setpoints is a violation of IEEE 279-1971.
The finding stated that this appeared to be a systematic problem that should be addressed.
RESPONSE
We do not agree with the conclusions of this finding.
The suction pressure switch setpoints were based on the existing head loss calculations AFJD-1 and AFJK-1 and approved design drawings.
A separate documented calculation specifically for the suction pressure switches was not per formed prior to the IDI nor was it deemed necessary.
A documented calculation (AFTd-02) was performed to verify to the IDI Team that the setpoint is consistent with the system design requirements.
No design i
' error in the setpoints was found which could be attributed to-the lack of a separate and specific calculation.
In general, instrument setpoint bases are existing engineering
, f( )
calculations, design drawings, and/or vendor-supplied component l
design data.
Separate setpoint calculations are performed I
for only those instrument sc-tpoints that cannot be determined from existing data.
Instrument setpoints determined from these sources are documented on prepared, reviewed, and approved i
instrument data sheets.
This method of establishing instrument setpoints does provide a documented design basis in compliance with IEEE 279-1971.
The auxiliary feedwater pump suction pressure switch is probably the most complex and most safety significant switch in the Sargent & Lundy (S&L) scope.
Discussions were held between the instrumentation personnel and the responsible engineers to determine the requirements for the switch.
To address the IDI concern, S&L will make a documented assess-l ment of the safety-related C&I instruments in the S&L scope to iden-tify instruments that are judged to be complex in application and scope.
A documented calculation will be provided for those instruments identified, if it does not already exist, to verify the adequacy of the setpoint, and will include a verification of the setpoint accuracy (Finding 6-7), and the reset valve (Finding 6-8), if applicable.
F6.3-1
h '
B/B
' FINDING 6-4:
CALIBRATION INTERVAL This finding addresses a discrepancy between the test reports (References 6.28 and 6.29) and the proposed techni-cal specifications regarding the calibration test interval for the auxiliary feedwater pump suction pressure switches (lPSL-AF051 ar.d 1PSL-AF055).
The test reports specify a test interval of 208 weeks, while the technical specifi-cations specify a test interval of 18. months.
RESPONSE
The instrument maintenance calibration test reports for 1PSL-AF051 and IPSL-AF055 were incorrectly marked with a calibration frequency of 208 weeks (4 years) instead of 18 months.
The test reports have been revised to reflect the appropriate i
calibration frequency.
Byron Instrument Procedures 2000-003 and 2000-004 provide administrative controls for calibration intervals.
Additionally, calibration requirements are being verified during the review of Byron technical specifications.
i I
D F6.4-1
B/B A(O FINDING 6-5:
SWITCH CONTACT OPERATION This finding addresses an apparent inconsistency between the contact action specified on the instrument data sheet for pressure switch 1PSL-AF055 and the corresponding electrical schematic diagram where the pressure switch low-low inter-lock is shown.
RESPONSE
We agree that the instrument data sheet incorrectly indicated the contact action.
Discrepancies between instrument data sheets and schematics occur during the design process and instrument data sheets, as a general practice, are revised as required.
Since the actual contact type specified for the pressure switch in question is an SPDT, it is not critical to specify the contacts as opening or closing on decreasing pressure.
Specifying the contact as opening or closing in this case would be equally correct.
It is important, however, to specify the contact action to occur at a decreasing or increasing setpoint and
_D this was done.
We would further note that the contact operation
'(l specified on the instrument data sheet does not control or s-l jeopardize the contact implementation required on the electri-cal schematic diagram for the correct equipment function.
For the particular case noted in the finding, we have, since the IDI, revised the auxiliary feedwater pump suction pressure switch instrument data sheet to be in agreement with the electrical schematic diagram.
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F6.5-1
B/B O(V FINDING 6-6:
TEST REPORT The Commonwealtn Edison test report for 1PSL-AF055 incorrectly states the direction of pressure change for the low suction pressure setpoint (Reference 6.29).
In addition, this test report has a typographical error in the number for
' wire AFlBLAS in that it was listed as AFlBLA3.
At the time this test report was inspected, Commonwealth Edison actions to correct the Sargent & Lundy instrument data sheet and Commonwealth Edison test reports had not been initiated using the instrument discrepancy report form.
Subsequent Commonwealth Edison actions to correct.the test report and the data sheet discussed in Finding 6-5, while technically correct, appear to introduce unnecessary complexity in the designation of vacuum setpoint values.
On the basis of our examination of other test reports (References 6.125, 6.126, 6.127, and 6.129), these errors did not appear to represent systematic weaknesses.
RESPONSE
As noted on page 6-10 of the IDI Report, actions were taken
-g to correct the instrument data sheet and the test report using
'( )
approved procedures.
Also, as noted in the IDI Report, no l
evidence was found to suggest a systematic weakness in this I
area.
l 1
l l.
F6.6-1
B/B O
FINDING 6-7:
SETPOINT ACCURACY REQUIREMENTS j
This finding addresses an alleged need for setpoint accuracy to be specified on the instrument data sheet for pressure switches 1PSL-AF051 and 1PSL-AF055.
The finding states that the absence of the setpoint accuracy indicates a systematic weakness and is not in compliance with IEEE 279-1971.
RESPONSE
We do not agree with the conclusions of this finding.
The instrument data sheet is the documented design basis for the instrument in compliance with the general requirements of IEEE 279-1971.
The instrument data sheet contains the specific information to establish the requirements the instrument vendor must meet and therefore establishes the design basis.
In general, the information on the data sheet is obtained from an engineering assessment of information contained in i
3 system calculations, design drawings, and/or vendor supplied
.)
component design data.
Instrument vendor specification sheets and/or catalog information are generally used as part of the assessment.
Setpoint accuracy is determined in a similar way.
The setpoint accuracy required is then used (along with other instrument requirements) in the review of instrument vendor catalog infor-mation to establish instrument selection.
Instruments selected and documented on the data sheets are vendor standard designs which are selected to envelop the system operating requirements an opposed to specifying special design characteristics (i.e.,
setpcint accuracy) which would meet a specific system require-ment.
Cee the response to Finding 6-3.
The instrument data sheet records specific information about the instrument, as well as, the manuf acturer and model number.
Vendor setpoint accuracy specifications, in addition to other design data and specifications, exist for that model number.
l This information supplements the instrument data sheet design l
information.
We conclude rhis is consistent with IEEE 279-1971.
F6.7-1
B/B l
FINDING 6-8:
BASIS FOR RESET VALUE This finding addresses an alleged need for a pressure switch reset value design basis for 1PSL-AF051 and IPSL-AF055.
Com-pliance to IEEE 279-1971 is cited as the requirement.
' RESPONSE We do not agree with the conclusions of this finding.
As stated in the response to Finding 6-7, the instrument data sheet is the documented design basis for the instrument in compliance with the general requirements of IEEE 279-1971.
The instrument data sheet contains the specific information to establish the requirements the instrument vendor must meet and, therefore, establishes the design basis.
In general, the information on the data sheet is obtained from an engineering assessment of information contained in system calculations, design drawings, and/or vendor supplied 1
component design data.
Instrument vendor specification sheets and/or catalog information is generally used as part of the assessment.
Reset values are determined in a similar way.
The reset value required is then used (along with other instrument requirements) i l
in the review of instrument vendor catalog information to establish instrument selection.
Instruments selected and documented on the data sheets are vendor standard designs which are selected to envelop the system operating requirements, j
as opposed to specifying special design characteristics (i.e.,
reset value) which would meet a specific system requirement.
Special design characteristics which would meet exactly the specific system parameter are not specified.
In this particular case, the pressure switch reset was specified as the "manuf acturer's standard" which enveloped the requirements.
Therefore, specific calculations to determine an exact reset value were not required.
See the response to Finding 6-3.
l Finally, the instrument data sheet records specific information about the instrument, as well as, the manufacturer and model number.
Vendor reset specifications, in addition to other design data and specifications, serve to supplement the instru-ment data sheet design information.
We conclude this is con-sistent with IEEE 279-1971.
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F6.8-1
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B/B n
(
FINDING 6-9:
PREOPERATIONAL TESTING A commitment on FSAR page 14.2-31 states that during the auxiliary feedwater system preoperational test, ' motor and diesel-driven pumps will be verified to start under any safeguard situation under any possible control lineup, including restart capability, from any control station, following a protective trip.'
Contrary to this commit-ment, no plan exists to test the actual operation of the pressure switch at its reset point under conditions where the pump is required to automatically take suction from the essential service water system.
Commonwealth Edison and Sargent & Lundy personnel indicated understandable reluctance to test the auxiliary feedwater system using essential service water because that system's relatively impure water would necessitate cleanup of the auxiliary feedwater system and the steam generators.
However, the test could be performed using water of better quality.
Commonwealth Edison and Sargent & Lundy also indicated
~
a belief that testina the individual components separately is adequate.
However, we could find no planned test to demonstrate that the piping from the essential service water system will carry full flow without excessive restriction.
It, S In addition, no test of the entire integrated system response to a need for switchover to the safety-related water source l
is planned.
The inconsistencies with the FSAR commitment t
should be resolved.
RESPONSE
The lack of planning to actually switch the suction of the
(
auxiliary feedwater pumps to the essential service water system l
on low suction pressure does not constitute a violation of l
the FSAR commitment quoted on page 6-11 of the IDI Report.
l The suction alignment valving consists of simple open/close valving and, as such, does not constitute control lineups l
in the sense of the FSAR discussion.
However, to clarify I
and put on record our intentions in this regard, page 14.2-31 of the FSAR has been revised via Amendment 43 (Sept. 1983)
(copy attached).
The IDI Team did indicate that they felt the test in question could be performed using " water of better quality".
We wish to point out that the essential service water system has been filled for some time with strained river water.
Even should extensive flushing and cleaning be done, it would be difficult to ensure that hideout contamination would not result in water chemistry that does not meet the stringent requirements for steam generators.
Secondary side water chemistry has been a matter of significant interest in recent times, its importance has gained wider recognition.
Moreover, the as
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F6.9-1
l B/B
(
piping from the essential service water system consists of a fairly short run of 6-inch pipe, containing two gate valves, fed from a 14-inch header.
(The pump suction headers are six inch lines).
This system is pressurized by the essential service water pumps.
The small confirmatory gain resulting from the test desired by the IDI Team is wholly outweighed by the potential risk to the steam generator chemistry associated with such a test.
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Attachment to Finding 6-9
{
B/B-FSAR AMENDMENT 43 SEPTEMBER 1983 TABLE 14.2-19 AUXILIARY FEEDWATER SYSTEM (Preoperational Test)
J Plant Condition or Prerequisite Prior to core load, prior to and during integrated hot func-tional testing, and during ECCS full flow testing, RCS will be at various temperatures and pressures.
Test Objective To verify the ability of the auxiliary feedwater system to respond to a feedwater demand under any plant condition.
Test Summary The auxiliary feedwater system will be tested prior to hot i
functional testing to verify pump performance over extended periods on recirculation, and at various flow rates.
Motor and diesel-driven pumps will be verified to start under any safeguard situation under any possible control lineup, including restart capability, from any control station, following a protective trip.
Control logic and interlocks for both manual and automatic operation and protective features for motor-and diesel-driven pumps and all power-operated j
valves will be verified for setpoint, indication, and alarms.
All motor-and diesel-driven pumps will be tested for five consecutive, successful cold starts per pump.
i All motor operated valves will be verified to position or reposition to the required lineup f rom any plant condition, i
safeguard situation or suction requirement.
Essential service water booster pumps attached to the diesel prime movers will be verified for flow and cooling requirements of the engine and cubicle cooler.
All flow limiting devices will be verified by line flow checks and identification tab data.
Acceptance Criteria I
The. auxiliary feedwater system supplies feedwater in accordance with Subsection 10.4.9.3.1.
14.2-31 4
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B/B C
FINDING 6-10:
STAMPING OF SAFETY RELATED DRAWINGS This finding addresses the absence of a " Nuclear Safety Related" stamp from a control logic diagram (M-4037-1AF04) which shows safety related equipment.
RESPONSE
We agree with the conclusions of the finding that the control logic diagram M-4037-1AF04 was not correctly stamped " Nuclear Safety Related".
We have corrected this minor error and reissued the drawing showing the " Nuclear Safety Related" stamp.
In addition, Sargent & Lundy has taken action to review the complete set of logic diagrams and the control and instrumen-tation diagrams for correct stamping and has revised, as required, those drawings which are not in conformance with the stamping requirements.
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FINDING 6-11:
IDENTIFICATION OF SAFETY-RELATED COMPONENTS This finding addresses an alleged omission in classifying certain safety related controls cabinet modules as safety related.
RESPONSE
We do not agree with the conclusions of this finding.
The channel test and relay card modules in question are contained in Control System Cabinets 1PA33J and IPA 34J.
Because of project practices which require instrument numbers only-for those instruments which need the documentation of setpoint and/or calibration data, these modules were not numbered.
In this case, instrument numbers are not required in order to provide a classification for the modules since the classi-fication is provided by virr.ue of the cabinet classification in the project Equipment List, classification of the instru-ment loop in the procurement specification and classification of devices on wiring and schematic diagrams.
We believe the current Sargent & Lundy documentation provides adequate classi-fication of the modules as safety related hardware, i kJO
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B/B I A)
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FINDING 6-12:
EQUIPMENT STATUS DISPLAY SYSTEM CRITERIA This finding describes a disagreement between the design criteria and the logic diagrams for deactivation of the Equipment Status Display.
RESPONSE
We disagree with the IDI Team's statement that the Equipment l
Status Display (ESD) system design criteria's statements regarding deactivation are contrary to IEEE 279 or the impli-cation that FSAR commitments have not been met.
Regulatory Guide 1.47, Section B (Discussion), specifically states that "certain safety-related functions of a nuclear power may be bypassed or made inoperable during the performance of periodic tests or maintenance".
The ensuing regulatory guide discussion describes requirements for an indication system based on by-passes or inoperability during testing and maintenance.
The ESD system is designed to meet these requirements and to meet IEEE 279 requirements in specific cases.
However, the generality regarding deactivation of the ESD after receipt of a safety injection signal, as listed in the design criteria DC-ME-07-BB, will be revised to reflect the i t C actual implementation as noted on the logic diagrams (i.e.,
the auxiliary feedwater system is not deactivated after receipt of a safety injection signal).
This specific design criteria will be revised since it provides important documentation for the ESD system, i
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i Section II Detailed Responses tc Unresolved Items i
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B/B
' p\\j UNRESOLVED ITEM 2-1:
DIESEL ENGINE EXHAUST PIPE The licensee should assure that the vulnerability of the auxiliary feedwater pump diesel exhaust pipe and its
~
hinged cap has been considered.
The revisions made to the exhaust piping of the emergency diesel generators should be reviewed for applicability to the auxiliary feed-water pump diesel exhaust piping.
4
RESPONSE
Rupture discs were added to the exhaust lines of the emergency diesel generators since a postulated tornado missile could potentially disable the exhaust line for each of the redundant diesels causing both trains to be inoperable.
Only one of the two auxiliary feedwater pumps is diesel driven and poten-tially vulnerable to tornado generated missiles.
The motor-driven auxiliary feedwater pump and components are entirely inside the auxiliary building.
Approximately 7 feet of auxiliary feedwater pump diesel exhaust piping is exposed above the auxiliary building roof.
It was our engineering judgment that this short length of 16-inch Schedule 40 pipe would not crimp completely closed from a l
\\
tornado missile and incapacitate the auxiliary feedwater pump diesel.
Sargent & Lundy has performed a detailed review to calculate the affect of the tornado missiles on the auxiliary feedwater pump diesel exhaust stack.
The review has confirmed our engineering judgment that the auxiliary feedwater system will not be incapacitated.
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UNRESOLVED ITEM 2-2:
MAXIMUM PIPING PRESSURE All systems shall be verified to have calculations or documentation in place to assure the maximum pressure is within the pipings capabilities.
RESPONSE
The maximum pressure for the safety-related systems are being documented in the individual system piping design specifications.
The maximum pressures are checked against the latest design minimum wall calculations ( 4-9-8 2) and Sargent & Lundy Mechanical Standard MES-2.5, Maximum Allowable Working Pressures to verify the adequacy of the affected piping.
We trust this additional information resolves this item.
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UNRESOLVED ITEM 2-3:
PIPE TIP DEFLECTIONS Sargent & Lundy Project Instruction PI-BB-38 states that subsequent to pipe whip restraint design the final pipe tip deflections are transmitted to Project Management Division for review.
This had not been done prior to the inspection and criteria for this review are not stated.
RESPONSE
Final pipe tip deflections became available during the IDI and were transmitted to the Project Management Division (PMD) as noted in the Inspection Report (IDI Report Reference 2.150).
This is consistent with the schedule for finalization of pipe whip restraint design.
Work in this area at late stages of construction is not abnormal because as-built dimensions can affect restraint calculations and design.
The Sargent & Lundy project instructions are written to stan-dardize repetitive design functions and to define departmental interfaces.
Project instructions are not written to define all design work.
Final disposition of the tip deflections is being coordinated by the Project Management Division within the mechanical department.
The final report will fully document i (~')
the procedures and results.
v l
This review is underway.
The tip movements inside containment have been superimposed on composite drawings to check for i
interference with mechanical and electrical equipment.
An additional field check is being made to insure that no field routed systems are within the tip deflection envelope.
Auxiliary Building breaks are treated by checking the location of the break to ensure that no safe shutdown systems are in l
the same area.
Because of the compartmentalized design of l
the Auxiliary Building, no field check is necessary.
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B/B UNRESOLVED ITEM 3-1:
ROD HANGERS AND PIPE REST SUPPORTS The licensee should either consider the use of realistic support stiffnesses in piping analyses, or develop more conservative acceptance criteria for rod hangers and pipe rest supports.
RESPONSE
There is a contradiction in the IDI Report, since on Page 3-9, the Report states "the use of infinite support stiffness met the licensing commitments and our audit calculations did not indicate any adverse effects for the sample we selected."
Thus, the IDI Team basically agrees with the procedure used as an acceptance criteria for rod hangers and pipe rest supports.
In light of the many conservatisms introduced into the design by the development of seismic inputs for the piping analysis, Sargent and Lundy believes that the " uplift check" method utilized is a more than an adequate engineering approach to justify the use of rod hanger and pipe rest supports.
I The alleged unpredictability of the seismic support loads l( )
is something that was never discussed or brought up during the IDI.
Sargent & Lundy subsequently reviewed Reference j
3.113, " Piping Stress Evaluation For Independent Verification of Sargent & Lundy Design SAFF-261-83," as performed by EG&G and could not substantiate "the 70% unpredictability of weight, thermal, and seismic loads" statement in the IDI Report attri-buted to Reference 3.113 on Page 3-6 of the Report.
Sargent & Lundy agrees with the IDI Team's conclusion on Page 3-9 of the IDI Report and the conclusion arrived at in Reference 3.113 which states,
...results of the study indicate that the analysis procedures used by Sargent & Lundy adequately examine piping stresses per ASME Code Criteria."
We therefore believe that adequate conservativeness exists and that all licensing commitments have been met such that this item is considered resolved.
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UNRESOLVED ITEM 3-2:
LATERAL VIBRATION OF STRUTS AND RODS S&L employs no criteria to evaluate the possibility of lateral vibrations of struts and rods and no criteria was available to evaluate the frequency of supports in the unrestrained direction.
RESPONSE
S&L's practice for designing supports is considered more than adequate for the following reasons:
1.
In conjunction with the applied piping loads, the self-excitation of the auxiliary steel provided for the purpose of supporting the piping is accounted for in sizing and designing the structural steel members.
2.
The criterion outlined in Paragraph 6.4.5 of Engineering Mechanics Division's " Lesson Plan for Training Personnel in Piping Analysis," EMD TP-1, Revision 3 requires an evaluation to include the mass effect of support hardware in piping system dynamic analyses.
By doing so, the generated loads contain the additional loads due to hardware self-l()
excitation.
l
'dased on the above discussion, S&L believes that a consistent I
support design practice does exist and the approach that S&L has adopted in addressing this issue is adequate.
We trust this explanation resolves this item.
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B/B C
UNRESOLVED ITEM 3-3:
DIESEL DRIVE STIFFENER The seismic qualification test determined that the coupling between the diesel engine and the gear box was loose due to excessive flexibility in the engine mounts.
Additional gussets were added to the gear box to resolve the problem, however, vendor drawings were not updated and issued to Sargent & Lundy.
It should be assured that this design change had been incorporated in the other diesel drive units (Byron 2, Braidwood 1, 2).
RESPONSE
The diesel engine supplier, Stewart and Stevenson, has confirmed in writing that the gussets were added to all four of the auxiliary feedwater pump diesels' gear boxes.
Stewart and Stevenson in-house production drawings were revised to show the subject gussets on 10-5-79.
Due to the small detail, it was impossible to show these gussets on the general outline and installation drawings submitted to Sargent & Lundy.
Sargent
& Lundy and Commonwealth Ediscn have verified that the gussets are installed on the diesel engines at the site, which should
}
resolve this item as requested.
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B/B
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UNRESOLVED ITEM 3-4 :
PIPE SUPPORT ADDED MASS Westing, house analysts do not consider the additional mass or weight that might be attached to piping by trapeze hangers, and we found no instructions for considering this e f f ec t.
It should be verified that there are no such cases in the Westinghouse scope or that any cases that might exist are being properly considered.
RESPONSE
Westinghouse analysis criteria does not require the inclusion of the weight of trapeze hangers in the piping model.
The effect is considered small, and furthermore, this type of support design is used infrequently by Westinghouse.
Clearly there has been no violation of any Westinghouse criterion or FSAR Gommitment associated with this unresolved item.
Westinghouse has reviewed a random sample of 377 small bore supports selected from a total Westinghouse scope of 6291 in the containment and auxiliary building.
The review found two trapeze hangers, one in Subsytem ISD01 and one in Subsystem ICV 25.
The weight of the trapeze in Subsystem 1CV25 is less i 0.
than one foot equivalent run of pipe, and therefore would have a negligible effect on the piping.
The weight of the trapeze in Subsystem ISD01 corresponds to approximately 3.5 feet of pipe.
It is one support of a total of 31 on this subsystem.
To demonstrate the effect of the trapeze weight, Subsystem ISD01 was analyzed with the additional mass of the trapeze support.
The anlaysis results show that all piping stresses remain within code allowables and that the changes in support loads result in interaction ratios less than code allowable stresses.
The subsystem designs for the two cases where trapeze supports were found would be the same with or without trapeze weights in the piping model.
Based on the results of the sampling which indicate a very small number of trapeze supports are used and the results of the test analyses that show the inclusion of trapeze weight did not change design, we conclude that the criteria as presently followed is sufficient for the piping analysis.
We trust that this additional information is sufficient to resolve this item.
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B/B Ip]
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UNRESOLVED ITEM 3-5:
FUNCTIONAL CAPABILITY CRITERI A Unresolved Item 3-5 addresses the difference in functional capability criteria for NSSS and BOP piping, stating that Westinghouse is using criteria approved for NSSS piping for analysis of BOP piping.
RESPONSE
I Westinghouse agrees with the IDI Team that this unresolved item does not represent any significant technical problem.
Westinghouse disagrees with the IDI Team that this unresolved item is an instance of Westinghouse meeting the licensing commitment appropriate to the NSSS rather than BOP.
The Westing-house functional capability criteria has been approved by th e NRC, as stated in the IDI report.
It is an acceptable method for addressing the NRC concern on the issue of demon-strating Class 2 and 3 functional capability.
By definition, the Westinghouse criteria are BOP criteria since the NSSS (the reactor coolant loop and primary equipment) contains only Class 1 piping.
Furthermore, Westinghouse has been using this criteria as a design basis for, all Class 2 and 3 stain-i less steel piping and not limiting it to a sampling process
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which would have satisfied the original NRC request.
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l To resolve this item, the FSAR will be amended to clarify the use of this criteria as defined in IDI Report Ref. 3.190 for Westinghouse scope of piping.
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B/B UNRESOLVED ITEM 3-6:
FORCE AND MOMENT DERECTIONS The unresolved Item concerns the assurance of the correct application of the orientation of allowable nozzle loads received by Westinghouse from Sargent and Lundy for the i
spray additive tank (ICS 0lT).
RESPONSE
The procedures applied to verify coordinate orientation of equipment loads (as described below) and the specifics per-taining to the loads-in the cited unresolved item (as described below) assure that the correct orientation is used in the equipment orientation.
The S&L coordinate system for allowable nozzle loads has been documented in S&L letter SLWC-5447, dated 12-1-83.
For all S&L-supplied equipment, the actual loads are sent to S&L for final approval.
These loads are transmitted in a well defined coordinate system.
The coordinate system has been documented in Westinghouse letter CAW-6625, dated 11-14-83.
The correct application of the proper coordinate system is assured by the final S&L verification.
te The coordinate system of S&L-supplied allowable loads for the spray additive tank was not defined for the three (3) force (Fx, Fy, F ) components and one moment (resultant moment, Me) limits provided.
(Note that the coordinate system definition is required for only forces since the resultant moment limit is independent of orientation.)
The actual forces on the tank were shown in the reviewed calculation to be less than the minimum force component limit; in fact, the resultant of the actual forces is less than the minimum force component limit (see Table F3.6-1).
Therefore, the S&L limits were met independent of the coordinate system.
The coordinate system for the allowable loads for'the spray additive tank is consistent with the coordinate system identified in SLWC-t 5447.
I In the design process, the receipt of nozzle load allowables for the piping analysis is not mandatory and is not always available for all equipment.
The evaluation of the nozzle loads from the piping analysis by Sargent & Lundy is the basis for establishing final acceptability of the equipment.
We believe this item is now resolved.
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! iO TABLE U3.6-1 SPRAY ADDITIVE TANK ALLOWABLE LOADS i
ACTUAL LOADS LIMITS 130 lbs }
Fy = 3500 lbs F
=
x l
270 lbs 316 lbs F.
F
=
Y l
2=
850 lbs F
100 lbs F3=
850 lbs
=
2 M
5000 in-lbs M
= 25,000 in-lbs
=
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B/B
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'T UNRESOLVED ITEM 4-1:
OPENINGS IN WALLS There are openings in the shear walls along Column Lines 10 and 26 which appear to be substantial in size.
It appearrd that the designers had made a judgment that the openings were not sufficiently large so as to change the distribution of forces in these walls.
However, we could find no evidence of how this judgment or other similar judgments were made.
The licensee should determine whether the judgment was valid for this opening and, in general, the bases upon which similar judgments have been made.
RESPONSE
For openings in walls, all interrupted reinforcing bars are replaced at the edges of the opening, and additional reinforcing is placed around openings to reduce any load stress concen-trations which could be present.
We agree that an engineering judgment was made that the openings in Column Lines 10 and 26 were not sufficiently large so as j
to change the distribution of forces in these walls.
This s
! j ('d) same type of engineering judgment was used in the design of other shear walls in the plant.
Also, there are cases where the original design engineer judged the redistributica of stresses around openings may have been significant.
In these cases, such as along the base of the shear walls along Column Lines 6 and 30, an in-depth analysis of the redistri-bution of forces around large openings was made, and the rein-forcing steel was located accordingly.
j In order to document the above-noted engineering judgment, l
additional calculations will be performed to show that redistri-bution of forces is not a significant effect for various cases of wall openings.
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l B/B LO UNRESOLVED ITEM 4-2:
TOP REINFORCING FOR SLABS In the design of slab 4AS53, the boundary condition where j
the slab framed into a wall was considered hinged, while the actual boundary conditions are such that a fixed support would be more appropriate for design.
Negative moment 4
'l steel equal to that at the continuous support was provided, and the potential problem was avoided since the designer supplied more than adequate. reinforcing steel.
- However, the Team was concerned since it found no criteria addressing this situation and this could lead to a situation where insufficient reinforcement would be provided.
In the Team's judgment the licensee should verify that adequate top reinforcement was provided for all non-continuous slab supports.
RESPONSE
As shown on Sargent and Lundy Drawings S-690 and S-790 (I.D.I.
References 4.64 and 4.77), negative moment steel, equal to i
that at the continuous support, was provided at the junction to the wall of slab 4 ASS 3.
This negative steel is not required I.
g by design according to Calculation 7.43 (I.D.I Reference 4.78),
but is provided as a standard detail on Sargent & Lundy drawings S-690 and S-790 to increase the factor of safety for slabs.
The design conservatively assumes that the junction of the 1'
slab and wall is hinged.
By assuming a hinge with no moment capacity, the maximum possible positive moment in the center of the slab is considered.
Steel is provided for this positive moment which is all that i
is required for a safe design.
Any negative moment steel added at a wall or other noncontinuoer support provides an i
additional factor of safety.
We trust this additional explanation t
resolves this item.
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i UNRESOLVED ITEM 4-3:
ANCHOR BOLT DIAMETER In November 1981, a Field Change Request was initiated by Hatfield Electric Company (Reference 4.53) which stated that no mounting detail was available and proposed a location for the two sets of racks as well as a mounting detail which was identical to that used in the calculation except it utilized 1/2-inch diameter expansion anchors instead of 1/4-inch anchors.
The drawing (Reference 4.52) was revised (Revision L) to provide for the use of 1/2-inch anchors on Unit 1 and 1/4-inch anchors on Unit 2.
A field check by the team indicated that for Byron,1/2-inch anchors had been used.
Upon questioning, it was determined that the 1/2-inch diameter, 7-inch long expansion anchors had been substituted in Unit 1 because of the need to have an anchor with sufficient length to penetrate through a 3-inch thick topping floor into the base concrete.
Currently, the drawing continues to call for 1/4-inch diameter anchor bolts for Unit 2.
Since this need may also exist for Unit 2, the licensee should resolve what will be required on Unit 2 and make any necessary revisions to the drawing.
I
RESPONSE
The design calculations for this mounting detail show that a 1/4-inch diameter expansion anchor is adequate to transfer the design loads.
Because the installation in question was made to a finish slab, a longer anchor was used as required i
by Form BY/BR/CEA.
This form requires the contractor to set l
expansion anchors into rough concrete.
Therefore, at the i
time of installation an anchor with the proper length was I
u se d.
If a similar installation is made to a finish slab in Un'it 2, the substitution of a longer anchor will also be made.
However, for purposes of clarity, drawings will be revised where applicable to specify the longer anchor.
We consider this item to be resolved.
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B/B UNRESOLVED ITEM 4-4 :
STARTING BATTERY SUPPORT RACKS CLEARANCE This item concerns a field inspection of the Auxiliary Feedwater Diesel Starting Battery Rack, which identified a space between the end of the battery and the end of the rack.
The Seismic Qualification Report for the batteries and racks had not yet been submitted, so no final resolution could be made as to whether the space was acceptable.
RESPONSE
Sargent & Lundy agreed at the time of the IDI to ensure that the final battery / rack installation is consistent with the arrangement used for its seismic qualification.
Subsequent to the IDI, the Seismic Qualification Report for the battery / rack was submitted.
The report indicated a spacer installed between the battery and the end of the rack.
Sargent &
Lundy has contacted the vendor to provide revised installation drawings and additional mounting hardware to make the installation agree with the Seismic Qualification Report.
If
/
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U4.4-1
B/B UNRESOLVED ITEM 4-5:
AS-BUILT REVIEW With the data developed from the execution of these procedares, Sargent & Lundy will undertake a program to review the structural design adequacy based on the as-built conditions.
It is noted that at the present time Sargent & Lundy envisions a 100% recheck program on loads from all types of hangers except for loads less than the limits provided for in PI-BB-34.
No separate procedure or instructon had been developed to address how the data being gathered are to be utilized and integrated to make the final as-built loading checks.
Considering the nature of the work to be done, the licensee should assure that such guidance is developed in the near future to support completion of the checks.
RESPONSE
A separate procedure exists in Calculations 18.1.2.5 (Revision 0 dated 3-16-83) and 18.2.9.5 (Revision 0 dated 3-14-83) which was developed to address how the data being generated via PI-BB-34 is to be utilized and integrated to make the final as-built loading checks.
This information was readily avail-i s(~;)
able at the time of the IDI but the IDI Team did not indicate i
that this was an item of concern.
The procedures are included in the Design Control Summaries of each item being checked.
There is no PI for final load check since once the loads are gathered according to PI-BB-34, there is no interdisciplinary work required.
The final load check is performed solely by the Structural Department.
We trust this addition 31 information resolves this item.
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B/B UNRESOLVED ITEM 5-1:
ANALYSIS CONCERNING LACK OF SEPARATION f
This item concerns specific open questions described in i
Notes 1, 2 and 3, and in Samples A, D and E of the report.
]
Note 1 concerns cable tray separation.
Note 2 indicates t
that a cable touching the sidewall of a cable tray is d
a violation of Hatfield Procedure #10.
Note 3 concerns touching of safety-related and non-safety-related conduits.
Sample A concerns support of a tubing bundle in the remote shutdown panel.
Sample D concerns safety-related and non-safety-related cables in close proximity in panel IPAllJ.
Sample E concerns safety-related and non-safety-i related cables in close proximity in Switchgear Bus 141 Cubicle 16.
RESPONSE
The response to this item is divided into six parts which
'i address each of the open questions:
J i
Note 1.
The cable tray separations identified with Note 1 were identified on Cable Separation Criteria Violation l
Reports prior to the IDI.
The IDI Team chose not
(
to acknowledge the existence or specific resolution in cable separation conflict reports.
Note 2.
IEEE Std-384 does not specify a separation distance from a cable-in-air to a cable in a raceway.
In 4
regard to Class lE/-Non-Class lE separation, Hatfield Procedure #10 requires a 12-inch distance from one cable-in-air to another cable-in-air.
This distance does not apply for a cable-in-air to a cable in a raceway (e.g., cable tray).
Hatfield Procedure #10 allows a cable-in-air to touch the sidewall of a cable tray.
Note 3.
There are safety-related and non-Safety-related con-duits in the auxiliary feedwater tunnel that touch each other.
There is in progress a program to address all potential instances of safety-related and non-safety-related conduits installed with less than 1-inch separation.
The detailed procedures for this program are still being developed but will consist essentially of (a) an onsite physical walkdown of a statistical sample of all safety-related conduits, (b) an engineering analysis of all identified instances where the conduit separation is less than 1 inch,
!O and (c) a modification of the conduit arrangement in any instance where the less than 1-inch separation 4
US.1-1
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B/B Dq cannot be justified by analysis.
This program did not result from the IDI.
The analysis will (where possible) show that the separation of less than 1-inch does not degrade the safety-related circuit.
The detailed procedures for the walkdown will be completed by December 15, 1983.
Sample A.
The tubing bundle has been reviewed by Westinghouse (Responsible for Seismic Qualification of the panel),
and an additional support will be installed.
Sample D.
All non-safety-related cables which could be in close proximity to safety-related cables in panel IPAllJ were identified in the Interface Review Report Index (i.e.: scheduled for analysis) at the time of the IDI.
The design provides for the maximum cable separation attainable, considering the limitations imposed by the panel size and the terminal block arrangement; where the " required" separation can not be attained, the lesser separation is analyzed.
Sample E.
It is noted that the Equipment Number " LAP 95ED" in the IDI Report is in error.
The correct equip-l l(v')
ment number is " LAP 05ER".
All non-safety-related j
cables which could be in close proximity to safety-related cables in Switchgear Bus 141 Cubicle 16 were identified in the Interface Review Report Index (i.e.: scheduled for analysis) at the time of the IDI.
i O
US.1-2
B/B
)
UNRESOLVED ITEM 6-1:
LOSS OF OFFSITE POWER TEST Commonwealth Edison Byron Station personnel stated that a complete test of each automatic initiation signal was scheduled during the engineered safety features pre-operational test.
A number of preoperational test pro-cedures (References 6.99, 6.100, and 6.101) and the loss of off-site power startup test procedure (Reference 6.102) were inspected at the Byron Station to determine how the engineered safety feature actuation system interface would be tested.
The loss of off-site power test will automatically i
initiate auxiliary feedwater and numerous other engineered safety feature systems when circuit breakers are opened to simulate a lose of off-site power; however, Step 9.1.20 of this procedure only requires verification "that at least one train of the above (listed) equipment is running."
This may have been an inadvertent omission from the test procedure acceptance criteria as the procedure was not yet signed.
Since both trains are required to start automatically under these conditions, the licensee should ensure that the test procedure requires verification that both trains start during the test.
RESPONSE
The engineered safety features actuation system tests and various tests of individual safety systems will provide full, and, in some cases, redundant testing of individual automatic initiation signals.
These will be done during preoperational testing.
The loss of offsite power startup test does not need to reconfirm each of these signals.
The requirement in Step 9.1.20 of this test is intended to ensure that adequate flow is maintained to the steam generators.
It is not intended to duplicate the preoperational test program.
8 However, the Chapter 14 FSAR commitment for the loss of offsite power test involves verification of diesel generator starting, and load sequencing.
Therefore, the test procedure should verify that the motor-driven auxiliary feedwater pump is properly sequenced onto the vital 4160V bus.
The procedure will be modified to conform to FSAR commitments.
t We believe that this response serves to resolve this matter.
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UNRESOLVED ITEM 6-2:
PRESSURE SWITCH QUALIFICATION This unresolved item describes the current, incomplete status of the equipment qualification for the auxiliary feedwater low suction pressure switches and the fact that complete documentation was not available for the incomplete test.
RESPONSE
Within the context of the whole equipment qualification program, it is agreed that the pressure switch qualification is incom-plete; however, we disagree with the inclusion of this particular component's qualification program as an unresolved IDI item.
The status of the qualification program for the pressure switches was fully explained to the IDI Team.
A brief discussion of the status of the pressure switch qualification follows:
Equipment qualification for the auxiliary feedwater pumps suction pressure switches was initiated in October 1982 for a previously specified instrument with a range of 0-100 psig.
l Subsequent to the beginning of this test (which includes aging),
l the pressure range of the switch was revised to 30 inches
-s l h of Hg to 20 psig.
The testing was allowed to proceed on the previously specified pressure switch model, with the intention of providing qualification of the current pressure switch model by similarity.
If this cannot be done, a new test would be required.
Since, at the time of the IDI, the test of the previous model was not complete, a final disposition for the current pressure switch model's qualification was not avail-able as explained to the auditor.
As of October 1983, the test report has been received and is being evaluated.
Sargent
& Lundy was fully aware of the status of the qualification program for the auxiliary feedwater pump suction pressure switches and will complete the program to meet the licensing requirements for the Byron /Braidwood Stations.
We consider this item to be resolved.
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U6.2-1
B/B Q'
UNRESOLVED ITEM 6-3:
CONTAINMENT ISOLATION REQUIREMENTS The Team found that FSAR Table 3.2-1 did not reflect a number of safety-related containment isolation require-ments that were currently reflected in the design criteria document of March 1983 (Reference 6.37).
This did not represent a finding because it is not uncommon for FSAR changes to follow design changes by a few months and the team did not determine whether or not an FSAR change was pisnned.
The licensee should assure that the inconsistency is corrected.
RESPONSE
The revisions to FSAR Table 3.2-1 to bring it into agreement with Revision 6 of CC-ME-01-BB
(" Classification Criteria of structures, Systems and Components"), dated 3/16/83 were included in FSAR Amendment 42.
Amendment 42 is dated May 1983 and i
was transmitted to the NRC by letter dated 5/31/83.
Had the IDI Team checked at the time they conducted their IDI, they would have found that an FSAR amendment which incor-ll(-)
porated the changes of Revision 6 of CC-ME-01-BB into FSAR Table 3.2-1 was in fact in preparation at that time and was issued for use to all FSAR holders shortly thereafter.
The
" inconsistency" has been corrected.
This item has no safety significance and in fact is characteristic of the normal FSAR revision process.
We consider this item to be resolved.
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U6.3-1
B/B
[O UNRESOLVED ITEM 6-4:
MANUAL START-STOP SWITCH This finding concerns the location of local start /stop controls for the motor driven auxiliary feedwater pump.
The controls are provided at the Remote Shutdown Panel.
The Westinghouse functional requirements specify local manual start /stop controls local to the motor driven pump.
The team did not determine the degree of need for the switch location reconmended by Westinghouse or the licensee's reason for the actual switch location.
RESPONSE
We agree with statements in the Report that the design meets current regulatory requirements and that the licensee is not required to follow every Westinghouse recommendation.
The Westinghouse recommendation was considered in determining the design location of the switch.
Sargent & Lundy and Commonwealth Edison Company determined that the local start /stop controls should be located at the Remote Shutdown Panel and that a control switch local to the motor driven pump is not required.
We consider this issue to be resolved.
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OBSERVATION 2-1:
CONDENSATE STORAGE TANK This observation recommended that a sizing calculation on the condensato storage tank be prepared to preclude the possibility of switchover to the relatively impure essential service water source for events of moderate frequency.
RESPONSE
Proposed Commonwealth Edison Technical Specifications require verification every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that the condensate storage tanks are maintained such that a minimum of 200,000 gallons of water are available for the exclusive use of the auxiliary feedwater system.
Each condensate storage tank holds 500,000 gallons of water which provides sufficient capacity for normal makeup and the minimum 200,000 gallons reserved for the auxiliary feedwater system.
No additional calculations are required.
We consider this item closed.
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B/B U(3 OBSERVATION 4-1:
SPREADING LIVE LOADS TO ADJACENT BAYS Section 10.0 of Design Criteria DC-ST-03-B/B (Reference 4.31) contained some details beyond those specified in the FSAR on procedures and assumptions to be made in con-sidering various load effects.
The Team considered one item as constituting an unconservative assumption.
This involved a statement in Section 10.2.1.1.3.1 regarding the input live loads from piping and cable tray loads within a given bay being spread onto two bays.
- However, since the actual loading was scheduled to be checked in the final load check, the statement would not lead to deficiencies.
Accordingly, this was not considered a Finding or an Unresolved Item.
It is mentioned as an item for licensee consideration with respect to revising the wording.
RESPONSE
l The design method in the Design Criteria for cable tray and
(
piping loads was apparently misinterpreted by the IDI Team.
l Loads shown are for all cable trays or pipes in a given direction
!(\\d~')
in a given area.
They are shown in one bay for simplicity, however, they represent loads in the entire area.
Also, the I
final load check will reevaluate the slabs for the actual, as built, loading condition to verify the adequacy of the original of loading assumption.
We trust this additional explanation resolves this matter.
l l
l l
o 04.1-1
B/B OBSERVATION 4-2:
WORDING OF CRITERIA On page 11-6 of Section 11 of Design Criteria DC-ST-03-B/B (Reference 4.31), the radical sign in equation 4a exends over the last term.
The terms of equations 4a
., g and 4b are apparently total required steel area for shear and flexural while N /f is that steel area due to gravity load which can be de8ucted.
It is not explained that the total areas for shear and gravity loads should be proportioned.
In Section 29.0 on page 29-3, spans 1 and le arereferredtoasbeingdefinedbyACI318-7k in which these spans are center-to-center spacing.
The method of analysis is specified as being from ACI 318-63 which defines clear spans rather than center-to-center-spans.
The above items are editorial in nature and are not considered findings or unresolved items.
They are mentioned as itens for which we recommend that the licensee consider clarifying the wording.
RESPONSE
A.
The radical sign in Equation 4a which extends over the last term is a typographical error.
The proper equation
<~-
d, was used in the design because the shear wall Design Control Summary and shear wall design in Calculation 7.12.6 use the correct formula.
The design criteria, page 11-5 will be revised to show the correct Equation 4a.
B.
The definition of N /f will be clarified in the design criteria to indicat8 that the total dead load is proportioned linearly to the respective sections in the wall (flange and web).
The shear wall design was performed using the correct method.
C.
The use of l and 1 in the design of slabs was used as i
9 a selection criteri3 to define the analysis method to be used.
Once the criteria was established, the applicable values for length were used depending on the criteria selected.
We trust this additional information resolves this item.
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04.2-1 1
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B/B C
OBSERVATION 4-3:
EDITING POSITION DESCRIPTION MANUAL The Sargent & Lundy Position Description Manual provides a job description for each job classification.
Position descriptions were reviewed from the department manager down into five of the nine divisions of the department.
In general, it was observed that Sargent & Lundy places high educational and experience requirements on engineering positions and that there is a graded increase in those requirements as the job responsibility increases.
Several items listed below were noted during the review.
(1)
None of the positions utilize professional engineering registration as a requirement.
(2)
The department manager is not required to have more experience than three of the key division heads.
(3)
The Structural Design Director, who is responsible for four divisions is not required to have experience equivalent to three of the key division heads not under his direction.
! ( f'/)
(4)
The heads of two of the divisions under the Structural Design Director do not have equivalent experience s_
requirements as three of the other key division heads.
(5)
A supervisor in the Analytical Division, and a Super-visory Structural Engineering Specialist in the Specialist Division are not required to have supervisory experience.
These items are observations and do not represent Findings or Unresolved Items, but are mentioned as items that Sargent
& Lundy may wish to consider with respect to editing its Manual.
RESPONSE
(1) Although the requirement for professional engineer regis-trati.cn'is not explicitly stated in the Position Description Manual, the Department Manager and Assistant Department Manager do have professional registration.
The position descriptions for the Department Manager and Assistant Depart-ment Manager will be revised to indicate that professional registration is required.
l (2) The experience requirement of 10 years is an appropriate I
limit for requiring a minimum experience level.
Beyond fr(
that, other qualities and capabilities which are not based on years of experience become more important.
Therefore, these positions all have the same limit.
4 04.3-1 l
l
B/B AbJ (3) The Structural Design Director position description has design experience equivalent to or more than that of the Engineering Division Heads he is responsible for.
The other two Divisions (Draf ting and Architectural Design) do not require a college degree and therefore require stronger emphasis on years of experience rather than college education, due to the nature of their responsibilities.
(4) Due to the different responsibilities between these Divisions, it is felt it is not necessarily appropriate to have equi-valent experience.
(5) The requirement for a Supervisor in the Analytical Division and the Engineering Specialist Division does not include supervisor experience because this could be the first supervisor's position this person is holding.
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B/B gx OBSERVATION 4-4:
HANGER LISTS AND LOAD SUHMARIES Instruction PI-BB-34 places the responsibility of preparing and updating hanger location drawings, hanger load summaries and the mechanical hanger lists on the Structural Department for the project.
Based on the team's judgment, the project instruction as written is believed to create the potential for errors and omissions since the Mechanical Department personnel would generally be more familiar with the piping hangers they designed than would the personnel from the Structural Department.
This is not a finding or an unresolved item but an item which the licensee may wish to consider.
RESPONSE
The present system for preparing and updating hanger load summaries is well established, is in accordance with existing procedures, and in our judgment, the most suitable way to perform the tasks.
The hanger drawing contains all load and location information required for design and installation.
This information constitutes the information required by the Structural Department for the load checks of the plant
, [\\ 'j structures.
The Mechanical Department generates an up-to-date listing of every hanger and its current revision, which is revised and issued on approximately a monthly basis (Mechanical Hanger List).
The Hanger Load Drawing (HLD) is a consolidation of load location information controlled by the Structural Department.
It includes loads from HVAC, cable tray, conduit, junction box, Westinghouse, NPS, and mechanical hanger loads.
The preparation of the HLD and hanger load summary (HLS) is an interim step in the overall structural design process.
Famil-iarity with the structure is the essential element in formulating j
the HLD and HLS since both are structural design tools.
Produc-l tion of an HLD and HLS require certain judgments that the l
Structural Department is qualified to make.
For these reasons, I
we believe the actual transfer of load and location infor-mation is best handled within Sargent & Lundy through the hanger drawings and drawing list.
We consider this item to be closed, i
04.4-1 l
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B/B
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(
OBSERVATION 4-5:
HANGER LOAD
SUMMARY
INSTRUCTION In the definition of the hanger load summary in Section 2.2 of instruction PI-BB-34, it appeared thac the intent is to also include those component supports whose load is applied to concrete slabs, walls, and framing.
- However, this was omitted from the instruction.
This is not a finding or an unresolved Item but is mentioned as an item recommended for licensee consideration in structuring the procedures.
RESPONSE
Instruction PI-BB-34 (I.D.I. Reference 1.31) does include component supports whose load is applied to concrete slabs, walls, and framing.
It is only the Hanger Load Summary (HLS) of Section 2.2 which applies only to steel members.
The HLS form is not required in the final load check of slabs and walls.
Section 4.1 of PI-BB-34 states that loads attached to slabs, concrete beams, walls, and steel are required.
Also, Table I Indicates what information is required for each of these items; thus, nothing has been omitted from the (f~)/
instruction.
Once the loads have been received by the Struc-s_
tural Department, the method of summarizing loads, other than those attached to steel, for the final load check is explained in the Design Control Summary for final load check of elabs, concrete beams, and walls.
Therefore, we do not consider a procedural change to be necessary.
1 N
N 04.5-1
B/B OBSERVATION 4-6:
FLOW DIAGRAMS The team noted that flow diagrams would improve instructions PI-BB-34, 35 and 37.
This is not a Finding or an Unresolved Item, but is mentioned as an item recommended for licensee consideration from the standpoint of improving procedural guidance and control.
RESPONSE
PI-BB-34 will be updated such that under Section 6, " Flow Chart",
the word "none" will be replaced by "none required."
A flow chart is not required in this PI since all flow of information is a simple, one-step process.
Various parties send the infor-mation outlined in the PI to the Structural Department.
A flow chart would be simplistic and of no value.
For PI-BB-35 and PI-BB-37, a flow chart will be added at a future revision to further clarify these instructions.
After this is completed, this item will be closed.
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B/B A
OBSERVATION 5-1:
PHYSICAL SEPARATION As noted in Section 7.7.2.2, there were instances where there was sufficient room to provide separation inside of cabinets rather than using analyses to justify the lack of separation.
It appeared, however, that virtually all instances may be resolved with analyses to justify the existing conditions.
This does not represent a finding or an unresolvad item because such analyses, if they are adequate, satisfy the licensing commitment and accepted industry standards.
Accordingly, the matter is mentioned as an item recommended for licensee consideration.
RESPONSE
It is Edison's position that two basic criteria must be satisfied relative to separation inside cabinets.
The first involves satisfying licensing commitments.
The second involves satisfying Edison's own concepts of safety, reliability and maintainability.
In specific instances, analyses are employed to satisfy both basic criteria as opposed to literal application of separation distances.
Edison finds either approach acceptable.
IO l
1 05.1-1 i
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k OBSERVATION 6-1:
EQUIPMENT STATUS DISPLAY POWER SUPPLIES This observation concerns the design of power supplies for the ESD System.
The power supplies are provided in accordance with the design criteria.
The Team observed that both power supplies are connected to the same division rather than from different divisions.
RESPONSE
The ESD SyGtem is classified non-safety-related.
Although the Byron design has two non-safety-related divisions, the physical location of the Equipment Status Display panel in the control room requires that cable entry be from the bottom, and only one division is located below the control room.
Both power supplies are taken from the same division because all cable in the lower cable spreading room is associated with one division.
The use of two power supplies from the same division provides more than very limited benefit as stated in the Report.
There are multiple sources of power to the 4-kV bus (including multiple j)
system auxiliary transformer and access to the diesel generator)
Nd that ultimately feed both power supplies.
The use of two power trains greatly reduces the possibility of loss of power due to a single component failure.
G 06.1-1
B/B iO-OBSERVATION 6-2:
EQUIPMENT STATUS DISPLAY QUALITY STANDARDS This observation recommends that additional emphasis be given to the documentation of the design for the Equip-ment Status Display System.
RESPONSE
The Equipment Status Display System is non-saf'ety related, Category II.
However, we will, as part of normal power plant design practice, provide a complete and formal documentation package.
The documentation package for the Equipment Status Display design will include both the hardware and software aspects of the system.
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06.2-1
1 B/B
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OBSERVATION 6-3:
PUMP INSTRUMENTATION LOCATION This observation recommends that the as-built installation of the auxiliary feedwater pump suction pressure switches be reviewed for ease of calibration and maintainability.
RESPONSE
Sargent & Lundy has reviewed the concerns of the IDI Team with the Byron Station Project Construction Department Lead Instrument Engineer and the Byron Station Instrument Foreman.
After inspection of the instrumentation associated with both the auxiliary feedwater pumps 1A and 1B, it has been determined that the instrumentc are accessible for maintainance and cali-bration.
We consider this issue closed.
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