ML20207B703
| ML20207B703 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 12/19/1986 |
| From: | Grimes B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | Walker R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| References | |
| NUDOCS 8612220349 | |
| Download: ML20207B703 (14) | |
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BB DEC N A10 : 5 DEC 191986 Docket No.:
5 J0-4 50-425 MEMORANDUM FOR:
Roger D. Walker, Director Division of Reactor Safety Region II FROM:
Brian K. Grimes, Director Division of Quality Assurance, Vendor, and Technical Training Center Programs Office of Inspection and Enforcement
SUBJECT:
V0GTLE ALLEGATION - CASE.NO. RII-84-A-0145; DESIGN CONCERNS g
Your memorandum to me dated September 11, 1986, forwarded a transcript taken from a confidential source involving various design-related issues regarding the Vogtle Electric Generating Plant and requested our assessment of those allegations. We have reviewed the subject transcript for the identification of the specific allegations and conducted a followup of inspection at the Los Angeles offices of Bechtel Western Power Corporation from October 20, 1986, to October 25, 1986.
Our assessment of the specific allegations is for-warded to you as an enclosure to this memorandum.
An inspection report has not been prepared since discussions with your Mr. Bruno Uryc, indicate that the confidentiality of the alleger would be better protected if our assessment is documented via memorandum.
If you have any questions regarding this matter, please contact R. Parkhill at FTS 492-9592.
e Brian K. Grimes, Director Division of Quality Assurance, Vendor, and Technical Training Center Programs Office of Inspection and Enforcement
Enclosure:
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Assessment of Allegation - Case No. RII-84-A-0145 Docket No.:
50-424, 425 50-530 Facility:
Vogtle Electric Generating Plant Units 1 and 2' Palo Verde 3 Nuclear Generating Station Assessment Team:
R. W. Parkhill, Inspection Specialist, IE (Team Leader)
H-B Wang,. Inspection Specialist, IE R. L. Pettis,. Inspection Specialist, IE E. W. Willhaus, Consultant, Harstead Engineering Assessment' Location:
Bechtel Western Power Corp., Los Angeles, CA Assessment Date:
October 20-26, 1986
Background
In response to a transcript of allegations received by IE via Region II, the above assessment team traveled to the Los Angeles offices of Bechtel Western Power Corporation the week of October 20, 1986, and performed interviews and reviewed applicable documentation relating to the allegations.
Summary The allegations were all technically oriented relating to design issues in the civil / structural, pipe stress and pipe support disciplines.
None of the allega-tions were substantiated in that none resulted in identification of safety significant issues. While the allegations contain factual information, the alleger did not understand the entire design process.
Our investigation of the allegations-indicates that the Bechtel design process adequately reconciled the discrepancies that the alleger indicated as safety concerns and would have corrected similar discrepancies. The design of a commercial nuclear power plant is by nature an iterative process due to the interactions necessary between the various technical disciplines.
During the course of designing a power plant it is not unusual that document inconsistencies exist.
The Bechtel design process is such that these inconsistencies are recognized and reconciled during the design finalization.
Discussion After thoroughly reviewing the transcript of allegations, the assessment team
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identified the following eight allegations for followup and resolution (they are discussed in the order in which they appear in the deposition):
1.-
Seismic Anchor Movement - Civil Discipline The allegation stated that the architect engineer (A/E) reduced the seismic anchor movements (SAM) by as much as 50% in order to qualify some piping analyses.
The IE inspector reviewed the A/E's calculations and design criteria regarding seismic analysis in category 1 structures which contained the piping systems mentioned in the allegation. Through calculational reviews and interviews of
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Q the A/E's analysts, the IE inspector did find in one instance where the SAM was reduced through a civil / structural memorandum to the pipe stress analysts (ref. 1).
In this memorandum, civil / structural allowed one significant SAM reduction which was for the gap between the valve house and the buried pipe.
This SAM was reduced from 0.5 inch to 0.143 inch for safe shutdown earthquake (SSE).
In reviewing the technical background of this reduction, it was observed that the calculated maximum horizontal SAM at the basemat of the valve house was 0.056 inch for operating basis earthquake (0BE).
The use of 0.1 inch by the pipe stress analysts as the SAM for OBE is almost twice the calculated value by civil / structural.
The value of SAM for OBE is assumed to be 70% of that for SSE, therefore, the SAM value used for SSE by the pipe stress analysts is 0.1/0.7 = 0.143 inches.
This value is significantly less than the original value of 0.5 inch, but it is still conservative compared to the calculated value of.056/.7 =.08 (Ref. 2, 9).
The IE inspector also reviewed other seismic analysis calculations of the containment structure (Ref. 3,4,5,6,7) to ensure that the basis of the SAM was justified. The interfaces between civil / structural and other disciplines were also reviewed to ensure that the documents transmitted involving SAM were fully understood and clear instructions were issued (Ref. 8).
The inspectors are satisfied that proper procedures were followed and clear instructions were issued to other disciplines regarding the use of SAM values.
This allegation is not substantiated.
2.
Trunnions The allegation stated that pipe supports utilizing trunnions were designed in unsymmetrical "imbalanced" configurations with the resulting localized stresses at the point of attachment to the pipe not being evaluated by the pipe stress analyst.
Additionally, the allegation stated that the localized stresses induced by the trunnions were not evaluated properly with regard to the close proximity of the trunnion to a stress riser (i.e. elbow, other welded attach-ment, etc.).
The systems specifically addressed in the allegation were main steam, feedwater and safety injection.
The IE inspector reviewed virtually every pipe support utilizing trunnions on main steam and feedwater, and a number on the safety injection system, which were in Bechtel's scope of design.
The pipe supports reviewed are listed in Reference 1 of this allegation.
In no pipe support reviewed was any example found that would lead to the " imbalance" situation referred to by the alleger.
In no pipe support reviewed was any example found of a trunnion "too close" to an elbow.
One support (V1-1305-56-H022 Rev. 4) was found in which a rectangular welded attachment was close enough to a valve to invalidate the method of analysis used to qualify the attachment.
However, this support (V1-1305-56-H022 Rev. 4) was on a non-safety related portion of piping and the staff informed Bechtel of this situation for their resolution.
This allegation is not substantiated.
3.
Flange Bending in Structural Steel Sections The allegation stated that Bechtel did not consider the effects of flange bending stresses on auxiliary support steel calculations for' main steam, main feedwater and safety injection systems.
In addition, the ef1*ects of multiple
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or " gang" supports were not evaluated in auxiliary steel calculations.
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above concerns were limited to the ASME Code Section III Class 2 and 3 applications.
The IE inspector reviewed selected pipe supports of the main steam, feedwater and safety injection systems in an attempt to identify those pipe supports utilizing stiffener plates. The stiffener plates.are used primarily to con-trol local flange bending and web buckling stresses in structural steel sections.
The pipe supports reviewed are identified in Reference 1 to this allegation.
F During this review, no instances were found in which local flange bending
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existed without the utilization of stiffener plates.
In cases involving the use of stiffener plates to control the local bending stresses, the inspector reviewed the calculations prepared by the Bechtel civil /
I structural group listed in Reference 2 to this allegation.
During this review t
no instances were found in which calculations failed to support the addition l
of stiffener plates.
All calculations were performed per Bechtel Instruction i
Memo #6 Revision 1 (Group 3) dated September 10, 1986 and titled "C/S Calcula-l tion Checklist." This checklist defines the procedure to follow to account for end reactions on existing building structural steel from pipe and rack i
type supports.
In addition, the checklist identified the V-SAMU organiza-tion as having the responsibility to evaluate local stresses in the existing _
l structural steel members in the presence of torsional loading created by pipe support attachments.
It should be noted here that this procedure was added to
' Desk Instruction #6 via Revision 8, dated May 29, 1985.
Bechtel's procedure to l
control and track the incorporation of each reaction imposed onto existing build-
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ing steel was reviewed by the inspector.
This comp' uter program " STRIVE" (Struc-tural Integrated Verification and Evaluation) is outlined in Instruction Memo #21
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(Group 3) dated April 30,-1985, and controls the tracking of all final loads i
from pipe supports onto existing building steel.
Included.in this load verifi-cation are the reactions of pipe supports and pipe racks at selected points of
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attachment to the building structure resulting from static and dynamic loads.
Support reactions are supplied to Bechtel C/S discipline by Bechtel Plant Design 3
and V-SAMU.
t The NRC inspector noted that no formal procedure existed to formally address 4
the calculation of local flange bending stresses; however, the calculations reviewed appeared to be consistent with accepted engineering practices.
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This allegation is not substantiated.
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l 4.
Building Settlement i
This allegation consists of two parts.
Part A states that the settlements calculated for the nuclear service cooling water (NSCW) tower did not consider the flooded condition.
Part B states that the NSCW tower settlement was revised i
from 3 inches to 1/2 inch in order to qualify piping systems running between the auxiliary building and the NSCW tower.
For Part A of this allegation the IE inspector reviewed the NSCW tower settle-ment calculation (Ref. 1,2) and found that the calculation was prepared in conjunction with the flooded condition.
The licensee has a monitoring system i
designed to record the settlements of safety-related structures including the l
NSCW tower. A recent report to the NRC (Ref. 3) indicates that the maximum j
recorded settlement of the NSCW tower is 3.6 inches which is still below the I
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Q calculated value of 4.7 inches.
The report also states that the settlements of the NSCW towers have stabilized in view of the incremental settlement during the last 12 months. This report is currently being reviewed by NRR.
For Part B of this allegation, the IE inspector reviewed civil / structural calculations concerning the justification of building settlements (Ref. 1,2),
field data regarding building settlement (Ref. 3), design interface information betn en civil / structural and piping analysis (Ref. 4), as well as the applicable portion of the pipe stress analysis of the NSCW system (Ref. 5).
Total building settlements for safety-related structures are listed in Vogtle FSAR Table 2.5.4-8.
l These total settlement values listed in the FSAR vary from 1 inch for small tank enclosures to 4.8 inches for deeply embedded concrete structures, such as the NSCW towers.
Building settlements used for pipe stress analysis are differen-tial settlements and are listed in Reference 4.
Furthermore, differential i
building settlements occurring before installation of piping are not considered in the pipe stress analyses.
Only differential building settlements occurring 1
subsequent to pipe installation are needed to perform pipe stress analysis.
It is appropriate for the architect engineer to use the differential building settlement as listed in Reference 4, rather than the total building settlements i
j as listed in the FSAR.
Additionally, Bechtel did not permanently support piping 1
j running between adjacent buildings until building settlement had stabilized.
After reviewing the civil / structural calculations for building settlement (References 1,2), the associated field report on settlement (Reference 3),
discipline design interface data (Reference 4), and the pipe stress analysis (Reference 5), the IE inspectors concluded that building settlement regarding NSCW piping was properly considered.
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Both parts of this allegation are not substantiated.
j 5.
Thermal Expansion Coefficients i
i The allegation stated that Bechtel had inappropriately used thermal expansion coefficients from a later ASME Code (i.e. 1980) than that which the piping design for Class 2 and 3 piping was based (i.e. ASME 1974 Edition through Summer i
75 Addenda). Use of these later Code values for thermal expansion coefficients l
would result in a less conservative piping analysis for the thermal loading l
evaluation.
This allegation was raised for Palo Verde as well as Vogtle.
The IE inspector confirmed that for Vogtle, the piping systems for Class 2 and 3 piping were designed to ASME III 1974 Edition through Summer 75 Addenda with thermal expansion coefficients taken from the Summer 79 Addenda to the 1977 ASME Code.
Even though the later Code has values of thermal expansion coefficients i
significantly less than those of 74 Edition, the inspection team confirmed l
with NRR that this is an acceptable approach.
Thermal expansion coefficients l
are a material property which has been more rigorously defined by the later Code and are not dependent on compliance with another section of the Code for their use.
Formal acceptance of the later Code's thermal expansion coefficients will j
j be documented via FSAR changes, since both Vogtle's and Palo Verde's FSAR do not identify the exceptions to the Code Edition and Addenda in effect for t
the piping systems.
NRR was informed of this issue via memorandum from B. K.
Grimes to B. J. Youngblood dated November 13, 1986.
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This allegation is not substantiated.
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6.
Main Steam Bypass Pipe Stress Modeling The allegation stated that Bechtel modeled the main steam bypass line incorrectly, in that not all pipe runs and branches in the computer stress analysis model ended in anchors, and that the complete bypass line was not included; further, that each of these actions are a violation of the applicable Code under which the analysis was done.
The IE inspector verified that the main steam bypass system is not in Bechtel's scope of design and had been in Southern Company Services scope of design since the early 1970s, which precedes the alleger's tenure at Bechtel.
- Hence, Bechtel could not have modeled the main steam bypass analysis incorrectly since it was not in their scope of work.
However, the IE inspector reviewed the associated technical issue as it applies to the main steam system, which is within Bechtel's scope of design, and evaluated how the main steam bypass is decoupled from the main steam system.
Upon reviewing the main steam piping analysis for outside of containment (i.e. X4CP-7073T), the IE inspector deter-mined that Bechtel had modeled only a portion of the bypass line. Approximately 120 feet of bypass piping was modeled and terminated at a 3 way pipe support restraint..This is not a Code violation since the Code allows the use of alternative methods of design.
Specifically, ASME Section III paragraph NC/ND-3673 for Class 2 and 3 piping only requires that the effects of branch lines be considered and the Code is silent on how this is to be accomplished.
Additionally, the team verified that Bechtel had criteria which controlled what size branch lines are included in the run piping analysis by reviewing Design Criteria DC-1017 Rev. 5.
In conclusion, the team verified that the Bechtel model did not terminate the branch run of the main steam line at an anchor, but instead used engineering judgment, terminated it at a 3-way restraint which was significantly distant from its connection to the main steam line.
The modeling of the main steam line to consider the effect of the main steam bypass was viewed by the inspection team to be acceptable, and consistent with industry practice.
This allegation is not substantiated.
7.
Seismic Anchor Movement - ME101 Input The allegation stated that the piping stress analysts were inputting the wrong value of seismic anchor movement (SAM) into Bechtel computer program for pipe stress analysis (ME101).
Specifically, the allegation stated that the piping analysts were inappropriately dividing the value of SAM from civil engineering by a factor of two in order to account for the ME101 program multiplying SAM by two. According to the alleger, the SAM from civil engineering represents 1/2 of the amplitude (peak to peak), therefore it should be input directly in ME101 without dividing by two.
The IE inspector reviewed Bechtel's method of controlling the piping analysts' l
input into ME101 regarding SAM.
The IE inspector reviewed Design Criteria
- 0C-1017, Rev. 5 which refers the analyst to Paragraph 3.6.5 of Bechtel's l
" Seismic-Interdiscipline Design Criteria" #DC-1005, Revision 1, specifically Tables 3 through 6.
These tables list the seismic relative displacements repre-senting 1/2 the total displacement range (i.e. from datum to peak).
This is obvious due to the footnote at the bottom of the table which indicates that the ME101 computer program assumes these values to be either positive (+) or negative (-).
Therefore, the IE inspector determined that these SAM values should be input directly into ME101 without any adjustment.
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Additionally, the IE inspector reviewed several piping analysis problems to see what values of SAM were used.
A review of Feedwater (inside containment)
Calculation #X4CP-7117; Auxiliary Feedwater Calculation #X4CP-7072EI; and the Nuclear Service Cooling Water System (NSCWS) Calculation #X4CP-7158.
All revealed proper treatment of SAMs, i.e., values were input directly from DC-1005 into ME101.
Conversations with Bechtel personnel assigned to the Vogtle project indicated that during 1983 discussions between Bechtel Civil Engineering and Pipe Stress, helped resolve any confusion over the proper treatment of SAMs.
Based on the above, this allegation is not substantiated.
8.
Snubbers The allegation stated that Bechtel did not consider that " oversized" snubbers on certain lines could cause these lines to become overstressed under thermal loading conditions.
This situation could occur because the snubber (s) had such a large capacity that the force generated by thermal expansion of the pipe would not be great enough to overcome the inertial resistance (1.0% of rated load) of the snubber to movement.
Thus the snubber would act as a rigid restraint causing the pipe to be overstressed.
The alleger was concerned that this situation would occur because of a change from full LOCA to " leak-before-break" criteria for the Class 1 piping ano specifically identified the reactor coolant system safety injection system and the chemical and volume control system as examples.
The IE inspector determined that this snubber issue is independent of the faulted load case (i.e. change from LOCA to leak-before-break) as alleged, but is dependent solely on the thermal loading condition.
In lieu of the systems identified in the allegation, the IE inspector decided to review the feedwater system between the steam generator and the containment penetration because:
(1) all of the feedwater system inside containment is within the analysts' scope of Bechtel, whereas portions of safety injection and chemical volume and control are within the scope of Westinghouse; (2) the feedwater system governing load is water hammer, the control of which requires several large capacity snubbers; (3) thermal loads are low compared to lines connected to the reactor coolant system; (4) snubbers are not used on reactor coolant system piping as stated by the alleger.
From the review of the feedwater line the IE inspector determined that the potential existed for the alleged situation.
Bechtel had not done analyses to ascertain the situation, but had either relied on engineering judgment, or had not considered the possibility at all.
When informed of the concern, Bechtel responded by rerunning thermal stress analyses for four piping systems, having included the appropriate snubber drag i
forces.
The lines analyzed were main feedwater from the containment penetra-tion to the No. 3 steam generator, and three auxiliary feedwater lines from l
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0 their respective containment penetrations to their respective steam generators.
No main steam line was analyzed because Bechtel was of the opinion that, the relative pipe size to snubber size is much larger than for main feedwater therefore, drag forces on main feedwater would produce higher piping stresses for the same size snubbers.
Bechtel indicated that the increase in stress, at previous high thermal stress points due to reanalysis using snubber drag forces, was no more than three percent for the four systems analyzed.
Bechtel also pointed out that in general, except for main steam piping, main feedwater and auxiliary feedwater other systems are not subjected to large hydraulic transient loads and therefore do not contain snubbers with capacities greater than those required for seismic design.
Furthermore, the critical high energy piping system thermal movements are evaluated during hot functional tests prior to fuel loading and during power ascension thermal growth and vibrating testing to ensure that the piping thermal movements are within established acceptance criteria.
Evaluations or physical modifications are performed if the criteria is not met.
Even though Bechtel had not analyzed the effect of snubber drag forces, later analysis indicates that the effect is negligible thus verifying their engineering judgment.
Therefore this allegation is not substantiated.
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REFERENCES FOR ALLEGATION 1.
1.-
Civil / structural memorandum BB-48087, "NSCW System; Revised Building Settlement & Seismic Anchor Movement".
D.L. Houghton to D. Capito, March 3, 1983.'
2.
Bechtel Calculation # X2CC4.10 "NSCW Valve House BSAP Computer Model" Rev. 1, 2-7-86 and BSAP Computer Run #AEVEPGB, Log #X2NSFF08, 11-14-84 Entitled "0BE Responses Spectrum Analysis of Valve House".
3.
Bechtel Calculation # X2CS10.1, " Design Criteria - Seismic", Rev. 1, 12-13-84.
4.
Bechtel Calculation # X2CS10.2, " Soil Properties", Rev. 1, 10-24-84.
5.
Bechtel Calculation # X2CS10.6.1, " Containment - Fixed Base Model",
i Rev. 2, 7-30-86.
6.
Bechtel Calculation # X2CS10.8.2, " FLUSH Model 2 Analysis", Rev. O, 7-20-79.
7.
Bechtel Calculation # X2CS10.9.1.8, " Final Response Spectra", Rev.1, 4
7-30-86.
8.
Design Criteria DC-1005, " Seismic - Interdisciplines", Rev. 1, 4-4-83.
9.
Bechtel Calculation # X2CS10.11.1, " Misc. - Relative Displacement Between Structures", Rev. 1, 10-20-86.
For Allegation 2.
1.
Pipe Support Drawings Reviewed:
V1-1301-001-H001 Rev. 7 V1-1301-002-H002 Rev. 3 1 Rev. 5 V1-1301-003-H002 Rev. 4 1 Rev. 5 4
V1-1301-004-H002 Rev. 4 3 Rev. 2
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V1-1301-006-H001 Rev. 4 (1)
l;k' VI-1301-007-H001 Rev. 2 a
2 Rev. 5 3 Rev. 2 4 Rev. 3 a
6 Rev. 4 7 Rev. 3 a
9 Rev. 3 3
10 Rev. 3 11 Rev. 4 14 Rev. 4 16 Rev. 4 22 Rev. 3 a
23 Rev. 5
'24 Rev. 4 25 Rev. 4 27 Rev. 4 30 Rev. 4 32 Rev. 3 33 Rev. 3 35 Rev. 3 55 Rev. 2 a
501 Rev. 9 V1-1301-008-H001 Rev. 3 2 Rev. 3 3 Rev. 3 4 Rev. 2 a
5 Rev. 3 6 Rev.<4 7 Rev. 5 8 Rev. 6 9 Rev. 5 a
10 Rev. 8 11 Rev. 6 12 Rev. 4 l
13 Rev. 3 14 Rev. 4 15 Rev. 5 16 Rev. 5 17 Rev. 4 i
18 Rev. 4 i
a 19 Rev. 4 21 Pev. 4 23 Rev. 5 27 Rev. 3 3
32 Rev. 3 39 Rev. 4 l
40 Rev. 6 i
41 Rev. 5 l
42 Rev. 5 j
43 Rev. 5 (2) i
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V1-1301-008-H045 Rev. 4 46 Rev. 3 47 Rev. 4 48'Rev. 3 49 Rev. 3 52 Rev. 3 53 Rev. 3 55 Rev. 2 56 Rev. 3 501 Rev. 8 502 Rev. 5 V1-1301-552-H001 Rev. 2 554-H001 Rev. 2 557-H001 Rev. 2 559-H001 Rev. 2 V1-1305-055-H001 Rev. 3 2 Rev. 5 7 Rev. 3 8 Rev. 4 9 Rev. 4 V1-1305-056-H002 Rev. 4 3 Rev. 3 5 Rev. 4 6 Rev. 3 7 Rev. 3 8 Rev. 3 9 Rev. 3 a
10 Rev. 2 11 Rev. 3 12 Rev. 3 13 Rev. 3 14 Rev. 4 15 Rev. 3 16 Rev. 3 17 Rev. 2 18 Rev. 2 22 Rev. 4 24 Rev. 3 25 Rev. 3 26 Rev. 2 27 Rev. 2 28 Rev. 3 29 Rev. 3 a
30 Rev. 2 V1-1305-057-H002 Rev. 5 4 Rev. 2 5 Rev. 3 8 Rev. 3
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V1-1305-058-H001 Rev. 8 V1-1305-060-H001 Rev. 6 V1-1305-061-H002 Rev. 4 3 Rev. 4 4 Rev. 4 7 Rev. 3 8 Rev. 4 9 Rev. 4 V1-1305-062-H001 Rev. 3 2 Rev. 6 V1-1305-064-H002 Rev. 6 a
9 Rev. 2 V1-1204-057-H007 Rev. 2 8 Rev. 3 13 Rev. 4 18 Rev. 4 26 Rev. 4 32 Rev. 5 For Allegation 3.
1.
V1-1301-148-H019 Rev. 2 20 Rev.'2 21 Rev. 2 V1-1301-212-H001 Rev. 2 2 Rev. 4 3 Rev. 4 4 Rev. 3 5 Rev. 4 6 Rev. 4 7 Rev. 3 i
8 Rev. 2 9 Rev. 4 10 Rev. 3 i-V1-1301-106-H001 Rev. 3 L
2 Rev. 3
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3 Rev. 3 4 Rev. 3 l
5 Rev. 3 6 Rev. 3 7 Rev. 2 V1-1301-002-H001 Rev. 5 2 Rev. 3 t
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V1-1305-058-H002 Rev. 4 3 Rev. 5 4 Rev. 3-5 Rev. 3 6 Rev. 3 a
7 Rev. 3 8 Rev. 2 V1-1305-060-H002 Rev. 4 3 Rev. 3 4 Rev. 3
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5 Rev. 4 6 Rev. 3 7 Rev. 3 V1-1305-061-H012 Rev. 1 a
013 Rev. 5 V1-1305-062-H004 Rev. 3 5 Rev. 3 6 Rev. 4 7 Rev. 4 8 Rev. 3 V1-1305-064-H003 Rev. 4 4 Rev. 4 5 Rev. 4 6 Rev. 3 7 Rev. 5 8 Rev 5 V1-1204-057-H023 Rev. 3 24 Rev. 4 25 Rev. 3 27 Rev. 3 28 Rev. 4 29 Rev. 3 30 Rev. 3 31 Rev. 3 33 Rev. 3 34 Rev. 3 35 Rev. 4 37 Rev. 6 V1-1204-057-H001 Rev. 5 2 Rev. 5 3 Rev. 4 4 Rev. 3 9 Rev. 2 10 Rev. 3 a
11 Rev. 4 12 Rev. 2 14 Rev. 3 (5)
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V1-1305-057-H015 Rev. 4 16 Rev. 4 17 Rev 4'
19 Rev. 4 20 Rev. 3 21 Rev 5 22 Rev. 4 36 Rev. 4 2.
V1-2301-071-H092 V1-1210-132-H008 V1-1301-148-H020 Rev. 2 V1-1301-212-H006 Rev. 4 V1-1301-212-H008 Rev. 2 For Allegation 4 1.
Bechtel Calculation # X2CF-S-106, " Settlement Analysis", Rev. 1, 1-23-86.
2.
Bechtel Calculation # X2CC3.25, " Settlement Calculations", Rev. O, 11-8-84.
3.
Vogtle Electric Generating Plant (VEGP) - Report on Settlement, August, 1986.
4.
Design Criteria DC-1017, " Pipe Stress and Pipe Supports Analysis i'
Criteria", Rev. 5, 3-19-86.
5.
Bechtel Calculation # X4CP-7158, NSCW Piping Analysis i
References for Allegations 5, 6, 7 and 8 are in text of assessment.
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