ML20127M956
| ML20127M956 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 05/17/1985 |
| From: | Ferg D CYGNA ENERGY SERVICES |
| To: | Colbert W, Youngblood B DETROIT EDISON CO., Office of Nuclear Reactor Regulation |
| References | |
| 83021.060, NUDOCS 8505230396 | |
| Download: ML20127M956 (40) | |
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Uv s 150 North Wacker Drme, Chicago, IL 60606 83021.060 312<236-5701 May 17, 1985 Director of Nuclear Reactor Regulation Attention: Mr. B.J. 'Youngblood, Chief Licensing Branch No. 1 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Mr. William F. Colbert, General Supervisor Nuclear Safety and Plant Engineering (342 NOC)
The Detroit Edison Company Enrico Fermi-2 Nuclear Operations Center 64 North Dixie Highway Newport, MI 48166
Subject:
NRC Review Questions and Comments Independent Design Verification Program Detroit Edison - Enrico Fermi Unit 2 Docket #50-341
Reference:
NRC Letter from B.J. Youngblood to Wayne J. Jens of Detroit Edison and L.L. Kamerzell of Cygna dated April 30, 1985
Dear Sirs:
Enclosed are Cygna's responses to the NRC questions contained in Enclosures 1, 2, and 3 to the referenced letter. Our response has been prepared in a format which will allow it to be inserted in the Final Report for the Detroit Edison Independent Design Verification Program on Fermi-2.
Please contact me if you require further assistance or clarification on this matter.
Very truly yours.
A cf. 8505230396 850517 PDR P
ADOCK 05000341 David A. Ferg PDR Project Manager DAF:od Enclosures (40 copies for NRC)
(20 copies for DECO) cc: M.D. Lynch (NRC, NRR-DOL) with Enclosure (2 copies)
J.G. Keppler (NRC IE, Region III) with Enclosure (2 copies) D 8 -
( i go 0.K. Earle (Deco) w/o Enclosure Boston Chicago Philadelphia San Francisco
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_I,N,STRUCTION_ S,HE,ET
- !-To insert-the enclosed printed material:into CYGNA's Independent Design ,
r ; Verification Progam Final Report'on Detroit' Edison's-Fermi-2 plant (Docket
- 50-341), perform-the.following steps:
'a) Remove page-7.7-65 and insert new pages 7.7-65, 7.7-65a and 7.7-65b; L
b) Remove pages 8.2-5 and.8.2-6 and insert attached pages 8.2-5, 8.2-Sa, 8.2-6 and 8.2-6a;
. c) Remove page 8.2-9 and insert new pages 8.2-9, 8.2-9a /~ -
and 8.2-9b; d) Remove pages 8.2-38 and 8.2-39 and insert new pages 8.2-38, 8.2-38a and 8.2-39; e) ' Insert new pages 8.2-41a and 8.2-41b behind page 8.2-41; f) Insert new page 8.2-46a behind page 8.2-46; g) Insert new page 8.2-63a behind page 8.2-63; h). Insert new page 8.2-67a and the attached nineteen (19) unnumbered pages from Dames & Moore behind page 8.2-67; Insert new page 8.2-72a behind page 8.2-72.
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I H e A
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Observation p(*h M TJ Record Review p
V iminmunmumunnn Attachment A Checklist No. Revision No, y PI Ol Sheet of 4
Observation No. PI-01-11 7 Yes No Valid Observation y Potential Finding y (PFR No. N/A )
Closed y Comments Further Cygna review has indicated that the postulated impingement and surge loads
, on the RHR system should not be considered since the source of the loads is the broken line to which the RHR line is attached. Detroit Edison will revise the design specification (3071-503) to reflect this. Cygna agrees with this evaluation, since the FSAR, Paragraph 3.6.5.1.1, states " Piping within the broken loop- shall no longer be considered part of the RCPB (reactor coolant pressure boundary)."
In evaluating faulted conditions in general, Reference 3.2, Article 4.5 states D "
...LOCA does not create temperature or pressure surges in the piping systems of any
( significance and therefore it is not evaluated for this event." This reference also states in Article 4.7 that " pipe stress due to Annulus Pressurization is not required to be included in the Code analysis and stress report." j Based on this information, this observation does not warrant any further investigation.
Supplemental review has revealed that a separate report from the ASME Design Report was generated to assess the impact of annulus pressurization on piping and structures. This approach had been agreed to with the NRC because annulus f pressurization was not in the original design basis for Fermi 2. Nevertheless, the piping supports are designed to accommodate the additional loads predicted by this analysis. The probable cause of this observation was the failure to update the DECO design specification to reflect the design basis commitments as stated above.
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o Approved By oject Mana er Date M M/
Detroit Edison Company; 8302$
7*7-65 Independent Design Verification' Program l
==u5 Observation M tOfd Record Review '
mmmmmmmimimi 7 Attachment A U
Checklist No. PI-01 R oision No. g Observation No. PI-01-11 Sheet p of 4 Yes No
~ Valid Observation X Potential Finding X (PFR No. N/A )
Closed X Comments Subsequent. information consisting of a) NRC letter to Detroit Edison, " Preliminary Evaluation of the IDVP Performed by Cygna Energy Services for the Fermi-2 Facility" dated March 27, 1984; b) meeting notes between the NRC, Detroit Edison and Cygna in Bethesda, Maryland on May 11, 1984; and c) Detroit Edison letter, EF2-72252 to the NRC dated September 27, 1984, indicates that the basis for resolving this observation was inadequate. There was a licensing commitment to the NRC by Detroit Edison to eva-
.luate the Fermi-2 design for structural integrity under combined annulus pressuriza-O tion and Design Basis Earthquake (DBE) loadings. (Detroit Edison Amendment 24 dated b June, 1979, in response to NRC Question 110.11 in Appendix E.5 of the Fermi-2 FSAR).
As indicated in Section 8.2.2.2, Cygna proceeded with a review of the Detroit Edison pipe stress evaluation for faulted loads to determine if annulus pressurization was properly considered for the in-containment RHR shutdown suction cooling element. In the course of this review, Cygna reported sufficient differences between the piping geometry analyzed for A/P loads and the as-built configuration which precluded our establishing structural integrity for the RHR piping under faulted load conditions.
Prior to the May 11, 1984 meeting in Bethesda, Md., Cygna's Project Team discussed elevating the observation to a Potential Finding Report in order to investigate further the significance of the differences. However, following the May meeting, the NRC, Detroit Edison and Cygna agreed that Cygna did not have enough information to
, proceed wit' our review and Detroit Edison committed to re-analyze the portion of the RHR' system .n question. The results of this re-analysis and a comparison of the as-built configuration of other large bore (NPS > 4") reactor coolant pressure boundary piping systems with the original annulus pressurization analysis input were presented to the NRC with' Detroit Edison's September 27, 1984 letter, EF2-72252. With minor weld size modifications, Detroit Edison was able to conclude that structural integrity would be maintained inthe Fermi-2 as-built configuration for faulted loads, including annulus pressurization. The April 30, 1985 letter from the NRC to Detroit Edison and Cygna indicates this issue has now been resolved without additional in-depth reviews by Cygna.
O Approved By Project Manager Date 11-11-83 Detroit Edison Company; 83021 7.7 65a Independent Design verification Program
M Observation M i'fd Record Review nmmumummmnmi Attachment A G
Checklist No. PI-01 Revision No. 1 Observation No. Sheet of PI-01-11 p 4 Yes No Valid Observation X Potential Finding X (PFR No. N/A )
Closed X Comments (Cont.d)
In response to the recent April 30, 1985 request from the NRC, Cygna evaluated the information contained in the Detroit Edison letter dated September 27, 1984 and con-curs with DECO's approach'and conclusions. Observation PI-01-ll can therefore be (O
L1 closed on the basis of this information and the fact that the NRC and Detroit Edison have bi-laterally resolved the issue of A/P loads with the added assurances that Cygna considers this resolution to be acceptable.
b,' Approved By Project Manager Date 11-11-83 Detroit Edison Comoany; 83021 ~
Independent Design Verification Program 7.7- 65 b
All of the observations assigned to Category A except Observations PI-01-11 and PS-01-03 received an expanded review by Cygna. Section 7.3, Exhibit 7-3 identifies to what extent the IDVP review was expanded. Observations
'PI-01-11 and PS-01-03 concerned the analysis of annulus pressurization (A/P) loads as a design. requirement for Fermi-2. Since A/P loads were not origi-nally considered by Cygna to be an actual design requirement on Fermi-2, the review was not expanded. Refer to Section 8.2.2 for further discussions con-cerning the A/P load issue. Observation EE-01-03 was by the nature of the observation expanded to review all safety-related loads which are sequenced on the diesel generator under accident conditions to ensure none would reduce the diesel generator voltage below 85%.
All of the observations assigned to Category B except Observations DC-01-05, O CC-01-12, DC-02-06, DC-02-07, DC-02-10, PS-01-04, PS-02-03 and ST-01-01 O
required an expanded review by Cygna. Again, Section 7.3, Exhibit 7-3 descri-bes the scope expansion conducted by Cygna to resolve these Category B obser-vations. To resolve Observation DC-01-05, all key design documents were reviewed by Cygna to ensure they had the proper QA level designation. A review of the personnel who acted as lead auaitors since 1978 was performed to resolve Observation DC-01-12. Sargent & Lundy's internal audit program was reviewed in depth to determine that there was no design impact on Fermi-2 due to Observation DC-02-06. Observations 00-02-07 and DC-02-10 were resolved by requiring DECO to perform a complete as-built analysis for all flued-head anchor structures and Sargent & Lundy to review all Fermi-2 pipe stress O Detroit Edison Company Fermi 2 Independent Design Verification niunui!ionnhinniin Final Report, TR-83021-1, Revision 1 8.2-5
reports and request the field to verify that as-built pipe supports are recon-ciled with the stress report results. Observatior. PS-01-04 concerned the com-parison of piping design loads for Operational Basis Earthquake (OBE) and Safe-Shutdown Earthquake (SSE) ground motion accelerations. As such, the observation did not require an expanded review because it inherently covers the seismic characteristics of the entire fermi-2 site. Observation PS-02-03 concerned a check to ensure seismic movements were within the working range of-spring hangers. Again, since the seismic movements were small (< 1/10") in both the RHR Cooling and RHR Service Water Systems, no expansion in review scope was necessary since the two systems are representative of other plant systems. However, Cygna requested Detroit Edison to review the remaining spring hangers to verify adequacy. Finally, Observation ST-01-01 ' involved the use of design summary sheets to incorporate the structural design criteria into each structural calculation on the Fermi-2 project prior to 1981. Even though ST-01-01 was generic to all of S&L structural activities, it had no generic implic' ' ions to the design process on Fermi-2 (refer to page 7.7-104 for further discussion).
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- l. l0f the twenty-six (26) observations assigned to Category C, eleven (11) required an expanded review to determine to what extent, if any, each obser-
-vation affected the Fermi-2 design. The scope expansions for Observations lPI-01-03,PI-01-07,PI-01-08,_PI-01-09,PI-03-05,PI-03-06,EQ-01-03, EQ-01-04, PS-00-04, ST-01-24 and ST-01-33 are described in Section 7.3, l l - Exhibit 7-3.
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Cyg'na determined that it was standard pract'ce for GE to use a default value for stress indices of 1.0 on small branch connections. Consequently
- c. Observation PI-01-06 required a generic resolt. tion involving GE pipe stress analysis techniques. For Observation PI-03-02, Cygna review all flued-heads to verify the omitted containment pressure stresses were negligible. Since
=., thermal movements are small for both the RHR Cooling and RHR Service Water elements and since both systems were representative of other high temperature
- Fermi-2 systems,_an. expanded' review for Observation PS-00-02 was not justified. In Observation PS-01-01, Cygna expanded the review until it was determined that GE had Yerified the shear lug design in the Class 1 pipe stress analyses. To resolve Observation DC-02-02, Cygna examined Sargent &
Lundy's method for specifying the use of. computer programs on the Fermi-2 pro-ject and checked this method to ensure the correct and proper programs were utilized in the design process. Review results were able to also demonstrate that Sargent & Lundy's method for calculating allowable loads on embedment plate stud bolts was sufficiently conservative to resolve Observation
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O Detroit Ediso.. Company
[ , , Fermi 2 Independent Design Verification
, Final Report TR-83021-1, Revision 1 8.2-6
ST-01-26. Observations ST-01-03, ST-01-05, ST-01-06, ST-01-09, ST-01-12, ST-01-13, ST-01-15, ST-01-16, and ST-01-19 are in the structural discipline and are unique only to the RHR Complex. Additional information associated with the resolution of Observations ST-01-03, ST-01-06, ST-01-09, ST-01-13, and ST-01-16 are provided in Cygna's responses to NRC Enclosure 3 Questions i
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(refer to Section 8.2.3) l O
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l V Detroit Edison Company Fermi 2 Independent Design Verification
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0bservation/ Reference PFR No. _ ____________
Description ._
Page _
DC-01-06 Root Cause: An incomplete review of the subject specification since the revision did not 7.7-31 have a P.E. certification. This was a random occurrence and appeared to be simply an over-sight on behalf of Detroit Edison Engineering.
Extent: No generic implications DC-01-07 Root Cause: Not applicable since observation was invalid. 7.7-32 DC-01-08 Root Cause: A lack of documented evidence that the Decroit Edison QA program with respect to 7.7-33 PFR-01 internal audits was being effectively implemented.
Significance: Without adequate assurance that the design control program was being effec-tively implemented, the quality and integrity of the Fermi-2 design could have been called into question. A comprehensive review indicated all elements of the design control program were evaluated during the course of the project. However, Cygna performed a comprehensive l review of the design control program elements including design input, design analysis, drawing control, procurement control, interface control, design verification, document control, design ,
changes, corrective action and audits. Cygna found sufficient assurances that all key aspects I of the design control program were evaluated during the Fermi-2 project duration and no poten-tial impact on safety exists.
Extent: Generic implications for the entire plant to the extent the design process could have been of questionable quality and a lax internal audit system might never have identified the extent of any weaknesses. However, Cygna determined in the course of their review that the internal audit and review activities were sufficiently extensive and thorough to conclude the Fermi-2 design process was adequate and performed with the requisite quality and integrity.
In addition, a DECO management directive was issued fo'r a comprehensive review of all the less-formal surveillance QA reviews and special audit programs to assure any corrective action items were formally documented, trackea and closed.
8.2-9 N
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~ Observation /- .
. Reference
-PFR No. Description ___
Page DC-01-09 Root Cause: A lack of management ~ attention and follow-up in reviewing audit results and 7.7-34; PFR taking appropriate action.to correct the deficiencies.-
Significance: The Fermi-2 design could have been adversely or unnecessarily impacted without timely and proper corrective action on design control audit-findings. However, a more formal, systematic. system for tracking audit findings and QA remedial / corrective action items has been initiated. An "Open Item Status Report - Engineering Quality Assurance" has been per-iodically issued to responsible Project Engineering personnel down to and including the Group Supervising' Engineers, which identifies individuals responsible for resolution along with expected dates of completion. Positive actions have been. implemented and in progress to resolve outstanding open audit findings and surveillence items and therefore no impact on plant safety.was found.
Extent: Generic implications for the entire plant to the extent the design process could have been of questionable quality due to a continued lack of corrective action on internal audit and surveillance' findings. However, Cygna'found that since Detroit Edison had imple-mented the tracking program for audit findings, significant progress had been made in resol .
ving open items. The monthly meetings with the President of Detroit Edison involving corpor-ate and Fermi-2 QA management provided for executive-level discussions on quality assurance matters and an appropriate forum to follow progress in closing the remaining open items.
With continued follow-up and management attention, Cygna expects the remaining open' items would be brought to a satisfactory, timely resolution without any impact on the design and safety of the Fermi-2 plant.
DC-01-10 Root Cause: A lack of documented evidence that the Detroit Edison QA program with respect to 7.7-35 PFR-03 contractor and vendor audits was being effectively implemented. Also, an audit schedule of A/E's which appeared too infrecuently. for continuous monitoring of supplier QA program imple-mentation.
Significance: Basically, it is Detroit Edison's responsibility to perform frequent audits of architect / engineers and engineering. consultants. They should maintain adequate documentation 8.2-9a I
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D %/ Xf Observation / Reference PFR No. Description Paae DC-01-10 of checklists and audit findings to provide added assurances that design control programs are PFR-03 being effectively maintained and implemented. With respect to architect / engineers,.however, (cont'd) the combined audit activities by Detroit Edison,.Sargent & Lundy'and Stone & Webster were determined to provide sufficient assurances that the design control programs at S&L and S&W were effectively and adequately implemented. An expanded review by Cygna of Fermi-2 engineering service suppliers including NUS, Nutect, Teledyne, Bechtel, Parsons, Griffels Associates, Hopper
& Associates and Multiple Dynamics Corporation, also disclosed that DEC0 performed the necessary audits to assure an effective implementation of each supplier design control program. Conse-quently, it was determined during the course of the review that this finding has an insignifi-cant affect on the overall Fermi-2 design and design control process.
Extent: Generic implications to the extent the design information and design control process from A/E organizations to Fermi-2 could have been of questionable quality and an insuf-ficient, infrequent vendor audit system might not have identified a weakness. Further review again confirmed that, in addition to scheduled Detroit Edison audits of Stone & Webster and Sargent & Lundy, the Detroit Edison QA organization acted as an observer of internal audits conducted by the architect / engineer QA departments. This provided Detroit Edison QA with a first-hand assessment of the degree of compliance for the audited activity against the A/E's program commitments. Additionally, Detroit Edison QA routinely reviewed A/E and engineering consultant internal audit reports. Cygna was again able to confirm that this finding had no impact on the safety of the Fermi-2 plant.
8.2-9b
l EXHIBIT 8.2.1-2 A ROOT CAUSE CLASSIFICATIONS U
Category Observation Comments A DC-01-01 See Section 7.3, Exhibit 7-3 l PI-01-11 Annulus pressurization piping loads PI-02-02 See Section 7.3, Exhibit 7-3 PI-03-04 See Section 7.3, Exhibit 7-3 PS-00-01 See Section 7.3, Exhibit 7-3 PS-01-03 Annulus pressuriazation support loads PS-03-01 See Section 7.3, Exhibit 7-3 PS-03-02 See Section 7.3, Exhibit 7-3 ST-01-02 See Section 7.3, Exhibit 7-3 EE-01-03 FSAR requirement on minimum motor starting voltage B DC-01-05 QA level designations DC-01-08 See Section 7.3, Exhibit 7-3 See Section 7.3, Exhibit 7-3 DC-01-09 DC-01-10 See Section 7.3, Exhibit 7-3 DC-01-12 Lead auditor qualifications DC-02-06 SRL internal audit files DC-02-07 Field design change requests DC-02-10 As-built field verification
(, PI-01-12 See Section 7.3, Exhibit 7-3 PI-02-03 See Section 7.3, Exhibit 7-3 PI-03-01 See Section 7.3, Exhibit 7-3 PS-01-04 Design specification revision required PS-02-03 Spring hanger seismic movements ST-01-01 S & L structural design criteria ST-01-30 See Section 7.3, Exhibit 7-3 ST-01-31 Concrete voids and exposed rebar C DC-01-03 See Section 7.3, Exhibit 7-3 DC-02-02 Computer program user requirements PI-01-03 See Section 7.3, Exhibit 7-3 PI-01-06 Branch connection stress PI-01-07 See Section 7.3, Exhibit 7-3 PI-01-08 See Section 7.3, Exhibit 7-3 PI-01-09 See Section 7.3, Exhibit 7-3 PI-03-02 Flued-head load cases indices PI-03-05 See Section 7.3, Exhibit 7-3 PI-03-06 See Section 7.3, Exhibit 7-3 EQ-01-03 See Section 7.3, Exhibit 7-3 EQ-01-04 See Section 7.3, Exhibit 7-3 PS-00-02 RHR pipng thermal movements PS-00-04 See Section 7.3, Exhibit 7-3 PS-01-01 Shear lugs for Class I piping b
U Detroit Edison Company
' Fermi 2 Independent Design Verification
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1 EXHIBIT 8.2.1-2 (cont'd) l
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ROOT CAUSE CLASSIFICATIONS Category Observation Comments C PS-02-04 Use of OBE vs. DBE loads ]
(cont'd) ST-01-03 RHR Complex design soil loading ST-01-05 Cooling tower frame analysis model !
ST-01-06 Basement reinforcing steel placement ST-01-09 Foundation wall rebar placement ST-01-12 Missing foundation walls loads ST-01-13 Reinforcing steel in beams G
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EXHIBIT 8.2.1-2 (cont'd)
ROOT CAUSE CLASSIFICATIONS O
(,' Category Observation Comments C ST-01-15 Shear wall overturning moments (cont'd) ST-01-16 Foundation wall design moments ST-01-19 Reservoir water effects ST-01-24 See Section 7.3, Exhibit 7-3 ST-01-26 Stud allowable load calculations ST-01-33 See Section 7.3, Exhibit 7-3 D PI-01-01 Long vs short radius elbows PI-01-02 Orientation of restraints 5810 & G16 PI-01-10 Shear lug input load error PI-02-01 Branch intensification factors PI-02-04 Restraint G01 geometry PI-02-05 Long vs short radius elbows PI-02-06 Lubrite plates in stanchions PS-01-05 Weld size error PS-02-02 Penetration sleeve gaps PS-02-05 Hanger E11-2189-007 internal brace PS-02-06 See Section 7.3, Exhibit 7-3 ST-01-10 Cooling tower slab load definition ST-01-28 Inconsistent section ST-01-32 Cantilevered slab loading EE-01-02 Conduit size drawing discrepency
( E DC-01-06 Missing PE certification DC-01-11 RHR Mechanical Design Document update DC-02-01 Seismic analysis report references DC-02-03 See Section 7.3, Exhibit 7-3 DC-02-04 S & L design review schedule DC-02-05 S & L pipe support design calculations DC-02-09 See Section 7.3, Exhibit 7-3 DC-03-01 Responsible engineer's signature DC-03-02 Receipt acknowledgement of drawings DC-03-03 Seismic report ccmment resolution DC-03-04 Filing of dispositioned DCN's PI-01-04 Snubber suppporting frame stiffness PI-01-05 Incorrect valve body weights EQ-01-02 Valve axial cyclic stresses PS-02-01 Support E11-2184-G01 gap size ST-01-04 RHR Complex thermal gradients ST-01-07 Cooling tower thermal gradients ST-01-08 Cooling tower slab thermal gradients ST-01-14 Shear loads on deep beam walls ST-01-18 Bedrock pressure grouting ST-01-20 Cooling tower seismic loadings ST-01-21 Cooling tower slab seismic loadings ST-01-23 DBE vs. OBE seismic design spectra ST-01-29 Bedrock pressure grouting EE-01-01 Circuit breaker interrupting rating Detroit Edison Company Fermi 2 Independent Design Verification i NIL A Final Report, TR-83021-1, Revision 1 8.2-39 llllll1lll1111llllllllllll1lll
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, l ADDITIONAL NRC REVIEW COMMENT:
"In reviewing your resolution of Observation No. PI-01-11, we conclude that this observation should be dispositioned by you as a potential finding which had a strong probability for a potential impact on plant safety. We believe this disposi-tion is in accordance with your criteria for your conduct of the Fermi-2 independent design verification program (IDVP). Accordingly, we find your resolution of PI-01-11 to be unacceptable.
To place this matter in proper perspective, we have found that your resolution of the other observations regarding the Fermi-2 piping systems reviewed in you IDVP, to be appropriate and acceptable. On this basis, we find that the inadequacy of your resolution of PI-01-11 to be an isolated matter which we attribute to the nature of the DECO commitments in its FSAR for annulus pressurization (AP) loads. Our re-review of these DECO commitments has indicated that they were somewhat ambiguous.
Accordingly, we required DECO to address this matter of the AP loads in combination with other loads including seismic loads. DECO submitted a letter dated September 27, 1984, in which it resolved our technical concerns on this matter, including the question of whether there were any generic implications arising from this issue.
However, in the context of your IDVP, we cannot conclude that Cygna's rationale for originally accepting apparent deviations by DECO from its FSAR commitments
[] regarding AP loads, is acceptable as noted above. Based on your evaluation of the AP V issue discussed in DECO's letter dated September 27, 1984. While your response should touch on the matter of the original ambiguity, we require that you primarily focus on the acceptability of DECO's conclusions regarding the present method of ana-lyzing the Fermi-2 piping systems subject to AP loads. Additionally, provide justi-fication if you do not reclassify PI-01-11 as a potential finding."
CYGNA RESPONSE:
Within the strict context of a design requirement, this issue has been somewhat ambiguous with respect to whether Fermi-2 safety-related piping systems attached to the reactor coolant pressure boundary be designed to withstand faulted conditions including annulus pressurization loads. Cygna notes, however, that Detroit Edison did. agree as an NRC licensing commitment, to evaluate whether the Fermi-2 design was able to maintain structural integrity under combined annulus pressurization and Design Basis Earthquake (DBE) loadings. (NRC Question 110.11 in Appendix E.5 of the Fermi-2 FSAR and subsequent Detroit Edison response via Amendment 24 to the FSAR sub- l mitted June, 1979). '
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CYGNA RESPONSE (Cont.d)
To address the matter at hand, it is now apparent based on a) information in the NRC letter to Detroit Edison, " Preliminary Evaluation of the IDVP Performed by Cygna Energy Services for the Fermi-2 Facility" dated March 27, 1984 (Reference 1); b) discussions in the subsequent meeting between the NRC, Detroit Edison and Cygna in Bethesda, Maryland on May 11, 1984; and c) results contained in the Detroit Edison letter to the NRC, " Annulus Pressurization Piping Load Re-evaluation" dated September 27, 1984 (Reference 2), that Cygna should have dispositioned Observation PI-01-11 (Page 7.7-65) as a potential finding. This would, in all probability, have been the action taken if Cygna had not stopped our review of thic issue in May, 1984. (Refer to Cygna project letter 83021.05B dated November 30, 1984).
As indicated above in our response to NRC Question 2 in Enclosure 2 of Reference 1, Cygna proceeded in April 1984, to determine if the Detroit Edison pipe stress eva-luation for faulted loads, including annulus pressurization, was properly considered for the in-containment RHR shutdown suction cooling element. Because of differences between the piping geometry analyzed for A/P loads and the as-built configuration, Cygna was unable to confirm structural integrity of the piping element under faulted load conditions. The NRC, Detroit Edison and Cygna agreed shortly after the May 20, 1984 meeting to resolve the issue without requiring an expanded review effort by Cygna. As an outcome of this agreement, Detroit Edison would confirm the validity of their previous assessment for structural integrity by re-analyzing the as-built recirculation and drywell RHR piping for combined A/P and DBE loads. Detroit Edison
-[] would also verify that the original analysis input was adequately represented in the LI as-built configuration for other large bore (NPS > 4") reactor coolant pressure boun-dary piping systems. The confirmation for structural integrity was provided in Reference 2 (a copy of which was recently requested by Cygna via project letter 83021.059, dated May 7, 1985).
Cygna has evaluated the information contained in Reference 2 and supports the conclusion that structural integrity of the as-built recirculation and drywell piping would be maintained provided the stresses resultant from a combined annulus pressuri-zation and DBE loading were within 3Sm of ASME code allowable values and all supports were within their Level D component ratings. Cygna also agrees that the results of the original analyses submitted in response to NRC Question 110.11 from Appendix E.5 of. the Fermi-2 FSAR for other reactor coolant pressure boundary piping systems would remain valid as long as the analysis input was reflected in the as-built con-figurations.
'In summary, Cygna. concurs with Detroit Edison's conclusions contained in Reference 2. We also support the NRC position that there are no generic safety implications on Fermi-2 resulting from the issue of annulus pressurization loads.
The Attachment A to Observation PI-01-11 (page 7.7-65) has been revised to indicated that Detroit Edison and the NRC resolved this issue without requiring an expanded review by Cygna, w/
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t J V ADDITIONAL NRC REVIEW COMMENT:
"In the first sentence of the last paragraph on Page 8.2-62, you stated that the shear capacity was calculated as the sum of the concrete strength and the shear friction contribution from the reinforcement. Your basis for this approach is not clear. Accordingly, provide justification for this approach. "
CYGNA RESPONSE:
In response to a Cygna observation, Sargent & Lundy re-evaluated the shear wall along column row E in the RHR Complex (see Sargent & Lundy Calculation 1.20.9, Rev.
2, dated 5-30-84). Maximum shear stresses were taken from a finite element analysis of the shear wall. For design purposes, the most highly stressed element in the shear was selected. Then, the shear strength was calculated as a combination of concrete and reinforcement strengths, with the latter determined using the shear friction provisions in Section 11 of ACI Standard 318.
Cygna considered the use of localized maximum stresses to be over conservative.
Also, Cygna did not support the application of shear friction to a potential diagonal tension crack. As a result, Cygna performed an independent check of this wall prior to issuing the IDVP report supplement. In our evaluation, the overall shear strength of the wall was based upon ACI 318, Section 11, "Special Provisions for Walls".
[] Cygna concluded that the shear strength of the concrete alone was greater than the
() applied shear. Therefore, the wall satisfactorily withstands the design shear loading using conventional design methods, without consideration of shear friction.
A N.,)
Detroit Edison Company AMg j Fermi 2 Independent Design Verification lillililllllllilllllillillllll Final Report TR-83021-1: Rev. 0 8.2-46a i
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V ADDITIONAL NRC REVIEW COMMENT "Although there are differences in the load combinations listed in Tables 8.2.3.1-1, -2, -3, and -4, we believe that these differences should have no signifi-cant effect on the final design of most of the seismic Category I structures. Our basis for this position is the assumption that these structures were designed in accordance with the load combinations described in the FSAR which DECO committed to
~
follow. However, we note in the attachment to Potential Finding Report (PFR) No. 8 and in Observation No. ST-10-21, that a reduction factor of 0.75 has been used for load combinations involving seismic loads. This is not in conformances with any of the load combinations listed in Tables 8.2.3.1-1, -2 and -3. Accordingly, provide an explanation why the 0.75 reduction factor was used in the structural design."
CYGNA RESPONSE:
As stated in'the IDVP report supplement above, Cygna performed several follow-
. up reviews of structural calculations to ensure that appropriate load combinations were being used. One ci the follow-up review items was the 0.75 reduction factor which was employed during an intermediate design phase. Cygna reviewed one-third of the final load reconciliation calculations and found all structural load combinations to be in accordance with FSAR. The 0.75 reduction factor was not applied in any of (D these final design packages.
U Based on this large sample, Cygna has concluded that the final design of seismic Category I structures is in conformance with the FSAR load combinations.
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'~] -Detroit Edison Company Fermi 2 Independent Design Verification gej . yj ,Final Report TR-83021-1: Rev. 0 8.2-63a lllllll!!!ll11111ll11111lll111
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() ~ ADDITIONAL NRC REVIEW COMMENT:
"You indicate that a factor of 1.5 is recommended in a Dames & Moore report.
Provide that portion of this report which discusses the bases for such a factor."
CYGNA RESPONSE:
Please find enclosed a portion of Dames & Moore report, " Static and Dynamic Soil and Rock Studies, Fermi II Nuclear Power Plant, for the Detroit Edison Company",
dated February 3, 1970. This enclosure contains the cover letter, pages 1 - 16, and a list of references.
The factor of safety of 1.5 for static conditions is shown on Table 1 (page 7) and explained in the third paragraph on page 12. A factor of safety of 1.1 is also recommended for dynamic conditions. As explained in the report, these safety factors are recommended to allow for variations in compaction of backfill and for residual pressure that may result from compaction operations.
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Detroit Edison Company L4[ ; g d Fermi 2 Independent Design Verification lililllillllllllllillilillllli Final Report TR-83021-1: Rev. 0 8.2-67a
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-+-- -The Detroit Edison Company 2000 Second Avenue Detroit, Michigan 48226 t
. Attention: Mr. Leonard Johnson, De',ign Engineer Gentlemen:
Ten copies of our " Report, Static and Dyna:,1c Soil and Rock Studies.
(A") Fe rn' i l Nuclea r Powe r Plant , for The DEUoit Edison Company" are herewith submitted.
4
- - The studies reported herein were planned in collaboration with Hessrs. Leonard Johnson and Joe Funston of The Detroit Edison Company and
~
Mr. Glen Chauvin of Sargent & Lundy. The -scope of work was outlined in our confirming proposal to The Detroit Edison Company dated December 18, l!!69.
The data presented in this report has been developed primarily to provide appropriate soil, rock and fill parameters for use in the structurel design of the Reactor and Auxiliary Building tructures. Datc. presented for the li-inch and smaller crusher-run fill material is applicable only to this material placed and compacted to a density equivalent to that of the test area
, ,. Investigated.
[
j lt has been a pleasure to undertake this program and we appreciate your continued confidence in our firm. Please contact us if you should have any question,'or comments regarding this report.
Yours very t ruly,
. DAMES & MOORE h' f
..( .m) George D. Leal L/ 'GDL:MFE:ew Y ' Ten Copies Snbmitted i-s . cc: -(S) Sargent & Lundy, Engineers l- -
Attention: Mr. Glen Chauvin
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REPORT i
.. . STATIC AND DYNAMIC SOIL AND' ROCK STUDIES FERMI II NUCLEAR POWER PLANT FOR .
THE DETROIT EDISON COMPANY L
INTRODUCTION r
, _ .This report presents the results of our static and dynamic soil-gj. .
] - and rock studies.at the Fermi ll Nuclear Power Plant and collates related data
.. previously presented in the Preliminary Safety Analysis Report (PSAR) and its bF Amendment. (References I and 2).: The purpose of this study was primarily to
! provide design-parameters for the structural analyses of the proposed Reactor
[ .
.' and Auxiliary-Buildings.
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The co:nbined Reactor and Auxiliary Building structure area has plan
'. i , dimensions of 154 feet by 205 feet. We. understand that the structure excava-l
.. tion will extend to elevation 534, approximately 18 feet into bedrock. The
- 1. ground surface level throughout the area of the proposed structures is -#
- 1.
- presently at- elevation 564.to 567 feet, which Is at or near the proposed foundatlon' leve1. for the Turbine BuiIding,'Radweste Building and Service
,., Building.- It is assumed that, at completion, the, final surface grade throughout.
I: the plant area will be at elevation 585 feet.
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, [L !I SCOPE
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.The specific scope of work for this study was as follows:
1 - Determination of the in-situ density of compacted fill
~
material which consisted of li-Inch and smaller crusher-run dolomite rock material,
, 2 Determination of the static modulus' of elasticity of
' ', the fill material.
} , 3 - Determination of the static modulus of elasticity of
~
the in-situ glacial till.
) .
), 4 - Determination of- the dynamic modulus of elasticity, shear modulus, Poisson's rat io, damping factors, and
' \( shear and compression wave velocities for the proposed
~
fill material.
5 - Determination of the variation of the modulus of
{ .
elasticity with depth for the in-situ rock under both static and dynamic loading, f
- 6' -Determination of the soil lateral pressure parameters
! for both the in-situ rock and the rock fill under the static and dynamic conditions.
During the course of our study,14r. Glen Chauvin referred us to the Sargent r, Lundy letter to The Det roit Edison Company datdd November 22, Ico8
' and requested thet our scope be extended as follows:
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7 - confirmation of previously submitted values of density, wave velocities, Poisson's ratio, shear modulus, and L ,/ \ .
- ' dar.;p i r.g f o r t he in-situ rock.
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- ' 't 8 - Determination of representative values of density, L
, wave velocities, Poisson's ratio, dynamic modulus of elasticity, shear modulus and damping for the
, in-situ glacial till.
9 . Confirmation of bearing capacity and settlement for structures founded on bedrock.
. 10 - Discussion of problem areas that may exist in
.3 , construction of future units.
~*
. 'In connection with this investigation, a test stockpile of controlled compacted fill was constructed at the direction of The Detroit Edison Compc,y.
We understcnd that this test area was placed and compacted in a manner similar to that which'will be used during final placement. Soil and Foundations.
,m
- J' Associates assisted The Detroit Edison Company in quality control during -
construction of the test area. The approximate location of the test area l's indicated on the Plot Plan, Plate 1.
Soll AND ROCK CONDITIONS GE NE Rt,t. :
A complete description of the geologic and hydrologic features of
,, the region and the site area is presented in Reference (1). Based on the 1results of recent investigations in the immediate plant area, Reference (2)
.was submitted to The Detroit Edison Company as an AT.endment to Reference (1).
The-Information contained in this section is largely condensed and extracted from the two referenced documents.
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SURFICIAL DEPOSITS:
1
- Subsurface conditions within the proposed plant area were investigated by drilling test borings at the locations shown on Plate 1, Plot Plan.
Approximately five feet of lacustrine peaty silts to clay had been removed f rom the site area at the time of our most recent investigation in November
- 1969. This-exposed the surface of the glacial till deposit at an average
.- elevation of 566; this is approximately six feet below the surface of the
- . g adjacent La ke E r ie . The till consisted of nearly imperraeable silty to sandy
. _ clays with varying amounts of gravel and cobbles. Occasional boulders, up to 18 inches in diameter, were encountered within the till near the bedrock surface. At random locations throughout the site, the lower one to five feet of' till graded to a clayey silt with some gravel and occasional boulders.
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- (.eJ- In addition to test borings reported in References (1) and (2), three .
o additional borings, 12 to 13 feet in depth, were drilled to recover undisturbed-X . .
.- semples of the glacial till. Logs of these borings are presented in the L
, Appendix of this report. The till exposed at the present surface of the excavation is hard in' consistency and grades very hard within several feet
. of _ the bedrock surface. .
BEDROCK:
A ~ complete description of the bedrock is presented in Reference (1)
{_ . .
L' and (2). Briefly, the bedrock consists of the Bass Islands Group of Sediments F *
.( predominantly dolomite) underlain by the Salina Group (shales and dolomites).
,Throughout the proposed plant area the surface of the Bass Islands Group is at. elevation 357 to 549 and the contact between the Bass Islands Group ard the Salina Grnup is at elevation 483 to 457 feet.
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Prior;to foundation ' installation for the Reactor and Auxiliary
~
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Buildings, the bedrock throughout the foundation area will be pressure grouted f rom foundation level (elevation 534) down to elevation 405 feet in accordance with procedures outlined in Reference (2).
FILL MATERidt:
Crusher-run rock material, li inch and smaller in size, will be used
~
Y, as backfill adjacent to the proposed Reactor and Auxiliary structures. This
. fill material will- be predominantly dolomite and will be quarried on-site.
We understand that the backfill material will be similar to the presently
. produced 1 inch and smaller crusher-run rock. Approxir,ately 60 percent of this material is less than one inch in size and it contains up to 20 percent i fines (materials passing the standard U.S. No. 200 Sieve).
( We understand that the fill material will be placed in icose harltontal lifts approximately 10 inches in thickness and that each lift will be compacted 7
.. by approximately 10 passes'with a vibratory roller similar to that used to 3
,, compact the test area constructed during the course of this study. A h description'of the test fill area is presented in the Appendix.
} FIELD AND LABORAT00.Y STUDIES
[.
j.,, Field and laboratory studies undertaken for this investigation I '
consisted of the following:
~
(a) Drilling three borings through' the glacial till to recover undisturbed samples; (b) Laboratory tests on undisturbed samples; i ,,,. (c) Construction of a test fill area; i e f V (d) Plate load tests on the glacial till and rock fill; fY and, V 4.
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- (e) Field seismic studies.
- Descriptions of these studies are presented in the Appendix to this report.
Some bedrock test data developed during prior investigations is also repeated in the Appendix to provide a collated summary of test data used for
~
the development of soll and rock parameters.
. DISCUS $10N AND RECOMMENDATIONS s
SUMMARY
. Based on our analysis of the results of field and laboratory testing, together with a review of published data, recommended design parameters for soll, rock and fill materials have been developed and are summarized in Table 1, Static and Dynamic Soil and Rock Values. A discussion of these values
, is presented in subsequent paragraphs.
As outlined in Reference (2), the ultimate bearing capacity of the foundation bedrock is estimated to be on the order of 300,000 pounds per
- square foot. The total settlement of the Reactor and Auxiliary Buildings is conservatively estimated to be on the order of 0.3 to 0.5 inches for an essumed applied pressure of 25,000 pounds per square foot.
Consideration has been given to construction difficulties that may occur in the design and construction of future units. Major problen areas would be associated with rock excavation by blasting, and possibly with dewa te ri ng . It is recommended that the feasibility of performing rock excava-tion for future units prior to the operations of Unit 11 be evaluated further.
Similarly, it is reco:cmended that records of blasting, grouting, dewatering,
'^- and other pertinent construction operations for fermi il be collated and cor.densed
~
into a post-construction report that would deal specifically with future const c-y
, tion probier.:s.
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c N/ TABLE I
-l STATIC AND DYNAMIC Soll N!D ROCK PROPERTIES CRUSHED IN-SITU BASS ISLAfrS ROCK FILL GLACI AL Till B ED F.CCK DENSITY (PCF)
Dry Density 139* 4% 125* 4% - 150* 10;.
! Wet Density 144* 5%' 140* 5%- -
Submerged Density 90* 3% 80* 3% llot 101 UAVE VELOC! TIES (FT./SEC.) s
[. Co press ion Wave 2500* 15% 7700* 7% 13000* IC%
Shear Wave 900* 25% 2200* 15% 7600* 15%
. PO!SSONS RATIO
", Stetic or Dyn:nic 0.4* 10% 0.45* 10% 0.24* 10% .
~
110Duttis or ELASTICITY (PSF)
Static 1.2 X 106
- 25% 0.5 X 106
- 20% 120 X 106 e ;93 Dynamic >4.0 X 106
- 30% 1.2 X 106 30% 183 X 106 533 increase Per Foot of Depth 0.48 X 106
- 25% 0.48 X 106
- 20;; O p.!
_ SHEAR MODULUS (PSF)
F' ( 3 - Dynsn.ic 1.4 X 106 e 393 g,g x 396
- 30% 72'X 106 : 50%
increese Fer Foot of Depth 0.17 X 106
- 25% 0.17 X 106 * ;gg g D'".Pl:'r. VALUES '(i'ERCCNT OF CRITICAL DAliPING)
. < Within Earthquake Levels 7% to 10% 5% to 8% 1%
, _l ATERAL PRESSURE (PSF /FT.)
Stctic-Rigid-Vall Above Veter 96* -
0 Stctic-Rigid Vall Suoterged 122f. -
63
' Static-Cantilever Well Above Water 32* -
0 Static-Cont ilever Ucii Subtrerged 80* -
63 D'ynamic-Rigid Wc il Above Wate r '320** -
0 Dynamic-Rigid Wall Belcw Watc-r 280** - -
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- / oA factor of' safety of 1.5 is recow. ended in the use of these values
- A factor of safety of 1.1 is recorrmended in the use of these values 1
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-STATIC AND DYNAMIC S0ll AND ROCK PROPERTIES:
Each of' the parameters presented in Table I is -discussed' belcrr.
. A brief description of the trethod of determining the values is given, and the range of variation is discussed.
Densitv - The densities given for the rock fill material were deter-mined from six relatively large scale density tests performed by Soll and
,' Foundations Associates in the compacted test fill. The individual test results Lare presented in the' Appendix. In determining the submerged density, the rock
+
fill material was assumed to have.a specific gravity equivalent to that of dolomite. The range of variation given is considered appropriate'for a controlled compacted fill of li inch and smaller crusher-run rock.
The densities for the in-situ glacial till and their rar.ge of varia-j tion were assessed from the moisture-density tests performed on undisturbed i i
semples. An appropriate specific gravity was used to determine the subnerged
~
, , . d' ens i t y .
,-_ Bedrock density and its range of variation were determined fro:. the results of mecsured densities of representative rock cores.
Vave Velocit ies - The compression and shear wave velocities presented In Table I for the compacted fill and the glacial till are the values measured
,, during this investigation. The in-situ rock velocities are the values measured
~
during prior studies and previously reported in Reference (1). The tabulated -
91 acial till velocities (Ve=7700 ft./sec., Vs=2200 ft./sec.) differ from the previously c.casured compression wave velocity (Ve=6500 f t./sec.) and cor.;puted shear wave velocity (V3 =2650 f t./sec.) which were reported in Reference (1).
in our opinion, the currently developed values are more applicable ir. that 7;A)
W they were measured in the specific plant area.
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, : 1 i:. M( The range of variation of wave velocities presented in Table I has
,.- been estimated from consideration of inherent uncertainties in metheds of
. measurement, and variation in grain size, density, and/or strength of the
.various materials.
Poisson's Rat io - The tabulated values of Poisson's ratio for thecompactedrocklfillandglacialtillwereccmputedfromthemeasured
, shear and compression wave velocities. Where possible, the load-settlement
,; data from plate load tests were compared to provide a further check on the
{ ;, values corrputed from the wave velocities.
Values for in-situ rock were -
i previously estimated f rom the seismic investigation reported in Referen:e (1).
i The range of -variation of Poisson's ratio v.ere estimated f rom consideration of probcble variations in wave velocities, probable variatiens a
'in grain si:c, density and/or. strength of the materials being considered.
' Static Modulus of Elesticity.- The tabulated static moduli of f
i elasticity for the rock fill and glacial till were determined f rom the load-
.,; settlecent behavior recorded during plate load testing, with Interpreretion of these data by the~ methods outlined in Reference (5).- computed values were s.
co.:. pared with published data (References 3,4,6,13,17,18) and minor
~
adjusteents'were made as necessary.
+
The variations of moduli with depth were deternined f rom the test
~
results using the. methods of Reference (5). The tabulated variations with
. .. ~ ~
depth should not' be used for depths of more than 50 feet. Based on resecrch of published data and a comparison of results with meduli values determined for. glacial 'till at other nuclear pcwor plant sites, it is recorrrended that the depth taken in computing the modalus of elasticity of the till be the
~
depth f rom the io.cest adjacent ground surface to the till layer being consider-ed, t
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.. v The modulus of elasticity of the bedrock was computed using the J. elastic moduli _information developed during testing but modifying the measured values on the basis of experience, RQD, vugs, discontinulties,' clay seans, and proposed grouting, to produce a modulus appropriate for the in-situ rock.
The tabulated value is applicable to'the Bass Islands Group of sediments and the range indicated covers variations that may result f rom variations within this bedrock group. No marked variation of modulus with depth or overburden o
pressure is expected for the bedrock.
.- The _ indicated range of values reflect the inherent errors of testing and analyses together with anticipated variations in properties of the varlocs materials.
Dynamic Modulus of Elasticity - The dynamic moduli for the compactc :
,) rock fill and glacial till were determined f rom clastic analysis of the
- results of the' field seismic investigation. The computed values were adjus ted to give values' which would be applicable for the anticipated strein levels
- which will be developed by the adopted earthquake levels. The results of the elastic analysis 'were also compared with the moduli ccmputed by the rethods of: Reference (5) using the rebound portions of the load-settlement curves..
~
When adjusted for strain level and confinement, the elastic analysis results and rebound . values compared well; thus the anticipated variation with depth
- co. puted by Reference (5) methods are considered appropriate.
The dynamic modulus of elasticity of the bedrock was determined by elastic analysis of the results of the' seismic investigation. -Computed
- ~
values were adjusted for strain level to give a value appropriate for the
,7-~g "9 routed in-situ rock within the adopted carthquake levels, in our opinion, the t.' ' '/
modulus will not vary markedly with depth or confining pressure.
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The range of values given reflects the accuracy of fleid measurerent
,7 and analysis together with the anticipated variations in grain size, density and/or strength' of the various materials.
^
Shear Modell - The shear moduli were computed using the elastic relationship between shear modulus, modulus of elasticity and Poisson's ratio.
The tabulated values of modulus of elasticity and Poisson's ratio were used
'L ,
and thus the shear moduli as tabulated' are appropriate for the adopted earth-
- . _q ua ke le ve l s . Similarly, the range of values reflects inherent uncertainties
. In methods of analysis and anticipated variations in grain size density and/:.-
strength of the various materials.
Damoino Values - For the compacted rock fill and the giccial till, an attempt was made to' determine damping f rom the behavior of load-unlocd cycle-
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Similarly, the energy losses of wave trains -develccei
(.
of the plate. load. tests.
in the seismic investigation were studied. Although these studies cave an-
" ; indication of the relative damping capacity of the two materials, a precise C 'astessment of damping'was not possible by these' methods. The tabulated values of damping are based largely on a review of available published dcta.
The dar. ping capacity of the bedrock was developed during prior studies reported in heference (1).
~
All of the tabulated damping values are expressed as a percentace c' c-itical damping.
Lateral Pressq-es - In cocputing lateral pressures appropriate fc.-
t,e compacted- rock fill, it was necessary to estimate the probable angle of iiternal friction of this material. Based on observations of the- ma te r ia !
g jg Fl aced in the field and based on research of available published data, the a .,'c
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o '. Interval f riction was assumed ec;ual to 40 degrees, s:
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< ' All static lateral pressure dcta presented in Table I are expressed v . -as' equivalent fluid pressures. For rigid walls, the tabulated values of laterol pressure _ are derived for the case of earth pressure "at rest." For cantilever walls, the tabulated values are derived for the case of." active" earth
- pressure..
.Dynanic lateral pres:tres for the rock fill and glacial till were t*-
vn 7
~ determined from'" passive" earth pressure theory allowing for the possible
'.h. -.'
- Increases in pressureishich could result from scismic accelerations. The i.
(- 1
' tabulated pressures will occur.only for that portion of_the substructure whicb .is out of_ phase with the adjacent backfill .during movement due to earthquake motion. : These pressures need not be considered -to act over the ent i re height ofL th'e subst ructure.
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For static- conditions, a factor. of sefety on the order.of 1.5 is
~
.reco . ended in -the use of the recom. mended desien values. This is to allo;
~
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' for veriations-In compaction of backfill and for residual pressures tht.t ry
>* result from co;.paction operations. For dyntmic conditions, a fect'or of f, safety on the order of 1.1 is recommen!ed .for similcr reasons.
It is _ cur opinion 'that static pressures -Impcscd by rock cn' rigid er
- cantile'ver walls above the grcund water level will be negligible. The !?teral
-Pressure in rock cuts belcu the water table will oc linited.to hydrosta:.ic.
-water. pressure. This assures that the walls will be poured dirc'ctly agoir.st i the' blast-excevated rock fcce. To assun. applicability of these criterie,
- it is recommended that the exposed rock wall be inspected by a qualified n
9eologist to insure that any rock masses which are loose or highly fractured
.are_ renoved or stabilized.
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Dynamic lateral pressures -in the bedrock will be controlled 6y the
.. rock-structure interaction during earthquake loading. To determine stress levels during seismic interaction, it is customary to construct a model and analyze the seismic beh'avior of the ground-structure system by finite elenent analysis. We assune that such a model can be con'structed using various rock parameters previously provided in this report. If this is not the case,.
~
we would be pleased to provide any additional data that might be required.
. 7, ROCK BEARING CAPACITY:
Data on the rock bearing capacity has. been presented in Rcference
-(2) and is repeated herein.
The ultimate bearing capacity of the foundation bedrock was evaluated, on a conservative basis, in accordance with methods described in y
) Rcierences (9) and (16). No consideration was given to the increase -In bearing -
4 capacity which will result from the grouting operations.
The strength of the foundation rock was evaluated by m-ans of rock
+ . compression tests. Considering this value to be appropriate for reck with en RQD ( Rock quality Designation) of 100, a reduction factor was selected based g
.q. on an assessment of the measured RQD values, informction en vug volume and size, fracture orientation and spacing, and presence of clay and shale secc.s.
. ., Application of this reduction' f actor produced a modified value approxicating the in-situ strength of the rock mass. On.this basis, the mininun ultimate bearing capacity' of the rock mass in the plant area is considered to be on.
I f the order of 300,000 pounds per square foot. Assuming a combined static and dynamic maxinun loading as high as 25,000 pounds per square foot, the factor n of safety. against foundation failure would exceed 12. Considering the rock strengthening by grouting operations, the factor of safety would be considerably
,. In excess of 12
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SETTLFMEllT:
l' Settlement data were also presented in Reference (2) and ~are repcoted'
};
below.
I)etailed design loads for. the Reactor 'and Auxiliary Buildings are
- presently not available. If the maximum unit pressure were to be as high as
- 25,000. pounds per square foot, it is estimated that the Reactor and Auxiliary
. Buildings would' undergo a maximum total settlement on the order of 0.3 to 0.5 e -
inches. This estimate has been computed using the elastic moduli inforcation
. developed during testing but modifying the measured values on the bcsis of experience, RQD, vugs, discontinuities, and clay seams to produce conserva-
. tIve def ornation moduli approprlate for the in-situ- rock.
7-'
i _
If applied unit pressures for the Reactor and Auxiliary Euildings
!_ fm are less than 25,000 pounds per square foot, actuel settlements would be f,(j
-~
reduced proportion'ately.
'~
FUTL%E UtilTS:
Construction of future units will be affected by the Fermi 11 unit =
t particularly with respect to rock excavation by blasting. Prior studies of
{' o.
viorations f rom blasting (References -7 and 8) ~ established tentative criterir I
l* for: shot-load versus distance f rom Fermi t . It was tentatively established 3.,, that not more than a 25 pound per delay shot load should be used at a miniru-
- distance of 400 feet f rom Fermi I and that.the total' shot load for delayed
['
- 1
' detonation be'linited to 1,000 pounds per shot. These limits are subject to review-and confircation or possible revision during production blastins fC'r y~
- 1. Fermi.II.
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a .0. if a: future unit.Is to be constructed immediately adjacent to
. . Fermi II, careful consideration shou!d be given to the planning of excavating ope ra t i ons . If possible, . rock ' excavation for the future unit might be conduct-ed. concurrent ly with that for Unit 'll . If this is not feasible, further studies will be required to establish blasting load.llmits.and other blasting criteria.
Planning. for a future unit should also consider the resulting
[" . unbalanced lateral pressures occurring due to backfill on one side of Unit II,
,,, and-open cut on the opposite side. As a' preliminary guide, it is estimated
.that the ' coefficient of friction preventing sliding of the Reactor foundation on:the grouted bcdrock will be on the order of 0.6 This estimated value should be checked when the grouted bedrock is exposed throughout the Fermi 11 foundation. area.
p 3j- The 'af fect of dewatering for future units should also be studied, .
with particular attention given to lateral pressure variations f rom submerced
- '- to non-submerged' conditions. Similarly, in the design of appurtenant earth-i
- 1 suppor'ted-buildings, possible settlement which ray result' f rom dewatering should be analyzed.
it is recommended that accurate records of grouting, dewetering, a
1 blasting and construction operations be kept during construction of Fermi ll.
r;. A_ post-construction report collating these data would contribute significantly
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to an assessment of' possible future construction problems.
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.The following Plates and Appendix are attached and complete this
. report:
1
_ Plate I- -
Plot Plan I 4
Q Appendix- -
Field and Laborat,ory Explorations
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,- 4 Respectfully submitted,.
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, ' DAMES & MOORE t *;
Ot f Y George D.. Leal
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'f/ REFERENCES T
I '
- l. Preliminary Safety -Analysis Report, Volume 1, Section 2-Site, Enrico Ferni Atomic Power Plant, Unit '2, The Detroit Edison Company
-(Published 1969).
- 2. Amendment to Preliminary Safety Analysis Report, Volu r.e 1,
- Section 2-Site, _ Enrico Fermi Atomic Power Plant, Unit 2 The Detroit Edison Company (Published 1970).
,. 3. Barkan, D.D., 'Dynamiesof Bases and Foundat iens , McGraw-Hill, Inc. ,
...- ,, 1962 v
J. 4. Bowles, J.E., Foundation Analysis and Desien, McGraw-Hill, Inc.,
- r. , 1968
- 5. Burmister, D.M., Prototype Load-Scaring Tests for Foundations of Structures, American Society for Testino and Meterials, Special Technical Publication No. 322, pp.98-119, 1963.
6.. Burmister, D.M., Physical Stress-Strain,'and Strength Responses of Granular Soils, A erican Society fer Testino and Meterirls, Special Technical
,.q . Publication No. 322, pp. 67-97, 1963. ,
~
17 . .. Dames & Moore, Reoort, Test Blas t i_no Proertm. Enrico Fe rmi
- Nuclear Po.ser Station,: Hear Monroe. Michican, for The Detroit Edison Company, July 2, 1969.
8 Dames & Moore, Report u Bitst-l_nduced Vibraticn s'nsitivitv Studv, Enrico Fe rmi Nuclea r Pc.ve r S tat ion. - Nea r ronroe. Mich icen, for The Detroit
- Ed i son Company , -_ Septembe r 29, 19os.
- ~
- 9. ;Duncan, N. and K. E. Hancock, The Concept of Contact Stress in the Assessment of the Behavior of Rock Masses as St ructural Foundations, Proceedines~of _t_he 1st Connress, Internationti Society of Rock Pecher.ieb,
- pp. 467 '+97, 1966.
- 10. - Hall, E. B., and B. B. Gordon, Triaxial Testing with Large-scale DHigh Pressure Equipment,' gerican Societv for Testino and Materiols, Special .
. Technical Publication No. 361, pp. 315-328, 1963. '
' 11. Hardin, B.O. _and F. E. Pichart, Jr., Elastic Wave Velocities in Granular. Soil' s , Journal of the Soil . che..;c<, and Feundation Division. P rec ce d_-
' ings of the A :erion Society of Civil Enqir+en;, voi . eg, SM), pp, 33-63, 19o6,
- 12. Leslie, .D. D., Large Scale Triaxial Tests on Gravelly Soils, Proc'eedinos of t_he Second Pan Amrican Conference on Soil Mechanics and Field
. Ennineerin,,.Vol. I, BraziI, 1963.
- p
'h ) ,
- ;13 Seed, H.B. and I. M. Idriss, influence of soil Conditions on Ground Motions. During Earthquakes, Journal of the Soil F.echenics and Foundat ions
- Division, Proceedinos of the American Societv 61 Civil Ennincers, Vol. 95, No. SMI, pp.99-137, 1969.
~"#
___________._.m_ _ _ _
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-q) 14 Sowers, G. F. , A. D. Ribb, C. H. Mullis, and A. J. Glenn, b"' . The Residual ' Lateral. Pressures Produced by Compacting Soils, 'Proceedinos of
_the Fourth International Conference on Soll Mechanics and Foundation Fncinearir ,
, .Voi, li, pp. 243-247,. London, l%7.
- 15. Soil and Foundations Associates, Report, Repetitive Plate Bearine Tes't Data, I inch Crushed Stone Fill . Enrico Fermi Atonic Pee:er Plcnt, for The Detroit Edison Company, January 3, 1970.
-16. Stagg, K. G. and O. C. Zienkiewica, Rock Mechanics in Encineering r- Practice, John Wiley & Sons, London, 1968
[.'
j f* . - 17. Terzaghi, K. and R. B.' Peck, Soil Mechanics in Enaineerino '
Practice, 2nd Edition, John Wiley & Sons, Inc., 1967.
b' : -- i' L *~ . '
- 18. Terzaghi, K., Theoretical soil Mechanics, John Wiley & Sons, Inc.,
l 1965.
- 19. Weissrann, G. F. and White, Small Angular Deflection of Rigid Foundations, Geotechnicue, Vol. XI, 1961
- 20. Weissmann, G. F. , Measuring the Modulus of Subgrade Soil l Reactions, Materials Research and Standards, pp. 71-74, February 1965.
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' ADDITIONAL NRC REVIEW COMMENT:
"In the original Observation No. ST-01-21, values of Mt equal to 2134.9 kip-feet and MEQ equal to 1255 kip-feet, are shown. Indicate how these values are related to the values shown in Table 8.2.3.6-1.
.CYGNA RESPONSE:
The values for Mt and MEQ, contained in the original Observation No. ST-01-21, represent the total applied overturning moments due to tornado and OBE loadings, respectively. Sargent & Lundy used these values in their analysis of the cooling towers.
The values shown in Table 8.2.3.6-1 summarize Cygna's finite element analysis of the cooling towers. The moments tabluated are the maximum moments (kip-feet /ft) developed within various elements in the finite element model. These moments should be considered local resulting moments. Of course, such local moments have much smaller magnitudes than the total applied overturning moments.
To show that the Sargent & Lundy and Cygna results are consistent, we have f') calculated the total overturning moments for tornado and OBE loadings using a) our v more detailed analysis input and b) Sargent & Lundy's analysis' approach. The resulting overturning moments are as follows:
Tornado: 2135 kip-feet
,0BE : 1237 kip-feet The primary reason for the minimal (< 2%) difference in OBE moment is due to Cygna's more detailed calculation of the cooling tower mass.
O
'~) Detroit Edison Company Fermi 2 Independent Design Verification y _3 y,3 Final Report TR-83021-1: Rev. 0 8.2-72a l111ll1111lll111111lllll1ll1ll
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