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%AR 5".1985 Docket No.: 50-323 MEMORANDUM FOR: Dennis Kirsch, Acting Director Division of Reactor Safety & Projects Region V James P. Knight, Acting Director Division of Engineering Brian Grimes, Director Division of Quality Assurance, Vendor and Technical Training Center Programs Robert M. Bernero, Director Division of Systems Integration Themis P. Speis, Director Division of Safety Technology FROM:
Hugh L. Thompson, Jr., Director Division of 1.icensing
SUBJECT:
DIABLO CANYON UNIT 2 - SSER 29 Enclosed is the final draft of SSER 29 for Diablo Canyon Unit 2 for your review and concurrence. SSER 29 is the staff evaluation of (1) the PG&E Internal Review Program (IRP) which determined the Unit 2 aoplicability of findings from the Unit I design verification program and (2) the PG&E. specific technical resolution of many of these concerns for Unit 2.
SSER 29 wil1 be provided to the ASLAB as input to its decision regarding the need for further hearings on the matter of design QA. We have in preparation the following additional reports:
SSER 28 - Allegations resolution status for Unit 1 and applicability for Unit 2 SSER 30 - Pipina systems and pipe supports for Unit 2 SSER 31 - Pesolution of issues from earlier SSERs, PGAE commitments, bases for license conditions, etc All of the above need to be issued prior to a licensing decision (PG&E schedule for fuel load is late March).
Contact:
H. Schierling, NRR
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Your review is specifically directed to Section 2.4 - Staff Evaluation and Conclusion, and those sections to which your staff provided input. Table 1 is the list of contributors. Table 2 identifies the only two open issues.
You should verify their status. At this time it does not appear that any issues in SSER 29 would preclude low power authorization or would result in a license condition.
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An advance copy of draft SSER 29 was provided to.each contributor on Marth 1, 1985. We request comments by COB March 6,1985 fn order to issue the report by March 8, 1985. Any questions should be directed to H. Schierling (x27100).
IsHugh L. Womd bn r., Director F Division of Licensing
Enclosure:
k As stated cc:
D. G. Eisenhut R. Vollmer F. Miraglia T. Novak C. Grimes R. Bosnak I
L. Rubenstein 4
G. Lear
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cc w/o enclosure:
P. T. Kuo H. Polk M. Hartzman A. Lee H. Walker J. Spraul J. Wermiel 0
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Table 1 Diablo' Canyon SSER 29 Contributors Section 1 H. Schierling Section 2 H. Schierling Section 3 H. Polk /P. T. Kuo Section 4 H. Polk /P. T. Kuo Section 5 H. Polk /P. T. Kuo Section 6 H. Polk /P. T. Kuo Section 7 H. Polk /P. T. Kuo Section 8 H. Polk /P. T. Kuo Section 9 H. Schierling (based on draft SSER 30)
Section 10.1 J. Wermiel 10.2 M. Hartzman 10.3 A. Lee /H. Walker Section 11 J. Spraul/D. Kirsch Section 12 H. Schierling e
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p Table 2 Diablo Canyon SSER 29 Open Issues 1$
Pipeway Structure:
Concern regarding analytical details of member connections; resolution -
(analysis and mods) to be completed prior to full power operation -
Section 2, Section 3.1 and Section 8.
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Turbine Building:
Concern regarding shear transfer mechanism; resolution (analysis and mods) to be completed prior to full power - Section 2 and Section 5.
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Supplement No. 29 E
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l Safety Evaluation Report f
related to the operation of i
Diaalo Canyon l\\ uc ear Power PlaK s
Units 1 and 2
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Docket Nos. 50-275 and 50-323
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ABSTRACT i
Supplement 29 to the Safety Evaluation Report for the application by the i
Pacific Gas and Electric Co. to operate the Diablo Canyon Nuclear Power Plant -
1 Unit 2 (Docket No. 50-323) has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission. SSER 29 reports on the f
applicant's Internal Review Program for the Unit.2 applicability and resolution of concerns that had been raised during the Unit I design verification by the Independent Design Verification Program, by the applicant's Internal Technical
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Program and by the staff of the U.S. Nuclear Regulatory Commission.
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DIABLO CANYON SSER 29 TC 4
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,n ra TABLE OF CONTENTS Page Abstract..........................................................
1.
INTRODUCTION AND BACKGROUND..................................
1.1 Unit 1 Design Verification..............................
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1.2 Allegations.............................................
1.3 Piping and Pipe Supports................................
1.4 Recent Unit 1 Licensing History.........................
2.
INTERNAL REVIEW PROGRAM FOR UNIT 2.....
2.1 Unit 2 Engineering Project Organization.................
2.2 Internal Review Program........................
InternalReviewProgramResults..............J.3:I..bf.f#d'b#"
2.3 2.4 Staff Evaluation and Conclusion...............
.3 : 2.. NS.C (C u c 8"* 5 M* U"N I 3.1./ Reidehen for (m esi 3.
SEISMIC DESIGN ASPECTS OF CIVIL STRUCTURES........
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CCNTAINMENTANNULUSSTRUCTURE.........................3..MAJolv*f%i Nd M, 3, ).
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4.1.1 Comparsion of Unit 1 and Unit 2...............
4.1.2 Unit 1 and Unit 2 Modifications...............
4.2 Scope of Review.........................................
4.2.1 Vertical Seismic Evaluation...................
4.2.2 Horizontal Sei smic Evaluation..............
4.2.3 20 Hz Cutoff.
4.2.4 Connection Evaluation........................
4.2.5 Member Evaluation.............................
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TABLE OF CONTENTS (Continued) c Pm 4.2.6 Tangential and Radial Beam Evaluation.........
4.2.7' Column Evaluation.............................
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4.2.8 Torsion Evaluation............................
4.3 Findings and Conclusions................................
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TURBINE BUILDING.............................................
5.1 % ). h b M r4.%.............
5.2 Scope of Review.........................................
5.2.1 Specific Calculations............................
5.2.2 Concrete Floor Diaphragm Strength Evaluation.....
4, 5.3 Findings and Conclusions................................
5.3.1 Independent Evaluations..........................
i 5.3.1.1 Shea r Wa ll Analy si s.....................
5.3.1.2 Buttress Mode1..........................
5.3.1.3 Floor Diaphragm at Elevation 119 Feet...
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5.3.2 Conclusions......................................
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6.
RACEWAYS.....................................................
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t 6.2 Scope of Review.........................................
6.3 Findings.=nd Conclusions...............
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BURIED C0NDUITS..............................................
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7.1 C _. ~_. e ',. 3C. k. 4r.e.d..g6. %. f. e. k...............
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Page 7.2 Scope of Review.........................................
7.3 Findings and Conclusions................................
8.
PIPEWAY STRUCTURE.......
8.1 Gra-n...I.o.fra.d4.d!.#.%...............................
8.2 Scope of Review...
8.3 Findings and Conclusions.
9.
PIPING SYSTEMS AND PIPE SUPPORTS.............
- 10. NON-SEISMIC DESIGN ASPECTS..................................
10.1 Systems....
10.1.1 Component Cooling Water System..................
10.1.2 Protection from Jet Impingement Oue to Moderate Energy. Li ne Breaks.....................
10.2 Jet Impingement Analyses..
10.3 9e+r=4c Equipment Qu} i fication..
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- 11. QUALITY ASSURANCE.............
12.
REFERENCES..............
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DIABLO CANYON SSER 29 TC e
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1 INTRODUCTION AND BACKGROUND On October 16, 1974, the staff of the U.S. Nuclear Regulatory Commission (NRC) issued its Safety Evaluation Report (SER) concerning the application of the Pacific Gas and Electric Company (PG&E) to operate the Diablo Canyon Nuclear Power Plant, Unit I and Unit 2 (Ref. 1). The SER was supplemented by Supplements No. I through 17 to present the staff's safety evaluation of ongoing reviews and additional developments and requirements that were applicable to both units.
The SER was further supplemented by Supplements No. 18 through 27 to present j
the staff's safety evaluation of certain issues that had been raised, principally j
with respect to Unit 1 (Ref. 4 through Ref. 13). The subjects addressed in these Safety Evaluation Report Supplements (SSERs) are the Unit I design verification (SSERs 18, 19, 20, and 24), allegations (SSERs 21, 22, 26, and 28), fire protection (SSER 23), piping and pipe supports (SSER 25), and matters concerning the Unit 1 full power license (SSER 27). Each of these subjects is discussed further below.
This is Safety Evaluation Report Supplement No. 29 (SSER 29) and addresses the staff's evaluation of the application, implementation and results of the Unit 1 l
design verification effort with respect to Unit 2.
SSER 29 also provides an j
overview of the staff's evaluation of the other matters listed above with respect
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to Unit 2.
4 1.1 Unit 1 Design Verification j
The Diablo Canyon Unit 1 design verification effort was initiated in late 1981 as a result of Commission Memorandum and Order CLI-81-30 (Ref.16) which suspended
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the Unit 1 low power Operating License No. DPR-76 (Ref. 24). The order required the completion of an Independert Design Verification Program (IOVP) for seismic service-related contract desigt activities prior to reinstatement of the suspended license.
In addition, the Direc ur of the Office of Nuclear Reactor Regulation requested an extension of the IOVP to non-seismic service-related contract de-sign activities and PG&E internal design activities (Ref.17). These had to be completed prior to issuance of a full power license for Unit 1.
The objective 02/27/35 1-1 DIABLO CANYON SSER 29 SEC 1 1
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of the IOVP was to verify the adequacy of the Unit I design because the staff-had questioned the adequacy of the implementation of the PG&E quality assurance program with regard to design activities. The IDVP was conducted under the i
management of Teledyne Engineering Services and by organizations independent of PG&E.
In late 1982 the IDVP was expanded to also include selective verification
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of construction quality assurance efforts.
In addition, PG&E initiated its own
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Internal Technical Program (ITP) under which appropriate actions were taken to l
resolve issues identified by the IOVP and, more importantly, which expanded the j
scope of design verification to include all_ seismic safety-related structures, l
systems and components. The staff safety evaluation of the adequacy of the l
IOVP and ITP were reported first in'SSER 18 in August of 1983.
The report also identified a number of issues which the staff required to be resolved at various stages in the Unit I licensing process, i.e., prior to fuel load, criticality, and exceeding the 5 percent power level.
SSERs 19, 20c and 24 presented the staff evaluation of the resolution of the issues in support of the various
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licensing actions.
The Atomic Safety and Licensing Appeal Board (ASLAB), on the basis of a heiring on motions to reopen the record on construction quality assurance, c;ncluded in ALAB-756, December 1983, that the Diablo Canyon record need not be reopened on this matter (Ref. 18).
Hearings were conducted by the ASLAB in late 1983 on the matter of design quality assurance on the basis of the IDVP and ITP and the staff evaluation in SSER 18. The ASLAB concluded in ALAB-763, March 1984, on the basis of the design verification program that Diablo Canyon Unit 1 adequately meets its licensing criteria (Ref.19).
The Board required (1) that appropriate technical specifications for the component cooling water system be imposed and (2) that appropriate jet impingement analyses be completed. The issues were i
resolved in SSER 16 and License Amendment No. 8 with respect to the first item item and in SSER 24 with respect to the second item (Ref. 20). ALAB-763 further stated that the license authorization previously granted for Unit 2 is not effective until the Board makes its findings on design verification for that unit.
Section 2 of this SSER describes the PG&E Internal Review Program which was established for Unit 2 to monitor and document the review, evaluation, and reso-i lution of issues that had been raised during the Unit 1 design verification.
02/27/S5 1-2 DIABLO CANYON SSER 29 SEC 1 q
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I 1.2 Allegations 3
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In early 1982, during the course of the Diablo Canyon Unit 1 design verification program, certain allegations were'made to the staff regarding the design and j
operation of the Diablo Canyon Unit component cooling water system and certain other design aspects. The staff ' evaluation of these particular matters is pre-sented in SSER 16. Since then numerous allegations have been made regarding l
the design, construction, and operation of the Diablo Canyon Nuclear Plant and the PG&E management of these acti-vities.
In late 1983 the NRC staff instituted l
the Diablo Canyon Allegation Mana'gement Program (DCAMP), which provided specific procedures for the evaluation of allegations and for their reso_lution (Ref. 21).
Many of the allegations were contained in affidavits and exhibits in support of
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petitions by the Government Accou,ntability Project (GAP) submitted to the NRC 1
q pursuant to the provisions of 10 CFR 2.206. As of the end of 1984 approximately j
1650 allegations had been received by the NRC.
The staff evaluation of the l
allegations with respect to their safety significance for the Unit I licensing i
decisions, is presented in SSERs 21, 22. 26, and 28.
l In late 1984, PG&E established an allegation review program to track all allega-
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tions identified in the NRC program, to determine their applicability and impact j
on Unit 2 and to ensure that all commitments or modifications resulting from
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the Unit 1 effort have been appropriately addressed and implemented for Unit 2.
l The effectiveness of this program depends largely on the information available
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to PG&E. The program was described in the PG&E Final Report of December 5, 1984 (Ref. 22).
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The staff is currently completing its evaluation of the applicability of all j
allegations to Diablo Canyon Unit 2.
The staff has found that in many cases j
the same evaluation and resolution is equally applicable to both units. The t
staff will document its conclusidn in a report prior to a Unit 2 license decision.
1.3 Piping and Pipe Support l
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Beginning in late 1983 allegations were identified to the NRC pertaining to j
piping and pipe supports, including the engineering and design practices within l
the Onsite Project Engineering Group (OPEG) at the Otablo Canyon site. The l
02/27/85 1-3 DIABLO CANYON SSER 29 SEC 1 i
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staff evaluation of the initial allegations was included in SSERs 21 and 22.
Because of the large number of allegations in this particular area, a special NRC review group was formed which investigated and evaluated these allegations on a more systematic basis. The review group included personnel from various NRC offices, regions, and many consultants. During the course of the investiga-tion, a member of the review group raised additional concerns which subsequently were also considered by the group.
In early April 1984, the group concluded that the issues raised in the allegations and by the review group member were not of sufficient safety significance to preclude the Commission to fully reinstate the suspended low power license for Diablo Canyon Unit 1 (Ref. 23).
The group also identified seven conditions which were included in the reinstated led power license and which had to be satisfactorily resolved and completed prior to issuance of a full power license (Ref. 24).
In mid-April 1984, the review group was supplemented and reoriented to inspect and evaluate the PG&E actions and responses resulting from their effort with regard to the seven license conditions. The effort was completed in July 1984 and the review group's safety evaluation is presented in SSER 25.
In addition, the staff performed an audit and evaluation of certain programmatic, aspects related to the engineering practices.
The results are documented in an NRC memorandum issued as Board Notification BN No.84-161 (Ref. 25).
The existing PG&E piping and support effort for Unit 2 was redirected in early 1933 to take into consideration the results of the then ongoing Unit 1 design verification effort and again in early 1994 to oe responsive to allegations in this area and to the Unit I low power license conditions.
The final result of the effort are presented in the PG&E submittal of January 31, 1985 (Ref. 26).
The results of the staff effort, including its review and evaluation of the PG&E Final Report, are presented in Section 9 of this report and the details will be the subject of SSER 30.
1.4 R cent Unit l_ Licensing History i
The staff review and evaluation effort since the beginning of the design veri-fication program for Unit 1 in late 1981 was almost exclusively in support of the Unit I licensing process for reinstatement of the low power license, OPR-76 02/27/85 1-4 OIABLO CANYON SSER 29 SEC 1
t (Ref. 31) and issuance of the. full power license DPR-80 (Ref. 27). The Commis-sion authorized fuel loading on November 8, 1983, and fully reinstated the suspended low power license effective on April 19, 1984 (i.e., criticality and operation up to 5 percent power). On August 10, 1984, the Commission authorized issuance of a full power license. The full power license was issued on Novem-ber 2,1985, after the lifting of a stay imposed by the U.S. Court of Appeals.
(Ref. 27),
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2 INTERNAL REVIEW PROGRAM FOR UNIT 2 1
- Diablo Canyon Unit 2'is essentially of the same design as Unit 1.
Some differ-ences in_the Unit 2 configuration are due to the mirror image layout and some 1
differences in the design are due to the later construction date. Specific differences exist in the nuclear steam supply system (core thermal rating, structural aspects of reactor vessel, overpressure protection feature), safety-j related structures (containment annulus, turbine missile shield for auxiliary.
feed pump, structural steel for fuel handling building, turbine building), safety-i related balance-of plant fluid systems (due to different power levels, unit specific arrangement and equipment ratings), and safety-related electrical systems l
(diesel generator and loads for de systems). The differences have been identi-fied in the common FSAR for Units 1 and 2 and were addressed, as appropriate, in the Safety Evaluation Report of October 1984 and Supplements 1 through 17.
2.1 Unit 2 Engineering project Organization The Diablo Canyon Unit 2 engineering effort since early 1982 has been performed by an organization within the Diablo Canyon Project (DCP), a PG&E organization extensively supplemented by engineering and management personnel from Bechtel power Corporation in San Francisco.
In general, the organizational structure is similar to the Diabic Canyon Unit 1 engineering organization, also part of the DCP. The Unit 2 Project Project Manager directs the effort which consists of (1) seismic design, (2) system design, (3) special projects, and (3) quality l
engineering.
In addition, the Unit 2 Internal Review Program (IRP) was estab-4 lished as an integral part of the Unit 2 engineering organization with the IRP director reporting to the Project Engineer, as explained below.
Some of the management in the Unit 2 engineering organization was also respon-sible for the same function in the Unit 1 organization, in particular at the j
group supervisor level. The technical staff in the Diablo Canyon Project, in general, was not assigned to a specific unit but performed the same function for bcth units.
In many cases a particular issue was analyzed and resolved by 1
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the same. individual for both units at about the same time.
In other cases, after-completing a set of' assignments on Unit 1 the same individual would perform the same' assignments on Unit.2. As the Unit 1 effort came to completion in the first half of.1984 the Unit 2 effort increased. The use of the same personnel
~1n the same assignments for both units assured a common understanding of the issue and more effective resolution. Due to the extensive effort in the areas of seismic'and piping and pipe support analyses, some personnel was used for only one unit; however the same management and technical supervision was re-sponsible for both units.
2.2 Internal Review program The PG&E Internal Review Program (IRP) was established in late 1982 to deter-mine the Unit 2 applicability of issues that had been identified as a result of the design verification effort and to assure that appropriate resolutions for Unit 2 were developed and implemented.
PG&E first referred to the IRP in a submittal of October 6, 1983, and subsequently described the program in detail and provided the results in submittals of July 31, October 19, November 2, December 7,1984 and February 21, 1985 (Ref. 28). The program was discussed with the NRC in a meeting on September 13, 1984 (Ref. 29).
The IRP included the issues that (1) resulted from the Unit 1 IOVP and ITP; (2)
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were identified by the staff and addressed in SSERs 18, 19, 20, and 24; (3) were the subject of License Condition 2.C.(11) in the Unit 1 Iow power license
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regarcing piping and pipe supports; and (4) were the subject of various allega-tions addressed in SSER 21. A total of 414 items were identified as resulting from or relating to the Unit i design verification effort.
They are listed in three tables of the various IRP submittals. (A few items were double counted as a result of the multiple sources above.) Some items-relate to a singular concern, others are broad in scope, for example the Unit i license conditions for pipe and support analysis.
The purpose of the IRP was to review all issues that had been identified during the Unit 1 design verification process with regard to their applicability.to Unit 2, to monitor and assure the resolution and completion of the applicable Unit 2 items, and to document the entire review and resolution process. The 02/27/85 2-2 DIABLO CANYON SSER 29 SEC 2 q
actual technical evaluation and resolution was achieved within the appropriate engineering discipline groups.
The IRP applied the following five-step process to identify, monitor and docu-ment the resolution of the items for Unit 2.
Step 1 The IRP Director made an initial assessment of each item for Unit 2 applicabil-ity and need for further action based on the following criteria:
1.
The item applies only to Unit 1.
No further IRP or engineering effort is required.
2.
The item was not an error or deviation on Unit 1 and therefore is incon-sequential. No further IRP or engineering effort is required.
3.
The item applies to a portion of the plant which is common to both units.
The resolution, including appropriate modification, was already implemented
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as part of the Unit 1 verification program. No further IRP, or engineering effort is required.
4 Unit I and Unit 2 are identical with respect to the subject item.
The Unit I resolution is equally applicable to Unit 2.
No further IRP or engineering effort is required.
5.
An already ongoing Unit 2 review / reanalysis by one of the engineering groups encompasses the concern raised by the Unit 1 item. Therefore, no further IRP or engineering effort with respect to the specific item is required.
The IRP identification and closeout of each item by one of the above criteria was reviewed and approved by the Unit 2 Project Engineer. Appropri-ate documentation for the closecut of each item was entered into the IRP files. Any item involving a physical modification for Unit I was not closed out by these criteria, except as under Criterion 3 above.
02/27/85 2-3 DIABLO CANYON SSER 29 SEC 2
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Fu-ther detailed review was initiated for the items that were not identi-fied and closed out during the initial ' review by one of the above criteria in Step 1.
Unit 1 background information was collected in a Unit 2 IRP re/iew package which was assigned for resolution to a Lead Review Entity (LRE), normally an engineering group leader. One review package could include more than one Unit 1 item for Unit 2 resolution. The items were combined in 195 review packages. The LRE performed a detailed review taking into consideration Unit 1/ Unit 2 uesign similarities and diffe-rences.
Individuals involved in the Unit I resolution of the item and in the original design and analysis for Unit 2 participated in the evaluation.
Step 3 Based on.the detailed review in Step 2, the LRE determined'the applicability of the Unit i resolution for each item:
(a) Unit 1 and Unit 2 are essentially identical with respect to the subject item: The LRE, in consultation with other engineering disciplines as appropriate, initiated implementation of the Unit I resolution to meet the Unit 2 design requirements, or documented the closecut in the IRP files in accordance with Step 1, Criterion 4.
The implemen-tation of the' Unit 1 resolution was monitored by the IRP.
(b) Unit 1 and Unit 2 were not identical with respect to the subject item:
The resolution of the subject item as developed under Step 2 is dif-ferent for both units. The LRE ensured that the resolution is consis-tent with license requirements and Unit 2 specific requirements.
If the resolution involved a change to plant operating procecures the effort is coordinated with Nuclear Plant Operations (NPO). All considerations and evaluations were documented and included in the review package.
02/27/85 2-4 DIABLO CANYON SSER 29 SEC 2
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3 Step 4 For all items tha't required a physical modification to Unit 2 normal engineering procedures were applied. Design change notices (DCN) and other appropriate engineering documents were issued. Completion of notification f
was verified, reviewed, and documented in accordance with established project procedures.
Step 5 l
The final implementation of the Unit 2 resolution was documented in the IRP files with an identificat' ion of all appropriate documents. The Unit 2
-effort, including modification as necessary, was considered ccmplete after the Unit 2 Project Engineer signed a completion sheet.
l 2.3 Internal Review program Results PG&E evaluated the 414 items that resulted from the Unit I design verification effort. As a result of the review under' Step 1, it was determined that 145 of these items did not require any further evaluation or modification in accordance with the criteria.
In Step 2 a total of 195 IRP review packages were assembled for the other 296 items. During further detailed evaluation under Step 3 by the Lead Review Entity and the cognizant engineering groups, it was determined that fer 184 of these items the Unit 2 resolution is the same as for-Unit 1, for 31 items anotner Unit 2 review activity, already in progress, encompassed the Unit 1 item, and for three items it was found that a different resolution was required.
In summary No further engineering analysis required, based 145 items on initial review Same resolution for both Units based on detailed 134 items review Unit 2 resolution encompassed by other activity 80 items Different resolution for Unit-2 4 items 02/27/S5 2-5 DIABLO CANYON SSER 29 SEC 2
The following are the three items with a different resolution for Unit I and Unit 2:
1 1.
Seismic qualification of main annunciator typewriter (IRP package-2-1049)'
2.
MELB protection shields for valves in AFW system (IRP package 2-8014) 3.
Jet impingement as result of HELB (IRP package 2-8049)
Details of the staff review of these items are presented in Section 10 of this report.
A physical modification for Unit 2 was required for 57 items, some of which had already been completed as a result of the Unit 1 design verification in areas common to both units. However, this breakdown can be misleading because the scope and magnitude-of the modification required for the individual _ items vary greatly. For example, the items for the piping and support effort related to the evaluation of the Unit 1 low power license conditions encompass many modifications. These items were monitored by the IRP, while the actual analyses and modifications were accomplished under PG&E's Piping and P'ipe Supports Review Program as discussed in Section 1.
On this. basis, these items were considered closed out with respect to the IRP.
PG&E submitted the IRP Final Report on November 2,1984 and further information by letters dated December 7,1984 and February 21,1985 (Ref. 28).
PG&E has stated that the engineering resolution for all items has been completed and necessary modifications are either complete or well underwa'y.
Inlthe Final Report PG&E has committed to complete all modifications prior to Unit 2 fuel loading.
2.4 Staff Evaluation and Conclusion The Diablo Canyon design verification program was required in late 1981 by-the Commission and by the Office of Nuclear Reactor Regulation specifically for-Diablo Canyon Unit 1 (Ref.16, Ref.17). While there was no expressed intent f
by the staff at that time to also require such a program for Diablo Canyon Unit 2 the staff 02/27/85 2-6 DIABLO CANYON SSER 29 SEC 2 l
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had always considered that the findings from the Unit 1 program would have to be evaluated for their applicability to Unit 2 and appropriate action would be taken. This intent was expressed by the staff during the hearings before the Atomic Safety and Licensing Appeal Board in late 1983.
In late 1981 when the Unit I low power license was issued (Ref. 30), that unit had been upgraded for the Hosgri event with respect to analysis, design and modifications. PG&E had devoted its resources to complete Unit I during 1981.
Therefore, the Unit 2 upgrade for the Hosgri event was lagging behind in analysis design and modifications by approximately one year. The subsequent Unit 1 design verification effort was therefore a verification of work that had been completed; During the design verification pro' gram for Unit 1 PG&E concentrated its efforts on this unit, in particular during 1982 and 1953. When the Unit 2 activities began to pick up in 1983, these efforts were not a verification of work that already had been completed, but were largely the initial Unit 2 Hosgri upgrade efforts. The results of the Unit I design verification program were considered at the same time.
Therefore, while the design verification for Unit I was con-centrated on specific issues, the same effort for Unit 2 more closely followed' normal engineering procedures.
As described earlier, the Diablo Canyon Unit 1 and Unit 2 are essentially of the same design.
Some safety-related structures are shared by both units, others are nearly identical. The units are a mirror image of each other which resulted in different layouts, primarily in piping systems and electrical raceways.
There are also some differences in the nuclear steam supply system and the balance of plant fluid systems due to the 5 percent difference in the core thermal power.
Because of this similarity the same design criteria, methodology and design process was applied to both units as described in the common Final Safety Analysis Report (FSAR) for both units and in the staff's Safety Evaluation Report (SER),
including Supplements No. I through 17.
~
The Independent Design Verification Program (IOVP) as required by the NRC for Unit 1 and the expansion of that program through the PG&E Interal Technical Program (ITP) for Unit 1 verified the design adequacy for Diablo Canyon Unit 1.
This included extensive reanalysis in particular with respect to seismic design considerations and in the area of piping and supports. Modifications were made 02/27/85 2-7 DIABLO CANYON SSER 29 SEC 2 m.
6
^
as necessary.
Because of the similarity between Units 1 and 2, the staff did not find it necessary to also require an independent design verification program for Diablo Canyon Unit 2.
This would have been a redundant effort. The staff did consider it necessary, however, that the results of the Unit 1 IDVP and ITP be equally applied to Unit 2.
At a meeting on September 13, 1985 with the NRC technical staff and management PG&E described the Internal Review Program (IRP) and discussed the status of the program. Most of the NRC personnel who participated in the review and evaluation of the Unit I design verification program were present at this meeting and likewise participated in the review and evaluation of the IRP and
~
its implementation.
The staff performed a number of audits and inspections at the PG&E Diabic Canyon Project offices in San Francisco and at the Diablo Canyon site (Ref. 32, 33, 34). The purpose of this effort was to verify the program-matic implementation of the IRP and to perform detailed technical evaluations of selected issues that had been raised during the Unit 1 design verification.
However, the staff effort was not limited in this regard as discussed later.
The staff audited the IRP to verify appropriate documentation and close-out of IRP packages. This included verification that the necessary decision, in accor-dance with the IRP process as discussed earlier, had been documented. The staff reviewed applicable DCP and IRP internal procedures and discussed them with IRP personnel.
The detailed technical audits included a review of calculation packages, drawings, computer codes, including inputs and outputs, and all other pertinent background and cocumentation. During these audits the staff discussed selected areas with PG&E, which in some cases led to specific requests for additionalinformation(Ref.35,36)(7, M Yhe staff audits and /
inspections with regard to the programmatic and technical aspects of the IRP was conducted by NRC staff from the Office of Nuclear Reactor Regulation (NRR)-
and from Region I and by NRC consultants.
In addition, NRC Region V staff performed inspections with respect to the overall Diablo Canyon Unit 2 project activi.ies and in specific areas as requested by NRR.
The staff reviewed and evaluated the technical aspects of issues in all areas within the scope of the Unit 1 design verification effect. The details are presented in Sections 3 through 11 and include the evaluation of the three items i
02/2S/85 2-8 DIABLO CANYCN SSER 29 SEC 2
.o 4'
than for Unit 1 (see Section 10). The detailed technical review and evaluation of the IRP by the staff was concentrated in two areas, (1) seismic evaluation of civil structures and (2) piping systems and pipe supports.
The staff evaluation of the seismic design aspects of civil structures is pre-sented in Sections 3 through 8.
It consisted of a detailed evaluation of (1) the applicability and resolution for Unit 2 of all concerns that were raised by the staff during the Unit I design verification; (2) the containment annulus steel structure; (3) certain aspects of the turbine building unique to Unit 2; (4) raceways to support Class 1A electrical cables a~nd wires; (5) buried elec-trical conduits that carry electric power and control cables from the turbine building to the intake structure; and (6) the pipeway steel frame structure attached to the outside of the containment, auxiliary building and turbine building which supports the main steam and feedwater lines. The staff evalua-tion was performed to verify that the applicable design criteria for Unit 2 have been met. The design procedures and methodologies for the above areas are essentially the same for both units.
As part of its design verification for Unit 1 the staff and its ' consultant (Brookhaven National Laboratory - BNL) had developed a separate model to evalu-ate the seismic design aspects of the Unit 1 annulus steel structure. Such a model was not developed for Unit 2.
However, the staff and its consultants did perform independent and evaluations of selected structural members in this regard to account for the Unit 1/ Unit 2 differences. The details are' discussed in Section 4.
The staff effort for seismic design aspects was expanded during the Unit 2 evaluation with respect to buried electrical conduits and the pipeway steel support structure as discussed in Sections 7 and 8 of this report. During the Unit 1 design verification effort the IDVP had verified the design of buried pipes.
For Unit 2 the staff selected buried conduits for its independent review.
The staff determined that the documentation for the design and construction of j
buried conduits was inadequate. PG&E has since initiated necessary documenta-l tion and verification for both units, including the opening of pull boxes for both units to verify adequate slack. The staff concludes that this concern i
is resolved.
03/01/85 2-9 DIABLO CANYON SSER 29 SEC 2 4
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The staff's evaluation of the Unit 2 pipeway steel support structure (designed by PG&E) was the result of a recent allegation regarding the seismic model for this structure. The staff determined that the concern relates to certain ana-lytical. details of member connections.in the' structure. The staff has not fully completed its evaluation of this seismic design aspect. Based on its evaluation of this issue thus far the staff expects that physical modifications, if any, will be minor. Furthermore, the pipeway structure is located outside containment.
Therefore, the staff has concluded that this issue need not be fully resolved prior to fuel loading or low power operation.
The staff requires that the com-plete resolution, including any necessary modification, be completed prior to i
full power operation. The staff is currently evaluating the applicability of this concern for the pipeway structure of Unit I which was designed by Westinghouse.
During the staff review of the Unit 2 aspects of the turbine building PG&E identified to the staff an issue regarding the analysis of a turbine building slab which did not account for the proper shear transfer mechanism. The staff review of this matter is still ongoing. The staff, based on its evaluation thus far, expects that the adequacy of the slab will be demonstrated analytically and no physical modification will be required.
The staff requires that this matter be resolved prior to full power operation.
The cetailed technical review and evaluation of the design and analysis of piping systems and pipe supports for Unit 2 was similar to that which was performed for Unit 1.
Further information is provided in Section 9 of this report and complete details are presented separately in SSER 30.
The staff effort included an evaluation of the resolution of (1) issues that had been raised during the Unit 1 design verification, (2) actions resulting from a license condition on this subject in the Unit I low power license, and (3) the Unit 2 appitcability and resolution of allegations related to piping and supports.
The NRC team, including consultants, audited in excess of 60 IRP packages and approximately 100 pipinc and pipe support calculations.
In addition, the staff met with a confidential alleger wno identified specific concerns regarding pipe supports.
These concerns and other anonymous allegations also were included in the staff's review effort (Ref. 38 and Ref. 39). The NRC team did conclude that the Unit 2 piping systems and pipe suoports meet the applicable design criteria.
03/01/S5 2-10 OIABLO CANYON SSER 29 SEC 2
Basedonitireviewandevaluation,includingauditsandinspections,thestaff finds that Ehe IRP was an effective program in identifying, documenting and re-solving the Unit I design verification issues with regard to Unit 2.
The staff also finds that the extensive use of the same technical and management staff for both units provided for consistency in the design verification program and the IRP. The staff concludes that the multiple review and evaluation of issues in the program provided for proper resolution of each issue. This included an initial assessment by the IRP Director, detailed evaluation by a technical group supervisor, consultation with appropriate other engineering disciplines and final review and approval of all resolutions by the Unit 2 Project Engineer.
4 02/27/85 2-11 DIABLO CANYON SSER 29 SEC 2
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3 SEISMIC EVALUATION OF CIVIL STRUCTURES
3.1 INTRODUCTION
The major structures at the Diablo Canyon Nuclear Plant site are the auxiliary building, intake structure, outdoor water storage tanks, and buried diesel oil tanks. They are common to both Units 1 and 2.
Th staff evaluation of the c ue i t seismic design of these structures has been report in SSERs 18, 19, 20, and A
A 24,ducia; the Ur't 1 r.ic...
The Unit 2 containment structure is practically identical to that of Unit I except that the annulus structure inside the con-tainment has some structural aspects and modifications which are unique to Unit 2.
ThestaffperformedseveralauditsofthePG&Ecalculationsfq;p'the t
Unit 2 annulus structure. The staff evaluation is reported in Section 4.0.
The Unit 2 turbine building is similar to that of Unit 1, but there are certain differences that impac,t the seismic response characteristics.
.The_ staff audited PG&E calculations for the Unit 2 turbine building. The staff evaluation is reported in Section 5.0.
PG&E performed seismic analyses for all Unit 2 raceways and supports using the same approach employed in the Unit 1 analysis. The weld strength of superstrut construction estaclished in the testing program during the Unit I review is also applicable to Unit 2 since the samples used in the testing program were taken from both units.
The staff audited the PG&E calculations as well as the design aids referenced in these calculations. The staff evaluation of Unit 2 raceways and supports is reported in Section 6.0.
Electrical cables between the intake structure and the turbine building are housed in conduits buried in the ground.
Some cables are in plastic pipe con-duits with a concrete cover on top of the pipes and others are encased in con-ThestaffauditedPG&E'sdocumentationoh;thedesignand crete duct banks.
(
construction of the buried conduits. The staff evaluation is reported in Section 7.0.
02/26/55 3-1 DIABLO CANYON SSER 29 SEC 3
-o.
The Unit 2 pipeway structure is similar to that of Unit 1.
The Unit 1 pipeway structure was analyzed and designed for the seismic events by Westinghouse Corporation. However, the Unit 2 pipeway structure was analyzed and designed by PG&E for the Hosgri earthquake and for DE and DDE earthquake by Nuclear Ser-vices Corporation. Therefore, different techniques were employed in the analy-ses for the pipeway structures in-Unit 1 and Unit 2.
The staff audited the PG&E (;n;f2 OnJZ calculations for the Hosgri earthquake and will review the, analyses for the DE x
and DDE earthquakes. The staff review and evaluation of the analyses for the D andDDhisongoing. The results will be provided in a future report. The X
staffevaluationontheanalysysperformedfortheHosgriearthquakeisreported in Section 8.0.
Evaluation 94 the analyses for the OE and DDE earthquakes will e
be reported _k e f u : 50 (qfer.
3.2 NRC taff] Concerns from Unit 1 v
As a result-of its review and evaluation fo the Unit 1 design verification effort the staff had identified 20 concerns related to the seismic design of civil structures. These concerns, called Open Items (OI),were identified in SSER 18 3
and further discussed and resolved in SSERs 19, 20, and 24. The staff has assessed the applicability of these items with respect to Unit 2 and evaluated the PG&E resolution as described in Table 2 of the various IRP submittals by PG&E (Ref. 28).
During the Unit 2 audits and inspections the staff also re-viewed and evaluated various aspects raised by th/se items, as appropriate.
(
A resolution was reached for each item in accordance with the IRP process dis L [#,Q'.
beh us ds er cussed in Section 2, either cy toentifying the commonality of an item for the-7 3
same or different approach where the units are different.
A 6c. sis f
3.2.1 Resolution [(of Open Item 0 for Common Structures p F
StaffO concerns regarding the PG&E analysis for the auxiliary building were
^
raised in OI 5, 6, and 7.
Staffff concerns regarding the PG&E analyses for the e
intake structure and the buried diesel oil tanks were raised in 01 25 and 26.
respectively.
Since the auxiliary building, the intake structure, and the buried diesel oil tanks are common to both units, the resolutions reached for Unit 1 are equally applicable to Unit 2.
The previous Unit I resolutions were documented in SSER 19 and SSER 20.
02/26/S5 3-2 DIABLO CANYON SSER 29 SEC 3 l
a.
- Auxiliary Building Floor Slab Flexibility (OI 5) 1 For Unit dhe staff requested PG&E to assess the assumptions used in the auxil-N iaryguilding seismic analysis to determine floor slab flexibility. PG&E modeled X
the entire auxiliary building (Unit I and Unit 2) with a stick model, whose characteristics were based on the assumption that the floor slabs were rigid as compared to the walls (i.e., the floor slabs behave as a rigid body). This item was resolved for Unit I by generating a 3-D finite element computer model of the building (Calculation 30.23.1.2.2) and performing static load analyses.
The results of these analyses demonstrated that the floor slabs were rigid as compared to the walls. Since the auxiliary building'is common to both u' nits, the @2--
item is resolved for Unit 2.
A 4
ACI Code Justification for Auxiliary Building (OI-6)
For Unit 1 the staff requested PG&E to justify 'the use of the ACI [ ode for 3
evaluating the floor slabs and walls of the auxiliary building. The concrete fcorslabsandwallsoftheauxiliarybuilg,ing,we vagatedagainstcriteria presentedinAppendix2AofthePhaseIPG&Eufbportraf.herthanthe1963ACI fode. The 1963 ACI [ ode has no specific provision for shear walls but does per-x mit criteria to be used which are based on test data.
Test data were developed by PG&E and presented in Appendix 2A of the PG&E Phase I Final Report, Design Verification Program ((E]NA review of the Appendix 2A criteria indicated that it is conservative relative to the shear wall provisions in the 1977 ACI
[oce.
Tne Unit 1 stress evaluation was therefore accepted by the staff. Since the auxiliary building is common to both Units and the criteria of Appendix 2A was used in the evaluation of the entire structure, the @.m.-
item is resolved-for Unit 2.
4 Soil Spring Influence on Auxiliary Building (OI-7)
Fcr Unit I the staff requested PG&E to assess the soil spring influence in the seismic response of the auxiliary building. A portion of the auxiliary building is founded on soil which is soft enough'so that soil / structure interaction effects are included. Questions were raised by the staff during the Unit I re-view regarding the soil procerties used to generate the interaction spring 02/E6/35 3-3 DIABLO CANYON SSER 29 SEC 3
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parameters.
In response to this concern, PG&E performed a6a~riEton3 para-a meter study showing that the seismic response of the structure was not sensi-tive to spring constant variations spanning values corresponding to plausible variation in the soil properties. Since the auxiliary building is a continuous structure common to both gnits, the @ Y em is resolved for Unit'2.
/
Intake Structure Lateral Forces (OI-25)
For Unit 1 the staff requested PG&E to fully evaluate the total lateral forces, the total resistance to sliding and the factor of safety against sliding of the-untake structure. The staff reviewed IOVP reports ITR-40 and ITR-68 ('Rev. O and 1), discussed the issue with the IOVP, reviewed documentation by Harding, Lawson Associates (HLA). and performed an independent analyses of sliding.
The staff concluded in SSER-20 that the intake structure is safe against sliding.
Since the intake structure is common to both units, no further review of this
.e -
(cfr) item is required for Unit 2.
f
- Buried Diesel Fuel Oil Tank (0I-26)
For Unit 1 the staff requested PG&E to perform additional analyses of the buried diesel fuel oil tanks. Additional analyses were performed by PG&E and verification studies wre conducted by Brookhaven National Laboratory (BNL), the staff consultant. Based upon these additionWitudies, the staff concluded
~
that all safety issues associated with the tanks were satisfactorily resolved.
Since the curied diesel fuel oil tanks are common to both units no furt'her re-view of this @ item is required for Unit 2.
M 3.2.2 Resolution ($f OpeB Itemslby Same Methodology [
l w
The following open items related to the analytical techniques employed in the PG&E analyses or the design codes used in the PG&E design l Aled, w, s, 4, 8, 1 2 3 Ad, ci
- 13. 14, 15, 23, 24, and 31. AlthougSgnarate analyses were performed for A
Unit 2,thefemployedthesameapproach, methodology (hpaswasusedforthe X
Unit 1 analysis. The staff audited the calculations for Unit 2 and found that the structures involved in these @ N ms are sufficiently similar to those 02/26/s5 3-4 DIABLO CANYON SSER 29 SEC'3 j
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of Unit 1 to justify this approach. Therefore, the staff concludes that the Unit 1 resolutions are equally applicable to Unit 2.
Ji Free-Hand Averaging of Spectra (OI-1)
For Unit 1 the staff requested PG&E to confirm that the free-hand averaging procedures for spectra used in the containment annulus structure are proper.
Fre4[handaveragingproceduresforsmoothingthefloorresponsespectrawere x
used in the low frequency range in which no structural frequencias exist. PG&E used this method for seismic evaluation of both units. Justification for the method was provided by PG&E during the Unit I review and was accepted by the staff. Similar justification for Unit 2 annulus is, therefore, not necessaryr'since 3awot w+s the methodology applied to both units.
43
- 20 Hz Cutoff Frequencies (OI-2)
The 20 Hz cutoff criterion was used in the horizontal seismic evaluations of the Unit I containment annulus for the Hosgri earthquake. Based on this criterion, potential amplification of responses in the frequency range of 20-33 Hz was neglected. This does not properly reflect the Hosgri criteria and thus the staff requested PG&E to provide a justification for the 20 Hz cutoff criterion. Thisd((F7'temwaslatterresolvedfortheUnit1 annulus e
structure in SSER 19.
PG&E performed a study for the Unit 2 annulus structure similar to the Unit I study to justify the 20 Hz cutoff criterion. The same modeling technique anc analytical approach were employed in this study. The staff has reviewed the study and found no anomaly in the results..The staff thus concludes that the 20 Hz cutoff criterion does not significantly alter the floor response spectra generated from the Unit 2 annulus structure analysis.
Further details are provided in Section 4 of this report.
_ AISC Code for Penetrations in Containment Shell (0I-3)
For Unit 1 the staff requested PG&E to assess the applicability of the AISC and ASME (Section III) 'Icdes to the design of penetrations in the containment shell.
/
a he In SSER 18 the staff stated that the use of the AISC fode for containment pene-(
tration analysis shcuid be justified. The resolution of this q{pejritem was 02/2e/55 3-5 OIABLO CANYON SSER 29 SEC 3
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presented in SSER-19. The containment design was shown to satisfy both codes for Unit 1.
The Unit 1 and 2 containment designs are very similar in general, and in particular in the vicinity of the penetration. Therefore since the Unit 1 design was shown to satisfy the requi'rement of both the AISC and ASME (SectionIII)hdes,thestaffconsidersthisitemtoberesolvedforUnit2.
4 Stress Levels at Openings in Containment Shell (OI-4)
For Unit 1 the staff requested PG&E to assess the adequacy of tne containment equipment hatch opening structure.
The staff stated that local yielding of steel plates around the opening in the containment exterior shell should be justified. Based upon the staff review of this item, as well as the IDVP re-view of Unit 1 the staff considered the plate analysis acceptable for Unit 1.
The staff reviewed the plate analysis during the audits for Unit 2.
The staff found that the analytical approach is identical to that for Unit I and the re-suits were within the code allowables. Thus, the staff concludes that the Unit 2 analysis is acceptable.
f.._
Seismic Input Motion to Fuel Handling Building (OI-8)
Me For Unit 1 the staff requested PG&E to document the use of auxiliary building A
motions as input to the fuel handling building seismic analysis. The fuel
&cvl handling building (FHB) is a steel structure,, : ":f r the auxiliary building, e
and is inciucec ir. a sticis model in the dynamic model of the auxiliary building.
Both the Unit 1 a 2 portions of the FHB are included in this stick model.
After the seismic analysis of the auxiliary building was completed, the motion at the base of the FHB stick was then used as input to the base of a-detailed model of the FHB. Two different models of the FHB were used. One model represented the five end bays (in either the Unit 1 or Unit 2 sides) and the other would represent the five middle bays. Torsional effects were accoun-ted for by connecting the cMumns of the FHB to a common base point which is located off the center of the FHB. Tne input motion is applied to this base point. The first model, representing the end bays, is therefore subjected to a
.i larger torsional component of loading than the second model since the columns J
of the first model are further from the center of the building than the column 02/26/55 3-6 DIABLO CANYON SSER 29 SEC 3 w
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of the second model. The input motion to the-FHB was reviewed during.the Unit 1 i
review and found to be acceptable. Since the FHB is supported by. the auxili-7 ary building which is common to both Units, this item is resolved for Unit 2.
+- Reduction of Degrees of Freedom for Fuel Handling Building (OI-9)
For Unit 1 the staff requested PG&E to justify the reduction in the number of dynamic' degrees of freedom used in the fuel handling building seismic analysis.
The; original fuel handling building (FHB) model retained 20 to 30 degrees of freedom for the dynamic analysis.
PG&E performed more detailed analyses which used V Cubstantially more degrees of, freedom.for the dynamic' analyses.
The x
The same model s[( see al so @y..., s.e.
analyses were judged to be adequate.
OI 8 above),
were used to represent both the Unit 1 and Unit 2 section of the FHBt Therefore, the same analysis is applicable to both units.
wel.e e.f %dil Or Turi:,a 3,;ld;y Roof Sys% (CE -l3L ForUnit*/thestaffrequested-PG&Etoconfirmthatthenumberof'degreesof~
freedom for the nodes above elevation 140 feet of all models for the roof truss is consistent with the turbine building response.
Four different verti-t cal models were used to model different areas of the Unit turbine building.
/
y The roof truss was modeled somewhat differently depending on whether the model was located near the middle section of the building rather than near the ends.
This was done to account for the fact that the' variation in the response betweenadjacentbaysisgreaterneartheendofthestructurethanitisnear l'
the middle. The differences between the models were justified since the calcu-lated respense for eacn of the models were shown to be consistent with the degrees of freedom retained in the middle. Since the Unit 1 vertical model re-suits were applied directly to Unit 2 (see OI 12), this item is resolved for Unit 2.
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Modal Combinations by SRSS in Turbine Buil ing (0I-14) 7
-ne staff identified a concern @ecardinkfor Unit 1 the acce tability of-alter-native procedures for modal combinations by the SRSS method 'n the turbine build-ing seismic analysis.
The Phase I Final Report (Ref.
) contained the state-ment that alternative procedures of load combinations were being reviewed.
In fact, all members of both units were evaluated using the.SRSS method of modal 02/26/S5 3-7 DIABLO CANYON SSER 29 SEC 3 i
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e combination. Some of the critical members were also evaluated using the double
-algebraic sum method. The staff found this approach acceptable. This item has therefore been resolved for Unit 2.
^
Allowable Bolted Joint Loads for Turbine building (OI-15)
For Unit 1 the staff requested PG&E'to justify the use of the AISC Code, 8th Edition in determining allowable stresses for roof connections in the turbine building.
The allowable joint loads for typical turbine building roof connections were based upon the 5th Edition of tne AISC Code rather than the 7th Edition. The 8th Edition permitted higher bearing stresses than the 7th Edition of the Code.
PG&E Calculation 65-T-004 required the bolts to be pretensioned to 70 percent of the yield strength thereby increasing the overall capacity of the joint.
The allowable joint ends in the 7th Edition neglect any frictional forces which occur between the plate of the joint.
Since the bolts are tightened frictional forces are generated between the plates. PG&E demonstrated that the addition of these frictional forces to the 7th Edition would result in a larger capacity than permitted in the 8th Edition. The application of the provisions of the 8th edition of the AISC Code was therefore accepted for Unit 1.
Since the joints are similar in Unit 2, this item is therefore resolved for Unit 2.
Cable Tra LQualification (0I-23) p
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02/26/85 3-8 DIABL' CANYON SSER 29 SEC 3 l
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4 COMBINATION OF THREE EARTHQUAKE COMPONENTS (0I-31)
For Unit 1 the staff. requested PG&E to clar fy ethods of combining directional
\\ftL i Dut l responses in the turbine building. The three earth" quake $ components were com-bined by the PG&E for the turbine-building by adding the full value of-the larg-est component to 40 percent of each of the other two components. Based on a study by PG&E this methud results in conservative values relative to SRSS com-bi ation except when two components are equal and the third component is zero.
In t us case the SRSS combination is about I percent higher than the method A
A used by PG&E. The staff accepted this method of combination for the turbine
-building during the review for Unit 1.
The same method was used for the eval-uation of the Unit 2 turbine building.
Therefore this item is resolved for Unit 2.
16 Oz$:d c
w Iwr F l Three cpen items (0I-10. @l and(0)12) were reviewed and evaluated idcatical during the staff's aucit of tne design and analysis of the turbine building as o h" discussed in-Section 5.
02/26/s5 3-9 DIABLO CANYON SSER 29 SEC 3
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Related to Turbine Building [ g,) f tp 3 9 Resolution ((bpenI 3.2.3 il V
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Load Combinations for Turbine Building (OI-10)
For Unit I the staff requested PG&E to justify the load-combination equation used to seismically qualify the. turbine building because the strength require-ment of the structural members in the turbine building was not based on the proper combination f dead, live, and seismic loads. This concern was based on Nat
?.4g) the Phase IA/lpor(t hich did not indicate that such load combinations were used.
During the staff audits member strength calculation for the Unit 2 turbine build-ing were reviewed. The ultimate capacity of the members were compared with the demand for the load combination of dead, live and seismic loads (Hosgri effects).
In particular, the 'cliewing calc;1aticr.s were reviewed:
65-T-235 shear. walls 65-T-111 columns 65-T-311 buttresses The comparison showed that the capacity of all members exceeded the demand.-
The staff considers this item resolved for Unit 2.
g Turbine Building Roof Truss Model (0I-11)
For Unit 1 the staff requested PGLE to provide further information regarding the methoc of moceling the roof truss by two generalized uniaxial members and obtaining individual truss member response from the unia.xial member model.
The. truss system used for the roof truss of Unit I was modeled as equivalent uniaxial members per the PG&E Calculation 64-T-275. This calculation demon-strated that the stiffness of the equivalent member was_close to that for the actual truss system. Truss member loads are derived fr6m the end displacements of-the equivalent member.
PG&E Calculation 64-T-291 showed that member loads cerived t r,ing the equivalent truss member are close to member loads fourd by using the actual truss configuratior. This was.shcwn by performing static analysesoftwomodelsg]oneusingtheactualtrussconfigurationandthe V
second using the equivalent truss member and end displacements for generating the internal member forces. The loads generated from the equivalent model 02/26/35 3-10 DIABLO CANYON SSER 29 SEC 3
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.t were shown to be close to but larger than.the loads generated using-the actual truss model.
'The roof. truss system.for Unit 2 is.similar to that.for Unit 1.
The same model l
was used for both upits. The staff reviewed the physical differences between the Unit [1 and 2 roof trusses and concluded that the difference were of local 7
A-4 4
. nature and would have minimweffect on the stiffness of the equivalent member.
A.
The. equivalent model of the Unit 2 turbine building roof truss is.therefore acceptable.
o Vertical Seismic Model-of Turbine Building (OI-12)
Fcr Ur.i. I the.staf f requeste:: PG1E justify the different models used in the vertical seismic analysis for the turbine building.
Four vertical models were used to evaluate the seismic response of the Unit 1.turb ne building. Questions wereraisedregardingthecouplingbetweenthesemodels.houplingeffectbetween
/
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the four models was shown to be not significant and the Unit I analysis was accepted.
The turbine building vertical analysis was not redone for Unit 2.
Calculation 64-T-349 evaluated the mass and stiffness differences between-Units 1 and 2.
Based on the local nature and small magnitude.of the differ-ences the Unit 1 vertical analysis is judged to be applicable to Unit 2.
The small differences between Units 1 and 2 were used to provide additional broaden-ing of the Unit 1 spectra to make them applicable to Unit 2.
The broadening was done to include the local frequency of the Unit 2 structure. This broaden-ing of the response spectra is judged to be adequate to account for the smail differencesbetweentheUnit1andUnit2[urbine uildings.
g 02/26/85 3-11 DIABLO CANYON SSER 29 SEC 3
I a
4.0 CONTAINMENT ANNULUS STRUCTURE 4.1 Introductio'n The containment annulus structure is a space-frame steel structure located in the annulus area formed between the crane wall and the containment wall in each unit.
Its function is primarily to provide supports to the miscellaneous pip-ing and equipment in this area.
In both units the general arrangement of the annulus structure consists of four floors between the base (elevation. 91 feet) and elevation 140 feet. The top floor (i.e., elevation 140 feet) is a reinforced concrete slab with some openings to provide communication with the lower floor.
The other floors are open structural steel framing to support the piping systems and other equipment. Only the floor at elevation 117 feet is covered with steel grating. The annulus is a seismic Category I structure and is designed for all three earthquakes DE, DDE and Hosgri event. The details of the PG&E reevalua-tion on the Unit 1 annulus structure are given in DCP Phase I Final Report (Ref. 40). The staff's evaluation of the Unit 1 annulus structure, which included the review of the DCP Phase I Final Report and an independent check by the staff's consultant, is contained in SSER 18 (Ref. 4). A description of the Unit 2 annulus structure evaluation is given in the following sections.
4.1.1 rj ts_1 Ard_2_
During a meeting held in Washington, DC on September 13, 1984, the staff re-quested the Pacific Gas and Electric Company (PG&E) to provide a set of drawings of the annulus structures for both units (Ref. 29). The objective was to draw conclusions regarding the similarity and differences between the two structures.
The following drawings were audited:
1.
Containment Annulus Structure - Additional Modifications - El. 140',
Unit 2 - 471099, Rev. 1 02/27/85 4-1 DIABLO CANYON SSER 29 SEC 4
~:
2.
Containment Annulus Structure - Additional Modifications - El 117',
Unit 2 - 471098, Rev. 1 3.
Containment Annulus Structure - Additional Modifications - El 106',
Unit - 471097, Rev. 1 4.
Containment Annulus Structure - Additional Modifications - El 101',
Unit 2 - 471096, Rev. 4 5.
Annulus Frame - EL-140' - Containment Structure (As-Built), - Unit 1 -
469344, Rev. 2 6.
Annulus Frame - EL-117' - Containment Structure (As-Built), - Unit 1 -
469343, Rev. 2 7.
Annulus Frame - EL-106' - Containment Structure-( As-Built), --Unit 1 -
469342, Rev. 2 8.
Annulus Frame - EL-101' - Containment Structure (As-Built), - Unit 1 -
469341, Rev. 3 From a comparison of the plan views of the platforms at different elevations the staff concludes that the general arrangement of the outside columns is similar' but not identical for both units. The structural details between the column lines are not the same.
For example, in comparing the plans at elevation 117-feet the Unit 2 configuration has more bracing members than Unit 1.
Further-more, additional columns between the major column lines are found in the Unit 2 annulus structure.
In addition to the above drawings, PG&E also supplied the staff with representative sketches of some of the dynamic models used in the seismic evaluations of the annulus structures for both units. These were:
1.
Dynamic Analysis - Racial Frame 1, Unit 2 - 1032C-1. Rev. 0 2.
Dynamic Analysis - Radial Frame 2, Unit 2 - 1033C-1, Rev. 0 02/27/55 4-2 DIABLO CANYON SSER 29 SEC 4 u ------ ------- -------- Z1 ^ 11-- l~lT l i-------- ----- l- ----- - T -- -- - - --------- - - - - - - - -
1---------------------
- ~ -..
a =:
= =. -
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e 3.
Dynamic Annalysis - Radial Frame 2.5, Unit 2 - 1034C-1, Rev. 0 4.
Dynamic Analysis - Radial Frame 3, Unit 2 - 1035C-1, Rev. 0 5..
Dynamic Analysis - Radial Frame 3.5, Unit 2 - 1036C-1, Rev. O L
S 6.
Dynamic Analysis - Radial Frame 4, Unit 2 - 1037C-1, Rev. 0 7.
Containment Annulus - Dynamic Model (DC) - Frame 1 - 625C-1 4
8.
Vertical Seismic Analysis for Annulus - Frame 2 - (DC #1) - 626C-1 f
9.
DC #1. Annulus Area Revised Radial Frame Model for Frame 2.5, Rev.' 627C-1 1
10.
DCP - Containment Time History Analysis Frame #3 - 628C-1
,12. DC #1 - Seismic Analysis Annulus Area Column Line 4 - 630C-1 t
The analysis procedures used in both units for the seismic evaluation are the same although the models are somewhat different.
By comparing these models it was determined that the Unit 1 models incorporate more single mass oscillators than those of Unit 2.
i 4.1.2 U i R anA Unit 2 Modificat g
,A During the evaluation of Unit 1, PG&E made several modifications to stiffen the annulus structure in order to reduce the response of'the floors. Based on the experience gained in the reevaluation of Unit 1, PG&E modified the Unit 2 annulus structure for the same purpose.
In general, the modifications in both units intiude: a) strengthening of members'and connections. b) addition of beams and bracing members, and c) addition of columns. These modifications were incorporated in the seismic evaluations for both horizontal and vertical directions. The additional bracing members were primarily inserted to reduce horizontal amplifications.
Strengthening of members and connections as well as l
02/27/S5 4-3 DIABLO CANYON SSER 29 SEC 4
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n
1 the use of additional columns, especially at 'the lower platforms, produced a re-duction in vertical amplifications. While the general philosophy behind these modifications is the same in both units, the distinct difference is that for
~
Unit 2 the reduction in vertical amplifications was achieved by using more columns than by strengthening of members.
It is realized that the additional columns in Unit 2 are more effective in increasing the vertical frequencies.
Furthermore, Unit 2 modifications involve more bracing members than Unit 1, par-ticularly, for the platforms at elevation 117 feet. The above modifications were verified during the audits conducted by the staff in October-November of 1984 as well as during a staff field walkdown in Unit 2 on October 4,1984 cry s'
(Ref. 32).
4.2 Sccee o' Review Subsequent to the NRC meeting with PG&E on Unit 2 on September 13, 1984, the staff and its consultants conducted two audits on the Unit 2 PG&E evaluations of the annulus structure. The first audit was conducted during October 1-5, 1984 and the second during November 7-9, 1984 (Ref. 32 & 33). After the first audit, the staff and consultants met with PG&E on October 18, 1984 (Ref. 32).
Additionally on October 4, 1984 the staff and consultants visited the plant site and conducted a walkdown in Unit 2.
During these audits and visits, the major items reviewed with regards to the Unit 2 annulus structure were:
a.
Ve-tical seismic analysis b.
Hori: ental seismic analysis c.
20 H: cutoff justification d.
Connection evaluations e.
Member evaluations f.
Torsional evaluations The staff reviewed the seismic analyses performed by PG&E in both horizontal and ve-tical directions to assess the adequa'cy of the models and the floor response spectra.
The review concentrated on verifying that the staff's con-cerns expressed in SSER 18 and results of the previous investigations from the Unit I annulus structure were considered in the Unit 2 evaluation. A sample 1
02/27/i5 4-4 DIABLO CANYON SSER 29 SEC 4
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of the calculations for members and connections were reviewed for compliance with the FSAR requirements. Details pertaining to the review of the above items follow.
4.2.1 Vstrtical Jeismic _Evaluatio2 The-same analysis procedures used for the Unit 1 annulus structure were followed by PG&E'for the analyses of Unit 2 for the Hosgri earthquake although the parti-cular dynamic models were different. The following sources were reviewed during the audits:
1.
Binder 1032C, File S2.17, Unit 2:
Centainment Arnulus. Vertical 2.
Calculation 1010C-6: Tangential Beam Frequency Analysis In general, for the Hosgri evaluation, the vertical floor response spectra were generated by using the modal superposition method. A set of 28 radial frame models were used which generally correspond to the locations at the column lines. These models are lumped parameter models which incorporate both the annulus structure and the crane wall. The tangential beams were also included by attaching single degree-of-freedom oscillators to the frame members to simu-late the dynamic characteristics of the tangential beams.
Individual beam fre-quency calculations were performed to obtain the dynamic properties of the tangential beams. The models and the analytical procedures employed by PG&E to develop vertical floor response spectra were similar.to those used for the Unit 1 annulus structure. The vertical spectra for Unit 2 annulus structure are contained in the Diablo Canyon Project internal Design Manual DCM-17.
The vertical responses for DE and DDE events were taken to be the same for both units. These responses were taken to be the 2/3 of the corresponding DE and DDE ero period ground acceleration (ZPA) in acc'ordance with the FSAR reautrements.
4.2.2 if I
i The seismic evaluations performed by PG&E for the Unit 2 annulus structure in the hori:ontal direction are based on the same procedures as those followed for 02/27/e5 4-5 DIABLO CANYCN SSER 29 SEC'4
-~ n_
Unit 1, although different analytical,models of the individual steel platforms
.were generated. Physical modifications were made in the plant to assure that the fundamental frequencies of each platforms and of each of its composite mem-bers are at least 20 Hz. These calculations'were done iteratively and the total number of iterations are 5,17, and 14 for.the. steel platforms at elevations 101, 106 and 117 feet, respectively. The Bechtel BSAP computer program was used for the frequency analyses of the floors. According to PG&E, the models which cor-respond to the final iteration incorporate the latest design changes-in the structure. The following calculations were reviewed for the frequency analysis:
1.
Diablo Canyon Unit 2, Annulus Structure, Calculations - 1002C-1 to 1002C-11 (EL 117')
2.
Diablo Canyon Unit 2, Annulus Structure, Calculations - 1003C-1 to 1003C-3 (EL 106')
3.
-Diablo Canyon Unit 2, Annulus Structure, Calculations - 1004C-1~to 1004C-3 (EL 101')
The above analytical studies by PG&E demonstrate that no frequencies below approximately 20 Hz exist in the horizontal direction for the three steel plat-forms. Based on the 20 Hz frequency cutoff criterion, the horizontal' floor i
~
spectra developed for the internal concrete structure are applicable to the annc'.us platforms.
Thus, subsystems and components attached to the annulus structure are designed for the same horizontal spectra derived for the internal concrete structure which are the same for both units. These spectra are docu-mented in' Design Manual DCM-17. As a result of these frequency studies, various Design Change Notices-(DCN) were issued to the field for physical modifications.
Such DCN's and the corresponding as-built drawings were reviewed during the audit of October 5-9, 1984.
PG&E has stated that all modifications issued for the Unit 2 annulus structure have been completed.
The DE and DDE floor spectra as documented in Design Manuals DCM-25 and DCM-30, i
respectively, were used for the qualifications of subsystems and components attached to the annulus structure in both units.
l 02/27/85 4-6 DIABLO CANYON SSER 29 SEC 4
~ ~.
n l
4.2.3 20_Hz Fritquency Cutoff _
,v During the reevaluation of the Unit 1 annulus structure the staff was concerned with the 20 Hz frequency cutoff criterion ((jh Ref. 4) because it did not reflect i
the Hosgri criteria. The staff requested PG&E to provide a justification for this criterion. Based on the information provided by PG&E this concern was resolved.for the Unit 1 annulus structure as discussed in SSER 19 (Ref. 5).
Although the general arrangements of the annulus structures in both units are
-similar, there are differences in structural details. Based on this fact, the staff requested PG&E to provide justification for the 20 Hz frequency cutoff criterion as applied specifically to the Unit 2 annulus structure.
The staff reviewed the following sources of information regarding the 20 Hz frequency cutoff criterion:
1.
Diablo Canyon Unit 2, Containment Annulus Structure, 20 Hz Frequency Study, Calculation No. 1140C-1.
2.
Calculation No. 900C-1, Rev. O, 20 Hz Frequency Study, Generation of Com-posite Damping, Multiple Support Method of Response Spectra.
3.
Calculation No. 900C-2, Rev. O, 20 Hz Frequency Study, Horizontal Model, Calculation of Column Stiffnesses at EL 106'.
4 Calculation No. 900-3, Rev. O, 20 Hz Frequency Study, Development of.
Horizontal Model, Interior Structure and Annulus Structure at EL 106'.
5.
Calculation No. 904C-1, Rev. O, 20 Hz Frequency Study - Horizontal Model, EW Earthquake Time-History Analysis (Newmark Hosgri), Output Accelerations at Annulus Pipe Supports for Lines K17-2314-3 and SG-3060-4.
6.
PG&E Nuclear Flant, Diabic Canyon Site Unit 2, W-3904 and W-5401 Hosgri Study.
A 02/27/55 4-7 DIABLO CANYON SSER 29 SEC 4
=
,a e
PG&E performed a parameter study to evaluate the effect of modes with frequen-cies greater than 20 Hz on the response of the annulus structure. The proce-dures used in this study were similar to those used for Unit 1.
It was shown that the modes with frequencies greater than 20 Hz did not have a significant effect on the floor response spectra.
R The staff requested PG&E to extend the 20 Hz cutoff frequency studies to include an evaluation of annulus members.
In~ order to demonstrate the effects on the.
annulus members, the increment in support loads due to the increased cutorf frequency was used to revise the utilization factors (the applied loads divided by the allowable loads) associated with the design of annulus members. The total stress in the members was then compared with the allowables per the FSAR.
ccc :ittent.
The old utilization factors were increased proportionally by the total increment in pipe support loads. This implies that the increments due to-seismic load from the 20 Hz frequency study were applied to all loads for the annulus member under consideration.
In actuality only the seismic stress of the member should have been increased. Obviously, this is a conservative approach.
Nevertheless, using this approach, the majority of the annulus steel members were found to satisfy the allowable stresses.
The calculated stresses in some members, however, did exceed the allowable stress.
For example, for beam #49, the new factor was found to be 1.018, i.e.,1.8% overstressed. These members could have remained qualified if only the increase in the seismic stresses was incorporated into the member stresses and then compute the utilization factor.
4.2.4 g ion g a_iuttcien_
The moment resisting plates of the radial beam connections are welded to the beam and column flanges while the moment resisting plates of the tangential beam connections are welded to the flange of the beam and to both web and flanges of the column. These connections are essentially rigid and, therefore, capable of transmitting axial forces, shears, and moments. The bracing member connections can transmit axial force and shear but not moment.
The strength check of these connections was carried out by using the Bechtel' computer program CONPAS. This program evaluates typical connections according to the AISC 1978 Code. As mentioned previously, the radial and tangential beams 02/27/85 4-8 DIABLO CANYON SSER 29 SEC 4
are connected to the column with moment connections while the beam to beam uses shear connections.
PG&E performed selective verification studies by which the results generated by COMPAS were compared with that obtained by hand computations.
The staff reviewed their verification and concluded that the computer generated results compared well with'the hand computation.
The bracing member strength was checked by combining axial force and bending moment about the z-axis. The staff reviewed detailed summary sheets No. 1006C-1 l
and 1006C-2. All design loads are much less than the allowable loads.
4.2.5 Mpmb x Evaluatlons _
The circular containment annulus structure consists of steel tangential beams.
radial beams and columns. The PG&E calculations reviewed by the staff show l
l that the strength checks of all the members were performed by the computer program GTSTRUDL.
For each member the static analysis results were combined I
with the separate seismic analysis results from the DE, DDE and Hosgri earthquakes.
The steel specification used internally by GTSTRUDL is the 1978 AISC specifi-cation. The allowable stress values for the DDE and Hosgri earthquakes re-flected the FSAR commitment.
The staff reviewed the following information in conjunction with the evaluation of annulus members:
A.
Tancertill B3,am Evaluatian:
A-1 Beam Location Drawing, EL 101', Quad 'B' (merbers B1266 and B1276)
A-2/A-3 Calculation No. 1005C-1, Rev. O, pgs ) esd Parameters for Code and Check by GTSTRUDL l
A-a Calculation No. 1005C-1. Rev. O, Rev. O, pg. 6, calculatfor to deter-mine the CODETOL parameter for GTSTRUDL (see page 2.2.5-130 of.
Attachment E)
A-5 Calculation No.1005C-1, Rev. O, p.11 (prep, sheet for GTSTRUDL l
input)
A-6 Beam Input - GTSTRUDL 02/14/55 4-9 OIABLO CANYON SSER 29 SEC 4
~
g
=
u.
_m.
A-7/A-12 Member forces for code check A-13/A-28 DE computer code check A-29/A-41 DDE computer code check-A-42/A-53 Hosgri computer code check B.
Radial Bnm f.p4uation:
c-x,e B-1 Computer. plot with beam identification, EL 101',; Quad 'B' B-2 Calculation No. 1005C-1, Rev. O, p. 10, parameter input code check B-3/8-8 Member forces for code check B-9/B-13 DE, DDE, and Hosgri code. check C. g Evalualions:
C-1/C-3 Calculation No. 1009C-1, Rev. O, pgs, 8, 9, 30 C-4/C-8 GTSTRUDL data for column R15 and Q15 4.2.6 Tancential and R_adial Beam Evaluations All tangential and radial beams are checked for. the combined' static and dynamic load effects. When torsional effects are included, separate checks were made as described below. A partial plan of the locations of the tangential and radial beams were included in Calculation No. 1005C-1. The calculations contained the criteria and input parameters for the GTSTRUDL input. Computer analysis and member strength check results were also performed by PG&E. The program GTSTRUDL generates the member forces. To perform the code checks.it uses these forces together with parameters such as allowable stresses, unsupported length, radius of gyrations, etc. to verify the adequacy of the beam based'on the' ISC.1978 specifications.
The staff reviewed the calculations and performed some independent checks.of the combined strest resulting from major forces of several members.
It Is con-cluded that the cede check procedure is satisfactory.
02/14/s5 4-10 DIABLO CANYON SSER 29 SEC 4 s.w..
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~p a-t 4.2.7 Colu3n ivalmation The exterior columns of the annulus frame are shown in Calculation No. 1009 C-1, Sheet No. 8.
The strength check of these columns was also carried out by the computer program GTSTRUDL. The member forces from the STRUDL analysis phase 'is utilized to -compute the design stresses.
The code check is based on the com-bined interaction formula given in the AISC specification. The staff reviewed the calculations and performed.some independent checks and found that the pro-cedure used for the strength check is satisfactory.
4.2.8 Torsion Evaluation Two catecories of torsional effect were evaluated by pG&E. One with rela-tively small torsion and the other with significant torsion.
Small torsion is defined as that with a magnitude less than 5 inch-kips.
In this case the combined stress of axial force, bending and shear is checked by GTSTRUDL and in general the stress is less than the allowable.
The torsional stress in this case is small enough so that there is no exceedance over the allowable.
Any torsion greater than 5 inch-kips is actually combined with those produced by the other force.
In this case more elaborate computat. ions were made. The staff reviewed these combination procedures and found it is satisfactory.
Beams with large torsional moments were checked in detail by hand computations.
Specifically, the following three beam computations were reviewed: H21290 at elevation 106 feet, A2216 at elevation 101 feet, and the beams F2232 and F3274 on column line 14 at elevation 106 feet. Beam H21290 is a W12X79 framed into a column at one end and attached to the reactor concrete wall at the other end.
PG&E assumed that the end conditions are torsionally fixed at the column and hinged at the wall. The distribution of the concentrated torsion at the interior of the span is performed according to a paper by J. G. Hotchkiss, (Ref 41). Warping effects were also considered. The results show that tne strength is satisfactory.
Beam A2216 is box shaped so that the warping effect due to torsion is negligible.
Although the distribution of the concentrated torsional moment to adjacent beams 02/14/85 4-11 OIABLO CANYON SSER 29 SEC 4
-l
--. - 4
m T
^
.. v i
I e
was based on an incorrect formulE, an independent computation done by the staff
~
indicates the strength is still ratisfactory.
In addition, PG&E calculation showed that the side plate would be extended to the face nf the column.
It is therefore concluded that the strength of the beam A2216 is satisfactory.
The same incorrect formula was us[ed for the analysis of the beams, F2282 and F3274. PG&E revised its calculations and computed the exact torsional moment based on continuity at Joint F388 instead of a fixed boundary condition as originally assumed. However, the effect due to unsymmetrical bending on the stress was neglected in the calculation. An independent-computation performed by the staff to include the unsymmetrical bending indicates that the stress of-the beams is satisfactory.
4.3 Findings and Conclusions Based on the staff's audit and detailed review of extensive information, the staff concludes for analysis and the design of the Unit 2 containment annulus structure the following:
a.
The analytical procedures used by PG&E for.the Unit 2 annulus structure are essentially the same as'those used for Unit 1.
Concerns raised during the Unit I annulus structure verification were considered and resolutions were implemented in Unit 2.. For the horizontal and vertical seismic evalu-atiors the models, methocologies and criteria followed by PG&E are found to be acceptable.
I b.
With respect to the 20 Hz frequency cutoff criterion, PG&E demonstrated in its studies that modes in' the range of 20-33 Hz do not have significant impact on the annulus stru:tural loads. Based en the margins associated with the design of the struc'ture, the annulus structure remains qualified for the additional amplifications. However, this conclusion does not imply generic applicability.
For other configurations the effect of the modes above 20 H: may not be the same, c.
The design of the individual members without torsions using the computer program GTSTRUDL is found to be satisfactory.
02/27/S5 4-12 DIABLO CANYON SSER 29 SEC 4 7.____-
d.
The strength check for the W-section beams with the relatively large tor-sion is found satisfactory. However the same strength check for a " box" shaped beam was incorrectly formulated for computing the distribution of the torsi.onal moments. This particular' beam and one other bean using the same formula are satisfactory when checked by a more appropriate formula.
f e
4 d
i i
i 1
0L 27/85 4-13 DIABLO CANYON SSER 29 SEC 4
l 4
5 TURBINE BUILDING 4
5.1 Introduction The turbine buildings for Units 1 and 2 form a single structure approximately 143 feet wide and 742 feet long, the long dimension of which runs in the N-S direction. The structure is a mill building type with the bents spanning in the short (E-W) dimension.
The columns begin at elevation 84 feet and terminate at elevation 193 feet, with steel roof trusses framing between the columns.
Interior steel frame and concrete diaphragm members are provided interior to the structure at elevations below elevation 140 feet. The primary E-W (short direction) stiffness of the structure is provided by concrete buttresses placed along the east and west walls of the structure, shear walls at the ends of the structure, and the top steel trusses. The N-S stiffness of the structure is developed from bracing in the steel superstructure and concrete shear walls in the east and west walls. An E-W expansion joint is provided about 344 feet from the south end of the building. This expansion joint structurally divides Unit 1 from Unit 2, with the Unit 2 half of the Turbine Building being south of Unit 1.
The differences between the two units are summarized as follows:
1.
The Unit 2 turbine building is two bays shorter than the Unit 1 building (about 54 feet).
l 2.
The On-Site Technical Support Center is located on the south end of the Unit 2 west wall and connected to the turbine building.
There is no such comparable structure attached to Unit 1.
3.
There are relatively small differences in equipment masses between the two structures.
4.
Modifications were made to locally increase the stiffnesses of the Unit 1 structure.
These same changes were also made to Unit 2 although in some 03/01/85 5-1 DIABLO CANYON SSER 29 SEC 5 J^
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cases the details of the changes varied slightly.
(Refer to SSER 18 page C.3-36 for a description of the modifications.)
5.
The buttress area long the west wall of Unit 2 is slightly different than the same area in Unit 1.
The differences between the two units are in general quite small and these differ-ences would have a small impact on the structural response.
The analysis of the Unit 1 turbine building was based on a three dimensional finite element model of the building.
Plate elements were used to represent the vertical shear walls and horizontal diaphragms while 3-D beam elements were used to model the beams, girders, and columns.
The roof system (trusses in both E-W and N-S directions) was modeled using an equivalent beam representing the roof truss properties. This beam connected the columns on the east side of the structure to those on the west side. A time history analysis was then per-formed to calculate floor response spectra while response spectrum methods were used to evaluate member loads caused by the seismic disturbances.
The same approach was used to analyze the Unit 2 turbine building after model parameters were modified to reflect the differences between the two units.
5.2 Scope of Review The emphasis of the staff's evaluation was focused on those aspects of the tur-bine building which are unique to Unit 2.
A general review of the PG&E analysis was made, followed by an audit on October 1 through 4,1984 (Ref. 32). A detailed review of the analytical approach, computer model, and design calcula-tions was made at this time. A site inspection of the turbine building was conducted on October 3, 1984.
Based upon the detailed review and inspection conducted during this audit, several specific topics were selected for further detailed evaluation. A series of design audits were held with PG&E on October 17, 1984, November 7 through 9,1984 and January 15 through 17, 1985 (Ref. 33 and 34).
Additional information and calculations were requested and reviewed by the staff during these audits.
Presented below is ~the staff eval-uation of specific calculations.
03/01/85 5-2 OIABLO CANYON SSER 29 SEC 5 w
m ww w
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i Calculation 64-T-275 L
This calculation presents a consolidated description of the Unit 1 models. The objective of the calculation was to assemble, in one place, all of the data that would be necessary for developing the Unit 2 model from the Unit 1 model. Mass calculations for all equipment and piping loads were included. ~ The analysis performed to determine the equivalent stiffnesses for the roof trusses was con-tained in the calculations.
Supplemental computations were made to verify that the equivalent roof stiffness matrices give similar results to those obtained when the actual roof truss members were analyzed for Unit-2.
1 Calculation 64-T-278 i
The differences between Unit 2 and Unit 1 were considered in this calculation.
1
_The major part of the calculation was concerned with computing the mass differ-
~
ences between the units.
Specific differences were identified with the mass i
values for each of the units tabulated.
The total mass differences were less than 5% with local differences being less than 10%.
Calculation 64-T-283 The model for Unit 2 was developed in this calculation.
The data contained in the above two calculations was the basis for the new model. A mirror ~ image of j
the Unit 1 model was used to form the geometry of the Unit 2 model with the i
Unit 2 model shortened to account for the two bay differences.
The mass differences developed in Calculation 64-T-278 were used to determine new model i
masses. New structural member properties were used when the structure is l
somewhat different.
Calculation 65-T-211 The north wall of Unit 2 was modeled with plane stress elements in the large i
model of the entire turbine building so that out of plane effects were not included in the model. This calculation evaluates the wall for out of plane-seismic loadings.
The wall was modeled with " equivalent beams." A response spectrum analysis was performed to obtain stresses in the wall, all of which 03/01/85 5-3 DIABLO CANYON SSER 29 SEC 5 i
h e
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. m wie
.--==w=v=
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- m*-
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~ * ' ' * '
-w-i+--t+1---
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==
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are shown to be within allowable limits, and reaction forces. The structure has been modified by adding members to distribute these reactions from the north wall (column line 19) to the concrete floor diaphragms which begins at column line 21.
During the plant inspection questions were raised as to
.the adequacy of this modification. This area was therefore identified as one requiring additional review.
Calculation 65-T-235 Stresses in the shear walls along the east and west end of the structure were evaluated in this calculation.
The total in plane shear forces, as determined from the response spectrum analysis, were applied to the shear walls.
This re-sults in a shear and moment requirement for the wall.
The loads were compared with the wall capacities.
Shear capacity of the wall is detiermined in accordance with the ACI Code as V = (s { + pf ) x area y
The moment capacity of the wall was based on ultimate strength methods consi-dering the wall section between steel column lines.
In one instance (east wall between column lines 28 and 29 at elevations between 119 and 140 feet) the moment capacity of the wall did not meet the demand.
In this case the area of the steel column was considered to be acting with the wall and the resulting moment capacity was found to be adequate.
Questions were raised at the audit as to the whether the connection between the column and wall would allow the column to act with the wall. This staff concern was therefore identified as requiring further investigation. The result of the staff evaluation is pre-sented in Section 5.3.1.1.
Calculation 65-T-111 The columns of the structure were evaluated in this calculation. The axial loads and bending moments about both axes were taken from the responses spectrum analysis.
The safety of the column was evaluated based on the AISC criterion:
03/01/85 5-4 DIABLO CANYON SSER 29 SEC 5
~y
.u.=.
.=
P C, M C, M x
y F+
p p
51
+
(1 p )M (I
F )M cr P
py px CV Cr In cases where the column does not meet this criterion the required member ductil-ity, p, was determined.
This was based on the DCP internal document "Bechtel Design Guide No. C-2.33 Simplified Inelastic Design of Non-Safety Related Structures," Rev. 2.
This criterion requires:
N N
P x
p + g- +
1p y
yx x
where M denotes the ductility ratio.
The largest ductility ratio found for the columns was 1.2.
This is further discussed in Section 5.3.1.1.
Calculation 65-T-311 The adequacy of the buttresses was evaluated in this calculation.
The buttress structure and steel column were treated by PG&E as a composite structure, with the stiffness of the composite structure lumped at the center line of the column.
The width of the buttress was substantial (29 feet).
The consequences of this shift of composite section center was identified as a staff concern that should be investigated further.
Detailed computations were performed for the shear walls of the buttresses and the caisson foundation system.
The strength of the shear walls was evaluated by methods consistent with the requirements of the ACI Code and found to be adequate.
Bearing capacity of the caisson required in these buttress calcula-tions were lower than the frictional capacity of the caisson, as determined from a tension pullout computation.
Concrete Floor Diaphragm Strength Evaluation During the January 15-17 audit PG&E discussed a problem that they discovered during the strength evaluation of the concrete floor diaphragms.
A construction joint runs along column line C from column line 19 to 20 at elevation 140 feet.
03/01/85 5-5 DIABLO CANYON SSER 29 SEC 5
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c A check of shear friction along this construction joint by PG&E indicated that a problem might exist.
Further studies are being performed by PG&E to determine i
whether the problem actually exists and if so what impact there would be on the overall behavior of the turbine building.
The large model of the structure will be modified so that shear cannot be distributed across this joint.
Solutions obtained from this model will then be used to determine the effect of such relative shear motion on the building's response. The nonlinear computer code, FINEL, will be used to determine the magnitude of the motion which could develop across this joint. The staff concludes that this issue need not be resolved prior to low power operation ~but must be satisfactorily resolved prior to full power operation.
5.3 Findings and Conclusions 5.3.1 Independent Evaluations As discussed above, the following three areas have been selected for further independent staff evaluation:
shear wall analysis; computer simulation of j
buttresses; and analyses of horizontal floor diaphragm at elevation 119 feet.
4' 5.3.1.1 Shear Wall Analysis
' Stresses in the shear walls of the turbine building were evaluated by PG&E in calculation 65-T-111, in which standard ACI Code methods were used to evaluate the wall shear and moment capacities.
In one instance (east wall between column lines 28 and 29 at elevations between 119 and 140 feet) the com-puted moment capacity of the wall did not meet the demand.
PG&E then assumed that the steel column acted with the wall and thereby developed a sufficient capacity.
This would require strain compatibility between the vertical strain in the wall and the axial column strain. The PG&E analysis did not account for the added strength that would be derived from the constraint provided by the girder at elevation 140 feet spanning between column lines 28 and 29. This girder would constrain the overall vertical deformation of the shear wall to be equal to the overall deformation of the column. The following independent analysis was performed to evaluate the extent to which this girder constraint would make the shear wall and column act as a unit.
03/01/85 5-6 DIABLO CANYON SSER 29 SEC 5
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The shear wall spans between the webs of the columns at lines 28 and 29.
The horizontal wall reinforcing bars (# 8 @ 12" each wall face) were welded to the web of the steel column.
The wall is continuous through the elevatio'n 119 feet level and butts into a horizontal steel beam at the elevation 140 feet.
The wall in question was subjected to a horizontal shear load of 1658 kips at the top of the wall.
A finite element computer model of the wall was made and solutions obtained with the SAP V computer program by the staff's consultant, Brookhaven National Laboratory (BNL).
Plane stress elements were used to model the shear wall and beam elements were used to model the columns and horizontal beam. The base nodes were fixed.
The column nodes were coupled to the wall nodes in the hori-zontal direction but were free to undergo relative vertical displacement.
The horizontal beam at the top of the wall was coupled to the wall for all degrees of freedom.
A horizontal shear load of 1658 kips was applied to the top of the wall and the solution obtained with SAP V.
The columns were found to have an axial force of 295.7 kips and a bending moment of 1064 kip-feet. The total moment to be car-ried across the elevation 119 feet level is 34,818 kip-feet.- The columns carry a moment of 8223 kip-feet so that the shear wall must carry the difference of 26,595 kip-feet. This moment demand of 26,595 kip-feet is less than the wall capacity of 31,150 kips as found in the PG&E Calculation 65-T-235. The shear demand of the wall is 1592 kips which is less than the wall capacity of 3645 kips. The staff concludes that the wall is adequate. Therefore, the girder provides a sufficient constraint so that the shear wall and column act together and the orginal concern is resolved.
5.3.1.2 Buttress Model Buttresses were provided to increase the E-W stiffness of the turbine building.
These buttresses are 2 feet thick reinforced concrete shear walls, 29 feet wide and span, practically from the foundation (elevation 85 feet) to elevation 119 feet. The buttresses were tied directly into the main columns of the build-ing.
The PG&E model of the turbine building includes the buttresses by lumping 03/01/85 5-7 OIABLO CANYON SSER 29 SEC 5
4 their stiffness with that of the column and placing the " composite" member at the center line of the column.
A plane frame model of a section of the turbine building centered about an E-W column-line was considered by BNL to investigate the effect the offset of the center of stiffness for the buttress on the building's response. Two models of the plane frame were made. The first model of the buttress was similar to the one used by PG&E.
The second model used plane stress elements to model the actual. geometry of the buttress.
The mass and stiffness properties of the models were taken directly from the PG&E model.
A horizontal acceleration loading of 1.0 g was applied-to each of the models and the solution was found using the BNL SAP V computer code.
A comparison of dis-placements and stresses is given below.
Comparison of Displacement and Stresses Parameter DCP Model BNL Podel Displ. at El-193' (inch) 8.63 9.69 Displ. at El-140' (inch) 1.506 1.978 Displ. at El-119' (inch) 0.086 0.133 Displ. at El-104' (inch) 0.045 0.063 Max. Stress in Column (ksi) 31.6 30.6 Buttress Shear Stress (ksi) 0.167 0.360 Buttress Normal Stress (ksi) 0.883 0.672 As can be seen the column stresses predicted with the PG&E model are more severe than those predicted by the more exact model.
The PG&E model predicts lower shear stresses and higher normal stresses than the BNL model. When these stresses are combined to obtain principal stresses, the PG&E model gives higher principal stresses than the BNL model.
Since stress evaluation is made against the principal stresses, the PG&E model is therefore conservative.
It may also be noted that the PG&E model is stiffer than the BNL model as shown by the smaller displacements.
Frequencies were determined for each of the models to assess this effect.
The first five frequencies are tabulated below.
03/01/85 5-8 DIABLO CANYON SSER 29 SEC 5
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Frequency Comparisons i
1.11 cps 1.05 cps 2
8.6 cps 8.3 cps 3
13.3 cps 10.6 cps 4
327.8 cps 3 (27.8 cps 4 5
(20.5cpsg 920.2, cps /
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For all modes, except the third, the variation in frequency between the two,
models.is well within the 10% broadening used to construct design spectra from calculated spectra. The actual design spectra was examined to determine the significance of the 20% variation in the third mode frequency.
It is con-cluded that there would not be a significant change in input to the structure for a frequency shift of this magnitude.
The' staff concludes that the PG&E
- modeling of the buttress is adequate.
5.3.1.3 Floor Diaphragm at Elevation 119 Feet The floor diaphragm at elevation 119 feet of the turbine building, between column lines C and E, is connected in the north-south direction to the vertical shear wall at column line 19 by means of struts which extend from the wall to the -
floor diaphragm.
The floor diaphragm was modeled by PG&E in the finite element model of tile turbine building as a series of plane stress membrane and truss elements.
Stresses from this computation, for both in plane compression and shear, were listed in calculation 65-T-421.
These stresses were compared with allowable shear stresses which were computed in Calculation 65-T-405.
The allowable shear stresses were based on elastic shear buckling formulae.
It was noted in this computation that the computed shear stresses listed by the PG&E calculations exceeded the allowable shears in some of the plate elements.
In addition, for those elements which were subjected to both compression as well I
i as shear, interaction between the compressive and shear stresses was not con-sidered when evaluating the buckling capacity of the plate elements.
The approach taken by PG&E was based on the assumption that the floor beam system supporting the diaphragm plates carries the in plane compressive forces, while the diaphragm plates themselves are subjected to only in plane shears.
03/01/85 5-9 DIABLO CANYON SSER 29 $EC 5
However, it was noted during the site inspection trip that the plates would be subjected to the in plane compressive forces since they are welded to the floor beams. To substantiate this observation, BNL undertook several finite element analyses of the floor diaphragm plates, using the SAP V computer program.
Single plate elements of the PG&E model were subdivided into smaller elements.
Approxi-mata stress and displacement boundary conditions were applied to these finer mesh models, and in plane stresses computed by the computer code.
In all. cases, it was noted that in plane compressive stresses were developed in the plate elements. This indicated that the behavior of the beam and plate system assumed by PG&E was not appropriate.
At the staff audit of November 7-9 and in the period following, additional cal-culations were provided by PG&E which refined their calculations and substanti-ated the diaphragm design. The primary recalculation was concerned with deter-mining the refined strut loads transmitted to the diaphragm along column line D and C3. The section of the vertical shear wall along column line 19 which contributed to the struct loads was recalculated.
It was found that the original tributory wall area to the nodal mass was too high at one node and too low at the adjacent nodes. New masses were calculated and used in a refined analysis of the diaphragm in the vicinity of the strut locations. These calcula-tions were reviewed during the January 15-17 audit.
It was determined from these calculations that considering the conservatism still remaining in the applied design loads (e.g., by applying peak accelera-tions to all mass point), the horizontal diaphragm will adequately perform its intended design function. An additional calculation was performed by PG&E to ensure that even if a restricted zone of instability did occur in the immediate area about strut line D, enough strength still remains in the diaphragm and in the wall at column line 19 to ensure that collapse would not occur due to seis-mic loadings.
The staff considers this issue resolved.
5.3.2 Conclusion Based on the completeness of the audits of the PG&E calculations, together with the independent checks performed by BNL, the staff determined that with one exception all staff concerns identified during the review regarding the turbine -
03/01/85 5-10 DIABLO CANYON SSER 29 SEC 5
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building have been resolved. The one remaining issue is the shear friction capacity of the horizontal concrete diaphragm at elevation 140 feet. This issue is described in the last part of Section 5.2.
Although PG&E has not completed its investigation on this issue, because of its localized nature and the limited
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number of Category I equipment in the surrounding area, the staff does not con-
' sider it to be an issue which would impact low power operation but must be resolved prior to full power operation.
03/01/85 5-11 DIABLO CANYON SSER 29 SEC 5
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T 6 RACEWAYS.
6.1 Introduction Class 1A electrical raceways are used to support and route all Class 1A elec-trical cables and wires used in the plant.
A raceway is a cylindrical pipe or conduit coritaining and protecting one or more electrical cables.
The raceways in turn are supported at points along their span by raceway supports which are attached to the structures. The raceway supports transmit the loads of the raceways to the structures and must be capable of performing this function during and after all seismic events, i.e., DE, DDE and Hosgri.
6.2 Scope of Review To assess compliance with the criteria the staff reviewed a sample of typical PG&E raceway support design calculations.
The staff also met with PG&E on October 18, 1984 at which time further requests for information were conveyed (Ref. 32). _ Finally, during November 7-9, 1984 an audit of the Unit 2 raceway evaluations was conducted at the PG&E offices in San Francisco (Ref. 33).
The documents listed at the end of this section were audited.
The sample al-X.
culations 2CG-91-1 and 2S-176 were reviewed in detail with every source number, i.e., spectral acceleration, allowable stress, etc., traced back to the appro-priate source document or PG&E Design Control Manual (DCM). The design aids were reviewed to assess their completeness, adequacy and ease of use.
For the design aids a random assessment of correctness was made.
Finally, at the staff's request, PG&E provided and/or developed selected ancillary calculations to substantiate various design calculation parameters.
The following is a list of documents reviewed and audited:
i Revision Title and Identification Revision Date Design Calculation - 2CG-91-1 0
9/22/83 Design Calculation 176 1
9/21/83 i
02/28/85 6-1 DIABLO CANYON SSER 29 SEC 6
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Design Calculation - 2SPOTWELD 0-9/07/84 Raceway Weight Check - Conduit K 2613 0'
11/08/84 Longitudinal Conduit Run - LCR 2 0
2/08/83 Raceway Support Design Aid 1
-0 8/03/84 Raceway Support Design Aid 2 0
8/04/84 Raceway Support Design Aid 11 0
9/08/84 Raceway Support Design Aid 29
- 0 11/23/81 Raceway Support Design Aid 31 16 5/11/84 Raceway Support Design Aid 33 12 7/30/82-Raceway Support Design Aid 34 17 6/26/84 i
Raceway Support Design Aid 35 9
8/03/83 Raceway Support Design Aid.35A 8
7/26/83 Raceway Support Design Aid 358 12 1/10/84 Raceway Support Design Aid 64 15 3/27/84 Raceway Support Design Aid 69 18 7/10/84 f'
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Qaids =DesignAid 6.3 Findings and Concerns The PG&E evaluation of raceway supports was found to be comprehensive.
Each support was qualified by either separate calculation or by generic enveloping calculations.
For each support or group of supports two calculations were in -
fact made. One to demonstrate adequacy for vertical and transverse seismic i
seismic loadings and other to the demonstrate adequacy for vertical and' longi-tudinal seismic loadings. The transverse evaluations typically involve a com-puter analysis of the support structure using the STRUDL computer code while the longitudinal evaluations were all' hand calculations. For either calcula-i tion the designer had available a number of design aids which summarize the j
applicable allowable stresses and design loadings. These design aids simplify the design task, speed the check effort and should reduce the incorrect use of design data.
In a typical design calculation for a single support type the first evaluation was made considering the generic support configuration and using generic (enveloping) design loadings for all supports of that type.
If some of the supports fail to meet criteria for those cor.ditions the calculation was refined to consider actual conduit weights,' support specific loadings and finally as built support configurations. 'In the design calculation reviewed (25-176) this 02/28/85 6-2 DIABLO CANYON SSER 29 SEC 6
a procedure was followed.
Items specifically verified and found to be adequate were:
generic weights, loadings and configurations a.
b.'
support specific weights, loadings and configurations c.
design aid use d.
unistrut spot weld strength check the consideration of actual bolt torque levels e.
f.
computer models In addition to verifying the source table for conduit weights, the staff requested PG&E to substantiate the listed weight for Conduit K2613. The com -
puter tabular listing of the raceway weights was verified by weight check calculations using the manufacturers weight data for the wires and conduit and the PG&E supplied number of wires in the conduit.
Some minor discrepancies were noted in the transverse evaluations.
Specifically some member section properties and joint relaxations specified in the computer model for raceway Support 25-176 were not appropriate.
Since these parameters would only impact the longitudinal evaluations, the errors do not affect the transverse qualification hand calculations and are of no significance.
If the same model had been used for the longitudinal evaluation, the errors could have been significant.
Seemingly, since the designer knew these parameters were of no importance, he specified them in an approximate fashion.
The staff finds the methods used by PG&E to evaluate the electrical raceways to be acceptable, 02/28/85 6-3 OIABLO CANYON SSER 29 SEC 6
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7.0 -BURIED CONDUITS 7.1 Introduction Buried electrical conduits carry the electric power and control cables from the turbine building to the intake structure. The length of this conduit run is l
approximately 1500' feet and generally follows the path of the water intake pipes.
The electrical conduits lie next to the water intake pipes and are located either-
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j immediately above or to one side of the water pipes.
1 The buried electrical conduits are typically 4-inch or 6-inch plastic conduits i
which in turn 'are placed in duct banks. -These duct banks are anchored in pull j
boxes located at intervals of 200 to 300 feet. Approximately half of the duct banks are buried in sand backfill and the remainder in concrete. Where the conduits lie in sand, the specifications require the sand to be compacted around the conduits and concrete cover placed above the conduits to protect the j
conduits from damage due to penetrations.
Flexible connections are provided at transition points of the conduits and at entrances to the pull boxes and
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buildings.
i j
The review of the buried electrical conduits was initiated at the NRC design audit l
from November 7 through 9. 1984, at which time a general description of the conduit configuration was provided to the staff and its consultants (Ref. 33).
]
Additional information was requested at that time, which was supplied late-in November 1984. More information was reviewed at a continuation audit at the PG&E offices from January, 15 through 17,1985 (Ref. 34). Additional infor-1 mation was provided to the staff on January 31, 1985.
l 1
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7.2 Scope of Review i
t To assess the adequacy of the design and construction of the electrical duct f
banks, the staff audited pertinent information in the following areas:
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03/01/85 7-1 DIABLO CANYON SSER 29 SEC.7 4
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(a) Detailed drawings of the structural configuration of the duct bank run from the turbine building to the intake structure, together with the relationship of the duct banks to the water intake conduits and bed rock levels.
(b) Estimates of peak relative displacements that can be anticipated at various support point locations along the duct bank run.
l l
(c) Information on the flexible capability of'the typical connection used at 1
the various connection points.
l (d) Capability of the electrical conduits to function in fully submersed situations in either ordinary ground water or sea water.
(e) Description of the procedures used to ensure that the electrical design is developed as intended, that is, the wire specified is in fact placed in the duct bank.
i (f) Information that the slack placed in the cable runs at the various pull boxes to ensures that relative movement can be sustained.
7.3 Findings and Conclusions Information on each of the items listed above was audited by the staff. The l
review is summari:ed as follows.
(a) Drawing SK-ECR-1 was developed from both as-built drawings of the conduit configuration and rock line and soil information obtained from various Harding Lawson & Associates (HLA) reports.
Both the as-built drawings and the HLA reports were reviewed. The detailed description of the conduits provided on the drawing was found to be adequate.
(b) Peak relative displacements of conduits were estimated by HLA for several postulated relative movements.
For example, it was assumed that the pull box moves with the bedrock during a seismic event while the electrical 02/28/85 7-2 OIABLO CANYON SSER 29 SEC 7 l
3 conduit moves with the backfill soil above. Displacement estimates were made using the SHAKE and TRIP computer programs as well as by means of a.
simplified design procedure. This simplified procedure is based on esti-mating peak axial strains in long buried structures from one dimensional wave analysis (Ref. 42). The maximum relative motion computed, including both horizontal and vertical potential movements, was 1.16 inches occurring over 1.5 feet which is the length of the flexible connection used at the conduit junction points. Test results show that the flexible connector used for the conduit can. easily accommodate motion of this magnitude.
(c) A typical flexible connector was tested by PG&E at the request of the staff.
Two separate tests were run.
In the first, one end of the connector was clamped in a vise, while the other was simply pushed up and rotated through an angle of 60. The lateral movement was greater than 8 inches.
In the second test, one end of the connector was again clamped in the vise.
The other end was lifted up but this time the end was not allowed to rotate. Relative displacements greater than four inches were developed.
In either test, no physical damage to the connector was noted. Although these tests were relatively crude, they show that the connector can sustain much larger movements than the conservative estimates mentioned above.
(d) Various cable specifications and typical test results were used by PG&E to indicate the capability of the electrical cables to operate in the adverse environments postulated at the site.
In particular, specifica-
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tions and data were utilized to indicate that the cable has long-term stability in the fully immersed condition in both fresh water and sea water.
Capability to perform in sea water is of particular interest at the Diablo Canyon site since the conduits are below sea level at the entrance to the Intake Structure.
(e) The staff reviewed a detailed description of the procedures employed by PG&E to ensure that tne cable specified in the Engineers Material Memo was in fact placed in the conduit. The process includes the development of purchase orders, specifications, factory inspection and testing, receipt inspection and testing at the site, cable installation and documentation.
02/27/65 7-3 OIABLO CANYON SSER 29 SEC 7
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The procedures employed are considered adequate to ensure that cable is placed as specified in the design.
(f) The staff requested that the cover plates of all pull boxes be removed and the cable connections within the pull boxes be physically inspected and photographed. The purpose of this inspction is to determine if in fact enough slack is provided in the electrical cables to accommodate the potential relative movements mentioned above.
By inspection of the photographs provided to the staff, slack significantly in excess of one foot is available in each cable run so as to provide adequate capabil--
ity for relative movement.
The staff concludes, based on its review and evaluation of engineering proce-dures and design and calculational documentation on buried electrical conduits, incuding audits and inspections, that both the design and as-built condition of the cables are adequate.
02/27/,55 7-4 DIABLO CANYON SSER 29 SEC 7
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8 PIPEWAY STRUCTURE 8.1 Introduction l
The pipeway structure is a steel frame structure attached to the outside of the containment wall as well as the auxiliary building and the turbine building.
lt has five major platforms at elevations 109, 114, 119, 127, and 138 feet.
l The two main steam lines 1 and 2 and feedwater lines 1 and 2 are supported by 1
the pipeway structure.
Large portions of the structure were shop fabricated and then joi~ned together in the field by bolted connections.
The pipeway structure is designed for DE, 00E and Hosgri earthquakes.
The Unit 1 pipeway seismic analysis was done by Westinghouse whereas the Unit 2 analysis was done by PG&E. A description of the Unit 2 pipeway calculation is given below.
l 8.2 Scope of Review PG&E performed the seismic evaluations for the pipeway structure using a three dimensional frame model. This model incorporated a single stick representing the containment wall.
This stick is coupled with the three dimensional assem-bly of beam and truss elements representing the pipeway structure plus the major piping systems.
Piping and piping supports were modeled with beam ele-ments.
This couoled model was used for seismic analysis for the Hosgri earth-quake. The evaluation for DE and 00E earthquakes was performed by Nuclear Services Corporation.
During January 15-17, 1985, the staff conducted an audit of the Hosgri analysis for the pipeway structure performed by PG&E.
The following documents were reviewed during the audit:
(1) Calculation No. 1113C-1, Rev. O, Procedure for Development of Geometry for BSAP Dynamic Computer Model.
02/28/85
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Computer Analysis Output.
(3) Calculation 1114C-1, Rev. O, Dynamic Analysis, BSAP Structural Mathematical-Model-Mode Data (including geometry and' constraints).
(4) Calculation 1114C-2, Rev. O, Dynamic Analysis, BSAP Structural Mathematical Model-Element Data (including material and section properties).
(5) Vol. 27: Computer Analysis, Vol. 27, Calc.1116C-1, Rev. O Dynamic Anal-ysis:
Frequency Run 3-3 Cycles #25 to #30 (with mode shape outputs)
(6) Vol. 25:
Computer Analysis Outputs. Calc 1116C-1, Dynamic Analysis:
Frequency Run 3-1 Cycles #25 to #30 (with mode shape outputs)
(7) Calculation 1141C-1, Design Guidelines for Static Evaluation of Pipeway Structure for Gravity and Hosgri Loads.
The coupled containment and pipeway structure model was used to obtain 142 modal frequencies between 0.92 Hz and 32.2 Hz. A time history dynaatc analysis was performed to generate the floor respons'e spectra. The floor response spectra were computed for various locations as required by the piping analysis group.
The input time history at the attachments of the pipeway structure to the auxiliary building and the turbine building was taken to be the same as the earthquake motion applied to the base of the containment stick model.
Five separate cases were considered:
two in the horizontal direction (N-S, E-W) using the Blume input time history with t = 0.04 seconds, two in the hort-zontal direction (N-S, E-W) using the Newmark input time history with t = 0.04 seconds and one in vertical direction using the Newmark input time history for t = 0.0.
The floor response spectra generated by the coupled containment and pipeway structure model were then used in piping evaluations which in turn produced loads which were used in a static analysis for member evaluation.
The static 02/28/65 I-2 OlABL7 CANYON SSER 29 SEC 8 l
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analysis utilized the dead load plus Hosgri combination. The members of the pipeway structure were then qualified based on this load combination. PG&E demonstrated that all members passed the stress check.
8.3 Findings and Conclusions
1 The following are the NRC staff findings and concerns on the pipeway structure:
The slotted holes provided to allow the relative motion between the pipe-a.
way and the auxiliary building were not properly accounted for in the structural model. Specifically, the nodes representi~ng the connection of the pipeway structure and the auxiliary and turbine buildings were modeled as free nodes in both global horizontal directions. Only the vertical motion of these nodes was restrained. The intended motion of these nodes, however, was to move along the direction of the framing beam. Similar modeling procedures were used for the end nodes of the radial beams of the pipeway structure which are framing into the radial beams of the pipeway structures which are framing into the turbine building.
Based on additional information provided by PG&E it is concluded that the modeling details of the connections to the turbine building will not impact the results of the analysis done for the pipeway structure based on this model. This is because the displacements in both directions can be accommodated by the slotted holes pended at their connections. With respect to the boundary condition of the nodes representing the pipeway structure framing into the auxiliary building, the pertinent displacement values should be provided in order to reach a conclusion as to whether or not there is sufficient clearance by the slotted holes to assure that these nodes are allowed to move in both horizontal directions and there-fore consistent with the assumptions used in the model of the pipeway structure, b.
As a result of the staff's review of the seismic evaluations performed by PG&E for the Hosgri event, it was found that the same input motion was applied at all support nodes of the dynamic model. Given the fact that 02/28/85
$[- 3 OIABLO CANYON SSER 29 SEC 4 3
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the pipeway structure is supported by different structures, i.e., contain-ment, auxiliary and turbine buildings, a choice of-a single input that reflect the effects from all nf these structures is required. PG&E selected the containment ground acceleration time history. Thus, the
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staff requested PG&E to provide the basis whdeh this selectiony h2ead
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for l W
In response to this request, PG&E provided additional information in order to justify the selection of the input used in the Hosgri evaluation of the pipeway structure. A comparison of vertical spectra from both the con-tainment and the turbine building at the location where the pipeway struc-ture is supported was made.
It shows that the containment acceleration spectra envelope of the turbine building at all frequencies of interest.
On the other hand, due to the pinned conditions at both ends of the beams framed between the pipeway structure and the auxiliary building no verti-cal input can be transmitted from the auxiliary building to the pipeway structure.
It is not clear, however, if vertical input can be transmitted from the auxiliary. building to the pipeway structure near the piping systems i.e., main steam line.
Thus, the attachments of the main steam lines in both the pipeway structure and the auxiliary building should be reviewed to assure that such vertical transmission of earthquake would not happen.
In this case, some spectra comparision would be necessary to provide the justification for the choice of the vertical input in the pipeway evaluations.
Based on the information submitted on January 31, 1,985, it is still not c.
clear that the time step of 0.01 sec used in the time history response calculation is adequate to properly compute the response spectra above-10 Hz.
Thus, this is still considered to be an open issue.
d.
The accidental torsion was accounted for by increasing the input amplitude by 6'..
The staff requested PG&E to justify this procedure. Normally the accidental torsional effects are accounted for by dynamic models in which the masses are set at certain eccentricity from the axis.
Instead of this procedure PG&E chose to increase the input by 6*.' and performed a dynamic 02/28/85
(-4 DIABLO CANYON SSER 29 SEC 8
q analysis using the original model without off-setting the lumped masses of the model. PG&E supplied additional information to justify that the 6%
increment in the input produces results which are comparable with those obtained by using a model in which eccentricities are included. Based on the information provided by PG&E the 6*4 increase was obtained from a study which was performed for the Unit 1 pipeway structure.
In this study it was concluded that the 6'; increase in the input
- earthquake can sufficiEntly cover the accidental torsional effects. This procedure was used in the Unit 2 pipeway structure af ter performing a comparison between the charac-teristics of the Unit 1 versus Unit 2 pipeway structures.
It was con -
cluded that the procedure used in Unit I was also applicable to the Unit 2 pipeway structure.
The staff concludes that this approach is acceptable.
In response to the staff's request, PG&E reviewed the strength capability e.
of the pipeway structure to accommodate the relative motions between structures and still remain within the allowable stress levels. The ver-tical motions between the various buildings supporting.he pipeway struc-ture are shown to be small.
They are not expected to alter significantly the stresses of the pipeway structure. On +' ' other hand, PG&E stated that the horizontal relative displacements can be accommodated by the connections provided with the slotted holes. Based on this information the staff concludes that the concern is resolved.
f.
Since the design of structural members of the pipeway structure could be controlled by the DE and/or DDE earthquakes, the DE and DDE aralysis should be provided to the staff for review.
Based on the detailed review of the PG&E calculations, the staff finds that the general approach used for the analysis of the pipeway structure by PG&E is adequate, subject to the resolution of the concerns identified above.
It is the staff's judgment that final resolutions of these concerns will not signifi-cantly alter the overa!1 response of the pipeway structure or structural modifi-catidn ca$Se9elformedwithlittledifficulty.
Therefore, the staff concludes Y
A that these issues need not be resolved for low power operation but must be resolved for full power operation.
02/28/85 l-5 DIABLO CANYON SSER 29 SEC 8
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4 9 PIPING SYSTEMS AND PIPE SUPPORTS The Unit 1 design verification effort included an extensive review and evalua-tion by the IOVP and ITP of the analyses for piping systems and pipe supports.
The staff evaluation was presented in SSER 18 which also identified a number of staff concerns. Their resolution was presented in SSERs 19, 20 and 24.
In late 1983 a number of allegations were identified to the NRC that pertained to small bore piping engineering practices, primarily these activities performed within the Onsite Project Engirieering Group (OPEG) at the Diablo Canyon site.
A team of NRC staff and consultants reviewed and evaluated the allegations, which also included audits and inspections at the site and at the San Francisco offices of the DCP.
The results were presented in SSERs 21 and 22.
Further allegations continued to be made by former employees in the area of piping and supports analyses and associated programmatic procedures. In addi-tion, concerns were raised by a member of the NRC team. This resulted in a license condition, consisting of seven specific elements, which was included in the reinstated Unit 1 low power license and which had to be met prior to issuance of a full power license (Ref. 24).
The scope of the effort by the NRC Peer Review Group was redirected to evaluate the PG&E efforts for the resolution of the license condition.
It included pipe system walkdowns and further ir.spections at the site and audits at the PG&E offices. The effort also included an evaluation of certain specific issues that had been raised by the NRC staff member with regard to the IDVP effort and the staff's IDVP conclusions in the area of piping and supports. An evaluation of certain programmatic concerns was performed by a separate NRC team.
The staff evaluation was presented in SSER 25 which concluded that the PG&E actions in response to the license condition satisfactorily met the require-ments of the condition, that the conclusions reached earlier with regard to the IDVP remained valid, and that the programmattc issues concerning onsite engi-neering have been resolved.
Further details on the last item were subsequently provided in a Board Notification (Ref. 25).
02/27/85 9-1 DIABLO CANYON SSER 29 SEC 9
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The above described efforts by PG&E and the NRC staff were directed specifically to the reinstatement of tfie Unit I low power license in April 1984 (Ref. 31) and issuance of the Unit 1 full power license in November 1984 (Ref. 27). As discussed in Section 2 of this report, as the efforts for Unit I came to 7
completion the same efforts increased for Unit 2.
The staff determined to perform for Unit 2 a similar broad based review effort in the area of piping and supports as had been performed for Unit 1.
The effort would include an evaluation of the resolution of (1) issues that had been raised "during the Unit I design verification, (2) actions resulting from the Unit i license condition, and (3) the Unit 2 applicability and resolution of allegations related to piping and supports.
In late 1984 an NRC review team was formed to review, evaluate, and audit the PG&E efforts in the area of piping and pipe supports. The details of the team's effort and its conclusions are described in a separate report, SSER 30 to the Safety Evaluation Reoort. The team included members of the NRC staff (Office of Nuclear Reactor Regulation and Region I) and consultants from national laboratories and private companies. The effort included audits of piping system and pipe support analyses, piping configuration checks, and hot piping system walkdowns and inspections at the Diablo Canyon Project Office in San Francisco and at the site, as appropriate. The review team also audited and inspected the programmatic aspects of the Internal Review Program and the Allegation Review Program, in particular with respect to piping systems and pipe supports.
The complete review of a piping analysis consisted of verification that the i
calculation package contained the documentation required in accordance with the PG&E procedures and an in-depth review of the calculation. This included a review of isometric drawings, pipe support attachment prints, seismic spectra
)
development records, seismic and thermal anchor movemen'ts, and computer input and output for all load conditions. Engineering judgements and assumptions were reviewed for correction and documentation.
Pipe supoort analyses were also checked for completeness and correctness. This included verification of 7
the correlation of support lu:ation, orientation and loads with those specified in the piping analysis and of the correlation of support drawings with the computer input.
j i
02/27/S5 9-2 DIABLO CANYON SSER 29 SEC 9 q
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s The analysis packages were chosen for review, both randomly and selectively.
The random selection provided a realistic sampling of all analysis work per-formed on Unit 2 piping systems. The selective process was used to check on typical problem areas to assure proper performance of critical evaluations. The NRC review team audited approximately one hundred specific piping systems and pipe support packages.
As a result of this audit, the team raised questions in the f'ollowing areas:
modules of elasticity stress intensification factors baseplates branch connection pressure design pipe and pipe support configurations welded attachment local stresses rigid support assumptions for fluid dynamic load analysis U-bolt piping on angle bolts Details on these matters and their resolution are presented in SSER 30. The
. review team has pursued the PG&E resolution to these issues and concludes that in all cases the resolution is acceptable.
In general, the team found some calculational discrepancies which, however, did not impact the satisfactory compliance with appropriate licensing criteria. Specifically, the team deter-mined that the license conditions regarding piping and pipe supports for Unit I have been satisfactorily addressed and resolved for Unit 2.
As part of the piping and pipe support effort, the review team observed the PG&E walkdowns, in the hot and cold condition, of (1) the letdown line from the reactor coolant system loop to the regenerative heat exchanger and (2) portions of the main steam piping. The team examined the basis for selecting the number and locations of the displacement measurement points and concluded that these are sufficient to validate piping behavior and detect unintended restraints.
The team also reviewed the procedures for conducting the walkdowns. Based on its review of previous walkdowns and the prescribed procedures and based on its observance of the walkdowns, the team concluded that acceptable engineering techniques were used for the walkdowns, including methods for determining the actual displacements, and that the walkdowns were performed in a competent 02/27/85 9-3 DIABLO CANYON SSER 29 SEC 9
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y manner to reasonably detect and justify those normal displacements.which deviated from calculated displacements.
The effort by the NRC team regarding piping systems and pipe supports also-included an extensive audit and inspection of the IRP process. The team reviewed approximately 60 IRP packages and 100 piping and support calculational-pa'ckages' not only with regard to their technical adequacy but also with respect to the IRP process.
In summary, the piping and pipe support review revealed some findings which
^
have all been resolved or deemed to be not of safety significance. These can be categorized as discrepancies with only an insignificant effect on' the par-ticular analysis. The piping system hot walkdown procedure review and onsite audit of the walkdowns of the two systems revealed that acceptable engineering techniques were employed and the walkdowns were concluded in a competent manner.
l The staff concludes that the issues raised during the Diablo Canyon Unit 1 design verification effort, allegations and in the Unit I low power license condition have been adequately addressed and resolved for Unit 2 and that the licensing 1
criteria in this regard have been met.
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10 NON-SEISMIC DESIGN ASPECTS 10.1 Systems The staff reviewed those areas regarding the non-seismic design for Diablo Canyon Unit 2 that had been identified during the Unit 1 design verification effort under the IDVP/ITP and which are included in the PG&E Internal Review Program (IRP) for Unit 2 as described in Section 2 of this report. The PG&E submittals of November 2 and December 7, 1984 identified the specific findings from Unit l'and described the application and resolution of those items with respect to Unit 2 (Ref. 28). The IRP implemented the Unit 2 design changes necessary in a similar manner to those that resulted from the design verifica-tion effort in Unit 1 as documented in SSERs 18, 19 and 20.
The findings from
~
these reports were considered by the staff with respect to their applicability to Unit 2.
PG&E identified the issue of protection against jet impingement from moderate energy line breaks as an issue where the resolution for Unit 2 was different from the Unit I resolution.
This matter is discussed in further detail in Section 10.1.2.
Based on its review the staff confirms that the concerns identified during the Unit 1 design verification program have been satisfactorily resolved for Unit 2.
The staff concludes that IRP properly incorporates the findings of the Unit 1 l
design verification program and adequately provides for those Unit 2 changes required to assure compliance with licensing criteria in the nonseismic design.
10.1.1 Component Cooling Water System As a result of the staff review of the component cooling water system (CCWS) as 1
documented in SSER 16, the staff imposed a technical specification on the operation of the CCWS with respect to the temperature of the ultimate heat sink,
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the Pacific Ocean.
The technical specification was included for Diablo Canyon 02/27/85 10-1 DIABLO CANYON SSER 29 SEC 10
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Unit 1 in the full power license (Ref. 27). The same technical specification is also applicable for operation of Unit 2 and will be included in the Unit 2 license.
10.1.2 Protection from Jet Impingement Due to Moderate Energy Line Breaks The PG&E submittal of November 2,1984 identified in Table 1 the concern for protection from jet impingement due to moderate energy line breaks which had been raised as item E0I 8014 during the Unit.1 design verification effort (Ref. 28). Differences ~from the Un'it I resolution were noted for Unit 2.
Specifically, the original IDVP concern involved improper protection from jet impingement as a result of postulated moderate energy line breaks for two auxil-iary feedwater system (AFWS) flow control valves and four AFWS level control valves.
Further details on the resolution of the concern for Unit 1 and Unit 2 were provided by PG&E in a letter dated February 21,1985 (Ref. _28).
Regarding the flow control valves, PG&E concluded for the Unit 1 concern that jet impingement shields were not required'to protect these valves as they were on the alternate long term AFWS water source supply line and their operation was not needed to assure AFWS safety function following the postulated moderate energy line break.
Further, the valves were equioped with hand wheels to facili-tate their subsequent operation, if necessary. Resolution of this item for Unit I was documented in SSER 18 (Ref. 4), and Followup Item 8 in SSER 20 (Ref. 6).
PG&E concluded in the IRP that the Unit I resolution was applicable to Unit 2.
The staff concurs with this conclusion.
Regarding the four level control valves, PG&E confirmed during the Unit 1 design verification program that adequate jet impingement shields were provided to protect the valves from the effects of the postulated moderate energy line break.
The Unit 1 resolution is documented in SSER 18. PG&E reanalyzed the above con-cern for Unit 2 with respect to the licensing criteria and concluded that impinge-ment barriers were not needed for these valves because sufficient redundancy in the AFWS is provided to assure its safety function following the postulated moderate energy line break. The staff concurs with this resolution.
- Further, because of the similarities in the design of Units 1 and 2 in this regard, the staff concludes that the above resciution could be applied equally to Unit 1.
02/27/85 10-2 DIABLO CANYON SSER 29 SEC 10
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s 10.2 Jet Impincement Evaluation PG&E has performed an evaluation for Unit 2 of the effects of jet impingement m safety related equipment inside containment due to high energy line-breaks (HELB).
In accordance with the resolution of Unit 1, the scope for Unit 2 included as sources those lines with a pressure greater than 275 psi or a temperature greater than 200 F, and which also meet the size and usage cri-teria of the Standard Review Plan. The staff evaluation for Unit I was presented in SSER 24 (Ref. 10).
As with Unit 1, PG&E determined that three additional categories of lines would-be affected:
Charging line from the containment penetration to the regenerative heat exchanger Reactor coolant pump seal injection lines Accumulator injection lines upstream of the check valves The PG&E evaluation of the effects of postulated breaks in these lines and concluded the following:
The charging lines in the first category is located in the pipe ~ tunnel and the regenerative heat exchanger room. Jet impingement effects are bounded by other high-energy lines in these areas.
The second category of lines cannot produce a jet even if the lines rupture.
These lines are located downstream of needle valves which significantly restrict the flow.
In the reverse flow direction, flow is restricted by the reactor coolant pump seals such that the pressure is less than 50 psi.
A rupture in the accumulator injection lines can produce a jet. Accordingly, these lines were added to the jet impingement walkdown and reviewed.
Due to the layout of these lines, there are no unacceptable consequences from the hypothetical rupture of these lines.
No physical modifications resulted from the consideration of these jets.
02/27/85 10-3 DIABLO CANYON SSER 29 SEC 10
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s The staff has reviewed and evaluated the PG&E evaluation and found it accept-able.
In addition to expanding the scope of the jet impingement review, PG&E also evaluated the applicability to Unit 2 of the findings and concerns by the IDVP E
during the design verification for Unit I regarding the effects of a HELB in Auxiliary Feedwater (AFW) System Line 594. This evaluation was performed as part of the Unit 2 Interval Re w Program under package IRP 2-8049 as described in detail in a submittal dated February 21, 1985 (Ref. 28).
The evaluation consisted of a detailed review of the layout of the AFW System in Unit 2 to determine the location of the postulated HEBL's and the effects on adjacent equipment, in particular the break which generated the concern in Unit 1.
PG&E determined that the Unit i resolution of the concern is not directly applicable to Unit 2 because the local pipe routing and location of supports of the AFW System at the location of the postulated break in Unit 1 is different from that in Unit 2.
As a result the breaks are not postulated at the same location for Unit 1 and Unit 2, and therefore a break location similar to that considered for Unit 1 does not exist for Unit 2.
The staff has reviewed this evaluation and finds it acceptable. PG&E also stated that the effects of postulated HELB's at other locations of the AFW System were evaluated as part of the Unit 2 overall high energy jet impingement effects review. The staff finds this acceptable based on the review performed for Unit 1.
10.3 Eouipment Qualification 10.3.1 Environmental Qualification As a result of the Unit 1 design verification program the staff had identified in SSER 18 four items (FI4, FIS, FI12, FI14) pertaining to the requirements of environmental qualification of electric equipment important to the safety for nuclear power plants as contained in 10 CFR 50.49. These concerns were resolved in SSERs 19, 20 and 24 The staff has reviewed the PG&E evaluation and resolu-tion for those items with respect to Unit 2 as described in the IRP documenta-tion (Ref. 28).
The staff concurs with the PG&E evaluation and concludes that the PG&E action resolves these concerns for Unit 2.
02/28/55 10-4 DIABLO CANYON SSER 29 SEC 10
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10.3.2 Main'AnnunciatorTypewriter Seismic Qualification During the Unit 1 design verification, the IDVP questioned the basis for the seismic qualification of the Unit 1 main annunciator typewriter. Upon further evaluation, it was verified that-the auxiliary building _ spectra were appro-priate and no corrective action was requested. During the IRP for Unit 2, PG&E determined that a different resolution was required for! Unit 2 as discussed in the PG&E submittals of November 2, 1984 and February 21, 1985 (Ref 28). The accelerations associated with the location of the Unit 2 annunciator typewriter exceeded the accelerations of the required response spectrum used for seismic qualification.
In order to correct the situation, the Unit 2 typewriter was relocated to a location equivalent to Unit 1, where seismic qualification was valid. The staff has reviewed _the information pro-vided with regard to che acceptability of the resolution. Based on the above, the staff has determined that the equipment is seismically qualified and the resolution is acceptable.
02/27/35 10 DIABLO CANYON SSER 29 SEC 10
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v AII CAPS 11 Quality Assurance The Quality Assurance (QA) Manual for the Diablo Canyon Nuclear Power Plant was issued by PG&E in January 1970, utilizing the proposed 18 criteria for QA. The criteria was issued in June of 1970 unchanged in Appendix B to 10 CFR 50,
" Quality Assurance Criteria for Nuclear Power Plants." The construction permit for Unit I was issued in April 1968 and for Unit 2 in December 1970. The QA Manual was to be used for Unit 1 to the extent possible and to be fully applied to all safety related activities and items for Unit 2.
With the formation in early 1982 of the Diablo Canyon Project (DCP), a PG&E organization including technical and management staff from PG&E and the Bechtel organization, the QA program was revised and based largely on the Bechtel QA program (Ref. 43). The staff reviewed the program and found that the procedures, require work and controls, when properly implemented, comply with the require-ments of Appendix 8 of 10 CFR 50 (Ref. 44). Thus, from 1969 until today PG&E has been committed to a QA program for Diablo Canyon Units 1 and 2 which meets NRC requirements.
In SSER 18 the staff concluded that " shortcomings found in and as a result of earlier QA programs (implementation) for certain design activities are being compensated oy verification of the design under the IDVP, that construction was done under acceptable QA controls, and that current corrective actions and the IOVP work itself are being performed in accordance with acceptable QA programs."
These conclusions are applicable to both Units 1 and 2.
Since the initiation of the design verification program in late 1981 numerous inspections have been performed in the area of quality assurance, principally by the NRC Region V Office. These inspections, in general, covered both units.
Sixteen vtalations pertaining to construction activities were issued since September 1981. The majority of these violations applied to a cross-section of construction errors, including welding of structural steel, piping, raceways, and HVAC supports.
PG&E was responsive to the notices of the violation.
02/27/S5 11-1 DIABLO CANYON SSER 29 SEC 11 a
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Considering the extensive inspections effort and the relatively small number.of i-minor violations, the staff finds that adherence to procedures was generally satisfactory.
The'constructioninspe'ctionsalsoshowedPG&Emanagement.tobefrequently involved in construction activities. The PG&E personnel generally had a good i
understanding of safety issues and worked towards resolution in a timely man-ner. The construction staffing appeared to be. adequate with identified positions filled on a priority basis. Training and qualification of PG&E/ contractor inspec-tion personnel has been improved in response to several of the previously mentioned violations related to contractor quality control.
Extensive NRC examinations were also performed as a consequence of large number of allegations. These allegations dealt principally with design and construc-tion activities, quality assurance, and quality control. Although over 1600-allegations were received by the NRC as of late 1984 from various sources, the great majority of these allegations were received by the staff since September 1983,- coincident with the Diablo Canyon Unit I readiness for fuel loading and low power testing.
The result of these examinations and investigations indicated that, while there l
may have been some lapses in the quality and management systems related to construction, the systems have worked reasonably well. The staff finds that there is reasonable confidence that the licensee and contractors have acted responsibly over the yea *s.
In conclusion, this extensive inspection effort demonstrated that the licensee has constructed the plant in substantial agree-ment with regulatory commitments and requirements.
I.
1 02/27/85 11-2 DIABLO CANYON SSER 29 SEC 11 v
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't REFERENCES 1.
U.S. Nuclear Regulatory Com ission,."Diablo Canyon, Units 1 and 2 Safety Evaluation Report," October 12, 1974.
2.
U.S. Nuclear Regulatory Comntission, "Diablo Canyon,. Units 1 and 2 Safety
~
Evaluation Report, Supplement 16," SSER 16, August 1983.
3.
U.S. Nuclear Regulatory Commission, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report. Suoplement 17." SSER 17. February 1984 4.
U.S. Nuclear Regulatory Commission, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report, Supplement 18," SSER 18, August 1983.
5.
U.S. Nuclear Regulatory Commission, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report, Supplement 19," SSER 19, October 1983.
6.
U.S. Nuclear Regulatory Comrsission, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report, Supplement 20," SSER 20, December 1983.
7.
U.S. Nuclear Regulatory Commission, "Diablo Canyon, Units 1 and-2 Safety Evaluation Renort. Supplemest 21." SSER 21, December 1983.
8.
U.S. Nuclear Regulatory Commission, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report, Supplement 22," SSER 22, March 1984.
9.
U.S. Nuclear Regulatory Commission, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report, Supplement 23," SSER 23, June 1984.
l 10.
U.S. Nuclear Regulatory Commission, "Diablo Canyon. Units 1 and 2 Safety Evaluation Report, Supplement 24," SSER 24, July 1984.
l i
02/27/85 12-1 DIABLO CANYON SSER 29 SEC 12
i,.
o' 11.
U.S. Nuclear Regulatory Commission, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report, Supplement 25," SSER 25, July 1984.
- 12. U.S Nuclear Regulatory Commission, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report, Supplement 26," SSER 26, July 1984.
- 13. U.S Nuclear Regulatory Ccmmission, "Diablo Canyon, Units 1 and 2 Safety Evaluation, Supplement 27," SSER 2'7, July 1984, 14.
U.S. Nuclear Regulatory Commissrion, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report, Supplement 28," SSER 28, in preparation.
15.
U.S. Nuclear Regulatory Co.mmission, "Diablo Canyon. Units 1 and 2 Safety Evaluation Report, Supplement 30," SSER 30,.in preparation.
16.
U.S. Nuclear Regulatory Cammission, Commission Memorandum and' Order CLI-81-30, " Order Suspending License," November 19, 1981.
17.
U.S. Nuclear Regulatory Commission, November 19, 1981, from Harold R.
Denton (NRC) to Malcolm H. Furbush (PG&E),
Subject:
Diablo Canyon Unit 1
- Independent Design Verification Programs.
- 18. ALAB-756,18 NRC 1340 (1983).
19.
ALAE-763, 19 NRC 571 (1934).
20.
U.S. Nuclear Regulatory Commission,
" Pacific Gas and Electric Company, Diablo Canyon Nuclear Power Plant, Unit 1 Amendment to Facility Operating License," Amendement. No. 8, April 13, 1984.
21.
U.S. Nuclear Regulatory Commission, November 17, 1983, from. William J.
Dir ks (NRC) te J. B. Martin (NRC). H. R. ' Denton (NRC), and R. C. DeYoung (NRC),
Subject:
Review of Allegations Outstanding on Diablo Canyon Nuclear Power Plant.
02/27/85 12-2 DIABLO CANYON SSER 29 SEC 12-y
_{'
4 d
22.
Pacific Gas and Electric Company, "Diablo Canyon Unit 2, Allegation Review Program, Final Report," December 5, 1984.
23.
U.S. Nuclear Regulatory Commission, Board Notification 84-084, April 18, 1984, Transmitting report of NRC review group Diablo Canyon piping issues, April 12, 1984.
24.
U.S. Nuclear Regulatory Cominission, "Diablo Canyon Low Power License DPR-76, Order Modifying License," April 18, 1984.
25.
U.S. Nuclear Regulatory Commission, Board Notification 84-161, "Diablo Canyon - Completion of Piping Review Activities," September 24, 1984.
26.
Pacific Gas and Electric Company, " Piping and Pipe Supports Review Program for Diablo Canyon Unit 2, Final Report," January 31, 1985.
27.
U.S. Nuclear Regulatory Commission, "Diablo Canyon' Full Power License DPR-80," November 2, 1984.
28.
Pacific Gas and Electric Company, "Diablo Canyon Unit 2, Design Review," October 6, 1983,
" Internal Review Program Summary and Status Report," PG&E Letter No.
DCL-84-276, July 31, 1984,
" Internal Review Program Summary and Status Report," PG&E Letter No.
CCL-51-332, Octcber 19, 1954,
" Internal Review Program Final Report," PG&E Letter No.
DCL-84-344, November 2, 1984,
" Internal Review Program Final Report," PG&E Letter No.
DCL-84-378, December 7, 1984,
" Internal Review Program - Supplemental Information," PG&E Letter No.
DCL-84-071, February 21, 1985.
29.
U.S. Nuclear Regulatery Commission, September 21, 1984, from H. Schierling (NRC) to G. Knighton (NRC),
Subject:
Summary of Meeting with PG&E on Diablo Canyon Unit 2 September 13, 1984.
02/27/85 12-3 DIABLO CANYON SSER 29 SEC 12
- ~ - -
e.ew-
, ~pws,.e n -
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I t
4 30.
U.S. Nuclear Regulatory Commission, "Diablo Canyon Low Power License, DPR-76," September 22, 1981.
31.
U.S. Nuclear Regulatory Commission, Memorandum and Order CLI-84-5, " Order Reinstating PG&E's Low Power License," April 13', 1984.
32.
U.S. Nuclear Regulatory Commission, November 29, 1984, from G. W.
Knighton (NRC) to J. D.
Shiffer (PG&E),
Subject:
NRC Trip / Audit Reports - Unit 2.
33.
U.S. Nuclear Regulatory Commission, January 3,1985, from G. W. Knighton (NRC) to J. D.
Shiffer (PG&E),
Subject:
NRC Staff Audits.
34.
U.S. Nuclear Regulatory Commission, January 31, 1985, from G. W. Knighton (NRC) to J. D.
Shiffer (PG&E),
Subject:
NRC Staff Audit.
35.
U.S. Nuclear Regulatory Commission, November 14, 1984, from G. W. Knighton (NRC) to J. D.
Shiffer (PG&E),
Subject:
Piping and Pipe Supports'-
Request for Additional Information.
36.
U.S. Nuclear Regulatory Commission, January 8, 1985, from G. W._Knighton (NRC) to J. D.
Shiffer (PG&E),
Subject:
Request for Additional Infor-mation on Footprint Loads.
37.
38.
U.S. Nuclear Regulatory Commission, December 20, 1984, from G. W. Knighton (NRC) to J. D.
Shiffer (PG&E),
Subject:
Pipe Support Concerns Expressed by Individual.
39.
U.S. Nuclear Regulatory Commission, January 17, 1985, from G. W. Knighton (NRC) to J. D.
Shiffer (PG&E),
Subject:
Anonymous Allegation on Diablo Canyon.
DengA WN{icAllpu Prog rua, 4
40.
Pacific Gas and Electric Company DCP Phase I, Final Report.
y c g i,,,.iqigg3 po, y o, Sg g gy ( p g., g) ;,
D G. ecs w (u u ), L (,J h 02/23/55 12-4 DIABLO CANYON SSER 29 SEC 12 7
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t c
- 41. " Torsion of Rolled Steel Sections in Building Structures," by John G.
Hotchkiss, AISC Engineering Journal, January,1966.
- 42. " Seismic Analysis of Structures and Equipment for Nuclear Power Plants,"
Topical Report No. BC-TOP-4-A, Rev. 3, Nov. 1974, Betchel Power Corp.
San Francisco,' California.
- 43. Pacific Gas and Electric Company, June 18, 1982, from P. A. Crane, Jr.
(PG&E) to F. J. Miraglia, Jr. (NRC),
Subject:
Diablo Canyon Project Quality Assurance Program.
Pacific Gas and Electric Company, August 13, 1982, from P. A. Crane, Jr.
(PG&E) to F. J. Miraglia, Jr. (NRC),
Subject:
Diablo Canyon Project Quality Assurance Program.
Pacific Gas and Electric Company, December 21, 1982, from P. A. Crane, Jr.
(PG&E) to G. W. Knighton, (NRC),
Subject:
Diablo Canyon Project Quality Assurance Program.
l 44.
U.S. Nuclear Regulatory Commission, January 26, 1983, from D. G. Eisenhut, (NRC) to P. A. Crane, (PG&E),
Subject:
Diablo Canyon Project Quality Assurance Program.
02/28/85 12-5 DIABLO CANYON SSER 29 SEC 12
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s( 3 [O 9hd NUCLEAR REGULATORY COMMISSION
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UNITED STATES 3*
.E WASHINGTON, D. C. 20555 f 4. [h h 12,1985 Docket No. 50-275 MEMORANDUM FOR: Thomas M. Novak, Assistant Director g gh for Licensing Division of Licensing
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FROM:
Don H. Beckham, Acting Deputy Director Division of Human Factors Safety
SUBJECT:
DIABLO CANYON POWER PLANT, UNIT 1 TEST PROGRAM CHANGES - LICENSE CONDITION 2.C.(3) 4 The licensee requested a change to its initial plant test program by letter dated March 11, 1985, from J. D. Shiffer (PG&E) to G. W. Knighton (NRC). The change is requested for Power Ascension Test No. 5.2, Radiation Surveys and Shielding Effectiveness.
The licensee has requested that radiation surveys at 90% power be deleted from their power ascension program on the basis that adequate radiation survey data has been taken at 30%, 50%, and 75% power to confidently predict radiation levels at 100% power. Furthermore, additional surveys will be made at 100% power to complete the test requirements. The licensee also points out that repeating the radiation surveys at 90% power would add unnecessarily to the radiation exposure of test personnel.
.\\
Regulatory Guide 1.68, Power Ascension Test No. 5.b.b, recommends radiation surveys only at 50% and 100% power. Based on this guidance and the predictability of radiation levels at full power, the staff concludes that the deletion of radiation surveys at 90% power is technically acceptable.
Therefore, pursuant to License Condition 2.C.(3) of License DPR-80, we hereby approve the deletion of radiation surveys at 90% power from the Initial Test Program of the Diablo Canyon Power Plant, Unit 1.
The reviewer for this change was Sam MacKay (X28147)/PSRB. The reviewer is not aware of any " Differing Professional Opinions" relating to this review.
Don H. Beckham, Acting Deputy Directo P Division of Human Factors Safety cc:
G. Knighton H. Schierling g sc g
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