ML16341A518

Safety Evaluation Report Related to the Operation of Diablo Canyon Nuclear Power Plant,Units 1 and 2.Docket Nos. 50-275 and 50-323.(Pacific Gas and Electric Company)

ML16341A518
Person / Time
Site:	Diablo Canyon
Issue date:	03/31/1985
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0675, NUREG-0675-S29, NUREG-675, NUREG-675-S29, NUDOCS 8503280011
Download: ML16341A518 (72)

Text

NUREG-0675 Supplement No. 29 Safety Evaluation Report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 Oocket Nos; 50-275 and 50-323 Pacific Gas and Electric Company U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation

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ABSTRACT Supplement 29 to the Safety Evaluation Report for the application by the Pacific Gas and Electric Company to operate the Diablo Canyon Nuclear Power Plant-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 PG8E Internal Review Program for the Unit 2 applicability and resolution of concerns that had been raised during the Unit j. design verification by the Independent Design Yerification Program, PGLE by the applicant's Internal Tech-nical Program and by the staff of the U.S. Nuclear Regulatory Commission.

Diablo Canyon SSER 29

TABLE OF CONTENTS Abstract.....

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Table of Contents...

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INTRODUCTION AND BACKGROUND......

1.1 1.2 1.3 1.4 Unit 1 Design Verification..

Allegatlons.......................

Piping and Pipe Supports.............

Recent Unit 1 Licensing History............

1-1 1-2 1-2 1-3 2.

INTERNAL REVIEM PROGRAM FOR UNIT 2...............................

2.1 Unit 2 Engineering Project Organization....

2.2 Internal Review Program....................

2.3 Internal Review Program Results 2.4 Staff Evaluation and Conclusion............

2.5 Ongoing Efforts..'

3.

SEISMIC DESIGN ASPECTS OF CIVIL STRUCTURES 2-1 2-1 2-4 2-5 2-8 3-1

3. 1 Introducti on..

3. 2 NRC Concerns from Unit 1..

3"1 3-2 3.2. 1 Resolution for Common Structures..

3. 2. 2 Resolution by Same Methodology..

3.2.3 Resolution Related to Turbine Buildi 4.

CONTAINMENT ANNULUS STRUCTURE................

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3-2 3-3 3-7 4.1 Introduction..

4-1 4.1.1 Comparsion of Unit 1 and Unit 2.....

4. 1.2 Unit 1 and Unit 2 Modifications.....

.2 Scope of Review.............................................

4 4-3 4.2.1 4.2.2 4.2.3 4.2.4 4.2.5 4.2.6 4.2.7 4:2.8 Vertical Seismic Evaluation.

Horizontal Seismic Evaluation..

20 Hz Cutoff.............

Connection Evaluation.

Member Evaluation........

Tangential and Radial Beam Evaluatio Column Evaluation........

Torsion Evaluation..

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4"3 4"4 4-5 4-6 4-6 4-7 4-8 4-8

4.3 Findings

and Conclusions.............,......................

4-8 Diablo Canyon SSER 29

'5.

'TURBINE BUILDING;..

TABLE OF CONTENTS (Continued)

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5.1 Introduction..

5.2 Scope of Review.

5.2. 1 Specific Calculations 5.2.2 Strength Evaluation for Concrete Floor at Elevation 140 Feet................

Diaphragm 5-1 5-2 5-2 5-4

5.3 Findings

and Conclusions..

5.3. 1 Independent Evaluations

5. 3. 1. 1 Shear Wall Analysis..

5. 3. 1. 2 Buttress Model 5.3. 1.3 Floor Diaphragm Elevation 119 5.3.2 Conclusions...

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5-4 5-4 5-5 5-5 5-7 5-8 6

RACEWAYS..............................

6-1

6. 1 Introduction.

6.2 Scope of Review.

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6.3 Findings

and Conclusions 7.

BURIED CONDUITS 6-1 6-1 6-2

7. 1 Introduction.

7.2 Scope of Review...

7.3 Findings

and Conclusions 8.

PIPEWAY STRUCTURE 7-1 7-1 7-2

,8. 1 Introducti on.

8.2 Scope of Review........

8. 3 Findings and Conc~ usions.

9.

PIPING SYSTEMS AND PIPE SUPPORTS 10.

NON-SEISMIC DESIGN ASPECTS.

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8-1 8-1 8-2 9-1 10-1

10. 1 Systems.................

10.1. 1 Component Cooling Water System..

10. 1.2 Protection from Jet Impingement Moderate Energy Line Breaks.....

10.2 Jet Impingement Analyses..;..........

10.3 Equipment qualification.

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Due to 10-1 10"1 10-1 10-2 10-3

'Diablo Canyon SSER 29

TABLE OF CONTENTS (Continued) 10.3. 1 Environmental guali'fication..

10.3.2 Main Annunciator Typewriter Seismic qualification....

11.

(UALITY ASSURANCE.

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REFERENCES..

Pacae 10-3 10-3 12-1 Diablo Canyon SSER 29

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 (PG8E) to. operate the Diablo Canyon Nuclear Power Plant, Unit -1 and Unit 2 (Ref. 1).

The SER was supplemented by Supplements No.

1 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'upplemented by Supplements No.

18 through 27 to present the staff's safety evaluation of certain issues that had been raised, principally with respect to Unit 1 (Ref.

4 through Ref. 13).

The subjects addressed in these Safety Evaluation 'Report Supplements (SSERs) are the Unit 1 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.

(Note:

SSER 28 and SSER 30 are in preparation and will be issued at about the same time as this report.)

This is Safety Evaluation Report Supplement No.

29 (SSER 29) and it addresses the staff's evaluation of the application, implementation and results of the Unit 1-design verification effort with respect to Unit 2.

SSER 29 also provides an overview of the staff's evaluation of the other matters listed above with respect to Unit 2.

l. 1 Unit 1 Desi n Verification 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 the Unit 1 low power Operating License No.

DPR-76 (Ref. 24).

The order required the completion of an Independent, Design Verification Program (IDVP) for seismic service-related contract design activities prior to reinstatement of the suspended license.

. In addition, the Director of the Office of Nuclear Reactor Regulation requested an extension of the IDVP to non-seismic serv'ice-related contract de-sign activities and PGLE internal design activities (Ref. 17).

These had to be completed prior to issuance of a full power license for Unit 1.

The objective of the IDVP was to verify the adequacy of the Unit -1 design in light of serious questions regarding the adequacy of the implementation of the PGKE quality assur-ance program with regard to design activities.

The IDVP was conducted under the management of Teledyne Engineering Services and by organizations independent of PG&E.

In late 1982 the IDVP was expanded to also include selective verification of construction quality assurance efforts.

In addition, PG8E initiated its own Internal Technical Program (ITP) under which appropriate actions were taken to resolve issues identified by the IDVP and, more importantly, which expanded the scope of design verification to include all seismic safety-related structures, systems and components.

The staff safety evaluation of the adequacy of the IDVP 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 1 licensing process, i.e., prior to fuel load, criticality, and exceeding the 5 percent power level.

SSERs 19, 20, and 24 presented the staff evaluation of the resolution of the issues in support of the various licensing actions.

Diablo Canyon SSER 29

The Atomic Safety and Licensing Appeal Board (ASLAB), on the basis of a hearing on motions to reopen the record on construction quality assurance, concluded 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, that on the basis of the design verification program, 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 resolved in SSER 16 and License Amendment No.

8 with respect to the first 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 in the Licensing Board's Initial Decision, LBP-82-70 (Ref. 37), 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-lution of issues that had been raised during the Unit 1 design verification.

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 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 the design, construction, and operation of the Diablo Canyon Nuclear Plant and the PG&E management of these activities.

In late 1983 the NRC staff instituted the Diablo Canyon Allegation Management Program (DCAMP), which provided specific procedures for the evaluation of allegations and for their resolution (Ref. 21).

Many'of the allegations were contained in affidavits and exhibits in support of petitions by the Government Accountability Project (GAP) submitted to the NRC pursuan ursuant to the provisions of 10 CFR 2.206.

As of the end of 1984 approximately 1650 allegations had been received by the NRC.

The staff evaluation of the allegations with respect to their safety significance for the Unit 1 licensing decisions, is presented in SSERs 21, 22, 26, and 28.

I late 1984, PG&E established an allegation review program to track all allega-tions identified in the NRC program, to determine their applicability and impac on Unit 2 and to ensure that all commitments or modifications resulting from the Unit 1 effort have been appropriately addressed and implemented for Unit 2.

The PG&E program does not address those allegations that are based on confiden-tial or anonymous information.

The program was described in the PG&E Final Report of December 5,

1984 (Ref. 22).

The staff is currently completing its evaluation of the applicability of all allegations to Diablo Canyon Unit 2.

The staff has found that in many cases the same evaluation and resolution is equally applicable to both units.

The staff will document its conclusion in Aa report prior to a Unit 2 license decision.

1.3 Pi in 'and Pi e

Su ort Beginning in late 1983 allegations were identified to the NRC pertaining to piping and pipe supports, including the engineering and design, practices within Diablo Canyon SSER 29

the Onsite Project Engineering Group (OPEG) at the Diablo Canyon site.

The 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 from fully reinstating the suspended low power license for Diablo Canyon Unit 1 (Ref. 23).

The group also identified seven conditions which were included in the reinstated low 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

PG8E, 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 a Board Notification in September 1984 (Ref. 25).

The existing PG8E piping and support effort for Unit 2 was redirected in early 1983 to take into consideration the results of the Chen ongoing Unit 1 design verification effort and again in early 1984 to be responsive to allegations in this area and to the Unit 1 low power license conditions.

The final results of the effort are presented in the PG8E submittal of January 31, 1985 (Ref. 26).

The results of the staff effort, including its review and evaluation of the PG8E Final Report, are presented in Section 9 of this report and the details will be the subject of SSER 30.

1.4 Recent Unit 1 Licensin Histor 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 1 licensing process for reinstatement of the low power license, DPR-76 (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, 1984, after the lifting of a stay imposed by the U.S. Court of Appeals (Ref. 27).

Diablo Canyon SSER 29 1-3

2 INTERNAL REYIEW PROGRAM FOR UNIT 2 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 others 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-related structures (containment

annulus, turbine missile shield for auxiliary feedwater pump, structural steel for fuel handling, building, turbine building), safety-related balance-of-plant fluid systems (due to different power levels, unit specific arrangement and equipment ratings),

and safety-related electrical systems (diesel generator and loads for dc systems).

The differences have been identified in the common FSAR for Units 1 and 2 and were addressed, as appropriate, in the Safety Evaluation Report of October 1974 and Supplements 1 through 17.

They are also discussed in PG&E submittals of October 6, 1983 and July 31, 1984 (Ref. 28).

2. 1 Unit 2 En ineerin Pro ect Or anization 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 Franc'isco.

In general, the organizational structure is similar to the Diablo 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 engineering.

In addition, the Unit 2 Internal Review Program (IRP) was estab-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 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 both units.

In many cases a particular issue was analyzed and resolved by 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 in 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 were used for only one unit; however the same management and technical supervision was re-sponsible for both units.

2.2 Internal Review Pro ram 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 Diablo Canyon SSER 29 2-1

submittal of October 6, 1983, and subsequently described the. program details 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 IDVP and ITP; (2) were identified by the staff and addressed in SSERs 18, 19, 20, and 24; (3) were the subject of License Condition 2. C.(ll) in the Unit 1 low power license regarding 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 1 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 1 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 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.

~Ste 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 as part of the Unit 1 verification program.

No further IRP or engineering effort is required.

4.

Unit 1 and Unit 2 are identical with respect to the subject item.

The Unit 1 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 closeout of each item was entered into the IRP e

Diablo Canyon SSER 29 2-2

files.

Any item involving a physical modification for Unit 1 was not closed out by these criteri a, except as under Criteri on 3 above.

~Ste 2

Further 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 l.

Unit 1 background information was collected in a Unit 2 IRP review 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 design similarities and diffe-rences.

Individuals involved in the Unit 1 resolution of the item and in the original design and analysis for Unit 2 participated in the evaluation.

~Ste 3

Based on the detailed review in Step 2, the LRE determined the applicab'ility of the Unit 1, 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 1 resolution to meet the Unit 2 design requirements, or documented the closeout 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 procedures the effort is coordinated with Nuclear Plant Operations (NPO).

All considerations and evaluations were documented and included in the review package.

~Ste e

,For all items that required a physical modification to Unit 2, normal engineering procedures were applied.

Qesign change notices (QCN) and other appropriate engineering documents were issued.

Completion of notification was verified, reviewed, and documented in accordance with established project procedures.

~Ste 5

The final implementation of the Unit 2 resolution was documented in the IRP files with an identification of all appropriate documents.

The Unit 2 effort, including modification as necessary, was considered complete after the Unit 2 Project Engineer signed a completion sheet.

Qiablo Canyon SSER 29 2"3

2.3 Internal Review Pro ram Results PG8E evaluated the 414 items that resulted from the Unit 1 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.

The other 269 items were assembled in Step 2 in 195 IRP review packages for further review.

During further detailed evaluation under Step 3 by the Lead Review Entity and the cognizant engineering groups, it was determined that for 185 of these items the Unit 2 resolution is the same as for Unit 1, for 81 items another 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 on initial review Same resolution for both Units based on detailed review Unit 2 resolution encompassed by other activity Different resolution for Unit 2 145 items 185 items 81 items 3 items The following are the three items with a different resolution for Unit 1 and Unit 2:

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, 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 PG8E s Piping and Pipe Supports Review Program as discussed in Section 1.

On this basis, these items were considered closed out with respect to the IRP.

PG8E submitted the IRP Final Report on November 2, 1984 and further information by letters dated December 7,

1984 and February 21, 1985 (Ref. 28).

PG8E has stated that the engineering resolution for all items has been completed and necessa'ry modifications are either complete or well underway.

In the Final Report PG8E has committed to complete all modifications prior to Unit 2 fuel loading.

Diablo Canyon SSER 29 2-4

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).

Mhile there was no expressed intent by the staff at that time to also require such a program for Diablo Canyon Unit 2 the staff 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 1 low power license was issued (Ref. 30), that unit had been upgraded for the Hosgri event with respect to seismic analysis, design and modifications.

PGLE had devoted its resources to complete Unit 1 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 program for Unit 1 PG8E con-centrated its efforts on this unit, in particular during 1982 and 1983.

Mhen the Unit 2 activities began to increase in 1983, these efforts were not a veri-fication of work that already had been completed, but were largely the initial Unit 2 Hosgri upgrade efforts.

The results of the Unit 1 design verification program were considered at the same time.

Therefore, while the design verifi-cation for Unit 1 was concentrated on specific issues, the same effort for Unit 2 more closely followed normal engineering procedures rather than verifi-cation of already completed work.

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 als'o 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.

1 through 17.

The Independent Design Verification Program (IDVP) as required by the NRC for Unit 1 and the expansion of that program through the PG8E Interal Technical Program (ITP) for Unit 1 verified the design adequacy for Diablo Canyon Unit 1.

This included extensive reanalyses in particular with respect to seismic design considerations and in the area of piping and supports.

Modifications were made as necessary.

The staff did not require an additional independent design veri-fication program for Diablo Canyon Unit 2 for the following reasons.

First, a

number of civil structures are common to both units and the Unit 1 design veri-fication applies equally to Unit 2.

Secondly, as discussed earlier, both units are nearly identical and the same design and analysis methodology was applied to both units.

Thirdly, the PGEE Unit 2 analysis effort since the beginning of the Unit 1 design verification effort was not limited to evaluating specific findings from the Unit 1 effort with respect to Unit 2, but, as in the case of Unit 1 involved the complete reanalysis of major structures and systems.

In this sense the Unit 2 effort can indeed be considered a design verification.

Furthermore, the staff review and evaluation of the Unit 2 seismic design also Diablo Canyon SSER 29 2-5

included some limited independent analyses and was broader in scope in the area of piping systems and piping supports.

The non-seismic analysis and design for Unit 1 and Unit 2 was one single effort and the same licensing criteria were applied to both units.

This effort had essentially been completed also for Unit 2 in late 1981 when the design veri-fication effort for Unit 1 was initiated.

Therefore, the findings from the Unit 1 design verification effort in non-seismic areas apply equally to both units and the staff considers it unlikely that concerns unique to Unit 2 do exist.

The staff has concluded that the Unit 2 non-seismic design has been adequately confirmed by the IDVP/ITP for Unit 1 as implemented through the IRP.

At a meeting on September 13, 1984 with the NRC technical staff and management PG8E 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 1 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 PGLE Diablo 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 and appro-priate bases had been provided.

The staff reviewed applicable DCP and IRP internal procedures and discussed them with IRP personnel.

The detailed techni-cal audits included a review of calculation packages,

drawings, computer codes, including inputs and outputs, and all other pertinent background and documenta-tion.

During these audits the staff discussed selected areas with PG8E, which in some cases led to specific requests for additional information (Ref.

35, 36).

The staff audits and inspections with regard to the programmatic and technical aspects of the IRP were 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 activities 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 that were identified as requiring a resolution for Unit 2 different 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 1 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 and wires; (5) buried elec-trical conduits that carry electric power and control cables from the turbine Diablo Canyon SSER 29 2-6

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 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 buried conduits was inadequate.

PG8E has since initiated necessary documenta-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 is resolved.

The staff's evaluation of the Unit 2 pipeway steel support structure (designed by PG8E) 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 full power operation.

The staff is currently evaluating the applicability of this concern for the pipeway structure of Unit 1 which was designed by Westinghouse.

During the staff review of the Unit 2 aspects of the turbine building PG8E 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 PG8E evaluation of this issue has not been completed and the staff review of this matter is still ongoing.

As discussed in Section 5 of this report, the staff requires that this matter be resolved prior to full power operation.

The detailed 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; however, the scope of the Unit 2 effort was larger than for Unit l.

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 veri-fication, (2) actions resulting from a license condition on this subject in the Diablo Canyon SSER 29 2-7

Unit 1 low power license, and (3) the Unit 2 applicability and resolution of allegations related to piping and supports.

The NRC team, including consultants, audited in excess of 60 IRP packages and approximately 100 piping and pipe support calculations.

In addition, the staff met with a confidential alleger'ho identified specific concerns regarding pipe supports.

These concerns and other anonymous allegations also were included in the staff's review effort (Ref. 38, 39).

The staff identified a concern pertaining to the pressure design adequacy of certain branch connections (Ref. 37).

PG8E has committed to identify all such branch connections and provide verification to show compli-ance with the applicable design code.

The staff finds this acceptable.

The staff concludes that the Unit 2 piping systems and pipe supports meet the appli-cable design criteria.

Based on its review and evaluation, including audits and inspections, the staff finds that the IRP was an acceptable program which was effective in identifying, documenting and resolving the Unit 1 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 verifica-tion 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 eval-uation 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.

2. 5 O~iEff This report identifies the following two areas where the PG8E and/or the NRC staff evaluation efforts are still ongoing:

1.

l~li B ilN PG8E is continuing the evaluation of the shear friction along a construc-tion joint at elevation 140 feet.

The staff concluded that the issue need not be resolved prior to the low-power operation but must be resolved prior to full-power operation (Section 5.2.2).

2.

~Pi ewa The staff is awaiting additional information pertaining to the seismic analysis of the Unit 2 pipeway to complete its evaluation.

Further audits will be conducted as necessary.

The staff has concluded that the issue need not be resolved prior to low-power operation but must be resolved prior to full-power operation (Section 8.3).

Diablo Canyon SSER 29 2-8

3 SEISMIC EVALUATION OF CIVIL STRUCTURES

3. 1 Introduction The major common civil structures at the Diablo Canyon Nuclear Plant are the auxiliary building, intake structure, outdoor water storage

tanks, and buried diesel oil tanks; The staff evaluation of the seismic design verification of these common structures has been reported in SSERs 18, 19, 20, and 24 and applies to both units.

Unit 1 and Unit 2 have individual fuel handling build-ings which are supported on a common floor in the auxiliary building.

The major unit-specific structures are the containment, including the annulus structure, and the turbine building.

They are essentially the same in design and construction.

The Unit 2 containment structure is essentially identical to that of Unit 1 except that the annulus structure inside the containment has some structural aspects and modifications which are unique to Unit 2.

The staff performed several audits of the PG&E calculations from the Unit 2 annulus structure.

The staff evaluation is reported in Section 4.0.

The turbine building for the Diablo Canyon Nuclear Power Plant is one contiguous building consisting of two separate civil structures, one for each unit, which are referred to as the Unit 1 turbine building and the Unit 2 turbine building.

These buildings are essentially the same; however, there are certain differences that influence 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 established in the testing program during the Unit 1 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-crete duct banks.

The staff audited PG8E's documentation of the design and construction of the buried conduits.

The staff evaluation is reported in Section 7.0.

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 the former Nuclear Services Corporation (now guadrex).

Therefore, different techniques were employed in the analyses for the pipeway structures in Unit 1 and Unit 2.

Diablo Canyon SSER 29 3-1

The staff audited the PG8E Unit 2 calculations for the Hosgri earthquake; the review and evaluation for the DE and DDE is ongoing.

The details of the staff evaluation, including the current status, on the analyses performed for the Hosgri earthquake is reported in Section 8.0.

3.2 NRC Concerns from Unit 1 As a result of its review and evaluation of 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 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 PG8E resolution as described in the various IRP submittals by PG8E (Ref. 28).

During the Unit 2 audits and inspections the staff also reviewed and evaluated various aspects raised by these

items, as appropriate.

A resolution was reached for each item in accordance with the IRP process discussed in Section 2, either by identifying the commonality of an item for both units or by providing a basis for the same or different approach where the units are different.

3.2. 1 Resolution for Common Struc'tures Staff concerns regarding the PG&E analysis for the auxiliary building were raised in OI 5, 6, and 7.

Staff concerns regarding the PG&E analyses for the intake structure and the buried diesel oil tanks were raised in OI 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 1 resolutions were documented in SSER 19 and SSER 20.

Auxiliar Buildin Floor Slab Flexibilit (OI-5)

For Unit 1 the staff requested PG8E to assess the assumptions used in the auxil-iary building seismic analysis to determine floor slab flexibility.

PG8E modeled the entire auxiliary building (Unit 1 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 1 by generating a 3-D finit'e 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 units, the item is resolved for Unit 2.

ACI Code Justification for Auxiliar Buildin OI-6)

For Unit 1 the staff requested PG8E to justify the use of the ACI Code for evaluating the floor slabs and walls of the auxiliary building.

The concrete floor slabs and walls of the auxiliary building were evaluated against criteria presented in Appendix 2A of the Phase I PG8E Final Report (Ref. 40) rather than the 1963 ACI Code.

The 1963 ACI Code has no specific provision for shear walls but does permit criteria to be used which are based on test data.

Test data were developed by PG&E and presented in Appendix 2A of the PG8E Phase I Final

Report, Design Verification Program.

A review of the Appendix 2A criteria indi-cated that it is conservative relative to the shear wall provisions in the 1977 Diablo Canyon SSER 29 3-2

ACI Code.

The Unit 1 stress evaluation was 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 item is resolved for Unit 2.

Soil S rin Influence on. Auxi 1 iar Buildin (OI-?)

For Unit 1 the staff requested PG8E 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 to produce some soil/structure inter-action.

questions were raised by the staff during the Unit 1 review regarding the soil properties used to generate the interaction spring parameters.

In response to this concern, PG&E performed a parameter study showing that the seismic response of the structure was not sensitive to spring constant varia-tions spanning values corresponding to plausible variation in the soil proper-ties.

Since the auxiliary building is a continuous structure common to both units, the item 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 IDVP reports ITR-40 and ITR-68 (Rev.

0 and 1), discussed the issue with the IDVP, 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 item is required for Unit 2.

Buried Diesel Fuel Oil Tank OI-26)

For Unit 1 the staff requested PG8E to perform additional analyses of the buried diesel fuel oil tanks.

Additional analyses were performed by PG&E =and verification studies were conducted by Brookhaven National Laboratory (BNL), the staff consultant.

Based upon these additional studies, the staff concluded that all safety issues associated with the tanks were satisfactorily resolved.

Since the buried diesel fuel oil tanks are common to both units no further re-view of this item is required for Unit 2.

3. 2. 2 Resolution by Same Methodology The following open items (OI) related to the analytical techniques employed in the PG&E analyses or the design codes used in the PG8E design:', 2, 3, 4, 8, 9, 13, 14, 15, 23, 24, and 31.

Although separate analyses and calculations were performed for Unit 2, they employed the same approach and methodology as was used for the Unit 1 analysis.

The staff audited the calculations for Unit 2 and found that the structures involved in these items are sufficiently similar to those of Unit 1 to justify this approach.

Therefore, the staff concludes that the Unit 1 resolutions are equally applicable to Unit 2.

Free-Hand Avera in of S ectra (OI-1)

For Unit 1 the staff requested PG8E to confirm that the free-hand averaging procedures for spectra used in the containment annulus structure are proper.

Diablo Canyon SSER 29 3-3

Free-hand averaging procedures for smoothing the floor response spectra were used in the low frequency range in which no structural frequencies exist.

PGLE used this method for seismic evaluation of both units.

Justification for the method was provided by PG8E during the Unit 1 review and was accepted by the staff.

Similar justification for Unit 2 annulus is, therefore, not necessary since the same methodology was applied to both units.

20 Hz Cutoff Fre uencies (OI-2)

The 20 Hz cutoff criterion was used in the horizontal seismic evaluations of the Unit 1 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 PGLE to provide a justification for the 20 Hz cutoff criterion.

This item was latter resolved for the Unit 1 annulus structure in SSER 19.

PGRE performed a study for the Unit 2 annulus structure similar to the Unit 1 study to justify the 20 Hz cutoff criterion.

The same modeling technique and 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 (OI-3 For Unit 1 the staff requested PG&E to assess the applicability of the AISC and ASNE (Section III) Codes to the design of penetrations in the containment shell In SSER 18 the staff stated that the use of the AISC Code for the containment penetration analysis should be justified.

The resolution of this item was presented in SSER 19.

The containment design was shown to satisfy both codes for Un'it 1.

The Unit 1 and 2 containment designs are essentially the same in the vicinity of the penetration.

Therefore since the Unit 1 design was shown to satisfy the requirement of both the AISC and ASME (Section III) Codes, the staff considers this item to be resolved for Unit 2.

Stress Levels at 0 enin s in Containment Shell OI-4 For Unit 1 the staff requested PG8E to assess the adequacy of the equipment hatch opening in the containment structure.

The staff requested 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 review of Unit 1 the staff considered the plate analysis acceptable for Unit -1.

The staff reviewed the design and analysis of the hatch opening during thy audits for Unit 2.

The staff found that the design and analytical approaches are identical to that for Unit 1 and the results were within the code allowables.

Thus, the staff concludes that the Unit 2 analysis is acceptable.

Seismic In ut Motion to Fuel Handlin Buildin OI-8)

For Unit 1 the staff requested PG5E to document the use of the auxiliary build-ing motions as input to the fuel handling building seismic analysis.

The Unit 1 and Unit 2 fuel handling buildings (FHB) are one contiguous steel structure Diablo Canyon SSER 29 3"4

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Allowable Bolted Joint Loads for Turbine Buildin OI-15)

For Unit 1 the staff requested PG8E 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 connec-tions were based upon the 8th Edition of the AISC Code rather than the 7th Edition.

The 8th Edition permitted higher bearing stresses than the 7th Edition of the Code.

PG8E 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 loads in the 7th Edition neglect any frictional forces which occur between the plate of the joint.

Since the bolts are tightened to 70 percent of yield strength, frictional forces are 'generated between the plates.

PG8E 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 there-fore accepted for Unit 1.

Since the joints are essentially identical in Unit 2, this item is therefore reso'Ived for Unit 2.

Cable Tra uglification OI-23 The staff identified a concern for Unit 1 regarding the qualification of the cable trays and the effect of the interaction between cable trays and their supports.

PG&E stated that the cable trays were qualified generically for the DDE and the Hosgri event.

Mhere the trays could not be qualified generically the as-built conditions were analyzed.

The supports were evaluated by two separate

analyses, one considered the support itself, the other considered a

coupled system.

The staff found the approach acceptable for Unit 1.

PG8E applied the same approach for the Unit 2 cable tray evaluation.

The staff finds this acceptable and the item is resolved for Unit 2.

Su erstruct Melds (OI-24)

For Unit 1 the staff requested PG8E to verify that the shear demand for the welds used in the superstrut construction met the allowable limit established in the testing program.

Since the samples in the test program were taken from both units the allowable weld strength of superstrut construction established in the tests is also applicable to Unit 2.

The staff also audited raceway cal-culations for Unit 2 and determined that the shear demand for the welds is within the established limits.

The staff concludes that this item is resolved for'Unit 2.

Combination of Three Earth uake Com onents OI-31 For Unit 1 the staff requested PG8E to clarify methods of combining directional responses in the turbine building.

The three directional earthquake components were combined by PG&E for the turbine building by adding the full value of the largest component to 40 percent of each of the other two components.

Based on a study by PG8E this method results in conservative values relative to SRSS combination except when two components are equal and the third component is zero.

In that case the SRSS combination is about 1 percent higher than the method used by PG8E.

The staff accepted this method of combination for the turbine building during the review for Unit 1.

The same method was used for Diablo Canyon SSER 29 3-6

the evaluation of the Unit 2 turbine building.

Therefore this item is resolved for Unit 2.

3. 2. 3 Resolution Related to Turbine Building Three open items (OI-10, OI-11, and OI-12) were reviewed and evaluated in detail during the staff's audit of the design and analysis of the turbine building as discussed in Section 5.

Load Combinations for Turbine Buildin (OI-10)

For Unit 1 the staff requested PG8E 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 of dead, live, and seismic loads.

This concern was based on the Phase I Final Report (Ref. 40) which did not indicate that such load com-

binations were used.

During the staff audits member strength calculations for the Unit 2 turbine building 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 following calculations were reviewed:

65-T"235 65-T-ill 65-T-311 shear walls columns buttresses The comparison showed that the capacity of all members exceeded the demand.

The staff considers this item resolved for Unit 2.

Turbine Buildin Roof Truss Model (OI-11 For Unit 1 the staff requested PGEE to provide further information regarding the method of modeling the roof truss by two generalized uniaxial members and obtaining individual truss member response from the uniaxial member model.

The truss system used for the roof truss of Unit 1 was modeled as equivalent uniaxial members per the PG8E Calculation 64-7-275.

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 from the end displacements of the equivalent member.

PG8E Calculation 64-7-291 showed that member loads derived using the equivalent truss member are close to member loads found by using the actual truss configuration.

This was shown by performing static analyses of two models, one using the actual truss configuration and the second using the equivalent truss member and end displacements for generating the internal member forces.

The loads generated from the equivalent model 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 was used for both units.

The staff reviewed the physical differences between the Unit 1 and Unit 2 roof trusses and concluded that the differences were of local nature and would have a minimal effect on the stiffness of the equivalent member.

The equivalent model of the Unit 2 turbine building roof truss is therefore acceptable.

The staff considers this item resolved for Unit 2.

Diablo Canyon SSER 29 3-7

Vertical Seismic Model of Turbine Buildin OI-12)

For Unit 1 the staff requested PGE.E to 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 turbine building.

guestions were raised regarding the coupling between these models.

The coupling effect between the four models was shown to be insignificant and the Unit 1 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 was 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 was judged to be adequate to account for the small differences between the Unit 1 and Unit 2 turbine buildings.

The staff considers this item resolved for Unit 2.

Diablo Canyon SSER 29 3-8

4 CONTAINMENT ANNULUS STRUCTURE

4. 1 Introduction The containment annulus str ucture 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, i.e.,

DE, DDE and Hosgri event.

The details of the PG8E reevaluation 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 Comparison of Unit 1 and Unit 2 During a meeting held in Washington, DC on September 13, 1984, the staff re-quested the Pacific Gas and Electric Company (PGSE) to provide a set of drawings of the annulus structures for both units (Ref. 29).

The objective was to draw conclusions regarding the similarities and differences between the two struc-tures.

The following drawings were audited:

1.

Containment Annulus Structure - Additional Modifications - El. 140',

Unit 2 - 471099, Rev.

1 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. 'nnulus 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 Diablo Canyon SSER 29

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.

Further, 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.

Fur-

thermore, 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 - Radial Frame 1, Unit 2 - 1032C-l, Rev.

0 2.

Dynamic Analysis - Radial Frame 2, Unit 2 - 1033C-1, Rev.

0 3.

Dynamic Annalysis Radial Frame 2.5, Unit 2 - 1034C-1, Rev.

0 4.

Dynamic Analysis - Radial Frame 3, Unit 2 - 1035C-l, Rev.

0 5.

Dynamic Analysis - Radial Frame 3.5, Unit 2 - 1036C-l, Rev.

0 6.

Dynamic Analysis - Radial Frame 4, Unit 2 - 1037C-l, Rev.

0 7.

Containment Annulus - Dynamic Model (DC) - Frame 1 - 625C-1 8.

Vertical Seismic Analysis for Annulus - Frame 2 - (DC'1) - 626C-1 9.

DC ¹1, Annulus Area Revised Radial Frame Model for Frame 2.5, Rev.

627C-1 10.

DCP - Containment Time History Analysis Frame ¹3 - 628C-1 ll.

DC ¹1 - Vertical Seismic Analysis for Frame

3. 5 - 629C-1 12.

DC ¹1 - Seismic Analysis Annulus Area Column Line 4 - 630C-1 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.

4. 1.2 Unit 1 and Unit 2 Modifications During the evaluation of Unit 1, PG8E 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, PG5E modified the Unit 2 annulus structure for the same purpose.

In general, the modifications in both units include:

(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 Diablo Canyon SSER 29 4-2

the use of additional columns, especially at the lower platforms, produced a re-duction in vertical amplifications.

Mhile 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 rather than by strengthening of members.

It was 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, particularly, 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 (Ref. 32).

4.4 ~ff 4 Subsequent to the NRC meeting with PGLE on Unit 2 on September 13, 1984, the staff and its consultants conducted two audits on the Unit 2 PG8E 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 8 33).

After the first audit, the staff and consultants met with PG8E 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 ~

b.

C.

d.

e.f.

Vertical seismic analysis Horizontal seismic analysis 20 Hz cutoff justification Connection evaluations Hember evaluations Torsional evaluations The staff reviewed the seismic analyses performed by PGLE in both horizontal and vertical directions to assess the adequacy of the models and the floor response spectra.

The review concentrated on verifying the staff's concerns expressed in SSER 18 and the results of the previous investigations from the Unit 1 annulus structure were considered in the Unit 2 evaluation.

A sample of the calculations for members and connections were reviewed for compliance with the FSAR requirements.

Details per taining to the review of the above items fol 1 ow.

4.2. 1 Vertical Seismic Evaluation, The same analysis procedures used for the Unit 1 annulus structure were followed by PG8E 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:

Containment Annulus, 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 Diablo Canyon SSER 29 4-3

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 PG8E 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 zero period ground acceleration (ZPA) in accordance with the FSAR requirements.

4.2.2 Horizontal Seismic Evaluation The seismic evaluations performed by PG&E for the Unit 2 annulus structure in the horizontal direction are based on the same procedures as those followed for 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 platform 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 PG8E, the models which cor-respond to the final iteration incorporate the latest design changes in the structure.

The following calculations for the Unit 2 annulus structure were reviewed for the frequency analysis:

1.

Calculations 1002C-1 to 1002C-ll (EL 117')

2.

Calculations 1003C-1 to 1003C-3 (EL 106')

3.

Calculations 1004C-1 to 1004C-3 (EL 101')

The above analytical studies by PG8E 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 spectra developed for the internal concrete structure are applicable to the annulus platforms.

Thus,.subsystems and components attached to the annulus structure are designed for,the same horizontal spectra derived for the internal concrete structure which ar'e 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 an'd 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, respectively, were used for the qualifications of piping systems attached to the annulus structure in both units.

Diablo Canyon SSER 29

4. 2.3 20 Hz Frequency Cutoff During the reevaluation of the Unit 1 annulus structure the staff was concerned with the 20 Hz frequency cutoff criterion (Ref. 4) because it did not reflect the Hosgri criteria.

The staff requested PG8E to provide a justification for this criterion.

Based on the information provided by PG8E 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 PG8E 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:

l.

Diablo Canyon Unit 2, Containment Annulus Structure, 20 Hz Frequency

Study, Calculation No.

1140C-1.

2.

Calculation No. 900C-l, Rev.

0, 20 Hz Frequency Study, Generation of Com-posite Damping, Multiple Support Method of Response Spectra.

3.

Calculation No.

900C-2, Rev.

0, 20 Hz Frequency Study, Horizontal Model, Calculation of Column Stiffnesses at EL 106'.

5.

Calculation No. 900-3, Rev.

0, 20 Hz Frequency

Study, Development of Horizontal Model, Interior Structure and Annulus Structure at EL 106'.

Calculation No.

904C-1, Rev.

0, 20 Hz Frequency Study - Horizontal Model, EM Earthquake Time-History Analysis (Newmark Hosgri), Output Accelerations at Annulus Pipe Supports for Lines K17-2314-3 and SG-3060-4.

6.

PG8E Nuclear Plant, Diablo Canyon Site Unit 2, M-3904 and M-5401 Hosgri Study.

PG8E performed a parametric 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 l. It was shown that the modes with frequencies greater than 20 Hz did not have a significant effect on the floor response spectra.

The staff requested PG8E to extend the 20 Hz cutoff frequency studies to eval-uate the annulus members.

In order to demonstrate the effects on the annulus

members, the increment in support loads due to the increased cutoff 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 commit-ment.

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, while in actuality only the seismic stress of the member should have been increased.

This is a conservative approach.

Nevertheless, using this approach, the majority of the annulus steel members Diablo Canyon SSER 29 4-5

were found to satisfy the allowable stresses.

The calculated stresses in some

members, however, did exceed the allowable stress.

For example, for beam ¹49, the new factor was found to be 1.018, i.e., 1.8X 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 Connection Evaluation 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 are connected to the column with moment connections while the beam to beam uses shear connections.

PG5E performed selective verification studies by which the results generated by CONPAS 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 and 1006C-2.

All desig'n loads were less than the allowable loads.

4.2.5 Nember Evaluation The circular containment annulus structure consists of steel tangential

beams, radial beams and columns.

The PG8E calculations reviewed by the staff show that the strength checks of all the members were performed by the computer program GTSTRUDL.

For each member the static analysis results were combined 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.

Tangential Beam Evaluation:

A-1 A-2/A-3 Beam Location Drawing, EL 101',

squad

'B'members B1266 and B1276)

Calculation No. 1005C-l, Rev.

0, pgs.

1 and 2, Parameters for Code and Check by GTSTRUDL Diablo Canyon SSER 29

Calculation No. 1005C-l, Rev. 0, Rev. 0, pg.

6, calculation to deter-mine the CODETOL parameter for GTSTRUDL (see page 2.2.5-130 of Attachment E)

A-5 Calculation No. 1005C-1, Rev.

0, p.

11 (prep.

sheet for GTSTRUDL input)

A-6 Beam Input - GTSTRUDL 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 Beam Evaluation:

B-l Computer, plot with beam identification, EL 101',

squad

'B'-2 Calculation No. 1005C-1, Rev.

0, p.

10, parameter input code check B-3/B-8 Member forces for code check B-9/B-13 DE, DDE, and Hosgri,code check C.

Column Evaluations:

C-1/C-3 C"4/C-8 Calculation No. 1009C-1, Rev.

0, pgs, 8, 9, 30 GTSTRUDL data for column R15 and $15 4.2.6 Tangential and Radial Beam Evaluation 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 PG8E.

The program GTSTRUDL computes 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 AISC 1978 specifications.

The staff 'reviewed the calculations and performed some independent checks of the combined stress resulting from major forces of several members.

It is con-cluded that the code check procedure is satisfactory.

Diablo Canyon SSER 29

4.2. 7 Column Evaluation The exterior columns of the annulus frame are shown in Calculation No.

1009 C-l, 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 categories 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 forces.

In this case more elaborate computations were made.

The staff reviewed these combination procedures and found them satisfactory.

Beams with large torsional moments were checked in detail by hand computations.

Specifically, the following three beam computations were reviewed:

H2129D at elevation 106 feet, A2216 at elevation 101 feet, and the beams F2282 and F3274 on column line 14 at elevation 106 feet.

Beam H2129D is a W12Ã79 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 the 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 was based on an incorrect formula, an independent computation done by the staff indicates the strength is still satisfactory.

In addition, PG&E calculation showed that the side plate would be extended to the face of the column.

It is therefore concluded that the strength of the beam A2216 is satisfactory.

The same incorrect formula was used 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 nonsymmetrical bending on the stress was neglected in the calculation.

An independent computation performed by the staff to include the nonsymmetrical bending indicates that the beams are satisfactory.

4.3 Findin s and Conclusions Based on the staff's audit and detailed review of extensive information, the staff concludes for the analysis and design of the Unit 2 containment annulus structure the following:

Diablo Canyon SSER 29 4"8

a 0 b.

The analytical procedures used by PG8E for the Unit 2 annulus structure are essentially the same as those used for Unit 1.

Concerns raised during the Unit 1 annulus structure verification were considered and resolutions were implemented in Unit 2.

For the horizontal and vertical seismic evalu-

ations, the models, methodologies and criteria followed by PG8E are found to be acceptable.

With respect to the 20 Hz frequency cutoff criterion, PG8E demonstrated in its studies that modes in the range of 20-33 Hz do not have significant impact on the annulus structural loads.

Based on the margins associated with the design, the annulus structure remains qualified for the addi-tional amplifications.

However, this conclusion does not imply generic applicability.

For other configurations the effect of the modes above 20 Hz may not be the same.

c.

The design of the individual members without torsion using the computer program GTSTRUDL is found to be satisfactory.

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 torsional moments.

This particular beam and one other beam using the same formula were checked by a more appropriate formula and found to be acceptable.

Diablo Canyon SSER 29

yN

5 TURBINE BUILDING

5. 1 Introduction The turbine buildings for Unit 1 and Unit 2 are two separate structures that form one contiguous building, 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 con-crete diaphragm members are provided interior to the structure at elevations

'elow 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 concr'ete 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).

2.

3.

The on-site technical support center is located at the south end of the westside wall of the Unit 2 turbine building and connected to it.

There is no such comparable structure attached to the Unit 1 turbine building.

It There are relatively small differences in equipment masses between the two structures.

4.

Modifications were made to locally increase the qtiffnesses of the Unit 1 structure.

These same changes were also made to the Unit 2 structure although in some 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 along the west wall of the Unit 2 turbine building is slightly different than the same area for the Unit 1 turbine building.

The differences are in general quite, small and these differences 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 usi ng an equivalent beam representing the roof truss properties.

This beam connected the columns on the east side of the Diablo Canyon SSER 29 5-1

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

~5*

f R

The emphasis of the staff's evaluation was focused on those aspects which are unique to Unit 2 turbine building.

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 calculations 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 evalua-tion.

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 evaluation of specific calculations.

5.2. 1 Specific Calculations Calculation 64-T-275 This calculation presents a consolidated description of the Unit 1 turbine building 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 turbine building 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 contained in the calculations.

Supplemen-tal 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 the Unit 2 turbine building.

Calculation 64-T-278 The differences between Unit 2 and Unit 1 turbine buildings were considered in this calculation.

The major part of the calculation was concerned with comput-ing the mass differences between the units.

Specific differences were identi-fied with the mass values for each of the units tabulated.

The total mass differences were less than 5X with local differences being less than lOX.

Calculation 64-T-283 The model for Unit 2 turbine building was developed in this calculation.

The data contained in the above two calculations was the basis for the new model.

A mirror image of the Unit 1 model was used to form the geometry of the Unit 2 model with the 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 masses.

New structural member properties were used where the struc-ture is different.

Diablo Canyon SSER 29 5-2

Calculation 65-T-211 The north wall of the Unit 2 turbine building was modeled with plane stress elements in the large model of the entire turbine building so that out-of-plane effects were not included in the model.

Calculation 65-T-211 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 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 begin 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.

The results of the additional staff evaluation is presented in Section 5.3. 1.3 below.

Calculation 65-T-235 Stresses in the shear walls along the east and west ends 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 demand for the wall.

The loads were compared with the wall capacities as determined below.

Shear capacity of the wall is deter-mined in accordance with the ACI Code as V = (s ~ +

pf ) x area c

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.

guestions 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 identified as requiring further investigation.

The result of the staff evaluation is presented in Section 5.3.1.1.

Cal cul ation 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 capacity of the column was evaluated based on the AISC criterion:

'm "x cr (1-p

)M P px C

N (lp-P

)M cr In cases where the column does not meet this criterion the required member ductility, 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

.Diablo Canyon SSER 29 5-3

+ +

P x

yx M

yx 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. l.

Calculation 65-T-311 The adequacy of the buttresses was evaluated in this calculation.

The buttress structure and steel column were treated by PG8E 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.

The caisson capacity required in these buttress calculations were lower than the frictional capacity of the caisson, as determined from a tension pullout computation.

The results of the additional staff evaluation is pre-sented in Section 5.3. 1.2 below.

5.2.2 Strength Evaluation for Concrete Floor Diaphragm at Elevation 140 Feet During the January 15-17 audit PGKE 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 lines 19 to 20 at elevation 140 feet.

A check of shear friction along this construction joint by PG8E indicated that a problem might exist.

Further studies are being performed by PG&E to determine whether the p'roblem 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.

PGSE is continuing its evaluation of this issue and the staff will continue its audit and review effort.

Because the concern relates to only one location and because only limited category I equipment is located in the affected area the staff concludes that this issue need not be resolved prior to low-power opera-tion but must be satisfactorily resolved prior to full-power operation.

5.3 Findin s and Conclusions 5.3. 1 Independent Evaluations As discussed

above, the following three areas were selected for further independent staff evaluation:

shear wall analysis; computer simulation'f buttresses; and analyses of horizontal floor diaphragm at elevation 119 feet.

Diablo Canyon SSER 29 5-4

5.3.1.1 Shear Wall Analysis Stresses in the shear walls of the turbine building were evaluated by PG8E in calculation 65-T-ill, 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.

PG8E 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 PG8E 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.

The shear wall spans between the webs of the columns at lines 28 and 29.

The horizontal wall reinforcing bars (8 8 8 12" each wall face) were welded to the web of the steel column.

The wall's continuous through the elevation 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 PG8E. model of the turbine building includes the buttresses by lumping their stiffness with that of the column and placing the "composite" member at the center line of the column.

Diablo Canyon SSER 29 5-5

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 PGLE.

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 PG8E 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 Displ. at El-193'inch)

Displ. at El-140'inch)

Displ. at El-119'inch)

Displ. at El-104'inch)

Max. Stress in Column (ksi)

Buttress Shear Stress (ksi)

Buttress Normal Stress {ksi)

DCP Model

8. 63

1. 506

0. 086

0. 045

31. 6

0. 167

0. 883 BNL Model

9. 69

1. 978

0. 133

0. 063

30. 6

0. 360

0. 672 As can be seen the column stresses predicted by the PG8E model are more severe than those predicted by the more exact model.

The PG8E model predicts lower shear stresses and higher normal stresses than the BNL model.

When these stresses are combined to obtain principal stresses, the PGKE model gives higher principal stresses than the BNL model.

Since stress evaluation is made against the principal stresses, the PG8E model is therefore conservative.

It may also be noted that the PG8E 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.

Frequency Comparisons Mode PG8 E model BNL model 1.11 cps 8.6 cps 13.3 cps 20.5 cps 27.8 cps 1.05 cps 8.3 cps 10.6 cps 20.2 cps 27.8 cps For all modes, except the third, the variation in frequency between the two models is well within the 10K broadening used to construct design spectra from calculated spectra.

The actual design spectra was examined to determine the significance of the 25K variation in the third mode frequency.

Because this Diablo Canyon SSER 29 5"6

frequency is in the flat portion of the design response spectra it is con-cluded that there would be an insignificant 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 purpose of these struts is to provide out-of-plane supports for the wall at column line 19.

This support is required for the N-S seismic loadings on the wall as shown in Calculation 65-T-211.

The floor diaphragm was modeled by PG8E in the finite element model of the turbine build-ing 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 as shear, interaction between the compressive and shear stresses was not considered when evaluating buckling of the plate elements.

The approach taken. by PG8E 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.

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-mate stress and displacement boundary conditions were applied to these finer mesh models, and the in-plane stresses were 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 cogributed to the strut 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 calculations 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 Diablo Canyon SSER 29 5-7

intended design function.

An additional calculation was performed by PG4E to demonstrate that even if a restricted zone of instability did occur in the immediate area about strut line 0, enough strength still remains in the dia-phragm and in the wall at column line 19 to demonstrate that collapse would not occur due to seismic loadings.

The staff considers this issue resolved.

5.3.2 Conclusions Based on the completeness of the audits of the PG8E 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 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 PG5E has not completed its investigation on this issue, because of its localized nature and the limited number of Category I equipment in the surrounding area, the staff does not con-sider it to be an issue which impacts low-power operation but must be resolved prior to full-power operation.

Diablo Canyon SSER 29 5"8

6 RACEWAYS

6. 1 Introduction Class lA electrical raceways are used to support and route all Class lA elec-trical cables and wires used in the plant.

A raceway is a cylindrical pipe or conduit containing and protecting one or more electrical cables.

The raceways are supported at points along their span by raceway supports which are attached to the structures.

The raceway supports transmit the raceway loads 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 ~62 6

To assess compliance with the design 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 Cal-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 and adequacy.

A random assessment of correctness was made for the design aids.

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 revi'ewed and audited by the staff:

Title and Identification Revision Revision Date Design Calculation - 2CG-91-1 Design Calculation - 2S-176 Design Calculation -

2SPOTWELD Raceway Weight Check - Conduit K 2613 Longitudinal Conduit Run -

LCR 2 Raceway Support Design Aid 1 Raceway Support Design Aid 2 Raceway Support Design Aid ll Raceway Support Design Aid 29 Raceway Support Design Aid 31 Raceway Support Design Aid 33 Raceway Support Design Aid 34 Raceway Support Design Aid 35 Raceway Support Design Aid 35A Raceway Support Design Aid 35B Raceway Support Design Aid 64 Raceway Support Design Aid 69 0

1 0

0 0

0 0

0 0

16 12 17 9

8 12 15 18 9/22/83 9/21/83 9/07/84 11/08/84 2/08/83 8/03/84 8/04/84 9/08/84 11/23/81 5/ll/84 7/30/82 6/26/84 8/03/83 7/26/83 1/10/84 3/27/84 7/10/84 Diablo Canyon SSER 29 6"1

6.3 Findin s and Conclusions The PG8E evaluation of raceway supports was found to be comprehensive.

Each support was qualified by either separ'ate calculation or by generic enveloping calculations.

For each support or group of supports two calculations were made:

one to demonstrate adequacy for vertical and transverse seismic loadings and the second to demonstrate adequacy for vertical and longitudinal seismic loadings.

The transverse evaluations typically involve a computer analysis of the support structure using the STRUDL computer code while the longitudinal evaluations were all hand calculations.

For either calculation the designer had available a number of design aids which summarized the 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 conditions the calculation was refined to consider actual conduit weights, support specific loadings and finally as built support configurations.

In the design calculation reviewed (2S-176) this procedure was followed.

Items specifically verified and found to be adequate were:

a e b.

C.

d.

e.f.

generic weights, loadings and configurations support specific weights, loadings and configurations design aid use unistrut spot weld strength check the consideration of actual bolt torque levels computer models In addition to verifying the source table for conduit weights, the staff requested PG8E to substantiate the listed weight for Conduit K2613.

The com-puter listing of the raceway weights was verified by weight check calculations using the manufacturers weight data for the wires and conduit and the PG5E 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 2S-176 were not appropriate.'ince 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, specifying them in an approximate fashion would be appropriate.

The staff finds the methods used by PGKE to evaluate the electrical raceways to be acceptable.

Diablo Canyon SSER 29 6-2

7 BURIED CONDUITS 7.j.

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 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 immediately above or to one side of the water pipes.

The buried electrical conduits are typically 4-inch or 6-inch plastic conduits which are assembled in duct banks.

These duct banks are anchored in pull 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 conduits from damage due to penetrations.

Flexible connections are provided at transition points of the conduits and at entrances to the pull boxes and bui 1 dings.

The review of the buried electrical conduits was initiated at the NRC design audit 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 in Novem-ber 1984.

More information was reviewed at a continuation audit at the PGIEE offices from January 15 through 17, 1985 (Ref. 34).

Additional information was provided to the staff on January 31, 1985.

2.2 ~ff 2 To assess the adequacy of the design and construction of the electrical duct

banks, the staff audited pertinent information in the following areas:

(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 bedrock levels.

(b)

Estimates of peak relative displacements that can be anticipated at various support point locations along the duct bank run.

(c)

Information on the flexible capability of the typical connection used at the various connection points.

(d)

Capability of the electrical conduits to function in fully submersed situations in either ordinary ground water or sea'ater.

Diablo Canyon SSER 29 7"1

(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.

Information that the slack placed in the cable runs at the various pull boxes is sufficient to ensure that relative movement can be sustained.

7.3 Findin s and Conclusions Information on each of the items listed above was audited by the staff.

The review is summarized as fol'lows.

(a)

(b)

(c)

(d)

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 8 Associates (HLA) reports.

Both the as-built drawings and the HLA reports were reviewed.

The detailed description of the conduits provi4ed on the drawing was found to be adequate.

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 conduit moves with the backfill soil above.

Displacement estimates were made using the SHAKE and TRIP computer programs as well as a simplified design procedure.

This simplified procedure is based on estimating peak axial strains in long buried structures from one dimensional wave analysis (Ref. 42).

The maximum relative motion computed, including both horizon-tal 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 discussed below show that the flexible connector used for the conduit can easily accommodate motion of this magnitude.

A typical flexible connector was tested by PG8E at the request of the staff.

Two separate tests were run.

In the first test, one end of the connector was clamped in a vise, while the other end 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.

Various cable specifications and typical test results were used by PG8E to indicate the capability of the electrical cables to operate in the adverse environments postulated at the site.

In particular, specifica-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.

Diablo Canyon SSER 29 7-2

(e)

The staff reviewed a detailed description of the procedures employed by PG8E to ensure that the 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.

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 inspection was 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 appears to be available in each cable run so as to provide ade-quate capability 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 cab1es are adequate.

Diablo Canyon SSER 29 7-3

8 PIPEWAY STRUCTURE

8. 1 Introduction 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.

It has five major platforms at elevations

109, 114,

119, 127, and 138 feet.

The two main steam lines 1 and 2 and feedwater lines 1 and 2 are supported by the pipeway structure.

Large portions of the structure were shop fabricated and then joined together in the field by bolted connections.

The Unit 1 pipe-way seismic analysis was done by Westinghouse whereas the Unit 2 analysis was done by PG8E.

A description of the staff evaluation of the Unit 2 pipeway structure is given below.

8.2 ~82 8

PG8E performed the seismic evaluations for the pipeway structure using a three dimensional frame model.

This model incorporated a

9 mass-single stick repre-senting the containment wall.

The adequacy of the stick model was demonstrated by a comparison of the floor response spectra generated by both, the stick model and an axi-symmetric finite element model of the containment exterior wall.

Good agreement was found between the floor spectra from both models.

The stick model of the containment was coupled with the three dimensional assembly of beam and truss elements representing the pipeway structure plus the major piping systems.

Piping and piping supports were modeled with beam elements.

This coupled model was used for seismic analysis for the Hosgri earthquake.

The evaluation for DE and DDE earthquakes was performed by Nuclear Services Corpo'ration.

During January 15-17, 1985, the staff conducted an audit of the Hosgri analysis for the pipeway structure performed by PGIIE.

The following documents were reviewed during the audit:

(1)

Calculation No. 1113C-l, Rev.

0, Procedure for Development. of Geometry for BSAP Dynamic Computer Model.

(2)

Vol. 21: 'omputer Analysis Output.

(3)

Calculation 1114C-l, Rev. 0, Dynamic Analysis, BSAP Structural Mathematical Model-Mode Data (including geometry and constraints).

(4)

Calculation 1114C-2, Rev.

0, Dynamic Analysis, BSAP Structural Mathematical Model-Element Data (including material and section properties).

(5)

Vol. 27:

Computer Analysis, Vol. 27, Calc.

1116C-1, Rev.

0 Dynamic Anal-ysis:

Frequency Run 3-3 Cycles ¹25 to ¹30 (with mode shape outputs)

(6)

Vol.

25:8 Computer Analysis Outputs.

Gale 1116C-1, Dynamic Analysis:

Frequency Run 3-1 Cycl'es ¹25 to ¹30 (with mode shape outputs)

Diablo Canyon SSER 29 8-1

(7)

Calcu1ation 1141C-l, 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 dynamic analysis was performed to generate the floor response 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 auxi 1iary building and the turbine bui 1ding 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 cases in the horizontal direction (N-S, E-W) using the Blume input time history with i; = 0.04 seconds, two cases in the horizontal direction (N-S, E-W) using the Newmark input time history with i; = 0.04 seconds and one case in vertical direction using the Newmark input time history for i: = 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 that were used in a static member evaluation analysis.

The static analysis utilized the dead load plus Hosgri combination.

The members of the pipeway structure were then qualified based on this load combination.

PGLE demonstrated that all members passed the stress check.

8.3 Findin s and Conclusions The following are the NRC staff findings and concerns on the pipeway structure:

a 0 As, a result of the staff audit in January 1985 it was found that the slotted holes provided to allow the relative motion between the pip'eway and the auxiliary building were not properly accounted for in the structural model (Ref. 34).

Specifically, the nodes representing the connection of the

'ipeway structure to the auxiliary building were modeled as free nodes in both global horizontal directions.

Only the vertical motion of these nodes was restrained.

Based on the audit it was concluded that the intended motion of these nodes was to move along the direction of the framing beam.

Similar modeling procedures were used for thy end nodes of the radial beams of the pipeway structure which are framed into the turbine building.

Based on additional information provided by PG8E 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.

PG8E has stated that the displacements in both directions can be accommodated by the oversized slotted holes at the 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 for the staff to determine whether or not there is sufficient clearance in the slotted holes.

This will assure that these supports are allowed to move in both horizontal directions and are, therefore, consistent with the assumptions used in the model of the pipeway structure.

As discussed below the staff evaluation will be com-pleted later.,

Diablo Canyon SSER 29 8-2

b.

As a result of the staff's review of the seismic evaluations performed by PG8E 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 the pipeway structure is supported by different structures, i.e., contain-ment, auxiliary and turbine buildings, a choice of a single input that reflects the effects from all of these structures is required.

PG8E selected the containment ground acceleration time history.

The staff requested PG8E to provide the basis for the selection of the input time history.

In response to this request, PG8E provided additional information in order to justify the selection of the input used in the Hosgri evaluation of the pipeway structure.

A comparison was made of vertical spectra from both the containment and the turbine building at the location where the pipeway structure is supported.

It shows that the containment acceleration spectra envelop the turbine building at all frequencies of interest.

On the other

hand, due to the pinned conditions at both ends of the beams framing between the pipeway structure and the auxiliary building no seismic input can be transmitted from the auxiliary building to the pipeway structure.

It is not clear, however, if any input can be transmitted from the auxiliary building to the pipeway structure through the piping systems,

e. g., main steam line.

Thus, the attachments of the piping systems in both the pipe-way structure and the auxiliary building should be reviewed to assure that such transmission of seismic loads would not occur.

C.

During the staff audit in January 1985 it was found that the integration time step used by PG8E to compute the response of the pipeway structure due to the Hosgri event is taken as 0.01 seconds.

PG8E was requested to justify the adequacy of this time step.

Based on the information submitted on January 31, 1985, it is still not clear that the time step of 0.01 seconds used in the time history response calculation is adequate to properly compute the response spectra above 10 Hz.

As discussed below the staff evaluation will be completed later.

d.

The accidental torsion was accounted for by increasing the input amplitude by 6X.

The staff requested PG8E to justify this procedure.

Normally the accidental torsional effects are accounted for by dynamic models in which the masses are set at certain eccentricities from the orthogonal axes.

Alternately PG8E chose to increase the input motion by 6X and performed a

dynamic analysis, using the original model without off-setting the lumped masses of the model.

PG8E supplied additional information to justify that the 6X increment in the input motion produces results which are comparable with those obtained by using a model in which eccentricities are included.

Based on the information provided by PG8E the 6X increase was obtained from a study which was performed for the Unit 1 pipeway structure.

In this study it was concluded that the 6X increase in the input earthquake can sufficiently cover the accidental torsional effects.

This procedure was used in the Unit 2 pipeway structure after performing a comparison between the characteristics of the Unit 1 versus Unit 2 pipeway structures.

It was concluded that the procedure used in Unit 1 was also applicable to the Unit 2 pipeway structure.

The staff concludes that this approach is acceptable.

Diablo Canyon SSER 29 8-3

e.

In response to the staff's request, PG&E reviewed the strength capability of the pipeway structure to accommodate the relative motions between struc-tures and still remain within the allowable stress levels.

The vertical motions between the various buildings supporting the pipeway structure are shown to be small.

They are not expected to alter significantly the stresses of the pipeway structure.

On the other hand, PG8E stated that the horizontal relative displacements can be accommodated by the connec-tions 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 analysis should be provided to the staff for review.

As discussed below the staff evaluation will be completed later.

Based on the detailed staff review of the PG8E calculations, the staff finds that the general approach used for the analysis of the pipeway structure by PG8E 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 significantly alter the overall response of the pipeway structure.

However, if structural modifications are necessary they can be performed with little diffi-culty.

Therefore, the staff concludes that this issue need not be resolved for low power operation but must be resolved for full power operation.

The staff will continue its review of this matter and will report the results at the completion of the evaluation.

Diablo Canyon SSER 29 8-4

9 PIPING SYSTEMS AND PIPE SUPPORTS The Unit 1 design verification effort included an extensive review and evalua-tion by the IDVP 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 those'ctivities performed within the Onsite Project Engineering 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 inspections 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 programmatic issues concerning onsite engi-neering have been resolved.

Further details on the last item were subsequently provided in a Board Notification (Ref. 25).

The above described efforts by PG&E and the NRC staff were directed specifically to the reinstatement of the Unit 1 low power license in April 1984 (Ref.

31) and issuance of the Unit 1 full-power license in November 1984 (Ref. 27).

As

'iscussed in Section 2 of this report, as the efforts for Unit 1 came to 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 1 design verification, (2) actions 'resulting from the Unit 1 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 Diablo Canyon SSER 29 9"1

in a separate

report, SSER 30 to the Safety Evaluation Report.

The team inc'luded 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 calculation package contained the documentation required in accordance with the, PG8E 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'hermal anchor movements, and computer input and output for all load conditions.

Engineering judgments and assumptions were reviewed for correction and documentation.

Pipe support analyses were also checked for completeness and correctness.

This included verification of the correlation of support location, orientation and loads with those specified in the piping analysis and of the correlation of support drawings with the computer input.

The analysis packages were chosen for review, both randomly and selectively.

The random selection provided a rea'listic sampling of all analysis work per-formed on Unit 2 piping systems.

The selective process was used to check on specific 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 ques-tions in the following areas:

modulus of elasticity stress intensification factors Mestinghouse piping support computer programs base plates with single row of anchor bolts branch connection pressure design field verification of pipe and pipe support configurations welded attachment local stresses rigid support assumptions for fluid dynamic load analysis structural anchor rotational stiffness U-bolt prying on angle members Details on these matters and their resolution are presented in SSER 30.

The review team has pursued the PGKE 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 cohpliance with appropriate licensing criteria.

Specifically, the team deter-mined that the license conditions regarding piping and pipe supports for Unit 1 have been satisfactorily addressed and resolved for Unit 2.

As part of the piping and pipe support effort, the review team observed the PG8E 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 Diablo Canyon SSER 29 9-2

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 f'r conducting the walkdowns.

Based on its review of previous walkdowns and the prescribed procedures and based on its observance of the walkdowns, the team concl'uded 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 manner to reasonably detect. and justify those normal displacements which deviated from calculated displacements.

The staff identified' concern pertaining to the pressure design adequacy of certain branch connections (Ref. 37).

PGKE has committed to identify all such branch connections and provide verification to show compliance with the appli-cable design code (Ref. 45).

The staff finds this acceptable.

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'nd 100 pipi'ng and support calculational packages not only with regard to their, technical adequacy.but also with respect to the IRP process.

In summary, the NRC team identified some concerns during the piping and pipe support review.

They have all been resolved or deemed to be not of safety significance.

These can be categorized as discrepancies with only an insignifi-cant effect on the particular analysis.

The piping system hot walkdown proce-dure review and onsite audit of the walkdowns of the two systems revealed that acceptable engineering techniques were employed and the walkdowns were con-cluded in a competent manner.

The staff concludes that the issues raised during the Diablo Canyon Unit 1 design verification effort, allegations and in the Unit 1 low power license condition have been adequately addressed and resolved for Unit 2 and that the licensing criteria in this regard have been met.

Diablo Canyon SSER 29 9-3

10 NON-SEISMIC DESIGN ASPECTS

10. 1

~Sstems 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 PG8E Internal Review Program (IRP) for Unit 2 as described in Section 2 of this report.

The PG8E submittals of November 2 and December 7, 1984 identified the specific findings from Unit 1 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.

PGIIE 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 1 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 the IRP properly incorporates the findings of the Unit 1 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 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, the Pacific Ocean.

The technical specification was irlcluded for Diablo Canyon 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 PGImE 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 EOI 8014 during the Unit 1 design verification effort (Ref. 28).

Differences from the Unit 1 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 PG8E in a letter dated February 21, 1985 (Ref. 28).

Diablo Canyon SSER 29 10-1

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 equipped with hand wheels to facili-tate their subsequent operation, if necessary.

Resolution of this item for Unit 1 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 1 resolution was applicable to Unit 2.

Based on its review of the IRP package documenting the PG&E eval-uation, 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.

Based on its review of the IRP package documenting the PG&E evaluation, 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 resolution could be applied equally to Unit 1.

10.2 Jet Im in ement Anal ses PG&E has performed an evaluation for Unit 2 of the effects of jet impingement on 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',

and which also meet the size and usage cri-teria of the Standard Review Plan.

The staff evaluation for Unit 1 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 PG&E evaluated 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.

Diablo Canyon SSER 29 10-2

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.

The staff has reviewed and evaluated the PG&E evaluation and found it acceptable.

In addition to exp'anding the scope of the jet impingement review PG&E also evaluated the applicability to Unit 2 of the findings and concerns by the IDVP during the design verification for Unit 1 regarding the effects of a HELB in Auxiliary Feedwater (AFM) System Line 594.

This evaluation was performed as part of the Unit 2 Internal Review 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 AFM 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 1 resolution of the concern is not directly applicable to Unit 2 because the local pipe routing and location of supports of the AFM 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 AFM System were evaluated as part of the Unit 2 over-all high energy jet impingement effects review.

The staff finds this acceptable based on the review performed for Unit l.

10.3 E ui ment uglification 10.3. 1 Environmental gualification As a result of the Unit 1 design verification program the staff had identified in SSER 18 four items (F-I4, F-I5, FI-12, FI-14) 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 these 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.-

10.3.2 Main Annunciator Typewriter Seismic gualification D~ing 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 Diablo Canyon SSER 29 10-3

relocated to a location equivalent to Unit 1, where seismic qualification was valid.

The staff has reviewed the information provided with regard to the acceptability of the resolution.

Based on the above, the staff has determined that the equipment is seismically qualified and the resolution is acceptable.

Diablo Canyon SSER 29 10-4

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 NRC s proposed 18 criteria for QA.

The criteria were issued in June of 1970 in Appendix B to 10 CFR 50, "Quality Assurance Criteria for Nuclear Power Plants."

The construction permit for Unit 2 was issued 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.

Mith 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, requirements and controls, when properly implemented, comply with the require-ments of Appendix B of 10 CFR 50 (Ref. 44).

Thus, from 1970 until today PG&E has been committed to a QA program for Diablo Canyon Unit 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 by verification of the design under the IDVP, that construction was done under acceptable QA controls, and that current corrective'ctions and the IDVP work itself are being performed in accordance with acceptable QA programs."

Based on the review of the IRP implementation, the staff has determined that this conclusion applies equally 'to Unit 2.

As part of its evaluation effort for the Unit 1 piping systems and pipe supports the staff also reviewed in mid-1984 programmatic provisions for current and future work by PG&E engineering organizations at the Diablo Canyon site (Ref. 25).

The effort included an evaluation of PG&E responses (Ref.

46) to staff requests for information (Ref. 47), audits of records and procedures and interviews of project personnel at the Project Offices in San Francisco and at the Diablo Canyon site.

The evaluation was specifically directed to the train-ing and quality assurance requirements and their implementation as related to the Onsite Project Engineering Group (OPEG).

The staff concluded with regard to the overall onsite project activities that (1) training programs are up to date and kept current, (2) engineering procedures are adequate and are being implemented, and (3) internal QA audits, responses and corrective actions are adequate and timely.

'urther, with regard to the piping system and support activities, specifically, the staff concluded that (4) the authority of the program for onsite design changes had been redefined, and Diablo Canyon SSER 29

(5) the design responsibility had been effectively transferred to the home offices.

The staff evaluation was performed as part of the Unit 1 pipe system and support evaluation.

However, it was based on audits and inspection of procedures and manuals that were applied by PG&E to both units.

The staff concludes that the above results, as described in detail in Reference 25, are equally applicable to both units.

Since the initiation of the design verification program for Unit 1 in late 1981, numerous inspections have been performed in the area of quality assurance and construction, principally by the NRC Region V Office.

In general, these inspections apply to both units.

While the nature of the inspections do not specifically address implementation of the IRP process, the results of the inspections for implementaion of the quality assurance program and construc-tion activities indicate that the design and analysis efforts for Unit 2 are adequate.

As a result of these inspection activities, sixteen violations were issued since September 1981.

The majority of these violations applied to a cross-section of specific construction errors, as compared to the approved design

features, including welding of structural steel, piping, raceways, and HVAC supports.

PG&E was responsive to the notices of the violation.

Considering the extensive inspections effort and the relatively small number of minor violations, the staff finds that adherence to procedures was generally satisfactory.

The construction inspections also showed PG&E management to be frequently involved in construction activities.

The PG&E personnel generally had a good 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 inspection 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 1 readiness for fuel loading and low power testing.

The result of these examinations and investigations indicated that, while there 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 years;-

In conclusion, this extensive inspection effort demonstrated that the licensee has constructed the plant in substantial agree-ment with regulatory commitments and requirements.

Diablo Canyon SSER 29 11-2

12 REFERENCES 2.

3.

4.

5.

6.

7.

8.

9.

10.

12.

13.

14:

15.

16.

U.S. Nuclear Regulatory Commission, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report," October 12, 1974.

U.S. Nuclear Regulatory Commission, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report, Supplement 16,"

SSER 16, August 1983.

U.S. Nuclear Regulatory Commission, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report, Supplement 17,"

SSER 17, February 1984.

U.S. Nuclear Regulatory Commission, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report, Supplement 18,"

SSER 18, August 1983.

U.S. Nuclear Regulatory Commission, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report, Supplement 19,"

SSER 19, October 1983.-

U.S. Nuclear Regulatory Commission, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report, Supplement 20,"

SSER 20, December 1983.

U.S. Nuclear Regulatory Commission, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report, Supplement 21,"

SSER 21, December 1983.

U.S. Nuclear Regulatory Commission, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report, Supplement 22,"

SSER 22, March 1984.

U.S. Nuclear Regulatory Commission, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report, Supplement 23,"

SSER 23, June 1984.

U.S. Nuclear Regulatory Commission, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report, Supplement 24,"

SSER 24, July 1984.

U.S. Nuclear Regulatory Commission, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report, Supplement 25,"

SSER 25, July 1984.

U.S Nuclear Regulatory Commission, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report, Supplement 26,"

SSER 26, July 1984.

U.S Nuclear Regulatory Commission, "Diablo Canyon, Units 1 and 2 Safety Evaluation, Supplement 2?,"

SSER 2?, July 1984.

U.S. Nuclear Regulatory Commission, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report, Supplement 28,"

SSER 28, in preparation.

U.S. Nuclear Regulatory Commission, "Diablo Canyon, Units 1 and 2 Safety Evaluation Report, Supplement 30,"

SSER 30, in preparation.

U.S. Nuclear Regulatory Commission, Commission Memorandum and Order CLI-81-30, "Order Suspending License,"

November 19, 1981.

Diablo Canyon SSER 29 12"1

17.

18.

19.

20.

21.

22.

23.

24.

25.

26.

27.

28.

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.

ALAB-756, 18 NRC 1340 (1983).

ALAB-763, 19 NRC 571 (1984).

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.

U.S. Nuclear Regulatory Commission, November 29, 1983, from M. J. Dircks (NRC) to Chairman Palladino, Commissioners Gilinsky, Roberts, Asselstine and Roberts,

Subject:

Plans for Dealing with Allegations to Diablo Canyon.

Pacific Gas and Electric Company,- "Diablo Canyon Unit 2, Allegation Review

Program, Final Report,"

December 5, 1984.

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.

U.S. Nuclear Regulatory Commission, "Diablo Canyon Low Power License DPR-76, Order Modifying License," April 18, 1984.

U.S. Nuclear Regulatory Commission, Board Notification 84-161, "Diablo Canyon - Completion of Piping Review Activities," September 24, 1984.

Pacific Gas and Electric Company, "Piping and Pipe Supports Review Program for Diablo Canyon Unit 2, Final Report,"January 31, 1985.

U.S.'uclear Regulatory Commission, "Diablo Canyon Full Power License DPR-80,"November 2, 1984.

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, "Diablo Canyon Unit 2 Specific Design Aspects" PG&E Letter No.

DCL-84-278, July 31, 1984, "Internal Review Program Summary and Status Report,"

PG&E Letter No.

DCL-84-332, October 19, 1984, "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.

r Diablo Canyon SSER 29 12"2

29.

30.

31.

32.

33.

34.

35.

36.

37.

38.

39.

40.

41.

42.

43.

U.S. Nuclear Regulatory Commission, September 21,

1984, from H. Schierling (NRC) to G. Knighton (NRC),

Subject:

Summary of Meeting with PG5E on Diablo Canyon Unit 2, September 13, 1984.

U.S. Nuclear Regulatory Commission, "Diablo Canyon Low Power License, DPR-76," September 22, 1981.

U.S. Nuclear Regulatory Commission, Memorandum and Order CLI-84-5, "Order Reinstating PG8E's Low Power License," April 13, 1984.

U.S. Nuclear Regulatory Commission, November 29, 1984, from G.

M.

Knighton (NRC) to J.

D.

Shiffer (PG8E),

Subject:

NRC Trip/Audit Reports - Unit 2.

U.S. Nuclear Regulatory Commission, January 3, 1985, from G.

M. Knighton (NRC) to J.

D; Shiffer (PG8E),

Subject:

NRC Staff Audits.

U.S. Nuclear Regulatory Commission, January 31, 1985, from G.

W. Knighton (NRC) to J.

D.

Shiffer (PGEE),

Subject:

NRC Staff Audit.

U.S. Nuclear Regulatory Commission, November 14, 1984, from G.

M. Knighton (NRC) to J.

D.

Shiffer (PG8E),"

Subject:

Piping and Pipe Supports-Request for Additional Information.

U.S.

Nuclear Regulatory Commission, January 8, 1985, from G.

M. Knighton (NRC) to J.

D.

Shiffer (PGEE),

Subject:

Request for Additional Infor-mation on Footprint Loads.

U.S. Nuclear Regulatory Commission, February 25, 1984, from G.

M. Knighton (NRC) to J.

D. Shiffer (PG8E),

Subject:

Piping Branch Connections-Request for Additional Information.

U.S. Nuclear Regulatory Commission, December 20, 1984, from G.

W. Knighton (NRC) to J.

D.

Shiffer (PG8E),

Subject:

Pipe Support Concerns Expressed by Individual.

U.S.

Nuclear Regulatory Commission, January 17,

1985, from G.

W. Knighton (NRC) to J.

D.

Shiffer (PG8E),

Subject:

Anonymdus Allegation on Diablo Canyon.

Pacific Gas and Electric Company, October 14, 1983, from J.

0. Schuyler (PGKE) to D.

G. Eisenhut (NRC),

Subject:

DCP Design Verification Program Phase I, Final Report.

"Torsion of Rolled Steel Sections in Building Structures,"

by John G.

Hotchkiss, AISC Engineering Journal,

January, 1966.

"Seismic Analysis of Structures and Equipment for Nuclea~

Power Plants,"

Topical Report No.

BC-TOP-4-A, Rev.

3, Nov. 1974, Betchel Power Corp.

San Francisco, California.

Pacific Gas and Electric Company, June 18,

1982, from P.

A. Crane, Jr.

(PG8E) to F. J. Miraglia, Jr.

(NRC),

Subject:

Diablo Canyon Project equality Assurance Program.

Diablo Canyon SSER 29 12-3

Pacific Gas and Electric Company, August 13, 1982, from P.

A. Crane, Jr.

(PG8E) to F. J. Miraglia, Jr.

(NRC),

Subject:

Diablo Canyon Project equality Assurance Program.

Pacific Gas and Electric Company, December 21,

1982, from P.

A. Crane, Jr.

{PG8E) to G.

W. Knighton,

{NRC),

Subject:

Diablo Canyon Project equality Assurance Program.

44.

U.S. Nuclear Regulatory Commission, January 26, 1983, from D.

G. Eisenhut, (NRC) to P.

A. Crane, (PG8E),

Subject:

Diablo Canyon Project equality Assurance Program.

45.

Pacific Gas and Electric Company, February 27, 1985, from J.

D. Shiffer (PG8E) to G.

W. Knighton (NRC),

Subject:

Supplemental Information on Design of Piping Connections.

46.

Pacific Gas and Electric Company, June 26, 1984, from J.

0. Schuyler (PG&E) to D.

G. Eisenhut (NRC),

Subject:

Diablo Canyon Unit 1 - Additional Information Regarding Piping and Supports.

47.

U.S.

Nuclear Regulatory Commission, June 20, 1984, from D.

G. Eisenhut (NRC) to J.

0. Schuyler (PG8E),

Subject:

Request for Additional Informa-tion Regarding Piping and Supports.

Diablo Canyon SSER 29 12-4

NRC FOAM 33S

<24Ml NREM 1102.

320), 3202 US. NUCLEAR REGULATORYCOMMISSION I, REPORT NUMBER IArrrFntdhF FIOC tdd Vol IVa,iltnyl NUREG-0675 Supplement No.

29 BIBLIOGRAPHIC DATA SHEET SEE INSTRUCTIONS ON THE REVERSE,

2. TITLE AND SUBTITLE Safety Evaluation Report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2

S,LEAVE BLANK 4, DATE REPORT COMPLETED B.AUTHOAISI MONTH March YEAR lbS5 B. DATE REPORT ISSUED MONTH March YEAR 1985 7, PERFORMING OAQANIZATIONNAME AND MAILINGADDRESS Ilncludt Zrp Codtl Division of Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D?C.

20555 B. PAOJECTITASKIWOAKUNIT NUMBER B. FIN OR GRANT NUMBER

10. SPONSORING ORGANIZATIONNAME ANDMAILINGADDRESS llncludtZrp Codtl Same as 7. above 11t. TYPE OF REPORT Safety Evaluation Report b, PERIOD COVER ED IInclurlttdtrtrl

'12. SUI'PLEMENTARYNOTFS Docket Nos.

50-275 and 50-323

13. ABSTRACT 1200 wocdt or Itrrl Supplement No. 29'o the Safety Eva3,uation Report for Pacific Gas and Electric Company's application for licenses to operate Diablo Canyon Nuclear Power Plant, Units 1 and 2

(Docket Nos.

50-275 and 50-323),

has been prepared by the Office of Nuclear Reactor Regulation 'of the U.S. Nuclear Regulatory Commi'ssion'.

This supplement presents the staff evaluation of the licensee's Internal Review Program for Diablo Unit 2 applica-bility and resolution of concerns that had been ra'ised during the Diablo Unit 1 design verification by the Independent Design Verification Program, the licensee's Internal Technical Program and the NRC staff.

ln. DOCUMENT ANALYSIS

~. KEYWORDS/OESCRIPTORS 15, AVAILABILITY STATEMENT Unlimited Ir. IDENTIFIERSIOPEN FND'ED TERMS IB.SECURITYCLASSIFICATION IPhil ptptl Unclassified IFhirrtporrl Unclassified

17. NUMBER OF PAGES

18. PRICE

r t