ML20213H083

Forwards Engineering Evaluation of Seismic Anchors for Electrical & Control Panels Re Several Recent Events Involving Inadequate Anchorage of Electrical & Control Panels.Usi A-46 Nearing Resolution

ML20213H083
Person / Time
Site:	Dresden, Davis Besse, Cooper, 05000000
Issue date:	10/16/1986
From:	Hebdon F Office of Nuclear Reactor Regulation
To:	Starostecki R, Vollmer R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE), Office of Nuclear Reactor Regulation
Shared Package
ML20213E759	List: IA-87-181, Forwards USI A-46 Value/Impact Analysis for Clarification & Addl Info to 860728 Example Regulatory Analysis.Part D, Seismic Qualification Cost Estimating Task, to NUREG/CR-3875 Also Encl ML20213E756 ML20213H083 ML20213H148 ML20214A051
References
FOIA-87-181, REF-GTECI-A-46, REF-GTECI-SC, RTR-NUREG-1030, RTR-NUREG-1211, TASK-A-46, TASK-OR AEOD-E611, NUDOCS 8611190076
Download: ML20213H083 (16)

Text

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/ 8,, UNITED STATES 8 p, NUCLEAR REGULATORY COMMISSION D  :

j WASHINGTON, D. C. 20655 k...*,/ October 16, 1986 AE0D/E611 MEMORANDUM FOR: Richard H. Vollmer, Deputy Director Office of Nuclear Reactor Regulation Richard W. Starostecki, Acting Deputy Director Office for Inspection and Enforcement FROM: Frederick J. Hebdon, Deputy Director Office for Analysis and Evaluation of Operational Data

SUBJECT:

DEFICIENCIES IN SEISMIC ANCHORAGE FOR ELECTRICAL AND CONTROL PANELS Enclosed is an engineering evaluation report concerning several recent events involving inadequate anchorage of electrical and control panels. The report specifically addresses recent events at Cooper, Dresden 2/3 and Da.v.is-Besse.

However, we also identified other similar events uncovered by inspections by Sandia for the USI A-45 (Decay Heat Removal) program and in the SCSS database.

Some of the events found by Sandia (a g., at AN0-1 and Cooper) are equally as significant as the ones in the rece' 'ERs.

In each of the recent events, alt w the plant design and installation specifications correctly specifieo die seismic anchorage, construction  !

deficiencies occurred that resulted in inadequate anchorages. The Cooper deficiencies were related to both emergency diesel generator switchgear electrical cabinets; at Davis-Besse the Class IE equipment electrical cabinets were missing anchor bolts; and at Dresden 2/3 none of the control panels in the control room were anchored.

l The USI A-46 (seismic adequacy of operating reactors) is nearing resolution.

As documented in the draft final reports for USI A-46 (NUREG-1030 and l NUREG-1211), the staff is proposing a detailed reinspection program for i

verifying the seismic design margin of all equipment necessary to bring a I plant to a safe, hot shutdown and to maintain it for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The reinspec-tion effort would be required by a generic letter and would be conducted after utility personnel receive training on reinspection procedures and evaluation techniques as developed by the Seismic Qualification Utility Group.

We have made the following suggestions for bringing the USI A-46 program to completion:

1. It is suggested tnat the Office of Inspection and Enforcement (IE) consider issuing an IE Infonnation Notice to all licensees and possibly owners of NTOL plants concerning the recent evidence that inadequacies continue to m

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. 1 exist in the seismic anchorage of equipment essential for accomplishing a safe, hot shutdown. Issuance of an Information Notice characterizing the recently identified problems in spite of previous, related reinspections and walk-throughs would provide details of the extent and variation of the continuing deficiencies and would emphasize this regulatory concern. This seems justified in light of the evidence of the inadequate response to IN 80-21. The Information Notice should discuss the recent experience, the potential safety implications, past and present regulatory attention to this concern (i.e., IN 80-21 and USI A-46) and reference the generic letter to be issued to resolve USI A-46.

~ -2. The fundamental assumption underlying the proposed resolution of USI A-46 c' is that gross deficiencies do not exist in the seismic anchorage of operating nuclear plant safety equipment. The operational experience seems

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to indicate that this assumption may not be valid. Thus, we suggest that the proposed generic letter be modified to require early walk-through inspection of operating reactors to verify the absence of gross seismic anchorage defects before the detailed reinspections are conducted.

3. We suggest that, prior to issuance of the generic letter, the scope of the applicability of the letter, which would exclude reactors that have been designed to the IEEE-344/1975 tandard, be reconsidered in light of recent experience that demonstrates construction and QA deficiencies are the root cause of inadequate anchorage rather than primarily design deficiencies.

4. We suggest that the generic, reinspection procedures be reviewed (e.g.,

reviewed by experts who are familiar with previous reinspections and walk-throughs) and/or documentation be developed by NRR to illustrate why the new program will overcome the limitations of the previous efforts to assure that limitations in walk-through inspections associated with

IN 80-21, the Systematic Evaluation Program (SEP), the Seismic Qualification Utility Group (SQUG) and the Senior Seismic Review Advisory Panel (SSRAP) activities are remedied by the generic procedures.

5. We suggest that NRR consider whether the proposed USI A-46 generic letter requirement to identify equipment categories essential for achieving a safe, hot shutdown could be identified prior to availability of the final reinspection procedures rather than serially after the procedures and training of inspection personnel have been completed. Parallel i development of equipment lists would somewhat expedite the reinspection efforts and would aid in accomplishing preliminary walk-downs for identification of gross deficiencies.

If you or your staff have any questions concerning the enclosed engineering evaluation report, please contact Neill Thomasson of my staff on extension x24431.

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Frederick bdon, J. k.DeputyN - Director Office for Analysis and Evaluation of Operational Data

Enclosure:

As Stated cc w/ enclosure:

D. Humenansky, OCM J. Sniezek, DEDR0GR E. Jordan, IE J. Partlow, IE.

R. Baer, IE N. Anderson, NRR C. Grimes, NRR A. Marchese, NRR G. Bagchi, NRR T. Chang, NRR T. Chen, NRR R. Gilbert, NRR W. Long, NRR A. DeAgazio, NRR J. O'Brien, RES D. Dubois, SRI (Cooper)

.W. Rogers, SRI (Davis-Besse)

L. McGregor, SRI (Dresden)

W. Kane, RI R. Walker, RII C. Norelius, RIII E. Johnson, RIV D. Kirsch, RV

AE00 ENGINEERING EVALUATION REPORT

• UNITS: Dresden Units 2 and 3 EE REPORT NO.: AEOD/E 611 Cooper Nuclear Station DATE: October 16, 1986 Davis-Besse Unit 1 EVALUATOR / CONTACT: N. Thomasson DOCKET N0S: 50-237/249 50-298 50-346 LICENSEES: Commonwealth Edison Company Nebraska Public Power District Toledo Edison Company NSSS/AE: General Electric /Sargent & Lundy General Electric / Burns and Roe Babcock & Wilcox/Bechtel

SUBJECT:

SEISMIC ANCHORS FOR ELECTRICAL AND CONTROL PANELS

SUMMARY

Three recent Licensee Event Reports (LERs) indicate that problems continue to exist in the adequacy of seismic anchorage of critical safety equipment in operating nuclear power plants. This has been a generic regulatory concern since 1980, when Information Notice (IN) 80-21 was published, and the issue was designated Unresolved Safety Issue (USI) A-46, " Seismic Qualification of Equipment in Operating Nuclear Power Plants." The 1980 Information Notice specifically suggested that licensees should perform walk-through inspections of the anchorage systems of certain specified categories of equipment, includ-ing electrical cabinets and control panels. In spite of IN 80-21 and indepen-dent walk-through inspections related to seismic anchorage of equipment at certain plants by different groups, the lack of seismic anchorage for electri-cal and control panels continues to be identified, including in some cases gross inadequacies. The near-term resolution of USI A-46 is aimed in part at remedying these residual defects. A generic letter is to be issued requiring walk-through inspections of operating nuclear plants using detailed generic checklists and procedures to verify the adequacy of seismic anchorage for all equipment necessary to accomplish and maintain safe, hot shutdown. Consequently, that solution appears to be adequate to assure that the issue is thoroughly

, addressed. However, in light of the evidence of continuing deficiencies, se suggest (1) that IE issue an updated IN characterizing recently identified deficiencies in order to emphasize the need to respond expeditiously to the generic letter to complete the reinspections, and to correct the deficiencies thus identified; and (2) that NRR modify the generic letter to require near-term walkdown of critical equipment to verify the absence of gross deficiencies before the planned detailed inspections.

INTRODUCTION The Systematic Evaluation Program (SEP) addressed the seismic adequacy of 12 older plants vis-a-vis current licensing requirements. Based on the subsequent findings of possible significant deficiencies in seismic anchorages and support of safety-related electrical equipment at several of these plants, 1

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i This document supports ongoing AE00 and NRC activities and does not represent the position or requirements of the responsible NRC program office.

an IE Information Notice (IN 80-21) was issued in May 1980 with specific focus on performing walk-down inspections of equipment anchorages in nuclear plants designed prior to 1971. Then, in 1980 NRR established the concern of seismic qualification of equipment in operating nuclear plants as an Unresolved Safety Issue (USI A-46).

Between late March 1986 and May 1986, three Licensee Event Reports (LERs) were received that documented inadequate seismic anchorage of electrical equipment in operating nuclear power plants. The initial case was at Davis-Besse 1 (LER 4 86-011) where cabinet doors on Cyberex class 1E equipment for essential instrument 120 VAC power were found to lack the required door bolts. The second LER concerned emergency diesel generator switchgear cabinets at Copper that were not fastened to embedded channels beneath the cabinets (LER 86-009).

The third and most extensive deficiency was found at Dresden 2, where it was determined that the control room control panels did not have positive anchorage to the floor (LER 86-009). In each instance, the deficiercy had existed since plant construction and was the result of installation errors, since the design drawings had specified seismic anchorage. As a consequence of these events, a review was initiated by the Office for Analysis and Evaluation of Operational Data (AEOD) to determine the extent of other similar deficiencies, regulatory requirements, history of addressing the question of seismic anchorage for electrical equipment, and the safety considerations of the deficiencies.

DISCUSSION Regulatory Requirements The fundamental regulatory requirements for seismic design of nuclear power plants are given in Appendix A to 10 CFR Part 100, " Seismic and Geologic Siting Criteria for Nuclear Power Plants" and General Design Criterion 2 of Appendix A to 10 CFR Part 50, " Design Basis for Protection Against Natural

, Phenomena." The regulations require nuclear power plants to be designed to assure functionality of certain structures, systems and components in the event of a design basis earthquake [a safe shutdown earthquake (SSE)]. One reason for such a requirement is to ensure the capability to shut down the reactor and maintain it in a safe shutdown condition during and following an SSE.

Specific guidance related to electrical cabinets, relays and motor control centers is also given in the Standard Review Plan (SRP) (NUREG-0800)

Section 3.10, Regulatory Guide (RG) 1.100, and IEEE Standards 344/1975, 501/1978, and 649/1980. These requirements and guidance have evolved over the history of commercial nuclear power development. Thus, criteria and methods for the seismic qualification of mechanical and electrical equipment have experienced considerable change. Consequently, for older operating reactors ** the capability of the plant and equipment to resist induced seismic loads and to perform essential safety functions following an SSE may vary significantly.

• • Those not reviewed against current design and equipment seismic qualification requirements, in particular IEEE 344/1975.

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The proposed resolution of USI A-46, scheduled for the Fall of 1986, is to issue a generic letter (GL) requiring reinspection using specified generic checklists and inspection procedures *** for the reinspection of seismic anchorage and mounts for all equipment necessary to bring the older power reactors (i.e., those not designad in accordance with IEEE 344/1975) to a safe, hot shutdown and to maintain it for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If such reinspection demonstrates the absence of significant deficiencies in the seismic anchorage of the critical equipment, the plant will be acknowledged to have adequate seismic design margins.

Regulatory Reviews All nuclear power plants have been designed and constructed to the seismic design requirements operative at the time of their construction permits. Similarly, the plants have implemented quality assurance (QA) inspections complemented by inspections.by NRC staff during construction. Although all plants were reviewed during their construction permit and operating licensing proceedings against prevailing seismic and QA criteria and requirements, regulatory guidance and requirements in these areas have experienced significant upgrading since the early 1970's.

In 1978, the NRC established the Systematic Evaluation Program (SEP) to review the designs of older operating plants to reconfirm and document their margins of safety. The SEP review assessed the significance of the differences between current technical safety requirements and guidance and those operative when a given reactor was licensed by comparing the as-built design to the Standard Review Plan (SRP), NUREG-0800, in 137 review areas. Seismic qualification of plant safety equipment was one of the review areas. As a result of the SEP reviews, IN 80-21 was issued to alert the nuclear industry of NRC's concern regarding the lack of engineering supports of safety-related electrical equipment in older plants, including motor control centers, switchgear, control room control panels and instrument panels.

In support of developing a resolution for USI A-46, an industry group, designated as the Seismic Qualification Utility Group (SQUG), was formed to provide input on seismic experience in the nuclear and non-nuclear industry.

Another group the Senior Seismic Review Advisory Panel (SSRAP) was created by SQUG to provide an independent review of the SQUG contractor work. A key element of the SQUG effort was to identify (1) seismic design practices in the nuclear and non-nuclear industries and (2) the extent of comparability of critical equipment in both types of facilities so that the extensive database of non-nuclear industry equipment experience in response to severe seismic motion might be translated to expected nuclear power plant response. In i conducting these activities, SQUG contractors and SSRAP performed walk-through inspections of Dresden 2, surveying eight classes of critical equipment and the seismic anchorages of the equipment.

The NRC research program to develop equipment fragility information also included some plant inspections of critical safety equipment seismic anchorage, as has the NRR program to address decay heat removal (USI A-45).

Recent Operational Events Davis-Besse 1 On February 28, 1986, Toledo Edison determined that, since plant licensing (1977), the cabinet doors on Cyberex class IE equipment (battery chargers, inverters, rectifiers, and distribution panels) for

• • • NUREG-1030 and its Regulatory Analysis--NUREG-1121.

essential instrument AC power were not bolted closed as required for seismic qualification. The deficiency was attributed to failure to develop adequate procedures for equipment installation. Apparently, installers during original construction did not recognize the seismic function of the bolts, given that the cabinet doors were hinged and latched; thus, during equipment installation the bolts were apparently removed on the belief that the bolts were for shipping purposes. The deficiency was recognized when Cyberex Corporation confirmed that the missing bolts were added during seismic qualification testing.

Each of the four 120 VAC distribution panels with the missing bolts supply one channel of the plant's (1) engineered safety features actuation system; (2) the Steam and Feedwater Rupture Control System; (3) the Reactor Protection System; and (4) the Anticipatory Reactor Trip System. According to Toledo Edison, loss of power to these systems results in a fail-safe condition since each system actuates on loss of two distribution panels, causing a reactor trip. However, as noted by the licensee, control and stabilization of the plant following such a trip become increasingly complicated as additional control panels are lost.

The licensee's corrective action was to install the required bolts on the cabinet doors and to revise plant maintenance procedures.

Cooper On March 18, 1986, the Emergency Diesel Generator (EDG) 2 switchgear generator output breaker cabinet at Cooper Nuclear Station was determined to lack anchorage to embedded channels beneath the cabinets. Although this reactor had been operating since 1974, the construction defect went unrecognized until plant staff found the deficiency while preparing for future modifications. In response to the deficiency, utility personnel inspected other floor-mounted switchgear and, on a sampling basis, inspected the anchorage of other safety-related equirment. A similar deficiency was found for EDG-1, but no other like problems with other equipment were found. The corrective action was to weld the cabinets to the embedded channels.

Dresden 2 and 3 On April 1, 1986, Commonwealth Edison determined that the Dresden Unit 2 control room control panels had not been positively anchored to the floor as required by the original design and installation instructions. The l deficiency was found during seismic evaluation of the control panels for addition of future instrumentation. The affected General Electric control panels are supported on structural steel channel sections, which were anchored to the concrete floor. The control panels were supposed to have been bolted to the steel channels via mounting holes in the front and rear of the control panels. However, the bolts had r:ever been installed. A similar condition was subsequently identified for the Unit 3 control panels.

The corrective actions at Dresden 2 and 3 were to install 1" x 1" SAE Grade 5 (ASTM A449) anchor bolts through the control panels to the structural steel channels or, in cases where interferences existed, to weld the control panel to the steel channel. Also, in order to prevent overturning, braces were added to the top of the control panel for the Standby Gas Treatment System.

Other Events The Sequence Coding and Search System (SCSS) LER database con-tained four other instances since 1981 of inadequate seismic mounts of safety-related electrical equipment and other related seismic inadequacies regarding seismic restraints other than anchorages. Each of these LERs is summarized in Table 1. l t

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Table 1 LIRs Involving Inadequate Seismic Anchorage for Electrical and Control Cabinets DATE PLANT LER EvtNT ROOT CAUSE PLANT STATUS COMMENT 3/18/86 Cooper 86-009 EDG-2 Switchgear 9 r. orator Apparent installation Reactor at 85% Power. Found in preparing output breaker ca?dnet not deficiency during original for future modiff-fastened to embed 191 etennels constresction. cation; other floor beneath the cabinet. Same mounted switchgear deficiency was found uc*.= inspected as well as inspection of EDG-1. samp1tng of other floor mounted safety-related equipment.

Bases were welded to channels embedded in the floor.

4/1/86 Cresden 2/3 86-009 Control panels in control roon Anch*.* u.it: <wver in- Unit 2 - 991 Power Found during seiselc lacked positive anchorage to stalled Unit 3 - Shutdown fuel evaluation of control the floor--bolts had never been removed. room panels for mounting installed as specified in original installation instruc-future instrumentation bolts; bolts installed tions during plant construction. and tack welds were Panels rested on structural steel added along with top channel section which was anchored braces.

to concrete floor.

2/28/86 Davis-Besse 86-011 Cyberex class IE equipment Bolting detail missing Plant shutdown cabinet doors for essential during development of A condition outside the instr. AC power not bolted design basis existing procedures. since plant startup.

Closed as required by seismic design (battery charges. Bolts insta11ed during inverters, recifiers, distribu-tion panels).

seismic qual. test were believed to be shipp-ing bolts.

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6-Table 1 (continued) .

DATE PLANT LER EVENT ROOT CAUSE PLANT STATUS COMMENT 12/13/82 Davis-Be.se 82-066 Auxiliary shutdown panel Design and fabrication error. Shutdown. Mechanics trouble-terminal board not seis- shooting a problem cally mounted.

found seismic mount missing--mounting screws had been removed.

11/19/85 Browns Ferry 85-016 Eight local reactor protec- Construction Configuration 1,2,3 85-020 tion panels on each unit Units 1-2 Refueling; Analysis showed Unit 3 never documented. 3 - shutdown as-built was seismic-not constructed per design ally qualified. Unit 2 drawings, had 2/8 panels not quellffed since anchor bolt mat'erial could not

be determined. Status of Unit 1 to be deter-mined.

4/18/85 Browns Ferry 85-013 Configuration of moua(s f*.' Change in intended use of Unspecified Panels used for startup i

i startup test instrue- ..ston panels from temporary use testing following panels relative to si.smic during startup testing to refueling and for qualifications is in deter- operational use.

minate. Panels are plywood special testing.

with sheet metal cabinet, Installation field improvised for it was fastened to the masonary wall originally intended to with the masonary anchors. be removed after Cables lay unsecured on main initial startup.

control room panels and could adversely affect safety-related panels / functions, i

1 Table 1 (continued)

DATE PLANT LER EVENT ROOT CAUSE PLANT STATUS ColeqENT 8/14/81 0.C. Cook 81-013 Switchgear cabinets, reactor Design deficiency in Not Specified Response to GL 81-14

! trip ano bypass breaker cabinets cabinet anchors. found problem (Aux.

i lacked sufficient seismic re- Feedwater System straints to prevent overturning Seismic Qual.)

during SSE.

1 8/15/83 0.C. Cook-2 83-073 Safety related battery chargers Nuts not installed during Not Specified not adequately seismically construction.

ancnored - anchor bolt nuts not installed as required by design.

! (LER 82-047 pevious similar occurrence.)

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.In each case the deficiency was related to construction installation errors.

The original designs correctly specified the seismic anchorages or other seismic support. The deficiencies had gone undetected since construction inspite of QA inspections, IN 80-21, and the high visibility for such concerns as indicated by its designation as USI A-46.

Other Evidence of Deficiencies In conjunction with staff efforts to resolve USI A-45 (Decay Heat Removal), NRC contractor staff from Sandia National Laboratory have conducted six walk-through inspections regarding 'seismic fragility of equipment essential for accomplishing decay heat removal. During these inspections, additional evidence was found of inadequate and missing seismic anchorage. For example, two plants lacked anchorage for 4Kv switchgear busses; two plants lacked anchorage on bus switchgear and motor control centers; in three plants the battery racks were inadequately supported or lacked anchors; and one plant had a deficiency like Dresden 2/3 in that its control room control panels had never been anchored to the floor. Cooper, one of the plants inspected by Sandia in 1984, reported a similar deficiency to the one described in the 1986 LER; the utility had performed inspections for such problems after the Sandia audit in 1984 and concluded that no other deficiences existed. Further, the Sandia audits illustrate a need for a careful well-thought out inspection procedure. For example, in one plant, Sandia determined that even though the cabinets appeared to be anchored at their base, the bolts did not penetrate through the steel channels to the floor.

Safety Significance of the Operational Events Each of the three recent LERs involving inadequate seismic mounts for critical electrical or control cabinets resulted in plant conditions that were associated with unanalyzed safety concerns. Each instance involved significant departures from the plant design basis and assumed conditions for the accident analyses in the plants' Final Safety Analysis Reports (FSARs). In those analyses, only a single failure is assumed as well as the availability of redundant or backup systems required to perform the essential safety functions.

In these events, redundant safety systems could be assumed to have failed in the event of an SSE since common-mode deficiencies were found (Cooper, switch-i gear for both EDGs; Davis-Besse, electrical cabinet doors for four redundant 120 VAC instrument channels; and Dresden's control room panels). Further, since multiple reactor safety systems would have been affected, the resulting incidents could have been more severe than those analyzed in the FSARs. However, final conclusions regarding the risk significance of these or any other cases of -

inadequate seismic anchorage of critical safety equipment must be determined by site / plant-specific design and as-built conditions. For example, lack of seismic mounts does not necessarily constitute deficiencies leading to failure of function because much of the equipment is laterally supported by other cabinets and the connecting cables will tend to prevent overturning. Nevertheless, because of the potential for common-mode failures of redundant systems and for simultaneous s

. failures of multiple safety systems, there is a strong basis to expedite resolution of this issue. As an indication of the urgency of this problem, in each case where plants were at power when the deficiencies were found (Cooper and Dresden 2), the licensee demonstrated recognition of a potentially serious safety concern by taking immediate actions to bring the plant to a safe shutdown condition. (Davis-Besse and Dresden-3 were already shutdown.)

Analysis of Regulatory Activities The adequacy of the seismic design, in particular the anchorage of critical safety-related equipment at older operating plants, has been the focus of numerous regulatory programs and reinspections since 1980. Some plants (e.g.,

Dresden 2 and 3) have been the focus of multiple inspections and walk-through inspections by independent groups focusing on the adequacy of seismic anchorage.

For example, during construction, utility QA/QC and NRC regional staff performed audits and inspectians; then the NRC and utility conducted audits with regard to the SEP program, which were followed by more detailed audits in response to IE IN 80-21. Then the SQUG conducted a walk-through to identify the types of electrical and control equipment used in nuclear plants and to ascertain the characteristics of the anchorage used for eight classes of the equipment.

Finally, the SSRAP conducted a walk-through to verify the SQUG findings.

Nevertheless, in April 1986, Comonwealth Edison identified the fact that the control room control panels had never been anchored to the floor, even though their response to IN 80-21 indicated that there was no deficiency in these anchors.

At least one other plant (Cooper) has been the subject of an earlier review (USI A-45 study) regarding as-built seismic fragility of electrical equipment.

Although this walk-through and evaluation occurred in 1984 when the utility conducted some related reinspections after Sandia found a switchgear mounting deficiency, Cooper found additional deficiencies associated with its emergency diesel generator switchgear cabinets in 1986.

In no case were we able to develop information regarding the specific procedures for or scope of the previous inspections and walk-throughs. Nevertheless, it is evident that significant deficiencies were missed. Therefore, in order to develop greater confidence in the proposed reinspection procedures and checklists for use in the generic resolution of USI A-46, we suggest that before they are implemented, the specifics of the procedures be reviewed by experts who are familiar with the previous reinspections and walk-throughs, and/or that documentation be developed to illustrate why the new program will overcome the limitations of the previous efforts.

The schedule projected by NRR for resolution of USI A-46 and its full implementation is to issue a generic letter in October 1986 followed by completion of generic reinspection procedures and checklists by 1987, trial applications by early 1987, and training seminars for utility personnel in the first quarter of 1987. Then, the utilities would develop lists of safety equip-ment essential to achieve a safe, hot shutdown following an SSE and conduct plant-specific walk-through reinspections. Final implementation of USI A-46 fixes would be scheduled as part of each plant's living schedule, with a target-for completion within 28 months of the generic letter. It is important to

i note that, as indicated in NUREGs-1030 and 1211, the fundamental assumption underlying the prcposed generic resolution of USI A-46 is that no gross deficiencies exist in the seismic anchorage of critical equipment. Further, the SSRAP recommended that prior to detailed reinspections, walk-throughs be conducted to verify the absence of gross mounting deficiencies (see NUREG-1030,

p. 2-30). We agree with the SSRAP recomendation. Thus, we suggest that the proposed generic letter be modified to require early walk-through inspection of operating reactors to verify the absence of gross seismic anchorage defects before the detailed reinspections are conducted.

The generic letter will provide the bases for regulatory action and utility reinspections. However, it only refers to continuing deficiencies and does not provide the details of specific, recently identified deficiencies in seismic anchorage of critical safety equipment that would clearly define the scope and clear presence of current problems, or the limitations of walk-throughs and reinspections. Thus, it is suggested that an updated Information Notice be issued as a supplement to the generic letter to illustrate the continuing inadequacies in the seismic anchorage of critical safety equipment.

l The adequacy of the reinspection procedures is critical to a successful resolution of this generic issue. Thus, reinspection without sound guidance i and QA controls would be a waste of time and resources. However, the implementation phase of detailed reinspection could be expedited by identifica-tion of the equipment essential for achieving a safe, hot shutdown while the reinspection procedures are completed. Thus, once procedures are available and

, staff trained, reinspections can proceed expeditiously. The results of the reinspections could provide a basis for decisions regarding the need for further interim safety measures and scheduling of long-term corrective actions.

USI A-46 was conceived and its proposed resolution is directed, based on an explicit assumption that modern seismic design and equipment qualification requirements (namely,IEEE 344/1975), underpinned by prelicensing review by the staff's Scismic Qualification Review Team (SQRT), provide reasonable assurance of seismic safety for those plants. Thus, the generic letter is proposed to include only the 70 operating units that were not designed in accordance with IEEE 344/1975. Whether or not this assumption is valid depends

on the overall thoroughness of the utility's QA efforts, NRC prelicensing inspection and the scope of the SQRT reviews. In light of the fact that the deficiencies discussed herein relate to construction, rather than design defects, we suggest that the proposed limitation of the reinspection to 70 operating

! reactors be reconsidered.

1 FINDINGS i

As a consequence of this engineering evaluation the following findings may be made:

Deficiencies, in some cases gross defects, in seismic anchorage due to absence of anchors exist in operating nuclear power plants despite a 1980 Information Notice on the same subject.

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• Deficiencies in seismic anchorage of electrical cabinets and related l control panels exist in redundant safety systems and could result in disabling of redundant and multiple reactor safety systems.

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Anchorage deficiencies continue to be identified by multiple, indepen-dent programs (LER, USI A-46. USI A-45 and fragility research program) and during reinspections by licensee and other parties.

The safety significance of the events involving inadequate seismic anchorage is dependent on site / plant-specific design and as-built conditions. However, when redundant equipment and/or multiple safety systems would be susceptible to loss of function due to a single seismic event, the potential for exceeding the design basis of the plant is increased.

The safety significance of the adequacy of the seismic anchorage is clearly indicated by the conclusions of the SQUG, SSRAP and staff that given adequate seismic anchorage, the seismic margins of the older operating nuclear power plants is deemed adequate. Conversely, without adequate anchorage, there could be serious concern with the seismic margins for the plants.

The identified deficiencies in seismic anchorages are construction related rather than design related.

CONCLUSIONS

1. There is continuing evidence of continued inadequacy, in some cases gross deficiencies, of seismic anchorage for critical safety equipment due to missing anchorage.

2. The resolution of USI A-46 generic issue is expected to result in a reinspection requirement and procedures that should lead to closure of this concern at older operating reactors.

SUGGESTIONS

1. It is suggested that the Office of Inspection and Enforcement (IE) consider issuing an IE Information Notice to all Ifcensees and possibly owners of NTOL plants concerning the recent evidence that inadequacies continue to exist in the seismic anchorage of equipment essential for accomplishing a safe, hot shutdown. Issuance of an Information Notice characterizing the recently identi-fied problems in spite of previous, related reinspections and walk-throughs would provide details of the extent and variation of the continuing deficiencies and would emphasize this regulatory concern. This seems justified in Ifght of the evidence of the inadequate response to IN 80-21. The Information Notice should discuss the recent experience, the potential safety implications, past and present regulatory attention to this concern (i.e.. IN 80-21 and USI A-46) and reference the generic letter to be issued to resolve US! A-46.

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2. The fundamental assumption underlying the proposed resolution of USI A-46 is that gross deficiencies do not exist in the seismic anchorage of operating nuclear plant safety equipment. The operational experience seems to indicate that this assumption may not be valid. Thus, we suggest that the proposed generic letter be modified to require early walk-through inspection of operating reactors to verity the absence of gross seismic anchorage defects before the detailed reinspections are conducted.

3. We suggest that, prior to issuance of the generic letter, the scope of the applicability of the letter, which would exclude reactors that have been designed to the IEEE-344/1975 standard, be reconsidered in light of recent experience that demonstrates construction and QA deficiencies are the root cause of inadequate anchorage rather than primarily design deficiencies.

4. We suggest that the generic, reinspection procedures be reviewed (e.g.,

reviewed by ex walk-throughs)perts who are familiar with previous reinspections andand/or do i

the new program will overcome the limitations of the previous efforts to assure that limitations in walk-through inspections associated with IN 80-21, .

SEP, SQUG and SSRAP activities are remedied by the generic procedures. '

5. We suggest that NRR consider whether the proposed USI A-46 generic letter requirements to identify equipment categories essential for achieving a safe, hot shutdown could be identified prior to availability of the final reinspection procedures rather than serially after the procedures and i

training of inspection personnel have been completed. Parallel development of equipment lists would somewhat expedite the reinspection efforts and would aid in accomplishing preliminary walk-downs for identification of gross deficiencies.

4 J

9

\

t

REFERENCES

1. Licensee Event Report (LER)86-011, Docket Number 50-346, Davis-Besse 1, March 27, 1986.

2. Licensee Event Report (LER)86-009, Docket Number 50-298, Cooper Nuclear Station, March 18, 1986.

3. LicenseeEventReport(LER)86-009,DocketNumber 50-237/249, April 29, 1986.

4. SECY-85-277, " Unresolved Safety Issue A-46, ' Seismic Qualification of Equipment in Operating Nuclear Power Plants,'" August 20, 1985.

5. Final Draft NUREG-1030 " Seismic Qualification of Equipment in Operating Nuclear Power Plants Unresolved Safety Issue A-46."

6. Final Draft NUREG-1211, " Regulatory Analysis for Proposed Resolution of ,

Unresolved-Safety Issue A-46 Seismic Qualification of Equipment of Operating Plants."

7. IE Information Notice Number 80-21, May 16,1980.

8. Letter from R. F. Janecek, Commonwealth Edison Company, to D. M. Crutchfield, NRR, USNRC

Subject:

"Dresden Station Unit 2 SEP-Positive Anchorage of Safety-Related Electrical Equipment NRC Docket No. 50-237."

i I

i l

mw. . - - - -

. a . m .. .

thesd M 0 805 Te4 phone (3011654 9260 TWX 7108249602 ATOMIC FoR DC June 25, 1984 2

i Mr. Victor Stello, Jr. -

Deputy Executive Director for ~

Regional Operations and Generic Requirements U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Stello:

The AIF Committee on Power Plant Design, Construction and Operation prepared an industry position paper regarding sels- -

nic' qualification of safety-related equipment in operating nuclear power plants and submitted this paper to Mr. Williaa Dircks on June 17, 1983. This paper enclosed as Attachment 1,  ;

concluded that historical and current seismic data demonstrate '

that seismic qualification of equipment in operating nuclear power plants is not a significant safety issue. The AIF Committee on Power Plant Design, Construction and Operation has been studying this issue for the past two years and most recently reviewed the draft Regulatory Analysis of USI A-46. ,

The comalttee's specific comments are enclosed as Attacheept 2 '

for your consideration. &

e In another industry effort, an owners nic Qualification Utility Group (SQUG) ,group called compiled the Sels-earthquake experience data from conventional power plant equipment and .

performed a detailed comparison with data from siellar equip-

, ment in nuclear power plants. The SQUG documented their con-clusion from the study in a letter to NRC's Harold Denton  !

dated February 17, 1984 which stated that "the functional and l

structural adequacy of properly anchored nuclear power plant "

equipment is not a significant safety issue". Thls conclusion  ;

was based on a study of eight categories of equipment in nuclear power plants.

The NRC Staff has sonitored the SQUG efforts closely and approved the selection of an independent third party review  !

group of seismic experts called the Senior Seismic Review and '

Advisory Panel (SSRAP). The SSRAP concurred with the SQUG concluslons for the eight categories of equipment that were i

I F6sS1205- A A/b- i

i L

• Mr. Victor Stello June 25, 1984 .

l l

! studied with caveats regarding anchorage, materials considera- '

) tions, and location relative to other equipments. The SSRAP i

did not make any judgements beyond the eight categories of -

l equipment that they reviewed.

i Harold Denton and other NRC Staff members met with SQUG and i

SSRAP on March 20, 1983. Mr. Denton concurred that " seismic qualification is not a significant issue" but that SQUG should

"give further thought to other classes of equipment and to r anchorage". However, the SQUG selected the eight classes of i

equipment to be representative of nuclear plant equipment.

i The SQUG data verifies that power plant equipment is inherently l

! rugged and that it is not susceptible to seismic damage when

! properly anchored. Anchorage of equipment during a seismic event is a recognized issue that the Industry has been addressing for some time. The industry has addressed the anchorage issue through the NRC's IE Bulletin 79-02 and IE Notice 80-21.

In addition, the NRC's consultant, Lawrence Livermore Labora- t j tories has tested electrical and mechanical equipment under j theSelsaicSafetyMarginsResearchProgram. As a result of 1 this seismic fragility testing, Lawrence Livermore has con- l cluded that the equipment is inherently rugged and will sur- l

, vive acceleration levels far in excess of building responses ,

associated with the Safe Shutdown Earthquakes regardless of l

! whether the equipment was qualified for seismic service. f 4 Futhermore, Lawrence Livermore has studied the SQUG effort and

concluded that it is feasible to address the seismic issue with j ,

' experience data. ,

1 l

j Relay chatter during a seismic event is a recognized issue that the industry has been addressins. However, relay chatter has ,

so many unique aspects that it Is difficult to make generic '

conclusions. We recognize that relays can chatter during an '

i earthquake and this has been considered in the design of  !

! electrical circuits and relay applications. IEEE Standard .

j 501 1978 recognizes that chattering will take place and j specifies the degree of chatter that is acceptable.

j As reported in the Regulatory Analysis, Brookhaven National l; 1 Laboratory (BNL) has performed a Probablisitic Risk Assessment j (PRA)under the NRC contract to r'esolve USI Task A 46.  ;

i i

i

Mr. Victor Stello -

3- June 25, 1984 The results of the PRA indicate that both core melt frequency and risk are dominated by st'ructural failures. BNL concluded, in part that, "...overall seismic induced structural failures and random failures (due to non-seismic causes) contribute more significantly than seismic induced equipment failure". BNL points out in their conclusions- that their analysis results basically indicate that upgrading equipment from the baseline (assumed fragility levels) does not introduce a marked reduction in risk.

We conclude from all the evidence that there is reasonable assurance that the seismic adequacy of equipment in nuclear plants is acceptable with adequate margin for a design basis earthquake. Furthermore, upgrading this equipment by documen-tation or increasing its resistance, as required by the pro-posed generic letter (Appendix A to the Regulatory Analysis),

is not justifiable based on a cost-benefit assessment.

Sincerely, ,

1 .

Joseph W. Williams, Jr.

Chairman, Committee on Power Plant Design, Construction and Operation ,

, JWW:bif Enclosure cc: Mr. Richard H. Vollmer - NRC Mr. Vincent S. Noonan - NRC , -

f p, , - ~ - -en-,- --:e--e-- - - - - - - - - - - ---,-----+-,,,-,-c--- - ---e ~ ,---o - -nn,----.--, -- ~-,-w,n-- - - , , , - - - , - - - , -

p..**. ,

Attachment 1 A NUCLEAR INDUSTRY POSITION REGARDING SEISMIC QUALIFICATION OF SAFETY-RMLATED EQUIPMENT IN OPERATING NUCLEAR POWER PLuf75

• 1. INTRODUCTfDN The purpose of this paper is to present a nuclear industry position concerning the seismic adequacy of safety-related equipment installed in operating nuclear power plants.

Since the advent of commercial nuclear power, seismic design and qualification criteria as well as the ana-j lytical and experimental methods used to verify seismic adequacy have undergone a significant evolution. In this

' regard, aqclear plant designs have included conservative enginetring practices to assure, among other things, compensation for uncertainties 10 describing seismic j' Icading and systes response. Operating plants have been designed, constructed, and maintained according to safety

,, requirements established at the time of their design and construction.

II. DISCUSSION A numbat of operating nuclear plants were issued operating licenses prior to February 20, 1971, -(the published date of the General Design Criteria (GDC) for Nuclear Power Plants). There is no substantive indi-i .

cation that industry practices and procedures developed and laplemented prior to the GDC for seismic qua} fication are inadequate. To the contrary, a recent studyL ),

addressing the seismic performance of equipment installed' in conventional power plants, found that the equipment performed very well during and after strong earthquakes.  ;

More recently a study published by Lawrence Livermore Laboratory evaluating the use of experience data to assess seismic performance of equipment in non-nuclear facilities during earthquakes has concluded that it is feasible to use seismic experience data to address seismic plants.(gguipment 37 qualification issues in operating s

p

{ . '* 4 I

The overwhelming majority of equipment in non-nuclear

! ~{

t power plants has been manufactured and installed over the years to proven industry design and testing standards. ,

Equipment which performed well in acn-nuclear facilities during severe eartnquakes is designed to the same industry standards as similar equipment installed in nuclear f acil-ities. The major differences between this equipment and current nuclear grade eMuipment of the same class, is its documentation to current standards. ,

Mistorical experience data has shown that equipment act vigorously seismically qualified by special test er

! analysis has nevertheless performed geceptably in earth-quakes. For example, an WRC report (8) which reviewed two tossil units experiencing an earthquake with a Richter [

magnitude of 6.6 documented that. *the two operating units  ;

4 were safely shutdown with no known malfunctions of elec-j trical control and instrumentation equipment during(pg after the earthquake". In addition, a recent study As j has shown that the structural integrity of anchored power plant equipment and components is not comproatsed in i strong earthquakes of up to 0.5g peak ground acceleration.

i Furthermore, typical power plant equipment operability is l not compromised in strong earthquakes with peak acceler.

ation of about 0.2g to 0.3g. It is entirely reasonable to ,

conclude that equipment will survive a severe earthquake

<-- because of th'e conservatism applied both in the design 1k of structures and equipment and in the generic design  !

i requirements to meet national standards.  ;

An important observation that een be made from fragility descriptions (4) is that most mechanical and electrical  :

i equipment is inherently rugged and will survive acceleration '

j levels far in excess of building responses associated with ,

i the Safe Shutdown Earthquake (55E) regardless of whether the equipment was qualified for seismic service. Iguipment l ,

and supports that fail in a structural mode, and which have i had specific designs to resist seismic loading, typteally have median capacities of several times the 55E level of  ;

l earthquake specified for design. l

!!!. POSITION

. A. Present day criteria! e.g., IEEE-344 Standard j requirements, should not be laposed on elder ,

' operating plants. Seismic information.is available i i to verify that industry practices and procedures are i

adequate and acceptable even wnen judged against j requirements of current standards.

l l

i

I* , .< . ".

^ '

3. Upgrading seisafe qualification for the purpose of

(. satisfying the latest criteria and documentation requirements is not justified because it will act reduce risks significantly.

C. The use of earthquake experience data on similar equipment is an acceptable means of judging the selsnic operability of equipment in an operatiaq plant.

IV. CONCLUSION Historical and current seismic data demonstrate that seismic qualification of equipment in operating nuclear power plants is not a significant safety issue.

REFERENCES

1. EOE Incorporated - Program for the Development of an t

Alternative Approach to Seismic Equipment Qualification, 1962.

2. Lawrence Livermore Laboratory - UCID - 19465 Draft.

October 1, 1982.

3. NUREG/CR-1665, ' Equipment Response at the El Centro Steam

.k

! Plant during the October 15, 1979, taperial Valley Earthquake,' prepared by Lawrence Livermore Laboratories for USNRC.

4. WUREG/CR-2405, 'subsysten Fragility Seismic Safety Margin l Research Program."

e l

(

I s

2 3

i

I 1

Attachment 2 1 i

Specific Comments on NRC Regulatory Analysis of USI A-46 Requirements Page 5, 1st paragraph

...the staff has concluded that it is unnecessary to perform explicit seismic qualification on the eight classes of equipment studies. This conclusion can be extended to other classes if additional data are collected and presented."

Comment:

We do not agree that it is necessary to extend the equipment studies to other classes of equipment. We do not believe that collecting data on additional equipment classes will more effectively demonstrate that

" equipment installed in nuclear plants is inherently rugged and not susceptible to seissile damage".

Page 5, 2nd paragraph i

...although equipment is inherently rugged and not susceptible to l seismic damage, failures due to seismic loads are likely to occur if equipment is not adequately supported or anchored."

1 Comment:

i We agree that equipment should be properly anchored. However, any

! programs for verification of anchorage that may be required should be

} allowed to take credit for verificat on activities that have already been i accomplished under other programs such as response to IE Bulletin 79 02 '

and IE Notice 80 21.

8 i

J l t t t

I

~

2-Page 10, Table 1, Cost Estimates of Seismic Verification Alternatives Comment 1 on Table 1:

The estimates for equipment replacement are very low. Following are some examples: , ,

(a) Air Conditioning Unit Replacement NRC - High Value $260,000 Actual Experience

- Equipment $225,000

- Installaticn $275,000

$500,000 (Double the NRC high estimate)

(b) Battery Racks Replacement l NRC High Value $ 5,000 Actual experience $ 15,000 per rack

. ($6,000 of that is installation)

(c) Transformer Replacement .

- NRC High Value $ 8 500.

, Actual Material and qualification 8 33,,000 Installation $ 37,000

$ 70,000 per transformer t

(d) Large Vertical Pump /

Motor Replacement

- NRC High Value $160,000

' Actual cost of just the motor .8600,000

( a s t e r i s t a n d 7q u a i f i c a t i o n -

not installation)

Actual Cost to rebuild re- $135,000 circulating pump motor (not including removal and replacement)

(e) Reactor Trip Switchgear Replacement NRC High Value $ 73,000 Actual experience $130,000 per reactor switchgear

Comment 2 on Table 1:

The average comparison costs are definitely low across the board except for equipment which is very similar. This comparison has to be made, for each piece of equipment, by a discipline engineer (mechanical, olectrical) and by an engineer familiar with structural dynamics. Even if no calculations are made, the engineers should understand well the dimensions, mass distribution, orientation and type of anchorage for the data base and the plant equipment. These activities cannot be comaleted within the range of cost shown. We estimate the real cost to be a)out 2-3 times higher than indicated in the table.

Page 11, paragraph under " Replacement"

" Estimates are primarily based on Process Plant Construction Estimating Standards by Richardson Engineering Services, Inc.... Estimates taken from the 1975 edition were increased by 30% to account for inflation."

(to equal to 1981 standards)

Comment:

We refer to the enclosed Handy-Whitman Index of Public Utility Con-struction Cost, Bulletin #116-Electric Indexes dated January 1, 1983, Section Nuclear Production. When using this bulletin, estimate in-creases from 1975 to 1983 are 83%, not 30%. Very rarely is Richardson used to do nuclear estimations. Its escalation rates are multiplied by 3 to 5 since it does not cover health physics, medicals, decontamination, respirator work, etc. Also nothing is mentioned about removal, decon-taminating and shipping offsite the components being replaced.

Page 13, Table on Estimated Costs to Licensee Comment:

The overall costs estimated by NRC are low. Typically for a PW?..

we anticipate a minimum of 40 electrical pieces of equipment and 70 cechanical 11eces of equipment would have to be evaluated /.nspected/

reported. 'ollowing are specific comments regarding the costs en page 13.

(a) " Define scope of system, subsystems, and components requi ed"-

This activity includes listing of systems and support systems, review of applicable one line diagrams and listing all compo-Then, all available drawings and/or vendor nents by(tag number.if any) are gathered and compared to the documentatio reports of the database, for proper classification of equipment. All these activities could cost aboat $15,000.

Our estimates for this activity range to a high of $100 000 if a system analysis must be done similar to Appendix R requirements.

(b)

" Compare database spectra with site spectra" - If comparison is for ground spectra, the NRC cost estimate is reasonable. If generic spectra aust be developed for the plant floors (based floor spectra, the total cost could be around $4,000.u -

(c) procedure electrical e must be developed for each category o uniform way.quipment so that the anchorage is inspected in a without We estimate the walk-through to cost about $35 000 considering the costs related to health physics training decontamination costs, scaffolding for on-line valves which can ,

not be inspected from platforms, deenergizing and reenergizing the electrical equipment, and need to have electricians dis-welds or anchor bolts. connect or push out of the way wires inside cabine easily double the amount above to about $70,000 - $100,00 Our low estimate is $20,000 for this activity.

(d)

"toRepairs estimateto anchorage'and supports" - This item is very difficult making easy, repairs to no more than 2-3 pieces of e Actual costs of recent repairs to anchorage and supports averaged $35,000 per pipe hanger and $47,800 per electrical support (material, engineering, and labor)

(e)

" Qualification or replacement of relays" - The replacement of relays would be the method opted by most utilities. This Furthermore, the cost does notassumes that approximately 12 to 20 gizing of cabinets and panels, testing and quality assuranceincl activlties.

times higher thanWe estiaste the total cost the NRC estimates, to be$50,000.

or about at least five low estimate for this activity is $10,000. Our (f).

" Miscellaneous modifications to components to fit experience data" - These support modifications appear to relate primarily to the systems.

Should additlonal supports or~ bracing be needed eight the NRC cost estimate allows for no more than four to supports.

(g)

" Procurement of additional experience or test data" -

Procurement of additional experience is an unknown factor.

5-Additional test data may sometimes be procurred from the vendor, however .this data is sold for about $10,000 to $40,000 per report. Tests on electrical devices can cost about $10,000 per test (including preparation of-test procedure and review of test) without counting the cost related to removing and rein-stalling the device. The cost to the 20 utilities in SQUG for experience data on eight categories of equipment has been $1 million or $50,000 per participating utility.

(h) " Preparation and submission.of r.eport to NRC" - We estimate t this cost more on the order of $50,000. Our estimates range from $5,000 to $100,000 depending on the level of detail re-quired and the degree of subsequent interaction with the NRC.

j Page 15. Schedule for Implementing the Proposed Requirements Comment :

The elapsed time from date of requirement should be tied to plant refueling cutage schedules rather than a set number of months after j the requirement is imposed.

Page 17, paragraphs (2) and (3):

i "BNL concluded, in part that, ...overall seismic induced structure j failures and random failures (due to non-seismic causes) contribute

core significantly than seismic induced equipment failure."

i "B'NL points out in their conclusions that their analysis results '

basically indicate that upgrading equipment from the baseline (assumed fragility levels) does not introduce a marked reduction in risk..."

Comment:

t Based 'upon the BNL report, the LLL report (NUREG/CR 1665), the SQUG

• report, the estimated cost to reverify'the seismic adequacy of certain equipment, and various pr grams that have previously been undertaken 1

to verify and correct equ puent anchorage and supports, we are of the cpinion that the program n the proposed generic letter is not cost

_and safety effective.

APPENDIX A. PROPOSED GENERIC LETTER (page 3-4)

For components not included in the data base, the licenssee may extend oxperience data base, provide test data or analysis or perform specific qualification procedures outlined in current licensing requirements.

Comment:

I This requirement implies requalification or replacement. According to the BNL study, this requirement does not introduce a marked reduction

, in risk.

k

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COST TRENDS OF ELECTRIC UTILITY CONSTRUCTION.

NORTH ATLANTIC REGION 1973:10 0

(

Cost INDet hWWHks 5- *

< O 2

cosste uevion ano sourutwt Q .

y = c c 2 c s e t e R -

= = x = 2 E .

a j t t t t t t t t e E z E a t E 3 1 Total Nat.All 5:ess Generation .. 10 10 10 10 13 15 18 19 21 19 18 18 19 19 18 1 2 Total Nnt. All St. & Nucleer Gen. . . - - - .

-- - - - - - - - - - - 2 3 Total Plant. All st. & Hydro. Gen. . . . - - - 10 12 15 18 19 21 20 19 19 19 19 19 3 4 4 5 STEAM PRODUCTION PLANT 5 6

7 8

Total Steam Production Nnt ......

Structurn & Improvements.Indt. ... 3ll Struct.& Improv Semi-Outdr. .... 311 8

8 8

9 9

11 il 15 16 18 17 18 18 25 20 19 18 17 18 18 18 19 19 18 18 18 18 6

7 8

9

(

9 Soikt Plant Equipment Coal Frd. 312 8 8 8 8 10 16 19 17 18 16 14 15 17 16 16 10 11 Boller Nnt Piping lastalled . . . . . . . 10 10 10 8 11 18 19 20 18 18 16 17 18 18 18 . 11 10 (

12 Turbogenerator Units . . . . . . . . . . . . . 314 9 9 9 9 13 14 17 19 22 23 20 19 19 19 19 12 13 Accessory Dectrical Equipment .... 315 14 14 14 14 16 18 21 26 27 27 26 26 26 26 26 13 14 15 Mac. Pow er Nnt Equipment . . . . . . 316 - - - - - - - - - - - - - - - 14 15 (*

16 NUCLEAR PROD. PLANT . 16 17 Total Nuclear Prod. Nat . . . . . . . . . - - - - - - - - - - - - - - - 17 18 Structures & improteaunts ...... .. 321 - - - - - - - - - - - - - - - 18 19 Reactor Plant Equipment ... ... ... . 322 - - - - - - - - - - - - - - - 19

-. 20 20 21 HYDRO PRODUCTION PLANT 21 22 Total H3 draulic Prod. Nnt ....... - - - 8 9 13 14 15 16 16 15 15 I6 16 16 22 23 Structurn & improvements ........ 331 - - - 9 11 16 17 18 20 18 18 18 19 18 18 23 24 Reservoirs. Dems & Waterways .... 332 - - - 7 9 13 15 15 17 17 16 16 17 16 17 24 25 water % hnis. Turbines. & Cens. . . . 333 - - - 7 9 il 12 13 14 13 12 12 12 12 12 25 26 26 27 OTHER PRODUCTION PLANT 27 28 Total Other Prod. Nat . . . . . . . . . . . - - - - - - - - - - - - - - - 28 (

29 Fuel Holders & Access. . . . . . . . . . . 342 - - - - - - - - - - - - - - - 29 \.

30 Cas Tur bogenerators . . . . . . . . . . . . . 344 - - - - - - - - - - - - - - - 30 31 31 32 TR ANSMISSION PLANT 32 33 Total Transnussion Plant . . . . . . . . . . Il ll 10 11 13 16 19 20 22 21 20 20 20 20 20 33 34 , Santion Equipmen,t . . . . . . . . . . . . . . . 353 15 15 15 15 17 20 25 27 30 30 28 28 29 29 30 34 35 Towers & Fistures . . . . . . . . . . . . . . . 354 8 8 8 8 11 15 16 16 16 16 15 15 16 16 16 35 13 36 36 Poles & Fis t ures . . . . . . . . . . . . . . : . . 355 6 6 6 6 7 3 12 14 15 13 12 13 13 13 37 Overhead Conductors & Devkes .... 356 15 14 13 13 21 23 27 28 29 21 20 21 21 22 22 37 ,

38 Underground Conduit ............ 357 6 7 7 7 8 10 14 16 17 16 15 16 16 15 15 38 l 39 Under8round Conductors & Dev. ... 358 12 11 10 11 16 18 21 22 23 18 18 21 20 20 20 39 40 40 41 DISTRIBUTION PLANT dl  !

42 Tc'ai Distribution Nat . . . . . . . . . . . ll 11 11 11 13 15 19 20 22 21 20 20 20 20 20 42 43 44 Station Eg eipment . . . . . . . . . . . . . . . 362 Poles. Towers & Flatures . . . . . . . . . . 364 15 6

15 6

16 6

16 6

17 7

20 9

25 10 27 ll 30 13 30 14 28 13 28 13 29 I4 29 14 28 43 14 44 (l i l

45 Overtwed Conductors & Devkes .... 365 12 11 10 11 17 19 22 23 23 17 16 17 17 18 17 45 46 Underground Conduit 366 8 s 8 8 9 12 16 18 20 19 17 18 18 18 17 46 47 48 Underer.Cond.& Dev In condt.

Une Tra nsformers . . . . . . . . . . . . . . . 368 367 12 42 11 42 11 42 11 42 16 42 19 45 22 61 23 64 24 68 19 70 19 62 22 to 21 62 21 61 21 57 48 47 kl 49 Pad Mounted Transformers . . . . . . . . 46A - - - - - - - - - - - - - - 49 50 Servkes.Overheed . . . . . . . . . . . . . . . . 369 ll 10 9 10 14 17 20 21 21 15 14 16 16 16 16 50 (

$1 Servien. Underground . . . . . . . . . . . . 369 10 10 ll 12 14 17 20 22 22 18 16 17 17 18 19 51 (

$2 Meters instalkd . . . . . . . . . . . . . . . . . . 370 31 31 31 31 31 35 39 44 46 49 46 44 43 42 42 52 53 Strut Ushting. Overhead . . . . . . . . . 373 - - - - - - - - - - - - 22 22 21 53 54

$5 Mass Arms & Luminaires last. . .... 373 Saint Ughting. Underground ....... 373 23 23 23 55 54

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5 5

6 18 18 19 19 18 17 17 19 19 20 22 22 22 22 24 24 24 24 24 28 32 36 39 40 6 7

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18 18 17 16 15 15 16 16 16 18 18 18 18 19 20 21 21 22 24 28 32 35 38 36 3g 8 9 16 16 16 16 15 14 14 16 16 17 19 19 19 20 21 21 21 21 21 24 27 32 35 37 9 10 10 C 11 12 19 19 19 19 22 19 19 23 19 22 17 21 16 22 16 25 16 26 16 26 18 29 18 30 18 30 19 30 20 30 20 30 21 30 21 30 20 30 23 35 26 42 28 45 29 47 32 11 47 12 49 13

( 13 26 26 27 27 26 25 25 27 27 28 30 30 31 32 32 32 32 31 32 37 43 44 46 14 - - - - - - - - - - - - - - - - - - - - - - 37 39 14 15 15

• 16 16 17 - - - - - - - - - - - - - - - - - - - - - - - - 17 18 C 18 19

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20 20 21 21 C - 22 16 16 16 16 15 14 I4 15 15 16 17 17 17 18 19 17 20 20 21 21 21 21 22 24 24 28 28 32 32 33 35 35 22 36 23 23 18 18 18 17 16 15 15 16 16 16 18 18 18 18 19 24 17 17 17 17 16 14 14 15 15 15 17 17 17 17 18 20 20 20 21 23 27 31 32 34 24 37 25 C 25 26 12 13 14 14 14 13 13 14 16 16 17 18 19 20 21 22 23 23 23 26 31 34 35 26 27 27 28 - - - - - - - - - - - '- - - - - - - - - - - - - 28

( 29 30 29 30 31 31 32

(. 32 33 20 20 22 20 20 19 19 20 21 22 23 23 23 24 25 25 25 25 35 26 35 29 39 34 46 38 49 39 52 42 33 56 34 34 29 29 31 30 29 28 29 32 32 33 35 36 36 36 37 37 37 35 16 16 16 16 16 14 14 15 15 16 17 17 17 18 19 20 20 20 21 23 27 30 32 34 35

( 36 37 12 21 12 23 12 25 12 21 12 19 Il 17 10 18 12 21 12 21 22 12 14 23 14 22 15 22 15 23 16 24 17 26 18 27 19 27 22 28 23 33 28 38 30 40 31 40 32 36 42 37 38 15 15 15 15 15 13 11 14 15 15 16 17 17 17 18 19 19 20 21 24 27 31 32 33 38 51 39 39 19 20 20 22 24 22 23 23 26 27 27 27 27 32 37 44 47

[% 40 21 23 19 19 18 19 40 41 41 42 20 20 21 20 20 19 19 20 20 20 22 22 22 23 24 25 25 26 26 29 34 37 39 41 42 44 46 48 50 43

( 43 44 27 13 26 13 27 13 27 14 27 13 26 12 26 12 28 13 28 13 29 13 31 15 31 15 32 15 18 33 16 19 33 17 19 33 18 21 33 19 21 32 21 22 34 22 22 39 23 26 28 30 31 32 32 32 33 44 34 45 45 17 18 20 17 16 14 15 17 17 17 19 18 46 18 16 16 17 17 18 20 20 21 21 22 23 23 23 26 29 33 34 36 46 f 18 18 17 17 13 33 39 46 50 53 47

( 47 48 20 53 22 52 24 56 20 55 20 53 18 SI 20 53 21 54 21 55 23 25

$9 23 61 24-61 24 61 27 63 28 62 28 58 28 58 28 56 65 81 84 87 91 48 49 - - - - - - - - - - - - - - - - - - - - - - 102 102 49 50 15 17 17 17 17 18 19 20 21 21 25 28 30 30 32 50

[ 17 18 15 14 13 14 15 15 16 20 20 24 24 25 26 26 30 34 37 38 39 51 b 51 18 18 18 17 17 15 15 17 17 18 21 21 52 42 42 42 42 42 42 44 46 48 48 48 48 48 48 49 49 49 49 49 54 61 65 70 70 52 53 20 22 22 22 21 20 21 22 22 23 24 24 23 24 25 26 26 27 27 30 36 39 42 44 53 u

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59 61 60 59 60 60 62 64 1 1 Total Plant.AB Seesa Genetstion ... 44 43 48 49 50 -55 58

- - - - - - - - - - - - - - 63 2 2 Totel Plant. All St.& Nucleer Gen. . . . 63 3 - '

44 47 48 49 54 57 58 59 59 59 59 60 61 3 Tos*!Plaat Al:St.& Hydro. Gen. ... . 43 4 '

4 5 5 STEAM PRODUCTION PLANT 64 62 62 62 63 64 6 44 44 46 48 49 56 61 63 64 6 Total Steam Production Plant ...... 57 7 7 Saructwas & Improvements Indt. .. 311 36 38 39 41 C 46 49 50 52 53 53 54 55 56 .'

56 58 58 56 57 57 58 60 8 8 Struct. & improv Semi-Outdr. . . . . . 311 38 4 d as 45 50 55 67 9 42 43 45 47 53 59 61 64 64 64 64 64 66 9 Boiler Plaat Equipment. Coal Frd. 312 41

-10 10 60 60 61 11 \ '

M 42 47 53 56 58 59 59 11 BoDer Plant Pipin8 l assaned ...... 35 37 40 51 69 68 67 68 69 12 '

314 52 52 $$ $7 58 68 75 80 79 75 12 Turboteneref or Unks . . . . . . . . . . . . . 67 58 59 58 60 65 13 I 60 62 62 66 70 73 72

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13 Accessory Electrical Equipment .... 315 56 57 Misc. Power Plan Equipment ...... 316 40 43 44 46 47 50 $3 M $7 58 59 60 60 62 63 14 '

I 14 15 15 .

16 16 NUCLEAR PROD. PLANT - - - - - - - - - - 64 17 .

17 Total Nuclear Prod. Plant . . . . . . . . - - - - - - - - - - 60 18 (7) 18 Structures & Improvements . . . . . . . . 321 - -

- - - - - 65 19 19 Reactor Plant Equipment . . . . . . . . . 322 - - -

20 '

20 21 HYDRO PRODUCTION PLANT \

21 36 38 40 di 43 47 50 52 54 55 55 $6 57 59 60 22 22 Total Hydraulic Prod. Plant ....... 55 56 57 23 38 38 39 41 43 46 49 50 52 53 53 54 23 Structures & Improvements . . . . . . . . 331 59 24 38 40 di 44 47 49 51 52 53 55 56 57 .

24 Reservoirs, Dems & Waterways .... 332 35 36 66 67 25 43 46 47 49 M 62 65 66 66 65 64 65 ,,

25 Water Wheels,Turbanas.& Gees. .. 333 41 26 26 27 j-27 29 OTHER PRODUCTION PLANT 28 Tosal Other Prod. Plant . . . . . . . . . .

Fuel Holders & Aceses. . . . . . . . . . . 342 71 61 74 72 28 62 29 74 30 f}

Gas Turbo 8enerstors . . . . . . . . . . . . . 344 30 31

'32 TRANSMISSION PLANT 33 . Total Transmission Plant . . . . . . . . . 46 48 50 $1 53 $7 58 60 60 60 58 58 58 61 64 33 31 32

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353 63 65 68 69 70 77 31 84 83 77 69 68 64 68 72 34 34 Ssation Equipment . . . . . . . . . . . . . . 59 62 35 54 56 57 35 Towers & Fistures . . . . . . . . . . . . . . 354 355 36 35 38 37 40 39 41 40 42 41 45 44 48 47 50 48 52 49 53

$1 51 53 54 55 57 36 (7J 36 Poles & F1stures . . . . . . . . . . . . . . . . .

46 48 50 53 56 61 65 64 62 63 63 64 59 63 66 37 37 Overhead Conductors & Devices ... 356 to 38 40 41 44 47 49 50 $1 53 55 56 58 f Conduit ............ 357 35 37 38 71 39 38 8':f ,.E 58 to 61 60 60 61 65 62 64 63 65 68 66 58 Under8round Conductors & Dev. ... 358 30 40 40 41 DISTRIBUTION PLANT 56 M $7 57 58 58 60 62 42 ,

44 45 48 49 50 53 54 , ,

42 Tosal Distribution Pinot . . . . . . . . . .

56 59 61 62 68 72 74 74 72 67 68 66 68 10 43 (

43 Santion Equipment . . . . . . . . . . . . . . . 362 SS 53 54 57 44 M 38 39 40 44 46 47 48 50 50 52 44 Poles. Towers & Fistwes . . . . . . . . . 364 35 49 49 48 50 $1 51 53 54 56 59 45 +

• 39 40 42 45 '

45 Overhead Conduct ars & Devices . .. . 365 37 48 50 52 54 55 57 58 60 61 46

• 39 40 42 43 46 46 Under8round Conduit ........... . 366 38 70 61 61 63 64 63 63 64 69 75 47 67 67 68 71 47 Under8r. Cond.& Develn Condt. . . .. 367 65 115 121 118 114 112 108 98 93 0 95 48 109 112 112 48 Une Transformers . . . . . . . . . . . . . . M8 103 103 102 99 95 94 95 90 90 49 102 102 102 102 102 102 102 '

40 Pad Mounted Transformers . . . . . . .. 368 102 45 47 47 48 49 51 54 50 37 39 39 42 45 43 43 50 Service + Overtend . . . . . . . . . . . . . . .. 369 35 44 42 42 45 46 48 52 51 45 44 44 46 45 43 l 51 Services.Under8tound . . . . . . . . . . .. M9 44 45 14 78 80 82 83 82 82 82 82 82 52 70 73 74 71 f 52 Meters Instaued . . . . . . . . . . . . . . . . .. 370 70 65 65 64 64 64 65 66 67 53 .

l 50 $4 54 57 62 '

Street U8 htin 8. Overhead . . . . . . . . .. 373 48 $1 68 54 53 70 70 66 66 66 65 65 67 ,

- - - 57 64 54 Most Arms & buninaires last. ..... 373 - 61 62 60 60 61 61 62 55 '

55 $steet U8 h tin 8.UA " .. ..... 373 46 46 47 51 53 54 59 61 56 56

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COST TRENDS OF ELECTRIC UTILITY CONSTRUCTION,

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5 6 66 68 70 74 79 87 95 100 104 119 129 137 140 144 149 155 159 166 175 183 193 199 210 219 6 64 68 75 84 92 100 106 118 124 128 126 131 132 137 140 148 154 162 174 182 187 192 7 7 59 61 6J M 77 85 92 100 105 125 135 139 134 138 138 143 144 154 160 169 183 190 197 201 8 8 62 71 9 69 71 73 76 81 88 95 100 104 121 133 141 145 152 155 162 167 175 183 191 202 209 220 230 9 10 4

] 10 11 12 63 70 M

72 68 72 71 75 77 80 88 89 97 98 100 100 102 102 113 112 121 116 124 128 128 136 135 137 138 146 149 153 152 156 159 163 168 174 175 IRO 188 189 189 195 198 210 209 11 218 12 13 66 71 74 77 82 88 94 100 105 l16 129 135 141 144 146 158 158 160 165 176 179 187 198 201 13 14 65 68 71 75 80 87 94 100 105 115 123 128 131 135 141 147 153 157 165 175 181 188 200 212 14

' 15 15 '

16 16 66 69 75 88 95 100 103 lid 121 127 132 136 141 I46 149 155 162 170 177 184 193 200 17 T 17 11 81 79 93 100 104 120 124 125 130 133 137 140 146 152 161 167 175 179 166 la J 18 62 64 67 13 83 87 89 96 100 103 lio 114 122 128 135 139 143 147 151 156 163 171 179 187 194 203 19 19 67 70 72 76 20 7 20

.21 21 22 63 65 68 73 79 86 94 100 105 116 123 129 130 135 138 143 146  !$3 160 170 178 186 194 200 22 23* 59 61 64 68 75 84 92 100 106 118 124 128 126 131 132 137 140 148 154 162 174 182 187 192 23 24 61 M 67 72 78 86 94 100 104 116 124 128 129 133 135 139 142 148 155 164 171 179 185 189 24 25 69 71 73 78 83 88 95 100 105 114 122 131 135 143 149 158 162 170 178 189 196 206 220 231 25 26

, 26 27 27 '

) 211 28 14 86 88 93 97 99 100 101 106 121 133 143 144 152 161 163 164 173 179 185 193 200 28 II 29 M M 68 14 81 88 M 100 103 116 122 129 135 141 144 150 156 IM 170 179 187 196 201 212 29 30 77 85 89 92 95 98 100 100 101 10 3 119 133 145 145 154 163 165 165 175 181 187 194 201 213 30 31

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31 32 32 33 M 10 72 76 82 88 92 100 107 123 135 141 144 148 149 158 157 160 166 174 185 193 203 207 33 34 74 78 80 83 87 39 92 100 108 124 142 149 151 152 153 164 166 172 179 183 194 201 208 215 34

) 35 36 37 65 60 69 68 62 71 71 M

71 77 69 77 82 74 88 88 80 97 94 86 98 100 100 100 104 llo 103 124 125 117 137 140 132 141 144 147 137 140 159 142 143 169 142 143 172 146 146 182 148 148 174 145 160 153 164 150 166 161 169 154 172 169 180 165 186 177 191 169 195 201 184 190 198 223 177 179 208 35 202 36 216 37 187 38 38 62 M 67 72 79 39 96 100 103 112 117 123 126 132 133 141 t

1 39 72 74 71 76 80 80 90 100 115 141 143 133 135 142 134 157 146 147 152 179 195 209 234 224 39 40 40 41 41 42 M 67 10 74 81 87 92 100 106  !!8 129 138 138 143 144 150 152 155 163 174 179 185 195 201 42 43 71 M 81 83 87 89 92 100 107 122 135 140 141 144 145 157 159 163 173 173 181 189 195 202 43 44 59 61 63 68 74 81 87 100 109 123 137 144 141 143 I43 148 151 156 165 177 185 191 197 207 44 45 61 65 68 74 85 93 96 100 103 114 130 144 154 163 166 175 169 161 170 178 186 196 210 210 45 46 66 70 87 94 100 104 117 122 123 127 12) 136 140 144 149 160 163 169 173 179 46

% 62 64 78 Ill 205 47

) 47 48 76 95 78 98 74 103 80 99 85 100 86 101 97 99 100 100 108 100 132 109 128 119 129 134 130 '133 132 133 136 140 137 146 146 150 146 154 155 158 187 IM 197 160 205 159 217 174 193 48 49 93 96 98 94 95 97 99 100 102 103 104 106 104 105 113 115 127 128 135 137 138 163 164 188 49 50 57 61 M 70 81 89 95 100 104 107 114 121 124 128 131 136 142 144 150 158 166 177 180 184 50 51 57 to N 68 73 77 86 100 109 119 l13 105 106 113 114 116 Ils 124 129 134 139 165 169 178 51 52 82 83 86 89 93 98 99 100 101 106 115 125 130 132 137 138 139 142 143 149 144 142 149 158 52 53 68 72 74 79 86 91 96 100 103 122 136 151 150 156 159 166 169 179 192 199 209 217 226 238 53

, 54 72 71 71 75 88 93 M 100 103 Il6 131 137 142 152 158 166 168 178 189 i94 203 218 230 243 54 55 67 74 71 75 86 92 97 100 102 122 133 151 151 157 161 169 172 182 1% 203 213 219 228 239 55

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COST TRENDS OF ELECTRIC UTILITY CONSTRUCTION.

E-1 cowitNWD NORTH ATLANTIC REGION 1973:100 (

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i 1982 b

i 1983 b

i 1984 b

i 1985 b

i 1986 b

i 1987 b

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4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 1 Total Plane.A8 Saenn Generation ... 218 222 ~ ' ~'

1 2 Total Plant.All St.& Nwlear Cea. ...

(

217 221 2 3 Total Plant.All St. & Hydro. Gen. . . . . 217 221 3 4 4 5 STEAM PRODUCTION PLANT 5 6 Total Seesm Prodwtion Plant ...... 225 230 6  ;

7 Sarwturn & Improvements.ladr. ... 311 191 199 7 8 Struct. & Improv.. Semi.Ouidr. . . ... 311 192 199 8 9 Boiler Nos Equipment. Coal Frd. ... 312 236 242 9 10 11 Boiler Plant Piping lastalled ....... 219 22M 10 11

(

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12 Turbogenerator Units . . . . . . . . . . . . . 314 228 231 12 13 Accessory Dectrical Equipment .... 315 229 233 13 14 15 Misc. Power Plant Equipment ......

16 NUCLEAR PROD. PLANT

• 316 224 2J2 14 15

(

16 17 Total Nuclear Prod. Plant .. ...... 212 217 17 (

18 Esructures & Improvements ... . . . . . J21 192 avb 18 (

19 Reestor Plant Equipment . . . . . . . . . . 322 214 220 19 20 to

(

21 HYDRO PRODUCTION PLANT 21

, 22 Total Hydraulic Prod. Plant ....... 203 209 22 23 331 191 199 23 24 , $stwtwes Reservoirs&Dems lmprovements

& Waterways . .......

.... 332 192 19n 24 /

25 Waler Wheels. Twbon, & Gens.. . . .

26 .

333 243 244 25 26

(

27 OTHER PRODUCTION PLANT 27 28 Total Other Prod. Plant . . . . . . . . . . .

l l 29 Fwl Holders & Access. . . . . . . . . . . .

30 Gas Twbogenerators . . . . . . . . . . . . .

342 344 222 22s 220 226 223 230 28 29 30

.{

31

.32 TRANSMISSION PLANT 33 Total Transaussion Plana . . . . . . . . . .

34 . Esation Equipmnt . . . . . . . . . . . . . . . 353 217 220 231 232 31 32 33 3d

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35 Te=ers & Fistwas . . . . . . . . . . . . . . . 354 205 211 35 ,4.

36 Poles & Fistwes . . . . . . . . . . . . . . . . .

37 Overhead Conductors & Devices ....

355 356 210 215 222 224 36 37

(

38 Underground Conduit ............ 357 199 204 38 30 Undergreand Conductors & Dev. ... 358 243 247 39 40 40 {

41 DISTRIBUTION PLANT 41 42 Total Distribution Plant . . . . . . . . . . .

43 Station Equipment . . . . . . . . . . . . . . .

44 Poin. Towers & Fistwes . . . . . . . . . .

45 Overhead Conductors & Devices ....

362 364 365 211 214 221 224 217 221 220 225 42 43 44 45

(

46 Underground Conduit ............ 366 lus 190 46 /

47 Under st. Cend. & Dev. In Conde. . . . .

48 Une Transformers . . . . . . . . . . . . . . .

367 368 205 206 201 206 47 48

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49 Pad Mounted Transformers .... . ... 368 1% 159 49 50 Serv ice.-Overhead . . . . . . . . . . . . . . . . 369 195 199 50 (

51 Services Underground . . . . . . . . . . . . 369 179 175 51  %.

52 Meters lastelled . . . . . . . . . . . . . . . . . . 370 176 is6 52 53 Street Ughting. Overhead . . . . . . . . . .

54 Mast Arms & Luminaires Inst. .....

55 Street Ushting Underground .......

54 .

373 373 373 24N 252 256 249 260 256 53 54 55 56

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