ML20245B573

Design & Const Problems at Operating Nuclear Plants

ML20245B573
Person / Time
Issue date:	03/31/1987
From:	Caroline Hsu NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD)
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ML20245B564	List: ML20245B563 ML20245B565 ML20245B573
References
TASK-AE, TASK-E707 AEOD-E707, NUDOCS 8904260163
Download: ML20245B573 (46)

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c AF0D Engineering Evaluation Report

• UNIT:

See Appendix EE REPORT N0.: AEOD/E 707 DOCKET NO.: See Appendix DATE: March 31, 1987 LICENSEE:

Multiple EVALUATOR / CONTACT:

C. Hsu

SUBJECT:

DESIGN AND CONSTRUCTION PROBLEMS AT OPERATING NUCLEAR PLANTS EVENT DATE: See Appendix

SUMMARY

Licensee event report 85-024-02 issued on June 9,1986 for Crystal River 3 l

describes an event involving deficiencies in design and construction that resulted in a potential common mode failure of the nuclear service closed cycle cooling water (SCW) system. The deficiencies were not detected by the plant 0A and QC program for design and construction, but rather detected following an indication of structural damage after the plant was in operation for some time. While the plant was operating at rated power, cracks were discovered on the concrete support pedestal for discharge piping from SCW system heat exchangers 1A and IB. Hairline cracking of the support pedestals for heat exchanger 1C and ID was also found.

Investigation into the cause of the cracking revealed an error in the computer piping analysis. The expansion joints in this system piping had been modeled incorrectly. While the piping reanalysis was being performed, an additional problem was discovered. A rigid seismic restraint used in the computer model for the system piping analysis was not included in construction documentation and, therefore, was never installed.

The missing seismic restraint was the result of an omission during transfer of the computer design output to restraint fabrication documents.

The piping analysis error could have led to a failure which would render both trains of the SCW system inoperable. The missing seismic restraint could have also led to a similar failure during a seismic occurrence. These were significant deficiencies in the design and construction of the plant which could pose a potential hazard to safe plant operation had these deficiencies remained uncorrected.

The event prompted a search of similar identified design and construction problems at other operating plants over the time period from January 1984 to September 1986. There were a total of 55 reports of deficiency that involved 34 plants. Of these 34 plants, 21 started comerical operation in 1970 or earlier, and the remaining 13 plants had less than two years of operation.

• This document supports ongoing AE0D and NRC activities and does not represent the position or requirements of the responsible NRC program office.

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. 1 The design and construction deficiencies described in these reports can be grouped into six categories:

(1) piping stress exceeding code limits (13 reports); (2) incorrect hardware or improper installation of hardware (16 reports); (3) lack of fire seals for electrical cable penetrations (5 reports);

(4) electrical wiring error (6 reports); (5) errors in electrical, instruments-i tion and control circuits (10 reports); and (6) electrical and control panels

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not seismically supported (5 reports). Of particular note are four reports in which the deficiencies could cause the affected systems to fail to meet the single failure criterion. At three units, the deficiency could lead to the loss of redundant cooling systems and at the fourth unit it could prohibit the reactor protection system scram capability by normal means.

Based on the review, the designed construction deficiencies identified in these events could be attributed to inadequate document controls, unreviewed safety and design condition or inadequate review during modification.

Inadequate document controls appear to be the problem associated with design change control, including undocumented design change and omission of design or modification items in construction drawings. This could lead to questionable field installation. Unreviewed safety and design condition could cause changes to design in conflict with the technical specification, plant FSAR and the existing procedural requirements.

Inadequate review during modification often provide incorrect data for a modification and would result in errors in implementation.

In some cases this also caused incomplete post modification testing.

Accordingly, it appears that the 55 reports represent a potential generic problem in that the design and construction deficiencies were not detected in the plant QA and QC verifications for compliance with the plant requirements.

This suggests that the plant QA and QC programs may not have been adequate to identify all existing design and construction problems, with the likelihood that there may be other undetected, significant design and construction deficiencies in operating plants.

In view of this safety concern, the following actions are suggested:

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IE consider issuing an information notice to address the findings of this st!!dy for feedback purpose.

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NRR and the regions should review the adequacy of current QA & QC programs used in plant modifications to verify that the plant is in conformance with design and construction requirements. Also, the findings of this study should be used as a reference for future inspection and review of changes in plant design and construction.

INTRODUCTION Licensee event report (LER) 85-024-02 issued on June 9,1986 from Crystal River 3 describes an event involving deficiencies in design and construction that resulted in a potential common mode failure of a system important to safety. The deficiencies were not detected in the plant quality assurance (QA) and cuality control (OC) progran for design and construction but rather detected following an indication of structural damaae after the plant was in operation for approximately nine years.

The review of this event led to initiation of a

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. search and evaluation of design and construction problems at other operating plants.

(See Appendix for description.) The intent of this evaluation was to determine the prevalence of such problems and to identify the generic implica-tions from these plant events.

The report from Crystal River 3 involves a major deficiency in design and construction that resulted in a potential common mode failure of the nuclear service closed cycle cooling water (SCW) system. The plant was operating at 96 percent reactor power when a cracked concrete support pedestal for discharge piping from SCW system heat exchanoers IA and IB was discovered.

Hairline cracking of other support pedestals was also found.

Investigation into the cause of the cracking revealed an error in the computer piping analysis; expansion joints in this system piping had been modeled incorrectly. While the piping reanalysis was being performed, an additional problem was discovered. A rigid seismic restraint used in the computer model for the system piping analysis was not included in construction documentation and, therefore, was never installed.

The missing seismic restraint was the result of an omission during transfer of the computer design output to restraint fabrication documents. These were significant deficiencies in design and construction which could pose a potential hazard to the safety of the plant during operation had these deficiencies remained uncorrected.

Computer piping analyses were rerun and the results indicated that neither the damaged pedestals nor the original (undamaged) pedestals could withstand the design loads. The piping analysis error could have led to a failure which would render both trains of the SCW system inoperable. The missing seismic restraint, combined with a seismic event, could have also led to a similar failure. The loss of both trains of this system would be a condition outside the design basis of the plant. The SCW system provides cooling for essential components or systems upon receipt of an engineered safeguards signal during both normal and shutdown operations.

Inoperability of the system could prevent safe plant operation.

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The search of design and construction problems was extended to a review of LERs reported from January 1984 to September 1986 in the Sequence Coding and Search System (SCSS) LER data base file. This review identified a total of 55 reports involving design and construction problems that have also posed a potential safety problem to plant operations. These problems identified were not detected by the plant QC inspection or QA programs during plant construc-tion or modification phase. Nor could most of the problems be detected by the plant testing programs such as preoperational, startup, or surveillar,ce tests, since the problems are latent deficiencies which could not be identified by testing without the presence of certain postulated conditions; for example, a seismic, flood or LOCA event. These 55 reports were from 34 plants; of these 55 reports, 36 were of original design and construction problems and the other 19 were results of modifications. A review of these reports indicates that these design and construction deficiencies can be grouped into six categories, f

These categories are:

(1) piping stress exceeding code limits; (2) incorrect i

hardware or improper installation of hardware; (3) lack of fire seals for

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electrical cable penetrations; (4) electrical wiring error; (5) errors in i

electrical, instrumentation and control circuits; and (6) electrical and f

control panels not seismically supported. Based on these categnries, a brief j

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I description of the 55 reports involving design and construction problems are listed in Tables 1 through 6 in the Appendix.

Each of these categories along with their causes, safety implications, corrective actions taken, and the need for follow-up actions are discussed in the followinn sections.

l DISCUSSION (1)PipingStressExceedingCodeLimits A significant number of design and construction deficiencies identified in this study were associated with safety-related system pipings in which the existing piping stress was found to exceed the design limits specified in plant technical specifications. These degraded conditions could fail the piping and result in inoperability of the associated sysM m during a seismic or LOCA ~ event. The piping deficiencies identified were not detected by the plant QC inspections or QA programs during the design and construction phase.

In each case the licensee discovered the stress deviations during a special review or inspection initiated due to reasons other than the implementation of a QC or QA program. A total of 13 reports involving design and construction deficiencies attributed to excessive piping stress are listed in Table 1.

The degraded conditions in seven of these reports had existed since original plant startup and the others resulted from system modification work performed some time after initial plant operation. These reports were from Oyster Creek 1, Dresden 2, Quad Cities 1, Palisades Browns Ferry 1, Cooper, Crystal River 3, Fermi 2, Davis-Besse 1 and LaSalle 1 & 2.

A review of these reports indicate that the piping deficiencies were attributed to one of the following factors:

Lack of design documents controls Unreviewed design condition i

Excessive gap in pipe support Inadequate design revies during modification The reports involving design and construction piping deficiencies which resulted from lack of design document control were identified from Dresden 2, Quad Cities 1. Crystal River 3 and Fermi 2.

The primary inadequacies in design document control are (1) undocumented design change, and (2) omission of design or modification items in construction drawings. The piping deficiency at Dresden 2 was a result of system modification which was performed in 1984. The modification was based on the original "as-built" drawings of 1978 instead of the updated drawings of 1982.

One of the two reports involving piping deficiency from Quad Cities I was also attributed to modification similar to the one fron Dresden 2.

The other report involved nine supports of the RHRSW system piping exceeding code stress allowables during SSE condition. The deficiency was caused by

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8 4 loading changes as a result of the previous addition of a cross-tie pipe.

The affected supports were to be modified when the cross-tie pipe was added, but remained unchanged because the cons'ruction drawings for the modification specified the new cross-tie pipe without indication of change in the supports.

For the report from Crystal River, a rigid seismic restraint in the nuclear service water system used in the computer model was not included in construction document and, therefore, was never i

installed.

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The report from Fermi 2 involved design calculations for safety-related pipings an( pipe hangers.

In this report, some of the design calculations had not been updated to reflect change documents that were issued after the as-built calculations were completed. As a result, five instances were identified during the reviews where deficiencies existed and physical changes in the plant were required.

Inadequate design document control could provide incorrect data for a

- modification and would result in errors. To prevent future occurrence of this type of problem during modification work, the licensee at Dresden 2 l

instituted changes to the architect-engineer guidebook to ensure review of existing design documents are performed when working on a pipe or pipe support and to provide assurance that the latest analysis and drawings are l

used.- Additionally, a system walkdown to confirm that the drawings represent the actual condition of the equipment is required.

Similarly, the licensee of Fermi 2 revised the plant engineering procedures to ensure stringent controls on verification checklist that design calculations used as design input basis are updated, revised and reconciled as required.

The three reports that were related to unreviewed design conditions were identified from Oyster Creek 1, Browns Ferry 1 and Cooper. The condition of a LOCA was not considered in the design of supports of a non-seismic connection to the ESW system piping at Oyster Creek 1.

This deficiency could prevent the ESW system from providing required flow to the contain-ment spray system in the event of a LOCA. Certain secondary containment piping penetrations at Browns Ferry I were not seismically qualified even though the requirements of seismic qualification for such penetrations are specified in the plant FSAR. At Cooper, the need for seismic restraints were not considered in the design of the expansion sleeve, crossover line and drainline for the gas treatment system.

The report from #itad Cities I was related to inadequate design review of a modification of certain pipinas attached to the torus.

In this modification, the engineer only qualified the added and the modified supports for the pipings. Supports for existing pipings, with stress found within the code limits, were not verified to be qualified and were subsequently found to exceed code stress allowables, and the affected systems rendered inoperable.

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. (2) Incorrect Hardware or Improper Installation of Hardware I

A total of 16 reports are identified in Table 2 in which design and con-l struction deficiencies were found to be associated with incorrect hardware l

or improper installation of hardware. These 16 reports were identified at 10 plants: Oyster Creek I, Browns Ferry 1, Pilgrim 1, Hatch 1 and 2, Millstone 2, Arkansas 2, LaSalle I and 2, Catawba 1, Diablo Canyon 1 and 2, and Crystal River 3.

The discrepancies in eight of these reports had existed since initial plant startup and the others resulted from inodifica-tion work performed some time after the plants were in commercial operation.

Discrepancies in three of these reports were discovered because of malfunc-tions of the affected equipments during surveillance testings. The other discrepancies were discovered when the plants were performing reviews, inspections or analyses for some special designated programs other than the plant QA or QC programs for design and construction (including the preoperational, startup, and post-modification tests).

Those discrepancies identified during reviews, inspections or analyses were subsequently determined to have caused degraded conditions which could prevent the affected systems from performing their proper safety function in certain postulated or operating conditions.

Reviews of these events indicate that the major causes of incorrect hardware or improper installation of hardware are inadequate quality control, inadequate document control, and human errors during hardware installation.

Five reports involving improper installation of hardware which resulted from inadequate installation quality controls were identified from Browns Ferry 1, Pilgrim 1, Hatch 1 and 2, and Arkansas 2.

In each report the deficiency was associated with modification work.

Inadequate administrative controls and review of design changes during the modifications appeared to be contributing factors for these deficiencies.

For the report from Browns Ferry 1, some of the battery racks for the diesel generators were not installed per design drawings, and the racks were declared seismically unqualified.

The improper installation deficiencies were due to the failure of the responsible engineers to take the appropriate steps to ensure compliance with design requirements during the modification.

A testing instrumentation on the HPCI turbine at Pilgrim I was improperly installed due to the licensee's inadequate control of the contractor personnel working on the installation. The missing outer clamp on a seismic restraint for the RHR service water pump at Hatch I was a result of inadequate installation and followup verification during a previous removal and reinstallation operation. At Arkansas 2, two fire dampers failed to close because of the intrusion of a sheet metal screw through the duct work blocking the motion of the dampers. The cause of the intrus1on of the screw was inadequate installation and modification

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, quality control due to the fact that there were no requirements to test these dampers for operability, and no functional test on these components were performed.

In the report from Hatch 2, the material for the bolts that secure the RHR service water pump column was over stressed due to a design modification for the support bracket of the pump column in 1982. When the modification was made, the need to change the bolting material was not recognized during

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the review of the seismic analysis.

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The design and construction deficiencies in these five reports were attri-buted to inadequate installation quality control during the modifications.

To improve some aspects of installation quality controls, the licensee at Pilgrim initiated a procedure that requires future contract work at the site to be administered by a designated indtvidual with plant-specific knowledge of administrative controls approval requirements.

Five of the 16 deficient reports identified were related to inadequate document controls. These reports were from Oyster Creek 1, Browns Ferry 1, LaSalle 1 and 2, and Catawba 1.

The primary inadequacies in document i

controls are (1) undocumented design or modification changes, and (2) plant FSAR requirements not incorporated in design drawings.

The human errors during hardware installations are (1) sloppy workmanship, (2) inadeouate followup verification, and (3) misunderstanding of design details. The reports involving improper installatfor of hardware which resulted from human error were from Crystal River 3. uyster Creek I and Diablo Canyon I and 2.

The report of Crystal River 3 was related to sloppy workmanship. Some of the bolts in a seismic anchor for the ductwork of the control complex HVAC system were either not long enough to properly secure the support to the wall, or the bolts were cut off and the heads were tack welded to the support frame. This gave the appearance of properly installed bolts when they were not. Approximately 90 of the 160 supports had such a bolting deficiency. This was an original installation problem. The ventilation contractor was responsible for his own quality control inspection.

No plant QC organization was involved in the in-stallation.

For the report from Oyster Creek I, the deficiency was attributed to inadequate followup verification.

Three hydraulic snubbers were found in-operable because the associated pipe clamp to shear lug " welds" were not put back after the clamps were removed to facilitate pipe inspection.

At Diablo Canyon 1 and 2, because of personnel misunderstanding of the design details, nuts on embedded plates of a main feedwater piping snubber support for Unit 2 steam generator 2-2 had never been installed.

In-spection of the corresponding equipment on Unit I disclosed a similar deficiency.

Both snubbers were determined to have been inoperable during previous plant operation.

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, (3) Lack of Fire Seals for Electrical Cable Penetrations Under the requirements of Appendix R to CFR Part 50, an adequate fire barrier should be provided for electrical cable penetrations to assure safe shutdown capability. Although the fire barrier requirement was also specified in the plant FSARs, five plants discovered that fire seals were missing in electrical cable penetrations. These cables supply power to i

safety-related systems and equipment. Table 3 describes these reports l

involving missing fire seals in cable penetrations.

The five plants are Limerick 1, WPPSS 2, Grand Gulf 1, River Bend 1 and Callaway.

For four of these plants the fire seals, which were found missino, had never been installed and were discovered missing during a special review program conducted sometime after the plants were in operation. The missing fire seals at River Bend 1 were identified during a walkdown inspection of the plant QC program. The cause of the missing seals at timerick 1, Grand Gulf I and Callaway appeared to be a deficiency in the construction program in that it failed to properly identify tiie conditions requiring the need of internal seals of cable penetrations. The failure to install the internal fire protection seals in the electrical cable penetrations found at the other two plants was due to unreviewed conditions in the original design for electrical cable fire protection seals.

Internal fire protection seals j

for electrical gutters at Limerick I were found missing when the plant was operating in Mode 2.

These cutters penetrate the fire barrier enclosure between the 4 KV switchgear room and the rear access corridor.

Three of the fire protection seals are subject to technical specification require-ment. The affected cables at Grand Gulf I were associated with redundant trains of systems required for the safe shutdown of the plant. This condition is not in compliance with Appendix R requirements of 10 CFR 50 and was determined outside the design basis of the plant. Systems and equipment affected were: the standby service water (SSW) system, the SSW reactor shutdown flow recorder, an RHR iockey pump, the emergency diesel generator fuel oil transfer pump, the reactor core isolation cooling system, the SSW pump house ventilation system, the safeguard switchgear and battery room ventilation system, the ECCS pump room ventilation system, and the power feeders and control cables to the ESF load centers. The report from Callaway 1 indicated that 23 fire barrier penetrations did not have the internal conduit seals installed. The missing seals were discovered during an inspection initiated as a result of a potential penetration design deficiency identified at the licensee's other nuclear plant.

Omission of the conduit seals in some configurations constituted violations of the limiting condition for operation as provided in plant technical specifications.

The unsealed cabled penetrations found in WPPSS 2 and River Bend I were due to the failure to consider conditions requiring fire seals in the l

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.3 9-original design. Thus, fire seals required for these penetrations were not provided in design documents. Penetration fire protection seals were not installed for the biological shield wall at WPPSS 2.

It was determined that the original design did not consider radiant heat from adjacent fire zones to possibly ignite the urethane foam separating the biological shield wall and primary containment vessel. The resultant fire could incapacitate.the electrical penetrations required for the safe shutdown of the plant. With the unit at full power operation, approximately 60 unsealed fire barrier penetrations in the fire barrier. wall were found at River Bend 1 during a walkdown in response to a licensee's finding in a quality assurance report. The missing fire seals constituted violations of plant technical specifications which requires penetration assemblies in fire barriers to be operable at all times.

(4) Electrical Wiring Error Table 4 lists six reports describing errors in electrical wiring. The wiring errors in four of the six reports had existed since original in-stallation and those in the other two reports resulted from system modifications. These six reports were from Nine Mile Point 1, Browns Ferry 1, Peach Bottom 2 and 3 and Limerick 1.

The electrical wiring errors at Nine Mile Point I and Peach Bottom 2 constituted a violation of cable separation criteria specified in the plant FSAR under the requirements of Appendix R of 10 CFR 50. The cables involved in the routing discrepancy at Nine Mile Point I were the control cables for containment spray raw water pump, cables associated with manual controls of all four core spray pumps, and cables feeding buses from offsite power. The cable discrepancies found at Peach Bottom 2 involved redundant power cables for two isolation valves lined up for one contain-ment penetration. These power cables shared a common cable tray which was in violation of the cshle separation criteria for the plant. The licensee of Peach Bottom determined that the cable discrepancy was the failure of the. responsible engineer to properly evaluate the isolation requirements of these valves during a modification. The licensee subsequently issued instructions which require responsible engineers to consult with the appropriate engineering group for additional expertise on any modification.

Two reports involving wiring errors were identified at Browns Ferry 1.

In one of these reports, during a schematic review of the reactor protection system (RPS), nine wires were found not routed in conduit internal to RPS panels as required.

This error could have led to a single failure pre-venting the RPS scram capability.

(In the LER, the licensee has stated that this condition could allow an internal panel fire to short the scram discharge volume drain valves and the CRD scram valves to 120V power cables, thus preventing a reactor scram.) The causes of errors were due to unclear notation and/or misinterpretation of notes on panel drawings.

. In the other report, the cables for the normal and the alternate control powe'r for 4Kv shutdown board "A" were found reversed. This error was attributed to a failure to properly terminate the wiring when the shutdown board battery was originally installed in 1973.

Inadequate labeling of the transfer switch also contributed to the error, j

i The report of Peach Bottom 3 describes an event which occurred when the I

unit operated at full power. A wiring error caused a blown fuse in an f

electrical circuit which resulted in isolation valves in the reactor water

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cleanup system automatically closing. The cause of the wiring error was 1

unknown. However, the failure to identify this error during previous installation inspections also contributed to this event.

The incorrect wiring at Limerick I caused the supply and return valves in the drywell radiation monitoring system to fail to close from a manual isolation signal during a unit startup. The wiring had been changed during a modification.

Prior to completion of the modification, a field engineer working on verification of the modification changed the wiring in accordance with the unrevised circuit print. The discrepancy was caused by inadequate document control.

The wiring discrepancies in these identified reports were not detected by the plant OC inspections during installation nor by post modification testing and had remained undetected.

Discrepancies in four of these reports occurred in original installation and those in the other two events were a result of modifications. These wiring errors were attributed to inadequate document controls, unreviewed safety conditions, unclear notation on installation drawings, or inadequate post modification testinc.

(5) Errors in Electrical, Instrumentation and Control Circuits There are ten reports identified in Table 5 where a design / construction deficiency has been found involving errors in electrical, instrumentation and control circuits for safety-related systems. These ten reports were from Oyster Creek 1, Browns Ferry 1 & 2, Salem 1, Crystal River 3, Hatch 1, Brunswick 1 Perry 1 and Palo Verde 1.

Except for two reports from Perry 1, the circuit errors were discovered when the plants were performing reviews or inspections initiated for some specially designated program some time after the plants were in operation. At Perry 1, the errors in the circuits were identified because of malfunctions of the affected eouipment.

None of these discrepancies were identified during the plant construction phase verification program. The errors identified during reviews or inspections were subsequently determined to have presented scenarios which could cause the affected circuits to misinterpret status of the associated systems, and could result in the system not performing according to the intent of its original design.

Six of the ten reports identified having errors in the circuits were attributed to certein cperating safety conditions not considered in the

. design of the circuits.

These six reports were from Oyster Creek I, Browns Ferry 1, Salem 1, Crystal River 3, Hatch 1 and Perry 1.

In almost every case, the deficiency in the circuits resulted from the designer not being fully aware of the design and operating parameters under which the associated systems were required to operate; even though, in some cases, the missing design conditions were the design criteria established in the plant FSAR or the bases of the technical specifications. The errors in the circuits identified in the reports from Oyster Creek 1, Browns Ferry 1 and Salem 1 were determined to result in the associated safety-related systems not performinp according to the original design intent of being single failure proof under certain postulated conditions.

The affected systems were the core spray system at Oyster Creek 1, several electrical board room cooling systems at Browns Ferry 1, and the euxiliary feedwater system at Salem 1.

The design deficiency of the emergency feedwater initiation and control circuit at Crystal River 3 would prevent the system from electrically switching into

" Maintenance Bypass." This could cause a spurious actuation of the system while performing troubleshooting. At Hatch, an error in the control circuit for the standby gas treatment system supply dampers was caused by improper setup during the initial installation.

The error would cause the dampers to fail-close on a loss of electrical power. This conflicts with the FSAR requirements.

Inadequate design consideration caused design error in the containment vessel and drywell purge system flow control logic at Perry 1.

As a result, a vacuum was created in containment, causing the opening of the containment vacuum relief valves.

To prevent future recurrence, preventivo measures during the design process are necessary to minimize the possibility of the deeigners not being aware of related postulated or operating conditions.

io achieve this, the licensee at Salem made an improvement in the design change program such that all modification packages are to be reviewed by a safety analysis group and to further enhance the design review process, a specific review is to be performed for safety grade / control grade inter-faces. The licensee of Perry plant also revised its current procedures governing the evaluation of design discrepancies to clarify the need for followup action to eliminate malfunctions resulting from errors in control logic.

i Four other reports involving control logic errors were attributed to inadequate design document control.

These four reports were from Browns Ferry 1, Brunswick ', Perry 1 and Palo Verde 1.

The error in the primary j

containment isolation system at Browns Ferry 1 was attributed to a design oversight which permitted issuance of a drawing error during the plant construction phase.

This error resulted in a combination of switch manipulations that could lead to an unanalyzed condition. The resulting condition requires manual operator action to prevent release of radioactive l

4 gases to the environment. At Brunswick 1, isolation logic was not in-stalled in the control building emergency ventilation system because the logic circuit was not shown in the system circuit drawings. This deficiency would result in failure to isolate the system from a high chlorine signal as required in the plant FSAR. The deficien.:1es identi-fied in the reports from Perry 1 and Palo Verde 1 resulted from modi fea-tions. The report from Perry 1 indicated that the RHR isolation va1<es did not receive a reactor vessel high pressure isolation signal as stated in the FSAR. This failure was due to inadequate documentation.

The isolation signal had been changed to a high drywell pressure isolation signal but the architect-engineer failed to update the FSAR and technical specifications to reflect the change.

Similarly, the cross connected pressure transmitters for the main steam atmosphere dump valves at Palo Verde 1 were also related to inadequate documentation. The cross connec-tion was a result of personnel mistakenly verifying the as-built plant conditions to drawings that did not reflect a modification that had been made. To prevent a recurrence of similar problems, the licensee of Perry initiated an engineering review to ensure consistency between actual plant condition, installation diagrams, the FSAR and technical specifications.

(6) Electrical and Control Panels not Seismically Supported A total of five reports involving electrical and control panels not seismically supported are identified. Table 6 lists the five reports which were from four operating stations, these being Dresden 2 & 3, Browns Ferry 1, 2, & 3, Cooper and Davis-Besse 1.

The deficiencies identified in those reports were design and construction related errors that had existed since initial construction. The electrical and control panels involved were the control room panels at Dresden 2 & 3, the startup test panels and local reactor protection system panels at Browns Ferry 1, 2, & 3, the diesel generator output switchgear cabinet at Cooper and cabinet doors on Class IE equipment for essential instrument 120 VAC power at Davis-Besse 1.

The affected control panels are Class 1E components and should be seismically qualified to assure functionability and structural integrity as required by " General Design Criteria 2" of Appendix A to 10 0FR Part 50. With the exception of the startup test panels at Browns Ferry, the original designs correctly specified the seismic anchorages or other seismic supports. However, the panels were not seismically supported and the deficiencies had not been detected since construction inspite of QA inspections.

The design and construction deficiencies described in these reports were identified in an evaluation initiated for some other specific program or incidentally discovered in design changes or inspections fer unrelated problems, some time after the plants were in commercial operation.

The deficiency at Dresden 2 & 3 was revealed during a seismic evaluation of

e 4 control room panels _for mounting of future instrumentation. The panels had not been bolted to a support beam as indicated in the original in-stallation instruction, but rather rested on top of the beam, since initial construction. An unusual event was declared and the plants were shutdown when the deficiency was revealed. Similarly, the control panels at Browns ferry 1, 2, & 3 (LER 85-020) and Cooper were identified as not being installed as required by the original construction drawings during initial installation and the discrepancies had remained undetected.

The deficiency at Browns Ferry was identified in a design analysis and that at Cooper in preparation for a future design change.

Seismic qualification.was not considered in the design of the startup test panels of Browns Ferry 1, 2, & 3 (LER 85-013).

Identification of this deficiency was a result of the TVA Regulatory Performance Improvement Plan. The cause of the cabinet doors at Davis-Besse 1 not being bolted closed as required by the seismic design was that the bolting detail was missed during the development of the station procedures.

In each case the de:#fciencies would reduce or prevent the panel's ability to remain func?ional in a seismic event. Thus, safe operation of the plant could.Save been jeopardized if controls in the panels failed.

FINDINGS AND ONCLUSIONS Based on the preceding discussion and related follow-up activities conducted for this study, the following findings and conclusions are provided.

1.

A total of 55 reports involving design / construction deficiencies associated with safety-related systems have been identified from 34 operating plants.

Among the 55 reports, 36 were of original design / construction problems and the other 19 were results of modifications. These design / construction deficiencies were not detected by the plant QA and QC programs (including the preoperational, startup, and post-modification tests) during plant construc-tion or modification phase and had existed since original plant startup or modification.

These were significant deficiencies in design and construc-tion which could have posed a potential hazard to the safety of plant operation had these remained uncorrected.

2.

These design and construction deficiencies were discovered under one of the following conditions:

Review, inspection or analysis initiated for some special designated programs, such as Systematic Evaluation Program, review of IE information notice, or other regulatory-related programs.

Review as a result of similar problems that have occurred on other systems or at other plants.

Malfunction or loss of operation of components or systems.

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. Evidence of structural degradation or damage.

Incidental recovery during review or inspection of unrelated items or problems.

3.

The deficiencies in these events identified could be grouped into six categories; they are:

a.

piping stress exceeding code limits b.

incorrect hardware or improper installation of hardware c.

lack of fire seals for electrical cable penetrations d.

electrical wiring error e.

errors in electrical, instrumentation and control circuits f.

electrical and control panels not seismically supported 4.

-These deficiencies were attributed to inadequate document controls, unreviewed safety and design conditions or inadequate review during modification.

Inadequate document. controls appeared to be the problem associated with design change control, including undocumented design change and omission of design or modification items in construction drawings.

j This could lead to questionable field installation. Unreviewed safety and

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design condition caused changes to design in conflict with the technical i

specification, plant FSAR and the existing procedural requirements.

l Inadequate review during modification often provided incorrect data and, in some cases, incomplete post-modification testing for modification and would result in errors in implementation.

5.

Most of the design and construction problems _ identified in the 55 reports i

are of latent deficiencies which'would not be detectable by the plant testing programs, such as preoperational, startup, post-modification, or surveillance testing, since certain postulated design conditions (e.g.,

seismic, flood, fire or LOCA) could not be ' simulated during the testing.

However, in the case of certain deficiencies involving electrical wiring

. error and control circuit error, an adequate post-modification testing 3

program could identify the deficiency and corrective action can be taken.

I

(

6.

For the deficiencies resulting from modifications, many plants appear to

)

take a positive attitude and attempt to identify the cause and implement corrective action. However, son of the plants considered the identified deficiencies to have been an isolated case on the basis that their current e dification procedures are more comprehensive than the ones used i

in the previous modifications.

7.

Most of the deficiencies in the reports identified were discovered when the i

plants were performing review or inspections initiated for some special dc*.ianated programs. Since these programs did not cover all operating F 'e s, th,mumber of design and construction events reported may not represent tne true extent of design and construction problems that existed in operating plants, and there are likely to be other undetected, signifi-cant design / construction deficiencies in operating plants.

1

SUGGESTIONS 1.

IE slic91d issue an IE information notice to inform licensees of the potential sefety problem concerning the undetected design / construction deficiency. This information notice should also address the six design / construction deficiencies and the causes identified in this study, which have occurred at operating plants in the past two years.

2., In view of the number of design / construction problems detected after plants were already in operation for some time and since the identifications were not the result of QA or QC programs for the plant construction phase, NRC staff should review the adequacy of cut' rent QA & QC programs used in plant modification to verify that the plant is in conformance with design and construction requirements.

This report could be used for guidance in future plant inspection and review of changes in plant design and construc-tion.

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