ML20212M485
| ML20212M485 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 12/13/1985 |
| From: | Architzel R, Milhoan J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | |
| Shared Package | |
| ML20212M484 | List: |
| References | |
| 50-285-85-22, NUDOCS 8608260264 | |
| Download: ML20212M485 (100) | |
See also: IR 05000285/1985022
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U.S. NUCLEAR REGULATORY COMMISSION
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0FFICE OF INSPECTION AND ENFORCEMENT
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Division of Quality Assurance, Vendor,
and Technical Training Center Programs
Report No.:
50-285/85-22
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Docket No.:
50-285
Licensee:
Omaha Public Power District
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1623 Harney Street
Omaha, Nebraska 68102
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Facility Name:
Fort Calhoun Station
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Inspection At:
Omaha Public Power District Engineering Offices,
Omaha, Nebraska
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Fort Calhoun Station, Blair, Nebraska
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Inspection Conducted:
September 16-20, 30, and October 1-8, 1985
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Inspection Team Members:
Team Leader:
R. E. Architzel, Senior Inspection Spec,ialist, IE
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Mechanical Systems:
G. J. Overbeck, Consultant, Westec Services
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Mechanical Components:
A. V. duBouchet, Consulting Engineer
Electrical Power:
G. W. Morris, Consultant, Westec Services
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Instrumentation &
Control:
L. Stanley, Consultant, Zytor Inc.
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Design Control:
A. Saunders, Reactor Engineer, IE*
M. Murphy, Reactor Inspector, Region IV*
R. Lloyd, Reactor Engineer, IE*
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Ralph E. Architzel
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Team Leader
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Approved by:
w [.
1[ 3 ~ ~ /.1/#/84~
ames L. M11hoan
Date
Lection Chief
uality Assurance Branch
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- Part time
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B608260264 860821
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ADOCK 05000285
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LIST OF ABBREVIATIONS
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American Concrete Institute
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American Institute of Steel Construction
American Society of Mechanical Engineers
ANSI
Amerian National Standards Institute
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American Society for Testing and Materials _
CFR
Code of Federal Regulations
Engineered Safety Features
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Final Safety Analysis Report
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Generation Station Engineering
Heating, Ventilation and Air Conditioning
IEEE
Institute of Electrical and Electronics
Engineers
Loss of Coolant Accident
Modification Request
NRC
U.S. Nuclear Regulatory Commission
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Nuclear Steam Supply System
Omaha Public Power District
OSAR
Operations Support Analysis Report
Piping and Instrumentation Diagram
Updated Safety Analysis Report
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1.
INTRODUCTION AND SUM 4ARY
1.1 INTRODUCTION
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.The following subparagraphs provide introduction to the ebjectives, format and
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focus of the Fort Calhoun Safety Systems Outage Modification Inspection.
1.1.1 OBJECTIVES
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This inspection was part of a trial NRC program being implemented to examine
the adequacy of licensee management and control of modifications performed
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during major plant outages.
The purpose of this portion'of the Safety Sys' tem ~
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. . Outage Modification Inspection Program was to examine, 'on' a sampling basis, the'
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detailed design and engineering which was required to support the outage.'
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1.1.2 REPORT FORMAT AND DEFINITIONS
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The areas examined during this inspection are addressed by discipline in the
following chapters.
Deficiencies, unresolved items, and observations are
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defined below and are included in an appendix to this report.
(1) Deficiencies
Errors, inconsistencies or procedure violations with regard to a
specific licensing commitment, specification, procedure, code or
regulation are described as deficiencies.. Follow-up action is required
for licensee resolution.
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(2) Unresolved Items
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Unresolved items are potential deficiencies which require more
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information to reach a conclusion.
Follow-up< action is required
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for licensee resolution.
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(3) Observations
Observations represent cases where it is considered appropriate to
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call attention to matters that are not deficiencies or unresolved
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items.
They include items recommended for licensee consideration but -
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No' licensee
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response is required.
1.1. 3 FT. CALHOUN PROJECT ORGANIZATION
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The Omaha Public Power District is the licensee for the Ft. Calhoun Nuclear
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Power Plant.
As such, Omaha Public Power District is responsible for the
design, construction and operation of the facility.
The utility holds
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responsibility for the overall plant design, with contract design support from
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Stone and Webster.
The original architect / engineer (Gibbs and Hill, Durham and
Richardson) is no longer under contract to the licensee. Other firms are
occasionally engaged for services.
Combustion Engineering designs and provides
the nuclear steam supply system.
The nuclear steam supply contract is managed
directly by Omaha Public Power District.
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1.1.4 INSPECTION EFFORT
The inspection was an interoffice NRC effort conducted with contractor
assistance. Team members were selected to provide technical expertise and
design experience in the discipli.nes listed.
Most of the team members had
previous experience as employees of architect-engineering firms or reactor
manufacturers working on large commercial nuclear power plants.
The others
had related design experience on commercial nuclear facilities, test reactors,
or naval reactors.
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Beginning on August 12, 1985, a portion of the inspection team devoted one week
to the initial study of background information and preparation of plans for the
inspection.
The week of August 19, 1985 was spent at the site and at the
licensee's office to become familiar with the respective organizations and
interfaces and to gather additional background material.
The majority of the
team inspection activities occurred at the Generating Station Engineering Offices,
Omaha, Nebraska, the weeks of September 16 and September 30, 1985. The
inspection activities concluded on October 8, 1985 with an Interin Status
Briefing.
The inspection team reviewed the organizations' staffing and procedures and
interviewed personnel to determine the responsibilities of and the
relationships among the entities involved in the design process.
Primary
emphasis was placed upon reviewing the adequacy of design details (or
products) as a means of measuring how well the design process had functioned
in the selected sampling area.
In reviewing the design details, the team
-focused on the following items:
(1) Validity of design inputs and assumptions
(2) Validity of design specifications
(3) Validity of analyses
(4) Identification of system interface requirements
(5) Potential indirect effects of changes
(6) Proper component classification
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(7) Revision control
(8) Application of design information transferred between organizations
(9) Design. verification methods
The team inspected four engineering disciplines within the project.
The four
disciplines were mechanical systems (Section 2), mechanical components
(Section 3), instrumentation and controls (Section 4), and electric power
(Section 5).
1.2 SUMARY - MECHANICAL SYSTEMS
The team reviewed two modification packages in detail and two additional
modification packages in part.
The packages reviewed included modifications
planned for the current outage and completed modifications.
The team identified a number of examples of deficient design activity.
Based
on the modification packages reviewed, these items can be categorized as
follows:
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(1)
In ccrtain instances, engineering input into post-modification
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fulfill the functional design requirements of the system or component
testing procedures was inadequate to confirm that physical modifications
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(Deficiency D2.1-7 and D2.2-3).
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(2) Seismic requirements were not properly addressed in modification
packages (Deficiency D2.2-3 and Observation 02.1-5). '
(3) In certain instances, errors found in the design details indicate that
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design verifications were not performed in sufficient detail to
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substantiate the design.
In addition, the design verification process
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was normally performed prior to site acceptance (following construction
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and testing of the modification) rather than as the design output
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documents were completed.
This practice, can needles' sly ~ place the '
. independent reviewer in a time critical role for plant restart.
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(Deficiencies D2.1-6, D2.1-7, D2.2-1, D2.2-2, D2.2-3, an'd D2.2-6).
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(4) Incorrect or inappropriate documents were being used as sourc'es of
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design, input with insufficient consideration given to the original
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plant design basis (Deficiencies D2.2-1 and D2.2-6).
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The team noted that Omaha Public Power District engineers expressed different
opinions as to the appropriate source documents for design input.
Generating
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Station Engineering personnel indicated that controlled copies of system
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descriptions are sources of design input and design criteria, but the
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operations staff associated with the upkeep of the system descriptions
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informed the team that the system descriptions are for operational use and
that the USAR is the source of design input and criteria.
The team identified
a weakness in the design control process in that original design bases and
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calculations have not been maintained in a' workable form and subsequent
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modifications have not been maintained as auditable design input documents.
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This condition is aggravated because design calculations are not maintained as
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living documents.
Instead design calculations are prepared and filed with the
modification package at the discretion of the design engineer.
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consequence, existing design calculations covering certain design attributes
are not readily retrievable or may not exist.
This may result in a lack of
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design traceability from design input through to design output and an
inability t'o determine the design bases of systems and components.
The team found that engineering judgments were frequently not documented when
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. . used as the basis for not performing a calculation. The lack'of this docu-
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mentation, including appropriate justification, results~'in the lack of~a ba~s'is
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input to
use of endesign output.
Omaha Public Power District procedures do not address the
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gineering judgment and the team has found excessive reliance on its use.
This is considered to be a significant weakness.
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1.3 SUPMARY - MECHANICAL COMPONENTS
The team performed a design review of selected modifications to piping systems
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and equipment at Fort Calhoun Station.
The team additionally reviewed
installed piping and equipment identified during a visit to Fort Calhoun
Station on September 20, 1985.
The team also reviewed the design basis for
balance of plant piping systems and equipment at Fort Calhoun Station.
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Om:ha Public Power District appears unable to access the design specifications
originally prepared by Gibbs, Hill, Durham and Richardson for Fort Calhoun
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Station, and has not prepared alternate controlled design documentation for
use by Omaha Public Power District design personnel.
(Deficiency D3.1-1).
In
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preparing the piping design for plant modifications, Omaha Public Power District
has not referred to the operating and accident temperatures and pressures originally
used by the architect engineer to analyze Safety Class 1 large-bore pipe, nor did
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Omaha Public Power District prepare alternate controlled pressure and temperature
data for use by Omaha Public Power District design personnel in lieu of
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retrieving the original design basis.
(Deficiency D3.1-2).
The team identified a
possible discrepancy between the generic spacing criteria which the contractor,
Peter Kiewit, used to install small-bore pipe during initial construction and the
minimum horizontal frequency criterion specified in USAR Appendix F for small-bore
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pipe ponerating or connected to the containment shell.
(Unresolved Item U3.1-3).
The team found that none of the six modification packages reviewed adequately
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documented the design basis for the installed piping and equipment.
The team
also found that none of the six modification packages reviewed adequately
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qualified the revised piping and equipment configuration by analysis, particularly
with respect to the Class 1 seismic' criteria detailed in USAR Appendix F.
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addition, Omaha Public Power District was not able to produce the seismic
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qualification of the junction box /unistrut support configuration which the team
identified during a visit to the site on September 20, 1985.
Of particular
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concern to the team are the numerous deficiencies identified in Omaha Public
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Power District modification request FC-C1-127.
Omaha Public Power District
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performed this modification in response to NRC Generic Letter No. 81-14.
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modification forms the basis for ensuring that the auxiliary feedwater system will
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function following the occurrence of earthquakes up to and including the safe shut-
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down earthquake for Fort Calhoun Station.
The team noted that a specific licensee
commitment to modify.an unstable valve operator prior to the end of 1981 had been
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made to the NRC, however the valve's support had not been modified.
1.4 SUMARY - INSTRUMENTATION AND CONTROL
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In the instrumentation and control area, the team found that most of the
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. design modifications were being accomplished in a controlled and technically
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acceptable manner. However, the individual items which were noted by the team
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appeared to be caused by certain common factors, such as:
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(1) Insufficient consideration of the original plant design basis
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(Observations 04.1-1 and 04.3-3).
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(2) Inadequate review of technical assumptions in the design (Observation
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04.2-1, Deficiency D4.3-1 and Unresolved Items U4.3-2, U4.4-1, and
U4.5-3).
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(3) Procedural weaknesses related to design verification (Deficiency D4.3-1
and Observation 04.3-3).
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In a number of instances noted by the team, the Omaha Public Power District
engineers did not indicate a sufficient awareness of the Fort Calhoun design basis
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when design modifications were developed and completed (Observations 04.1-1 and
04.5-2 and Unresolved Item 04.4-1).
Retrievability of some design basis infor-
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mation appeared to be a significant obstacle, and there was evidence of a strong
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dependence on the Updated Safety Analysis Report for such information.
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The team noted a strong dependence on use of checklists in the design, and
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design verification processes. While use of checklists is necessary, this
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practice does not appear to be sufficient to properly con. sider technical
-assumptions made by the design engineer in the design process.
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1.5 SlM ARY - ELECTRICAL POWER
The team reviewed three modification packages which were being' prepared for
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the present outage.
None of these packages had been completed, having only
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reached the Final Design Package stage.
The Construction Package including
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detailed design drawings, instructions and procedures, had not yet been prepared.
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~ The team 1.dentified problems in interpretation of the USAR requirements and
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referenced standards (Observation 05.1-3) because of what appeare'd to be . t
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insufficienttraceabilityoftheexistingplanttothe~or;iginalde'signbasis-
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for cable sizing.
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The team identified problems in the lack of review of, or inadequate justifi-
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cation for acceptance of, manufacturer's data or manufacturer's computer
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analysis for fire protection wrapping for raceways.
(Deficiency D5.2-1)
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The team identified inadequate coordination between the DC distribution
breaker and the new battery charger.
(Unresolved Item US.1-2)
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The team identified inadequate checking and verification of input and
assumptions used in the recent battery sizing calculation.
(Deficiency
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D5.1-1)
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In general, the team found the check and verification process to be highly
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dependent upon the experience of the reviewer.
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1.6' SlMARY - DESIGN CHANGE CONTROL
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Selected portions of the design change process were reviewed by the team.
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These portions included the preparation and approval of safety evaluations,
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the processing of emergency modifications, and an overview of documentation
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of design inputs.
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Several deficient conditions were identified concerning the performance'of
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'safet'y analyses pursuant to 10 CFR 50.59.
These include:
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(1) The licensee's procedures did not require safety anal'ysis of
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non-safety related changes to the facility as described in the USAR.
The
team identified a number of facility changes which did not undergo 10 CFR
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50.59 safety analysis (Deficiency 06.1-1).
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(2) Proposed facility changes accomplished as emergency modifications
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were not subject to 10 CFR 50.59 safety analysis before the affected
system was relied upon for facility operation (Unresolved Item U6.1-2).
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In addition the team also identified several weaknesses in the licensee's
process of performing safety analyses.
These include not specifically
requiring a review to ascertain if technical specifications were affected
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by ching:s, cnginsors not routinely availing themselves of NRC's position
regarding the safety analysis (Safety Evaluations), and not
differentiating between safety analysis required by 10 CFR 50.59 and
written determinations that plant review committee review items do not
constitute unreviewed safety questions (Observation 06.1-3).
The team identified a weakness regarding what the team considers excessive time
to complete engineering and design work associated with emergency modifications
(Observation 06.2-1).
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The team also identified as a deficient condition improper design control
associated with an emergency modification.
A modification to the
auxiliary feedwater system was only partially completed in 1980. The partial
completion was not identified until engineering review of the modification three
years later, and not corrected until the current outage.
This was not with-
standing an engineering determination that the modification violated 10 CFR 50,
Appendix A, General Design Criteria 57 (Deficiency D6.2-2).
Deficiencies and
concerns relating to the design change process were also identified on a
discipline-specific basis.
2.
MECHANICAL SYSTEMS
This portion of the inspection evaluated the mechanical systems design
aspects of plant modifications with emphasis on technical adequacy:
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capability of the modified system to perform all of its safety functions
as prescribed by its design bases and associated safety analyses, the
post-modification testing to demonstrate that the modified system will
perform its safety function, traceability of the design input through to
design output, and the design verification process.
The team reviewed two modifications in detail and two additional
modifications in part.
Initially the inspection concentrated on planned
modifications, MR-FC-83-158 and MR-FC-84-144; however, completed
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modifications, MR-FC-78-43 and MR- FC-81-21B, were reviewed during the
latter stages of the inspection to determine if concerns developed during
the review of planned outage modification package were also evident in
completed modification packages.
2.1 MECHANICAL DESIGN ACTIVITY TO SUPPORT CURRENT OUTAGE
The team examined summaries of the modifications planned for the 1985
outage and selected four modifications with the potential for significant
engineering activity. These four modification packages were further
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evaluated at Omaha Public Power District's offices to identify those
packages suitable for detailed review.
A candidate modification package was one which altered the manner in
which the system or component was operated and which required engineering
calculations, drawing revisions, procurement specifications, and
procedural revisions to implement.
Modification MR-FC-83-158 was
selected for detailed review.
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MR-FC-83-158 is a normal modification (as opposed to an emergency or
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minor modification; Deficiency D6.2-2 addresses the classification and
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handling of this modification request) to install an air accumulator on
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each of two valves, YCV-1045 A and B. These instrument air accumulators
were to be installed to permit remote manual isolation in the event of a
steam generator tube rupture with a concomitant loss of non-safety-related
instrument air.
Steam is supplied to the auxiliary feedwater turbine pump
from a steam header fed by two steam branch lines, one from each steam
generator. The steam is supplied from each branch line up to downstream
isolation valve YCV-1045 through normally closed isolation valves YCV-1045A
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and YCV-1045B.in the branch lines. These isolation valves are pneumatically
operated and can be remote manually operated from the control room.
In the
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origina1 system design, valves YCV-1045 A and B were designed to fail open on
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loss of instrument air, and valve YCV-1045 was designed to fail closed.
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modification, MR-FC-78-43, was initiated on an emergency' basis in 1978 to
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redesign the valve operator for YCV-1045 and replace it with a fail open-
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. operator.. In addition to replacing the valve actuator of YCV-1045,9:aii accumu-
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- .lators wer~e to be added to the valve actuators for YCV-1045 A and B to permit
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' the remote manual isolation of these valves from the control room in the
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event of a steam generator tube rupture with a concomitant loss of
non-safety-related instrument air. However, these air accumulators were
not installed prior to returning the plant to power.
During close-out of
MR-FC-78-43, a new engineering evaluation and assistance request was
initiated and became MR-FC-83-158 to install the air accumulators.
The
close-out of emergency modification MR-FC-78-43 and safety evaluation
which permitted plant operation are described in section 6.2 of this
report and Deficiency D6.2-2.
The Final Design Description for.MR-FC-C3-158 states that each
accumulator is to be sized to provide aire to the valve actuators for one
hour.
However, design calculations do not exist to confirm sizing of the
air accumulators.
A calculation does not exist which demonstrates that a
sufficient stored volume of pressurized air will be available to close
YCV-1045 A and B assuming a loss of instrument air.and minimum initial
accumulator pressure.
The valve is a spring actuated to open, and
sufficient air pressure must be provided to overcome spring pressure
and approximately 1000 psi differential pressure across the globe valve
during closure.
The team was informed that a sizing calculation was not
performed for this modification package; instead the design engineer
indicated that he referred to calculations in a completed modification and
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' fo'r' review the sizing calculations used by the design engineer; however,
these calculations could rot be found during the inspection and, therefore,
were not available for review (Deficiency D2.1-1).
Appendix F of the USAR describes criteria for the seismic design of structures
and components, including instruments and controls.
The Critical Quality
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Element List identifies those structures, systems, and components which are
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safety-related.
These documents form the basis for the design requirement that
the air accumulators and associated valving and tubing be safety-related and
seismic designed.
Contrary to these requirements, the procurement
specifications for check valves used to isolate the accumulators from the
non-safety-related instrument air header, and the isolation valves used to
isolate the accomlators from the valve actuator and instrument air header for
maintenance, do not specify seismic requirements (Deficiency D2.1-2).
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In addition, the team found that a vendor exception to the specification's
storage requirement for the accumulator isolation valves was not reflected in
the procurement document (Observation 02.1-3).
The team found no documentation in the modification file concerning the
seismic and stress analyses for the air accumulators.
The team was referred to
a generic calculation not applicable to a particular modification package.
The
team found that the control and use of generic calculations are not described in
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the Generating Station Engineering procedures and are the exception rather than
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the rule.
(Observation 02.1-5).
In examining the verification process used to confirm the final design,
the team determined that MR-FC-83-1SS'was not treated as a normal modification
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in accordance with established Generating Station Engineering Procedures
Manual.
This manual describes the responsibilities of persor.nel, the types of
modification requests, the information to be included in preparation of a
modification package, and steps to document field changes and close-out.
The
team found that a construction package was prepared even though the design
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verification of the final design package had not been completed.
The team
noted that the actual implementation of the design verification process does
not differentiate between normal and emergency modifications, although
Generating Station Engineering procedures indicate that third party
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verifications will be accomplished as the design package for a normal
modification is completed.
The team was informed that the procedures do not
have this intent but rather permit design verifications to be performed prior
to site acceptance.
The team was informed that Standing Order G-21 was the
governing procedure; however, this order does not indicate that completion of
third party verifications of a normal modification package can be delayed until
site acceptance (Deficiency D2.1-6).
The team determined that post-modification testing for this change was
inadequate to confirm that YCV-1045 A and B could be shut against a
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differential pressure of approximately 1000 psi and remain shut for one
hour with the accumulator air volume alone.
The post-modification testing
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procedure closed the valves with no differential pressure and with instrument
air header pressure instead of accumulator air pressure prior to commencing the
testing.
The only acceptance criterion is that the valves must remain shut for
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one hour. As a consequence, the test is a static air pressure test with no
acceptance criterion provided for acceptable air leal. age (Deficiency D2.1-7).
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Review of USAR sections, flow diagrams, and systea descriptions for
various systems was performed to understand the design bases for these
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systems and the impact the design change had on these systems.
The team
found incorrect information on the main steam system flow diagram (Deficiency
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D2.1-8) and in the auxiliary feedwater system description (Deficiency D2.1-9).
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--.
-.
.
--
. _ .
_
i
.
.s
-
.
.
The most significant error was the incorrect representation of the piping
arrangement associated with main steam isolation valve bypass valves and the
<.
auxiliary feedwater steam warmup lines, and as a consequence, the incorrect
'
identification of safety class boundaries.
A safety-related portion of the
system may have been considered as non-safety-related.
- The team also determined that the installation package for MR-FC-83-158
did'not reference a generic support spacing procedure for the installation of
the instrument air lines, and did not specify the radial location of the Hilti
.l
bolts which restrain the air accumulators.
This deficiency is described in
section 3.2 and Deficiency D3.2-2 of this report.
i
J.
?
' The team also examined modification MR-FC-84-144, because it was identified as
a modificati~on package that could be incorporated into the~ plant simultaneously
~
'
with MR-FC-83-158. MR-FC-84-144 is a normal modification involving the
~
~
'.
replacement of solenoid valves for YCV-1045 A and B.
'The: team' identified'no
-
"
. . . .
.P
. ~ concerns with this relatively simple modification to: replace a 'comporient with a
..
[..
' like component.
However, the team noted an apparent discrepancy between
-
MR-FC-84-144 and MR-FC-83-158 with respect to the use of fluorocarbon elastomer
material.
Specifically, the design package for MR-FC-84-144 indicates that
Viton, an E. I. duPont de Nemours trade name, is not recommended for
,
i
'
-
application in high radiation areas and the modification is to replace the
'
solenoids with ones which do not contain that material.
However, MR-FC-83-158
permits Viton to be used as a seating material in the safety-related
1
i
applications without proper consideration of material compatibility (Unresolved
Item U2.1-10).
2.2 DESIGN ACTIVITY ASSOCIATED WITH COMPLETED MODIFICATIONS
i
The team selected modification MR-FC-81-218 for review because it is
i
similar to modification MR-FC-83-218, which replaced fail close pneumatic
actuators with actuators that fail open.
The replacement actuators were
installed on valves HCV-438B and HCV-4380.
These valves are containment
l.
isolation valves located outside containment in the component cooling
.
water supply and return lines associated with the reactor coolant pump
~j
lube oil coolers and seals.
Like MR-FC-81-218, this modification added
instrument air accumulators to these valves to permit the operator to
maintain th'e valves closed until operator action could be taken to manually
close the valves locally.
In addition, the modification added a component
, cooling water pressure low signal in series with a containment isolation
3
. .. . actuation signal such that the presence of both signals -is necessary to close
~
-
the valves.
-
-
-The modification file contained a calculation which used~ incorrect and
E
'unconservative design input to demonstrate that the air accumulator had
~
,
sufficient capacity.
In the calculation, the volume of ~ air stored in the
accumulator is over estimated by 335 percent and the available air pressure is
assumed to be equivalent to the maximum instrument air system pressure instead
of the minimum pressure.
In addition, the calculation does not consider system
leakage or the period of time that the valve must remain shut. The team noted
that surveillance testing is not performed to demonstrate the capability of the
safety-related portion of the instrument air system to close these valves and
to maintain them closed for a finite period of time without loss of function,
j
9
.
.
..
_
.
-
. _ _
. .-
_
-
- . _ . . -
.
-
..
.
-
.
As a consequence, the implicit assumption of zero leakage is not conservative
and realistic (Deficiency D2.2-1).
HCV-4388 and D are containment isolation valves which are open following
an accident and must be capable of being shut throughout the course of
,
the accident. As a consequence, the air accumulators and associated
.
piping and valves are seismic Class I and safety related.
However, the
'
team found that seismic and quality requirements were not properly addressed in
the modification package.
Specifically, the team found a purchase order for
seismic qualification analysis of the replacement actuator and valve assembly
-
did not invoke the requirements of 10 CFR Part 50 Appendix B.
As a
j
consequence, the analysis was not performed in accordance with the service
1
organization's quality assurance program.
The team also found that the
installation / test procedure did not reference a generic support spacing
. procedure for the installation of the instrument air tubing.
In addition, no
calculation existed at the time the modification was completed to confirm that
the as-constructed, air accumulator, including base plate and Hilti bolts, was
>
adequately sized to withstand expected seismic loadings (Deficiency D2.1-2).
l
Like the post-modification testing procedure for MR-FC-83-158, the post
modification testing procedure for MR-FC-81-21B did not require the use
.
of the pressurized volume of the accumulator to shut the valves and only
?
a static air pressure test was performed.
In addition, the team found no
documented basis for the acceptance criterion that the valves remain shut
for twenty minutes (Deficiency D2.2-3).
The team also identified information missing from modification file
MR-FC-81-218.
No records of third party review were found by the team.
(Observation 02.2-4).
.
Portions of the compressed air and component cooling water systems were
reviewed to understand the design basis and the impact the design change
had on these systems.
The team found the compressed air system
description was not updated to include valves HCV-4388 and D on a list of
.
valves equipped with instrument air accumulators,
The team noted that
,
site acceptance of this modification indicates that the system
description had been updated.
The site acceptance of this modification
was complet'd in May 1983 and the compressed air system description was
e
-
most recently revised in April 1985.
The component cooling water system
description was not revised to correctly describe the change incorporated
in the logic circuits to shut HCV-4388 and D.
The team noted that site
acceptance for this modification was completed in May 1983 and that the
system description had not been updated since December 1981 (Deficiency
02.1-9).
During the team's review of the installation and test procedure, the team
found two instrument air header isolation valves not depicted on an
instrument air diagram showing riser details.
During a field inspection,
t
the team confirmed that the valves are installed in the plant (Deficiency
D2.2-5).
The team also reviewed the safety evaluation included in the Final Design
Description for modification MR-FC-81-218.
The team found the safety
evaluation was based on an incorrect assumption and analysis
methodology.
The safety analysis did not refer to original design
calculations and the qualitative argument used reflects an incorrect
10
-
.
gr
--s-
--
-we-
w
,, ve
,_m
w,-wv-,,..m-y=,.g--
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y
y-
,-w*,-w,-,-.--i.--.---tm.ev----w--
-v
e-
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- - ,--
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.
.
. :
_
understanding of the heat transfer phenomenon between heat removal
systems. The safety analysis contains an unsubstantiated and
inappropriate assumption concerning operator action to secure heat: loads
under certain accident conditions.
Although the basis of Technical Specification 2.4 contains incorrect information concerning the heat
removal capacity of the component cooling water heat exchangers, it was
not revised (Deficiency D2.2-6).
3.
MECHANICAL COMPONENTS
_
This portion of the inspection evaluated selected modifications to piping
, ,
. systems and equipment at Fort Calhoun Station from piping analysis, piping:and:
-
_
- equipment support standpoint.
Six modification requests were included in the
'
. , , ,
. scop.e of.this review.
Several issues relating to piping / equipment supports -
- - -: :
- -
were identified during a plant tour.
'
' ' ' ~
-
-
- -
3.1 Design Basis
--
-
,
The team reviewed information relating to the design bases of Fort Calhoun
i
Station Critical Quality Elements, Class I and II components, and their
relationship to ASME classifications and other code requirements.
.
The team could not obtain the design specifications governing the
i
procurement, design, fabrication and installation of the balance of plant
(supplied by other than the steam supplier) piping systems and equipment for
for Fort Calhoun Station that were issued by Gibbs, Hill, Durham and
- Richardson, the architect-enginee.' (A/E).
Instead of the design specifi-
cations issued by the architect-engineer, Omaha Power Public District
'
Generating Station Engineering (GSE) uses the original procurement specifi-
cations which formed the basis for the design specifications issued by the
architect engineer.
However, these procurement specifications are not
-
controlled documents (Deficiency D3.1-1).
As an example, the piping design
specification of record at Fort Calhoun Station is Gibbs & Hill Piping Specifi-
cation H-1, as noted in General Note 10 of the piping and instrumentation
diagram symbol list.
However, Omaha Public Power District uses Technical
Specificati.on No. 1/ Piping, one of 46 design specifications contained in Omaha
l
Public Power District Contract No. 763, Section H, which detailed the procure-
ment, design, fabrication, installation and testing for much of the balance of
'
(
plant piping and equipment.
- -
.
. The team could not obtain the operating and accident' temperature and pressure
-
data originally prepared by the architect-engineer for piping systems at Fort
-
Calhoun Station during the inspection.
The original piping analyses performed
. . by the architect engineer do not appear to be accessible.
In 1979, in order
-
, to perform reanalysis of large-bore safety class piping systems in response
t
l
to IE Bulletin 79-14, Omaha Public Power District collated operating and
l
accident temperatures from the FSAR and from analytical and operating data
l
for use.by Gilbert / Commonwealth.
However, Omaha Public Power District never
L
controlled this document (Deficiency D3.1-2).
The team is concerned that
l
Omaha Public Power District may be using this uncontrolled temperature data to
j
perform modifications to the installed piping.
I
Small bore pipe was supported in accordance with the generic spacing criteria
j
developed by the architect-engineer and detailed in Peter Kiewit (the
11
.
c-
.,
3,
-- - - .
_
. - . - . , , - , . - - _ _ , , , , , _
,,,.,,,.,4--
y--
-,-.,,,,,,,_-._,v,.,---,,-_7.-.,._,,,,.-,..,,,..,_,,ym_
, - - , - . , . - -
y,
, _ . , .
y, pen
c
.-y,.,-,
-
-
_ . .
- _ . .
- _
_
--
_
-
__
.
..
.
-
.
constructor) Contract 763/ Group I and Group II Piping Systems / Recommended
Procedure for the Support and Seismic Restraint of Piping two Inch and Smaller.
The tabular data in this procedure details allowable pipe support spacing in
the intake structure, auxiliary and containment buildings as a function of
pipe diameter (1 in. to 2 in.) for small-bore piping subjected to combined
dead load, thermal and seismic loads. The architect-engineer was required to
!
provide piping thermal displacements and reactions to the contractor.
Omaha
Public Power District was not able to access this information for the team.
This issue had previously been identified by the NRC Region IV office
(Inspection Report 85-03).
.
The seismic criteria embodied in the small-bore pipe generic spacing criteria
are based on minimum natural frequencies of 6 cps horizontal, 18 cps vertical,
for the auxiliary building and containment.
The minimum natural
frequencies specified for the auxiliary building and containment in the
procedure are in agreement with the seismic criteria detailed in USAR
,
l
Appendix F, Subsection F.2.2.2.
However, USAR Appendix F specified a
more stringent minimum natural frequency of 12 hertz (rather than 6
hertz) horizontal for piping runs which penetrate or are connected
to the containment shell, as a consequence of a slight amplification in
equipment acceleration response to the normalized ground response spectra at
l
approximately 6 hertz.
Omaha Public Power District could not confirm that
small-bore piping systems penetrating or connected to the containment shell
are in compliance with this minimum frequency criterion (Deficiency D3.1-3).
<
l
The team also reviewed the procurement specifications for balance of plant
valves with respect to the USAR Appendix F seismic criteria for Class I
'
piping and equipment.
The Omaha Public Power District Critical Quality Element
List notes that the valve specifications were developed by Omaha Public Power
i
District and the architect-engineer, and that the valve specifications can
be obtained by referring to the original contract documents.
The team examined
-
the design specifications contained in Omaha Public Power District Contract
No. 763, the original contract document used to procure the bulk of the
balance of plant valves for Fort Calhoun Station.
No seismic criteria are
-
detailed in these specifications. Therefore, documentation is not available to
establish seismic qualification of these valves.
Adequacy of equipment seismic
qualification documentation, including valves and operators for older plants, has
already been identified by the NRC.
This issue is being tracked for resolution
as Unresolved Safety Issue A-46.
The team reviewed draft NUREG 1030, Seismic
Qualification of Equipment in Operating Nuclear Power Plants, and the associated
Regulatory Analysis published in the Federal Register (FR 85-21054) on
September 4, 1985.
The team noted that Fort Calhoun Station is one of the plants
identified for NRC review pursuant to this Regulatory Analysis (Observation
03.1-4).
,
'
3.2 PIPING / EQUIPMENT REVIEW
'The team reviewed modification request FC-84-61, which will enable the periodic
removal of safety injection relief valves S1 209, 213, 217 and 221 for setpoint
testing.
The team noted that MR-FC-84-61 does not reference the source of design input
used in the analysis and did not reference various applicable design bases nor
document engineering judgement that such references were not needed.
!
(Unresolved Item U3.2-1).
12
- . - -
-
--
. - . -
.
- _ _ -
. - - . -
.
_
.
..
. . . . . .
_ , . .
-
..
i I
.
'
.
..
.
.
-
,
Modification request FC-83-158, which provides air accumulators with check
.
valves for valves YCV-1045A and B, does not reference a generic support spacing
F
procedure for the installation of the instrument air lines, and does not specify
.
.the radial location of the Hilti bolts which restrain the air accumulators
(Deficiency D3.2-2).
The team reviewed a second modificiation package which
was similarly deficient (reference Deficiency D2.2-2).
Modification request FC-84-162 redesigns two containment ventilation duct
supports to improve personnel and equipment access.
However, the Omaha Public
-
Power District calculation does not consider Design Bases Accident thermal
loads of 288 degrees F.
In addition, the revised duct support configuration,
u
i
which consists of. a horizontal angle and a brace, is not analyzed for the
g
'
combination of vertical seismic load and transverse horizontal seismic load
~(Deficiency D3.2-3).
I'...,Duringa.sitevisitonSeptember
20, 1985, theteamidentifieda' junction. box;
--
~
- ~ 'which. supplies power to the operator for valve YCV-10458.
The junction box is
_
.. restrained by a pair of unistrut supports which are in turn supported by
-
--
' conduit.
Omaha Public Power District could not identify a seismic analysis
j
which qualifies this configuration to the governing seismic provisions of USAR
Appendix F (Deficiency D3.2-4).
Modification request FC-83-83 replaces the containment pressure switches which
feed the engineered safeguards system high containment pressure logic matrices.
-
The team reviewed the seismic qualification of the pressure switches and the
y
associated switch supports and Hoffman boxes.
The Omaha Public Power District
W
-
calculation performed to qualify the support for the pressure switch does not
reference the vendor drawing for the pressure switch, preventing confirmation
j
of the switch dimensions and weight used in the analysis.
In addition, the
Hoffman box shown on the Omaha Public Power District arrangement drawing was
=
2
not identified in the Omaha Public Power District calculation.
(Observation
1
03.2-5).
E
-
Modification request FC-84-92 does not adequately implement or reference the
'
I
design basis for Fort Calhoun Station.
This modification contracted for the
.
i8
design and fabrication of nozzle dams for the hot and cold legs of the steam
generator, to enable refueling to proceed concurrently with primary head work
-
such as eddy current examinations.
The steam generator nozzle dams were
-
designated as Critical Quality Elements (CQE) on the Omaha Public Power
-
a
1
'
District nozzle dam purchase order; and were, therefore, subject to the
.
3
governing Class I seismic criteria detailed in USAR Appendix F.
Howeve~r, the
E
...
~
Omaha Public Power District contract to the nozzle dam vendor, Nuclear Energy-
)
~ . Services, did not specify any seismic provision, and Nuclear Energy Services
j__
did not perform a seismic analysis (Deficiency D3.2-6).
U
During a site tour, the team examined auxiliary feedwa'ter steam feed valve
'-
YCV-1045B because of a planned modification to add an air accumulator to the
,
g
valve's air operator supply line.
The team questioned the operator's existing
support arrangement in that it was supported by a thin rod attached to a stair
,
3
post and did not appear seismically qualified.
--
The team reviewed modification request FC-81-121, which Omaha Public Power
y
District performed in response to NRC Generic Let;er No. 81-14, Seismic
9
Qualification of Auxiliary Feedwater Systems.
Gilbert / Commonwealth performed
i
_
-
g
13
_
.
?
- - - - - - - - -
-
.
.
...
.
-
-
.
-. _
_
_
..
._
i
- .
.
.
a walkdown of the auxiliary feedwater system at Fort Calhoun Station for Omaha
)
Public Power District, and identified four major seismic deficiencies, one of
'
wt'ich involved unstable valve operator supports.
Gilbert / Commonwealth
specifically noted that the valve operator for valve YCV-1045B was unstable, and
recommended that the existing rod restraint to replaced with a strut.
Gilbert /
Commonwealth recommended the addition of a number of support, for the steam
drive and condensate portions of the auxiliary feedwater piping' associated with
pump FW-10.
Gilbert / Commonwealth also recommended that a detailed stress
analysis of the auxiliary feedwater system be performed to confirm that the
addition of restraints to the auxiliary feedwater system would not result in
.
'
excessive thermal loads.
Omaha Public Power District Generating Station
Engineering (GSE) elected to perform the recommended piping analysis.
The team
noted the following deficient conditions relating to this analysis.
(1) The valve operator for valve YCV-1045B is currently restrained by a
rod which is anchored to a stairpost fabricated from a steel angle; the
strut substitution recommended by Gilbert / Commonwealth was not
implemented;
(2) The Omaha Public Power District as-built drawing does not detail
either the valve operator or the existing valve restraint;
(3) The vendor drawing for valve YCV-1045B could not be accessed to
verify the valve and operator weights, or the operator offset dimension
used in the Omaha Public Power District piping analysis;
,
(4) The valve operator restraint was not modeled in the Omaha Public
!
Power District piping analysis;
-
i
(5) Omaha Public Power District could not access a summary of the pipe
stress due to combined dead, thermal and seismic loads in the vicinity of
the valve;
(6) Omaha Public Power District could not access a summary of the
reactions due to combined dead, thermal and seismic loads for the
'
supports adjacent to the valves.
Based upon cursory examination of the
computer output, the supports appear to be overloaded;
(7) The Omaha Public Power District computer runs are not referenced in
the modification request and are therefore not auditable; and,
(8) As noted in Deficiency D3.2-7, the licensee specifically committed to
correct the unstable valve operator on YCV 1045B by the end of 1981 in a
,
July 14, 1981 letter to the NRC.
In addition, the NRC project manager
documented a telephone conversation with the licensee, confirmin'g completion
,
of this action (correction of unstable valve operators), in the NRC letter
'
forwarding the Safety Evaluation for Generic Letter 81-41, Seismic Qualification
of AFW Systems.
Contrary to these commitments, valve YCV-1045B and the adjacent piping and
supports were not adequately analyzed to the governing seismic provisions of
1
USAR Appendix F (Deficiency D3.2-7).
i
14
i
- .
- . - - . -
-
- - -
-
.
- -
--
_ _ _ _
__
__
_
-
.
4.0 INSTRUMENTATION AND CONTROL
The. team reviewed instrumentation and control design modification packages for:
(1) consistency with Fort Calhoun design basis requirements; (2) conformance
with applicable regulatory criteria and FSAR commitments; (3) technical. ,
adequacy of the chosen design approach, and (4) completeness of design details
and independent verification reviews relative to Omaha Public Power District
procedures.
During the Fort Calhoun outage inspection program, the team reviewed the
.
following fourteen instrumentation and control design modification packages:
Undervoltage Protection;
.
Reactor Coolant Hot Leg Level Indication ~;
_
Bypass or Trip of ESF Channels Without Jumpers;
~~
.
.
HEPA Filter Differential Pressure Indica ~ tion;
.
-
~
Containment Pressure Switches;
'
~~~
-
Transfer of P250 Points to the ERF Computer;-
--
~
High Power Rate of Change Trip Alarm;
-
Fuse Protection for Certain Limit Switch Circuits;
Replacement of Safety-Related HFA Relays;
Delta Temperature Power Process Loops;
Thermal Margin / Low Pressure Trip Drawer Modification;
Addition of Main Feedwater Valves to S/G Isolation;
Replacement of Component Cooling Flow Element, and
Redundant Fusing for Alternate Shutdown Circuits.
The technical approach and design content provided in five of the desigr)
modification packages were satisfactory based on the team's review w'ith
Omaha Public Power District personnel. The team had no further questions
regarding design modifications FC-81-64, FC-84-96, FC-84-152, FC-84-179,
and FC-85-80.
4.1 ANNUNCIATION OF REACTOR PROTECTIVE SYSTEM TRIP BYPASSES
-
Annunciation of reactor protective system trip bypasses to the control
room operatpr has been a long-standing requirement, but may conflict with
more recent human factor recommendations regarding operator displays.
Design modification FC-84-46, which involved conversion of a high rate of
change of power trip alarm annunciator to be a high rate of change trip
,,
. enable. annunciator, was developed to address the human factor considerations._
,
The original Fort Calhoun design met the trip bypass indication commitment
by causing one annunciator window to be illuminated except for brief periods
during plant startup or shutdown. During the prop.osed conversion of this
. annunciator to a " dark-board" concept as recommende'd by human factor con-
siderations, the design modification eliminated the existing trip bypass
annunciation.
The proposed design modification did not address or attempt to resolve
the conflict in annunciation requirements for this particular trip bypass.
(Observation 04.1-1).
15
..
- .
- -
. _ -
..
.
.
Near the end of the inspection, Omaha Public Power District personnel were
considering a slight modification to two neutron monitoring system alarms
that would permit concurrent satisfaction of both the IEEE commitment and
the desired dark-board annunciator concept.
4.2 ANALYSIS OF SAFETY-RELATED INSTRUMENTATION ACCURACY
,
1
Analyses involving safety-related equipment, or critical quality elements
-
(CQE), are required to meet commitments to ANSI N45.2.11 and applicable
licensee procedures.
The team reviewed a supporting technical analysis
for FC-84-140, which involved replacement of temperature dectectors and
related instruments for delta temperature power process loops used by the
reactor protective system.
,
The OSAR-85-83 analysis prepared by Technical Services was not identified as
being either safety-related or related to a Critical Quality Element and did
not contain the calculation formula used to derive analysis results.
The team
was unable to independently review this analysis, and required assistance from
Omaha Public Power District personnel to confirm the correctness of the
!
calculation results.
The team also noted that Technical Services procedure N-TSAP-5 did not
'
contain all of the requirements applicable to safety-related calculations
as described in Generating Station Engineering procedure B-9 (Observation
04.2-1).
4.3 DESIGN ASSUMPTION IDENTIFICATION, DOCUMENTATION, AND CONFIRMATION
The identification and documentation of technical assumptions made during
.
the design process, and a timely confirmation of their validity, are
important design control and design verification elements specified in
ANSI N45.2.11. A number of design modification packages were examined for
implicit and explicit design assumptions.
-
The team identified a weakness in the Omaha Public Power District
l
'
instrumentation and control design process regarding the identification,
documentation, and confirmation of design assumptions.
Fuse protection of certain solenoid-operated valve limit switch circuits,
as described by design modification FC-84-74A, did not identify and
resolve an implicit technical assumption regarding the coordination of
two types of fuses (Deficiency D4.3-1). This omission is not in
l
- accordance with the design evaluation requirement in Generation Station
Engineering procedure B-2. A need to confirm the coordination of two
different fuse types had been noted on a design checklist by a
third party reviewer; however, the design package did not provide any
indication that the coordination had been confirmed or that it was
appropriate.
'
For design modification FC-81-102, involving engineered safety feature
keylock bypass switches, the need to specify an appropriate combination
of keylock cylinders and bypass keys to cugment plant administrative
16
_
-g.mr.r--,y--g-
g-y+,-
---s-
,,,,,-
,,-,y-,
, - ,-,,-,
-
,.-g-,,,,.,,,w
,,,,,,-m--y--,,----,
, - --+--, . - + -- - . .
.w-----...
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-w,r,-,.--
- .w-
w-
- -isr
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.
.
.
,_
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.
..
.
.
controls was not identified by the design engineer (Unre~ solved Item
'
U4.3-2). The final design package technical description and design
evaluation sections, required by procedure B-2, did not contain all of
the equipment requirements necessary to establish an unambiguous design
'
. configuration. Such design provisions would help assure that only one
,
channel could be bypassed at any given time.
-
4.4 CONTROL ROOM PANEL WIRING SEPARATION
The team examined field cable and control room panel wiring separation
criteria and implementation practices.
The team observed that Omaha Public Power District has been attempting to
~ .'
, achieve current industry separation implementation practices for some
. .
y
!;,,
. recent design modifications. The team noted that separation criteria -
-
-
-C
a- - . existed on drawings for cabling external to the control ~r~ooin ~ anels, and
~
~
p
observed its implementation during visits to the plant. An FSAR
commitment made in 1970 stated that physical separation ~ of individual
~
. channel components and wiring would be maintained wherever practicable.
The team determined that achievement of internal wiring separation within
panels was a General Electric responsibility, but was unable to locate
j
documented criteria during the inspection. The team observed that Omaha Public
Power District had developed definitive separation criteria for internal panel
wiring and harnesses over the past several years, and had implemented
appropriate separation criteria for some panel modifications.
The team identified an ambiguity with the Omaha Public Power District
position regarding the minimum separation distance of safety-related
. wiring from both redundant safety related wiring and non-safety-related
,
wiring within control room panels (Unresolved Item U4.4.1).
l
l
Implementation of design modification package FC-77-40 for undervoltage
l
protection was questioned by the team because of a lack of separation
among safety-related wiring for redundant channels emerging from separated
'
-
barrier enclosures within panel CB-4.
The need to separate these wires or to
provide a justification analysis was not addressed. The design of this
modification does not meet current Omaha Public Power District separation
practices, and is not in accordance with a USAR section 7.3 commitment for
separated and segregated engineered safeguard controls.
Similarly, the
. . .
c FC-81-102: design modification did not maintain adequate separation distances-or
-
-l justify the. separation distance between safety-related a~nd no~n-safety-related
-
--
,
-wiring for.the planned addition of keylock bypass switches for three engineered
safety feature process variables (Unresolved Item U4.4-1). The design of this
.
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modi.fication does not meet current Omaha Public Power District intended
- separation practices, and the need for an analysis to justify the association
.
of Class 1E wiring with non-Class 1E wiring to the annunciator was not
identified.
An analysis has not been performed to demonstrate that Class
1E circuits have not been degraded below an acceptable level, and the
acceptability of use of braided conductor wiring has been assumed but has
not been demonstrated. In this particular modification, the non-Class 1E
wiring to the annunciator provides a common link among all four redundant
ESF channels, and does not appear to comply with the USAR section 7.3
l
commitment for separated and segregated engineered safeguard controls.
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4.5 CONFORMANCE WITH PROCEDURES
The team reviewed a number of design modifications to assess the degree
of conformance with established Omaha Public Power District procedures.
For design modification FC-82-178, which involved the addition of differential
pressure indicators to a number of HEPA filter units, the design engineer did
not comply with Generating Station Engineering procedure A-9 which required
that a drawing sepia be issued to alert other individuals that a change was in
process (Deficiency D4.5-1).
-
During the plant walkdown, the team identified a battery fuse block
enclosure constructed of masonite and fiberboard in each of the battery
rooms (Unresolved Item U4.5-3). This enclosure was not identified as a
significant combustible in the fire hazards analysis. Omaha Public Power
District has not confirmed that this material is not a significant
combustible for the published fire hazards analysis.
5.0 ELECTRIC POWER SYSTEM
The electrical modification packages scheduled for installation at Fort
Calhoun during the 1985 outage were _in varying s.tages of completion
'
during this inspection. Of the major packages reviewed, none had
progressed from the Final Design Package to third party review at the
~
start of the inspection.
By the second week of the inspection some
third party reviews had been performed but still no construction packages
had been prepared even though the outage was then in progress.
The team
- looked at two modification packages scheduled for this outage in detail
and a third package in overview.
The team also looked in detail at an
-
additional' modification package completed in an earlier outage that had a
direct impact on one of the 1985 modifications reviewed by the team.
5.1 DIRECT CURRENT SYSTEM MODIFICATIONS
t
The team reviewed changes to the power sources and loads as detailed in
modification package MR-FC-84-119.
This package included changes to the
station batteries (which would reduce the available capacity), the
battery chargers (which would increase their capability to carry the
steady state de load), and the instrument inverters (which would add
additional load considerations on the de system).
The team revisved the effect which removing two cells from the batteries
would have on tr.e required capacity of the battery.
Omaha Public Power
District correctly based this determination upon sizing calculations
performed using a higher permissible limit on cell discharge voltage. As
part of this review, the team also reviewed completed modification
MR-FC-79-03, Replacement of Station Batteries.
The battery profile
established in the 1979 modification formed the sole basis for the latest
calculation input.
The' team found no justification for using this
earlier unchecked input data (Deficiency D5.1-1).
i
The team reviewed the replacement of the old battery chargers with new,
larger battery chargers and found that insufficient consideration had
been given to the interface between the DC Switchboard and the larger
charger.
Under the new design, the battery may not be able to be recharged
following a test or design basis discharge (Unresolved Item U5.1-2).
18
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The team reviewed the power cable sizing changes required because of the
larger equipment being added to the system and noted that no~ standard
interpretation of cable sizing for cable routed in cable tray existed for
Fort Calhoun (Observation 05.1-3).
5.2 RACEWAY FIRE PROTECTION
The team reviewed the analysis performed to determine the effect the fire
protective. wrapping would have on the enclosed cables.
This work was to
-
be performed under modification package MR-FC-85-25 in response to Appendix R
concerns regarding redundant power supplies for the pressurizer heaters being
routed through the same fire zone.
Omaha Public Power District proposed to reroute the existing BUS 3 motor .
. ; control center feeders through conduit and proposed to provide a.one hour fire
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1p'Interam E-50A" fire protection materials.'otection for these conduits by wrapping th
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The team determined that Omaha Public Power District used cable derating
factors obtained from the 3M Company based upon an internal 3M Company
computer program.
The team also determined that Omaha Public Power
District personnel were not familiar with the computer input or output
data supplied by 3M.
Omaha Public Power District failed to verify the
computer code used by 3M or even request 3M to supply correlation test
data before they were questioned on this by the team (Deficiency 5.2-1).
Regarding this item the team concluded, based upon subsequent test data
supplied to Omaha Public Power District by 3M, that the original derating
factors used by Omaha Public Power District were not in agreement with
the test resulti. Based upon the original modification package design,
degradation to power cables could have resulted to these motor control
centers feeders and to other power cables in similarly fire protected
raceways in the future.
5.3 LOAD CENTER TRANSFORMER REPLACEMENT
.
The team reviewed the effect changing load center transformers would have
on the elec'trical power transformers.
Replacement of these transformers
is scheduled for the 1985 outage under modification package MR-FC-84-105.
.
The team reviewed the specification data for the new transformers.
~ '
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- Manufacturing of the Fort Calhoun transformers had not been completed so
that no production test data was available for review. ~The team noted
,,
. .that the responsible engineer had reviewed the transformer impedance
requirements for compatability with the existing switchgear and had
discovered a problem with the transformers that had been replaced in an
earlier outage.
6.
DESIGN CHANGE CONTROL
The team performed a partial review of the Omaha Public Power District
design change process and safety evaluations associated with outage
modifications.
A limited review of completed emergency modifications was
also conducted.
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6.1 SAFETY EVALUATIONS
The team evaluated final design package safety evaluations for adequacy
in accordance with the requirements of 10 CFR 50.59.
Safety evaluations
were reviewed for modifications planned for this outage and for previous
work accomplished as emergency modifications.
The team reviewed Fort Calhoun Station Standing Order No. G-46,
'
" Evaluation of Procedures, Procedure Changes, Tests and Experiments for
Safety Evaluation and Status as an Unreviewed Safety Question." This
procedure provides guidance for the preparation of written safety
evaluations.
The procedure includes the subset of items listed in
Technical. Specification 5.5.1.7.b which requires the Plant Review
Committee to render determinations in writing with regard to whether or
not items constitute unreviewed safety questions. Other procedures
reviewed by the team also detail requirements for the performance of
safety analysis.
Generating Station Engineering procedure B-2,
" Production of Design Description and Evaluation," for example,
stipulates that safety analysis must be performed for safety-related
Critical Quality Element structures, systems and components.
The procedure additionally requires safety analysis for certain
non-safety-related modifications if such modifications have a reasonable
possibility of damaging safety-related components.
The team noted several problems with the licensee's procedures regarding
the performance of safety analysis / evaluations.
One problem noted by the team was the failure of the licensee to perform
safety analyses for certain non-safety-related changes, as required by 10 CFR 50.59.
10 CFR 50.59 allows licensees to change the facility as
described in the FSAR, without prior NRC approval, provided the change
doesnotinvolve?fsthangeinTechnicalSpecificationsoranunreviewed
safety question.
10 CFR 50.59 is also permissive for procedures (as
described in the FSAR) and for tests and experiments (not described in
described in the FSAR).
An unreviewed safety question is deemed to be
involved (a) if the probability of occurrence or the consequences of an
accident or malfunction of equipment important to safety previously
evaluated in the safety analysis report may be increased; or (b) if a
,
possibility for an accident or malfunction of a different type than any
evaluated previously in the safety analysis report may be created; or (c)
if the margin of safety as defined in the basis for any technical
specification is reduced.
The team found that safety evaluations were not being done in all cases
for non-Critical Quality Element (non-safety-related) systems that were
' described in the USAR. .Five non-Critical Quality Element modifications
were noted to have final design packages issued with no safety evaluation
accomplished:
1
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Modification
Request No.
Title
_
_
483-175
Feedwater Regulating System Instrumentation
Replacement
485-008
Boric Acid Addition System
4748-057
Power System Stabilizer
483-174
Reactor Regulating System Steam Dump and Bypass
-
Alarm
483-90
Replace LP Feedwater Heaters
.
Each of the above affected systems or equipment were described in the
USAR in sufficient detail that completion of the codifications would
require changes to USAR text, drawings or tables tc accurately represent
the~ newly changed systems.
A safety evaluation is required by 10 CFR
-
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J 50.59 for modifications such as the above even though the equipment and
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systems are non-safety-related (Deficiency D6.1-1).
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One of the major jobs being accomplished during the 1985 outage ~at Fort -
Calhoun was the installation of~ dams in the steam generator primary head
nozzles.
These were contracted to Nuclear Energy Services, Inc. by the
!
Technical Services Group of Omaha Public Power District and will allow
]
. refueling to proceed simultaneously with steam generator tube inspections.
. The team requested to review the 10 CFR 50.59 safety analysis associated
.
with the use of these nozzle dams.
The team was informed that the dams were
considered a tool and not a modification, and therefore that Generating
Station Engineering had not been required to perform a 10 CFR 50.59
safety analysis.
The team was told that a safety analysis would be
performed by Technical Services in accordance with Standing Order G-46
prior to use of the nozzle dams.
The team was concerned regarding the
i
diverse nature of the safety analyses' performed pursuant to 10 CFR 50.59
and those performed to implement Technical Specification requirements,
and questioned the licensee's methodology which results in bypassing the
'
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responsible design organization (Generating Station Engineering) for
performing safety evaluations for this type of facility change.
.
l
The team found also that final design package 10 CFR 50.59 reviews were
i
not accomplished and documented in all cases by Generating Station Engineering
design engineers prior to accomplishment of emergency plant modifications and
subsequent plant operation.
Three cases were noted; modifications for
correcting DC grounds on Critical Quality Element Safety Injection valves
.-
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(MR 484-84) and diesel generator speed sensing power supply modificati.ons (MRs
483-129 and 483-152) (Unresolved Item U6.1-2).
-
. .The team also noted that in general, the safety evaluations reviewed
tended to be quite brief with limited detail and analysis provided in
them. They simply answered the three questions posed by 10 CFR 50.59
with very little explanation provided to give the team confidence that
l
all safety concerns were being analyzed.
In some cases, where final
design package and construction package safety evaluations were
'
accomplished, the wording was identical.
This indicated to the team that
there was a lack of independent consideration between the final design
and construction safety evaluations as required by Omaha Public Power
District procedures.
In addition, unless these thought processes are
adequately documented, each' reviewer in the approval circuit must
reconstruct and reanalyze scenarios that may have already been
accomplished by the design engineer.
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,-,
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,-.y.
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Anothsr problem concerns the lack of a procedural requirement to evaluate
whether or not a proposed change (modification, procedure, test or
experiment) involved a change in the Technical Specifications; and
therefore, whether or not it can be implemented without prior NRC
approval.
The team identified one modification which internally
l
interprets the technical specifications to prevent the creation of
malfunction of a different type than previously analyzed in the FSAR.
Specifically, the replacement of the vital ac inverters creates a
possibility of powering the non-interruptible (battery powered) buses
which the inverters supply from interruptible (off-site power or diesel
_
generators) sources.
If more than one of these buses was powered from
interruptible power the plant would be in an unanalyzed condition.
To
address this concern, the engineer stated that if an inverter is in
bypass (bus powered from interruptible ac) the inverter would be
considered inoperable per the Technical Specifications.
The possibility
of a different type of accident may have been created even though the
associated safety evaluation contains an interpretation of Technical
Specifications which, if implemented, would lessen the probability of
occurrence of such an accident (Observation 06.1-3).
6.2 FINAL DESIGN PACKAGES
The team conducted a limited evaluation of final design packages of
planned outage modifications for adequacy with regard to the licensee's
implementation of its commitments to ANSI N45.2.11 design control
requirements.
Emergency modifications were also reviewed for issuance of
,
after-the-fact final design packages.
After review of numerous final design packages for this outage, the team
,
considered that design inputs were not clearly specified in the packages.
The final design packages were formatted with headings of Design Basis,
Technical Description and Design Analysis but the text generally
represented a narrative account of the problem and' solution with no clear
specification of design inputs.
ANSI N45.2.11, to which Omaha Public
,
Power District is committed, requires in part that "The design input be
specified...to provide a consistent basis for making design decisions,
.
accomplishing design verification measures, and evaluating design
!
changes." Design inputs were not found to be clearly spelled out in the
final design packages to meet this requirement.
In addition, the
Checker's Checklist-Design Package, provided in Generating Station
l
Engineering Procedure B-2, asks "Are design inputs correctly selected and
!
incorporated into the design?".
The team considers this step of the
,
checking process cannot be easily or accurately accomplished if design
inputs are not clearly specified as required by the ANSI N45.2.11.
-
During review of emergency modifications, the team noted that what it
considers excessive time had been taken to issue the final design
'
packages associated with six modifications.
Omaha Public Power District
procedure G-21 allowed completion of emergency modifications in the plant
prior to completing the final design package but provided no guidance for
timely issue of after-the-fact design packages.
Two of the six modifcations
.had after-the-fact design packages issued, but one was 11 months and the
other 42 months after completion of the plant modification.
The remaining four
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emergency modifications did not have after-the-fact design packages issued at
,
the time of this ins'pection.
These plant modifications had been completed for
~
a period ranging from 16 to 32 months.
Even though ANSI N45.2.11.does not -
' _
specifically address timely closure of work packages,~ it is. desirable from a
_
system acceptance standpoint to reduce the amount of time to prove that an
emergency modification has been implemented properly.' A tracking system did
exist for projects not closed out but insufficient management attention
appeared to be directed toward review of the listing and closecut of the
items (Observation 06.2-1).
-
One of the modifications (MR 83-158) being accomplished during the
'
current outage arose from the inadequate implementation of an earlier
modification performed on an emergency basis.
During review of
. . ~
modification MR 83-158, which adds accumulators to air operated (fail open)
-
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st'am ~ supply valves (YCV 1045 A/B) to the steam driven'AFW pump, the team -
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.
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determined that the modification had been initiated t'o allow closeout of -
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an'other modification (MR 78-43) performed on an emergency basis 1.n March
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1980.
When the closeout review of MR 78-43 was conducted in October
_
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1983 (a period of time which the team considers excessive), Generating
'
Station Engineering determined that the modification, as installed,
i
violated the General Design Criteria 57 requirement for the ability to
remote-manually isolate a closed system penetrating containment.
i
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A portion of original modification which added accumulators to the steam
supply valves to address containment isolation concerns had not been
accomplished.
The licensee closed out the original modification and
started a new modification to provide the accumulators.
The
i
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"after-the-fact" 10 CFR 50.59 safety evaluation _ determination that no
unreviewed safety question existed was based upon future work to be done
!
on another modification.
In January 1985, after additional internal
discussions between licensee site, Technical Services and Engineering
~
personnel, a decision was made to prioritize the installation of the
accumulators to be accomplished in the Fall 1985 refueling outage and to
revise emergency procedures to alert operators to the potential need to
locally isolate the AFW steam supply line following a steam generator
tube rupture.
A determination was made that an unreviewed safety question did -
not exist and that this was not a reportable event (Deficiency D6.2-2).
.
O
e
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7.0 BACKGROUND
7.1 MEETINGS
Interim Status Briefing
_
An interim status briefing on the status cf the safety systems outage modifi-
cation inspection program was conducted on October 8, 1985 at the conclusion
of the design portion of this program.
The following persons attended the
meeting:
-
Name
Title
Organization
O rchitzel
Teaiii Leader
NRC - IE
P. Surber
Sect. Mgr. - GSE
K. J. Morris
Manager - QA
OPPD ~
D. Wittke
V.P. - Engineering
J. K. Gasper
Mgr. Administrative Sycs.
R. L. Andrews
Div. Mgr. - Nuclear
Production
W. C. Jones
Vice President
J. J. Fisicaro
Supervisor - Nuclear
Regulatory and Industry
Affairs
J. C. Barker
Team Leader, Outage Insp.
NRC - IE
J. E. Konklin
Chief, Special Programs
NRC - IE
Section
B. Grimes
Director, Div. of Q.A.,
NRC - IE
Vendor & Tech. Training
J. L. Milhoan
Chief, Licensi.ng Se.ction,
NRC - IE
Quality Assurance Branch
G. Overbeck
Mechanical Systems
NRC - WESTEC
A. H. Saunders
Reactor Engineer
NRC - IE
M. Eidem
Mgr. - GSE Mech.
R. L. Jaworski
Section Manger - Tech.
Services
T. L. Patterson
Manager - Technical Support
,
S. K. Gambhir
Manager - GSE Electrical and
Nuclear
L. Stanley
Instrumentation & Control
NRC - Zytor, Inc.
A. V. Du Bouchet
Mech. Comp.
NRC - Cons. Engr.
S. W. Ky
NSC, Senior Researcher
Nuclear Safety
Center, Korea
G. Morris
Electrical Consultant
NRC - WESTEC
R. Lloyd
Reactor Engineer
NRC - IE
i
M. E. Murphy
Project Inspector
NRC'- RIV
!
W. Gary Gates
Manager - Fort Calhoun
Station
During the Interim Status Briefing, the team presented the significant
findings which had been identified during the inspection.
Periodic
management briefings were also held during the inspection.
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7. 2 PERSONS CONTACTED
Name
Title
-H.2Tiden
Lead Mechanical Engineer
__
Organization _ __
D. Ecklund
Mechanical Design Engineer
L. Gundrum
Mechanical Design Engineer
R. Eurich
Mechanical Design Engineer
S. Gambhir
Manager, Electrical & Nuclear
Engineering
-
P. Surber
Section Manager
J. Fisicaro
Supervisor-Nuclear Regulatory
and Industry Affairs
B. Livingston
Manager GSE Document Control
--
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- J. Albers
Document Control
R. Lewis
R. C. Kellogg.
Supervisor - Mech./GSE
.
Supervisor - Mech./TS
M. E. Eidem
Manager - Mech./GSE
T. L. Patterson
Manager - Technical Support
R. L. Jaworski
Section Manager /TS
J. R. Tucker
Electrical Design Engr./GSE
J. E. Bentzinger
Supervisor - Procurement QA
H. L. - Little
Supervisor, Electrical
H.J. Faulhaber
Supervisor, Electrical
L.W. Jackson
Lead Engineer, Electrical /I&C
.
W.C. Gartner
Senior Engineer, Electrical /I&C
'
R.P. Clemons
Senior Engineer, Electrical /I&C
B.R. Briganti
Engineer, Electrical /I&C
R.R. Ronning
Engineer, Electrical /I&C
N.B. McShannon
Senior Designer
.
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R.W. Coen
Senior Designer
E. Erickson
Design Verification Consultant
R. Mehaffey
Supervisor I&C/ Electrical Technical
Services
D. Haas
Mechanical Design Engineer
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D. Deboer
Mechanical Design Engineer
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LIST OF DEFICIENCIES, UNRESOLVED ITEMS AND OBSERVATIONS ___
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Item
Title
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D2.1-1
(Deficiency)
Lack of Design Analysis to Support Sizing of
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Air Accumulators for Valves YCV 1045 A/B.
D2.1-2
(Deficiency)
Seismic Requirements no't Specified in
MR-FC-83-158 Procurement Documents
02.1-3
(Observation)
Vendor Exceptions to Specifications not
Reflected in Procurement Document
2.1-4
N/A
Ites Number not Used
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...
02.1-5
(Observation)
Procedural Error Caused Se'ismic an'd litres's'
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Analysis for MR-FC-83-158 Not To Be Filed In
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Modification File
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D2.1-6
(Deficiency)
Failure to Follow Procedural Requirements for
~
a Normal Modification Resulting in Lack of
Required Design Verification Review
D2.1-7
(Deficiency)
Incomplete Installation / Testing Procedure in
Construction Package for MR-FC-83-158
D2.1-8
(Deficiency)
Incorrect Information on Flow Diagram for
Main Steam System
D2.1-9
(Deficiency)
Incorrect System Description Statements
'
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U2.1-10
(Unresolved)
Use of Fluorocarbon-Elastomer Material in
High Radiation Environments
D2.2-1
(Deficiency)
Incorrect Design Input in Calculation
.
Associated with MR-FC-81-21B
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D2.2-2
(Deficiency)
Incomplete Consideration of CQE and Seismic
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Class I Requirements for Portions of
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MR-FC-81-21B
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D2.2-3
(Deficiency)
Incomplete Installation / Testing Procedure
Performed for MR-FC-81-21B
!
02.2-4
(Observation)
Incomplete Modification File for a Completed
Modification
,
!
D2.2-5
(Deficiency)
Incorrect Information on Instrument Air
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Diagram
D2.2-6
(Deficiency)
10 CFR 50.59 Safety Evaluation Based Upon an
Incorrect Assumption and Analysis Methodology
A-1
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Item
Title
D3.1-1
(Deficiency)
Balance of Plant Design Specifications
D3.1-2
(Deficiency)
Design Temperatures for Safety-Related Piping
U3.1-3
(Unresolved)
Small Bore Pipe Support Spacing
03.1-4
(Observation)
Seismic Qualification of Valves Installed in
--
Class I Piping Systems
U3.2-1
(Unresolved)
MR-FC-84-61 Design Input Source and Use
D3.2-2
(Deficiency)
MR-FC-83-158 Installation Procedure
D3.2-3
(Deficiency)
MR-FC-84-162 Calculation
D3.2-4
(Deficiency)
Junction Box Supports
03.2-5
(Observation)
Containment Pressure Switch Seismic
Qualification
D3.2-6
(Deficiency)
Steam Generator Nozzle Dams
D3.2-7
(Deficiency)
YCV 1045B Valve Restraint
04.1-1
.(Observation)
High Power Rate of Change Trip Bypass
04.2-1
(Observation)
Delta T Power Loop Analysis
D4.3-1
(Deficiency)
Limit Switch Circuit Protection by Fusing,
MR-FC-84-74A
U4.3-2
(Unresolved)
ESF Bypass Switch Keylock Provision,
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MR-FC-81-102
04.3-3
(Observation)
Procurement Requirements on Equipment Vendors
U4.4-1
(Unresolved)
Design Basis Physical Separation Within Panels
D4.5-1
(Deficiency)
Drawing Changes by Procedure A-9, MR-FC-82-178
04.5-2 ,,(Observation)
Flow Element Design Basis Conditions
U4.5-3
(Unresolved)
Battery Room Fire Hazard Analysis
D5.1-1
(Deficiency)
Battery Sizing Calculation
US.1-2
(Unresolved)
Battery Charger /DC Bus Coordination
A-2
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Item
Title
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05.1-3
(Observation)
Power Cable Sizing Criteria
05.1-4
(Observation)
Pre-operational Test Requirements
05.1-5
(Observation)
Inverter Sizing without Analysis
05.1-6
(Observation)
Design Interface Control
.
D5.2-1
(Deficiency)
Fire Wrap Protection for Cable Raceways
.D6.1-1
(Deficiency)
Safety Evaluations for Non-Safety-Related
Systems Described in the USAR
'
U6.1-2
(Unresolved)
Safety Analyses for Emergency Modifications
06.1-3
(Observation)
Vital AC Inverter Bypass Mode
,
i
06.2-1
(Observation)
Untimely Closeout of Emergency Modifications
D6.2-2
(Deficiency)
Modifications to AFW Turbine Steam Supply Valves
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.
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.
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. _ _ _ _ __
..
.
-
.
D2.1-1 (Deficiency) Lack of Design Analysis To Support Sizing Of Air
Accumulators For Valves YCV 1045 A/B
DESCRIPTION:
Steam supply to the Auxiliary Feedwater turbine pump is upplied
from a steam header fed by two steam branch lines, one from each steam
generator.
The steam header is normally pressurized up to isolation valve
YCV-1045 through normally open isolation valves YCV-1045A and YCV-1045B in the
branch lines.
These isolation valves are pneumatically operated and can be
remote manually operated from the control room.
In the original system
-
design, valves YCV-1045 A and B were designed to fail open on loss of
instrument air, and valve YCV-1045 was designed to fail close.
A modification
,
request, MR-FC-78-43 (Reference 1), was initiated on an emergency basis in
1979 to redesign the valve operator for YCV-1045 and replace it with a fail
open operator.
This modification request was initiated in September, 1978
after the turbine pump failed to start during operability testing because of
4
inadvertent closure of an instrument air supply valve to YCV-1045 actuator
(LER-78-030).
To enable remote manual isolation in the event of a steam generator tube
rupture with a concomitant loss of non-safety-related instrument air, air
accumulators were to be added to the valve actuators for YCV-1045 A and B;
however, these accumulators were not installed prior to returning the plant to
power operation.
During closeout of FC-78-43, a new engineering evaluation
and assistance request, EEAR FC-83-158 (Reference 2), was initiated to install
the air accumulators.
In a January 15, 1985 memorandum (Reference 3), MR
FC-83-158 was scheduled for completion during the Fa11-1985 planned outage.
The Final Design Description (Reference 4) states that each accumulator will
be sized to provide air to the valve for one hour.
To assess the implementation
of the design process for modifications, the team reviewed the sizing
calculations.
Design analysis does not exist to confirm sizing of the air accumulators.
The
i
team found that a calculation does not exist which demonstrates that a
-
sufficient stored volume of pres'surized air will be available to close
YCV-1045 A and B assuming a loss of instrument air and minimum initial
'
accumulator. pressure.
The valve is spring actuated to open, and sufficient
air pressure must be provided to overcome spring pressure and approximately
.
1100 psi differential pressure across the globe valve during closure.
The
'
team was informed that a sizing calculation was not performed for this
modification. package.
The design engineer indicated that he referred to
,
calculations in a completed modification and used engineering judgement to
'
conclude that the current design was adequate.
The team found no
documentation of the engineering judgement and requested for review the sizing
calculations referred to by the design engineer.
These calculations were not
available during the inspection.
l
BASIS:
The licensee committed to implement ANSI N45.2.11 (Reference 8) for
design activities associated with modifications of safety-related structures,
systems, and components.
Contrary to the requirements of this standard, a
design analysis was not performed in a planned, controlled, and correct
manner. In addition, the design activity was not traceable from design input
through to design output.
A-4
.
.--,.-,___-__-e.---
- .
.
..
.
.
REFERENCES
1.
Document Control File fer MR FC-78-43, Failure Mode of YCV-1045.
2.
EEAR FC-83-158, Air Accumulators for YCV-1045 A/B, November 8, 1983.
3.
OPPD Memorandum TS-FC-85-42H, Review of Failure Mode Modification on
YCV-1045 A/B Steam Supply Valves to Steam-Driven AFW Pump, FW-10,
January 15, 1985.
.
4.
OPPD Final Design Description MR-FC-83-158, Air Accumulators for YCV-
1045 A/B CQE, Rev. O, February 14, 1985.
5. ' OPPD Generating Station Engineering Procedures Manual', Revision of
August 1985.
-
6.
Regulatory Guide 1.33, Quality Assurance Program Requirements (Opera-
' tion), Rev. 2, February 1978.
7.
ANS-3.2/N18.7, Administrative Controls and Quality Assurance for the
Operational Phase of Nuclear Power Plants Revision of N18.7-1972,
February 19, 1976.
8.
ANSI N45.2.11, Quality Assurance Requirements for the Design of Nuclear
Power Plants, 1974.
l
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.
D2.1-2 (Deficiency) Seismic Requirements Not Specified In MR-FC-83-158
!
DESCRIPTION:
The team examined the procurement documents for MR-FC-83-158 to
determine if appropriate requirements had been included.
The air accumulators and associated tubing and valves serve a post-accident
function to close YCV-1045 A and B.
These control valves are classified as
seismic Class I in accordance with Appendix F of the USAR; therefore, the air
-
accumulators and associated valves and tubing are considered seismic Class I.
The procurement specifications for isolation and check valves (References 2
and 3) do riot specify seismic requirements.
The team noted that third party
' design verifications (References 4 and 5) of these two specifications concluded
that the design inputs were correctly selected and incorporated into the
design.
BASIS:
Omaha Public Power District has committed to implement the guidance of
ANSI N45.2.11 (Reference 8).
ANSI N45.2.11 requires that the applicable
codes, standards and regulatory requirements be properly identified and
properly addressed.
Contrary to this requirement, the design verifier did not
ensure that the seismic requirements were included in the procurement
i
l
documents.
)
l
i
REFERENCES
!
1.
OPPD Critical Quality Elements (C.Q.E.) List, Rev. 2, May 24, 1985.
j
2.
OPPD Purchase Order No. 72533, SS-6C-10 NUPRO 3/8" Check and Relief
i
-
Valve, August 15, 1985.
l
3.
OPPD Purchase ~0rder No. 70650, SS-1KS6 WHITEY 3/8" Forged Body, Shut-
off Valves, July 19, 1985.
.'
'4.
OPPD Design Verification Checklist-Specifications for MR-FC-83-158,
I
Instrument Air Check Valve Specifications, September 19, 1985.
l
5.
OPPD Design Verification Checklist-Specifications for MR-FC-83-158,
.
Instrument Air Isolation Valve Specification . September 19, 1985.
6.
Regulatory Guide 1.33, Quality Assurance Program Requirements (Opera-
.
tion), Rev. 2, February 1978.
7.
.ANS-3.2/N18.7, Administrative Controls and Quality Assurance for the
Operat.ional Phase of Nuclear Power Plants Revision of N18.7-1972,
February 19, 1976.
8.
' ANSI N45.2.11, Quality Assurance Requirements for the Design of Nuclear
Power Plants, 1974.
.
.
A-6
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.
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.
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.
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.
.
02.1-3
(Observation) Vendor Exceptions to Specifications Not Reflected in
'
DESCRIPTION: Modification request, MR-FC-83-158, is a normal modification to
-
-
install an accumulator on each of two valves, YCV-1045 A'and B.
The Critical
Quality Elements List (Reference 1) states that instrument air system air
accumulators and associated piping and valves that su'pply' air to valves that
must function post-accident are considered critical quality elements.
Because
YCV-1045 A and B perform a post-accident safety function, the instrument air
accumulators and associated tubing and valves are also considered Critical
Quality Elements.
The team examined the procurement documents for MR-FC-83-158 to determine if
the vendor proposed equipment met or exceeded specifi~ cation ~ requirements.
-
The specification for Critical Quality Element for in'strument is'oTation valve's'
'
~
. req' ired that the storage be in compliance with ANSI N45.2.2, 'Le' vel-C an'd that
' ' ~~
~ u
~
'
'
the packaging be in accordance with vendor procedure WS-23 with engineer
'
~
approval required.
In a letter (Reference 2), the ve'ndor took exception to
!
the storage requirement stating that his valve supplier does not attempt to
conform to ANSI N45.2.2, Level C.
Despite this exception to the specification
.,
.
requirements, the purchase order (Reference 3) was issued to the vendor
indicating that packaging, shipping, storage and handling shall meet or exceed
i
the requirements of ANSI N45.2.2, Level C.
j
The team believes that documentation of the acceptability of the vendor
j
exception would enhance the conformance of the equipment with present procure-
j
ment document requirements.
REFERENCES
1.
OPPD Critical Quality Elements (CQE) List, Rev. 2, May 24, 1985.
2.
Omaha Valve & Fitting Company letter from S. Pendleton to H. Frazier
(OPPD), July 17, 1985.
'
3.
OPPD Purchase Order No. 70650, SS-1KS6 Whitey'3/8" Forged Body,
.
Shut-off Valves, July 19, 1985.
4.
OPPD Quality Assurance Plan 4.1, Procurement Process, Rev. O, September
r
1, 1984.
.
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.
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.
.
.
02.1-5 (Observation) Procedural Error caused Seismic And Stress Analysis For
MR-FC-83-158 Not To Be Filed In Modification File
DESCRIPTION:
The seismic and stress analysis (Reference 1) for the air
accumulators was prepared initially on August 12, 1985.
Since then the
analysis has been checked, third party reviewed, revised, rechecked, and was
ur.dergoing a second third party review during the design-inspection.
-
The team found no documentation associated with this calculation package in
-
the modification file, and it is uncertain that the appropriate documentation
would have been filed in the modification file because the modification file
number was not identified on the memorandum (Reference 2) routing the
calculation for final review.
Omaha Pubife Power District does not maintain calculations as living documents
.
(i.e., calculations are not maintained in a calculation file and kept current
,
as plant systems, structures, components are modified throughout the life of
the plant).
Instead, calculations are performed, as required, for each
modification.
As a consequence, the modification file is the only controlled
location for retention of design calculations.
The inclusion of design
calculations is left to the discretion cf the Design Engineer.
The team was informed that the subject calculation was a " generic" calculation
not applicable to a particular modification package.
As such the generic
calculation should have been referenced in the modification file.
The team
)
found that the control and use of generic calculations are not described in
the General Station Engineering procedures.
The team found that calculations
1
of this nature were the exception rather than .the rule.. For the modification
packages reviewed, the team found no references to generic calculations.
A better control process for generic calculations would allow enhanced retrieval,
use and revision of such calculations.
1
REFERENCES
.
1.
OPPD Calculation, Generic Air Accumulators using Propane Tanks Built to
DOT Spec. 4BA-240, Rev. O, August 12, 1985.
2.
OPPD M.emorandum from Department Manager to B. R. Livingston, Design
Review Generic Air Accumulator Calculations, September 10, 1985.
,
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.
.
d
D2.1-6 (Deficie.7cy) Failure to Follow Procedural Requirements For A Normal
Modification Resulting In Lack Of Required Design Verification Review
DESCRIPTION:
The Generating Station Engineering (GSE) Proce'dures Manual
describes the responsibilities of Generating Station Engineering personnel,
the types of modification requests, the information to be included in
p aparation of e modification package, and steps to document field changes and
closeout.
Three types of modifications are described.
These are normal,
emergency, and minor.
A minor modification does not involve any Critical
.
Quality Element (CQE) components.
A normal modification involves the
-
preparation of a preliminary design package (optional), a final design package
including third party review, and a construction package with third party
t
'
review.
For an emergency modification request, the same procedure is applied
- '
.
' exce~pt that certain approvals may be accomplished by telephone and the
~
-
'
. completion of the documentation may be accomplished following' completion of '
F
.
-
. f the modification.
For emergency and normal modifications ~, ~the preliminary
~
. design package is normally waived.
After an emergenc~ modification is ~
, ..
y
installed, preparation of an "after-the-fact" (ATF) final design ~ package and
~~
subsequent reviews in accordance with the normal modification are performed.
Modification request, MR-FC-83-158, is a normal modification to install an
,
!
accumulator on valves YCV-1045 A/B.
This modification was initiated in 1983
l
to correct partial completion of another modification accomplished on an
emergency basis in 1980. On February 19, 1985, the final design package
(Reference 1) for this modification was sent for third party review.
On
February 26, 1985, the third party reviewer completed his review and
i
determined that the final design package was not in compliance as documented
i
on a design document verification record (Reference 2).
On June 10, 1985, the
,
construction package (Reference 3) was sent to the Plant Manager for approval.
,
MR-FC-83-158 was not treated as required for a normal modification in
accordance with Design Procedure B-2 (Reference 4).
The team found that a
'
construction package was prepared even though the design verification of the
final des' n package had not been completed.
For a normal modification, the
.
team was
formed that the preparation of a construction package prior to
completion of the final design package is an accepted practice.
From
interviews, the team determined that it was not uncommon for design
l
verifications to be completed after normal modifications had been installed.
'
It appears that this practice is similar to that used for emergency
l
modifications.
This situation was further aggravated by the Design Engineer who made a
-
~
!
determination that the construction package did not require third party review
!.-
and who signed a memorandum (Reference 5) for the Dep~artment Manager stating
l,
th~at a third party review was not required.
!
BASIS:
Contrary to Generating Station Engineering Design Procedure B-2 Item
',
2.5.3, which states that after approval of the final design package, for
normal. modifications only, the Design Engineer will prepare the Construction
l
Package, a construction package for a normal modification was prepared and
completed prior to approval of the final design package.
A Construction
4
Package Design Verification was not performed, contrary to procedure item
2.7.3, which states that a design verification review was required if the
construction package involved the installation of Critical Quality Element
components.
i
A-9
,
l
l
'
l
...
.
-
.
.
REFERENCES
1.
OPPD Memorandum GSE-FC-85-66 (M7-C), Final Design Review of MR-FC-83-158
" Air Accumulators for YCV-1045 A/B", February 19, 1985.
2.
OPPD Design Document Verification Record / Routing Sheet, Project / Design
Modification FC-83-158, Routing No. 373, February 26, 1985.
3.
OPPD Memorandum GSE-FC-85-480 (M8), Construction Package for MR-FC-83-
158, June 10, 1985.
4.
GSE Design Procedure B-2, Production of Design Description and
Evaluation, Rev. 1/84.
5.
OPPD Memorandum from Department Manager to B. R. Livingston, Design
Review for MR-FC-83-158, June 11, 1985.
_
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_ _ _ _ _ _ _ - - _ __ _ - _ _ - - _ _
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.-
-
,
,
..
.
.
D2.1-7 (Deficiency) Incomplete Installation / Testing Procedure in Construction
.
.
Package for MR-FC-83-158
DESCRIPTION: Modification request MR-FC-83-158 is a n'ormal modification to
install an accumulator on each of two valves, YCV-1045 A and B.
These
instrument air accumulators were to be installed to permit tHF remote manual
)
isolation in the event of a steam generator tube rupture with a concomitant
i
loss of non-safety-related instrument air.
YCV-1045 A and B are normally
closed steam admission valves located in steam bra.nch lines feeding the
s
auxiliary feedwater turbine pump, and they fail open on loss of instrument
air.
These control valves are classified as seismic Class I and as Critical
Quality Elements (i.e., safety-related).
-
YCV-1045 A and B are 2-inch globe valves which may be required to shut against
.
.:.~~a diffe~rential pressure of approximately 1000 psig. 'The post-modification
"
'.
... testing procedure (Reference 1) does not test this design function.
During
.
the installation, YCV-1045 A and B are closed; therefore, the valves are
closed prior to commencing post-modification testing.
The first step of the
'-
~
test, Step 6.6, pressurizes the installation with normal instrument air supply
causing the actuator above the diaphragm to be filled.
Step 6.7 opens the
valve handwheels of. valves YCV-1045A and YCV-1045B; however, the valves remain
in the closed position because air has not been vented from above the
Step 6.8 directs that the installation be isolated from the normal
"
instrument air header using the root valve and that the actuator should be
monitored for one hour to ensure the valves remain shut with air supplied by the
accumulators alone.
As a consequence, the test procedure does not use the
pressurized volume of the accumulator to shut the valves. . In addition, no
testing adjustment is made to test the capability of the 2-inch globe valves to
shut against high differential pressures, nor is an acceptance criterion
-
provided for acceptable air leakage.
The only acceptance criterion is that the
valves must remain shut for one hour.
The team also noted that Step 6.8 requires the pressure of air in the
.
accumulator to be noted if the valve does not open.
However, there is no
'
pressure gauge on the accumulator or intervening piping.
BASIS:
Omaha Public Power District has committed to Regulatory Guide 1.33
(Reference 5) which endorses ANSI N18.7 (Reference 6).
This standard requires
!.
'
that modifications which affect functioning of safety related structures,
systems, or components be inspected and tested to coriffrm that the
~
'
modifications or changes reasonably produce expected Yesults and that~the
-
~
- ' change does not reduce safety of operations.
These t'est procedures are to
'
-
include appropriate quantitative or qualitative acceptance criteria for
'
determining that important activities have been satisfactorily accomplished.
Contrary to these requirements, the test procedure would not have confirmed
that the modification produced expected results and did not have an acceptance
criterion for acceptable air leakage.
REFERENCES
1.
OPPD Installation Procedure MR-FC-83-158, Air Accumulators for YCV-1045
A/B, Rev. O, June 10, 1985.
2.
OPPD Checker's Checklist - Construction Package MR-FC-83-158, Air
Accumulators for YCV-1045 A/B, June 10, 1985.
A-11
.
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. , _ , , , , - - ~ - - - -
-
..
,
-
.
,
A.
OPPD Memorandum from Department Manager to B. R. Livingston, Design
Review for MR-FC-83-158, June 11, 1985.
4.
OPPD Design Verification Checklist - Work Instructions and Test
Procedures, MR-FC-83-158 Air Accumulators for YCV-1045 A/B, February
26, 1985.
5.
Regulatory Guide 1.33, Quality Assurance Program Requirements
(Operation), Rev. 2, February 1978.
_
6.
ANSI N18.7/ANS 3.2, Administrative Controls and Quality Assurance for
the Operational Phase of Nuclear Power Plants, February 19, 1976.
.
9
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-
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.
.
~
D2.1-8.(Deficiency) Incorrect Information On Flow Diagram For Main Steam
.
System
.
. DESCRIPTION:
During the course of the inspection, the team reviewed design
~
.
aspects of various modifications with respect to the'information contained on
'
system flow diagrams.
The following inconsistencies and errors were
identified during the team's review of the steam system flow diagram.
~
System Description III-2 (Reference 1), steam system description, states that
-
the main steam isolation bypass valves, HCV-1041C and HCV-1042C, are
horizontally mounted, motor-operated, 2-inch globe valves.
These valves are
piped into the valve body of their respective main steam isolation valves.
Review of the valve drawings (Reference 2) shows that steam' pass'es through the-
~
.
_:
T bypass line upsteam of the disc associated with the niain steam isolation
'
~
.
. valves, f. lows through HCV-1041C and back into the valve body of the main steam
-
- - -
.
. isolation valve downstream of the valve's disc, HCV-10410'is' cracked'open~and'
_ ~
' ' -
3
.
steam. flows through the non-return valve (HCV-1041B)/ When pre'ssdres and
-
'
-
temperatures have equalized, the main steam isolation #1ve is opened.
The
L
' main steam isolation valves and the non-return valves are within the Class 2
boundary.
4
s
-
Flow Diagram 11405-M-252 (Reference 3) incorrectly represents the piping
4
arrangement associated with the bypass valves and the auxiliary feedwater
steam warmup lines.
The drawing indicates that the piping to the bypass
,
valves taps off the upstream side of the disc and returns to the upstream
side, versus the correct return to the area between the main steam isolation
valve and its associated reverse flow check valve.
In addition.the piping
connected downstream of the bypass valve is indicated as non-safety by a flag.
This piping also supplies the warmup line for the auxiliary feedwater steam
headers, which is also incorrectly indicated as non-safety.
These lines tap
off the downstream side of the bypass valves and are piped to the downstream
side of YCV-1045 A and B through normally open isolation valves MS-336 or
MS-337.
The portion of the piping from the main steam ~ isolation bypass
i
isolation valves to either MS-336 or MS-337 and the associated branch line to -
l
the main steam isolation valve body is incorrectly depicted as non-safety.
The team noted that Omaha Public Power District's Critical Quality Elements
(CQE) List (Reference 4) is a system level Q-List which relies in part on the
correct classification on system flow diagrams.
~ _ ' ;During the inspection, Omaha Public Power District acknowledged the error and
_
committed to correct the flow diagram.
C"
"
-
BASIS:
Omaha Public Power District committed to Regulatory Guide 1.64
(Reference 5) which endorses ANSI N45.2.11 (Reference 6).
This standard
' requires that personnelause proper and current drawings and design inputs.
Contrary to this requirement, the steam system flow diagram was not correct or
current with the as-installed arrangement in the plant.
REFERENCES
1.
OPPD System Description III-2, Steam System, Revised August 16, 1984.
2.
Schutte & Koerting Company Drawing 69-XC-20, Assembly Drawing 28X24-600**
Main Steam Trip Valve, Rev. 8, March 5, 1985.
A-13
-
-
. - - - - - --. . . - - - -
- - -
- - - - -
- .
- -
.
.
.
.
-
-
3.
OPPD Drawing Numbsr 11405-M-252, Flow Diagram Steam, Rev. 33, June 21,
1984.
4.
OPPD Critical Quality Elements (CQE) List, Rev. 2, May 24, 1985.
5.
Regulatory Guide 1.64, Quality Assurance Requirements for the Design
of Nuclear Power Plants, Rev. 2, June 1976.
6.
ANSI N45.2.11, Quality Assurance Requirements for the Design of Nuclear
Power Plants, 1974.
.
.
~
.
A-14
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.
.
D2.1-9 (Deficiency) Incorrect System Description Statements
DESCRIPTION:
During the team's review of various modification packages, Omaha
Public Power District's system descriptions were examined to confirm system
design bases.
The following errors or inconsistencies were identified in the system
descriptions reviewed.
Auxiliary Feedwater System Description III-4 (Reference 1) states
~
a.
that steam admission valves, YCV-1045 A and B, each have 3/8 inch
unvalved bypass lines that serve to maintain warm steam lines up to
.FW-10 isolation valve YCV-1045.
Contrary t'o this description, the
"
_
' team determined that the source of warming steam to the ~ auxiliary -
.
feedwater turbine line downstream of YCV-1045 A and B passes through - -
'
' -
.. .
.normally open manual valves MS-336 and MS-337, respectiVely.
These: -
.
--
valves are shown on Steam System Flow Diagram (Reference'2) sand:
'
related physical piping diagrams.
-
b.
Modification MR-FC-21B changed valves HCV-438B and HCV-438D (outside
containment component cooling water isolation valves for the reactor
,
'
- ' coolant pump lube oil coolers and seals) from failed closed to fail
open and added air accumulators to permit the valve operator to keep
the valves closed until action could be taken to isolate the lines
manually.
The Station System Acceptance form (Reference 3) for this
modification indicates that the system description had been updated.
Contrary to this indication, Compressed Air System Description
III-10 (Reference 4) does not include valves HCV-438 B and
D on the list of valves equipped with instrument air accumulators.
l
The team noted that site acceptance of this modification was
completed in May 1983 and that the Compressed Air System
,
Description was most recently revised in April 1985.
'
Modification MR-FC-81-21 developed a component cooling water
,
c.
pressure low signal and added it to the control circuits for valves
HCV-438 B and D, such that these valves remain open except when a
containment isolation signal and a component cooling water pressure
low signal are simultaneously present.
The Station System
Acceptance form indicates that the system description had been
updated.
Contrary to this indication, Component Cooling Water
- - System Description I-7 (Reference 5) omits the low pres'sure-signal'
'
,
.
.g3
-
and states that "CIAS closes containment isolation valves
'
,
HCV-483A/B/C/D, thus isolating CCW flow to the reactor coolant
L
-
pumps.
The team noted that site acceptance for this modification
"
'
was completed in May 1983 and that the system description had not
-
been updated since December 1981.
=
The team determined that system descriptions are maintained by site -
engineering personnel, and during a site inspection the team was informed by
site engineering personnel that the descriptions had numerous errors.
BASIS:
Omaha Public Power District's Quality Assurance Plan 5.1 (Reference 6)
l
requires that those organizations participating in activities affecting safety
'
shall be made aware of, and use, proper and current instructions, procedures,
drawings, and engineering requirements for performing the activity.
Contrary
l
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.
r
,
,
to this requirement, design descriptions, available for use as design input,
were incorrect or not updated following completion of modifications.
REFERENCES
1.
OPPD System Description III-4, Auxiliary Feedwater System, Rev. 5,
August 16, 1984.
2.
OPPD Drawing 11405-M-252, Flow Diagram Steam, Rev. 33, June 21, 1984.
-
'
3.
OPPD Station System Acceptance form J for MR-FC-81-21B, CCW Isolation
to RCP's, May 4, 1983.
4.
OPPD System Description III-10, Compressed Air, Rev. 5, April 10, 1985.
5.
OPPD System Description I-7, Component Cooling Water System, Rev. 3,
December 3, 1981.
6.
OPPO Quality Assurance Plan 5.1, Control of Plant Design and Modifica-
tions, Rev. O.
.
e
e
O
O
O
A-16
_
.
..
.
.
.
U2.1-10 (Unresolved Item) Use of Fluorocarbon-Elastomer Material in High
Radiation Environments
-
DESCRIPTION:
During the team's examination of modification packages, the
~
.
material compatibility with expected environments was' ' reviewed for
modification packages MR-FC-84-144 and MR-FC--83-158.
~
The first modification package involves the replacement of solenoid valves for
YCV-1045 A and B.
The Final Design package (Reference 1) states that the
-
existing solenoid valves are acceptable for the application but have Viton
material as elastomer seals.
Viton is an E.I. duPont de Nemours trade name.
It is often described as a fluorocarbon elastomer and has the chemical
designation as vinylidene fluoride and hexafluoro propylene.
The Final Design
.
.
package. indicated that Viton is not recommended for applicatio'n in rauiation
~
'
_:.:.
areas therefore, all solenoid valves containing this' material are to be
~
. E
- ..' remove;d from service.
Likewise, the Engineering Evaluation and~ Assistance Request:
'~
"' '...~.:.(Reference 2)' indicates that the solenoid seals should b'e' changed to 'av'ofd
~
~
~
- stocking.Viton.
Omaha Public Power District's Technical-Services organization
~
'
concurred with the modification (Reference 3), indicating this modification
will ensure no Viton parts are stored, since these parts are not to be
installed in a radiation area.
Modification package MR-FC-83-158 is a normal modification to install air
accumulators on the same valves YCV-1045 A and B, auxiliary feedwater turbine
steam admission valves.
These instrument air accumulators were to be
installed to permit the remote manual isolation of a steam generator in the
event of a tube rupture with a concomitant loss of non-safety-related
instrument air.
This modification includes the installation of instrument air
check valves to isolate the safety-related instrument air accumulators from
the non-safety-related instrument air headers.
The team examined the
procurement specifications for MR-FC-83-158 and determined that Viton was
being used as a seating material in the safety-related instrument air check
valve.
The procurement specification (Reference 4) for the safety-related instrument '
air check valves permits the use of'Viton as a seat material.
The team noted
that the original specification (Reference 5) specified Buna "N" as a seat
material; however, the valve's supplier took exception to this seat material
and stated in a letter (Reference 6) that Viton would be supplied instead.
Based upon this exception the specification was revised to include Viton as an
' acceptable seat material.
CC2
~
During the inspection, the licensee was unable to explain the disparity
.between the two modifications except to indicate that the radiation dose'at
~
'
' the location of the solenoids and the safety-related check valves was
sufficiently low that Viton would be an acceptable seating' material.
The
licensee pointed out that the procurement specification for the instrument air
check valves was prepared besed upon a specification previously used in
another application.
The licensee stated that environmental conditions were
not revised downward because the values specified were conservative.
Therefore, the specification identifies the radiation environment as 3.0 E6
rads even though the expected condition is apparently lower.
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.
During the insp'ction, the team did not determine if Viton had been used in
e
other instrument air applications or other safety-related applications.
The
team requested a listing of all modifications which installed air
In response to this request and at the end of the inspection,
the licensee produced a short list of possible modifications apparently
generated upon the recollection of various engineers.
Independently, the team
identified a modification where air accumulators and instrument air check
valves had been added; however, the team could not determine the seat material
used in the check valves because the procurement specification was not
-
included in the modification file (See Deficiency D2.2-4).
. BASIS:
Criterion III of 10CFR50 Appendix B (Reference 7) requires that design
control measures be applied to insure compatibility of materials.
Omaha
Public Power District has committed to Regulatory' Guide 1.64 (Reference 8)
which endorses ANSI N45.2.11 (Reference 9).
The standard requires verifiers
to confirm that the specified parts are suitable for the required application
and that specified materials are compatible with the design environmental
conditions to which the material will be exposed.
Contrary to these
requirements, an unacceptable material may have been used in a high radiation
environment.
REFERENCES
1.
OPPD Final Design for MR-FC-84-144, Replacement of the Solenoid Valves
for YCV-1045A and YCV-1045B, Rev. O, March 20, 1985.
2.
OPPD Engineering Evaluation and Assistance Request Nc. FC-84-144,
Upgrade of YCV-1045A Solenoid, August 23, 1984.
-
-
3.
OPPD Technical Services Review and Evaluation for Modification Request
No. FC-84-144, Upgrade of YCV-1045A Solenoid, December 17, 1984.
4.
OPPD Instrument Air Check Valve Specification.for MR No. FC-83-158, Rev.
1, August 15, 1985.
5.
OPPD Instrument Air Check Valve Specification for MR No. FC-83-158, Rev.
O, May 24, 1985.
6.
Omaha Valve & Fitting Company letter from S. Pendleton to H. Frazier
(0 PPD), August 13, 1985
t
7.
10 CFR 50 Appendix B, Quality Assurance Criteria for Nuclear Power Plants
and Fuel Reprocessing Plants
8.
Regulatory Guide 1.64, Quality Assurance Requirements for the Design of
Nuclear Power Plants, Rev. 2, June 1976.
9.
ANSI N45.2.11, Quality Assurance Requirements for the Design of Nuclear
Power Plants, 1974.
A-18
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.
D2.2-1 (Daficiency) Incorrect Design Input In Calculation Associated With
MR-FC-81-21B
. DESCRIPTION:
Modification MR-FC-81-21B is a completed modification which-
~
replaced fail close pneumatic actuators with actuators that fail open.
,
The replacement actuators were installed on valves HCV-438B and HCV-438D.
These valves are containment isolation valves located outside containment in
the component cooling water supply and return lines associated with the
reactor coolant pump lube oil coolers and seals.
This completed modification
added instrument air accumulators to these valves to permit the valve operator
_
to maintain the valves closed until operator action could be taken to manually
close the valves.
Because this modification is similar to a modification
I
planned for the 1985 outage (i.e., MR-FC-83-158), the team reviewed
~
modification MR-FC-81-21B to determine if errors and ' discrepancies found
.
during the review of modification MR-FC-83-158 were systematic.
'
-
The modification file contained a calculation sheet (Refe'rence 1) which
~
' concluded that the air accumulator had sufficient volume.
This calculation ~
was reviewed and the following discrepancies noted:
-
The calculation states that the accumulator is a 20 pound propane tank
a.
o
i
and that the volume is 4423 cubic inches.
The source of this information
'
is not referenced.
A 20 po'und propane tank, typically, has a volume of
j
approximately 1320 cubic inches.
The team believes that the volume of
stored air used in the calculation is overestimated by 335 percent.
b.
The calculation assumes that the air pressure is at 100 psig.
The
reference for this assumption and justification for its use is not
documented.
The instrument air system pressure will range between 80 and
.
'100 psig per the compressed air system description (Reference 2).
The
'
assumption that the air accumulator is fully charged at maximum
instrument air pressure is not conservative and inappropriate for an air
accumulator sizing calculation.
!
c.
The calculation does not consider system leakage or the period of
^
time that the valve must remain shut.
The valves operated by these
accumulators are fail open containment isolation valves.
In the event of
a need-to close these valves, they would have to remain shut for the
duration of the accident or until operator action is taken to manually
shut the valve.
The implicit assumption of zero leakage is not
_
. conservative and unrealistic.
The team noted that the air ai:cumulator
.
.
1
~ . installation was not properly tested after modification (See Deficiency
~~
D2.2-3) and that surveillance testing is not performed to demonstrate the
. capability of the Critical Quality Element (i.e., safety-related) portion
~ of the instrument air system to close these valves and maintain them
closed for a predetermined period of time without loss of function.
d.
The calculation sheet is not signed by a checker.
Instead reference
,
!
is made to see a B-2-2 Form.
However, the B-2-2 form is not attached or
included in the modification file.
I
i
BASIS:
Omaha Public Power District committed to Regulatory Guide 1.64
l
(Reference 3) which endorses ANSI N45.2.11 (Reference 4).
This standard
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r quires .that calculations includa dnfin:d objectives, identification of
design inputs and their sources, and documentation of assumptions and
identification of those assumptions which need confirmation at a later date.
Contrary to these requirements, the calculation contained incorrect and
inappropriate assumptions without identification of their sources or
justification for their use.
4
REFERENCES
_
1.
OPPD Generating Station Calculation Sheet for Modification File
MR-FC-81-21B, Accumulator Tanks, March 26, 1983.
2.
OPPD System Description III-10, Compressed Air, Rev. 5, April 10, 1985.
3.
Regulatory Guide 1.64, Quality Assurance Requirements for the Design of
Nuclear Power Plants, Rev. 2, June 1976.
4.
ANSI N45.2.11, Quality Assurance Requirements for the Design of Nuclear
Power Plants, 1974.
.
l
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i
.
!
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_
D2.2-2 (Deficiency) Incomplete Consideration Of CQE And Seismic Class I
Requirements For Portions of MR-FC-81-21B
_
DESCRIPTION:
Modification MR-FC-81-21B is a completed modification which
replaced fail close pneumatic actuators with actuators that fail open.
the
replacement actuators were installed on valves HCV-438B and HCV-4380.
These
valves are containment isolation valves located outside containment in the
component cooling water supply and return lines associated with the reactor
coolant pump lube oil coolers and seals.
This completed modification added
-
instrument air accumulators to these valves to permit the operator to maintain
the valves closed until operator action could be taken locally to manually
~
close the valves. This modification was completed in 1983 and was similar to
.
a modification planned for the 1985 outage (i.e., MR-FC-83-158).
,
.In.USAR Appendix F, the component cooling water system is classified as a
',
~
seismic Class I system.
In the Critical Quality Elements' List, air-
,
accumulators and associated piping and valves that supply ~ air to valves that
must function following an accident are identified Critical-Quality Elements
-
'
based upon the classification and operating requirements of the valves that
they supply.
HCV-4388 and D are valves which are open following an accident
and must have the capability to be closed throughout the course of the
accident.
As a consequence, the air accumulators and associated piping and
valves are seismic Class I and Critical Quality Elements because they must
remain functional during and following an accident to shut HCV-438B and D.
,
The team reviewed the seismic qualification of components installed during
this modification.
The team found that seismic requirements were not properly addressed in the
modification package.
The following discrepancies were identified:
!
it.
Purchase Order No. 56600 (Reference 1) was issued to an engineering
organization to confirm that the valve and operator assembly
supplied by a manufacturer was seismically qualified without
invoking the requirements of 10 CFR Part 50 Appendix B.
In
.
addition, the purchase order did not invoke the requirements of 10 CFR Part 21 and was not identified as applicable to critical quality
elements.
As a consequence, the engineering organization did not
complete the computer analysis in accordance with their Qualit/
Assurance Manual and identified this to Omaha Public Power Dittrict
' :
in an August 1983 letter (Reference 2).
Because the procurerent was
. . . .
.
not' considered to involve services for a Critical Quality' Element, a
"
'
Quality Assurance representative did not review the purchase order.
b.
The installation / test procecure (Reference 3) did not reference Fort
'
Calhoun criteria (Reference 4) for routing and support of seismic
instrument tubing.
No calculation existed at the time the modification was completed to
c.
confirm that the as-constructed air accumulator, including base plate
and Hilti bolts, was adequately sized to withstand expected seismic
loadings.
A 1983 calculation (Reference 5) in the completed
modification file does not address seismic considerations.
A
subsequent generic analysis (Reference 6) performed in 1985 appears to
A-21
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,
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.
confirm the configuration is adequate; however, confirmation is required
to verify that the installed configuration is the same or is bounded by
that analyzed.
During the site visit the team conversed with a site engineer regarding
seismic requirements for the instrument air system.
The engineer, who stated
he was responsible for installation of air accumulators, erroneously stated
that no portion of the instrument air system was required following an
accident, and therfore that there was no need for seismic installation.
BASIS:
Contrary to 10 CFR 50 Appendix B Criterion IV, the licensee did not assure
that applicable regulatory requirements, design bases, and other requirements
which are necessary to assure adequate quality are suitably included or
-
referenced in documents for procurement of services.
Omaha Public Power District committed to Regulatory Guide 1.33 (Reference 7)
which endorses ANSI N18.7 (Reference 8) for quality assurance program
requirements for operating reactors.
This standard requires that each
procedure contain instructions in the degree necessary for performing a
.
required task by a qualified individual without direct supervision and that
-
they contain appropriate references.
Contrary to these requirements the
procedure did not address the installation requirements for seismic tubing.
Omaha Public Power District has committed to implement ANSI N45.2.11
(Reference 9) for design activities associated with modification of
safety-related structures, systems and components.
Contrary to the
requirements of this standard, a design analysis was not performed in a
planned, controlled, and correct manner.
REFERENCES
1.
OPPD Purchase Order No. 56600 to Stevenson & Associates, CERTIVALVE
Computer Program Analysis for HCV-438B, May 11, 1983.
-
2.
Stevenson & Associates letter No. 83C2220 from W. Djordjevic to
W. Weber (OPPD), CERTIVALVE Analysis of Fisher Valve HCV-4388 for
Ft. Ca.1houn Station, August 5, 1983.
3.
OPPD Revised Design Description Appendix 7.4 for MR-FC-81-21B,
Installation Procedure for HCV-438B/D Air Accumulators, Rev. 2,
March 26, 1983.
4.
Stone & Webster Report J.0. No. 13007.65, Guideline for the
Installation of Tubing and Tubing Supports for Seismic Instrument
Systems, March 1982.
5.
OPPD Generating Station Calculation Sheet for Modification File
MR-FC-81-21B, Accumulator Tanks, March 26, 1983.
.
6.
OPPD Calculation, Generic Air Accumulator Calculations, Rev. 1,
September 10, 1985.
7.
Regulatory Guide 1~33, Quality Assurance Program Requirements (Oper-
.
ation), Rev. 2, February 1978.
8.
ANSI N18.7/ANS 3.2, Administrative Controls and Quality Assurance for
the Operational Phase of Nuclear Power Plants, February 19, 1976.
9.
ANSI N45.2.11, Quality Assurance Requirements for the Design of Nuclear
Power Plants, 1974.
A-22
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. D2 g..3-(Deficiency) Incomplete Installation / Testing Procedure Performed For
e
.
~MR-FC-81-21B -
t
DESCRIPTION:
Modification, MR-FC-81-21B is a completed modification which
'
replaced fail close pneumatic actuators with actuators that fail open.
The
. replacement actuators were installed on valves HCV-438B and HCV-4380.
These
valves are containment isolation valves located outside containment in the
component cooling water supply and return lines associated with the reactor
coolant pump lube oil coolers and seals.
This completed modification added
-
instrument air accumulators to these valves to permit the valve operator to
maintain the valves closed until operator action could be taken locally to
manually close the valves. This modification was completed in 1983 and was
- similar to a modification planned for the 1985 outage (i.e., MR-FC-83-158).
The~ team reviewed the post-modification testing accomplished in view of the
team's concerns expressed in Deficiency D2.1-7.
~
-
..
. .The post-modification test procedure (Reference 1) did not Vequi'r~e' the u'se'of
'
the pressurized volume of the accumulator to shut the valves.
The
-
installation and test procedure closed HCV-438B and D, then isolated air from
the instrument air header by closing valves IA-174 and IA-175.
In this
configuration, only a. static test was conducted.
The test procedure did not
require the use of the pressurized volume of the accumulator to shut the
valves. The acceptance criteria was to ensure that the valves remained shut
'
for twenty minutes; however, the team found no documented basis that twenty
minutes was a sufficient period of time to identify the need to manually close
these valves and to physically have a plant operator perform the required
action locally at the valves.
BASIS:. Omaha Public Power District has committed to Regulatory Guide 1.33
(Reference 2) which endorses ANSI N18.7 (Reference 3).
This standard requires
that modifications which affect functioning of safety-related structures,
systems, or components be inspected and tested to confirm that the
modifications or changes reasonably produce expected results and the change
does not reduce safety of operations.
These test procedures are to include
.
appropriate quantitative or qualitative acceptance criteria for determining
that important activities have been satisfactorily accomplished.
Contrary to
these requirements, the test procedure would not have confirmed that the
modification produced expected results and did not have acceptance criteria
for acceptable air leakage.
REFERENCES
P~~
~
~
1.
OPPD Setpoint/ Procedure Change No. 10320 for EEAR FC-81-21B/SRDC0
83-27.
Appendix to EEAR FC-81-21B/SRDC0 83-27 Test Procedure for
Accumulators, March 28, 1983.
2.
Regulatory Guide 1.33, Quality Assurance Program Requirements
(Operation), Rev. 2, February 1978.
3.
ANSI N18.7/ANS 3.2, Administrative Controls and Quality Assurance for
the Operational Phase of Nuclear Power Plants, February 19, 1976.
A-23
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.
.
02.2-4 (Observation) Incomplete Modification File for a Completed Modification
DESCRIPTION:
Modification MR-FC-81-21 is a completed modification which
replaced fail close pneumatic actuators with actuators that fail open.
The
replacement actuators were installed on valves HCV-438B and HCV-4380.
These
valves are containment isolation valves located outside containment in the
component cooling water supply and return lines associated with the reactor
coolant pump lube oil coolers and seals.
This completed modification added
instrument air accumulators to those valves to permit the valve operator to
maintain the valves closed until operator action could be taken to manually
close the valves.
Because this modification is similar to a modification
planned for the 1985 outage (i.e., MR-FC-83-158), the team reviewed
modification MR-FC-81-21B to determine if errors and discrepancies found
during the review of modification MR-FC-83-158 were systematic.
In reviewing this completed modification, the team identified information
missing from the modification file as follows:
a.
Although a Generating Station Engineering Calculation Sheet
(Reference 1) was included in the modification file, there was no
record of a third party review or checking of the calculation.
The
team did not find a verification checklist demonstrating a
third party review or a completed Form B-2-2 documenting a checker's
review.
b.
Although instrument air check valves were procured, no indication
exists of a procurement specification or of a third party
verification.
Improved implementation of the licensee's document control methods should
preclude incomplete files and enhance the design control process.
REFERENCES
.
1.
OPPD Generating Station Calculation Sheet from Modification File
MR-FC-81-21B,4ccumulatorTanks, March 26, 1983.
2.
Regulatory Guide 1.64, Quality Assurance Requirements for the Design
of Nuclear Power Plants, Rev. 2, June 1976.
3.
ANSI N45.2.11, Quality Assurance Requirements for the Design of
Nuclear Power Plants, 1974.
,
,
4.
GSE Administrative Procedure A-9, Document Control, Revised August
1983.
5.
GSE Administrative Procedure A-2, Modification Request Development,
Revised January 1984.
6.
GSE Design Procedure B-11, Design Verifications, Revised April 1982.
A-24
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..
.
..
,
.D2.2-5 (Deficiency) Incorrect Information On Instrument Air Diagram
DESCRIPTION:
The team reviewed the portions of the instrument air system
while evaluating post-modification testing requirements for planned
.
.
modification MR-FC-83-158 and completed modification MR-FC-81-218.
Ins $ument air header isolation valves, IA-175 and IA-176, were used during
the installation and testing of modification MR-FC-81-21B.
However, these
valves do not appear on Omaha Public Power District drawing 11405-M-264
(Reference 1).
It appears that these valves were overlooked when preparing
the drawing or incorrectly deleted.
During a field inspection the team confirmed that the valves are installed in
the plant.
.
BASIS:
Omaha Public Power District has committed to Regulatory Guide 1.64
~2
(Refer'ence 2) which endorses ANSI N45.2.11 (Reference 3).
This standard
requires that documents, including changes, be reviewed for adequacy and
approved for release by authorized personnel.
Contrary to this requirement, a
document was released which did not depict the as-installed piping / valving
arrangement in the plant.
~
REFERENCES
1.
OPPD Drawing 11405-M-264, Instrument Air Diagram Riser Details, Rev. 4,
June 27, 1984.
2.
Regulatory Guide 1.64, Quality Assurance Requirements for the Design
of Nuclear Power Plants, Rev. 2,' June 1976.
3.
ANSI N45.2.11, Quality Assurance Requirements for the Design of
Nuclear Power Plants, 1974.
.
.
5
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D2.2-6 (Deficiency) 10CFR 50.59 Safety Evaluation Based Upon An Incorrect
Assumption and Analysis Methodology
DESCRIPTION:
Modification MR-FC-81-218 is a completed modification which
replaced fail close pneumatic actuators with actuators that fail open.
The
replacement actuators were installed on valves HCV-4388 and HCV-4380.
These
valves are containment isolation valves located outside containment in the
component cooling water supply and return lines associated with the reactor
coolant pump lube oil coolers and seals.
This completed modification added
instrument air accumulators to these valves to permit the valve operator to
maintain the valves closed until operator action could be taken locally to
manually close the valves.
In addition, the modification added a component
cooling water pressure low signal in series with a containment isolation
actuation signal such that the presence of both signals is necessary to close
the valves.
Asaresugtofthismodification,thepost-LOCAheatloadwas
increased by 3.15 x 10 BTU / hour which corresponds to the heat load from the
reactor coolant pump seal and lube oil coolers.
A safety evaluation (Reference 1) was included in the Final Design Description
(Reference 2) for the modification.
This evaluation concluded that (a) the
modification would not increase the probability of an occurrence or the
consequences of an accident. from the analysis previously done in Volume 4,
Section 9.7 of the Fort Calhoun USAR, (b) the modification would not create
the possibility of an accident or malfunction other than those analyzed in
Volume 4, Section 9.7 of the Fort Calhoun USAR, and (c) the modification would
not reduce the margin of safety as defined in the basis for technical
specifications since this is not a basis for a technical specification.
This modification was completed and site accepted (Reference 3) in May of
1983.
Based upon the team's review of this safety analysis the following
conclusions were made:
The safety analysis performed by an Omaha Public Power District Design
o
Engineer did not refer to original design calculations.
The lack of
.
original design analyses or their unavailability did not result in the
performance of new calculations, instead the Design Engineer used a
qualitative argument based upon USAR statements.
The qualitative argument used by the Omaha Public Power District
o
Design Engineer does not reflect a correct understanding of the heat
.
transfer phenomenon between heat removal systems.
Specifically, the team
found that the qualitative argument implicitly assumed that the designed
heat removal capacities of equipment coolers and heat exchangers are
independent of each other and therefore can be added and subtracted to
determine heat removal capacity between systems.
.
The safety evaluation contains an unsubstantiated and inappropriate
o
assumption concerning operator action to secure heat loads under certain
accident conditions.
o The basis of technical specification 2.4 contains incorrect informa-
tion concerning the heat removal capacity of the component cooling water
heat exchangers.
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Although the Design Engineer stated that the margin of safety as defined in
.. '
~ the basis for a technical specification was not reduced, it appears that he
j .
recognized that the basis of technical specification 2.4 contained information
.
which was incorrect and required revision regardless of the proposed
!
. modification.
This is evident by comments made in the safety analysis of the
-
1
Final Design Description.
The basis of technical specification states that
4
threecomponentcoolingheagexchangershavesufficientcapacity(withample
reserve) to rrmove 420 x 10 BTU / hour following a loss-of-coolant accident.
However,thesafetyanalygisinthemodificationfileindicatesthattheheat
4
removal value of 420 x 10 BTU /hourcorrespondstoghecapacityofthe
containment air coolers (two units rated at 70 x 10
6
BTU / hour / unit) and
cooler / filters (two units rated at 140 x 10 BTU / hour / unit), not the component
1 .
.. cooling water heat exchangers.
' removal capacity of the component cooling water system is 402 x 10Thesafetyeval
-~
!
BTU / hour
p fassuming three of four heat exchangers and two of three pumps are available.
- ,-
'
6
~
. ..
To address the addition of 3.15 x 10 BTU / hour from the reactor coolant pump s'eal
-
'. .: '
'
' nd lube oil coolers and the apparent existing error in the basis of the
a
technical specification, the design engineer presented the following rationale
,
forconegudingnosafetyimpact.
First, he noted that the USAR states that
280 x 10 BTU / hour of heat removal capacity is assumed in the containment
pressure and temperature analysis and that both the containment spray system
j
and the containment air cooling system are each designed to remove heat in
)
excess of this value during post-LOCA conditions.
Second, he states that the
i
containment spray system is independent of the containment air cooler system
4
(i.e., as long as either system is available the containment heat removal
function will be satisfied).
Third, he eliminates from further consideration
heat loads from the containment spray system to the component cooling water
system by noting that component cooling water is only supplied to the shutdown
.'
cooling heat exchangers to remove containment spray system heat loads upon
receipt of a recirculation actuation signal which occurs later in the
i
- accident, when air cooling loads are significantly reduced.
Based upon the
l
foregoing, the design engineer appears to have concluded that the post-LOCA
)
. heat load see'n by the component cooling water system can be as high as 425 x 106
BTU /hourimmediatelyfollowinganagcidentifallcontainmentaircooling
6
coils perform as designed (425 x 10 E6 BTU / hour based up n the sum of 420 x 10
6
BTU /hourfromcontainmentaircoolingcoils,3.g5x10 BTU / hour from the reactor
coolant pump seal gnd lube oil coolers, 0.3010 BTU / hour from charging pump
coolers, 1.05 x 10 BTg/hourfromthesafetyinjectionandcontainmentspraypump
..
coolers, and 0.30 x 10 BTU / hour from the control room air conditioning).-
l' . ' Because this value exceeds the designed heat removal capacity of the component
'
~
!
.
. cooling water heat exchangers, the design engineer assumed in the safety
j.
evaluation that if all containment air cooling system units operated as
6
!
. de' signed that the operator would select one cooling unit rated at 70 g 10
!
'
BTU / hour and isolate it to reduce the post-LOCA heat load to 355 x 10 BTU / hour.
.
!
This value is then within the heat removal capacity of the component cooling
water system.
Based upon the documentation in the modification file and the lack of any
t
reference to design analysis in the safety analysis, it appears that no
,
i
comparison was made to original design calculations.
The team determined that
design calculations are not controlled by Omaha Public Power District as
i
living design documents but are filed when performed with the modification
package.
The team was informed that original calculations performed by the
architect-engineer during construction may or may not be available because the
A-27
.
- ' .
.
_,
.
'
..
..
-
- - _ _
.--
..
.
-
.
original design was performed in the late 1960's and that the information that
is available is in storage.
The team found that some original architect-
engineering information was located in files within Generating Station
Engineering's document control area and some information was located outside
the building in commercial storage.
However, the team determined that these
files were not organized into a workable source of original design information
for assessing the original design basis.
As a consequence, the team found
that this information was not generally used by design engineers working on
modification packages.
It appears that the Omaha Public Power Design Engineer
resorted to using information contained in the USAR without confirming its
accuracy and also used qualitative judgements to conclude that the
modification was adequate.
To implement this safety evaluation, the Design Engineer prepared proposed
revisions to the basis of technical specification 2.4 and USAR sections 1.4,
6.3, 6.4, and 9.7.
The team determined that all of the USAR sections were
revised in accordance with the Design Engineer's incorrect safety analysis.
In spite of the Design Engineer's assumption of operator action, the emergency
operating procedure for a LOCA (Reference 4) was not revised.
The emergency
procedure does not instruct or caution the operator to secure one containment
air cooling unit if all cooling units start as designed and off-site power is
available.
As a consequence, the USAR does not agree with the emergency
operating procedure.
A Document Update Checklist (Reference 5) completed by
the Design Engineer indicates that operating instructions and emergency
procedures do not require revision even though his safety evaluation and USAR
revisions require such action.
The team found that controlled copies of the Technical Specifications in the
-
-
control room and in Generating Station Engineering's reference library still
contained'an unrevised basis for technical specification 2.4.
Specifically,
theheatremovalcapagityofthecomponentcoolinggaterheatexchangersis
described as 420 x 10
BTU / hour instead of 402 x 10 BTU / hour.
BASIS:
Omaha Public Power District has committed to Regulatory Guide 1.64
-
(Reference 6) which endorse ANSI N45.2.11 (Reference 7).
This standard
requires that design changes be reviewed and approved by the same groups or
organizations which reviewed and approved the original design documents.
When
an organization which originally was responsible for approving a particular
design document is no longer available, the ANSI N45.2.11 Standard states that
the plant owner shall designate a new responsible design organization which
may be the owner's own engineering organization and that the designated
. organization shall have access to pertinent background information, have
demonstrated competence in the specific design area of interest and have an
adequate understanding of the requirements and intent of the original design.
Although the Omaha Public Power District's engineering organization can be
designated as the new responsible organization, it is required to have
an adequate understanding of the requirements and intent of the original
design.
Contrary to these requirements, Omaha Public Power District's Generating
Station Engineering organization did not have access to original design
analyses nor did not prepare comparable design analyses in the absence of such
design analyses. Instead, a qualitative argument was employed based upon an
incorrect understanding of the heat transfer phenomenon between heat removal
A-28
_ __. .--
.
-
-_ .
- .-_ -- - - .
.
.--__ --
.
.
.
-
.
,.
systems.
,
i
10CFR 50.59 permits licensees to make changes, conduct testing or experiments
as described in the safety analysis report without Commission approval, unless
such action involves a change in the technical specifications or an unreviewed
safety question.
An unresolved safety question is defined, in part, to occur
if the margin of safety as defined in the basis for any technical specifica-
tion is reduced.
Contrary to this requirement the licensee did not identify
that the basis of a technical specification was incorrect.
REFERENCES
1.
OPPD Safety Analysis, MR-FC-81-21B Component Cooling Water Isolation
- -
to Reactor Coolant Pump Seals Appendix 7.3 Safety Analysis, Rev. O,
August 30, 1982.
!
- -
2.
OPPD Revised Final Design Description, MR-FC-81-21B Component Cooling
Water Isolation to Reactor Coolant Pump Seals, Rev.1, December 6,
1982.
!
3.
OPPD Station System Acceptance, EEAR/MR-FC-81-21B CCW Isolation to
RCPs, May 5, 1983.
!
4.
OPPD Emergency Procedure EP-5, Loss of Coolant Accident, Rev. 24,
April 4, 1985.
5.
OPPD Document Update Checklist for SRDC0 81-239, Completed between
l
November 17 and November 18, 1981.
1
l
6.
Regulatory Guide 1.64, Quality Assurance Requirements for the Design
!
of Nuclear Power Plants, Rev. 2, June 1976.
7.
ANSI N45.2.11, Quality Assurance Requirements for the Design of Nuclear
l
Power Plants, 1974.
,
l
.
!
'
l
!
i
?.
l
!
A-29
!
l
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, m r ,,
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e,m
-,
-sm ww- ww.,mm,,
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,
-
.
e
D3.1-1 (Deficiency) Plant Design Specifications
DESCRIPTION:
Section "H" of Omaha Public Power District Contract No. 763
contains 46 design specifications which governed the analysis, design,
fabrication and testing of balance-of plant piping systems and equipment
procedure for Unit No. 1, Fort Calhoun Station.
The design specifications contained in Omaha Public Power District Contract
No. 763 do not constituto a cont. rolled design document.
These specifications
have not been revised or distriLated to the design staff.
The corresponding
specifications issued by the architect-engineer, Gibbs, Hill, Durham and
Richardson, which derived from the specifications detailed in Contract No.
763; have also not been revised or distributed to the engineering staff in a
controlled manner.
These latter specifications were design input for piping
at the plant.
As an example, Omaha Public Power District's Piping and
Instrumentation Diagram 11405-MECK-1 notes that:
"All piping shall be in
accordance with the requirements of the latest issue of Gibbs & Hill Piping
Specification H-1."
However, Omaha Public Power District could not access
this document during the inspection.
.In the absence of the design specifications issued by the architect-engineer,
the design specifications contained in Contract No. 763 appear to be used as
the defining design document for much of the plant piping and equipment.
BASIS:
The Omaha Public Power District Quality Assurance Manual (which
implements Omaha Public Power District commitments to ANSI N45.2.11) requires
that:
(1) " Applicable design inputs, such as design bases, regulatory requirements
codes and standards, shall be identified, documented and their selection
reviewed and approved.
Changes from specified design inputs, including
the reasons for the changes, shall be identified, approved, documented
and controlled," (Chapter 5.1 of Plant Design.and Modifications, Section
4.2, Design Inputs, Subsection 4.2.1.), and:
-
(2)' " Methods shall provide for relating the final design back to the source
i
of design input.
This traceability shall be documented."
(Chapter 5.1,
Section 4.2, Subsection 4.3.3.)
'
Contrary to these requirements, plant , design specifications are not being
l
controlled.
!
)
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i
A-30
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-.
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.-
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D3.1.2 (Daficiency) Design Temperatures for Safety-Related Piping
- _
_
DESCRIPTION:
In order to qualify a piping system, either by explicit or
generic analysis, the imposed loads, which include consideration of operating
and accident temperature, must be defined.
-
In 1980, Omaha Public Power District provided temperature data to
'
Gilbert / Commonwealth to reanalyze a number of safety-related piping systems in
'
the Ft. Calhoun plant in response to Bulletin 79-14. Generating Station
Engineering (GSE) verbally requested that Technical Services (TS) collate the
operating and accident temperatures for the safety-related piping in Ft.
Calhoun.
Technical Services subsequently transmitted this data to Generating
Station Engineering on a marked-up set of piping and instrumentation diagrams.
~,
' Technical Services compiled this temperature data from tne FSAR, and from
~~~
.
. analytical and operating data.
Omaha Public Power District subsequently
;';. transmitted the set of marked-up piping and instrumentation ~ diagrams to '
~
~
'
.' ~ ~ Gilbert / Commonwealth for use in their reanalysis. The marked-up set of piriing -
-
i'
'
'and instrumentation diagrams is not a controlled document.~
~
The original analysis temperatures which the architect-engineer (Gibbs, Hill,
Durham'and Richardson) originally used to perform. piping analysis was not
accessed in preparing the transmittal of information to Gilbert / Commonwealth.
Neither the licensee nor the team could determine if the operating and
accident temperatures which Gilbert / Commonwealth used to reanalyze a number of
safety-related piping systems were consistent with the temperature data
j
originally used to qualify these piping systems.
The operating and accident temperatures detailed on the marked-up piping and
i
instrumentation diagrams were used for all reanalysis work performed by
Gilbert / Commonwealth, and may have been used subsequently by Omaha Public
i
Power District for modifications to the installed piping.
BASIS:
The transmittal and use of uncontrolled temperature data is contrary
1
to the following requirements of the Omaha Public Power District Quality
Assurance Manual:
.
l
(1) Chapte.r 3.1, Document Control, Section 4.0, Requirements and Controls,
)
Subsection 4.1.1, which notes, in part, that:
"The preparation, issue and
change of documents that specify quality requirements or prescribe activities
!
affecting quality shall be controlled to assure that correct documents are
j
being employed";
M
- -
1
!
(2) Chapter 5.1, Control of Plant Design and Modifications, Section 4.2,
'
,
Design Inputs, Subsection 4.2.1, which notes that:
" Applicable design inputs,
such as design bases, regulatory requirements, codes and standards, shall be
-
identified, documented and their selection reviewed and approved.
Changes
i
from specified design inputs, including the reasons for the changes, shall be
identified, approved, documented and controlled";
,
!
(3) Chapter 5.1, Section 4.3, Design Process, Subsection 4.3.3, which notes
-
.that:
" Methods shall provide for relating the final design back to the source
of design input.
This traceability shall be documented";
!
,
l
A-31
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^
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-
- -
- - -
.- -. -
-
. .. --
-. -
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.
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(4) Chapter 5.1, Section 4.4, Interface Control, Subsection 4.4.4, which
notes that:
" Procedures shall be established to control the flow of design
information between Divisions / Departments.
Design basis information
transmitted from one Division / Department to another, shall be documented and
controlled.
Transmittals shall identify the status of the design basis
information or documents provided and, where necessary, identify incomplete
items which require further evaluation, review, or approval. Where it is
necessary to initially transmit design basis information orally or by other
informal means, the transmittal shall be confirmed by a controlled document,"
~
and;
,
(5) Chapter 5.2, Calculational Analysis, Section 4.0, Requirements and
Controls, Subsection 4.1.1, which notes, in part, that:
" References and
calculation inputs shall be identified, and shall be traceable to their source
documents to permit subsequent verification."
i
4
e
,
/
A-32
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-.
_
- --- - - -
- .--
- - - - _
-.
. - - -- --- - --= -
-
- -
_-
_
-
-_.
_ _ - .
.-
- __
-
- -
.
..
.
.
U3.1-3 (Unresolved Item) Small Bore Pipe Support Spacing
DESCRIPTION:
Piping two inches and smaller was field routed'for Unit No. 1,
Fort Calhoun Station.
Peter Kiewit Sons' Co., the contractor performing the
field routing, based restraint spacing and type on the technical criteria
detailed in " Recommended Procedure for the Support & Seismic Restraint of
Piping 2 Inch and Smaller." This procedure was developed by the contractor on
the basis of technical data developed by the architect-engineer, Gibbs, Hill,
. ...
"
Durham and Richardson, Inc. The contractor's support spacing criteria differ
from the seismic criteria detailed in the USAR for piping penetrating the
containment.
--
- As noted in the procedure under the heading entitled,' SEISMIC DESIGN CRITERIA:
'
_
~'"The calculation method used in determining seismic forces is based on the
-
" _ . .~ ..premisethatthepipingasrestrainedfallswithinth[e'rligidrange'which ~
~
.
. .
'. precludes the possibility of the piping going into resonance with the imposed
~
-
-
~
~
' '
'and/or building response frequency, thereby allowing the use of conservative
.
seismic acceleration design factors.
The minimum natural frequency as
designated by the engineers is 20 cycles per second in the horizontal
direction and 60 cycles per second in the vertical direction for the Intake
Structure, and 6 c
and Containment." ps horizontal,18 cps vertical for the Auxiliary Building
USAR Appendix F, subsection F.2.2.2, notes that:
"The first step in seismic
analysis of piping was to position seismic restraints closely enough to ensure
that the natural frequency of piping in the auxiliary building and containment
building was 6 hertz horizontally and 18 hertz in the vertical direction."
However, USAR subsection F.2.5 specifies a more stringent criterion of 12
(rather than 6) hertz in the horizontal direction:
"Therefore, for those
piping runs which penetrate the containment shell or are otherwise connected
to it, the spacing of restraints was such as to assure a lowest dominant
natural frequency of 12 Hz horizontally and 18 Hz vertically for the pipe run .
up to the first point of full fixity."
BASIS:
Based on the available documentation small-bore Class I pipe connected
to or penet' rating the containment may.not meet USAR seismic criteria,
considering the discrepancy between the contractor's su
and the seismic criteria detailed in the USAR for smallpport spacing criteria
bore (Class I) pipe
..;:
penetrating or connected to the containment shell.
No additional field
' routing procedures or analyses were available which address increased
.
horizontal rigidity for these piping systems.
!
4
i
I
I
1
4
A-33
i
_ _ _ _ ,_ _ . _ _ _ _ _ _
_
-
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-.
.. -
- - . _
_-. _
-
- - -
-
_
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_
__
_-
- - _ _ -
..
-
,
.
-
{
03.1-4(Observation)SeismicQualificationOfValvesinstalledInClassI
!
Piping Systems.
DESCRIPTION:
USAR Appendix F categorizes structures, piping systems and
equipment either as Class I or Class II (a plant-unique designation, not an
ASME classification) for purposes of seismic qualification at Fort Calhoun
Station.
The team examined the design process relating to valves to ascertain
if seismic considerations had been incorporated. Seismic qualification of
valves installed in Class I piping systems could not be determined by the
team.
-
As noted on page 1.4 of the Critical Quality Element List (Reference 1):
"Until the present ASME Section III component code was developed, items such
as pumps and valves were not specifically covered b
"In order to apply codes which were, in many cases,y a nuclear code," and:
written for piping only,
the architectural engineering firm (Gibbs, Hill, Durham and Richardson) and
Omaha Public Power District developed the specifications for pumps and valves.
The local designations are Class A-(safety class) and Class B (non-safety
class). The specifications for these classes can be obtained by referring to
the original contract documents." The team examined the design specifications
contained in Omaha Public Power district Contract 763, the original contract
documents used to procure valves for Fort Calhoun Station (References 2-6),
excluding those supplied by the NSSS.
No seismic criteria are detailed in
these specifications. Valves for the Ft. Calhoun plant were constructed in
accordance with piping codes ASA B31.1 and USAS B31.1.
Subsection F.2.2.1 of the USAR lists the following methods of analysis which
were applied to Class I structures, systems, and equipment:
The natural frequency of vibration of the structure or component was
a.
determined.
b.
The response acceleration of the component to the seismic motion was-
taken from the response spectrum curve at the appropriate natural
frequency and damping factor.
S' tresses resulting from the combined influence of normal loads and
c.
the additional load from the design earthquake were calculated and
checked against the limits imposed by the design standard.
d.
Stresses and deflections resulting from the combined influence of
normal loads and the additional loads from the maximum hypothetical
earthquake were calculated and checked to verify that deflections
would not produce rupture.
t
The NRC has already identified that for older plants, documentation is not
always available to establish seismic qualification of components.
A program
is in place to examine this issue on some older plants (post Systematic
Evaluation Program plants, including Fort Calhoun Station).
This program for
resolution of Unresolved Safety Issue A-46 (Reference 7) will include a review
of seismic capability of valves and valve operators.
A-34
. . _ -
.
.
..
-
.
REFERENCES
1.
Omaha Public Power District, Fort Calhoun Station - Unit No. 1 Critical
t
Quality Elements (C.Q.E.) List, Rev. 2, Issued May 24, 1985.
~
2.
OPPD Contract No. 763, Technical Specification No. 10, Safety and
Relief Valves.
,
t
3.
OPPD Contract No. 763, Technical Specification No. 11, Manual Globe,
j
Gate and Self-Actuated Check Valves.
4.
OPPD Contract No. 763, Technical Specification No. 12, Manual Ball,
'
j
Butterfly and Saunders Patent Valves.
!
5.
OPPD Contract.lto. 763, Technical Specification No. -13, Motor Operated '
!
Valves.
6.
OPPD Contract No. 763, Technical Specification No. 14, Air Operated
Valves.
7.
NUREG -1030 (Draft); Seismic Qualification of Equipment in Operating
,
Nuclear Power Plants and Regulatory Analysis for Proposed Resolution of
Unresolved Safety Issue A-46, published September 4, 1985 (FR85-21054),
,
i
,
i
i
-
i
.
i
A-35
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- . -
. - - -
-
. -
- .
- .
- . . . -
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_ _ _ _
.
_
_
_
_
_ ~ _ _ _ _
. - _
._.- _
.
.
.
.
-
.
,
U3.2-1 (Unresolved Item) MR-FC-84-61 Design Input Source and Use.
i
DESCRIPTION:
The team reviewed modification request FC-84-61, which installed
'
unions to facilitate the periodic removal of safety injection relief valves SI
,
209, 213, 217 and 221 for setpoint testing.
USAR Appendix F defines the
safety injection system as a Class I system.
i
The Final Design for MR-FC-84-61 (Reference 1) does not reference:
i
(1) The source of analysis temperatures and pressures used as calculation
input for the portion of the safety injection system to be modified;
(2) The governing load combinations detailed in USAR Appendix F, Table F-1;
4
(3) The vendor drawing for the safety injection tank relief valves;
,
l
(4) The design basis for the seismic qualification of the safety injection
system, and;
)
(5) Existing support locations and types for the portion of the safety
.
injection system to be modified.
!
'
With respect to item (1), only the rated pressure at maximum operating
temperature is specified for the safety injection tank / relief valve system;
.
the pressure is specified as 265 psig in Section 6.0 of the Final Design, and
'
as 275 psig in Section 7.0 of the Final Design. With respect to items 2-5, the
team found that the engineer did not analyze seismic effects and did not
document his judgement that such analysis was not required.
BASIS: The Omaha Public Power District Quality Assurance Manual which
.
1
implements the licensee's commitments to ANSI N45.2.11, requires that:
,
j
(1) " Applicable design inputs, such as design bases, regulatory requirements,
codes and standards, shall be identified, documented and their selection
)
reviewed and approved", (Chapter 5.1, Control of Plant Design and
. Modifications, Section 4.2, Design Inputs, Subsection 4.2.1);
)
(2) " Analyses shall be sufficiently detailed as to purpose, method, assumption,
design input, references and units such that a person technically qualified in
'
j
the subject can review and understand the analyses and verify the adequacy of
the results without recourse to the originator," (Chapter 5.1, Section 4.3,
Subsection 4.3.5).
l
1
Contrary to these requirements, the licensee has not identified the source of
.
design input and sufficiently documented analyses including engineering
'
'
judgements.
REFERENCES
1.
OPPD Final Design for MR-FC-84-61, Form GSE-B-2-2, Revision 1, dated
,
January 15, 1985.
i
i
2.
Peter Kiewit Sons' Co. Contract 763, Group I & Group II Piping Systems
4
l
Recommended Procedure for the Support & Seismic Restraint of Piping
'
2-Inch and Smaller.
,
.
l
A-36
1
gn-,-,
w-
-
_.
, , . , - - - . _ , , - - , _ ,
n
w
,w_,,,-
--
.wn
w
-
._
_
_ _ . _ _
,
..
-
.
D3.2-2 (Deficiency) MR-FC-83-158 Installation Procedure
-
,
i
DESCRIPTION: . . Support design for tubing and small-bore piping, which addresses
- - the governing seismic criteria, is normally performed on a generic basis.
,
Support spacing is then accomplished in accordance with generic design
2
guidelines, instead of a detailed physical or isometric drawing. The
installation procedure for modification request FC-83-158 does not address
support spacing requirements for tubing.
Modification request FC-83-158 provides air accumulators with check valves for
.
valves YCV 1045 A and B.
These valves are on the steam feed to the turbine
i
auxiliary feedwater pump FW-10, and are fail open valves. The Final Design
provides~a sketch which schematically locates the new tubing, valves and air
) -
accumulator with respect to the existing air set and root valve.
However, the
t
-
- installation procedure does not reference a generic support spacing procedure.
-
-
- The team notes that Stone & Webster prepared such a guideline for Omaha' Public
~
.
PowerDistrictin1982(Reference 1),whichprovidesgenericroutingand
support criteria for seismic instrument piping. The team also notes that the
radial location of the Hilti bolts which restrain the air accumulator was not
defined in Section 6.3 of the installation procedure.-
,
I
'
Subsequent to the preparation of the installation procedure Omaha Public Power
District perfomed a- calculation which seismically qualifies the air
!
j
accumulator support configuration (Reference 2). That calculation specifies a
9-inch radial location of the Hilti bolts with respect to the centerline of
i
the air accumulator.
'
BASIS: The licensee comitted to ANSI N18.7 (Reference 3), which requires
'
that each procedure contain instructions to the degree necessary for performing
_
'
a required task by a qualified individual without direct supervision, and
that the procedure contain appropriate references. Contrary to this require-
.
ment, the procedure did not address the installation requirements for seismic
l
tubing.
.
~
.
!
!
REFERENCES
'
1.
Stone & Webster guideline," Guideline for the Installation of Tubing and
Tubing Supports for Seismic Instrument Systems," J. O. No. 13007.65,
j
dated March 3, 1982.
2.
OPPD Calculation, " Generic Air Accumulators using Propane Tanks Built to
DOT Spec. 4BA-240," Rev. O, 6 pp., dated August 15, 1985; Rev. IR,
1 page, dated September 11, 1985.
3.
ANSI N18.7/ANS 3.2, " Administrative controls and Quality Assurance for
the Operational Phase of Nuclear Power Plants", c bruary 19, 1976.
e
A-37
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E
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.
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___
. _ _ _
__ ._
.
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i
D3.2-3 (Daficiency) MR-FC-84-162 Calculation
'
DESCRIPTION: A calculation prepared to qualify a modification to an existing
ventilation duct support must consider the specified design and operating
,
'
conditions, as well as the applicable seismic provisions of USAR Appendix F.
The team reviewed modification request FC-84-162, which redesigns two
containment ventilation duct supports to improve personnel and equipment
The containment heating, ventilation, air conditioning' ductwork is
access.
categorized as Class I equipment.
Team review of the calculations filed with
the modification indicates that:
(1) Thermal loads were not considered.
Specification No. 21
-
(Reference 1) specifies a design basis accident temperature of 288
degrees F.
'
(2) The natural frequency of the duct is computed on the basis of linear
beam theory, which is unconservative for the intent of this
calculation which is to establish that the duct has a fundamental
,
frequency in the rigid range; i.e., greater than 33 Hz.
(3) The 4 in. x 4 in. x 1/4 in. horizontal support angle is sized on the
basis of a-bending moment which is half the magnitude of the
critical bending moment in the angle (4954 vs. 10710 in-lb).
(4) The combination of vertical seismic and horizontal seismic loads in
the transverse direction of the horizontal angle was not considered.
There are no supporting calculations which justify this assumption.
!
.(5) The new supplementary steel is to be painted; however, Specification
.
No. 20 (Reference 2) requires that supplementary steel be
galvanized.
There was no documentation addressing this deviation
from the specification.
,
BASIS:
The Omaha Public Power District Quality Assurance Manual which
,
implements the licensee's commitment to ANSI N45.2.11 requires that:
(1) " Applicable design inputs, such as design bases, regulatory requirements,
codes and standards, shall be identified, documented and their selection
reviewed and approved", (Chapter 5.1,
Control of Plant Design and
Modifications, Section 4.2, O. Design Inputs, Subsection 4.2.1).
(2) " Methods shall provide for relating the final design back to the.
~
source of design input.
This traceability shall be documented",'
(Chapter 5.1, Section 4.3, Design Process, Subsection 4.3.3).
Contrary to its commitments, the licensee did not control design inputs nor
relate the. final design back to the source of design inputs.
'
REFERENCES
-
1.
OPPD Contract No. 763, Section "H", Technical Specification No. 21,
Reactor Containment Ventilation, Air Cooling and Filtering Equipment.
2.
OPPD Contract No. 763, Section "H", Technical Specification No. 20,
Heating, Ventilating and Air Conditioning Equipment.
,
,
A-38
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D3.2-4 (Deficiency) Junction Box Supports
DESCRIPTION:
The team inspected valve YCV-1045B, which is on the steam feed
to turbine auxiliary feedwater pump FW-10, during a plant walkdown conducted
on September 20, 1985.
The team noted that junction box JB-432A, which supplies power to the operator
for valve YCV-1045B, is restrained by a pair of unistrut supports, which are
in turn supported by conduits EB-4943, EB-9494 and EB-9127.
The team
questioned the support configuration for this junction box, in that unistrut
supports are normally used to support conduit and conduit are generally not used
.
as supporting membars, Omaha Public Power District could not produce a seismic
analysis which qualifies this configuration.
, BASIS:
USAR Appendix F requires that appurtenances to Class I systems be
seismically qualified to the Class I standards detailed therein.
As noted in
Subsection F.1.3 of USAR Appendix F:
"All supports associated with Class I
equipment are to be designed to Class I standards, i.e., in accordance with
the seismic criteria detailed in USAR Appendix F".
Contrary to USAR
commitments, the licensee has not demonstrated seismic qualifications of the
above equipment.
.
e
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A-39
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03.2-5 (Observation) Containment Pressure Switch Seismic Qualification
DESCRIPTION: Modification request FC-83-83 was prepared to evaluate alternate
replacement switches for the containment press .e switches which feed the
engineering safeguards system logic matrices.
The team reviewed the seismic
qualification of SOR pressure switches A-D/PC-742-1 and A-D/PC-742-2 (Eight
pieces), and the seismic qualification of the associated switch supports and
Hoffman boxes, with the exception of the seismic qualification of the
supporting instrument rack.
'
Theteamfoundtheseismicqualificationtestsofthepresurhswitches
performed by ACTON on behalf of SOR to be acceptable (Reference 1).
The Omaha Public Power District calculation performed to qualify the WT2x6.5
support for the pressure switch does not reference the drawing for the
pressure switch, preventing confirmation of the switch dimensions and weight
used in the analyses (Reference 2).
In addition, the Hoffman box shown on the
Omaha Public Power District arrangement drawing (Reference 3) was not
identified in the Omaha Public Power District calculation and was not included
in the analysis:
The team believes that a more thorough analysis for this modification would
provide additional assurance of the equipment's seismic qualification.
REFERENCES
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..,
1.
ACTON Report No. 17344-82N-D entitled, " Qualified Pressure Switches for
Nuclear Service / Test Report for Nuclear Qualification," dated February 4,
1983.
,
2.
SOR Drawing, entitled " Dimension Drawing 12N6-BB4-NX-C1A-JJTTX6, Drawing
No. 8215-783, Rev. 1," dated November 29, 1984.
.
3.
OPPD Drawing entitled, " Pressure Switch Mounting Arrangement, Drawing No.
B-4091, Rev. 0", dated May 21, 1985.
.
.
e
A-40
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D3.2-6 (Deficiency) Steam Generator Nozzle Dams
DESCRIPTION:
Omaha Public Power District contracted for the fabrication of
removable pipe plugs (dams) for the hot and cold leg pipes of the steam
generator (MR-FC-84-92) to enable refueling to proceed concurrently with
primary head work such as eddy current examinations.
As such, these dams are
the boundary of the reactor coolant system during such refueling operations.
The steam generator nozzle dams are designated as critical quality elements on
Purchase Order No. 7234, and are, therefore, subject to the seismic provisions
detailed in USAR Appendix F for Class I equipment.
However, Omaha Public
Power District Contract No. 1453 to Nuclear Energy Services (NES), the nozzle
~
dam vendor, did not specify any seismic criteria, and NES did not perform a
seismic analysis.
-
BASIS:
The Omaha Public Power District fai'ied to ensure that the steam
generator nozzle dams were qualified to the governing seismic provisions.
USAR Appendix F, Section F.1.3, lists the reactor coolant system as a Class I
System.
.
o O
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D3.2-7 (Deficiency) YCV 1045B Valve Restraint
'
DESCRIPTION:
During the inspection, the team toured the plant to examine
various equipment scheduled for modification during the upcoming outage. The
team examined auxiliary feedwater steam feed valve YCV-1045 B, an air operated
valve scheduled to have an accumulator added to ensure the valve operator's
2
'
ability to close the valve following a steam generator tube rupture with
concomitant loss of the non-safety instrument air system.
The team noted
that the valve's operator was questionably restrained by a thin rod attached
to a stair post, and therefore examined the seismic qualification of the
piping subsystem containing YCV-1045 B.
!
,
In response to NRC Generic Letter No. 81-14, " Seismic Qualification of
Auxiliary Feedwater Systems," (Reference 3), Omaha Public Power District
engaged Gilbert / Commonwealth to perform an evaluation of the auxiliary
i
feedwater system at Fort Calhoun Station.
Omaha Public Power District actions
j
taken to bring the. system into compliance with IE Bulletin No. 81-14 are
'
contained in Omaha Public Power District Modification Request No. FC-81-127.
~
MR-FC-81-127 summarizes the four. major seismic deficiencies identified by
,
Gilbert / Commonwealth.
One of these, Item 2 of Form B states:
" Valve
'
Operators on Small Bore Piping.
The current operator supports were found to
i
t
be unstable.
Modification work will involve removing the existing support
rods and replacing them with a more stable support by the end of 1981."
i
'
Gilbert / Commonwealth specifically noted that the valve operator for valve
YCV-1045B.is unstable in the transverse direction, and recommended that the
existing rod restraint be replaced with a strut in the transverse direction
(Reference 1, attachment 2).
Gilbert / Commonwealth recommended the addition of
a number of supports for the steam drive and condensate portions of the
l
auxiliary feedwater piping associated with pump FW-10.
Gilbert / Commonwealth
also recommended that a detailed stress analysis be performed to assure that
the additional supports do not have a detrimental thermal impact on the system
(Reference 1, attachment 2).
Omaha Public Power District elected to perform
the required analysis, using computer codes TPIPE and NUPIPE.
,
The licensee's response to Generic Letter 81-14 (Reference 1) identified the
four major seismic deficiencies and included as attachments the Gilbert /
Commonwealth Inspection Evaluation Report and the calculation tabulating
support discrepancies.
The licensee letter specifically committed to removing
the existing support rods on the unstable valve operators and replacing them
with more stable supports by the end of 1981. The attachments indicated
support number AFW-15 was unstable.
This is the support for YCV-1045B, which
the team questioned during their walkdown.
The NRC letter fowarding the
safety evaluation for Generic Letter 81-14 concluded that the seismic
'
qualification of the auxiliary feedwater system was acceptable provided that
the four corrective actions committed to by the licensee were taken.
One of
these actions (Item B) was to replace existing support rods for valve
operators with more stable supports.
Apparently because the committed
completion date was before issuance of the NRC letter, a telephone call was
made to the licensee to verify completion of these actions.
The letter
documented the NRC's understanding that these actions have been completed.
Notwithstanding these commitments and attempts at verification, the team
noted the following:
g
A-42
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-_ , , . , . , , _ , _ , _ _ , _ , _ _ _ _ _ . _ _ ,
_ _ - , _ _ _ .
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,'(1) T.he valve operator for valve YCV-1045B is currently restrained by a
-rod which is affixed to a stairpost; the strut substitution
committed to by Omaha Public Power District was not implemented;
,
(2) The Omaha Public Power District as-built drawing does not show
either the valve operator or ths existing rod restraint
(Reference 4);
[
(3) The vendor drawing for valve YCV-1045B could not be obtained to
j
verify the valve and operator weights, and operator offset dimension
1
with respect to the valve centerline;
.
(4) The valve operator restraint w:: not ;;;cdeled in the stress analyses;
(5) Therearenocalculationswhichcombinedeadioad[thermaland
seismic stresses in the vicinity of the valve to confirm.the
structural adequacy of the adjacent pipe;
'
'(6) There are no calculations which combine deadload, thermal and
seismic loads for the adjacent supports; based on cursory
' examination of the computer output by the team, the supports appear
to be overloaded, and:
1
(7) The TPIPE and NUPIPE computer runs are not referenced and are
i
therefore not adequately controlled.
!
BASIS: The primary basis for this deficiency is USAR Appendix F, which
requires that Class I piping systems and equipment be seismically qualified to
the Class I standards detailed therein.
In addition, t.e licensee's
commitments in response to Generic Letter 81-14 were not implemented.
, REFERENCES
1.
OPPD (W.C. Jones) letter to the NRC (D.G. Eisenhut) response to
.
Generic Letter No. 81-14, dated July 14, 1981.
2.
NRC (R'.A. Clark) letter to OPPD (W.C. Jones), fowarding a Safety
Evaluation Report in regard to Generic Letter 81-14 to OPPD dated
February 10, 1982.
-
3.
NRC Generic Letter 81-14, Seismic Qualifications of Auxiliary Feedwater
Systems, dated February 10, 1981.
4.
OPPD Drawing, " Fort Calhoun Station /CQE Piping Isometrics / Seismic Sub.
System #MS-4099A," Drawing No. D-4318, Sh. 1, Rev. 1, dated August 26,
1985.
A-43
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04.1-1 (Observation) High Power Rate of Change Trip Bypass
DESCRIPTION:
A high rate of change of power trip is operable within the
reactor protective system between 10 x 10 4 percent power and 15 percent power.
In the original design for Fort Calhoun, a bypass annunciator was automatically
illuminated except during plant startup or shutdown between these two611mits.
This type of annunciation complies with IEEE Std. 279-1968 (Ref. 1) in that,
" ..if the protective action of some part of the system has been bypassed or
.
deliberately rendered inoperative for any purpose, this fact shall be
continuously indicated in the control room."
This continuously illuminated annunciator is in conflict with a more recent
!
recommendation for a " dark board" during normal plant operation (Ref. 2).
Consequently, a design change .eodification (Ref. 3) was proposed to convert
the annunciator to indicate the state of a, "...high rate of change of power
trip enabled." With this change, the annunciator is illuminated only when the
trip function is effective, and is " dark" during those periods when the trip
function has been automatically bypassed.
,
The final design package did not identify the IEEE Std. 279-1968 design basis
requirement for this automatic bypass annunciation. As a result, no
justification was provided to address the conflicting requirements placed on
this particular annunciator. The safety analysis, prepared in accordance with
i
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10 CFR50.59, stated that no alarm signal was eliminated; however, the automatic
indication of a protective system bypass has been eliminated by this change.
(Absence of the " trip enable" does not necessarily indicate bypass of the trip
function because of annunciator logic changes incorporated with the modification
request).
The team believes that if the licensee had thoroughly examined this change
the potentially conflicting design objectives could have been satisfactorily
addressed and resolved.
The team understands that OPPD is giving further
consideration to implementation of this design modification.
t
.
REFERENCES
1.
IEEE Std. 279-1968, Criteria for Protection Systems in Nuclear Power
Generating Stations, section 4.13.
i
2.
NUREG/CR-3217, Near-Term Improvements for Nuclear Power Plant Control
,
Room Annunciator Systems, April, 1983.
'
3.
FC-84-46, High Power Rate of Change Trip Alarm, Rev. O, 3/6/84.
,
,
A-44
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04.2-1 (Observation) Delta T Power Loop Analysis
DESCRIPTION:
A set of replacement resistance temperature detectors has been
purchased from Conax along with associated transmitters from Foxboro to
replace existing Rosemount detectors and transmitters in the reactor coolant
hot and cold leg loops. Technical Services was requested to evaluate the
impact of this new instrumentation on the reactor protective system inputs to
the thermal margin / low pressure calculator drawer.
The Technical Services evaluation was provided in an OSAR-85-83 report dated
September 26, 1985. The team determined that:
,
(1) The analysis involved safety-related RPS channel inputs, but was
-
not specifically identified as being either safety-related or a
'
calculation involving critical quality equipment;
(2) The analysis presented input values and final results, but did not
provide the calculation formula nor the entry of input values into
the formula needed for traceability;
~
In addition, the team noted that applicable Technical Services procedure (Ref.
3) does not contain the Critical Quality Element identification requirement of
a similar Generating Station Engineering procedure (Ref. 5).
The team believes that the licensee's various responsible design organizations
should implement design changes in a consistent fashion, and that in this
particular instance the Technical Services procedure lacks desireable controls
contained in the Generating Station Engineering procedure.
REFERENCES
~
1.
FC-84-140, De1ta T Power Process Loops.
2.
OSAR-85-83, Uncertainty Evaluation for MR-FC-84-140, 9/30/85.
3.
Technical Services Procedure, N-TSAP-5, Operations Support Report
-
Documentation, Rev. 1, 5/85.
4.
Technical Services Procedure, N-TSAP-8, Analysis Verification,
Rev. 1,
5/85.
5.
Generating Station Engineering Procedure, B-9, Technical Calculation
Production, Checking, and Approval, 1/84.
A-45
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D4.3-1 (Deficiency) Limit Switch Circuit Protection by Fusing, MR-FC-84-74A
DESCRIPTION:
Subject to postulated submergence, nine safety-related valves
had their solenoids relocated to higher elevations during the previous refueling
outage. The valve position limit switches were also discovered to have not
been qualified for submergence during third party review. However, they were
not relocated at the time to avoid additional mechanical complexity. Omaha
Public Power District decided to provide low-current fast-acting fuses in the
indicator light portion of the valve control circuits to retain operability of
the solenoid, even though position indication may be lost or may become
ambiguous to the operator.
For the selected design (Ref. 1), a technical assumption that the indicator
light fuse (shown as 0.25 amperes in the final design modification package and
as 0.50 amperes in the construction design modification package) would
interrupt a fault current before the solenoid fuse (10 or 15 amperes) was not
justified, particularly since the expected range of circuit current
interruption is outside the range of values specified in the manufacturer's
,
catalog. The design engineer did not provide fuse coordination data in either
of the desigr. pack' ages; however, the first design package checker's checklist
(Ref. 2) had a notation to " verify fuse clirves" and the subsequently revised
design package checker's checklist required that a calculation of fuse
coordination be provided in the design package.
The design package did not
provide any indication that the coordination had been confirmed or that it was
appropriate.
The team's review of this catalog data (Ref. 3 and 4) indicated that the
circuit interruption time differential between the two fuses may be only 10
milliseconds. Unintended circuit interruption by the larger valued fuse would
prevent electrical operation of the solenajds for these nine valves, and would
,
impair the control room operator's capability to remotely close HCV-238 and
HCV-239 valves for long term core cooling.
i
F
On September 17, Omaha Public Power District requested BUSSMANN to confirm
~
i
the fuse coordination of the selected fuses, and a favorable response (Ref. 5)
has been received. A demonstration test is planned by BUSSMANN to confirm
their analysis assessment.
BASIS:
An identified technical assumption had not been verified during the
preparation, review, and approval of the final design modification package in
violation of a design evaluation requirement on page B-2.6 of Omaha Public
l
Power Di' strict Procedure B-2 and Omaha Public Power District's commitment to
section 4.2 of ANSI N45.2.11-1974.
.
REFERENCES
1.
MR-FC-84-74A, Fuse Protection for Certain Limit Switch Circuits.
2.
Design Package Checker's Checklist, FC-84-74A, Rev. O, 5/31/85.
3.
BUSSMANN " MIN" fuse catalog, 10 and 15 ampere ratings.
3 ._
4.
BUSSMANN "KTK" fuse catalog, 0.25 and 0.50 ampere ratings.
5.
BUSSMANN letter, Fuse Coordination, 9/30/85.
A-46
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.U4.3-2 (Unresolved Item) ESF Bypass Switch Keylock Provision, MR-FC-81-102
DESCRIPTION:
To simplify the means for bypassing specific engineered safety
_
feature channels, keylock bypass switches are being implemented into the trip
channels for pressurizer low pressure and steam generator low pressure (Ref.
1). Electro Switch Series 20 switches and Hoffman NEMA enclosures with
cylinder locks have been selected for this purpose.
The purchase order (Ref. 2) for metal enclosures to house these bypass
switches requested that cylinder locks and keys be provided, but did not
,
specify the lock combinations needed to assure that only one channel would be
!
' bypassed at any given time. An Omaha Public Power District Form B (Ref. 3) had
requested that different keys be used for the individual trip and bypass '
-
functions. Relevant design guidance was provided in the reactor protective
_:-
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3 system description (Ref. 4) for keylock bypasses in that, "...each trip bypass
~~
.
.has a-different lock cylinder combination; however, corresp'on' ding trips in
-
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.each of the four protective channels have the same cylinder combination...with
one key provided for each trip type." Thus, for the reactor protective
system, administrative controls on keylock switches were enhanced by an
' engineering thought process and hardware differences in the keylocks.
The final design modification package contained no requirement for keylock
cylinder combinations and the number of keys needed to control bypassing of
individual trip channels.
BASIS: The above configuration appears to violate Omaha Public Power District
procedure B-2 pages B-2.5 and B-2.6 in that the technical description and
design evaluation did not contain all of the equipment requirements necessary
to establish an unambiguous design configuration.
REFERENCES
1.
MR-FC-81-102, Bypass or Trip of ESF Channels Without Jumpers, Rev. O,
8/14/85.
2.
Purchase Order 98505-CB, 8/5/85.
3.
OPPD Form B, Technical Services Review and Evaluation, 4/25/83.
-
4.
Ft. Calhoun RPS System Description, Rev. 4, 5/24/84.
5.
USAR section 7.3.1.
.
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A-47
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04.3-3 (Observation) Procurement Requirements on Equipment Vendors
DESCRIPTION:
Engineering groups within nuclear steam supply and architect
engineer organizations that specify and procure safety-related equipment
typically require that: (1) plant specific technical requirements for
equipment performance be specified to the vendor, and (2) expected equipment
performance be confirmed and documented by the design engineer in a traceable
manner.
For a number of design modifications reviewed by the team, Omaha Public Power
District design engineers have chosen to specify procurement of a vendor
product by catalog number in conjunction with submittal of vendor
qualification test reports where applicable. This approach tends to place a
greater degree of conformance responsibility on the Omaha Public Power
District design engineer rather than on the equipment vendor.
In addition, a
complete identification of relevant performance requirements may not be
accomplished.
The team has noted two problems in the procurement approach selected by Omaha
Public Power District for the modification packages reviewed; namely, (1)
important equipment performance requirements have not always been identified,
and-(2) a confirmation that the vendor's product satisfies the Fort Calhoun
need has not been documented in a traceable manner for design verification
purposes. For example,
(1) When isolators are used between Class 1E and non-Class IE
circuits, one requirement is that the isolator perform its
required function during the application of the maximum
credible voltage to its output terminals (typically 120 volts
AC and 125 volts DC). This requirement was not identified to
TEC for Critical Quality Element isolators between safety-related
signals and the. emergency response facility computer system. The TEC
test plan did not include a test of the isolator for application of
these voltages to the output terminals. In addition, test reports
were not required to be submitted for Omaha Public Power District
,
review.
(2) When keylock switches are used to bypass protective system
or ESF trip functions, design control over the cylinder lock
combination and number of keys is needed to augment plant
administrative controls (see Unresolved Item 4.3-2).
These aspects
were not included in the purchase order for MR-81-102.
(3) A new vendor was selected to provide replacement pressure
switches for the containment high pressure measurement.
An identification of a requirement for pressure boundary
integrity would appear prudent since these switches serve
as an extension of the containment boundary prior to being
isolated. (Remote manual isolation is provided).
!
The team believes that consideration should be given to improving the
effective use of documents that can provide design requirements for the design
engineer.
Table B-2-1 in Generating Station Engineering procedure B-2
>
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provides a framework for assuring that design inputs have been considered in
the design process. Other pertinent sources for design requirements include
the Fort Calhoun USAR and individual system descriptions.
In addition, an item by item comparison of Fort Calhoun technical require-
ments with equipment performance data contained in vendor reports would
provide better traceability that the procured equipment meets the plant's
requirements.
REFERENCES
1.
FC-83-83, Containment Pressure Switches.
2.
FC-83-109, Transfer of P250 Points to the ERFC.
3.
FC-85-62, Replacement of Component Cooling Flow Element.
4.
FC-81-102, Bypass or Trip of ESF Channels Without Jumpers.
.
.
A-49
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U4.4-1 (Unresolved Item) Design Basis Physical Separation Within Panels
DESCRIPTION:
The team has reviewed both current and previously implemented
design modifications that involved the physical separation of safety-related
cables outside of control room panels and the separation of safety related and
l
non-safety-related wiring within these panels. The requirement for separation
'
was qualitatively stated in the " Independence" section of IEEE Std. 279-1968.
The Fort Calhoun plant has committed to meet IEEE Std. 279-1968.
For recent plant modifications, a definitive quantitative design basis for
physical separation of redundant safety-related internal panel wiring
harnesses is stated as six inches or a physical barrier by an Omaha Public
Power District wire list form. During construction, the plant had a
requirement for separation within panels (Ref. 3 and 8) and a separation
requirement between safety and non-safety cables (Ref. 9). A commitment for
-
separate and segregated routing of~each engineered safeguard control channel
has been made (Ref. 10).
The Fort Calhoun design basis for physical
separation is stated in a qualitative (i.e. functional) manner on design
documents and in the USAR.
However, a quantitative design basis needed to
provide measurable acceptance criteria does not appear to be stated in a
complete and unambiguous manner leading to certain potential deficiencies. For
example:
(1) Separation of original wiring within panels is defined only to
the field interface terminal blocks; redundant safety wiring
separation for plant modifications has been specified, but
separation of safety to non-safety internal panel wiring or the
installation of barriers has not been imposed. A Final Safety
Analysis Report Appendix G commitment made in 1970 stated that
physical separation of individual channel components and wiring is
maintained wherever practicable;
(2) Achievement of internal wiring separatiori within panels was a
,
General Electric responsibility (Ref. 3); however, the criteria
used to accomplish this separation could not be located during
the inspection. Original wiring within the panels appears to be
Vulkene 600 volt insulation over multi-strand conductors. Visual
.
inspection of control room panels by the team indicated that
separation appeared to have been incorporated in the original design;
(3) For recent modifications, internal panel wiring uses single
conductor wiring covered with an 85% coverage braid (Ref. 6).
The equivalence of this wiring to meet an acceptable distance or
barrier criterion has been assumed; however, it has not
been demonstrated;
(4) While redundant Engineered Safety Features components implemented
for undervoltage protection (Ref. 2) were mounted within separated
barrier compartments, their wiring external to these compartments is
in direct contact within control room panel CB-4, and;
(5) The proposed addition of trip bypass switches (Ref. 1) has
safety-related wiring on three wafers of each switch and non-
safety related wiring to an annunciator on the fourth wafer.
A-50
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An analysis to confirm that Class IE circuits would not be
degraded below an acceptable level due to their proximity to
non-Class 1E circuits was not provided in the design package.
'
BASIS: The routing of redundant safety-related wiring in direct contact from
-
an internal panel compartment barrier configuration violates the wiring
harness separation requirement imposed by the Omaha Public Power District wire
list form. No analysis has been provided to justify the lack of separation for
this wiring. This is a violation of the USAR commitment for separation of
engineered safeguard features controls.
,
A quantitative criterion for separation of safety-related and non-safety-
related internal panel wiring could not be located. For the ESF bypass' switch
-
-modification, no separation criteria for safety to non-safety internal panel
wiring have been applied, and no analysis has been performed in~ lieu of
-
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. quantitative separation criteria to assure that Class:1E ' circ'uits have not
~
.
been' degraded below~an acceptable level. Since the non-safety wiring provides
a potential common link among redundant channels, this is a violation of the
USAR commitment for segregation of engineered safeguard features controls.
REFERENCES
1.
MR-FC-81-102, Bypass or Trip of ESF Channels Without Jumpers,
Rev. O, 8/14/85.
2.
MR-FC-77-40, Undervoltage Protection, Rev. O, 8/13/78.
3.
GHDH Letter NY-762-313 to GE, " Contract 762 Panel Wiring," 8/31/70.
4.
IEEE Std 384-1981, Criteria for Independence of Class 1E Equipment
and Circuits.
l
-
5.
IEEE Std 420-1982, Design and Qualification of Class 1E Control
Boards, Panels, and Racks Used in Nuclear Power Generating Stations.
6.
Purchase Specification GSEE-0505, Alpha Wire Corporation, Rev. O,
i
4/28/77.
7.
OPPD Procedure, GSEE-0516, Requirements for Installation of Electric
Cables at Fort Calhoun Station, Rev. 2, 7/19/85.
8.
OPPD Contract to General Electric, page H 1-7.
,
9.
USAR page 8.5-3, item (i) and USAR figure 8.5-1 note 22.
10.
USAR page 7.3-1, section 7.3.1.b.
,
A-51
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- - . . -
- .
-
. - . -
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04.5-1 (Deficiency) Drawing Changes by Procedure A-9, MR-FC-82-178
'
,
DESCRIPTION:
Generating Station Engineering Procedure A-9 (Reference 2)
specifies that when an existing drawing needs revision during preparation of a
modification request, the design engineer is to request a sepia of the
drawing.
This requirement provides control of the drawing during
consideration of the design modification, assists in documentation of the "
as-built status, and alerts other users to coordinate any desired changes with
other drawing (facility) changes under consideration.
Drawings 11405-M-1 and 11405-M-2 (References 3 and 4) were modified to
incorporate air filter differential pressure switches without use of the sepia
control process specified by procedure A-9.
The drawings were directly
modified based on engineering sketches provided with the final design
modification request package.
I
BASIS:
Compliance with the provisions of engineering procedure A-9 was not
provided during the development and implementation of a final design
modification package.
REFERENCES
.
1.
MR-FC-82-178, HEPA Filter DP Indication, Rev. O, 1/23/84.
2.
GSE Procedure A-9, Document Control, section 2.3.3.4, 8/83.
'
3.
P&ID Drawing 11405-M-1, Containment Heating, Ventilation, and Cooling,
Rev. 48, 8/13/85.
4.
P&ID Drawing 11405-M-2, Auxiliary Building Heating and Ventilation,
i
Rev. 37, 11/84.
-
1
4
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. --_-_. .,.
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04.5-2 (Observation) Flow Element Design Basis Conditions
.
. DESCRIPTION:
A replacement flow element was specified for she component
~
,
~
-cooling water system flow measurement.
In this specification,' maximum
temperature, minimum and maximum humidity, integrated radiation dose, and
seismic acceleration values were identified.
The specified values for the mechanical flow element were not consistent with
design basis conditions specified for associated electrical equipment in the
GSEE-0802 general requirements specification for controlled access areas of
the auxiliary building, as shown in the following tabulation:
Parameter
FE-498 Specification
GSEE-0802. Specification
f
~^'
Humidity
40 to 100%
15 to 100%
Temperature
120 degrees F
40 to 122 degrees F
-
6
7
Integ. Radiation
3x10 rads
2x10 rads
-
'
Horiz. Seismic
3g
.27g (USAR)
~
Vert. Seismic
3g
.13g (USAR)
During the inspection, it was determined that the temperature value was
i
obtained from an original Gibbs, Hill, Durham and Richardson instrument
specification sheet, the humidity range from other unidentified Fort Calhoun
,
'
documents, the radiation dose from either the equipment qualification program
status report for components in the vicinity of this flow element or from an
obsolete inside containment radiation value, and the seismic acceleration
values from an engineering judgement assessment.
On a technical basis, the
teamconcernregardingtheradgationdosevaluewassatisfiedasth9 expected
dose is predicted to be 4 x 10 rads.
Although the actual values specified for the flow element are technically
acceptable, the team questioned the control process which generated the
specifications.
A consistent design basis for operating condition values,
except where additional margin is deemed necessary on a case-by-case basis,
should be used for equipment procurement. Differences from this plant basis
should be identified and justified in the final design package in accordance
-
with the design input provisions of ANSI N45.2.11-1974.
REFERENCES '
1.
FC-85-62, Replacement of Component Cooling Flow Element.
-.
2.
GSEE-0802, General Requirements for CQE (Class 1E) Electrical-
Equipment Required for Use in Controlled Access Areas ~of~the Auxiliary
-
'
Building Outside Reactor Containment, Rev. O, 7/14/80.
"
3.;
System Component Evaluation Work Sheet, item PCS-412, sheet 4-62, Rev. O,
'4/8/85.
-
4.
.GHDR Orifice Plates and Flanges Specification Sheet, item FE-498, sheet
1.14B, Rev. 1, 2/18/75.
A-53
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,
U4.5-3 (Unresolved Item) Battery Room Fire Hazard Analysis
DESCRIPTION:
Approximately five years ago, an extensive fire hazard analysis
was performed at the Fort Calhoun Station. For each of the two battery rooms,
the significant combustibles were identified as the plastic battery cases,
polystyrene separators between the battery cases, and a small amount of
electrical cable insulation (Ref. 3).
During a plant walkdown the team identified a fuse block enclosure constructed
of masonite with a fiber board cover in each of these rooms (Battery room
arrangement drawings ,(Ref. I and 2)).
BASIS:
The existence of a wooden fuse block enclosure was not identified in
the fire hazard analysis of the battery rooms (fire areas 37 and 38). The team
could not locate a description of the test for significance determination for
combustible materials used by Omaha Public Power District in the fire hazard
analysis.
It is indeterminate whether this material is a significant
combustible with respect to the published fire hazards analysis.
REFERENCES:
1.
Arrangement Drawing 11405-E-73, Switchgear, Diesel-Generator, and
Electrical Penetrations, Elevation 1011 feet, Rev. 26, 3/22/83.
2.
Arrangement Drawing D-4168, section X-X, Rev. O, 1/22/84.
3.
Fort Calhoun Safety Evaluation Report, page 5-23.
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A-54
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D5.1-1 (Deficiency) Battery Sizing Calculation
-
DESCRIPTION:
The battery must be sized to provide sufficient capacity to
~
supply the direct current loads under all operating conditio~n's without its
voltage dropping below a specified minimum value.
Battery voltage under
-
discharge conditions is determined by the state of charge remaining in the
cells, the number of cells which make up the battery, and th'e rate of
discharge.
Battery voltage under charge conditions is based upon the number
of cells and the battery charger setting in volts per cell.
In order to
reduce the maximum voltage of the battery under charge conditions, Omaha
Public Power District removed 2 cells from the 60 cell battery (Reference 1).
Removing cells from the battery reduces the battery's capability to supply the
~
required load current without the battery voltage falling below the minimum
i '~ .
acceptable voltage. The battery size must be checked ~u' der this lower
-
n
capacity condition.
-- -
-
The battery size was checked by Omaha Public Power Distri~ct'using the
.
'
calculation method recommended by IEEE 485 (Reference 2).
The battery cell
-
.[
data was correctly used at the lower permissible discharge rate to limit the
-
cell discharge voltage to 1.81 volts per cell (or 105 volts for 58 cells).
However, the team noted that the battery current discharge profile was the
same profile used in the 1979 sizing calculation (Reference 3) that was
originally used to purchase replacement batteries.
This profile had not been
,
updated even though increased de loading was responsible for planned
replacement of the 200 ampere battery chargers with 400 ampere units.
In an
attempt to determine the adequacy of the 1979 profile, the team found:
The load table used to construct the discharge profile was
-
composed of general loads without supporting references to
substantiate detailed loads.
No justification was documented for failing to include major
-
loads such as switchgear control power or diesel generator
field flashing.
.
The 1979 calculation did not contain any documented check or
-
verification of the required discharge profile.
BASIS: Generating Station Engineering procedure B-9 (Reference 4) requires
, the checker to confirm that assumptions have been justified.
Generating
Station. Engineering procedure B-11 (Reference 5) also ~r~e~ quires the third party
,
,
reviewer to confirm that the calculation assumptions have been' justified.
The
-
l
inputs and assumptions used in the latest battery sizing calculation were not
verified.
REFERENCES
^~
1.
OPPD Modification MR-FC-84-119, Battery Charger and Inverter Replacement.
2.
IEEE Std. 485, Sizing Large Lead Storage Batteries for Generating
Stations and Substations, Recommended Practice For.
3.
OPPD Modification MR-FC-79-03, Replacement of Station Batteries.
4.
OPPD Procedure GSE-B-9, Technical Calculation Production, Rev. 8/85.
5.
OPPD Procedure GSE-B-11, Design Verification, Rev. 8/85.
l
I
A-55
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U5.1-2 (Unresolved Item) Battery Charger /DC BUS Coordination
DESCRIPTION:
Battery chargars are provided to maintain the battery in a fully
charged condition and to recharge the battery following a discharge.
The
4
battery chargers are also the source of the steady state DC power required for
normal plant operation.
Because of load growth, the original battery chargers
were operating in their overload region.
Omaha Public Power District decided
to replace the existing 200 ampere Exide battery chargers with 400 ampere
battery chargers from Power Conversion Products (Reference 1).
The original
battery charger would provide a current limit at 110% (220 Amperes).
As part of the modification to replace the battery chargers, Omaha Public
Power District also planned to replace the existing 225 ampere breakers at the
DC switchboard with 400 ampere breakers (Reference 2).
Omaha Public Power District stated that the 400 ampere breaker was the largest
breaker size that would fit in the existing switchboard.
Omaha Public Power
District failed to demonstrate that the new breaker was compatible with the
new battery charger.
Omaha Public Power District made the assumption that the
battery chargers would limit the charging current to less than 400 amperes
-
based upon the nominal charger voltage and required steady state load.
l
The team noted that the instruction manual provided with the battery charger
i
stated that the battery charger would go into current limit in the range of
110 to 125% of rated full load.
This would mean that in attempting to
!
recharge a discharged battery, the battery charger would attempt to provide up
to 500 amperes.
This amount of current would trip the 400 ampere breaker
located at the DC switchboard.
Omaha Public Power District could not produce
any evidence at the time of the inspection to demonstrate that the current
limit feature of the battery charger could be adjusted to limit the current to
below the 400 ampere breaker trip point.
BASIS:
Omaha Public Power District procedure GSE-B-2 (Reference 3) was
prepared consistent with commitments to ANSI N45.2.11 (Reference 4)
-
,
requirements for design change control.
Procedure section 2.5.2 requires that
'
when specific setpoints or limitations are imposed by other systems or
components,. such setpoints or limitations are to be clearly stated.
Neither
the checker or the third party reviewer questioned the compatibility of the
new battery charger with the existing switchboard, including the new larger
circuit breaker.
REFERENCES
1.
OPPD Modification MR-FC-84-119, Battery Charger and Inverter Replacement.
2.
OPPD Dwg. D-4341, Issue A, Inverter Replacement and Battery Charger
One Line Diagram.
3.
OPPD Procedure GSE-B-2, Production of Design Description and Evaluation
Nuclear Modifications.
4.
ANSI N45.2.11, Quality Assurance Requirements for the Desigr. of Nuclear
Power Plants.
A-56
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,
.
,
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.
05.1-3 (Observation) Power Cable Sizing
DESCRIPTION:
Power cables are sized to consider allowable ampacity, voltage
.
drop and short circuit conditions.
The allowable ampacity for a power cable
is dependent upon:
1) Type of insulation on the conductor.
2) Ambient temperature seen by the raceway.
3) Proximity to other cables in the raceway.
Three types of raceways are in general use in power plants.
These are
underground ducts, conduits, and cable trays.
The Insulated Cable Engineers
'
Association (ICEA) has established allowable ampacities for power cable
_ addressing the above considerations in their standards (References 1 and 2).
. . , The Fort Calhoun Updated Safety Analysis Report Section 8. 5'.'4 refers only 'tci
~ ~ 'J -
.the IPCEA (currently ICEA) Publication No. P-46-426 cable ampacity limits.
-
~
'
-
However, the vast majority of power cables are routed in cable tray.
The
cable selection analysis used in the original plant design is not available to
Omaha Public Power District engineers, so it is left to the individual
engineer to interpret how this standard applies to cables in cable tray each
time a new cable is routed or an existing cable rerouted.
Also the USAR
makes specific reference to Article 430-22.of the National Electrical Code
(Reference 3, which pertains to motor loads) but makes no reference for
similarly derating any other continuous duty loads.
Design of cable sizing would be enhanced if Omaha Public Power District
obtained the specific criteria used to size cable routed in cable tray and
documented those criteria for future use in a cable sizing guidance document.
REFER.ENCES
~
1.
ICEA Publication No. P-46-426, Power Cable Ampacities.
-2.
ICEA Publication No. P-54-440, Ampacities in Open-Top Cable Trays.
3.
' NFPA-70, National Electrical Code.
4.
USNRC Regulatory Guide 1.64, Quality Assurance Requirements for the
.
Design of Nuclear Power Plants.
5.
ANSI N45.2.11, Quality Assurance Requirements for the Design of
Nuclear Power Plants.
!
6.
MR-FC-84-119, Battery Charger and Inverter Replacement
7.
MR-FC-85-25, Fire Wrapping of Power Cables.
l
I
A-57
.
.
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1,-w-
--.-g
,,y.,
,-.y-..
.,--- ,- - ------ , , , . .
,.m,.e,-.---..v
-m--
w
,.- .,
-,.-e-,.m,-.,,--.---r,---
, - - -
-- - - - - - -
-
-
l
.
.
.
-
.
05.1-4 (Observation) Pre-operational Test Requirements
DESCRIPTION:
The direct current system consists of large stationary
-
batteries, battery chargers, distribution equipment and the de loads.
The
equipment and components connected to the cc system are designed to operate
over a specified range of voltage above and below the system nominal voltage.
The system minimum voltage is restricted by the allowable minimum battery cell
discharge voltage in accordance with the battery capacity sizing calculations.
The system maximum voltage is determined by the setting of the battery charger
required to maintain the battery fully charged.
Because of the industry problems experienced with high de system voltage
affecting connected de components, Omaha Public Power District decided, in
modification package MR-FC-84-119 (Reference 1), to reduce the system maximum
voltage to 135 volts by removing two of the sixty cells that make up each of
the station batteries.
With two less cells in the string, the battery charger
voltage could be reduced and still maintain the desired volts per cell
charging voltage.
Also, as part of this same modification package, the battery chargers
themselves would be replaced with larger battery chargers from a different
equipment manufacturer.
During the course of the inspection, the team determined that the battery
chargers were originally manufactured for the Marble Hill Station and
purchased from Public Service Company of Indiana complete with all
documentation (Reference 2). Team review of the original test data for this
equipment. revealed that the High DC Voltage alarm setpoint was factory set at
150 volts.
No revised setup procedures were available at the time of the
inspection.
The team also noted that no surveillance procedures were
available for the disconnected cells.
A lower setpoint for the High DC Voltage alarm setpoint would provide the
desired protection for the de system.
Surveillance procedures are also needed.
for the spare cells.
REFERENCES ~
1.
OPPD Modification Package MR-FC-81-119, Battery Charger and Inverter
Replacement.
2.
OPPD Purchase Order 079927, Model 35-130-400 Power Conversion Products
Battery Chargers.
A-58
.
--
.
.
.
05.1-5 (Observation) Inverter Sizing Without Analysis
DESCRIPTION:
Safety and non-safety-related instrument.ation at Fort Calhoun
are fed from four 125 Vdc to 120 Vac inverters.
These de powered inverters
provide a source of uninterruptible power for the four safety-related and two
non-safety-related instrument buses.
The original design of this system
included provisions that would permit two safety-related instrument buses to
be tied together in the event that one of the four inverters was taken out of
service for maintenance.
Whenever this tie was made the inverters would run in their overload region,
because of increased loading on these buses since the original design. Also,
. because of the age of the original inverters, Omaha Public Power District
decided to replace them with new units.
--
The new inverters were purchased from existing stock from the Tennessee Valley
~
-
.
.
- Authority.
However, Omaha Public Power District was'only able to obtain
safety-related inverters of a smaller size than the original design (Reference
1,2).
Omaha Public Power District justified using the smaller inverters for
. the safety buses by making a one time measurement of the running load on each
of the instrument power buses during normal plant operation (Reference 3) and
by purchasing two additional inverters for the two non-safety-related
i
instrument buses (Reference 4).
No analysis was performed of the connected
load or the effect off-normal plant operation would have on the load
requirement because Omaha Public Power District felt that the apparent margin
,
j
over running load would compensate for any off-normal operating transients.
I
. A well controlled design would establish what specific loads are connected to
the instrument inverters and determine the individual demand loading under
different conditions of plant operation.
REFERENCES
1.
OPPD USAR Fig. 8.1-1, R3 7/85, Simplified One-Line Diagram, Plant
Electrical System.
.
2.
OPPD Drawing D-4341, Issue A, 6/85, Inverter Replacement and Battery
Charger One-Line Diagram.
'
3.
Teleph'one Call, H. Faulhaber (GSE) to J. Foley (Ft. Calhoun) 8/2/85.
4.
OPPD Purchase Orders No. 07828 and No. 08008, Elgar Inverters.
.
.
A-59
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05.1-6 (Observation) Design Interface Control
DESCRIPTION:
Major plant modifications require interface between different
design organizations.
The design control procedures in use at Omaha Public
Power District require that interfaces be considered as part of the design
modification process (Reference 1).
Two areas of design interface were examined by the team during this review:
(1) Equipment mounting requires review and verification by the
structural department to ensure that safety-related equipment is
- mounted in accordance with the vendor's seismic qualification.
This
review must also verify that building floor loading is acceptable.
(2) Equipment heat loss and ventilation requirements must be reviewed by
the mechanical department to ensure that any additional heat loads
are accounted for.
The design modification packages reviewed contained reference to these
.. concerns but the design process had not proceded far enough for the structural
W $ rtment to perform any review (Reference 2).
No input to the mechanical
department had been performed at the time of the review.
The team concluded that a system does exist which generally questions
inter-disciplinary interface.
However; the team is concerned that no
inter-disciplinary review had yet been performed by either structural or
mechanical departments for the electrical modifications that were scheduled.
for installation during the 1985 outage.
REFERENCES:
1.
OPPD Design Procedure GSE-82, Production of Design Description and
Evaluation of Nuclear Modifications.
2.
OPPD Internal Memorandum GSE-FC-85-850(E15), 9/4/85, Inverter Replacement,
Seismic Suoports.
,
,
A-60
.
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.
.
.
.
4
D5.2-1-(Deficiency) Fire Wrap Protection for Cable Raceways
.
DESCRIPTION:
Cables for redundant systems are separated from their' redundant
counterparts to meet the fire protection requirements of 10 CFR Part 50
Appendix R (Reference 1).
Where the required separation distance cannot be
'
maintained, fire barriers may be used.
During a previous NRC inspection
,
(Reference 2), a lack of sufficient separation was noted for the pressurizer
heater power cables.
'
In response to the previous inspection finding, Omaha Public Power District
i
prepared modification package MR-FC-85-25 to move the feeder cables to three
Bus 3 motor control centers (power source for the pressurizer heaters) from
'
cable. trays and reroute these cables in conduit.
These conduits were to be
. protected with a 3M fire wrapping system.
Because fire ~ wrap' reduces the~ heat
,
t
.
transfer from the. cable through the conduit, Omaha Public Power District
-
'
requested 3M to provide cable derating factors for the Omaha Public Power
-
- :
District application.
3M~ responded with derating factors developed using a-
'
heat transfer computer program developed by 3M for this type of application
(Reference 3).
Based upon these derating factors, and the general cable
ampacity design margins described in the Updated Safety Analysis Report
(Reference 4), Omaha Public Power District justified the use of existing cable
sizes.
Omaha Public Power District did not determine the actual loads on the
,
1
motor control center feeder cables in this analysis, nor did they attempt to
justify the 3M computer generated derating factors.
In an attempt to verify the 3M program, the team independently estimated the
!
derating factor required for the 3M fire wrap by using the heat transfer
i
-
method developed by Neher and McGrath (Reference 5).
This method was the
basis for the Insulated Cable Engineers Association cable ampacity standard
- '
(Reference 6) referenced in the USAR.
The team's estimate for required cable
derating was higher than that suggested by 3M.
l
Omaha Public Power District was not able to explain the computer inputs and
>
outputs used and/or developed by 3M.
In response to the team's questions,
Omaha Public Power District requested from 3M a verification of their computer
program. 3M supplied a report (Reference 7, developed in response to the
' team's concerns) based upon test data to justify the required derating
factors.
The 3M test data indicated a required derating factor almost twice
that determined by the 3M computer program and subsequently used by Omaha
Public Power District in the design modification analysis.
~
~
~
--
BASIS:
Generating Station Engineering procedure B-9 (Reference 7) requires
-
that computer calculations be checked.
Generating Station Engineering
~
procedure B-11 (Reference 8) checklist B-11-1G requires that computer
calculations be verified. The checker failed to question the need to know the
j
actual load current on the motor control center feeder cables.
Both the
checker and the third party reviewer failed to verify the computer program
'
developed by 3M or confirm that it had been verified.
A-61
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.
.
...
l
,
,
.
.
.
.
REFERENCES
1.
10CFR50 Appendix R, Section III, G.
2.
USNRC Inspection Report 50-285/83-12.
3.
3M letter 7/11/85, Computer Ampacity Determinations.
4.
USAR Section 8.5.4.
5.
Neher & McGrath, AIEE Paper #57-660, October 1957.
6.
IPCEA P-46-426, Power Cable Ampacities.
7.
3M Report, Ampacity Considerations, October 1985.
8.
OPPD Procedure GSE B-9, Technical Calculation Production, Rev. 8/85.
9.
OPPD Procedure GSE-B-11, Design Verification, Rev. 8/85.
l
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1
A-62
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.- _.-
.
- - . - _ _ . , , _ . , ,
. . - . _ . ,
,
_
_ . _
. .
, . _ ,
. , _ _ - , _ _ _
_ . . - -
-._.------.__v_.--.,
_
_
,.
,
j
.
.
.
.
.D6.1-1 (Deficiency) Safety Evaluations For Non-Safety-Related Systems
Described In The USAR
,
. DESCRIPTION:
Non-safety related final design packages'were reviewed in
- .'
conjunction with the USAR to determine if safety evaluations as required by
10 CFR 50.59 were required, and if so, were properly accomplished and
-documented in the final design packages.
Each of the final design packages
in this review were evaluated against the USAR descriptions to determine if
..
changes to USAR text, drawings and figures would be required as a result of
~
these modifications.
If so, as required by 10 CFR 50.59, a safety
evaluation of the design change would be required to determine if an
.
.unreviewed safety question existed even though these were non-safety related
systems. 10 CFR 50.59 does not differentiate between safety and
-
non-safety-related systems.
~
~
~
-
.
.0maha Public Power District procedure Standing Order G-21, Station
~
' Modification Control, discusses safety evaluations for both final design
-
U
'
packages and construction packages.
This procedure only requires the
preparation of a safety analysis in a Final Design Package if safety related
<
equipment is involved or impacted.
By virtue of G-21 procedure requirements
.for the Planner, a construction package 5-fety evaluation should always get
accomplished, however, the team considers that a construction package safety
evaluation is for construction related activities and does not fulfill
10 CFR 50.59 requirements for a safety analysis of the design aspects of
design changes unless those design attributes are specifically addressed.
G-21 also indicates that the final safety analysis of the design is to be
part of the final design package.
Review of non-safety-related final design packages revealed that five
modifications planned for completion during this outage did not have safety
evaluations in the final design packages.
Each of the affected systems or
>
equipment were described in the USAR and it appeared that completion of the
modifications would require changes to USAR text, drawings or tables to
l
accurately represent the newly changed systems or equipment.
The
-
modification packages and affected areas of the USAR are discussed as
-
follows:
MR 483-175 Feedwater Regulating System Instrumentation Replacement
.
This modification will replace the existing Feedwater Regulatory System
-
-
. '
~1
(FWRS) with an entirely new system.
The new system will use two'~ '
'
~ separate controllers, one for downcomer level and one Yor' flow error.
It will also have automatic control above 5% power.
.At the time of the inspection, USAR section 7.4.3 specified use of a
three element controller using steam flow, feed flow and downcomer
level.
USAR Figure 7.4-5 depicted a three element controller.
USAR section 7.4.3 specified manual control below 15% power.
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MR 485-008 Boric Acid Addition System
This modification will modify the phosphate addition system to add
boric acid for control of intergranular stress corrosion cracking
F
and steam generator tube denting.
!
At the time of this inspection, USAR section 10.2.2 specified
,
that, " Chemicals are added to the feedwater upstream of SG
feedwater pumps for oxygen scavenging and ph control." The team
noted that this modification will be adding more chemicals for
other reasons.
USAR P&ID M-253 showed piping entry points for " phosphate feed see
>
P&ID M-265."
The team noted that this modification will be adding
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boric acid at these entry points and not phosphate treatment.
In
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addition, the new concept of use of boric acid as a chemical
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control agen.t in the steam generators did not undergo a safety
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analysis by engineering.
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MR 474B-057 Power System Stabilizer
This modification will add a stabilizer to the main generator Alterex
Exciter to stabilize generator power in case of fluctuations in the
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Mid-Continent Area Power Pool (MAPP).
!
At the time of this inspection, USAR section 8.2.1 had a network
stability analysis.
The team noted that this modification will enhance
Fort Calhoun Station's ability to handle network fluctuations. USAR
section 10.2.4 discussed generator field excitation by the Alterex
Excitation System.
The team noted that this modification will add a
power system stabilizer to the Alterex Exciter.
MR 483-174 Reactor Regulating System Steam Dump and Bypass Alarm
This modification will make wiring changes and additions within the'
main control boards to provide operators with status of system
conditions for dissipating excess NSSS stored energy following a
At the time of this inspection, USAR section T.4.4.2 discussed system
design with a list of what the system consisted of as well as system
inputs and system outputs.
The team noted that this modification will
change wiring and add indicators to the control room.
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MR 483-90
Replace LP Feedwater Heaters
This modification will replace existing CuNi low pressure feedwater
heaters and drain cooler tube bundles with stainless steel units.
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At the time of this inspection, USAR Figure 10.2-6 depicted flow,
temperature and heat transfer data based on the existing CuNi units.
The team noted that this modification will introduce changes to this
data.
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BASIS:
10 CFR 50.59, in part, allows changes to facility as described in
'
'the Safety Analysis Report (SAR), without prior NRC approval, if the
~
modifications do not introduce an unreviewed safety question.
The Omaha
Public Power District procedure for 10 CFR 50.59 reviews for unreviewed
safety questions is Standing Order (50) G-46, Evaluation of Procedures,
Procedure Changes, Tests, and Experiments for Safety Evaluations and Status
as an Unreviewed Safety Question.
The result of such reviews is a written
safety evaluation addressing the three question criteria presented by 10 CFR 50.59.
Safety evaluations are required by 10 CFR 50.59 when USAR text,
drawings or tables are changed by facility modifications.
Each of the
modifications discussed above were described in the USAR in sufficient
detail that changes to the USAR would be warranted.
Accordingly, safety
.
,
. evaluations should have been accomplished in the final design packages.
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U6.1-2 (Unresolved Item) Safety Analyses For Emergency Modifications
DESCRIPTION:
Omaha Public Power District procedure Standing Order G-21
allowed emergency modifications to be accomplished by plant personnel with
telephone approvals and also allowed issue of final design packages after
completion of the modifications.
The team reviewed emergency modifications
to determine if design safety evaluations had been accomplished prior to
operation of the modified systems.
Generating Station Engineering (GSE)
procedure B-2 requires that a safety evaluation be accomplished for all
critical quality element (CQE) equipment, and, as discussed in deficiency
D6.1-1, 10 CFR 50.59 requires a safety evaluation for design changes to
facilities as described in the USAR.
Review of emergency modifications revealed that the following CQE emergency
modifications had been accomplished and the affected systems relied upon for
operation without a final design safety evaluation.
MR 484-84
DC Grounds on Critical Quality Element (Safety Injection)
Valves
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MR 483-129 Diesel Generator Speed Sensing Power Supply
MR 483-152 Diesel Generator Speed Sensing Power Supply
These modifications were completed in May 1S84, September 1983, and October
1983, respectively.
Presence of construction package safety evaluations was
not looked for in each of these cases since, as discussed in deficiency
D6.1-1, construction package safety evaluations do not necessarily satisfy
design aspect requirements for final design package safety evaluations.
The
design process followed for emergency modifications was a simplified process
performed by.the plant engineers and not the responsible design
organization.
The team considers that a period.of time between system
,
modification and completion of the design safety evaluation was acceptable
provided a safety evaluation by the responsible design organization was
completed prior to relying on the system for plant operation.
.
BASIS:
These modifications all involved critical quality element equipment.
In accordance with Generating Station Engineering procedure B-2, a safety
,
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evaluatio'n of the~ design is required for all critical quality element
structures, systems or components.
10 CFR 50.59 requires that safety
i
evaluations be performed for proposed changes to the facility as described
in the FSAR.
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06.1-3.(Observation) Vital AC Inverter Bypass Mode
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. DESCRIPTION:
The licensee was experiencing several problems with the vital ~
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AC inverters.
MR-84-119 (Reference 1) was initiated to correct these
problems.
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In the original design, the four instrument AC buses (A), (B), (C), and (D)
were powered from static inverters which in turn were powered from redundant
Class 1E batteries.
Non safety-related vital AC buses AI-42A and AI-42B
were supplied through isolation transformers from vital ac buses C and D,
respectively.
There were no provisions for powering these vital ac buses
from the station's Class 1E 120 volt ac (interruptible) distribution, nor
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'were any automatic transfer devices incorporated in the design.
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..The proposed design for tne station uninterruptible vital'ac~incorpo~ rated
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.' automatic transfer of a vital bus to interruptible 120~vacupon' loss'of
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. power by a static transfer switch and provisions to manually transfer the
bus to interruptible ac power by means of an inverter bypass switch.
The
proposed design also provided two new non-safety related inverters dedicated
to buses AI-42A and AI-428.
These inverters contain similar bypass and
automatic transfer provisions.
The original design also contained circuit breakers and associated cabling
^
which allowed powering two vital buses from the same inverter (A and C
together or B and D together) for purposes of maintenance on the inverters,
etc~
The new design does not retain such a feature for the
non-safety-related vital ac buses.
During review of the Final Design Package for MR-84-119, the team identified
several questions regarding the adequacy of the associated 10 CFR 50.59
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safety analysis.
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Licensee procedures do not specifically require evaluations and
documentation of whether or not Technical Specifications are affected by
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proposed facility changes performed pursuant to 10 CFR 50.59.
The team noted that several technical specifications are involved with the
vital ac inverters.
These include specifications 2.7.1(i) for the four ac
instruments buses, specifications 2.7.1(j) for the non-safety vital ac
.
buses,,and specification 2.0.1(2) for emergency and normal power sources.
The team noted that the licensee's proposed design has created new
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. possibilities for powering the subject buses from oth~er than their normal
... .;uninterruptible power supplies.
The licensee did not-incorporate hardware
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' . features which would preclude powering more than one of these buses from
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.interruptible power.
In evaluating whether or not the possibility of an
accident or malfunction of a different type than evaluated previously in the
USAR was created, the licensee determined that:
"No new failure models are created if no more than one inverter is
operated in BYPASS at anytime.
Operation of more than one inverter in
the BYPASS mode could create a failure mode which has not been
previously evaluated. Technical Specification 2.7 permits plant
operation with one inverter inoperable.
This is justified because all
engineered safety feature and reactor protection system channels revert
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to or can be placed in a two-out-of-three logic.
Ht. wever, if two
inverters were simultaneously operated in BYPASS, a loss of offsite
power combined with failure of a single DC bus could result in complete
loss of instrument AC until a diesel had started.
This failure mode
has not been analyzed in the USAR.
Therefore, in order to maintain
consistency with the existing USAR and Technical Specifications, the
BYPASS mode must be defined as inverter failure."
The team noted that the technical specifications do not specifically refer
to inverters, therefore a question is raised regarding the meaning of
" inverter failure." The team also considered that internal interpretations
of technical specifications in a safety analysis are not an appropriate
method of implementing changes in that such conditions should be clearly and
unambiguously addressed in the Technical Specifications.
During this
inspection the team identified one other deficiency (D2.2-6) wherc an
internal' safety evaluation statement, requiring the operator to secure one
of the containment coolers if all four start following an accident, had not
been incorporated into plant procedures.
The team noted that although the interpretation that a bypassed inverter is
considered inoperable would reduce the probability that more than one
inverter could be powered from interruptible sources, it does not eliminate
the possibility.
Operator error, potential common mode failure, etc., could
result in more than one vital ac bus powered from interruptible ac sources.
In addition, only interruptible backup power is now available to the non-
safety-related vital buses.
This aspect was not addressed in the safety
analysis.
The licensee did not consider the basis for technical specification 2.0.1(2)
in this analysis.
This technical specification broadly allows consideration
that individual systems, components, and devices are operable if either the
7
normal or emergency power source is operable, and allows the governing
electric power technical specification limiting condition for operation to
control continued plant operation.
.
The team asked Generating Station Engineering personnel whether or not they
consulted the NRC Safety Evaluation Report, documenting NRC review and
acceptanc'e of the FSAR, during the preparation of the safety analysis.
The
team was informed that although certain Generating Station Engineering
personnel read specific safety evaluations when issued by the NRC (for
example, with Technical Specification changes), copies of the safety
evaluations were not maintained by engineering nor consulted during licensee
safety evaluation preparation.
The team obtained a copy of the original
'5afety Evaluation Report, dated August 9, 1972, and noted that several
aspects of the proposed inverter modification could at least be questioned
in light of the evaluation.
For example, the report stated that the
licensee had originally proposed automatic transfer devices for both the ac
and de systems, however, these features were eliminated following NRC
concerns regarding jeopardizing redundant power sources.
The team noted
that the vital ac buses which are normally de powered through an inverter
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will new have an automatic transfer to ac power.
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A technical specification change to clearly address the availability
requirements for the new inverters, and review of a comprehensive set of NRC
safety evaluations by engineers performing safety analyses would minimize
the potential for introduction of an unreviewed safety question.
REFERENCES
1.
Modification Request FC-84-119, Battery Charger and Inverter
Replacement Lowering Terminal Voltages and Battery Discharge Breakers,
Final Design Review dated August 20, 1985.
2.
Fort Calhoun Standing Order No. G-25, Evaluation of Procedures,
" Procedure Changes, Tests and Experiments for Safety Evaluations and
Status as an Unreviewed Safety Question," issued November 24, 1984.
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06.2-1 (Observation) Untimely Closeout Of Emergency Modifications
DESCRIPTION:
Omaha Public Power District procedure Standing Order G-21
addresses emergency modifications and allows issuance of the final design
package by. the Generating Station Engineering design engineer after work has
been completed on the emergency modification.
These design packages are
called "after-the-fact" design packages.
Emergency modifications were evaluated by the team for length of time
required to issue after-the-fact design packages since completion of the
modification.
Omaha Public Power District procedures provide no guidance
for timely issuance of after-the-fact design packages.
Six emergency modifications were noted to have excessive issue times for
after-the-fact design packages. These are summarized below:
Time Required
Date Mod.
Design Package
To Issue Design
MR No.
Installed
Issue Date
Package (Months)
484-83 Hydrogen Purge
May 1984
April 1985
11
Fan Shielding
480-93 Replacement of
March 1981
October 1984
42
483-07 Alternate Load
Jan. 1983
None Yet
32+
Cell for FH-1
483-129 DG-2 Speed
Sept. 1983
None Yet
24+
Sensing Power Supply
483-152 DG-2 Speed
October 1983
None Yet
24+
Sensing Power Supply
,
484-84 DC Grounds on
May 1984
None Yet
16+
SI Valves
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Although ANSI N45.2.11'does not specifically address closure times for work
packages, it is clear from a system acceptance standpoint that an excessive
length of time for the responsible design organization to issue their design
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documents is unsatisfactory.
Because emergency modifications were usually
done by plant personnel and not by the responsible design organization,
timely performance of the independent design activity would help to assure
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that th'e installed emergency modification was in fact satisfactory from a
design standpoint.- This would allow final acceptance of emergency
~ modifications by the System Acceptance Committee much sooner.
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D6.2-2 (Deficiency) Modifications to AFW Turbine Steam Supply Valves
DESCRIPTION:
While performing an operability check on the s' team driven
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auxiliary feedwater pump (FW-10) in September,1978, it was found that an
instrument air supply valve to the YCV-1045 operator had been inadvertently
closed with the result being that YCV-1045 was in its failed closed position
.
and FW-10 was inoperable.
Subsequently, EEAR FC-78-43 was processed which
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recommended that the YCV-1045 valve operator he redesigned to fail open,
1
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assuring maximum auxiliary feedwater availability upon loss of air.
The
memorandum later additionally recommended the addition of air accumulators to
the YCV-1045A/B valve operators, which have always been fail open valves.
.This would enable remote manual isolation of a tube-ruptured steam generator
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upon loss of instrument air as per Criterion 57, App. A, 10 CFR 50, which
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' containment.
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The original proposed modification changed the failure ~ mode of the' air
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operated AFW turbine steam admit valve (YCV-1045) to fail open from fail
closed. Generating Station Engineering evaluated the emergency status of
the modification in June 1979 and determined that the proposed modification
would violate containment isolation requirements.
MR 78-43 was revised to
required addition of safety-related accumulators to the steam supply valves
(YCV-1045A/B).
On March 21, 1980 work was actually completed on YCV-1045 on
an emergency basis making the valve fail open; however, no accumulators were
installed on YCV-1045 A and B.
In October 1983, during Generating Station Engineering review of the
"after'-the-fact" Final Design package for emergency MR 78-43, they
discovered that the accumulators had not been installed on YCV-1045A and
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YCV-1045B.
Generating Station Engineering recommended that the plant make
arrangements to install accumulators as soon as possible as the operation of
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these valves would be required during a steam generator tube rupture
!
(Reference 5).
The station initiated a low priority (priority 4) request
for a " minor" modification on December 27, 1983. During November 1983,
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Generating Station Engineering completed the "after-the-fact" Final Design
package for MR 78-43. This package, dated November 8, 1983, noted that air
accumulators were to be installed in the future under a new MR 83-158.
The
safety analysis included in the Final Design Package discusses the need to
isolate the steam supply following a steam generator tube rupture and for
"
M . ' containment isolation provisions, and notes the abili~ty o'f YCV-1045A 'and B ^
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~ to accomplish this with accumulators.
The unreviewed safety question
evaluation addressed whether the probability of occurrence or the
..
consequences of an accident or' malfunction of equipment important to safety
previously evaluated in the Safety Analysis Report may be increased.
The
evaluation concluded that since this modification changed the failure
position of valve YCV-1045 and added air accumulators to valves YCV-1045A
and YCV-1045B, this will not affect the results of the main steam line break
analysis.
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However, the team noted (as acknowledged in the Final Design Package for MR
78-43) that the accumulators had not been added and it was improper to
conclude that no unreviewed safety question existed based upon work to be
performed at some future date.
During review of the station's Engineering Evaluation and Assistance Request
in December 1984, Technical Services identified the improper " minor"
classification of the modification request and recommended upgraded priority
from 4 to priority 1.
The Technical Services review also resulted in
discussions between Technical Services, Licensing, Generating Station
Engineering and Plant personnel addressing these concerns.
!
As a result of these discussions (reference 6) the licensee decided to
install the accumulators at the next refueling outage, or the next (earlier)
.
outage of sufficient duration.
The licensee also evaluated operation of the
plant during the previous six years in light of the requirements of General
,
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Design Criterion 57 (reference 1).
The licensee concluded:
"From 1979 until the present, the requirement of Criterion 57, App. A, 10
-CFR 50 has been met under normal plant operation, or in the event of a steam
generator tube rupture.
It is only when the postulated tube rupture occurs
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in conjunction with a loss of air that the requirament for remote manual
isolation of the affected steam generator cannot be met.
However, during
this six year interval, local manual isolation of the two-inch steam line
has always been possible and this operator action would have been performed
in the event of a tube rupture coupled with loss of air as per Emergency
Procedure EP-30, Step D.8.d.
The plant has determined that time required for
this operator manual action is not excessive and will insert an appropriate
statement regarding this requirement in EP-30, to reinforce operator's
>
awareness of this required action to isolate the potential radionuclide
leakage path via FW-10 steam feed."
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The licensee further determined that the operation of the plant with the
described configuration did not constitute an unreviewed safety question and ,
was not reportable.
The team vert'fied that the emergency procedures had
been revised as recommended.
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The team determined that the following design control inadequacies relating
!
to failure t'o incorporate a portion of an approved modification, excessive
length of time to process a completed emergency modification, and basing a
.
10 CFR 50.59 safety evaluation for a completed facility change on work yet to
'
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be performed were a deficient condition.
The team also considers that
l
operation of the facility as described was an unreyiewed safety question in
that the possibility had been created wherein remote manual isolation of the
'
AFW steam supply might not be possible following a steam generator tube rupture
if offsite power was lost or the non-safety-related instrument air system was
otherwise unavailable.
'
The team also noted that the USAR does not accurately reflect the as-built
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configuration for the containment penetrations which are associated with
YCV-1045A and B.
USAR Table 5.9-1 shows these penetrations (M-94 and M-95)
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as Type IVD, containing a single power operated valve whose normal position
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is open, fails closed, and accident position is closed.
The team noted that
although this depiction is correct for the main steam isolation valves, the
~
Afit steam supply taps off on the upstream (containment) side of the main
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steam isolation valves, and that these valves are normally closed, fail
open, and accident position is open.
In addition, the main steam isolation
.
valve bypass valves are not shown. These USAR errors had apparently not
been identified and corrected during licensee reviews of these
modifications, notwithstanding the licensee's concerns for compliance with
BASIS:
10 CFR 50.59 requires evaluations of proposed changes to the
facility to be made without prior NRC approval, to ensure an unreviewed
safety question does not exist.
A proposed change involves an unreviewed
' safety ' question if the consequences of an accident previously analyzed in
.
.the FSAR.is increased or if a malfunction of a different type than evaluated
,
'previously in the Safety Analysis report may be created.
10 CFR 50, Appendix A, Criterion 57 - Closed system isolation valves,
requires that, "Each line that penetrates primary reactor containment and is
neither part of the reactor coolant pressure boundary nor connected directly
to the containment atmosphere shall have at least containment isolation
valve which shall be either automatic, or locked closed, or capable of
remote manual operation."
REFERENCES
1.
CFR 50.59, 50.73(a)(2)(B)), Part 50 Appendix A, General Design
Criterion 57.
2.
USAR Table 5.9-1, Containment Isolation Valves, penetrations M-94, 95.
3.
Modification Request FC-83-158, Air Accumulators for YCV-1045A/B, EEAR
dated 11/8/83.
4.
Emergency Modification Request FC-78-43, Analysis of Failure Mode of
YCV-1045, Safety Analysis dated April 20, 1979.
5.
Memorandum from M. Eidem to W. Gates, Failure Mode of YCV-1045,
dated October 27, 1983.
6.
Memorandum from R. Jaworski to W. Gates, Review of Failure Mode
.
Modifications on YCV-1045A/B Steam Supply Valves to Steam-Driven AFW
Pump, FW-10, dated January 15, 1985.
,
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