ML20154K131
| ML20154K131 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 08/19/1988 |
| From: | Imbro E, Parkhill R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20154K120 | List: |
| References | |
| 50-285-88-200, NUDOCS 8809230229 | |
| Download: ML20154K131 (73) | |
See also: IR 05000285/1988200
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U.S. NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION l
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Division of Reactor Inspection and Safeguards
Special Inspection Brar.ch
Report No.: 50-285/88-200
Docket No.: 50-205
Licensee: Omaha Public Power District
1623 Harney Street
Omaha, Nebraska 68102
Facility Narre: Fort Calhoun Station
Inspection at: Omaha Public Power District Engineering Offices, Omaha, Nebraska
Inspection Conducted: February 18, March 21-25, and April 4-8, 1988
Inspection Team Members:
Team Leeder: R. W. Parkhill, Senior Operations Engineer, NRR
Assist. Team Leader: M. E. Murphy, Region IV
Mechanical Systems: G. J. Overbeck, Consultant, ERC International
Mechanical Components: A. V. DuBouchet Consulting Engineer
Instrumentation and Control: J. M. Leivo Consulting Engineer
Electrical Power: G. W. Morris, Consultant ERC International
Attended Exit Meeting:
J. P. Jaudon, Deputy Director, DRS, Region IV
A. B. Beach, Deputy Director DRP, Region IV
E. V. Imbro Section Chief, NRR
T. F. Westenn'an, Section Chief, Region IV
P. H. Harrell, Sr. Resident Inspector, Fort Calhoun
l T. Reis, Resident Inspector, Fort Calhoun
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l /&*/ ?. k' v,1-S e
! RonaldY.ekhlh
Parkhill Date
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Team Leader
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Approved by: OMU
E. V. Inibro, Chief Late
Team Inspection Development y
and Appraisal Section
8609230229 88o916
gDR ADOCK 05000285
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Fort Calhoun Station
550MI Design Reinspection
February 18, March 21-25, and April 4-8, 1988
1. Background
The NRC Office of Inspection and Enforcement conducted a safety systems outage
modification inspection (SS0MI) of the Fort Calhoun Station in the fall of
1985. That inspection was initiated to examine the adequacy of licensee
management and control of modifications durino major outages. The associated
inspection report, 50-285/85-22, for the design portion of the SS0MI was issued
' on January 21, 1986; and an enforcement conference was held in the regional
offices on July 10, 1986. NRC regional management requested that the Division
of Reactor Inspection and Safeguards within the NRC's Office of Nuclear Reactor
Regulation conduct a reinspection to assess tne licensee's corrective actions
associated with all of the original SSOMI's design findings.
2. Purpose
The primary purpose of this inspection was to verify the adequacy of the
corrective actions initiated by the licensee in response to the 47 design-
related findings identified in Inspection Report 50-285/85-22. This assessment
included an evaluation of the generic implications of each finding as well as
the specific resolution. The secondary purpose of the inspection was to
perform a programatic review of the licensee's design-basis reconstitution
program.
3. Personnel Contacted
The following is a listing of key personr.el contacted during the inspection: <
Name Position
J. Fisicaro ,
Supervisor, Nuclear Regulatory and Industry Affairs (NRIA)
5. K. Gambhir' Section Manager, Generating Station Engineering (GSE)
H. M. Tackett Consultant to NRIA
M. Eidem Manager, GSE Mechanical and Nuclear i
D. Mangan Consultant, GSE
R. Parsons Design Basis Project Engineer, GSE
D. Deboer Engineer, GSE
R. Lewis Supervisor, Mechanical Engineering, GSE
H. J. Faulhaber ilectrical Engineering Manager, GSE
S. Miller Battery Engineer, GSE
D. Morris Load Coordinator, GSE
R. Ronning Cable Engineer, GSE
R. Clemens Cable Engineer, GSE
W. C. Gartner Lead Electrical Design Engineer, GSE
J. Tucker Sr. Design Engineer Electrical Engineering, GSE
S. Crites Sr. Designer, Mechanical Engineering, GSE
J. Lechner Fire Protection Engineer, Fort Calhoun Station
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4. Sumary of Open Items / Licensing Issues
The NRC inspection team reviewed a large number of calculations, drawings,
specifications, and other design documents during the 10-day inspection as
further detailed in this report. On the basis of this .eview, the team was able
to close 39 of the 47 original findings identified in Inspection Report
50-285/85-22. The findings that remain open are 02.1-1 and D2.2-3 in the
mechanical syst ems disciplire, U3.1-3 and D3.2-7 in the mechanical components
discipline, D4.3-1 and U4.4-1 in the instrumentation and controls discipline,
and 05.1-1 and 05.2-1 in the electric power systems discipline.
Attachment A provides a comprehensive description of the resolution and status
of all the SS0MI design findings in Inspection Report 85-22 including a
description of the correction actions necessary to resolve the oper. Items.
The two mechanical system items that remain open identify concerns relating to
the adequacy of the functional testing performed on the air accumulators for
air-operated valves. These air accumulators are attacned to the instrument
air lines to allow various safety-related air-operated valvae, to achieve a
safety-related position in the event of loss of the instrument air system.
Through review of the actual test procedure and test results, the inspection
team found that the licensee did not adequately perform postmodification
testing to ensure that the accumulators and their associated check valves
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would function as designed. Specifically, the functional test did nut dupli-
cate accident conditions, nor were the test data reviewed to correlate them
to accident conditions. Also, the test instrumentation was not arranged to
measure initial accumulator pressure and system leakage. To resolve this
finding the licensee is asked to review and validate all air accumulators
tested to date, complete its program for evaluation of the use of air accumula-
tors including the development of functional test criteria and surveillance
testing, and provide the necessary training to all affected personnel.
The two mechanical component items that remain open are associated with main-
tenance of the Updated 3afety Analysis Report (USAR) and corrpliance with USAR
commitments. The team found that the licensee's alternate seismic criteria for
small-bore piping'need to be submitted to the NRC for formal review since they
relax current USAR comitments (see Unresolved Item U3.1-3 for details). Also,
the team noted that the licensee is not complying with the USAR criteria
associated with valve actuator seismic accelerations, nor has it developed an
alternate criterion. As a consequence valve actuator seismic accelerations have
not been considered and valve operability during and after a seismic event is
uncertain (see Deficiency 03.2-7 for details).
The two instrumentation and controls items that remain open involve noncom-
pliance with a USAR conmitment and relaxation of applicable code requirements.
First, the team found that the licensee was not meeting a USAR comitment to
nrovide position indication for containment isolation valves during all modes
of plant operation incluoing accident scenarios. Second, the team questioned
the licensee's basis for physical separation of control wiring within panels.
Specifically, the licensee's original pNetice of comen routing of redundant
safety-related divisions and comraon routing of safety-related and nonsafety-related
wiring was inappropriately continued for modifications to safety-related panels
that were recently procured to the requirements of Regulatory Guide 1.75 and
Institute of Electrical and Electronics Engir eers (IEEE) Standard 304. The team
believes that such modifications may represent an unacceptable relaxation of the
criteria stipulated when the panels were procured.
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The two electrical power system open items involve nonconservatims in the battery
sizing calculation and the fire wrap cable ampacity derating calculation, in
regard to the former, the licensee neglected to consider battery end-of-life
effects. To resolve this issue, the licensee has comitted to obtain the
battery manufacturer's assistance to establish the capacity remaining in the
batteries. To resolve the latter open item, the licensee needs to correct the
cable derating calculation to include proper consideration of a specific manu-
facturer's cable data, application of the correct thermal conductivity for the
fire wrap, and justification for one power feecer cable to a motor control
center being above the allowable ampacity.
In addition to the eight open items remaining from the original SSOMI, four of
the items that were closed for the purpose of this inspection remain open
licensing issues to be resolved by the licensee and NRR's project directorate.
These four open licensing issues involve noncompliance with USAR seismic design
requirements for valves, electric motor operators, junction boxes and pressure
switches and the adequacy of the piping analyses performed in response to NRC
Bulletin 79-14 See Open Items 03.1-4, D3.2-4, 03.2-5, and 03.2-8. These
issues are discussed below,
in regard to Observation 03.1-4, the inspection team found that the seismic
qualification of certain safety-related components was not in accordance with
the current licensing comitments in Appendix F of the USAR. The team found
that the seismic design or test criteria for various valves (e.g., motor-
operated, air-operated, ball, manual gate and globe, relief, safety) and
electric motors specified in Omaha Public Power District Contract 763 were not
provided in their associated design specifications. The licensee infonned the
team that the lack of seismic design criteria for valves, motors, and other
components would be resolved as a part of Unresolved Safety Issue (USI) A-46.
However, during a teleconference between the inspection team, members of NRC's
Mechanical Engineering Branch (MEB), the NRC project manager and the Region IV
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staff, the team was informed that MEB's program to resolve the seismic qualifi-
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cation issues addressed in USI A-45 assumes that safety-related equipment
confonns with the USAR comitments. In a letter from NRC (J. Calvo) to OPPD
l (K. J. Morris) dated July 28, 1988, the NRC staff accepted, subject to certain
l conditions. OPPD's proposal to delay the resolution of seismic qualification
of equipment until the resolution of USI A-46.
The original issues associated with Findings 03.2-4 and 03.2-5 documented
that the Unistrut supports for the junction box to valve YCV 1045B did not
meet the seismic provisions of USAR Appendix F; and the replacement pressure
switches in Modification Request MR-FC-83-83 warranted a more thorough analysis
to substantiate seismic qualification, respectively. The team reviewed ano
found acceptable these specific issues during this inspection of the imple-
mentation of corrective actions. However, the licensee stated that the generic
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issue of seismic qualification of all other safety-related junction boxes and
pressure switches will be included in its program for resolving USI A-46.
Therefore, it is necessary that this comitment be addressed by the licensee in
its submittal pertaining to the A-46 program.
During the programatic review of the licensee's design-basis reconstitution
program for the auxiliary feedwater system, the team identified concerns
relating to the adequacy of the licensee's 79-14 program (Deficiency D3.2-8).
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l The intent of this inspection was not to review the licensee's 79-14 program.
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However, in reviewing the design-basis reconstitution program, the team tried
to deterinine whether certain design attributes in the area of piping analysis
would be addressed by the design-basis reconstitution program and whether the
licensee had data available to support these attributes. The team identified
three design attributes that should have ' een reviewed under the 79-14 program,
but were no'.: 0) consideration of seisnac anchor movements between struc-
tures, (2) consioiration of equipment nonle thermal movements, and (3)
considera. ion of friction loads for pipe supports. in a very limited review
of piping analysis attributes, three problems were found. Therefore, the team
believer that a more detailed evaluation of the 79-14 program performed for the
licenset by Gilbert /Comonwealth should be conducted by NRC.
5. Assessmerc of Changes Since SSOMI (85-22)
a. Mech)nical Systems
Since the 550MI, the licensee has initiated actions that should substan-
tially improve modification packages if implemented in accordance with
procedures. These actions include the following:
(1) issuing general engineering guidelines for the preparation of specifi-
cations, design packages, installation and test procedures, and
10 CFR 50.59 safety evaluations
(2) strengthening of Standing Order G-21 to ensure the timeliness of
third-party review for normal modifications
(3) developing a new administrative procedure to govern the use and
documentation of engineering judgment in place of detailed calcula-
tions or design analyses during the preparation of modification
packages
(4) revising Administrative Procedure A-5 to clarify that technical
exceptions from the recomended bidder must be formally withdrawn
by the bidder or the the licensee's purchase .'ccument and/or the
specififation must be reconciled to include the exceptions
(5) revising Administrative Procedure A-11 to control calculations
and to require tb indexing of all calculations
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(6) strengthening of Design Procedure B-11 to require multidisciplinary
l review of modification packages
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j The team reviewed these changes and found that they address the weaknesses
- in the licensee's modification control program originally identified by the
! SSOMI team. However, continued weaknesses exist in the use of the USAR as a
i source of design input and in sostmodification testing. In recognition of
this weakness, a design-basi! recons'.itution program has been initiated by
the licensee. Interim use of the USAR until design-basis documents are
available is acceptable considering the precautions included in the various
implementing procedures regarding the need for reliable design input.
Post-modification testing continues to be a weakness. During the reinspec-
tion the team reviewed functional testing performed on air accumulators and
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identified errors that suggest that inadequate attention to proper functional
test and appropriate acceptance criteria is still a weakness,
b. Mechanical Components
The team concludes that the licensee's program, which it is developing, should
enhance its design staff's ability to implement design modifications to
piping, equipment, and supports in a controlled manner. The team's assessment
is based on its review of
(1) the licensee's responses to the deficiencies, unresolved items,
and observations which the team documented in NRC
Inspection Report 50-285/85-22
(2) the licensee's program to regenerate the design basis for the safety-
related pipi19 systems at Fort Calhoun
(3) General Engineering Guide GEG-3, "Preparation of Design Packages"
With respect to item (1), the team specifically noted that to address
Deficiency D3.1-1, the licensee withdrew an uncontrolled des.g i specification
and engaged Stone & Webster to regenerate the system design basis for the
safety-related piping systems at Fort Calhoun. The team also noted that to
address Deficiency D3.1-2, the licensee withdrew uncontrolled design tempera-
ture data and engaged Applied Power Associates to document the operating and
accident temperatures for the safaty-related piping systems at Fort Calhoun.
With respect to item (2), the team believes that the licensee's commitment
to regenerate the design basis for the safety-related piping systems at Fort
Calhoun will provide a controlled and retrievable design basis for the safety-
related piping, equipment, and supports installed at Fort Calhoun.
With respect to item (3), the team reviewed General Engineering Guide GEG-3,
"Preparation of Design Packages," which the licensee engaged Sargent & l. undy
to prepare. ,T,he team believes that the procedure if properly followed,
should erable the licensee's design staff to implement design modifications
in a co., trolled manner.
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In ramary, despite its concerns documented elsewhere in this inspection
report, the team believes that the licensee has made measurable progress since
tne last SSOMI by committing to create a controlled and retrievable design
basis for safety-related piping, equipment, and supports and by enhancing the
ability of its design staff to implement design modifications in a controlled
manner,
c. Instrumentation and Control
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l The i!am noted several areas of programmatic improvement relative to the
l initial SSOMI findings. The licensee has developed more explicit proce-
i dures which should promote better engineering and control of design changes,
i provided personnel are adequately trained and the procedures are consistently
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tollowed.
One notable example of this improvement was the newly developed General
Engineering Guide GEG-3, "Preparation of Design Packages." This document
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appears very thorough in addressing the necessary technicol attributes of
a design change End encourages an engineering thought process in developing a
modification. Before this guide was developed, design attributes were treated
more superficially in Design Procedure B-2.
For example, GEG-3 provided an outline of a root cause detenmination
process, which should promote better problem definition and design-basis
development for changes. It also provided a comprehensive list of design
attributes for consideration, with specific reference to supporting
engineering guides, e.g., human factors review and ALARA (as low as
reasonably achievable) analysis guidelines. Guidance also was provided
for systems interaction analysis and 10 CFR 50.59 reviews. The structure,
as well as content of the document, was in a practical, readily usable
form; it appeared less likely that significant attributes would be omitted
in the change process, if GEG-3 were conscientiously used.
The team noted that several of the original SS0MI findings might have been
less likely if all of the guidelines had been available for developing the
modifications inspected. For example, the definition of environmental and
process conditions for procurement specifications might have been nm
accurate and consistent; separation requirements and design basis might
have been more evident to the preparers of the modifications; the defini-
tion and solution of the problem for the containment isolation valve limit
switches in Deficiency 04.3-1 might have been more consistent.
The team encouraged the licensee to complete these engineering guides,
which together with a documented design basis should promote significant
improvements to the modification process,
d. Electric Power Systams
Programs were just being established that would ensure a detailed review
and verification of future modifications. These new programs also would
require interdiscipline review. General Engineering Guideline GEG-9,
"Electrical Systems Interaction" prepared by Sargent & Lundy in February
1988, had not'been issued formally but promised to provide strong guidance
to the design engineer. Recently prepared General Engineering Guideline
GEG-27, "Safety Evalu stions," and GEG-28, "Preparation of Installation and
Test Procedures," if properly implemented could have eliminated many of
the original observations in NRC Inspection Report 85-22. Revisions to
Standing Order 21 "Station Modification Control," Revision 29, November
1987, and the issuance of GEG-3, "Preparation of Design Packages," November
1987, should strengthen the modification control procedure. Revisions to
Design Procedure B-11, "Independent Design Verification," February 1988,
should also ensure the accuracy of the design packages.
6. Design Basis Reconstitution Program (DBRP)
Recognizing that tne licensee was in the initial stages of developing its DBRP
and that the f*C has not established specific guidance for such programs, the
team perfortred a programmatic review of the DBRP, on a sampling bat s, to hssess
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whether all the necessary design-basis documentation had been identified (but
not necess'rily regenerated) within the mechanical components and instrumentation
and control (l&C) disciplines. The team sampled some of the available products
of the program to assess the completeness of necessary design attributes, but
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did not attempt to assess the depth and technical adequacy of the identified
design-basis documents.
The DBRP is intended to establish comprehensive and up-to-date compilation of
design records. It will document the original design basis, eliminate the need
for using the USAR as a design-basis document, provide documentation of calcula-
tional inpi ts and assumptions, provide licensing comitment tracking, and
eliminate the use of uncontrolled documents. There are two types of design
basis documents: system design-basis documents and plant-level design-basis-
documents. The system design-basis documents are documents which contain a
comprehensive listing of design information for a specific system. Plant-level
design-basis documents are documents which contain generic design information
applicable to multiple systems or which are not system related but contain
design basis infonnation for the facility (e.g., pipe stress criteria, seismic
criteria, welding criteria).
The team's review included the scope of the DBRP as reflected by the current
list of candidate systems, a sample of the system level design-basis document
(DBD) for the auxiliary feeriwater system, and the intended scope of forthcoming
plant-level DBDs. The review of the plant-level DBDs in the !&C area consisted
primarily of interviews and discussions with the team's counterparts to assess
their intentions, since details of the plant-level DBDs generally were not yet
available. In the mechanical components area, the team assessed the licensee's
ability to pri. pare the plant-level DBD in accordance with USAR comitments
by compiling a list of design loads and asking the licensee to confirm that
each load type was controlled and retrievable through its document retrieval
system, since plant-level DBDs had not yet been prepared.
Regarding the scope of the program, the team noted that the following safety-
significant systems seemed to be missing: the reactor protection / engineered
safety features actuation system, the loose parts monitoring system, and the
offsite dose assessment information processing and display system as described
in hUREG-0654. Discussions with the licensee indicated that it also intends
to include instrumentation specified in Regulatory Guide 1.97 and h0 REG-0737
in the program, ev,en though they are not explicitly listed.
Regarding the auxiliary feedwater system DBD, the team was generally impressed
by the overall structure and content of the document and the number of attri-
buter addressed in depth. As one example, the team liked the presentation of
design requirements in addition to that of design implementation; this approach
seemed effective in capturing subtle design insights that could be lost if the
"why" as well as the "how" of the design were not adequately documented. The
team did identify some design attributes it expected to see, but was unable to
find, in the auxiliary feedwater system DBD. These attributes included
requirements for instrument channel range, accuracy, and repeatability;
requirements for alarms; requirements for trending, recording, and archiving
data; and interlock requirements imposed by equipment vendors to assure proper
- operation of the equipment.
Regarding the plant-level DBDs in the I&C area, the team understood that
several engineering guides, standards, analyses, and studies / evaluations were
planned. On the basis of its discussions with the licensee, the team expected
that in addition to attributes explicitly listed in the program documents, the
following were or would be included in the program:
(1) instrument channel uncertainty calculations and techniques
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(2) design basis for internal flooding (e.g., pipe breaks)
(3) outdoor-temperature ranges (the team understood that no safety-related
instrumentation and controls were exposed to the outdoor enviroment)
(4) evaluation of the plant annunciator system to forra a design basis for the
system
(5) I&C grounding and shielding
(6) inclusion of fuse type and fuse irating in the planned instrument bus
load study
(7) instrument and tubing installation specifications
In reviewing the plant-level DBDs in the mechanical components area, the team
found that the auxiliary feedwater (AFW) piping analyses of rer.ord did not
incorpordte an evaluation of the following items, which are detailed in
Deficiency D3.2-8:
(1) a USAR commitment to consider the effects of relative seismic displace-
ments between the containmet,t shell and the auxiliary building on the
connected AFW piping,
(2) consideration of the effects of the steam generator nozzle thermal
displacements on the connected AFW piping
(3) consideration of the effects of the AFW pump turbine inlet nozzle thermal
displacements on the connected piping
(4) consistent consideration of friction forces for pipe support design
As a result, the team concludes in the mechanical component's area that the
licensee's current ability to prepare plant-level CBDs such t.s CS-51, "Seismic
Criteria," and ME,10. "Pipe Stress and Supports," was hampered by the lack of
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controlled documents that specified code of record and design criteria and by a
j USAR that contained several provisions considered obsolete by the licensee.
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l In general, the team was favorably impressed by the licensee's efforts to date.
! It also sensed from its discussions that the licensee seems motivat*d to
establish a thorough and useful design basis to be maintained for the life of
the plant. ;
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ATTACHMENT A
STATUS OF FINDINGS
FINDING NO. STATUS TITLE
Mechanical Systems Discipline
02.1-1 Open La;k of Design Analysis To Support Sizing of
Air Accumulators for Valves YCV-1045 A/B
D2.1-2 Closed Seismic Requirements not Specified in
MR-FC-83-158 Procurement Documents
02.1-3 Closed Vendor Exceptions to Specifications Not
Reflected in Prw urement Document
2.1-4 N/A Item Number Not Used
02.1-5 Closed Procedural Error Caused Seismic and Stress
Analysis for MR-FC-83-158 Not To Be Filed in
Modification File
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D2.1-6 Closed F611ure To Follow Procedural Requirements
for a Normal Modification Resulting in Lack
of Required Design Verification Review
D2.1-7 Closed Incomplete Intta11ation/ Testing P.ocedure in
Construction Package for MR-FC-83-158
02.1-8 Closed Incorrect Infonnation on Flow Diagram for <
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Main Steam System
D2.1-9 ,
Closed Incorrect System Description Statements
U2.1-10 Closed Use of Fluorocarbon-Elastomer Material in
High Radiation Environments
D2.2-1 Closed Incorrect Design input in Calculation
Associated With MR-FC-81-218
i D2.2-2 Closed Incomplete Consideration of CQE and Seismic ;
l Class 1 Requirements for Portions of
MR-FC-81-21B
D2.2-3 Open incomplete Installation / Testing Procedure
Performed for MR-FC-81-21B
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02.2-4 Closed Incomplete Modification File for a Completed
i Modification
D2.2-5 Closed incorrect Information on Instrument Air
Diagram
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FINDING NO. STATUS TITLE
D2.2-6 Closed 10 CFR 50.59 Safety Evaluation Based Upon
Incorrect Assumption and Analysis Methodology
MechanicalComponentsDischline
D3.1-1 Closed Balance of Plant Design Specifications
D3.1-2 Closed Design Temperatures for Safety-Related Piping
U3.1-3 Open Small Bore Pipe Support Spacing
03.1-4 Closed * Seismic Qualification of Valves Installed in
Class I Piping Systems
U3.2-1 Closed M -FC-84-61 Design Input Source and Use
D3.2-2 Closed MR-FC-83-158 Installation Procedure
D3.2 3 Closed MR-FC-84-162 Calculation
D3.2-4 Closed * Junction Box Supports
03.2-5 Closed * Containment Pressure Switch Seismic
Qualification
D3.2-6 Closed Sceam Generator Nozzle Dams
D3.2-7 Open YCV a%5B Valve Restraint
D3.2-8(new) Closed * Auxiliary Feedwater Piping Analysis Design
input Loads
Instrumentation and Controls Discipline
04.1-1 Closed High iower Rate of Change Trip Bypass
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04.2-1 Closed belta T Power Loop Analysis
D4.3-1 Open Limit Switch Circuit Protection by Fusing,
MR-FC-84-74A
U4.3 2 Closed ESF Bypass Switch Keylock Provision,
MR-FC-31-102
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04.3-3 Closed Procurement Requirements on Equipment Vendors
U4.4-1 Open Design Basis Physical Separation Within Panels
C4.5-1 Closed Drawing Changes by Procedure A-9, MR-FC-82-178
- Finding closed for this inspec'.lon report, but the associated licensing issue
needs to be resolved between the licensee and NRR project directonte
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FINDING NO. STATUS TITLE
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04.5-2 Closed Flow Element Design Basis Conditions
V4.5-3 Closed Battery Room Fire Hazard Analysis
Electrical Power Systems Discipline
05.1-1 Open Battery Sizing Calculation f
US.1-2 Closed Battery Charger /DC Bus Coordination
05.1-3 Closed Power Cable Sizing Criteria
05.1-4 Closed Pre-Operational Test Requirements
05.1-5 Closed Inverter Sizing Without Analysis '
05.1-6 Closed Design Interface Control
05.2-1 Open Fire Wrap Protection for Cable Raceways
Design Change Control
06.1-1 Closed Safety Evaluation for Non-Safety-Related
Systems Described in the USAR
06.1-2 Closed Safety Analyses for Emergency Modifications
06.1-3 Closed Vital AC Inverter Bypass Mode
06.2-1 Closed Untimely Closecut of Emergency Modifications
06.2-2 Closed Modifications to AFW Turbine Steam Supply Valves t
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(0 pen) Deficiency 02.1-1, Lack of Design Analysis To Support Sizing of
Air Accumulators for Valves YCV-1045 A/B
BACKGROUND
When the modification request, MR-FC-78-43 was initiated in September 1978, an
air accumulator sizing calculation was not performed to demonstrate that a
sufficient stored volume of pressurized air would be available to close valves
YCV-1045 A and B assuming a loss of instrument air and minimum initial accumu-
lator pressure. In response to the finding, the licensee functionally tested
the accumulators to ensure that their size was adequate.
STATUS OF FINDING
During the reinspection, the team reviewed the functional test, performed on
January 5, 1986, to assess the adequacy of design. The test had the following
weaknesses and differed from the modification final design package as follows:
(1) The test acceptance criterion was based on holding valve YCV-1045 A or
B shut for 30 minutes, even though the final design description of the
modification package indicated that the accumulators were designed to
supply air to keep the valve closed for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The team was informed
that the time was relaxed from I hour to 30 minutes on the basis of an
engineering judgment that the distance from the control room to the
location of the valves, in Room 81, is very short. The design engineer
discussed the issue with operations personnel, and they concluded that 30
minutes was long enough. Field Change #2 relaxed the hold time and when
it was approved an oversight occurred so that the final design description
was not updated.
(2) The functional test did not duplicate the accident condition, and the
acceptance criterion was not altered accordingly. Specifically, valves
YCV-1045 A and B were not shut against system pressure and the worst-case
lowest accumulator pressure was not duplicated. In spite of these short-
comings, the licensee determined that the existing postmodification test
produced satisfactory results by compring the minimum air pressure
necessary to close and hold the valve in that position for 30 minutes with
the initial accumulator pressure corrected for worst-case conditions.
A related issue was the adequacy of operating procedures to warn of the
potential loss of instrument air to the valve operators. Abnormal Operating
Procedure AOP-17. "Loss of Instrument Airg" Revision 5, February 11, 1986,
contained a note warning the operator that valves YCV-1045 A and B are equipped
with air accumulators and remain operable for at least 30 minutes, after which
they fail open on loss of accumulator air reserves. The team noted that
Emergency Operating Procedure E0P-04, "Steam Generator Tube Rupture," Revision
02, February 1, 1988, and Abnormal Operating Procedure A0P-24. "Steam Generator
Tube ,1upture (PPLS Blocked)," Revision 1, October 8, 1986, did not have a
similar warning. The team was informed that Emergency Operating Procedure
E0P-20. "Functional Recovery Procedure," Revision 3. March 14,1988, referred
the operator to Procedure A0f-17. Procedure E0P-20 provided the operator
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actions for events during which a diagnosis was not possible, for two or more
events occurring simultaneously, or for events during which emergency guidance
was not available. The team confirmed that under Maintenance of Auxiliaries
Procedure MVA-3, "Recovery of Instrument Air," Revision 2 March 14, 1988, the
operator was referred to Procedure A0P-17. Because Procedure E0P-20 referred
the operator to the AOP containing the caution, the team's concern is satisfied.
Another related issue was the testing of the instrument air check valves which
isolate the accumulator and associated air-operated valve from the instrument
air system upon a loss of instrument air. The licensee has submitted a revised
inservice inspection program plan, dated December 15, 1987, to the NRC and has
identified the two check valves (Tag Nos. IV-1045 A-C and IV-1045 B-C). The
revised plan requires these valves to be functionally tested to the open and
closed positions.
This item will remain open pending the (1) completion of a formal calculatior,
to document the adequacy of the postmodification test when corrected for
worst-case conditions and (2) revision of the final design description for
modification MR-FC-83-158 to correct incorrect information.
PELATED GENERIC ISSUES
See Deficiency 02.2-3 regarding the generic issues related to the adequacy of
the overall testing program for the air accumulators and the actions by the
licensee required for closure.
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(Closed) Deficiency 02.1-2, Sesimic Requirements Not Specified in
MR-FC-83-158 Procurement Documents
BACKGROUhD
The seismic requirements for the manual and check valves for modification
MR-FC-83-158 were omitted from procurement documents even though these valves
in conjunction with air accumulators and associated tubing serve a post-
accident function to close valves YCV-1045 A and B. At the time of the 1985
SSOMI, neither design analysis nor documented engineering judgment existed to
substantiate that the subject valves were seismically qualified.
STATUS OF FINDING
During the reinspection, the licensee provided documentation of the engineering
judgment that NUPRO check valve M/N SS-6C-10 and small manual valves purchased
for the above modification were seismically qualified (GSE-FC-88-596, memorandum
from M. E. Eidem to MR-FC-83-158 file, April 5, 1988). The team reviewed this
engineering judgment and found that it was an acceptable basis for concluding
that the valves were seismically qualified by similarity for approximately
6g's. However, the licensee could not justify the seismic qualification for
100g's as stated in the April 15, 1986 response to the original SSOMI report.
Since the subject valves are qualified for the their intended purpose, the team
suggests that the licensee carefully review future submittals with an increased
attention to accuracy and less attention to dramatic emphasis.
RELATED GENERIC ISSUES
A generic issue identified during the SSOMI was the use of undocumented
engineering judgment in lieu of design analyses. During the reinspection, the
licensee issued a new policy A-14, "Use of Engineering Judgment," Revision 0,
April 1988. This policy outlines the requirements and documentation necessary
when using engineering judgment in place of detailed calculation or analysis in
preparing modification packages, 10 CFR 50.59 safety evaluations, and
engineering reports. This procedure, when implemented, should improve trace-
l ability of final oesign back to design input. This new policy satisfies the
team's concern.
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See Observation 03.1-4 for the generic issue related to equipment seismic
qualification.
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(Closed) Observation 02.1-3,VendorExceptionstoSpecificationsNotReflected '
in Procurement Document
BACKGROUND
Although the supplier of components for a safety-related modification took
exception to the storage requirements of a procurement specification,
documentation did not exist of an engineering review and acceptance of the
vendor's exception. The observation suggests that the vendor's exception
should have been evaluated and, if acceptable, the specification should have
been revised to reflect the acceptable alternative.
STATUS OF FINDING
During the reinspection, the licensee revised Administrative Procedure A-5,
"Procurement of Material and Labor," Revised April 1988. This revision speci-
fied that the technical exceptions received from prospective suppliers should
be addressed in the documented evaluation of bids. For a specification
pertaining to critical quality equipment all technical exceptions from the
recomended bidder must be formally withdrawn by the bidder or from the licen-
see's purchase document and the specification must be reconciled to include the
exceptions. In addition, all changes to purchase documents and specifications
must be reviewed and approved in the same manner as that required for the
original document. This revision to Administrative Procedure A-5 is responsive
to the team's concern.
RELATED GENERIC ISSUES
There is no related generic issue because the specific issue did not result in
a safety concern and the programrutic change addresses the general issue.
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(Closed) Observation 02.1-5, Procedural Error Caused Seismic and Stress
Analysis for MR-FC-83-158 Not To Be Filed in
Modification File
BACKGROUND
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During the 1985 SSOMI, the team found that a seismic and stress analysis
"Generic Air Accumulators Using Propane Tanks Built to 00T Spec. 4BA-240,"
Revision 0, August 1985, for the air accumulators was not filed with the
modification package. At that time, the licensee did not maintain calcula-
tions as living documents i.e., calculations maintained in a calculation file
and kept current as plant system, structures, components were modified through-
out the life of the plant. Instead, calculations were done, as required, for
each modification and the modification file was the only controlled location
for retention of design calculations. The team made the observation that a
better control process for calculations would allow enhanced retrival, use, and
revision of design calculations.
STATUS OF FINDINGS
During the reinspection, the team reviewed Administrative Procedure A-11,
"Calculation Numbering and Revision Control." Revised March 1987. The proce-
dure was revised to require the indexing of all calculations, with a separate
log book for each station both fossil and nuclear. In addition, the main frame
computer could be searched using the GSE-1 search program to identify the file
location of a calculation. Current and past revisions of calculations were
available on microfilm in Document Control. The team examined the log book for
the Fort Calhoun Station and witnessed a demonstration of the GSE-1 search
program for calculations.
The current method of filing calculations should onsure that calculations can
be retrieved and revised. The team observed that old calculations had been
added to the index as well as new ones. The computer aided retrieval system
appeared to be a very effective tool.
RELATED GENERIC 155VES
The related generic issue of control of calculations is addressed above.
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(Closed) Deficiency D2.1-6, Failure To Follow Precedural Requirements, for
Normal Modification Resulting in Lack of Required
Design Verification Review
BACKGROUND
Normal modification MR-FC-83-158 was not treated in a manner consistent with
the requirements of Design Procedure B-2 "Production of Design Description and
Evaluation," Revised January 1984. During the 1985 SSOMI, the team found that
a construction package was prepared even though the verification of the final
design package had not been completed. This situation was further aggravated
by the design engineer who determined that the construction package did not
require third-party review and who signed a memorandum for the department manager
stating that a third-party review was not required.
STATUS OF FINDING
Although the licensee maintained that no deficiency cxisted, the team confirmed
during the reinspection that the following actions had beea completed.
Revision 27 to Standing Order G-21, dated April 10, 1987, was issued and
included the following changes:
(1) Definition 1.4.25 was included to address independent design verification.
(2) Paragraph 5.6.8 was revised to include the requirements for completion
of the independent design verification before the contruction package is
approved by the Manager, Fort Calhoun Station. In addition, this para-
graph was revised to state that all field changes or procedural changes
and any additional calculations or analyses must be completed before the
system is accepted.
These revisions addressed the team's concern that construction of a normal
modification will not occur before third-party review is completed unicss
approved by site and engineering management.
i Administrative Procedure A-2, "Modification Request Development," was revised
in February 1988 to include the following change:
(1) Section 2.4 requires that design verification of all design documents and
installation and testing procedures be completed before the final design
,
package is accepted by the Manager, Fort Calhoun Station. Exceptions may
l be granted by joint agreement between the Manager, Fort Calhoun Station and
the Generating Station Engineering (GSE) section manager for certain
specialized cases,
i for the purposes of construction package approval and system acceptance, any of
l the following is considered a completed third-party review (independent design
i verification):
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l (1) Third-party review in accordance with GSE Procedure B-11 that is completed
j and signed off as being "in compliance."
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(2) Third-party review in accordance with GSE Procedure B-11 that is completed
and signed off as being "in compliance except as noted" provided reviewer's
comments have been reviewed and resolved by the design engineer. To
resolve comments requires a field change / procedural change incorporating
any changes suggested by the third-party reviewer and a letter written by
the design engineer to the GSE Manager providing coment and resolution.
(3) A letter from the design engineer to the department manager stating that
any field changes will not result in design deficiencies that will
preclude the modification from performing its intended function. In
this case, the department manager will review the letter and document
the approval. Examples are field changes that are justified by engineer-
ing judgment based on safety margins provided in the original design or
because of their similarity to other approved designs. This option is
used for design verification of emergency modifications. This approach
does not mean that third-party reviews are not performed; instead it means
that the design engineer will ensure that the third-party reviews have
been completed for each field change and that their net effect will not
cause a design-deficient condition to exist.
Design Procedure B-22, "Independent Design Verification," was Revised February
1988 to include the following changes:
(1) Paragraph 4.2 was amended to require all modifications to have a
multidiscipline review.
(2) Paragraph 1.0 was clarified to indicate that independent multidisciplinary
design verifications will be dore for design changes including, those
pertaining to critical quality equipment (CQE), limited CQE, non-CQE, and
fire protection system and components.
Although the licensee in response to inspection Report 50-285/85-22 maintained
that the observation was not a deficient condition or violation of its proce-
dures, the need for performing design verifications has changed and the method
used for performing these reverifications has been improved. On the basis of a
programmatic revidw of the above-referenced procedures, the team's concerns
have been satisfactorily addressed; therefore, this item can be closed.
RELATED GEtiERIC ISSUES
The related generic issue was addressed in conjuction with the resolution of
the aforementioned finding. ;
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(Closed) Deficiency D2.1-7, Incomplete Installation and Testing Procedures
in Construction Package for MR-FC-83-158
BACKGROUND
The postmodification testing procedure did not provide for the testing of the
design function of the air accumulators to shut valves YCV-1045 A and B against
a differential pressure of approximately 1000 psig. During installation, these
valves were closed and never cycled as part of postmodification testing. The
air supplied by normal instrument air header was used to pressurize the actua-
tors instead of air supplied by the accumulators alone. In addition, no
acceptance criterion defined acceptable air leakage. After the NRC audit, a
new functional test was performed.
STATUS OF FINDING
Weaknesses identified in the functional test performed on January 5, 1986,
are identified in STATUS OF FINDING for Deficiency D2.1-1. The functional test
corrected the weakness stated in the deficiency with respect to initial closure
of valves YCV-1045 A and B with the normal instrument air supply; however, it
did not contain adequate acceptance criteria to confirm that the modified
system would perform under worst-case accident conditions.
Revision 17 to Standing Order G-21, dated April 10, 1987, was issued to
(1) reference Standing Order G-19 for testing requirements; (2) define the
requirement for stating the postmodification system / component performance
-
requirements for preparing, performing, and evaluating the test; (3) specify
the person responsible for evaluating modification-related test results; and
(4) ensure that systems are tested and test results are approved before the
systems are returned to service after they have been modified,
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The operations engineer and the onsite review committee are responsible for
ensuring that the postmodification system / component performance requirements
are adequately stated, sufficient steps are included for perfoming the test,
responsibility for evaluating the test results is clearly defined, and adequate
assurance exists that testing results are approved before a system is returned
to service following a modification.
On the basis of the revision to Standing Order G-21, the team finds that
procedural requirements appear to exist to prevent the use of incomplete
installation and testing procedures. Weaknesses in the functional test will be
corrected as part of closure of Deficiency D2.1-1; therefore, this item can be
closed.
RELATED GENERIC ISSUES
The weaknesses in the second functional test contributed to the team's concern
that implementation of postmodification testing requirements at Fort Calhoun is
generally weak and, therefore, additional licensee attention is needed.
Specific action is described as part of Deficiency 02.2-3, which is still open.
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(Closed) Deficiency D2.1-8, Incorrect Information on Flow Diagram for Main
Steam System
BACKGROUND
During the 1985 SS0MI, the team identified an error'on the drawing showing
the piping arrangement associated with the main steam isolation bypass valves
and the auxiliary feedwater steam warmup lines. The drawing indicated an
incorrect arrangement where the piping to the bypass valves taps off the
upstream side of the disc and returns to the upstream side, instead of the
correct arrangement where the return is between the main steam isolation valve
and its associated reverse flow check valve. This error apparently caused the
piping connected downstream of the bypass valve to be indicated as being
nonsafety-related.
STATUS OF FlhDING
During the reinspection, the team reviewed Drawing 11405-M-252, "Flow Diagram
Steam," Revision 49, dated February 17, 1988, and found that it correctly
shows the piping arrangement associated with the bypass valves and the auxiliary
feedwater steam warmup lines.
RELATED GENERIC ISSUES
On the basis of its review, the team finds that this error appears to be
isolated and does not warrant a broader review.
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(Closed) Deficiency D2.1-9, Incorrect System Description Statements
BACKGROUND
During the 1985 SS0MI review of various modification packages, the team
examined the licensee's system descriptions to confirm system design bases and
found errors in three system descriptions. In response to this deficiency, the
licensee stamped all volumes of the system descriptions as "Uncontrolled
Document, For Information Caly."
STATUS OF FINDING
,
During the reinspection, the team confirmed that Notebooks 51 and S4 containing
Volumes 1 through 3 of system descriptions had been stamped "Uncontrolled '
Document For Information Only." On the basis of the evidence that notebooks
containing old system descriptiens are marked for information only, this item
is closed.
RELATED GENERIC ISSUES '
x related generic issue is the inability to access the design-basis information
when impleu nting design change. One of the goals of a Design-Basis Reconstit9-
tion Project, a program initiated by the licensee, is to organize design-basis
records in such a way that a set of system-oriented design-basis documents (DBDs)
can be generated froin these records. DBDs will be developed for safety systems
and systems that may affect operation of safety systems. In addition, plant-level
documents also will be developed to address such generic subjects as seismic and
fire protection. These DBDs will reflect the current design condition of the
plant, combined with a limited historical perspective and the justification for
the current plant configuration or generic subject area. The DBDs will be
controlled documents to be revised as the plant configuration changes as a
result of modifications.
The team briefly examined the Program Plan for Design-Basis Reconstitution
Project, Revision 2. dated July 2, 1987. This program, if implemented in
accordance with the objectives stated in the Program Plan, should result in a
complete set of new design-basis documents. This program is scheduled to
produce approximately 10 plant-level documents and 30 system-level documents
and is expected to be completed by April 1990.
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(Closed) Unresolved Item U2.1-10. Use of Fluorocarbon-Elastomer in a High
Radiation Environments !
BACKGROUND
The procurement specification for safety-related instrument air check valves
associated with rnodification MR-FC-83-158 pemitted the use of Viton as a
seat material. Normally this material is not used in high-radiation environ-
ments. The original specification scecified Bena "N" as a seat material;
however, the valve's supplier took exception to the use of this seat material
and stated in a letter that Viton would be supplied instead. On the basis of
this exception the specification was revised to include Viton as an acceptable
material even though the radiation environment remained at 3.0 E6 rads in the
procurement specification.
The team was concerned that (1) the use of Viton may not have been appropriate
for the instrument air application reviewed and (2) Viton may have been used in
other instrument air or safety-related applications even though a high-
radiation environment may exist.
STATUS OF TINDINGS
During the reinspection, the team confirined that the integrated dose has been
calculated and that the integrated dose (0 to 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br />) is 2.067 E2 rads from
the containment atmosphere and 7.686 E2 rads from Room 69. The subject OPPD
calculation is ES-86-10. "Post-Accident TID for Room 81 Calculation #64,"
Revision 1, dated September 1986. It appears that the total integrated dose
over a 1000 hour0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> period is less than or equal to 9.753 E2 rads. This value
is significantly less that the specified value of 3.0 E6 rads used in the pro-
curement specification. Therefore, the use of Viton as a seat material appears
to be warranted in the instrument air check valve application discussed above.
The team coilfirmed that General Engineering Guide GEG-3, "Preparation of Desigq
Packages," Revision 0. December 1, 1987, provides guidelines for determining
the environmental conditions for new equipment. The guide specifies-that
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all pertinent envi'ronmental conditions for new equipment such as pressure, -
temperature, and radiation (along with the duration of exposure) should be
documented in the Environinental Conditions section. Mild-environment para-
meters should be taken from ETS-001, and harsh-environment parameters should be
taken from the EEQ Manual. Also, GEG-3 suggests that the preparer of a design
package describe how the processes, components and equipment installed as part
of the modification arc suitable for the application and compatible with exist-
ing materials and process conditions.
The use of Viton as seat material for the instrument air check valves in Roon
81 is warranted based on acditional information provided by the licensee.
Guidance has been provided that should help in preventing recurrence of
spe:ifying and using naterials that are incompatible with the environmental
conditions. Therefore this unresolved item is closed.
- RELATED GENERIC ISSUES
j Since it was demonstrated that there was no misapplication of materiel, the
- licensee has not determined where else Viton has been used in the plant and,
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in particular, in other safety-related instrumer.t air check valve applications,
Because the original preparer of the modification package used a specification
pertaining to another application and the supplier of small air check valves
typically uses Viton setts, the possibility exists that Viton has been used in
an inappropriate application. However, the team believes that this likelihood
is very small on the basis of the fol)owing: (1) the integrated radiation dose
to Viton valve seats can be further lowered because of the shielding provided
by the valve's metal body and (2) the typical application would not require the
, check v61ves to continue to functior, fo; long periods ef ter an accident and
therefore the integrated dose over the period of required operation would not
adversely affect safety function of the check valves.
On the basis of the foregoing, the team conc 1cdes no generic concert, exists.
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(Closeu) Deficiency 02.2-1, Incorrect Design Input in Calculation Associated
With MR-FC-81-21B
BACKGROUhD
During the 1985 SSOMI, the team found that the modification file contained a
calculation theet that showed that the air accumulator had sufficient volume
but that had the fellowing discrepancies:
(1) The volume of stored air used in the calculation was overestimated by 335
percent,
,
(2) The calculation assumed that the ai. pressure is 100 psig, even though the
instrument air system pressure will range between 80 and 100 psig,
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(3) The calculation did not consider system leakage or the period of time that t
the valve must rcmain shut.
(4) The calculation sheet was not signed by a checker and a B-2-2 foria was not '
attached to the calculation or included in ts' modification file.
In response to this finding, the licensee completed c second calculation.
STATUS OF FIhDING
Ouring the reinspection, the team reviewed OPPD Calculation FC2007,
"Accumulator Sizing," Revision 1, March 26, 1987. In the new calculation, a
more appropriate accumulator volume of 1320 cubic inches was used. The minimum '
, air pressure used was 80 psig, the lower limit of the instrument air system
!
pressure. With these initial conditions, the accumulator was demonstrated to
l be sized properly with a factor of safety of 1.4 on accumulator final pressure.
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In a memorandum dated March 21, 1988, GSE-FC-88-509, from D. K. Haas to M. E.
Eidem, "Documenting Minimum Instrument Air Pressure " the minimum int.trument
air pressure was determined to be 76 psig. This pressure correlates to the !
closing of valve PCV-1753 at a setpoint of 80 psig plus or minus 4 psig
to isolate plant air to maintain instrument air pressure. If this new value
I is included in the calculation, the factor of safety is changed frem 1.4 to
'- 1.34.
l The calculation sheet for the new calculation was signed by the checker, as [
were the GSE-B 2-2 form and each page of the calculation. r
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On the basis nf the foregoing, the team's-concerns were resolved and the item
is closed.
RELATED GENERIC ISSUES
1he related generic issues associated with functionelity and testing of air
accumulators are addressed in conjunction with Deficiencies 02.1-1 and 02.2-3. ;
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(Closed) Deficiency D2.2-2, Incomplete Consideration of CQE and Seismic <
Class I Requd.rements for Portions of MR-FC-81-21B
BACKGROUND
During the 1985 SSOMI, the team found that the seismic requirements were not
properly addressed in the modification package for completed modification
MR-FC-81-21B and identified the following discrepancies:
(1) A purchase order was issued to confirm that the valve and operatcr
assembly supplied was seismically cualified without invoking the require-
ments 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.
(E) The installation / test procedure did not reference Fort Calhoun criteria
for routing and support of seismic instrument tubing.
(3) An undocumented engineering judgment was made that the installed configura-
tion of the air accumulator, including base plate and Hilti bolts, was the
same or was bounded by a 1985 generi_c seismic analysis.
')
STATUS Of FINDING
The licensee confirmed that the subject procurement document was not intended
to be used as a basis for seismic design nor was it intended to invoke the
requirements of 10 CFR Part 50, Appendix B. Because of similarities between
the old and new actuators, it was assumed, on the basis of an engineering
judgment, that a new seismic analysis was not required. However, a seismic
analysis was performed by another vendor and confirmed acceptable seismic
accelerations for VM ves HCV-438 B and D (Stevenson & Associates, Inc. Report,
"Seismic Analysis of Fisher Controls. Fort Calhoun Nuclear Service Valves,
Valve Tag Nos. HCV-438 B & D." November 1985).
The licensee initiated a program to confirm the seismic qualification of air
accumulatom specified as critical quality equipment after water had gotten
into the ' m rument air header. This program confirmed that an analysis was on
file and that the as ir. stalled configuration for many plant air accumulators
was boJnded by the analysis however, the air accumulators for valves HCV-438 B
and HCV-434 0 were not included in the program. During the reinspection, the
licensee performed a field verification and confirmed that the accumulators for
these w ives were seismically mounted (GSE-FC-88 611 memorandum from M. E.
Eidem to MR FC-81-21B file, April 6, 1988).
The team examined documentation on special training sessions on routing
and supporting seismic instrument tubing, and sizing air accumulators and
providing seismic support (GSE-FC-86-1510, "Training Sessions " December 16,
1986).
RELATED GENERIC ISSUES
The licensee's reliance on undocumented engineering . judgment contributed to
the team's concern about excessive use of engineering judgment. During the
reinspection, the licensee issued a new Administrative Procedure A-14, "Use
of Engineering Judgment," Revision 0, April 1988. This policy outlined the
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requirements and documentation necessary when using engineering judgment in
place of detailed calculation or analysis in preparing modification packages,
10 CFR 50.59 safety evaluations, and engineering reports. This procedure,
when implemented, should improve. traceability of final design back to design
,
input. This procedure satisfies the team's concern.
The seismic design adequacy of other plant equipment is being addressed in '
Observation 03.1 4. ,
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ - _ - _ _ _ _ _ _
_-
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(0 pen) Deficiency D2.2-3, Incomplete Installation / Testing Procedure
Performed for MR-FC-81-21B
BACKGROUND
During the 1985 550MI, the team found that the postmodification test procedure
for modification MR-FC-81-21B did not require the use of the pressurized volume
of the accumulator to shut valves HCV-438 B and D. The installation and test
procedure called for the closing of valves HCV-438 B and D using instrument
air, then the isolation of air from the instrument air header t,y closing valves
lA-174 and IA-375. In this configuration only a static test was conducted.
l
The acceptance criterion was to ensu.e that the valves remained shut for 20
minutes; however, the team found no documented basis that 20 minutes was
sufficient time to identify the need to manually close these valves and to have
a plant operator perform the required action locally at the valves.
In response to the team's finding, a functional test was completed before
] the end of the 1987 outage.
STATUS OF FINDING
During the reinspection, the team reviewed Maintenance Order 872293 "Air Accumu-
lator Testing on HCV-4388, HCV-4380, HCV-238. HCV-239, HCV-240, and HCV-712A."
The stated purpose of this procedure was to provide instructions for the functional
testing of valves with air accumulators specified as critical quality equipment
to ensure a safe shutdown in case of loss of offsite power coincident with a
design-basis accident, or to mitigate the consequences of such an accident.
Tests of valves HCV-438B and HCV-438D were performed on May 14, 1987.
The functional test did not duplicate the accident conditinn, and the testing
- instrunentation was not arranged to permit confirmation of the design. ,
, Specifically, the test procedure called for the installation of a pressure gage
- between the valve actuator and its air set (i.e., pressure regulator). There-
fore, throughout the test the pressure gage only measured the air set pressure
of approximately 31 to 33 psig. In addition, the initial instrument air header
presure was not recorded, even though it could vary from 100 to 76 psig.
,
Likewise, the pressure in the accumulator after the valve was closed was not
'
recorded, nor was the pressure remaining in the accumulator after the valve was
held shut for 30 minutes. Therefore, system leakage could not be evaluated.
3 Because unknown system leakage could have been excessive, and minimum instru-
~
ment air header pressure would not have existed to hold the valve shut (i.e.,
30 psig) under worst-case accidant conditions.
A memorandum from the licensee documentad that the actions of an operator were
timed as he simulated manual isolation of valves HCV-438B and D (FC-1669-86,
'
memorandum from W. G. Gates to J. K. Gasper, "Integrated Regulatory Requirement
Log !!em No. 860192," November 13, .986). The licensee found that the operation
was completed within 6 minutes and thus satisfied the design requirement to
l allow for 20 minutes of holding air.
!
This deficiency remains open pending completion of the following actions:
- (1) Functional testing of valves HCV-438 B and D when they are shut ard as they
remain shut for 20 minutes under worst-casa accident conditions.
A-19
. --. .- _ . . - - - - - . - .
. _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _
.
(2) Completion of the program initiated to provide a comprehensive evalua-
tion of systems that depend on air accumulators for proper functioning
during an accident. Completion includes development of criteria for
functional testing and surveillance testing.
RELATEL GENERIC *SSUES
A related ceneric issue is the adequacy of air accumull'ars on all safety-
relatedvalves. In response to this concern, the licensee initiated a
program to
(1) determine the operating criteria for the valves during each applicable
postulated accident (i.e., operating pressure and tempe*ature, length of
time af ter an initiating event when valve operation will commence, and
length of time the valve operator must function)
(k) cetermine the criteria for functional testing of each valve operator '
identified
(3) develop appropriate periodic testing to ensure that the systems
continue to function as required
OPpD Memorandum TS-TC-88-120 from R. C. Kellogg to J. J. Fisicaro, dateo
February 16, 1988 states that Operations Support Analysis Report (0SAR) 87-10
was prepared to document the findings related to the air accumulators. A total
of 84 valves were identified as being equipped with air accumulators and 38
were safety-related. It is expected that the OSAR 87-10 report will b2 i
completed and reviewed by June 1, 1988, and that $11 safety-related
accumulators will be tested before or during the 1988 outage.
Another related issue is the procedural control to prevent the use of
'
incomplete installation and testing procedures. Standing Order G-21 was
revised (Revision 27, April 10, 1987) to (1) reference Standing Order G-19 for
testing requirements; (2) define the requirement for stating the postmodifica-
tion system / component perfomance requirerents for preparing, perfonning, and
evaluating the test; (3) specify the person responsible for evaluating modifi-
cation-related test results; and (4) ensure that systems are tested and test
results are approved before the systems are returned to service after they have
been modified.
The operations engineer and the onsite review committee are respor.sible for
ensuring the postmodification system / component perfennance requirements are
adequately stated, sufficient steps are included for perforning the test,
responsibility for evaluating the test results is clearly defined, and adequate
assurance exists that tet, ting resultr, are approved before a system is returned
to service following a modification.
,
i GSE Engineering Manual, Section GEG-3, Paragraph 5.9.2, stated that "test
l procedures shall be included to test the system to assure its function in
'
actual operation." In addition, the verification process specifically
addressed functional testing of modifications. B-11 Checklist F, Section 5.P.
"Installation / Test Procedures," Item 21 asked, "Does testing demonstrate, as
cicse to practical, the no-mal, abnormal and emergency function can be
l
l A-20
l
_ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ ___ _
- .
accomplished?" These requirements should ensure that all design changes are
adequately functionally tested after installation.
On the basis of the revision to Standing Order G-21, the team finds that proce-
dural requirements appear to exist to prevent the use of incomplete installa- '
tion and testing procedures. In addition, the licensee has issued for trial
use General Engineering Guideline GEG-28, "Preparation of Installation and Test
'
Procedure," Revision 0, February 1988.
In spite of procedural requirements and guidelines, the recurrence of weak-
nesses in postmodification testing indicates a need for continued licensee
attention.
On the basis of the foregoing, the following additional items are required to
<
close cut the related generic issues:
(1) training on the development of functional testing and test acceptance
criteria
(2) reassessment of completed functional testing of air accumulators performed ;
as part of OSAp. 87-10 '
'
(3) completion of all functional testing of safety-related air accumulators by
- the end of the next refueling outage.
'
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- *
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(Closed) Observation 02.2-4, Incomplete Modifi;ation File for a Completed
Hodification
'
BACKGROUhD
During the 1985 S$0MI, for completed modification MR FC-81-21B, the team ;
identified information that was missing from the modification file, including
(1) records of third-party review or check of a calculation and (2) a specifi-
cation for the procurement of safety-related check valves or of a third-party
verification.
STATUS OF FINDING
In response to this finding, the licensee strengthened procedures and developed !
'
guideline documents for the preparation of modification packages. During the
reinspection, the team evaluated the following documents:
'
(1) General Engineering Guide GEG-3, "Preparation of Design Packages,"
l Revision 0, December 1,1987, should prevent recurrence of an incomplete
modification file for a completed modification because it prescribes a
modification design process. GEG-3 gives a standard approach for the
- preparation of a design package. Detailed engineering guidelines and
associated checklists are provided to assist the design engineer in
preparing the design package and documenting various design features,
assumptions, and design inputs.
Section 5.5 discusses design analysis required for each phase of a design
package. For example, in a final design issue, this section will present
j an overview of all the analyses that would be dono to support the design
modification.
! Section 5.5.2 discusses the procurement specifications. In a final design
issue, all procurement specifications for engineered equipment will be
listed.
(2) Design Procedure 8-11. "Independent Design Yerification," Revised
i February 1988 l
j (a) Paragraph 4.2 requires that all modifications have a multidiscipline [
j review, !
! (b) Paragraph 1.0 states that independent multidisciplinary design
verifications will be done for design changes including those
! pertaining to critical quality equipment (CQE), limited CQE, non-CQE,
and fire protection system and components.
(3) Administrative Procedure A.11. "Calculation Numbering and Revision !
Control," Revised March 1987, which was revised to require the indexing of ,
l. all calculations, with a separate log book for each station (fossil and t
nuclear). In addition, the main frame computer can be searched using the !
GSE-1 search program to identify the file location of calculations. l
Current and past revisions of calculations are available on microfilm in '
Document Control. The team examined the log book for Fort Calhoun Station
and witnessed a demonstration ef the GSE-1 search prcgram for calculations.
1
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A-22
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._ n. -~-.
-v- - .,n---------~----_,s.- .- - _ - - - - - - - . - - , - . - - - - - , ,
--
. , . - -
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _
i
On the basis of the foregoing, this item is closed.
RELATED GENERIC !SSUES
The prograntnatic aspects of this issue are discussed above and adequately
address the generic issues associated with the content of modification
packages.
.
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A-23
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i
(Closed) Deficiency D2.2-5, Incorrect Information on Instrument Air Diagnm
BACKGROUND
'
Instrument air header isolation valves, IA-175 and IA-176, were used during
the installation and testing of mod'fication MR-FC-81-21B. However, these
valves did not appear on OPPC Drawint, 11405-M-264, the piping and instrumenta-
tion diagram (P&lD) for the instrument air system. During a field inspection
the team confirmed that the valves were installed in the plant.
STATUS OF FIhDING
.During the reinspection, the licensee revised P&lD 11405-M-264, Sheet 3
"Instrument Air Diagram Riser Details PalD." Revision 12. dated April 11, 1988,
to include missing instrument air isolation valves IA-174 and IA-175. '
RELATED GENERIC ISSUES
Because the licensee had taken no action regarding on this deficiency before
the team's reinsper. tion and because of the licensee's previous position that
these types of valves that were not specified as critical quality equipment
(CQE) were shown at the discretion of the design engineer and his/her super-
visor, the team was concerned that other non-CQE valves in the other systems
important to safety were not shown on the systerr P&!D. In an OPPD memorandum
- (GSE-FC-88-627 from M. E. Eidem to S. K. Gambhir, W. G. Gates, anc J. J.
Fisicaro, dated April 12,1988), the licensee indicated their int:nt to revise
the P& ids or other related drawings to si.a manual and check vahes whose
operation or misoperation could affect the function of comporents served by the
instrument air system,
i As a general practice, P&lDs for CQE and non-CQE systems should accurately
reflect the as-installed condition. Although the licensee's response does not
!
appear to address systems other than the instrument air system, the team found
no instances where valves were omitted from other P& ids. Therefore, on the
basis of the licensee's stated intention for the instrument air system, this
! generic issue is closed.
1
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A-24
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_ _ _ . . . _ _ _ _ . _ _ . _ . _ . . ___._
_._-___ _._._ _____._ ____._ __._.____ _ __ _ ._ _
_ -
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
. l
(Closed) Deficiency D2.2-6, 10 CFR 50.59 Safety Evaluation Based Upon an
Incorrect Assumption and Analysis Methodology
BACKGROUND
Modification W.-FC-81-21B caused the post-loss-of-coolant-accident (LOCA) heat
load to increase by 3.15 million Stu/ hour. During the 1985 SSOMI, the team
found that a safety evaluation included in the final design description was
weak for the following reasons:
(1) The safety analysis performed did not refer to original design calcula-
tions. The lack of original design analyses or their availability did not
result in the performance of new calculations.
(2) The qualitative argument used did not reflect a correct understanding of
the heat transfer phenomenon between heat removal systems.
(3) The safety evaluation contained an unsubstantiated and inappropriate
assumption concerning operator action to secure heat loads under certain
accident conditions.
(4) 1he basis of Techni.al Specification 2.4 contained incorrect infomation
concerning the he* removal capacity of the component cooling water (CCW)
heat exchangers.
STATUS OF FINDING
.
During the reinspection, the team closed this item on the following basis:
(1) The margin of safety in the Technical Specifications was substantially not
affected because of the high margin between the new post-LOCA heat load
(maximum design loads on the CCW system following a LOCA plus a new heat
load resulting from allowing the reactor coolant pumps to operate) ano the
heat removal capecity of the CCK' system.
(2) Controlled Copy No. 58 of the Technical Specifications was reviewed, and
the team confirmed that Technical Specificatica Section 2.4, "Containment
'
Cooling," has been revised to read, "Three component cooling heat
exchangers have sufficient capacity to remove 402 million Btu /hr following
a luss-of-coolant accident."
RELATED GENEPIC ISSUES
One of the major concerns related to this deficiency was the apparent lack of
original design analyses or their availability. The lack of a design analysis
did not result in new calculations. Thus, a design engineer had to rely on
qualitative arguments instead of quantitative bases. The team reviewed the
following guidance and procedures provided to the design engineer to assess if
this weakness had been corrected:
(1) GEG-3, "Preparation of Design Packages * Revision O. December 1, 1987
requires that each functional requirement be referenced to a design-basis
I
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cocument, ktere the original design basis is missing, applicable analysis
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A-25
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'. .
4
should be documented to support the design value used for the modification
or other justification should be provided.
(2) Standing Order G-21, "Station Modification Control," Revision 29 November
19, 1987, and GSE Design Procedure B-2, "Production of Design Description
and Evaluation Nuclear Modifications," Revised December 1987, identified the
following as reference documents for detemining the design basis of an
existing system, component, or structure for the purpose of modifica-
tions and safety evaluations:
- Updated Safety Analysis Report in conjunction with any pending
change
- Technical Specifications including Basis section
- safety evaluation reports for Technical Specifications amendments
' design drawings
Additionally, if the calculations or actual design requirements were not
available, the design changes will be based on the assumption that no
,
design margins exist unless otherwise justified on the basis of conserva-
- tive assumptions and/or field verification. Alternatively, the calcula-
tions will be redone to establish design margins, and design bases will be
recreated on an as-needed basis.
3. OPPD Nuclear Production Division Policy / Procedure E-1, "10 CFR 50.59
Safety Evaluations " guides the preparer in identihing and reviewing
plant-specific design, operating, and technical documents that describe
the affected structures and system components and their res'ective safety
function (s). These documents include, but are not limited to, design
basis documentation and calculations, related design changes, related
licensee event reports, previous safety analysis, operating instructions /
procedures, surveillance tests, and system descripticns.
(4) General Engin'dering Guide GEG-27, "10 CFR 50.59 Safety Evaluation,"
Revision 0, February 1988, states that design-basis information, analyses,
'
i
and supporting system interaction evaluations necessary to perfom a
safety evaluation were developed as part of other sections of the design
- package. The purpose of the safety evaluation section is to abstract
the salient conclusion and supporting information developed in those
sections to develop a logical presentation of the potential safety issues
involved with the modification.
(5) Administrative Procedure A-11. "Calculation Numbering and Revision
Control," Revised March 1987, was revised to require the indexing of all
- calculations with a separate log book for each station (fossil and
- nuclear). In addition, the main frame computer can be searched using the
- GSE-1 search nrogram to identify the file location of calculations.
! Current and past revisions of calculations are available on microfilm in
! Document Control. The team examined the log book for Fort Calhoun Station
and witnessed a demonstration of the GSE-1 search program for calculations.
A-26
i
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On the basis of the foregoing, the team found that the licensee had stressed
the need to refer to controlled sources when performing a safety evaluation and
had provided a means of recovering old enslyses. Although the USAR typically
is not considered a design document but is instead a compilation of design
and accident information, the interim use of this document until design-basis
documents are available should be with the awareness that some of the
information may no longer be valid.
In general, programmatic activities since the SS0M) should result in improved
safety evaluations. The specific concerns pertaining to this deficiency have
been resolved. Thus, the related generic issue is considered closed.
4
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- ._. . _ _ . - . _ , , _ _ - . . _ _ , . _ _ . - - - . . . , _ _ . _ _ - _ - _ - _ _ _ . _ _ _ , _ _ _ _ _ _ , _ . _ _ . _ _ - . . _ _ _ _ . , , _ -
____ _ ____ _ _- _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
'. *
.
(Closed) Deficiency 03.1-1, Balance-of-Plant Design Specifications
BACKGROUND
Deficiency D3.1-1 documented the licensee's use of the design specifications
contained in OPPD Contract No. 763 to define the design basis for
balance-of-plant piping and equipment. However, the licensee did not control
Contract No. 763 in accordance with the requirements of its quality assurance '
manual.
STATUS OF FIND!teG
To address Deficiency D3.1-1, the licensee withdrew Contract No. 763 from use
as a design document. It is documenting the regenerated system functional
basis for each safety-related piping system at Fort Calhoun in a system design-
basis document (SDBD). Each SDBD references a series of plant-level design-
basis documents (PLDBDs) such as PLDBD-CS-51, "Seismic Criteria," and
PLDBD ME-10. "Pipe Stress and Supports," which specify the governing design
criteria for the system piping, equipment, and supports.
On February 18, 1988, the licensee provided the NRC with a list of the SDEDs
and PLDBDs under preparation. At that time, the licensee indicated that the
PLDBDs would be completed by January 1989, and the SDBDs would be completed by
January 1990. This item is therefore clnsed.
, RELATED GENERIC ISSUES
The licensee's inconsistent use of design input data to implement modifications
j to existing piping, equipment and supports is a generic issue related to
Deficiency 93.1-1. However, the team believes that the use of PLDBDs such as
'
PLDBD-CS-51, "Seismic Criteria," and PLDBD-ME-10. "Pipe Stress ard Supports "
should enable the licensee's design staff to specify design input data for use
in future modifications in a consistent manner.
%
! 0
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A-28
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!-___-. .- .__. ,_,-- . ._ ._ - - -_ _ _ - _ _ _ __ .. - . _ - - _ - -
_ _ ________
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(Closed) Deficiency D3.1-2 Design Temperatures for Safety-Related Piping
BACKGROUND
In response to IE Bulletin 79-14, the licensee engaged Gilbert /Comonwealth
,
(G/C) to reanalyze a number of safety-related piping systems at Fort Calhoun
Station. Deficiency 03.1-2 documented the licensee's transmittal of operating
and accident temperature data to G/C in 1980 which the licensee did not control
in accordance with its quality assurance manual.
STATUS OF FINDING
To address Deficiency D3.1-2, the licensee engaged Applied Power Associates
! (APA) to document the operating and accident temperatures for the safety-
l related piping systems at Fort Calhoun Station. APA's repo/t is entitled
I "Documentation of the Source of Operating and Accident Pipe Temperatures / Fort
Calhoun Station," APA Report No. AOR 103.83.07, dated October 1987. The
licensee has documented APA's report as Operations Support Analysis Report
(0SAR)86-11. Deficiency D3.1-2 is therefore closed.
The reinspection team notes that OPPD Memorandum GSE-FC-87-1838, dated
November 10, 1987, summarized the licensee's comparison of the operaling
temperature differentials documented in OSAR 86-11 with the temperature
differentials that it transmitted to G/C in 1980. The memorandum documents
temperature differentials that were larger than the temperature differentials
that the licesee originally provided to G/C for parts or all of the following
piping systems:
(2) safety injection system
(3 containment spray system
(4 component cooling system
(5 chemical and volume control system
(6 raw water system
The licensee identified significant increases in the temperature di Herentials
for portions of the chemical and volume control system, but noted that no
analysis documentation existed for this small-bore piping.
The team noted that NRC Region IV documented the licensee's lack of thermal
Analysis documents for safety-related small-bore piping in NRC Inspection
Report 50-285/85-03, dated May 23, 1985. The team asked the licensee to
provide the NRC with the details of its proposed corrective action program to
address the open issues pertaining to esfety-related small-bore piping at FSrt
Calhoun Station (see Unresolved Item U3.1-3).
The licensee also identified significant increases in the temperature differen-
tials for five piping subsystems in the raw water system. Stone & Webster
(S&W) was reviewing the load capacities of some of the pipe supports in the
affected piping subsystems. OPPD Memorandum GSE-FC-88-461, dated March 11,
1988, indicated that S&'.!'s review was scheduled for ccmpletion by April 15,
1988. On the basis of the results of S&W's review, the licensee may reanalyze
the affected raw water piping subsystems.
A-29
_ _-_____-__- -___ _
_ _ _ _ . - . ~ _ . - . _ . . . - . . . . - . _ _ _ - _ - _ - . . _ . - . . ._
o 4
,
1
RELATED GENERIC ISSUES
The licensee's documentation and transmittal of design information are generic
issues related to Deficiency D3.1-?. OPPD Memorandum TS-FC.86t182, dated
.
P, arch 10, 1986, indicates that OSARs documented and developed in accordance '
with Technical Services Procedures N-TSAP-5 and -6 were controlled documents ,
that meet the requirements of the licensee's quality assurance manual. The
! team, therefore, believes that design infonnation which the licensee docurents
and transmits in accordance with these technical services procedures will meet
<
the requirements of the licensee's quality assurance manual.
l
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A-30
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ __ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - -
.. .
(0 pen) Unresolved Item U3.1-3, Small Bore Pipe Support Spacing
BACKGROUND
Unresolved Item U3.1-3 documented the licensee's use of a nomograph to
field-route small-bore piping that did not implement the 12-Hz minimum horizon-
tal frequency criterion stipulated in USAR Appendix F. Section F.2.2.2, for
piping f.onnected to the containment.
,
STATUS OF FINDING
The licensee's program to address the lack of design-basis documents for
safety-related small-bore piping was sumarized in OPPD Memorandum GSE-FC-88-500,
dated March 21, 1988. The licensee intended to address Unresolved Item U3.1-3
as part of the progrsm that was sumarized in the memorandum. The purposes of
the program were to
(1) develop in-house software for piping analysis, based on design specifica.
tions prepared by OPPD
(2) prepare and licenu a set of seismic criteria as an alternative to the
current USAR seismic criteria
(3) formulate program to address issues pertaining to safety-related
small-bore piping using items (1) and (2).
With respect to item (1), the licensee purchased a version of the SUPERPIPE ,
computer program and associated software from Impell Corporation. Impell had
already benchmarked the in-house version of SUPERPIPE, and the program and
associated software would soon be accessible on the licensee's computer facili-
ties. The sof tware requirements which OPPD specified for the computer programs
were documented in an OPPD report entitled "System Requirements Specification
and Software Design Description for SUPERPIPE and Supporting Programs, Version
22C," Revision R01.0, dated November 3, 1987.
,
With respect to item (2), Impell generated a set of alternate seismic criteria
for the licensee that were documented in two Impell reports entitled "Alternate
Seismic Criteria & Methodologies for Fort Calhoun Station," Volume 1. "Criteria
& Methodologies," Report No. 01-1390-1650 Revision 0, dated January 1988; and
"AlternMe Seismic Criteria & Methodologies for Fort Calhoun Station," Volume
2. "Jus;ification of Criteria & Methodologies " Report No. 01-1390-1650,
Revision 0, dated January 1988. ;
OPPD Memorandum GSE-FC-88-385, dated March 1,1988, indicated that the generat-
ing station engineering staff had issued the referenced Impell reports to the
licensing staff for eventual submittal to the NRC.
,
With respect to item (3), OPPD Memorandum GSE-FC-88-506 noted that the licensee's
formulation of a program to address the issues related to small-bore safety- '
related piping was contingent on NRC's acceptance of the alternate seismic
criteria that Impell prepared for the licensee. The licensee planned to
implement the program sumarized in the memorandum in the sumer of 1989 and to j
t.omplete work by June 1, 1991.
A-31
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_ _ _ , , . s-_ . ,_ . __ . _ . . . _ . . _ _ _ _ _ _ . . _ _ , _ _ _ _ _ . _._ _ _ _ _ _ ,
..
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.
The reinspection team noted that the licensee's latest response to the SSOMI,
dated April 10, 1987, did not sumarize its corrective action program for ,
safety-related small-bore piping.
l
The team also noted that the alternate seismic criteria that Impell prepared
for the licensee specified a number of design criteria that represent relax 6-
tions of current USAR commitments. For example, Impell proposed that piping
vibratory modes be combined by random vibration principles, rather than by the
square root of the sum of the squares (SRSS), and that ASME Code Case N-411-1
.
'
damping values be used to analyze safety-related piping systems instead of the
O.5-percent domping currently specified for the seismic analysis of safety-
related piping systems in USAR Appendix F Table F-2.
! Unresolved Item U3.1-3, therefore, remains open pending the licensee's
4
submittal to the NRC of the proposed methodology, scope, and time frame of its '
corrective action program to address the issues related to safety-related
small-bore piping. The team noted that several of the alternate seismic
criteria in the referenced Impell reports represent relaxations of current USAR
criteria.
!
RELATED GENERIC ISSUES 1
i
The licensee's inconsistent implementation of the seismic criteria contained in
USAR Appendix F is the generic issue that is related to Deficiency 03.1-2. The :
- team recomends that the licensee review USAR Appendix F and replace seismic
! criteria that is no longer implemented with comparable criteria that it
considers compatible with carrent design practice.
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_ _ _ _ _ _ _ _ _ _ - _ _ _ _
- _ _ _ -
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(Closed) Observation 03.1-4, Seismic Qualification of Valves Installed
in Class ! Piping Systems
BACKGROUND
Observation 03.1-4 noted a lack of documentation supporting the seismic
qualification of valves and valve operators installed in safety-related piping
systems at Fort Calhoun Station. However, the licensee is a member of a utility
group that is addressing the lack of seismic qualificat!cn documents for some
components in older plants in response to NRC Unresolved Safety Issue (USI)
A-46, "Seismic Oualification of Equipment in Operating Plants."
STATUS OF FINDING
The licens e was resolving the valve seismic qualification issue and related
issues addressed in US! A-46 through the Seismic Qualificaticn Utility Group
(SQUG). The NRC was currently reviewing a draft SQUG report entitled "Generic
Seismic Qualification Procedure for Nuclear Plant Equipment," dated March 27,
1987. A memorandum to SQUG members frou R. E. Schaffstall of KMC, Inc. (SQUG
Status Review Mweting, Implementation Schedules for USl A46, dated March 8,
1988) indicated that SQUG would issue Revision 0 of its report to member
utilities for review by April 30, 1988, and to the NRC by May 31, 1988. The
memorandum indicated that the NRC would issue a generic safety evaluation
report (SER) by June 30, 1988. The NRC was expected to request plant-specific
schedules from SQUG members within 60 days of the issuance of the SER.
The reinspection team participated in a telephone conference on April 6, 1988,
with NRC Mechanical Engineering Branch (MEB) and Region IV staff to discuss the
seismic qualification requirements for the safety-related manually operated
valves installed at Fort Calhoun Station. During the conference, the MEB staff
informed the team that MEB's program tc resolve the seismic qualification
issues addressed in USI A-46 assumed that installed safety-related eouipment was
in conformance with current plant USAR licensing criteria. As i,'idicated in
Observation 03.1-4, the licensee could not confirm that safety-related valves
and valve operators installed at Fort Calhoun Station were qualified in accord-
ance with the seismic criteria specified in USAR Appendix F for u fety-related
equipment. In a letter from NRC (J. Calvo) to OPPD (K. J. Morri;) dated
July 28, 1988, the NRC staff accepted, subject to certain conditions OPPD's
proposal to delay the resolution of seismic qualification of equipment until
the resolution of USI A-46.
The team is, therefore, closing Observation 03.1-4 for the purposes of the SSOMI,
since the licensee's program to qualify the safety-related valves and valve
operators installed at Fort Calhoun Station is to be resolved with the NRC's
Office of Nuclear Reactor Regulation.
RELATED GENERIC ISSUES
The lack of seismic qualification of equipment installed in older plants to
current licensing criteria is the generic issue related to Observation 03.1-4
In response to Unresolved Safety Issue A 46, the licensee is attempting to
verify the seismic adequacy of other active mechanical and electrical equipment
in adoition to valves and valve operators.
A-33
. - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
.
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(Closed) Unresolved Item U3.2-1, MR-FC-84-61 Design Input Source and Use
BACKGR0thD
Unresolved item U3.2-1 documented the licensee's inadequate references to the
design basis and the undocumented use of engineering judgment in Modification
Request MR-FC-84-61.
STATUS OF FINDING
OPPD Memorandum GSE-FC-86-770 ("Documentation of Engineering Judgment: Union
, Mass on Safety Injection Tank," dated August 6, 1986) indicates that the
installation of the 1-inch, 2.5-pound union above the safety injection (SI)
tank would induce negligible additional forces, moments and stresses on the $1 :
tank nozzle. Unresolved Item U3.2-1 is therefore closed.
RELATED GEhERIC ISSUES
.
The adeouate preparation of modification request packages and the adequate
j cocumentation of engineering judgment are generic issues related to Unresolved
- Item U3.2-1.
.
The reinspection team reviewed General Engineering Guide GEG-3, "Preparation of
Design Packages," Revision 0. December 1, 1937, which Sargent & Lundy prepared
,
for the licensee, and believes that the procedure provides adequate guidance
to licensee personnel involved in the preparation, documentation, and review of
design packages for Fort Calhoun Staticn. ,
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The licensee's interin instructions to its design staff to document the use of
engineering judgment were detailed in OPPD Memorandum LIC-00-060, dated March 7,
1986. A procedure entitled "Use of Engineering Judgment," Revision D was
issued in April 1968.
The team believes that the licensee's proposed procedure, when implemented in
I conjunction with the plant-level design-basis documents that the licensee is
preparing, should substantially reduce the incidence of undocumented engineer-
ing judgment in design modification packages. ,
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___ _____ _ _ _ _ _ _ _ _ _ _
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(Closed) Deficiency D3.2-2, MR-FC-83-158 Installation Procedure
BACKGROUND
Deficiency 03.2-2 documented the licensee's failure to specify adequate seismic
support spacing criteria for instrument tubing in Modif t:ation Request
- MR-FC-83-158.
STATUS OF FINDING
The licensee's initial response to the SSOMI documented in Attachment A to OPPD
Letter LIC-86-106, dated April 15, 1986, qualified the subject tubing by
analysis. During the reinspection, the team confinned that the licensee had
revised Modification Request MR-FC-83-158 to incorporate an OPPD calculation
entitled "Tubing Support Distance Cales/ Air Accumulators for YCV-1045 A & B."
Revision 0, dated December 18, 1985. Deficiency D3.2-2 is therefore closed.
RELATED GENERIC ISSUES
.
- $
4
The adequate preparation of modification request packages is a generic issue
- related to Deficiency D3.2-2.
l
The reinspection team reviewed General Engineering Guide GEG-3, "Preparation of '
Design Packages," Revision 0. December 1, 1987, which Sargent & Lundy
prepared for the licensee, and believes that the procedure provides cdequate
guidance to licensee personnel involved in the preparation, documentation, and
review of design packages for Fort Calhoun Station.
The licensee also revised Part !! of Fort Calhoun Station Unit No. 1 Standing
Order G-30. "Field Changes to Modification Construction Drawings," on April 4,
1987. The team believes that implementation of the revised procedure, which
governs field changes to modification design documents such as drawings an(
'
work instructions, should ensure the documented qualification of field-routed
instrument tubing in plant modifications.
.
.
A-35
.
_ _ _ _ _
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(Closed) Deficiency 03.2-3, MR-FC-84-162 Calculation
BACKGROUhD
Deficiency D3.2-3 documented the licensee's inadequate qualification of two
redesigned containment ventilation duct supports in Modification Request
MR-FC-84-162. The team also concluded that painting the redesigned support
steel structures, as specified by the licensee, instead of galvanizing them
was not in accordence with the licensee's requirements.
The team closed part of Deficiency D3.2-3 on the basis of the additional
calculations that the licensee prepared during the S$0K1 and the licensee's
initial response to the 550MI documented in Attcchment A to OPPD Letter
LIC-86 106, dated April 15, 1986. In its response, the licensee noted that tr.c
calculations that it prepared after the 550Mi audit confirmed the adequacy
o'l the redesigned duct supports. However, the team kept Deficiency D3.2-3 op'?n
pending the licensee's preparation of a centrolled document that specified shop
and field surface preparation of seismic Ctegory I materials.
STATUF OF FINDlhG
To address the reraining open item in Deficiency 03.2-3, the licensee engaged
Sargent & Lundy to prepare a cocument the.t specified shop and field surface-
ps aparation procedures for sessmic Category I materials in the containment,
Tha licensee documented Sargent & Lendy's report as CPPD GSE Engineering
Standard CTS-3, "Selecting, Specifying, Applying ar.d Inspecting Paint ar.d
Coatings," which the licei,see approved for use or. August 27, 1987. Deficiency
D3.2-3 is therefore closed.
RELATED GENERIC ISSUES
The adequate preparction nf modificat1on request packages is a generic issue
related to Deficiency 03.2-3.
The reinspection team reviewed General Engineering Guide CEG-3, "Preparation of
Design Packages," Eevision 0. December 1, 1987, which Sargent & Lundy prepared
for the licensee, and believes that the procedure provides adequate guidance to
- licensee personnel involved in the preparatier,, documentation, and review of
design packages for Fort Calhoun Station.
,
A-36
____ ______________ _.
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _
- *
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(Closed) Deficiency 03.2-4, Junction Box Supports
BACKGROUND
Deficiency D3.2-4 documented Unistrut supports for the junction box to valve
YCV-1045B that did not meet the seismic provisions of USAR Appendix F'.
STATUS OF FINDING
The reinspection team closed Deficiency 03.2-4 on the basis of the licensee's
initial response to the SSOMI documented in Attachment A to OPPD Letter
LIC-86-106, dated April 15, 1986. In its response, the licensee noted that it
.
had installed a qualified seismic support for the junction box during the 1985
refuelina outage. The team reviewed Modificatjan Request MR-FC-85-201 to con-
~
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firm the licensee's design, qualification, and installation of the new supports
for the junction box and adjacent conduit. Deficiency D3.2-4 is therefore
closed.
RELATED GENrRIC ISSUES
s .
'
The possibility that other installed junction boxes and conduit may lack
supports that meet the seismic provisions of USAR Appendix F is the generic
issue related to Deficiency D3.2-4.
J
OPPD Memorandum GSE-FC 88-6?2, dated April 7, 1988, indicated that the licensee
planned to address this g'eneric concern under the scope of Unresolved Safety
Issue A-46. Therefore, ihis item becomes an open licensing issue to be
resolved between the licensee and NRC's Of fice of Nuclear Reactor Regulation.
,
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(Closed) Observation 03.2-5, Containment Pressure Switch Seismic Qualification
BACKGROUND
!
Observatiun 03.2-5 noted that the replacement pressure switches in Modification
Request MR-FC-83-83 warranted a more thorough analysis to provide additional
assurance of the equipment's seismic qualification.
STATUS OF FINDING
1r. response to Observation 03.2-5, the licensee revised modification Request
MR-FC-83-83 to incorporate an equipment qualification review checklist for the
replacement pressure switches and a calculation entitled "Junction Box Hounting
Support Adequacy /MR-FC-83-83," Revision 0, dated December 13, 1985. Observa-
tion 03.2-5 is therefore closed.
RELATED GEhERIC ISSUES
The possibility that other installed replacement pressure switches may lack
complete seismic documentation is the generic issue related to Observation
03.2-5.
OPPD Memorandum GSE-FC-88-622, dated April 7, 1988, indicated that the licensee
planned to address this generic concern under the scope of Unresolved Safety
Issue A-46. Therefore, this item becomes an open licensing issue to be
resolved between the licensee and NRC's Office of Nuclear Reactor Regulation.
.
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(Closed) Deficiency D3.2-6, Steam Generator Nozzle Dams
BACKGROUND
<
Deficiency D3.2-6 documented the licensee's procurement of seismic Category I ;
critical quality equipment (CQE) removable steam generator nozzle dams '
without requiring the vendor to qualify the nozzle dams to the seismic provi-
sions of USAR Appendix F.
i
The team closed part of Deficiency 03.2-6 on tne basis of the licensee's
initial response to the SSOMI documented in Attachment A to OPPD Letter :
LIC-86-106, dated April 15, 1986. In its responte, the licensee noted that the
nozzle dam vendor had prepared a seismic analysis that qualified the steam
generator nozzle dams. However, the team kept Deficiency 03.2-6 open pending
the licensee's preparation of a procedure that would establish guidelines for !
the licensee's procurement of CQE materials and services.
STATUS OF FlHDIhG
'
To address the remaining open item in Deficiency 03.2-6, the licensee issued -
Technical Services Procedure N-TSAP-14. "Determination and Procurement of CQE
- and Limited CQE Items and Services, December 1986. Deficiency 03.2-6 is
therefore closed.
RELATED GENERIC ISSUES
>
'
The possibility that the licensee procured other CQE equipment without
,
specifying the required seismic provisions of USAR Appendix F is the generic
j issue related to Deficiency D3.2-6.
To addrrss this generic concern, the licensee reviewed a majority of the
<
requisitions issued since 1982 to confirm that the procurement specifications
for hardware, services, and sof tware were properly prepared, or that procure-
- ment specifications that were inadequately prepared did not result in the
procurement and use of inappropriate or deficient hardware, services, or
software. *
-
J
The licensee identified three purchase orders that required quality assurance l
certification. OPr0 memorandum TS-FC-87-170B, dated July 21, 1987, summarized
the review that the licensee and Combustion Engineering conducted to confirm
,
that these purchase orders met the licensee's CQE requirements.
!
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- . .
(0 pen) Deficiency D3.2-7, YCV-1045B Valve Restraint
BACKGROUND
Deficiency D3.2-7 documented the following deficiencies in the calculation of
record for the auxiliary feedwater piping subsystem in the vicinity of control
valve YCV-1045B:
(1) The valve operator was restrained by a rod attached to a stairpost.
The licensee had not implemented a comitment to the NRC to replace the
strut.
(2) The licensee's as-built drawing did not show either the valve operator or
the existing rod restraint.
(3) The vendor drawing for the valve operator could not be obtained to verify
the valve and operator weights and operator offset dimension.
(4) The valve operator restraint was not modeled in the stress analysis.
(5) There were no calculations that combined deadweight, thermal, and seismic
pipe stresses in the vicinity of the valve.
(6) There were no calculations that cembined deadweight, thermal, and seismic
loads for the supports adjacent tr. the valve. The supports appeared to be
overloaded.
(7) The computer runs were not referenced and, therefore, were not adequately
controlled.
With respect to item (1), the licensee reanalyzed the piping subsystem with the
valve operator restraint removed, and concluded that the operator restraint was
not required on the basis of the pipe stress levels and support loads in the
vicinity of the valve (study run CYG1 was executed on December 14, 1985, to
address the NRC's,SSOMI concerns; computer run CFSR was executed on April 9,
1987, to be the computer run of record). The licensee removed the valve
operator strut on January 19, 1987, via Modification Request MR-FC-86-89.
The team noted, however, that the licensee has a licensing comitment to restrain
control valve operators that induce seismic bending stresses greater than 1500
psi in the supporting pipe. As noted in USAR Appendix F. Section F2.2.2:
'
Special seismic restraints (ei':her rigid or snubbers) were provided
cn control valve mechanisms to prevent overstress when the control
mechanism forms a mass center outside the pipe center line and gene-
rates over 1500 psi bending stress on the piping system due to earth-
quake G loading.
The control valve operator strut was originally installed to satisfy the USAR
Appendix F criterion, since the architect-engineer (A/E) computed a seismic
bending stress of 10.686 psi in the supporting pipe resulting from the e cen-
tric mass of the valve operator. The valve operator weighed 197 pounds and had
an offset dimen; ion of 30.5 inches with respect to the centerline of the
supporting 2-inch Schedule 80 pipe.
A-40
,
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Computer run CFSR computes a total seismic bending stress in the supporting
pipe of 10,149 psi. However, the licensee did not implement the referenced
'
USAR Appendix F criterion since the seismic restraint was removed even though
the seismic bending stress induced in the pipe by the valve operator exceeded
1500 psi.
The licensee maintained that the piping configuration was adequately qualified :
in the vicinity of the valve, since the pipe stresses were within allowable
limits and the loads induced in the adjacent pipe supports were within design
capacity. However, the reinspection team noted that the USAR Appendix F '
.
'
criterion, which the A/E originally specified to limit the magnitude of the
seismic stresses induced in the supporting pipe, additionally limited the
magnitudes of the valve operator accelerations and stresses when the valve body '
1
was also seismically restrained. In newer plants, valve operator functionality
i is confirmed by comparing the seismic accelerations computed at the location of i
l the operator center of gravity with the maximum allowable operator accelera-
,
tions specified by analysis or test.
The team therefore noted that the licensee could not guarantee the functionality
of control valve YCV-10455 or other control valves during a seismic ev et
unless it invoked the USAR Appendix f 1500-psi criterion, or a comparRe
criterion. '
The team noted, however, that the licensee was a member utility in the Seisnic
, Qualification Review Group (SQUG), which is addressing Unresolved Safety Issue ;
(USI) A-46, and that US! A-46 addresses the generic seismic qualification of ;
, motor-operated valves in older plants,
'
l The team recommended that the licensee review USAR Appendix f and replace
,
seismic criteria that it no longer implements with comparable criteria that it
- considers corapatible with current design practice. The licensee should conduct
I this review before it prepares plant design-basis documents (PLDBDs) such as
a PLDBD-ME-10. "Pipe Stress and Supports," and PLDBD-CS-51, "Seismic Criteria,"
! which it had scheduled for completion by January 1989, and which it intended to
!
reference in the system design-basis documents that it was also preparing.
,
! The licensee also should seek NRC review of the alternate seismic criteria that
Impall preparad, amend the USAR accordingly, and incorporate these additional
, provisions in the PLDBDs (see Unresolved Item U3.1-3 for a discussion of the
! licensee's proposed alternate seismic criteria).
I
j With respect to item (2), the licensee has not yet updated OPPD Drawing D-4318,
,
sheet 1 of 3. Revision 2, nated June 26, 1986, to incorporate the computer
j
rodel node point for the valve operator. The drawing was both the as-built
- drawing and the piping stress isometric drawing for the auxiliary feedwater
,
piping suosystem. ;
)
With respect to item (3), the licensee did not have a controlled document that
J specified the weight of valve YCV-1045B and the weight and offset dimention of
the valve operator. The licensee retrieved a document prepared by the A/E
1 that tibulated valve data, but this document was not controlled. The licensee i
a indicated that uncontrolled design data retrieved from the A/E will be
addressed as part of its design-bas;s reconstitution program.
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.
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For modifications to existing designs, the licensee should verify valve data by
accessing the appropriate valve vendor drawings or by corresponding with the
valve vendor.
With respect to item (4), the licensee removed the valve operator strut so
that the computer model was consistent with the as-built configuration.
With respect to item (5), computer runs CYG1 and CFSR provided nomal and i
accident load combination pipe stresses at the valve location which were below
code allowable valves.
With respect to item (6), the licensee provided the team with the following
pipe support calculations that Gisbert/Comenwealth prepared as part of
MR-FC-81-127 and that qualified the pipe supports adjacer.t to valve YCV-1045B:
i (1 Calculation AFW-50, MSSP-50, Revisicn 0, dated October 2, 1981
(2 Calculation AFW-14, MSSP 14. Revision 0, dated September 28, 1981
(3 Calculation AFW-49, MS$p.49, Revision 0, dated October 12, 1981
,
'
The team noted that the licensee generally could not access the original
calculations that the A/E prepared to generically qualify the pipe supports on
safety-related small-bore (2-inch and less) piping. The team recomendi d that
the licensee's evaluation of the generic load capacities of safety-rela 1ed -
small-bore pipe supports be addressed as part of the licensee's program to
resolve open issues pertaining to safety-related small-bore piping. As noted
in Unresolved item U3.1-3, the team recormended that the licensee preset t this
program to the IRC. -
With respect to item (7), the licensee had not completed the design verifica-
tion of computer run CFSR executed on April 4,1987, which was the computer
run of record that supports Modification Request MR-FC-81-127.
- t
The team noted that the licensee has the following two SSOMI licensing comit-
- ments to the NRC that required that the calculations to support Modification
- Request MR-FC-81-)27, as well as other redifichtions, be completed by May 1,
1988: -
,
,
(1) licensing comitment RRD Item 87-278, which t equired action to close out
emergency modifications requiring seismic updates
t
. (2) licensing comitment RRD ltem 87-145, which required the licensee to
address and resolve the accessibility of computer program analyses. ,
4 STATUS OF FINDING
s
item (1) remains open pending until tim licensee qualifies valve YCV-1045B and
4
other control valves to the current USAR Appendix F 1500-psi criterion or to a
) coeparable criterion that it considers is consistent with current design
l
practice.
I Item (2) remains ooen until the licensee revises the referenced drawing to '
incorporate the node point for the valve operator.
Items (3), (4), (5) and (6) are closed. l
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Item (7) rem 4tns open until the licensee completes the design verification of
, computer ryn CFSR.
! RELATED GENERIC !$$UES
i
!
The licensee's inconsistent implementation of the seisaic criteria contained in l
USAR Appendix F is the primary generic issue that is related to Deficiency ;
"
D3.2-7. ,
1. The team recomends that the licensee review USAR Appendir. F and replace
seismic criteria that it no longer implements with comparablo criteria that it I
t
4
< considers compatible with current design practice. !
! ;
) Maintenance end use of design information and proper use and documentatien v
- engineering judgment are additional generic issues related to Deficiency 03.2-7. ;
I
- The team believes that the use.of plant-level design-basis documents (PLD80s)
l such as PLDBD-CS-51, "Seismic Criteria," and PLDBD-ME-10. "Pipe Stress and i,
? Supports " should enable licensee design staff to maintain design information ;
l in accordance with the requirements of the licensee's quality assurance manual. {
4 *
The use of General Engineering Guide GEG-3, "I'reoaration of Design Pcckages," ,
! Revision 0. December 1, 1987, which 3 argent & Lundy prepared for the licensee, >
1 should enable licensee design staff to use design information in a controlled !
, manner. '
i
! Finally, the team noted that an OPPD prc:edure entitled "Use of Engineering i
!
Judgment" was in draft form and was scheduled to be issued by May 1, 1988. The ~
! team believed that the procedure, when implemented in conjunction with the
j
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plaid level dssign-basis documents that the licensee was preparing, should ;
substantially reduce the incidence cf undocumented engineering judgment in i
design modification packages, i
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(Closed - New Item) Deficiency D3.2-8, Auxiliary Feedwater Piping Analysis
Design Input Loads
BACKGROUND
During the week of April 4, 1988, the team reviewed the auxiliary feedwater
(AFW) system design-basis document SDBD-FW-AFW-117, which Stene & Webster
prepared for the licensee, and which the licensee issued for trial use on
March 1988. The licensee was preparing a series of plant-level design-basis
documents (PLDBDs) that the AFW system design-basis document (and other SDBDs)
will incorporate by reference. These PLDBDs should specify the design-basis
loads required to qualify the AFW system.
Since it could not review the PLDBDs, the team ssked the licensee to confirm
that several USAR Appendix F licensing commitments .nd piping code design
requ' verits were currently controlled and accessible and that it had
' t. d these criteria in the AFW piping analyses of record.
. ,censee engaged Gilbert /Comonwealth (G/C) to analyze the AFW system in
response to Office of Inspection and Enforcement Bulletin 79-14 and performed
additional analysis to address Generic Letter 81-14.
,
FINDING
,
The team found that the associated stress analysis did not consider seismic
anchor movements between structures, equipment no:zle thermal movements, and
friction loads for pipe supports. Specifically,
(1) G/C did not consider the effects of seismic anchor movements (SAMs) on the
AFW piping supported between the containment and auxiliary buildings, as
required by Section F.2.2.2 of USAR Appendix F. The AFW piping in the
vicinity of containmcat penetration M-97 was shown on OPPD Piping Isomet ic
Drawings D-4236, Revision 1. .heet 1 of 1, and D-4238, Revision 2, sheet
7 of 7. The USAR did not tabulate St"s, and the licensee could not
provide the team with a design docu.unt that tabulated SAMs or specified
consideration of SAMs.
(2) G/C did not consider the effects of the rteam ger.erator nozzle thennal
displacements (TAMS) on the piping shown on OPPD Piping Isoraetric Drawing
D-4236, Revision 1, sheet 1 of 1. G/C's reanalysis of the AFW piping
subsystem resulted in an erroneous replacement of a snubber adjacent to
the steam gencator nozzl1 with a stcut. In 1986, licensee plant staff
noted that the strut was uamaged. Licensee dssign staff determined that
G/C had not considered the stean generator TAMS. and comouted the TAMS
in OPPD Calculation FC 001502, dated May 1, 1987. The licensee reanalyzed
the affected piping and prepared Modification Request MR-FC-87-23 to remove
the strut.
(?) The licensee did not consider the effects of the turbine-driven AFW pump
turbine inlet TANS on the piping shown on OPPD Piping Isonetric Drawing
D-4318, Revision 2, she , 1 of 3. As a consequence, the turbine inlet
nozzle thermal loads tabulated in Attachment 6 of the AFW SDBD did not
include the effects of TAMS. With respect to items (2) and (3), the
'acensee could not provide the team with a design document that specified
consideration of squipment TAMS.
A-44
_. . -
_ . . .- - .- -
-- -
. .
(4) The licensee indicated that piping friction forces for pipe support design
were evaluated on a case-by-case basis, but that it could not provide the
team with a uesign document that specified consideration of pipe support
friction loads.
Item (1) was an example of the licensce's failure to implement a licensing
cccmi tmt nt.
With respect to items (2) and (3), Section 1-719.7.3 of USAS B31.7, the piping
code of record for the AFW system, required consideratior of equipment nozzle
thermal displacements.
Finally with respect to item (4), Piping Specification No.1 in Contract No. 763
(an uncontrolled document) required the design of pipe supports in accordance
with the criteria of Sections 120 and 121 of USAS B31.1, which specified
consideration of friction effects on pipe supports.
RELATED GENERIC ISSUES
The team reviewed only three design attributes that would have been addressed
in the PLDB0s had these documents been available and identified problems in
all three areas. Therefore, the adequacy of the 79-14 Program implemented by
G/C for the licensee was questionable and was viewed by the t2am as a failure
to nieet licensing comitments. This issue is closed for the purpose of this
inspection report, but remains an open licensing issue to be resolved between
the licensee and the NRC's Office of Nuclear Reactor Regulation.
.
A-45
i
_ - _ _ _ _ _ - _ _ - _ - - .
-- - - - _ - - - - -
.
.
(Closed) Observation 04.1-1, High Power Pate of Change Trip Bypass
BACKGROUND
Observation 04.1-1 concerned a design modification (FC-84-46, "High Power Rate
of Change Trip Alarm," Revision 0, March 6, 1984) intended to change the
operation of a reactor protection system bypass alarm so that the alarm would
actuate only when the corresponding trip function (high power rate of change)
was in effect and not automatically bypassed. The existing alarm was contin-
uously active below 10 E-4-percent power and above 15-percent power; during
these conditions the trip bypass was autcmatically in effect, and the existing
alarm was thereby actuated. The licensee's basis for the modification was tu
improve the "black board" characteristics of the annunciator system during
normal plant operation. However, the SSOMI team had observed that the "black
board" basis in this case conflicts with the requirement of IEEE Std. 279-1968,
"Criteria for Protection Systems in Nuclear Power Generating Stations " Section
4.13, that continuous indication of any protective action bypass be provided in
the control room. The team had noted that the final design package anC the 10
CFR 50.59 analysis had not identified the design basis specified in IEEE Std. 279.
STATUS OF FINDING
Fc11owing further review by the licensee after this finding was identified,
this modification was cancelled (GSE-FC-85-1276, "Resolution of CAT / MOT Team
Findings for MR-FC-84-46." December 31,1985).
RELATED GENERIC ISSUES
To address the remaining generic concerns, the team requested that the licensee
review all annunciator modifications performed under the "black board" improve-
ment program to verify that no other reactor protection / engineered safety
features actuation bypass alarms had been modified in conflict with the
requirement of IEEE Std. 279 cited above. In addition, the t2am asked the
licensee to demonstrate that its design-basis reconstitution program and the
subsequent design.: basis verification will address the requirements of IEEE Std. 27L for bypass indication, noting that the reactor protection system was not
included in the licensee's design-basis document (DBD) priority list
(GSE-FC-88-17. "lRR Log No. 870172 Select and Prioritize Candidate Systems for
Design Basis Document Development," January 7, 1988) or plant-level P.
(GSE-DB-88-17, "Attachment II, Plant Level Design Basis Document," Ap i. 7,
1988). ,
In r 3ponse to the team's request, the licensee reviewed the 17 "black board"
mod 1/ications and provided a sumary of these modifications to the team
("Review of ' Blackboard' Modifications Associated With the Fort Calhoun Station
Annunciatur Upgrade Program," received from W. Gartner, April 5,1988). The
team agreed with the licensee's conclusion that none of these r'odifications
involved protection system bypass indication; therefore, the team's generic
coacern about past riodifications is resolved. Regarding the team's concern
about future modifications, the licensee provided a letter f rom the contractor
st.pporting the DB0 development effort (Stone & Webster Letter 0B-100, from
Beach tu S. Gambhir, "Co.itrol Circuit Bypass Indication Design Basis Project,"
April 5, 1988); this letter stated that the design requirements for bypass
indication would be included in the contractor's effort and that the licensee
should ensure that other organizations developing D20s for Fort Calhoun include
A-46
_ _ _ _
_ _ . - .. _ _ _ _ _ -.
-- . .. . _ _
. .
. . .
these requirwients wherever applicable. On the basis of team's understanding
that- the licensee will ensure that these requirements of IEEE-Std. 279 are also
appropriately addre ':'A if other contractors are involved, this ictin is closed.
,
f
-
4
r
i
f
2
4
o
P
1
i
6
1
t-
.
[
A-47
. - _ , - - _ - - . _ _ _ _ _ .
.. .
(Closed) Observation 04.2-1, Delta T Power Loop A^alysis
.rLCKGROUND
Observation 04.2-1 concerned the evaluation by the technical services staff
of replacement resistance tei. rature detectors (RTDs) and temperature trans-
mitters (FC-84-140, "Delta T Puer Process Loops") provided for measurtment
of reactor coolant system hot- and cold-leg temperatures; these instruments
provide inputs to the reactor protection system. The SS0MI team found that
the evaluation had not identified the channels as safety-related or as having
calculations involving critical quality equipment (CQE), the analysis
(05AR-85-83, "Uncertainty Evaluation for MR-FC-140," September 30,1985)
presented input values and final results without a traceable calculation, and
the applicable technical services prN-dure (N-TSAp-5, "Operations Support
Analysis Report Documentation," Reviuan 1, May 1985) did not contain the
CQE 'dentification requirement of a .amilar GSE procedure (Generating Station
Engineering Procedure B-9, "Technict Calculation Froduction Checking, and ,
Approval," Ja.1uary 1984). The tear's generic concern was the consistant
implementation of design changes among the licensee's various responsible
design organizations.
STATUS OF FINDING
During the inspection, the licensee retrieved the calculation showing the
square-root-of-the-sum-of-the-squares (SRSS) method te determine channel
uncertainty (Technical Services Procedure N-TSAP-6, Fom 6-1, "Analysis
Objectives and Methods Record," page 2, Revision 1, May 1985); the reinspection
team found that this document provided c traceable calculation and method.
RELATED GENERIC ISSUES
Regarding the team's SSOMI concern about determining the safety cignificance
of evaluations by the technical services staff after the SS0MI, the licent,ee
revised Procedure N-TSAP-5, "Operations Support Analysis Report Docu.ientation,"
Fom 5-1, Revision 2 March 1986, to include a checkoff for safety-related/CQE
system components . This action resulted in consistent CQE identification
requirements for GSE and technical services staff and resolved the team's
generic concern.
On the basis of the above clarification and corrective action, this item is
closed,
t
,
A-48
. _ _ _ _ _
-.
.
_
. _ _ . ._ - _ _
.- .
.
.. .
(0 pen) Deficiency D4.3-1, Limit Switch Circuit Prctection by Fusing,
MR-FC-84-74A
BACKGROUND
Deficiency D4.3-1 concerned the isolation of electrical faults caused by
postaccident submergence of limit switches for iine safety-related pilot
soleno:d-operated valves. The licensee had provided low-current, fast-acting
fuses in the indicating light branch of the valve control circuits with the
intent of retaining valve operability and sacrificing position indication,
since the limit twitches had not been rigorously qualified for submergence.
Successful isolation of the faulted limit switch depends on coordination of
the solenoid circuit fuse and the limit switch branch fuse. The SSOMI team
had found that the design package (MR-FC-84-74A, "Fuse Protection for Certain
Limit Switch Circuits," and Design Package Checker's Checklist, FC-84-74A,
Revision 0, May 31, 1985) had not substantiated that the fuses were enordina-
ted. The team had also observed from the manufacturer's catalog data that
the circuit interruption time differential may be only 10 milliseconds or
less (Bussmann "MIN" fuse catalog, 10 and 15 ampere ratings, and Bussmann
"KTK" fuse catalog, 0.25 and 0.50 ampere ratings), which would not ensure
reliable isolation of the faulted limit switches. As a result, continued
operability of charging line isolation valves HCV-238 and HCV-239 could not
be ensured for long-term core cooling following an accident. The team's
generic concern was that an identified technical assumption (i.e., that the
fuses would coordinate properly) had not been verified during development of
the modification design package.
STATUS OF FIND!hG
After the SSOMI, at the licensee's request, the manufacturer conducted five
repetitive tests to demonstrate successful coordination of the 10-ampere
Bussmann "MIN-10" and 0.5-ampere Bussmann "KTK-1/2" fuses. Bussmann reported
that the overall test results were successful [ letter from S. R. Coble
(Bussmann) to R. Clemens (OPPD), "Fuse Coordination," dated October 10,1985];
however, the licensee was unable to retrieve a vendor test report documenting
the test methodology, conditions, and certification. In addition, a subsequent
field modification had changed the control circuit fuse rating from 10 ar.. peres
to 7 amperes to protect replacement electrical penetration feodthrough assem-
blies that had been added to the circuit (Calculation Sheet MR-FC-84-74A, "Fuse
Protection for Containment Limit Switches, Field Change No. 6 to SRDC0-85-31,"
November 13,1985). The replacement fuses were also changed from the MIN-10 to
the KTK-7 type. Thus, the final configuration of the fuses is different from
the configuration tested (with respect to rating and type), and supporting
documentation was not evident for the tested configuration. Accordingly, the
team still has conce ns about the demonstrated ability of the fuses to success-
fully coordinate under design-basis conditions.
The team identified two additional concerns with respect to USAR requirements
governing the design basic dor seven of the nine valves that are the subject
of this finding. The first concern involves the basis for sacrificial isola-
tion of the position indication limit switches. Specific USAR comitments for
containment isolation (USAR Section 5.9 page 5.9.-5, Revision 3. July 1987)
require that "the status of all automatic [ containment isolation) valves, open
or closed, is indicated in the control room." The intended sacrificial isola-
tion violates this requirement for the seven containment isolation .alves.
A-49
.. .
.
The second concern regards a USAR comit. ent (USAR Section 5.9 page f.9-5,
Revision 3, July 1987) to "incorporate fail-safe provisions" for automatically
operated valves. Valves HCV 438A and C (component cooling water supply to the
reactor coolant pumps) as well as the charging line valves previously cited do
not meet the USAR comitment, since these valves are energized to close [0 PPD
Drawing (schematic diagram) 11a05-EP-438, Revision 10]. Apart from the USAR
discrepancy, the team is primarily concerned about the potential for defeating
containment isolation if the fuses for the two component cooling water valves
should fail to coordinate. The charging valves are of less concern in this
regard because they would not be required to close until later in the accident,
the fuses are in the control room, and more time would be available to replace
upstream fuses (if necessary to restore operability).
To address its concerns about fuse coordination and the discrepancy in the USAR
containment isolation commitment, the team asked the licensee to provide a basis
whereby postaccident operation will not be unduly compromised by loss of posi-
tion indication due to postaccident flooding or by failure of the subject fuses
to coordinate properly. The following points need to be addressed:
(1) the means available to the operator to detect and respond to these
situations in a timely fashion
(2) the required and allowable operator response time, based on the
flooding scenarios
(3) assurance that any operator action is appropriately governed by
existing procedures
(4) resolution of any inconsistencies between USAR Section 5.9 and the
as-built design with respect to position indication and "fail-safe
provisions" required by the USAR
(5) the degree of dependence on successful fuse coordination
This item remains ,open pending either revision of the USAR requirements cited
above with a basis provided for these exceptional valves, or rigorous
demonstration of the qualification of these limit switches for postaccident
submergence.
RELATED GENERIC ISSUES
There are no additional generic issues assDciated with this finding other than
those previously discussed.
A-50
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. . .
.
(Closed) Unresolved Item U4.3-2, ESF Bypass Switch Keylock Provision,
MR-FC-81-102
BACKGROUND
Unresolved Item V4.3-2 concerned apparent inconsistencies in requireraent
documents for a modification that would ada keylock bypass switches to
engineered safety features actuation channels for pressurizer low pressure and
steam generator low pressure (MR-FC-81-102, "B
Without Jumpers," Revision 0, August 14, 1985)ypass . The or Trip of
SSOM1 ESF
team Channels
had noted that
the final design modification package contained no requirement for keylock
cylinder combinations and the number of keys needed to control bypass of
individual trip channels. Cylinder locks and keys were stipulated by the
purchase order for the switch enclosures (Purchase Order 98505-CB, August 5,
1985), but Technical Services Review and Evaluation Form B (OPPD form B.
"Technical Services Review and Evaluation," April 25,1983) requested that
different keys be used for individual trip and bypass functions; the latter
requirement was not reflected in the purchase order or the final desi;n modifi-
cation packsge. The team's generic concern was that OPPD Design Procedure B-2,
"Production of Design Description and Evaluation," January 1984, appeared to
have been violated in that the technical description and design evaluation did
not appear to contain all of the requirements necessary to establish an
unambiguous design configuration.
STATUS OF FINDING
Af ter the 550MI, the Generating Station Engineering staff recomended cancel-
lation of this movification [GSE-FC-87-2098, "RRD Item 860199, Bypass Switches
(MR-FC-81-102)," December 30,1987]. Although the team believes that the
proposed change could be an 16provement in the means provided for administra-
tive control of channel bypass it acknowledges that the original design basis
does not require manual bypass capability using keylock switches; the basis
for channel inoperability requirements is provided in Amendment 88 to the
Technical Specifications (OPPD Docket No. 50-285, Fort Calhoun Station,
4 Amendment 88, May ,9. 1985) and does not require the proposed switches.
RELATED GENEP,IC ISSUES
Regarding the SSOMI team's generic concern about the design modification process,
the licensee developed GEG-3, "Guidelines for Preparation of Design Packages,"
Revision 0, November 1987. This procedure provides more uniform, specific, and
comprehensive guidelines for unambiguously defining the intended design con-
figuration. For example, Section 5.6.16 requires that the modification
preparer evaluate the possibility of operator error involved with the change.
On the basis of the licensee's improved design procedures and its intent to
train personnel in their effective implementation (GSE-FC-87- D30, Subject:
IRR Log Item 87-0185, Program Plan for Updating / Improving Existing Procedures,
December 15,1987), this item is considered closed.
A-51
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_ _ _ _ _ _ _
.. .o
(Closet') Obsarvatior. 04.3-3, Procurement Requirements on Equipment Vendors
BACKGROUND
Observatiun 04.3-3 concerned the inconsistent dccumentation of equire.ent
performance requirements and of vendor compliance during the procurement
process for certain Class IE isolation devices (FC-83-109, "Transfer of P250
Points to the ERFC"), keylock bypass switches (resolved under Item U4.3-2), and
replacement pressure switches for containment high-pressure channels (FC-83-83,
"Containment Pressure Switches"). The SSOMI team had noted that consideration
should be given to improving the effective use of design requirement documents
as inputs to the design engineer and to improving traceable verification that
the vendor performance data meet the design requirements.
STATUS OF FINDING
Regarding the specific design requirements for the isolation devices and pres-
sure switches, the team reviewed the licensee's procurement file and specifica-
tions (157-TR 02, "Test Report on Isolation Testing and Measurements of the TEC
Model 156 Se.-ies Isolators, Including Shorts Opens, and 120 Vac Fault With
Fuses Shorted," Revision 2 September 12, 1961, and GSE File No. 14915
Specification No. 6.20, Sheet 1, Revision 3) to ensure that the instruments
were pro:ured to appropriate specifications; no discrepancies were found.
PELATED GENERIC ISSUES
In response to the team's generic concerns about the documentatson of perfor-
mance specifications and conformance, the Generating Station Engineering staff
improved these aspects of its procurement process by issuing General Engineer-
ing Gcides GEG-3, "Guideline for Preparation of Design Packages," Revision 0,
November 1987, and GEG-2, "Guideline for Preparation of Procurement Specifica-
tions," Revision 0, August 1987; and by revising QADP-12, to improve its receipt
inspection process by the addition of a checklist. These documentt M ovide
more specific and generally comprehensive technical and procedural guidance for
developing and verifying design requirements. However, the team was unable to
detennine from these documents how electrical isolation requirements (a design
attribute of this finding) would be established for specifications. The
licensee advised the team that isolation requirements were being addressed by
Electrical Design Criteria EDC-1, "Electrical CQE Equipment Independence
Criteria," Revision 0, March 1988, and the team was given a draft copy. Since
this document nad not been issued, the team did not formally review it, but it
noted that Section 7 addresses electrical isolation. On the basis of the
licensee's commitment to establish a specific design basis for electrical
isolation for guidance in the preparation of specifications, this item is
closed.
After the reinspection, EDC-1 was issued and the electrical isolation require-
ments were included. The team made several observations regarding this
document, which are presented in the discussion of Unresolved Item U4.4-1,
- Design Basis Physical Separation Within Panels."
A-52
.. .,
(0 pen) Unresolved item U4.4-1, Design Basis Physical Separation Within Panels
BACKGROUND
Unresolved Item U4.4-1 concerned the apparent lack of a quantitative / measurable
design basis for the separation of redundant safety-related internal panel
wiring and the separation of safety-related and nonsafety-related internal
panel wiring. The S50MI team had reviewed both current and previously imple-
mented design modificatians. As specific concerns, th? team identified the
basis for the use of braid as a separation barrier (Purchase Specification
GSEE-0505, Alpha Wire Corp., Revision 0, April 28,1977); the basis for
allowing internal panel wiring for redundant divisions to be in contact
(MR-FC-77-40 "Undervoltage Protection," Revision 0, August 13, 1978); the
basis for allowing a multiple wafer switch to serve as a separation barrier
between safety-related and nonsafety-related panel wiring (MR-FC-81-102,
"Bypass or Trip ESF Channels Without Jumpers," Revision 0, August 14,1985);
and violation of a panel wiring separation requirement imposed by the GSE
Wire List Fonn (draf ting form for direction of construction), Note 2.
STATUS OF FINDING
A comitment made in 1970 in Appendix G of the Final Safety Analysis Report
for Fort Calhoun requires that physical individual channel components and
wiring be physically separated wherever such separation is practicable; IEEE Std. 279-1968, "Proposed IEEE Criteria for Nuclear Power Plant Protection
Systems," Requirement 4.6, requires that redundant channels "shall be
independent and physically separated to accomplish decoupling of the effects
of unsafe environmental factors, electric transients, and physical accident
consequences documented in the design basis, and to reduce the likelihood of
interactions between channels during maintenance operations or in the event of
channel malfunction." The team found no analysis demonstrating that the wiring
configurations cited above meet these specific requiremsats to which Fort
Calhoun had been licensed.
Regarding the violation of the wire list form requirements prohibiting common
harnessing of redundant trains, the licensee explained that the form was a
document used for providing wiring instructions for the construction effort
and the footnoted requirements could not be achieved when modifying original
panels and were not considered a consistent design basis.
Regarding the use of multiple wafer switches as a separation barrier between
safety-related and nonsafety-related panel wiring, the licensee stated in its
response [LIC-86-106, "SSOMI (Design) 50-285/85-22," Attachment A, Item UA.4-1,
April 15, 1986) that this practice is consistent with Section 7.2.2.1, para-
graph 3 of IEEE Std. 384-1981, ("IEEE Standard Criteria for Independence of
Class IE Equiprent and Circuits"), which allows separation of isolation device
input / output wiring to be less than six inches if the separation is not less
than the distance between input and output terminals. The team finds this
i acceptable only if the ability of the isolation device to withstand maximum
credible levels of voltages and fault currents has been demonstrated by
analysis and test (also a requirement of IEEE Std. 384-1971). The team found no
such analysis or test. Consequently ~, the team believes the design basis is
incomplete for supporting this practice.
A-53
._ _ _
_ _ _ __ _ _ _ _ _ _
. _ _
.. ..
\
Regarding the practice of common routing / bundling of panel instrumentation and
control wiring and the use of metallic braid as a barrier, the licensee
retrieved the minutes of meetings with the control room panel vendor (Meeting
minutes, Gibbs & Hill / General Electric NID at Gibbs & Hill office, September
21-29, 1970 Scope: [ Panel] Separation Criteria) during the followup inspec-
tion. These minutes indicated that safety-related Danel wiring must incor-
porate a copper braid shield and overall jacket to permit the common bundling
of wiring for as many as four redundant safety divisions. The team found no
analysis to support this use of the braid as a barrier. Discussions with the
licensee and a review of the wiring specifications indicated that the braid is
ungrounded and provides 85-percent coverage. The team concludes that the braio
does not appear to provide a significant barrier to propagation of either
electrical faults or localized combustion, although it may provide mechanical
protection during installation or modification. Accordingly, the team finds
that the licensee has not demonstrated a design basis for the use of this braid
as a separation barrier.
The licensee issued EDC-1, "Electrical CQE Equipment Independence Criteria,"
Revision 0, March 1988, after the 550M1 followup inspection. The licensee
gave the team a draft of these criteria during this inspection. The draft
document indicated that the practice of commonly bundling the braided wire
has apparently been continued for wiring modifications to safety-related
panels more recently procured to the requirements of Regulatory Guide 1.75
and IEEE Std. 384(e.g., the alternate shutdown panel). The team finds that
this common bundling practice represents an unacceptable relaxation of the
criteria stipulated when the panels were procured unless a basis were estab-
lished justifying exceptions.
The approved EDC-1 examined after the followup inspection deleted specific
reference to this wiring modification practice, but did not appear to
prohibit common bundling of different safety division wiring; the document
equivocally states, "Wires of different classification are not generally
bundled together." Again, the reinspection team found no analysis or test
to support any exceptions to Regulatory Guide 1.75 and IEEE Std. 384, which
were the separation criteria levied when these newer panels were procured.
The team did not formally review EDC-1 beyond the scope of the concerns of
this finding; however, it made the following cursory observations regarding
several other provisions of the document that could affect instrumentation and
control design:
, (1) EDC-1 permits the use of circuit breakers actuated solely by overcurrent
,
as well as fuses to be used as isolation devices; these are exceptions
(2) Where credit for fuse coordination is permitted under the fort Calhoun
design basis, sufficient margias of coordination must be rigorously
l demonstrated, documented, and maintained. Deficiency 04.3-1, "Limit
Switch Circuit Protection by Fusing MR-FC-84-74A." supports this generic
observation.
(3) Separation criteria for instruments and instrument lines must also include
requirements for hazardous areas (e.g., areas containing high energy lines,
major rotating apparatus, or other potential sources of high energy),
where the specified minimum distances may not be sufficient.
A-54
- _ _ _ _ _ .
.. .
.
In summary, the team finds that the licensee has made significant progress in
reconstituting the electrical independence criteria that apply to the original
design as well as to more recent modifications. Specific and quantitative
criteria are provided in EDC-1. However, the team concludes that the required
analyses and tests supporting the practice of common or proximate routing of
internal instrumentation and control panel wiring of redundant safety divisions
(and safety /non-safety divisions) are not evident. This is especially important
in the case of modifications to newer panels to ensure their design basis is
not degraded by subsequent modifications. Accordingly, this unresolved item
remains open pending the licensee's establishment of an acceptable design basis
for this practice. This design basis must be appropriately supported by analyses
or tests that consider maximum credible faulted conditions on the circuits of
interest.
RELATED GENERIC ISSUES
ho additional generic issues have been identified other than those just
discussed.
..
i
A-55
_ _ _ _
.. .,
(Closed) Deficiency D4.5-1 Drawing Changes by Procedure A-9, MR-FC-82-178
BACKGROUND
Deficiency D4.5-1 concerned a modification re
DP Indication," Revision 0, January 23, 1984)quett (MR-FC-82-178,
for adding "HEPA Filter
local air filter
differential pressure gauges, for which a sepia print of the piping And
instrumentation diagram (P&lD) was not issued for developing the modification.
This apparently did not conform to Generating Station Engineering Procedure
A-9, "Document Control," Section 2.3.3.4, August 1983, which requires the
design engineer to request a sepia print of an existing drawing that requires
, revision (such as the P&ID).
STATUS OF FINDING
In its response and in discussions with the followup team, the licensee clari-
fied the intent and meaning of this provision of Procedure A-9. Sepia prints
are issued to (1) provide a mechanism for showing the proposed design change
daring the development of the modification without changing the document of
record until the modification is installed and "as-built" and (2) inform all
who use the drawing of record about a pending modification that may affect the
drawing.
Regarding the first purpose, the team agrees that in this case, the modifica-
tion was adequately defined by the engineering sketches provided with the
design package and the P&ID sepia print was not required to describe the change.
Regarding the second purpose, the team asked how users of the P< would be
informed of the change. The licensee cited the requirement of Standing Order G-21
"Statior Modification Control," Form J. Revision 23, April 10,1987, that P&lDs '
(and all other control room drawings) must be updated before system accep-
tance; this would ensure that users are informed of the change.
The licensee also clarified the meaning of the Procedure A-9 requirement,
"when an existing drawing needs revision during the preparation of an MR,
a request for a sepia of that drawing is made," by stating that the intent
is to limit sepia print production to only those drawings that need revision
to install the modification, not to all drawings that may eventually need
revision. In this particular example, the licensee judged that the modifica-
tion was simple and did not have a significant impact on the P&lD.
On the basis of the licensee's clarification of the intent and meaning uf
Procedure A-9, the team concludes that this item is closed.
RELATED GENERIC ISSUES
The team does not agree with the portion of the licensee's response stating
that the modification "did not have significant impact on the P&lD," since
the addition of the two instrements represented a significant change in
functionality, even though the change itself was comparatively minor. As a
matter of good practice, a consistent threshold of "significant drawing
impact" should be defined and maintained when issuing sepia prints under
Procedure A-9.
A-56
\
,. ..
(Closed) Observation 04.5-2, Flow Element Design Basis Conditions
BACKGROUND
Observation 04.5-2 concerned discrepancies in specified environmental
conditions for replacement flow elements in the component cooling water system.
The values specified were not consistent with design-basis conditions specified
in GSEE-0802, "General Requirements for CQE (Class IE) Electrical Equipment
Required for Use in Controlled Access Areas of the Auxiliary Builoing Outside
Reactor Containment," Revision 0, July 14, 1980.
STATUS OF TINDING
The team had determined that although the actual values were technicall/ accep-
table, a generic concern remained.
RELATED GENERIC ISSUES
In response to the team's generic concern, the Gcaerating Station Engineering
(GSE) staff improved this aspect of its procurement process by issuing General
Eng19eeri ig Guides GEG-3, "Guideline for Preparation of Design Packages,"
Revision 0, hovember 1987, and GEG-2, "Guideline for Prepar4 tion of Procurement
Specifications," Revision 0, August 1987. These documents provide mure specific
and generally comprehensive technical and procedural guidance for developing
and specifying system performance requirements (Section 5.4.2 of GEG-3) and
environmental conditions (Section 5.4.3.1 of GEG-3).
On the basis of these programmatic improvements to the specification process,
this item is closed.
.
l
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A-57
..-. . -. , - - _ _ . - . _ - .__. _ _ _ . . ... - - - . . _ . - .
.. .
.
(Closed) Unresolved Item U4.5-3, Battery Room Fire Hazard Analysis
BACKGROUND
Unresolved item U4.5-3 concerned a masonite/ fiberboard fuse block enclosure
located in each battery room that had not been identified in the battery room
fire hazard analysis (FHA); since the coatin;. specifications could not be deter- l
mined, the combustibility of the material also could not be detennined.
l
,
STATUS OF FINDING
In response to this findinc, the licensee painted the enclosures of concern
withfire-retardantpaint[MainteranceOrder 92 858044, December 23, 1985 l
(completed December 27,1985), and performed a new FHA for Fort Calhoun ;
Station ["Fire Areas 37, 38 (Battery Rooms 1 & 2)," Revision 0, October 12, i
1987]. In reviewing the documentation of these corrective actions, the
reinspection team noted the following:
(1) The maintenance order (MO) for the repainting did not indicate a stock
number or other unique identifier for the coating used, although when
signing off on the completion of the repainting, craft personnel noted
that fire-retardant paint had been used [M0 92 858044, December 23, 1985
(completed December 27,1985)]. The team asked the licensee now assurance
was provided that the proper fire-retardant paint had been used. The
licensee referred the team to a memorandum that listed acceptable
fire-retardant coatings (FC-730-85, meinorandum listing acceptable
fire-retardant coatings, dated May 1985) and stated that this memorandum
,
is the basis for the application of fire-retardant paints.
On this basis, together with consideration of the notation craft personnel
had written on the MO, the team concludes that the repainting was done
correctly. The team further concludes that more positive identification
of the coating material should be provided on the M0, and understands that
the licensee intends to do so.
2. Through an oversight in the updated FHA, consideration of the combustible
loading of the enclosures in question had still not been included. The
licensee corrected the updated FHA ["Fire Areas 37, 38 (Battery Rooms 1 &
2)," Revision 1 April 7, 1988] during the inspection to include the
enclosures, and there was no significant effect on the analysis or
conclusions.
I
RELATED GENERIC ISSVES
i
The team considered the generic implications of the oversight in the FHA and
reselved them as follows:
(1) The oversight appu rs to be an isolated instance involv ng comparatively
low combustible loading (much lower, for example, than tae battery case).
(2) The fuse block enclosures are in separate fira areas.
(3) This type of fuse block enclosure would not be expected to reflect comen
practice in safety-related areas of the plant.
A-58
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(4) The licensee's housekeeping procedures have been improved to provide more
specific restrictions and checklists regarding potential fire hazards por.ed
by such materials as wood (Standing Order G-6, "Housekeeping," Revision 18
May 15, 1987).
(5) Preparation of design packages is now governed by GEG-3, ("Guideline for
Preparation of Design Packages," Revision 0, November 1987, Section 5.6.1,
"Fire Protection," which requires that fire-protection requirements be
considered.
On this basis, this item is closed.
.
A-59
_
.. . .
(0 pen) Deficiency D5.1-1, Battery Sizing Calculation
BACKGROUND
Deficiency DS.1-1 concerning the battery sizing calculation developed for the
1984 modification, MR-FC-84-119, used an unverified 1979 load profile without
justification. In resaonse to this SS0MI team findino the licensee, aided by
an outside consultant Applied Power Associates (APA)),, revised the de load
profile. This new load profile will increase the required ampere-hourt removed
from the battery by 21-percent and 19-percent on de buses 1 and 2, respectively.
STATUS OF FINDING
The APA battery sizing calculatio; contained two assumptions that the reinspec-
tion team wanted confimed. Tne first involved the trinimum battery temperature.
Cell temperature affects battery capacity. The APA calculation assumed a mini-
mum temperature of 70'F. Historic monthly surveillance data supplied by the
licensee for 1984 and 1987 indicated that the battery temperature had remained
above the minimum temperature used in the calculation. The battery surveillance
procedure, ST-DC-1, contained an acceptance criterion for maximum cell tempera-
ture of 90*F; however, no surveillance criterion existed for reinimum temperature.
In response to this concern, ST-DC-1 was revised (April 8, 1988) to alert opera-
tions personnel to the minimum allowable battery temperature.
The second assumption questioned in the sizing calculation involved the lack
of a correction factor for battery capacity degradation. This implied that
the fully charged battery capacity remained above 100-percert. The team
reviewed the battery perforinance test performed in 1985 following the battery
modification and the service test perforcied during the 1987 refueling outage in
an attempt to confirm the battery capacity. The team was not able to confim
the existing capacity because of inconsistent test data and incorrect test
acceptance c,iteria. In response to this concern, the licensee corrected the
surveillance test acceptance criteria. The team was also informed that the
licensee is in the process of obtaining the battery manufacturer's (EXIDE)
assistance to analyze the test results in order to establish the degree of
capacity remaining in the batteries. The licensee indicated to the team
that the results of this analysis will be factored into the battery calcula-
tions and surveillance procedures.
This item will remain open pending the licensee's confirmation that the
battery's existing capacity is greater or equal to the capacity required in the
calculation.
RELATED GENEP.lC ISSUES
A generic issue related to this finding concerns the independent review program.
The licensee has issued Design Procedure B-11, which satisfactorily addresses
independent design verification concerns.
A-60
._.._ _
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_ - _ _ _ . . _ _ _ __
. . ._ -. -.
F
.. . .
(Closed) Unresolved Item US.1-2, Battery Charger /DC Bus Coordination
BACKGROUND
During the 1985 refueling outage, the licensee replaced the original 200-ampere
battery charger with a new 400-ampere unit. The new charger had the capability
to limit current up to 500 amperes (when recharging a discharged battery or
other de transient). The new de breaker corinecting the charger to this dc bus
was only rated at 400 amperes and could trip at 500 amperes. No test or setup
procedures had been written before the 1985 S50M1 took place. In response to
this finding, the licensee prepared Test Procedura M0-871643, which reduced the
current limit value to 380 amperes to ensure that the bus breaker would not
trip on battery recharge.
STATUS OF FINDING
The team reviewed Test Procedure M0-871643 and found it acceptably ensured that
the current limit of the battery charger was reduced to approximately 380
amperes.
RELATED GENERIC ISSUES
A generic issue related to this finding is the adequacy L7 postmodification
test procedures. The licensee recently had General Engineering Guides GEG-9,
"Electrical System Interaction," and GEG-28, "Preparation of Installation
and Test Procedures," prepared to address these concerns on a generic basis.
..
<
A-61
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.
. . . .
(Closed) Observatien 05.1-3, Power Cable Sizing Criteria
BACKGROUND
During the 1985 S50MI, the team noted that the licensee had no forwal cable
sizing criteria other than a vague reference in the Updated Safety Analysis
Report to industry standards (Insulated Power Cable Engineers Association)
that did not specifically apply to the installation conditions proposed for
modification MR-FC-84-119. In response to the team's imm.ediate concern, the
licensee prepared Calculation FC-00476, "Cable Sizing Calculations for
Modification Request MR-FC-84-119," December 20, 1985, which addressed voltage
drop and ampacity for the affected cables.
STATUS OF FINDING
The team considers that the application of the generic power cable sizing
guidance is sufficient to close this item.
RELATED GENERIC ISSUE <
The licensee contracted with Stone & Webster Engineering Corp. (SWEC) to
develop a generic design procedure for determining cable ampacity. The oraft
procedure, which was presented to the licensee on March 24, 1986, was still
undf r review when the team returned for the followup inspection. In response
, to .he generic concern raised again during the followup inspection, and fol-
l lowing review by Sargent & Lundy Engineers (S&L), the licensee issued a
ger<eric cable sizing guidance document (EEG-2, "Guideline for Power Cable
Sizing," Revision 0, April 1988).
'
.
A-62
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, - - _ - . _ . - - - _ - _ - - - _ _ _ _ - - _ _ , _ _ - - _ _ _ - _ _ - _ _ _ _ - - . . - _ _ _ . -
.
. . .
(Closed) Observation 5.1-4, Pre-Operational Test Requirements
BACKGROUND
The SS0MI team found that Modification Request MR-FC-84-119 "Battery Charger
and Inverter Replacement" had not contained any requirement to set the battery
charger's high voltage alarm or to provide for maintenance of the two cells
removed from each safety-related battery. In response to this item, that
alarm was reduced to 140 volts and documented in a postmodification test
conaucted on November 14, 1985. The licensee stated that it will not maintain
the spare cells as critical quality element (CQE) items; therefore, no separate
maintenance is required on the spare cells.
STATUS OF FINDING
l
The team reviewed Test Procedure M0-871643 (May 11, 1987) and verified that the
de high-voltage alarn was still set at 140 volts. The alarm response procedure,
OP-10-A15, also identifies the high de voltage alarm as set at 140 volts.
However, the team noted that the battery surveillance procedure, ST-DC-2 issued
on January 23, 1987, did not contain alarm acceptance criteria. In response to
this observation, the licensee prepared Maintenance Procedure MP-EE-22A on
March 29, 1988, to verify the battery charger alann setpoints on a periodic
basis. The team considers this response acceptable.
RELATED GENERIC ISSUE
in response to the generic concerns related to the preparaticn of modification
installation and test procedures, the ii.ensee issued Engineering Guide GEG-28
in February 1988.
.
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. _ _ . ,_ _ - .
I . . . .
(Closed) Observation 05.1-5, Inverter Sizing Without Analysis
BACKGROUND
During the 1985 SSOMI, the team observed that the licensee had replaced the
original inverters with new smaller-size units without comprehensively
analyzing the inverter loading. In response to the team's concern, the
licensee prepared an inverter sizing calculation for Modification Request
MR-FC-84-119, March 5, 1986.
STATUS OF FINDING
The team reviewed the inverter sizing calculation and noted that the
continuous load calculated for all four inverters was below the continuous
rating of 7.5 kVA. The team noted that the licensee recognized that the
potential load of 7.478 kVA on inverter A left little margin and the
calculation suggested some nonsafety-related load transfer.
The team has been told that the electrohydraulic control load (panel AI-50)
was moved from safety-related inverter A to neasafety-related inverter 2 by
modification MR-FC-86-42 to ease the loading on inverter A. The team finds
this response adequate to resolve the imediate concern.
RELATED GENERIC ISSUES
in response to the generic issue raised by this finding, the licensee had an
electrical system interaction guideline, GEG-3, prepared in February 1988
to ensure that electrical system loading is considered in modifications. The
Generating Station Engineering electrical department has designated an electri-
cal luad coordinator, and recent modifications have been reviewed by engineering
personnel to identify electrical load changes. Electrical load study update
forms are now issued for each modification before, closecut. In addition, other
load studies will be performed on the 480-volt and 4160-volt systems as part
of the design-basis reconstitution program.
'
.
A-64
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. .
(Closed) Observation 05.1-6, Design Interface Control
BACKGROUND
During the 1985 SSOMI, the team noted that Electrical Modification Request
MR-FC-84-119 had not been reviewed by the structural reviewers or by the
HVAC reviewers. In response to this concern, the licensee implemented a
multidisciplinary independent review program.
STATUS ON FINDING
The team reviewed Design Procedure B-11. "Independent Design VerificStion,"
February 1988, and considers the procedure acceptable.
RELATED GENERIC ISSUE
Multidisciplinary review of modifications is being addressed on a generic
basis by General Engineering Guide GEG-3, "Preparation of Design Packages."
_
b
A-65
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.
.
. .
(0 pen) Deficiency 05.2-1, Fire Wrap Protection of Cable Raceways
BACKGROUND
During the 1985 SS0MI, the team noted that the licensee had been using unverified
derating factors for cables in conduit wrapped with a firt-protective material.
These factors had been obtained from the material manufacturer's (3M) computer
program, but the computer program had not been verified and gave results that did
not agree with actual test data. The licensee responded that the team had only
reviewed preliminary calculations and identified other calculations that wa.e
being developed in accordance with its procedures for developing calculations
pertaining to critical quality equipment (CQE). The licensee identified those
calculations as follows:
'
(1) FC-85-25-001, "Load Study MCC 3A1, 3B1, 3C1," Revision 1, October 7, 1985
(2) FC-85-25-002, "Cable Ampacity Deratings - Power Feeder Cables for MLL 2.a),
3B1, 3C1," Revision 1, October 9, 1985 (for cable conduit)
(3) FC-85-25-003, "Cable Ampacity Derating - Directly Wrapped Power Cables,"
Revision 1, February 4,1986
STATUS OF FINDING
The team reviewed Calculation FC-85-25-003 covering the installation of fire
wrapping directly on the power cables. The calculation was based on cable
characteristics (resistance and diameter) from Insulated Cable Engineers
Association (ICEA) etandards rather than from the specific cable data used at
Fort Calhoun as detailed in the Rockbestos Cable Schedule (Dwg. #W-LIST,
File #47122). Also, the licensee had assumed unconservative thermal conduc-
tivity at 350*F for the fire wrapping; this assumption was not appropriate
for the Fort Calhoun application with 90'C rated cable (i.e., 200'F). The
team used standard Neher McGrath heat transfer methods (similar to those used
to develop the ICEA ampacity standards) and estimated that the errors noted
above would result in the allowable ampacity of the fire-wrapped motor control
center MCC-3C1 feider cable (EA 140) dropping below the licensee's calculated
cable load developed in Calculation FC-85-25-001. This item will remain open
pending revision of the cable derating capacity calculation and the providing
of justification to the NRC for the existing cable.
RELATED GENERIC ISSUES
l No additional generic issues were related to this item since the calculations
l
reviewed were all inclusive.
i
i
A-66
t
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. - . .
(Closed) Deficiency D6.1-1, Safety Evaluation for NonSafety-Related
Systems Described in the USAR
BACKGROUND
Deficiency D6.1-1 concerned
systems or >.quipment described infive nonsafety-related
the Updated Safety Analysis modifications
Report (USARaffecting)
that lacked a safety evaluation in their design packages. The USAR had to
be revised to reflect the modifications. The SSOMI team had noted that
10 CFR 50.59 requires a safety evaluation to be included if the USAR would
be changed by facility modifications; no distinction can te made in this
regard between safety-related and nonsafety-related systems.
STATUS OF FINDING
In response to this deficiency, the lit.ensee has taken several corrective
actions:
'
(1) Safety evaluations for the five modifications in question have since
been performed (LIC 87-086, April 10, 1987).
(2) Appropriate USAR changes have been made regarding the modifications
(RRD 87-0122).
(3) A comitment has been made in tne Fort Calhoun Station Design Basis
Reconstitution Project Program Plan, July 2,1987, page 31, to review
modification packages for safety-related modifications and nonsafety-
related modifications that could affect safety-related systems; this
review will include all modifications installed after the license is
issued and is intended to confirm the adequacy of safety evaluations.
'
(4) OPPD Procedures A-2, B-2, and G-21 were revised to require safety evalua-
tions for all modifications affecting facilities or procedures described
in the USAR, including drawings.
(5) A revised prdcedure for preparing safety evaluations has been
developed and implemented (GEG-27. "10 CFR 50.59 Safety Evaluation,"
Revision 0, February 1968).
(6) Procedure GEG-3 requires a safety evaluation during the preparation of ,
modification packages (GSE-DB-88-17, "Attachment II, Plant Level Design '
Basis Document," April 4, 1988).
I'/) A training program regarding the 10 CFR 50.59 process was conducted for
the licensee's staff.
Although tne team did not perform a technical review of all corrective actions,
on the basis of the programmatic changes cited above, this item is closed.
RELATED GENERIC ISSUES
2
No additional generic issues have been identified other than those discussed
above,
i
I
A-67
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(Closed) Unresolved Item U6.1-2, Safety Analyses for Emergency Modifications
BACKGROUND
During the 1985 SSOMI, the team found a number of emergency modifications to
critical quality element (CQE) items for which a safety evaluation was not
included in the final design package. The licensee admitted that although
these modification packages did contain a safety evaluation for the construc-
tion phase, no safety evaluation could be located for the design phase.
STATUS OF FINDING
The licensee revised Standing Order G-21 "Station Modification Control," to
require that safety evaluations be included for all phases of construction and
design. The licensee has also stried that it has reviewed all previous emer-
gency modifications and has prepared safety evaluations on all related to
critical quality elements.
The team reviewed the applicable Section 2.3, "Emergency Modification Requests,"
of Standing Order G-21 (page 19, April 10, 1987) and confirmed the requirement
for preparing safety evaluations during both the design and construction phases
and including them in the final design package. In addition, General Engineering
Guide GEG-27 provides guidance on safety evaluations for all types of
modifications.
RELATED GENERIC ISSUES
No additional generic issues were related to this item.
'.
1
A-68
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. 0
(Closed) Observation 06.1-3, Vital AC Inverter Bypass Mode
BACKGROUND
Observation 06.1-3 concerned a deficiency in the safety evaluation included with
Modification MR-FC-84-119 ("Battery Charger and Inverter Replacement, Lowering
Terminal Voltages and Battery Discharge Breakers"): the safety evaluation failed
to address any effects on the Technical Specifications of operating one or more
instrument buses in the bypass mode. The licensee's original interpretation
was that the instrument bus would only be considered tioperable if it was powered
from an inverter; the licensee changed its first interpretation to define operable
as "being powered from the bypass transformer" (TS FC-86-807, December 1,1986);
in GSE FC-87-507 (March 31, 1987) the licensee required the instrument bus to
be powered from an inverter if it is to be considered operable.
STATUS OF FINDING
Operating Instruction Ol-EE-4-6, Revision 35. September 24, 1987, prohibits
more than one safety-related instrument bus from being powered in the bypass
mode when the reactor coolant system temperature is above 300'F. The reinspec-
tion team agrees with the present interpretation of instrument bus operability
requiring that the safety-related inverter be operable and connected to the
instrument bus.
RELATED GENERIC ISSUES
No additional generic issues were related to this observation.
.
'e
A-69
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_,
r ,
,
- .,
(Closed) Observation 06.2-1, Untimely Closecut of Emergency Modifications
BACKGROUND
During the 1985 SS0MI, the team reviewed the closecut of emergency modifications
and found that six modifications lacked the after-the-fact design packages as
long as 42 months after the modification was installed. The licensee responded
that marked-up drawings were prepared and forwarded to the Generating Station
Engineering (GSE) staff so that it could prepare the after-the-fact design
packages (FC-1339-86, September 2,1986).
STATUS 01 FINDING
Section E.3 of Standard Order G-21, "Station Modification Control," Revisicn 29,
November 19, 1987, states that the final design package for emergency modifica-
tions must be (1) independently verified before the system or equipment is
placed in operation and (2) independently reviewed by the System Acceptance
Committee (SAC) within 14 days after the system or equipment becomes opera-
tional. The reinspection team considers this acceptable. Three of four open
emergency modification items identified by the SSOMI team were verified closed
by review of Memorandum FC-1339-86. The fourth modification was verified as a
component change not requiring a modification.'
REl.ATED GENERIC ISSUES
The regional followup inspection (IR 50-285/88-02) verified that the use of
emergency modifications had been severely restricted since the 1985 550MI. A
review by the regional staff of six selected recent emergency modifications
indicated that the final design packages for all six had been issued.
.
A-70
F'.
. . .
(Closed}DeficiencyD6.2.2, Modifications to AFW Turbue Steam Supply Valves
BACKGROUND <
From March 1980 to January 1985, the licenses failed to meet the requirements
of 10 CFR 50.59: the licensee made a change to the facility as described in
the Updated Safety Analysis Report (USAR) but failed to conduct and document
a review to detennine that the change did not involve an unreviewed safety
question. The change to the facility involved the modification of the auxi-
liary feedwater (AFW) punp turbine comon steam admit valve (YCV-1045) from
the "fail close" to the "fail open" design mode (the change was com91eted
in March 1980) without the addition of a safety-reiated air accumulator
system for the individual "fail open" steam supply valves (YCV-1045 A and B).
The inability to close the "fail open" steam supply valves on the loss'nf
nonsafety-related instrument air would result in an additional fission
product release path, not analyzed in the USAR, for a steam generator tube
rupture incident. Therefore, the change involved an unreviewed safety
question because the consequences of an accident previously evaluated in the
USAR may have been increased.
In addition, the USAR incorrectly reflected the as-brilt configuratior, for
the containment penetrations associated with YCV-1045 A and B. USAR
Table 5.9-1 shows these penetrations (M-94 and M-95) as Type IVD. Type IVD
contains a single power-operated valve that is normally open, fails closed,
and whose accident position is closed. The SSOMI team noted that although
this depiction is correct for the main steam isolation valves, the M W steam
supply taps off on the upstream (containment) side of the main steam isolation
valves, ano that '.bese valves are normally closed, fail open, and have an open
accident positio1. In addition, the main steam isolation valve bypass valves
are not shown.
STATUS OF FINDING
This finding is closed for the following reasons:
(1) The teum has tonfirmed that the air accumulators have been installed
and tested (see D2.1-1).
(2) FSAR Table 5.9-1 has been deleted and surarseded by Figure 5.9-13,
Sheets 1 through 65. Penetrations n . M-95 are depicted on
2
Figure 5.9-13, Sheets 62 and 63, ap9e w y. This figure shows the
correct configuration and valve piu'*..", e ing nonnal, failed, and
accident conditions.
'
(3) (a) OPPD huclear Production Division Policy / Procedure E-1, "10 CFR 50.59
Safety Evaluation," and (b) General Engineering Guide GEG-27." 10 CFR 50.59
Safety Evaluation," Revision 0, February 1988, have beer, prepared since the
'
SSOMI. The procedure and the guidance document have been reviewed and
should enhance the depth and completeness of 10 CFR 50.59 evaluations.
(4) Standirg Order G-21, "Station Modification Control." Revision 29,
November 19, 1987, states in Section 2.3 that the final design packagc-
for emergency modifications must be independently verified before the
system or equipment is placed in operation and must be independently
reviewed by the System Acceptanco Committee (SAC) within 14 days after
'
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_ __ _
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l
the system or equi > ment becomes operational. In essence, the modification
control procedure las been revised to maintain administrative control of
"emergency" modifications to the same level as "nomal" modifications.
The modification control procedure now requires that safety evaluations
which address both the design and the installttion and testing aspects of -
the modification be completed before the start of construction. These
safety evaluatior.s are required to be reviewed by engineering personnel,
whc are responsible for the design bases, before the start of construc-
tion.
(5) The licensee reviewed those emergency modifications that had modification
completion reports (MCRs) completed. [hote that a modification completion
report and System Acceptance Committee review fom she"ld be submitted to
the Genersting Station Engineerin. staff (GSE) within two weeks following
the accep.ancs of a modification. The review was conducted to confim
that two safety evaluations had been performed: one as part of the design
package and one as part of the installation and construction phase. The
licensee reviewed 71 emergency modifications which represented the set of
,
'
emergency mndifications installed since the modi'ication process started
and that have modification completion reports as of April 10, 1987. The
teas selected three modifications (FC 3' 35, FC-84-83, and FC-85-161) and
confirmed that two safety evaluations were included in the modification
file. The team did not assess the technical adequacy of these safety
evaluations.
(6) The team examined documentation that indicates all completed CQE-related
emergency modi.~ications installed since initial full-power operaJon, and
have an MCR received by April 10, 1987, that do not require scismic
updates, have been closed. All known drawings and applicable plant
procettures have been updated (GSE-FC-87-1032, memorandum from S. Gambhir to
J. Fisicaro, "Closecut of RRD ltem No.87-277 " dated June 24,1987).
(7) The team confirmed that personnel performing activities associated with
10 CFR 50.59 evaluations had been trained. The team compared the indivi-
duals who attended 10 CFR 50.59 Safety Evaluation Training (00D
- Memorandum FC2T-2y7-88, dated April 6, 1988) witn *.he personnel qualified
4
for performing reviews of safety evaluations (FC-1986-87, menorandum fruc
T. L. Patterson to PRC Chaiman, dated January 4,1988).
'
RELATED GENERIC ISSUES
The related generic issues are included in the preceding discut.on.
1
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