ML20205J880
| ML20205J880 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 03/23/1987 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20205J791 | List: |
| References | |
| 50-456-86-65, 50-457-86-49, NUDOCS 8704010443 | |
| Download: ML20205J880 (33) | |
See also: IR 05000456/1986065
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
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Report No. 50-456/86065(DRP);50-457/86049(DRP)
Docket Nos. 50-456; 50-457
License Nos. NPF-59; CPPR-133
Licensee:
Commonwealth Edison Company
Post Office Box 767
Chicago, IL 60690
Facility Name: Braidwood Station, Units 1 and 2
Inspection At: Braidwood Site, Braidwood, Illinois
Inspection Conducted: December 1, 1986 through February 14, 1987
Inspectors:
NRC
T. M. Tongue
W. J. Kropp
T. E. Taylor
EG&G Idaho, Inc.
B. Barnes
Approved By:
W. S. Little, Director
Braidwood Project
Date
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Inspection Summary
Inspection on December 1, 1986 through February 14, 1987 (Report No.
50-456/86065(DRP); 50-457/86049(DRP))
Areas Inspected:
Routine, unannounced safety inspection of activities with
regard to licensee action on previously identified items; allegations;
regional recuesti licensee event reports; operational safety; Unit 2 tour
which incluced construction activities in the areas of welds, electrical,
and mechanical; report review; and meetings, training, and other activities.
Results: Of the eight areas inspected, no violations were found in seven
areas, one violation was identified concerning reportability of ESF and RPS
actuations (Paragraph 4).
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DETAILS
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1.
Persons Contacted
Comonwealth Edison Company (CECO)
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Corporate Personnel
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B. Thomas, Executive Vice President
C. Reed, Vice President, Nuclear Operations
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T. J. Maiman, Vice President, Projects
D. Galle, Assistant Vice President and General Manager
K. Graesser, Division Vice President
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D. J. Scott, Operations Manager, NSD
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D. Farrar. Director, Nuclear Licensing
W. Shewski, Quality Assurance Manager
B. M. Saunders, Nuclear Security Administrator
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- S. C. Hunsader, Nuclear Licensing Administrator
Braidwood Personnel
- M. J. Wallace, Project Manager
- E. E. Fitzpatrick.. Station Manager
- C. W. Schroeder, Station Services Superintendent
- K. L. Kofron, Production Superintendent
- D. L. Shamblin, Assistant Project Manager
- M.' E. Lohmann, Project Startup Superintendent
- W. E. Vahle, Project Construction Superintendent
C. J. Tomashek, Project Startup Superintendent
G. E. Groth, Project Field Engineering Manager
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- D. E. O'Brien, Assistant Superintendent - Operations
- L. E. Davis, Assistant Superintendent - Technical Services
- D. E. Paquette Assistant Superintendent - Maintenance
P. Cretens, Assistant Superintendent - Work Planning and Startup
- R. D. Kyrouac, Station Quality Assurance Superintendent
- E. L. Martin, Station Quality Assurance Superintendent
- P. L. Barnes, Regulatory Assurance Supervisor
R. M. Preston, Quality First Director
T. F. Hallaren, Administrative Services Director
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- M. Takaki, Regulatory Assurance - Licensing Group
- L. W. Raney, Supervisor Nuclear Safety Group
- E. Wendorf, PCD Electrical Supervisor
- R. Legner, Senior Operating Engineer
R. J. Ungeran, Operating Engineer
G. Masters, Operating Engineer
R. Yungk, Operations Engineer
F. D. Willaford, Security Administrator
T. C. Meyer, Station Fire Marshall
- T. W. Simpkin, Regulatory Assurance - Operating Group
- R. C. Bedford, Regulatory Assurance - Licensing Group
- H. D. Pontius, Regulatory Assurance - Licensing Group
- P. C. Zolan, Quality Assurance - Construction
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J. F. Phelan, Project Field Engineer
- B. Peacock, Tech Staff
- K. Buzek, Tech Staff
- T. J. Lewis, Startup Staff
- S. H. Stapp, Quality Assurance - Operation
Sargent & Lundy (S&L)
- M. Bielman, Engineering Analyst
NRR Headquarters Personnel
- J. Stevens, Licensing Project Manager
- V. Benaroya, Facilities Operation Branch
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The inspectors also talked with and interviewed other licensee employees,
including members of the technical and engineering staffs, startup
engineers, reactor and auxiliary operators, shift engineers and foremen,
electrical, mechanical and instrument personnel, contract security
personnel, and construction personnel.
- Denotes those attending one or more exit interviews conducted on
January 23, 29, February 6, and 13,1987, and informally at various
times throughout the inspection period.
2.
Licensee Action on Previously Identified Items
a.
Open Items
(Closed) 456/85015-05; 457/85016-05:
In the past, calibrated tools
or measuring devices used for electrical inspections or installa-
tions were not always recorded for specific components. An example
is the electrical penetration installations identified in violation
456/83018-06F. The licensee's electrical site contractor,
L. K. Comstock, issued Nonconformance Report (NCR) 3419 to identify
deficiencies in their calibration program. Calibrated tools or
measuring devices included in the scope of NCR 3419 were:
wrenches, thermometers, dynamometers, welding machines, rod ovens,
volt amp meters, micrometers, wire crimpers, and wire strippers.
The inspector reviewed the disposition of NCR 3419, which was
closed August 16, 1986, and noted no problems. The inspector also
determined that the violation pertaining to the torquing of the
electrical penetrations was closed in Inspection Report 456/86005.
Based on the review of the disposition of NCR 3419 and the closure
of violation 456/83018-06F this item is considered closed.
(Closed) 456/85015-09; 457/85016-07:
The licensee's new Quality
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Control Inspector Reinspection Program (QCIRP) and the resolution
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of rejected items identified during original QCIRP were to be
reviewed. The original QCIRP, which was not governed by a detailed
established program, had identified numerous deficiencies during the
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reinspection to evaluate the Pittsburgh Testing Laboratory (PTL) QC
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inspectors. These original deficiencies were documented on PTL
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Nonconformance Report (NCR) 182,-Revision 3.
This NCR was closed
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on September 4, 1986. The inspector reviewed the disposition and
supporting documentation and noted no problems. The inspector,
along with an NRC Region III specialist, reviewed the revised QCIRP
which was described in; licensee's Procedure PM-11, Revision 2,
" Quality Control Inspector Reinspection." This review did not note
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any programmatic concerns. The inspector and regional specialist
reviewed the results of the QCIRP. The regional specialist reviewed
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the welding discrepancies noted during the QCIRP and found none to
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- be design significant. The resultant reductions in capacity did not-
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impact on the structural integrity of the plant in that adequate
design margins were maintained. This matter is considered closed.
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(Closed) 456/85053-03: The draft Technical Specifications did not
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identify surveillance requirements for the batteries utilized in the
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starting of the diesel driven AFW pump, IAF01PB. When this item was
identified the licensee initiated action to develop surveillance
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requirements in the plant procedures. These procedures have been
issued and define the daily, monthly, and quarterly surveillance
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requirements for these batteries.
In regards to surveillance
requirements in the Technical Specifications, it was determined by
NRR, with the inspector's concurrence, that specific surveillance
requirements for the AFW batteries were not needed in the Technical
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Specifications. However, one issue remains in regards to performing
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a specific surveillance on the AFW batteries. That issue pertains
to a need for performing a periodic discharge test on the batteries
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to evaluate their capacity.
Presently there is no regulatory
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requirement to perform a discharge test on the nickel cadmium AFW
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batteries. The licensee is evaluating the r.eed for discharge
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testing. Until such time as this' evaluation is complete, discharge
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testing of the AFW batteries will be considered an Open Item
(456/86065-01(DRP)).
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(Closed) 456/85053-04: The diesel AFW pump engine startup panel,
IAF01J, has a switch for selecting which one of the two battery
banks is utilized ter starting the diesel AFW pump. Two procedures
reviewed for lining up the AFW system for standby operation and for
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emergency operation did not adcquately address operator actuation of
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this switch to use the other battery bank in the event the diesel
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electric motor starter failed to crank,
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CECO has resolved this problem by revising the two procedures (Bw0P
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AF-7 and Bw0A ELEC-5) questioned by the NRC inspector.
Procedure
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Bw0P AF-7, Revision 52, Page 3, Item 12, approved October 9, 1986,
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and Procedure 18w0A, ELEC-5, Page 7, Item 3d, approved August 25,
1986, were each revised to prompt the operator to select the
alternate battery bank if the diesel starter motor fails to crank.
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A review of these procedure revisions indicates that the issue has
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been completely resolved; the words used to prompt the operator to
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switch from one battery bank to the other in the event of a failure
to crank are, in both cases, are clear and easily interpreted.
This
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issue is closed,
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(Closed) 456/85057-02(DRP);457/85053-01(DRP): Routine NRC safety
inspections occurring between December 1, 1985, and February 1, 1986,
disclosed a conflict between two testing procedures (BwMP 3100-037
Revision 0, and BwVS 4.2.2-1, Revision 0,. including Temporary Change
No. 044) being used to test Braidwood Unit 1 pressurizer code safety
valves. ' Procedure BwMP 3100-037 required the leak test fluid to be
nitrogen, while Procedure BwVS 4.2.2-1 required the valves to be
' leak tested using water pressurized with nitrogen. The Nuclear Work-
Request (NWR) packages clearly established that the testing of the
code safety valves should be in accordance with both Procedures BwMP
3100-037 and BwVS 4.2.2-1.
This issue was resolved by:
(1) creation of Revision 2 of Procedure BwVS 4.2.2-1 which
eliminated the conflict in leak test fluids. Nitrogen is
specified as the leak test fluid for both this procedure and
Procedure BwMP 3305-028 which replaced Procedure BwMP 3100-037.
(2) revision of Maintenance Work Request Procedure BwAP 1600-1,
Revision 10, to require the use of a two part traveler (CECO
Forms 86-5165(S),8-79,and86-5218,10-79) whenever multiple
procedures are to be used.
In addition, special training was
given to cognizant maintenance personnel to sensitize them to
the need to look for conflicts between multiple procedures and
to use the written text of the traveler package to resolve and
eliminate these conflicts.
A sample review of. CECO training records showed that a special
course was given to maintenance personnel to train them to properly
implement procedure BwAP 1600-1. A part of that training was
reported by the' Quality Control Supervisor to include the use of a
traveler to resolve questions and conflicts arising from.the use of
multiple procedures.
Based on the above, this matter is closed.
(Closed) 456/86016-03(DRP): Quality of Release to Operations (RTO)
Evaluations. Due to the magnitude of the RTO review process, the
inspector expressed concern about the assurance that this process
would be conducted in a manner consistent with that expected of
nuclear plant operations. The inspector. conducted the following
review:
Preparation for actual review of RTO packages is conducted in
accordance with the two main procedures-governing the assembly
and content of RTO packages. These were reviewed in detail.
The procedures are " Instructions for System Turnover To
Operations" (PSU-01, Revision 6) and " System Turnover for
Operation" (BwAP 1205-11, Revision 3). Based upo
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review of these documents and discussions with li.n a thorough
censee
representatives, it appears that the licensee has developed
these procedures as a result of lessons learned earlier at
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other stations.
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These procedures appear to be well thought out; however,
resulting from.this review, it is suggested that Procedure
PSU-01, Revision 6, Section 2.0 (References) refer to Procedure
BwAP 1205-11.
It is noted that Procedure PSU-01 does call out
Procedure BwAP 1205-11 on Page 4, Section 4.0, Item 9.
This
omission is not a significant problem.
For this review, it was necessary to. review the RTO packages
with these procedures in hand. Though it is not apparent at
first, the RTO packages were found to be assembled in a very
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consistent format so that a person who is familiar with this
format could quickly locate and track the sequence and
chronology of signoffs and departmental approvals for most
any system or piece of equipment within that system.
" Station Review," Section 6. for each of the following RTO
packages was reviewed first:
RC-11
Reactor Coolant Air Operated Valves and Motor
Operated Valves
AP-13
Auxiliary Power Above 480 Volts (ESF)
SI-10
Safety Injection
SI-11
Safety Injection Accumulators
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SI-12
Safety Injection - Flow Balance
The inspector conducted a systematic review and determined that
checklists had been checked off and that signatures and dates
appeared in the proper blanks of each form. The inspector also
examined each list of work that remained to be done at the time
of acceptance of the system. This review is summarized by the
table shown below.
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Page
RTO
Numbers
-Description
Package
Within
of Forms
Volume
Section
and Materials
No.
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Reviewed
RC-11
1 to 92
BwAP 1205-11T6
93 to 99
BwAP 1205-11T2
100 to 102
BwAP 1205-11T3
103
BwAP 1205-11T4
104 to 110
BwAP 1205-11T5
AP-13
1 to 54
BwAP 1205-11T6
55 to 72
BwAP 1205-11T2
73 to 74
BwAP 1205-11T3
75 to 79
BwAP 1205-11T4
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BwAP 1205-11T5
There was no 11T5 form found.
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SI-10
1 to 84
BwAP 1205-11T6
85 to 94
BwAP 1205-11T2
95 to 99
BwAP 1205-11T4
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100 to 103
BwAP 1205-11T3
SI-11
1 to 19
BwAP 1205-11T6
20 to 25
BwAP 1205-11T2.
26 to 29
BwAP 1205-11T3
30 to 35
BwAP 1205-11T4
36
BwAP 1205-11T5
SI-12
1 to
BWAP 1205-11T6
11 pages
BwAP 1205-11T2
w/o numbers
17 pages
BwAP 1205-11T3
w/o numbers
5 pages .
BwAP 1205-11T4
w/o numbers
1 page
BwAP 1205-11T5
w/o number
All questions raised were adequately answered by licensee
personnel. The next review of the same RTO packages
concentrated upon Section 8.0 which dealt with the following
five topics:
Flush Package Documentation
Initial Pump Run Documentation
Initial Compressor Run Documentation
Initial Fan Run Documentation
Balancing Reports
- Not all packages had a section on each one of the above five
topics. Some of these topics are not applicable to each RTO
package and are hence not required.
Section 8 of RC-11 covers only Topic 1 and is divided into the
three sections shown below:
Flushing procedures and Forms - pages 1-51
Flush sign-offs - pages 53-76
Flush data sheets - pages77-117
Nothing was found inconsistent or missing from reviewing these
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documents.
Section 8 of SI-10 contains only topics 1 and 2.
There is
apparently no need for a Section 8 in SI-11. The next section
reviewed was the listing of " Master Deficiencies" in Section
4.0 of RC-11.
Each of the 256 listed deficiencies is
sequentially numbered at the top of the page.
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All questions were satisfactorily answered by licensee personnel
and it appears that the RTO process was conducted in an acceptable
manner. This matter is considered closed.
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(Closed) 456/86038-04:
In the area of _ controlling loss of stored
items and materials, the licensee has established an inventory of
safety-related items that would be conducted on a two year cycle.
This appears to be a rather long period of time between planned
reviews of the inventory for determining loss of stored items and
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materials. This policy was to be reviewed by the licensee to
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determine if it was timely.
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After thoroughly reviewing the problem, the licensee sees no need
to complete the inventory cycle more often than every two years.
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Over the four year period that this two year inventory cycle of
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safety-related equipment has been in effect, no instances of loss
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of items and materials have been identified . The licensee feels
the existing system is both adequate and cost effective. Should
loss become a problem, the licensee-has agreed to again consider
the need for more frequent inventory cycles. While reviewing this
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open item, the inspector noted that:
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(1) the existing two year inventory cycle of safety-related items
and materials is not done at the end of the 24th month, but
rather on a continuous monthly basis.
Therefore, there is
the potential to detect losses on a continuous monthly basis.
(2) The seriousness of a loss of, or the immediate detection of a
loss,-is further lessened by the licensee's " Combined System
Inventory" program where they have a computer listing of spare
parts at their other neighboring power plants. These
neighboring plants have either common spare parts or have parts
that can be interchanged. The following other neighboring
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plants are able to supply stock inventory items to Braidwood
upon request in a matter of hours:
Interchangeability or
Commonality of Spare
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Plant Name
Parts Inventory
Byron
Nearly 100%
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Zion
About 60%
LaSalle
About 30%
Quad Cities
About 30%
Dresden
About 30%
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In consideration of all of the above, this item is considered
closed.
(Closed) 456/86038-06: Special electronic equipment and
instrumentation parts, such as printed circuit boards are usually
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stored in a Level "A" storage area. The licensee has these items
presently stored in a Level "B" storage area -(inside and heated).
In ' order to verify that the licensee is following the storage
-requirements for Level "A" and Level "B" storage of hardware, the
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following was done:
(1) Reviewed Procedure BwWP 800-1, Revision 1. " Levels and
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Conditions of Storage. This licensee document was reviewed
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by the inspector with particular attention to the physical
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limits'of environmental conditions for Level "A" and Level
"B" storage. This licensee document references the storage
requirements of ANSI 45.2.2-(1972),-the licensee's Quality
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Assurance Manual (QP 13-51 and QP 10-54) and Regulatory Guide
1.38, Revision 2.
No problems were noted,
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(2) Reviewed licensee documents which showed that the technical
staff has, where needed and requested, reviewed storage
requirements for components and parts to determine if Level
"A" or Level "B" storage is required. The licensee provided a
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' series of documents and notes dating back to November 23, 1983,
illustrating that they have in the past reviewed the storage.
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requirements for components or parts in accordance with
Procedure BwWP 800-1, Revision 1.
As of February 3, 1987, no
hardware was in the Level "A" storage area, but documents were
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reviewed showing that two hardware items currently on order
will be stored in the Level "A" area upon receipt. These items
were both printed circuit boards for security systems with
controlled environment shelf lives of 40 years.
Five other
documents -also describing items requiring level "A" storage,
but which are not currently stored, were also reviewed. No
problems were noted.in this review.
(3) -An inspection of the Braidwood service building Level "A"
storagearea(locatedwithintheQAvault)andtheLevel"B"-
storage area (located within the storeroom on the third floor)
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was made to see that the requirements of Procedure BwWP 800-1,
Revision 1, were being met.
The Level "A" temperature and
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humidity, recorders were inspected and found to be in operation
and were determined to be affixed with up to date calibration
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stickers. Because the Level "A" storage room contained no
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hardware at the time of the inspection, there was no require-
ment to log (on a weekly basis) the temperature and humidity
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readings on Form BwWP 800-1A2 that is part of Procedure BwWP
800-1, Revision 1.
No problems were noted in this review.
It has been determined that licensee personnel are continuing to
maintain and use separate storage areas for Levels A, B, C, and D
in accordance with ANSI Standard 45.2.2; CECO QA Manual, QP 13-51
and QP 10-54; and Regulatory Guide 1.38, Revision 2.
For equipment
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or materials where the storage level is questioned, the licensee
technical staff performs an evaluation, on a case by case basis,
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to determine the proper storage level.
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- Based on the above information, this matter is considered closed.
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(Closed) 456/86038-07:
10-CFR 50.59 specifies that annually the
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licensee shall submit to the NRC, a report containing a brief
description of any changes, tests, or experiments not described in
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the FSAR. The licensee's Technical Specification, Section 6.9.1.4
and site administrative procedures'did not include the requirement
for this annual report. Braidwood Unit 1 received it's license
to load fuel on October 17, 1986. The inspector reviewed the
licensee's General Surveillance Computer Program and noted that the
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first annual report is identified to be due October 17, 1987. This
annual reporting requirement is listed in Nuclear Stations Division
Directive.(NSDD) A01, Revision 3, dated June.2, 1986. This
directive is also cross referenced in the General Surveillance
Computer Program. This item is considered closed.
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(Closed) 456/86038-09: A routine safety inspection conducted during
August and September of 1986 concentrated upon Section 6.5.1 of the
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Braidwood Technical Specification pertaining to offsite review and
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investigation. Minor concerns were identified; however, licensee
personnel were able to correct all but one concern prior to the
exit meeting at the conclusion of the inspection. This exception
was the revision control process for the procedure manual used by
the offsite review and investigative function.- The specific concern
was that the revision number and date did not appear on each page
of the Department of Nuclear Safety Organization and Administrative
Manual.
Pending review by the NRC of the adopted review control
process, this was an open item.
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In response to this open item, the licensee has changed the revision
control process for.the procedure manual used for offsite review and
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investigation. A February 5, 1987 review of Revision 0 (dated
January 1, 1987), of the Department of Nuclear Safety Organization
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and Administration Manual showed that all five pages did have the
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revision number and date typed in the upper right corner of each
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page. This item is closed.
b.
Unresolved Items
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(Closed) 456/83016-02(DRP): While examining equipment in Braidwood
Unit 1 Containment, the NRC inspector noted that all four containment
coolers had damaged cooling fins'. The damage apparently resulted
from placement of construction scaffolding within the coolers.
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Though the fins on the cooling coils were observably damaged, it
was not possible to determine by observation whether or not related
damage had occurred to the reactor containment fan coolers (RCFC)
service water cooling coils until operational leak testing of the
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cooling coil pressure boundaries had been completed.
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This problem has been resolved by performing the steps listed below:
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(1) Construction scaffolding was removed and the work area was
cleaned. This occurred around September 18, 1984.
(2) Cooling coil nozzles were repaired and the damaged cooling fins
were repaired by straightening (combing). After repairs were
complete, temporary plywood protective covers were installed to
protect the cooling coils and fins from future damage. These
actions were completed by December 11, 1985, and are documented
by Phillips Getschow Company (PGCo) Field Change Orders (FCOs)
IVP-8013, IVP-8014, IVP-7424, and IVP-7425.
(3) Leak testing of the cooling coil pressure boundaries was
completed and no cooling coil leakage was detected. PGCo
Pressure Test Reports, No. 20,000 and No. 8335, plus 40
separate RCFC cooling coil drawings, each with signed and
dated pressure test completion and acceptance stamps,
document completion of this task; all leak testing was
complete by September 5, 1986.
A January 28, 1987 field inspection of all units except, IVP01CA,
showed the cooling units to be clean and serviceable. The access
door to unit IVP01CA was bolted shut and the unit was operating,
making inspection impossible. Based upon review of the documents
cited above and a satisfactory field inspection of the damaged
units, this issue is closed.
(0 pen)-456/85004-01(DRP);457/85004-01(DRP): Water Presence in
Containment Tendons. The licensee examined all tendons and those
with water were recorded for re-examination. By letter dated
February 2,1987, Sargent & Lundy recommended to Commonwealth Edison
that the examination be conducted with the Unit 1 containment
inservice inspection. The letter also suggested that an additional
horizontal tendon be sampled in addition to the normal sample of one
dome, one vertical, and one horizontal tendon.
If possible, the
samples will be from tendons that have shown a presence of water.
This matter will remain open for followup on the ongoing tests and
inspections.
(Closed) 456/85032-01; 457/85031-01: The inspector identified
deficiencies in the auxiliary building and requested that the
licensee evaluate the welded and bolted plate connections for
specification conformance and safety significance. The licensee
stated that nonconformance reports (NCR) mcy already exist for the
beam-angle plate installations. The licensee reviewed the specific
areas identified by the inspector in the unresolved item to address
both the welded and bolted structural connections. As a result of
this review, the structural steel site contractor, G.K. Newberg
(GKN), issued GKN NCR Number 213-1362 to identify concerns with six
,
connections.
These connections were dispositioned "Use-As-Is" for
e
four connections and " Rework" for the other two connections. Prior
'
to the issuance of this unresolved item, GKN was in the process of
mapping structural bolting and welding inspections for specification
11
.
._~
_ . _ _ .
_ _ _ ___ _ _ _ . _ _ _
._ ._
_
.-
,
..
conformance. .Therefore,'any other similar type concerns have been
identified-through this mapping process and resolved utilizing
established nonconformance procedures. While performing this
mapping process, GKN. identified that certain inspection records
could not be located. The licensee issued NCR 838 on September 2,
1986 to identify that-certain structural steel connection inspections
-were missing. During the processing of this_NCR, the licensee
identified the following causes on NCR 838 for the missing records:
E
(1)' The welded connections installed by Mid City were inspected by
Mid City with a 10% overview by Pittsburgh Testing Laboratory
(PTL). The Mid City documentation turnover may not have
included all of these inspections.
(2) The PTL field inspector documented the connections by
identifying the connections inspected on an inspection drawing.
His final inspection reports did not, in all cases, positively
identify all the connections that were inspected.
+
(3) During the September 1979 shutdown of-the Braidwood site, it!
was noted that some of the inspection drawings were misplaced.
,
These drawings were recreated from-final inspection reports.
-l
All the attributes _ inspected may not have been recreated from
the final inspection reports.
i
i
(4) Since the beams were not originally numbered by Sargent & Lundy
(S&L), each structural steel fabricator. numbered the beams.
Therefore, identical beams within a division could have the
!
same mark number.
PTL used the mark number to track their
,
I
-bolting inspections. This duplication of mark numbers within a
i
division could cause confusion as to which connection had been
i
inspected.
Based on the above, the licensee concluded that although
,
- .
documentation of inspections could not be located, it was not as a
i
result of a lack of inspections and acceptance of the work
performed. The disposition of-NCR 838 was to accept "use-as-is"
i
these connections which were missing inspection records. This
disposition was reviewed by S&L in which they concurred with the
licensee's "use-as-is" recommendation. The results of S&L's review
,
I
was documented in a letter dated October 2, 1986. The basis of
j
S&L's concurrence for accepting the 1985 connections with missing
{
inspection documentation was that 54 had been reinspected under the
Quality Control Structural Steel Review (QCSSR) and 19 were
4
reinspected under the Braidwood Construction Assessment Program
(BCAP). S&L noted that none of the discrepancies identified under
the reinspection programs were of a nature or frequency to warrant
any further corrective action. S&L further stated that the
structural steel installation at Braidwood Station was found to be
.
of adequate quality. To determine the effectiveness of S&L's review
of the reinspection results noted above, an NRC Region III
specialist reviewed the 19 connections reinspected under BCAP. The
j
results reviewed were documented in the following S&L calculations:
1
4
12
1
. ,
, ,_-.-
-.
~ - - .
_,,
.-. - -.-~~ --. ---- ...
--- - ---.
- ..-
-
.
.
CSR-I-S-002-001-1
CSR-I-S-002-033-1
CSR-I-S-002-034-2
CSR-I-S-002-034-3
CSR-I-S-002-034-4
CSR-I-S-002-034-5
-CSR-I-S-002-034-6
CSR-I-S-002-037-1
CSR-I-S-002-037-2
CSR-I-S-002-041-1
CSR-I-S-002-041-2
CSR-I-S-002-041-4
CSR-I-S-002-044-1
CSR-I-S-002-044-2
CSR-I-S-002-044-3
CSR-I-S-002-044-4
CSR-I-S-002-045-1
CSR-I-S-002-051*
CSR-I-S-002-101*
CSR-I-S-002-110-1
CSR-I-S-002-141-1
CSR-I-S-002-146-2
CSR-I-S-002-147-1
CSR-I-S-002-147-2
CSR-I-S-002-152-1
CSR-I-S-002-152-2
CSR-I-S-002-153-1
CSR-I-S-002-153-3
- No discrepancies.were identified during the reinspection.
~
The regional specialist determined that the S&L evaluations were
adequate and there were no discrepancies which would warrant further
reinspection of connections which have missing inspection
documentation. Based on the disposition of NCR 838 and the NRC
Region III Specialist's review of S&L calculations, this matter is
considered closed.
c.
Violations
(Closed) 456/85032-02:
Flexible metal hoses for instrument 1-PT-544
and 1-PT-545 had traps formed. The licensee tracked the deficiencies
with the flex hose traps on Phillips Getschow (PGCo) Nonconformance
Reports (NCR) 4388 (1-PT-544) and 5054 (1-PT-545). Also, a similar
deficiency had been identified by PGCo on the flexible hose for
1-LT-559, which was documented on PGCo NCR 5105. These NCRs have
been closed. Further corrective actions by the licensee included
the following:
For instrumentation flexible hoses (Unit 1) within the ASME
Code jurisdictional boundary, Sargent & Lundy (S&L) has
performed a walkdown under cold plant conditions. A functional
13
1
.-
,
.
review of the results of this walkdown was perfomed by S&L
and Engineering Change Notices (ECN) 27101 and 28632 were
issued to clarify flexible hose configuration acceptance.
criteria.
For instrumentation flexible hoses (Unit 1) within
the ANSI B31.1 Code jurisdictional boundary classified as
' Class H (safety-related), PGCo submitted as-built drawings to
S&L. S&L'then performed a functional review of the hoses to
determine whether reroutes were necessary. The acceptance
criteria in ECNs 27101 and 28632 were also applicable to ANSI
B31.1Property "ANSI code" (as page type) with input value "ANSI</br></br>B31.1" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., Class H, installations.
The inspector reviewed ECNs 27101 and 28632 and found the acceptance
,
criteria well defined for inspection purposes. The inspector also
'
verified that NCRs 4388, 5054, and 5105 were properly closed and
that the dispositions required the reworking of the flex hoses to
-
preclude traps.
The licensee's corrective action to avoid further violations
i
included the issuance of ECN 27101, which clarified flexible hose
configuration acceptance criteria, and developing a procedure
(PGCP-3, " Flexible Metal Hose Installation") to define actions to be
taken during hot plant conditions for collection and evaluation of
data for flexible hoses. The inspector reviewed Procedure PGCP-3,
Revision 6, and noted no problems. To determine the effectiveness
of the corrective action, the_ inspector performed a general tour of
the plant to determine if there were any instrument-flexible hoses
installed with traps. The inspector noted no problems during this
tour. Therefore, based on the implementation of the licensee's
corrective actions and the fact that the corrective action appears
effective, this item is considered closed.
(Closed) 456/85045-01(DRP);457/85044-01(DRP):
Inadequate
Corrective Actions on Housekeeping and Equipment Protection.
Between September 3 and November 1, 1985, NRC inspectors determined
that, contrary to the Commonwealth Edison Quality Assurance Manual,
corrective actions were not promptly implemented for the three
issues listed below:
(1) Correctiveactions(instructiontocontractors),inresponse
to NRC concerns over protecting systems and components from
intrusion of foreign matter, such as dust and moisture, were
insufficient and resulted in exposure of the internals of
reactor protection electrical components in the auxiliary
electrical room and control room panels to deposition of
significant dust from nearby ongoing masonry work.
(2) Numerous damaged flexible electrical cable conduits were
observed during a plant tour.
The damaged conduits exposed
the cable insulation to cuts and gashes from the sharp edges
of the conduit and also forced the cable to support the conduit.
14
_
.-
- (3) An accumulation of oil was found inside the bearing
thermocouple junction box for centrifugal charging pump
1CV01PA. This caused the thermocouple cable insulation to be
immersed in oil and resulted in potential degradation of the
insulation.
In addition, this oil represented a potential
fire hazard to the pump.
The CECO response to the above. violation is documented in a
December 13, 1985 letter to NRC Region III.
The first of the three examples described above dealt with the
' failure to protect systems and components from foreign matter,
such as dust and moisture from nearby construction activities.
The licensee acknowledged.that housekeeping and equipment
protection measures were inadequate for the specific areas
described by the violation. These areas have since been
thoroughly. cleaned and vacuumed and are now not in violation of
the CECO Quality Assurance Manual.
To reduce the probability of~ recurrence of.this issue, the licensee
instructed construction contractors, on October 19, 1985, to seal
off these types of construction arens to prevent dust and dirt from
spreading to clean areas. This order had apparently not been fully
implemented.as of October 28, 1985 when the NRC inspectors verified
that nothing had been done to correct these conditions.
Further-
more, CECO's contractor personnel have been assigned on an ongoing
basis to maintain control panel cleanliness in all areas of the
plant. Additionally, openings through the control room and
auxiliary electrical equipment room ceilings from the overhead cable
spreading rooms have been equipped with temporary seals to prevent
debris from entering the panels.
Regarding the second example where numerous damaged flexible
electric cable conduits were observed, the licensee described it's
three ongoing programs to identify and repair damaged flexible
electric conduits. These three programs covered construction.
system testing, and plant area turnover phases of the project and
each identification, repair, and reinspection included specific
steps to detect and repair electric cable damage which may have
occurred incident to flexible conduit damage. Though occasional
instances of damaged flexible conduits may still be observed. CECO
felt the existing _three programs were adequate to cover the
detection and associated repair of all such deficiencies.
-
The third example was similar to the first because it involved
unwanted foreign matter found within an electrical enclosure. More
specifically, oil was detected inside a thermocouple junction box
for charging pump ICV 01PA. The referenced letter documents that oil
found within the thermocouple junction box has since been removed
and the junction box has been restored to original condition and
returned to service as of December 12, 1985.
15
4
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4
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A =
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4
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-
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.
4
.
-
Rece'nt observ'ations in Unit-1. indicate that housekeepong is
improving with the ongoing cleaning and painting in preparation
'
for operation. Observations in Unit 2 indicate a positive trend;
however, since it is still under construction, the concern. remains
high for system and component protection. This issue is considered
closed,
i
d.
10 CFR 50.55(e) Reportable Items
(Closed) 456/85005-EE: High Energy Line Breaks in the Auxiliary
Building. This item was previously discussed in Inspection Report
g
456/86016. The licensee has completed the necessary modifications
'
to automatically isolate the steam generator blowdown (SD) and
'
auxiliary steam system (AS) in event of a pipe rupture. The
licensee's proposed modification was reviewed by the Office of
,
'
Nuclear Reactor Regulation (NRR). This review was documented by
NRR in Safety Evaluations pertaining to the system aspects and the-
.
.
electrical aspects of the design. The results of NRR review were
'
!.
documented in a February 6, 1987 letter to the licensee. NRR
'
concluded that the SD and AS systems for protection.against
i
postulated piping failures in the auxiliary building were acceptable
and met the requirements of General Design Criterion (GVC) 4.
The
inspector verified these modifications were installed and the
'
appropriate testing was conducted. The inspector also verified that
procedures BwAR 0-31-C8, BwAR 1-15-E11, and BwAR 2-15-E11 relative
to this issue were developed, reviewed, and issued. This matter is
!
considered closed.
!
e.
Licensee Action on IE Bulletins
(Closed) 456/85001-BB; 457/85001-BB: Steam Binding of Auxiliary
1
Feedwater Pumps. Compliance for the Braidwood plant is based
upon following the lead .taken earlier by the Byron plant. This
,
review meets the requirements of Temporary Instruction TI2515/69.
The licensee's Architect-Engineering Firm, Sargent & Lundy (S&L),
reviewed IE Bulletin 85-01 relative to the Byron and Braidwood
L
plants. The S&L conclusion was that even though steam binding of
'
the AFW pumps is very unlikely because of the mechanical designs
used in these two plants, modification of operating procedures, as
,
!
suggested by the NRC, is prudent because it cannot be guaranteed
!
that steam binding will not occur. Based upon this recommendation,
the licensee's Project Engineering Department recommended that
Braidwood Station implement an operating procedure similar to Byron
!
Operating Abnormal Procedure (now B0A SEC-7, Revision 51) and
i
include a temperature check of the AFW pumps at the AF005 valves.
1
Isolation and cooldown of a loop to preclude steam binding of the
AFW pumps is initiated by either or all of the following:
(1)
If, as a result of a once per shift temperature check of valve
AF005, the temperature is observed to be greater than 130 F,
}
then Procedure Bw0A SEC-7, Revision 51 must be entered and
i
16
- - - , , - . -
.-
- . - -
- . . . . - , . - - - _ . . - . _ - . - - - - . - - - . -
,
- _ -
.
. - . -
- - -
.
.
- - - --. - - -.
4
.
4
>
...
followed. This procedure'provides the actions required to-
L
,
prevent steam binding of auxiliary feedwater pumps and.
1-
potential water hammer of auxiliary feedwater piping due to .
check valve leakage. The second step of this procedure
'
requires a check of series' downstream valve AF004 to determine
if it has a temperature of over 130 F.
If over 130 F, then
"
Step 3 initiates closure of the affected auxiliary feedwater
~
pump discharge isolation valves.
(2)
If steam vapor is released upon opening auxiliary feedwater-
,
piping high point vent (s), then Procedure Bw0A SEC-7, Revision
51 is entered and followed as described above.
.
(3) Braidwood Operating Procedure Bw0P AF-2, Revision 51 addresses
I
shutdown of the auxiliary feedwater system after its operation
to check and record the AFW pump discharge temperatures 30
minutes after the system has been secured.
If these
'
temperatures are determined to have increased since the AFW
was secured, the Shift Supervisor must be notified since the
increased temperatures are indicative of steam generator check
valve back leakage which could cause steam binding of the AFW
pumps. The Shift Supervisor will then opt to enter Procedure
Bw0A SEC-7, Revision 51 if the temperature rise is determined.
,
This item is closed.
1
(Closed) 456/86003-8B; 457/86003-88:
Potential Failure of Multiple
ECCS Pumps Due to Single Failure of Air-0perated Valve in Minimum
'
,
j
Flow Recirculation Line. By letter, dated November 14, 1986, the
licensee stated that a survey was conducted and documented by memo,
dated October 18, 1986. The safety injection system was identified
i
to have a common pump recirculation motor operated valve (ISI 8813).
This valve is required by technical specifications in Modes 1, 2,
i
and 3, to be de-energized, open, and it's position verified once per
i
shift. This bulletin is considered closed.
1
.
3.
Allegations
(Closed) RIII-86-A-0167: On October 2, 1986, the Senior Resident
1
'
Inspector (SRI) (0perations) received a concern from an individual
pertaining to the control of electrical cable. The individual stated
j
that millions of feet of electrical cable were missing and the concern
was related to prudency. The individual wrote memos to L.K. Comstock
!
(LKC) management and had not seen any corrective action. The individual
i
stated that the cable was missing from warehouses and none of the
installed cable had been removed from the plant.
The .indiv.idual further
'-
stated that this concern was identified to the licensee's Construction
Superintendent, Project Manager, and Quality First. Also, the individual
>
stated that the memos were stopped on the desk of an L.K. Comstock (LKC)
,
employee. On October 16, 1986, the individual again contacted the Senior
'
l
Resident Inspector (Operations).
In summary, the individual stated that
there were no safety problems with Braidwood construction; however, the
i
f
i
17
1
{.
.
- - - , - - - _ .
____,__.___,_.__,-,._.__..,.m._,,.._._..___
_ _ . _ _ . _.._,_._ ~.
-
-
- .
_
.
._
_ _ _ _ _
_-
.
r
-individual'had reservations.about the Quality First team. By letter,
'
dated October 21, 1986, the NRC Region III office requested the
,
licensee to conduct a review of the individual's concerns utilizing
an organization which is independent of the Construction. Superintendent,
LKC Project Manager, and Quality First. The results of this review
would then be reviewed by an inspector during a future onsite inspection.
Subsequently, the individual contacted the SRI-(Operations) on
c.
November 14, 1986 to discuss the NRC October 21, 1986 letter to the
licensee requesting a review of the individual's concern. The individual
reiterated that electrical cable was missing from storage and that the
individuci was concerned with Quality First. The individual also stated
that the Braidwood plant was built safely.
.
In a November 24, 1986 letter to Region III, the individual clarified
previous information.. The following clarifications warranted further
.
!
review by the NRC. These clarifications were:
someone was trying to cover up the report of missing cable by laying
the individual off for lack of work,
the entire package of memoranda and computer reports, which the
,
individual had given to the licensee's Project Manager, was stopped
at a LKC Manager's desk,
,
the storage requirements of cable onsite, which has nothing to do
'
with a safety concern, but under 10 CFR 50 are a valid NRC concern,
the indiv.idual never questioned, and does not now question, anything
regarding the safety or quality of the construction of Braidwood.
i
The concern was about missing cable.
4
'The results of the licensee's investigation of the individual's concerns,
-
I
were reviewed by the Senior Resident Inspector (Construction). The
licensee's investigation consisted of contacting.the individual by phone
on October 19, 30, and November 3, 1986. The licensee's investigation
was documented in a memo to the Project Manager and summarized as
follows:
There is sufficient quality control during cable installation to
assure that cable damaged during storage or damage due to improper
'
storage would be identified prior to installation.
1
.
If there were errors in reel records, traceability would not be
'
affected since traceability is established by Quality Control by
utilizing pull cards.
4
Currently existing cable inventories were deemed sufficiently
accurate to identify future cable needs based on Unit 2 completion.
The individual had no concern with Quality First or the Project
Construction Superintendent as stated during the phone conversations
of October 19 and 30, 1986.
!
l
18
_ _ .-_ _ - . _
.
. .
. .
. - . .
.
. _ _ _ _ _ . - _ . _ _ , _
.
.
The Senior Resident Inspector (SRI) (Construction) reviewed the
clarifications documented in the individual's November 24, 1986 letter to
the NRC Region III office and the results of the licensee's
investigation. The results of the review were as follows:
Individual's November 24, 1986 letter:
The individual believes that someone was trying to cover up the
a.
report of missing cable by terminating the individual for lack
of work.
Since the concern of missing cable is not a safety or quality
related matter, the individual's termination was not subject to
investigation by the inspector. The reason missing cable from
storage is not a safety or quality concern is explained in
Paragraph (d).
b.
The entire package of memos and computer reports, which were
given to the licensee's Project Construction Superintendent was
stopped at an LKC Manager's desk.
The inspector reviewed a memo, dated March 3, 1986, from LKC's
Project Administrator to distribution within LKC. This memo
was distributed to LKC's Project Management and Engineering
organizations. This memo requested that these organizations
review a handwritten memo received by the licensee's Project
Construction Superintendent pertaining to recommendations in
tracking cable inventory. This memo appeared to be written by
the alleger. A handwritten note in the margin, written by the
licensee's Project Construction Superintendent, requested LKC
management to review the handwritten memo's recommendation
which pertained to cable inventory. Based on the review of
this documentation the inspector could not substantiate this
portion of the allegation.
The storage requirements of cable onsite, which has nothing to
c.
do with a safety concern, but under 10 CFR 50 are a valid
Concern.
The inspector determined that the storage requirements for
cable reels is "outside." This type of storage is known as
" Level D."
Level D storage is an acceptable storage level for
cable reels.
In addition, it is a requirement for Quality
Control inspectors to inspect cable for damage prior to
installation.
Based on this inspection requirement and the
fact that the cable reels are allowed to be stored outdoors,
the inspector has no concerns in the method of storage of cable
reels at Braidwood.
In regards to manufacturer supplied cable,
the inspector does not have a concern with storage since this
type of cable is shipped and stored in either a drum or on
pallets.
A small quantity of cable was furnished by equipment
manufacturers.
This type of cable is also required to be
inspected for damage by Quality Control prior to installation.
19
I
.
.
d.
The individual never questioned, and does not now question,
anything regarding the safety or quality of the construction of
Braidwood. The concern was about missing cable.
It appears from the November 24, 1986 clarification that the
individual was aware that missing cable is not a safety or
quality concern. The inspector concurs with the individual for
the following reasons:
(1) All cable is procured " safety-related." Therefore, any
missing cable turning up later in construction phase and
installed would still have the appropriate test reports
substantiating its quality.
(2) The storage requirements for cable reels are those defined
for " Level D" storage. There are four levels of storage;
A, B, C, and D.
Level A has the most stringent storage
requirements (inside with temperature and humidity
control) and Level D has the least requirement (outside).
Therefore, if cable reels were missing from the assigned
storage area (Level D) the cable reels would then be
stored in a storage area that was equivalent (Level D) or
better (Level A, 8, or C).
Summary of Licensee's Investigation
.
a.
There are quality control inspections during cable installation
to assure that cable damaged during storage would be identified
prior to installation.
The inspector concurs with the licensee's conclusion. LKC
procedures require that cable be inspected for damage prior to
installation.
b.
If there were errors in cable reel records, traceability could
not be affected since traceability is established by Quality
Control by utilizing pull cards.
The inspector concurs with the licensee's conclusion. Cable
pull cards, which are required for every safety-related cable
pull, identify the source of the cable. Also, since all cable
, .is procured " safety-related," a receipt inspection of the cable
is performed. This receipt inspection includes a verification
that there are test records traceable to the cable reel being
received.
c.
Currently existing cable inventories were deemed sufficiently
accurate to identify future cable needs based on Unit 2
completion.
The inspector did not review this portion of the licensee's
investigation since the inventory of cable is not a quality or
safety concern.
20
.
.
.
d.
The individual had no concern with Quality First as stated
during a telephone conversation of October 19 and 30, 1986.
The inspector reviewed the documentation in the licensee's
Quality First file generated for the individual's concern.
Since missing cable is not a safety or quality concern, Quality
First actions of having the licensee's Project Construction
Department investigate the individual's concern was deemed
appropriate by the inspector.
e.
The individual's concern about the storage for cable reels is
not a safety concern since the acceptable storage for cable
reels is outside (Level D).
The inspector concurs with the licensee's conclusion. See
Paragraph (C) under the previous section of this report titled
" Individual's November 24, 1986 letter."
Based on the results of reviews documented above, the inspector concludes
that there are no safety concerns associated with this allegation and the
actions taken by the licensee appeared appropriate. Therefore, this
allegation is considered closed.
(Closed)RIII-86-A-0199: On December 22, 1986, the Senior Resident
Inspector (SRI) Operations was contacted by two Quality Control (QC)
inspectors from General Electric-MCIS with a concern that due to a
labor / management dispute, their inspections were being impaired. These
two QC inspectors were performing inspections for the site electrical
contractor, L.K. Comstock (LKC). These QC inspectors stated that the
labor / management dispute pertained to the overtime and absenteeism
policies. As a result of being frustrated or upset with the labor /
management dispute, the QC inspectors stated that one of them tore
up quality documents and the other did not properly carry out a rejcction
form for a motor operated valve (MOV). On December 23, 1986, the SRI
(Operations) had contact with the licensee's Construction Superintendent.
During this contact the SRI (Operations) informed the Construction
Superintendent of the potential concern expressed by the two QC
inspectors. The Construction Superintendent stated that he had been
unaware of any safety-related implications associated with the
labor / management dispute. The Construction Superintendent had previously
contacted the Resident Inspector (Operations) on December 22, 1986 to
brief the NRC on the labor / management dispute between the QC inspectors
and General Electric-MCIS.
Subsequent to the telephone contact with the SRI (Operations) on
December 23, 1986, the Construction Superintendent directed the LKC QC
Manager to meet with the inspectors. The purpose of this meeting was
to inform the inspectors that if any of them felt the current labor /
management situation was affecting their ability to do quality work
they should identify themselves. This meeting was convened in the
afternoon on December 23, 1986. The inspectors were instructed that if
1
they did not want to disclose themselves at the meeting that it could be
done at their convenience in a more private manner.
This meeting was
21
_
_
..
.
documented in a memo to file from the LKC QC Manager. This memo stated
there was no response from any individual either after the meeting or
in-private.
To determine if the labor / management dispute was affecting the work
performance of other QC inspectors the SRI (Construction) interviewed
nine QC inspectors and one document reviewer. .The following is a list
of the interviewee's certifications and experience:
Individual
Experience
Certifications
- 1
1 year at Braidwood
Level II in Weldinc
,
Configuration, anc
Concrete
Anchors
4 years at Clinton
in weld inspection
and Quality
- 2
2 years QC at
Level II in Welding,
Braidwood
Configuration Termination,
and Cable Pulling
6 years at Byron
,
- 3
31/2 years QC at
Level II in Weldinc
,
Braidwood
Configuration, anc
Receipt
Inspection
- 4
3 years QC at
Level II in Recei)t
j
Braidwood
Inspection, Cali3 ration,
Termination, and Cable
Pulling
- 5
Document Reviewer
,
- 6
3 years 10 months
Level II in Cable Pulling,
!
QC at Braidwood
Conduit, and Cable
Tennination
- 7
1 year QC at
Level II in Termination
l
Braidwood
Previous experience
l
at Millstone III
,
- 8
3 years QC at
Level II in Cable Pulling
Braidwood
and Conduit
- 9
1 year 10 months
Level II Welding and
at Braidwood
Configuration
..# 10
2 years at Braidwood
Level II Conduit, Welding
and Configuration
'
22
.
.
The following are the questions asked each individual and a summary of
their answers:
a.
Was each individual aware of the meeting convened on the afternoon
of December 23, 1986 by the LKC QC Manager?
All the individuals stated they were aware of this meeting and noted
no one coming forward with a concern that the labor / management was
affecting the quality of their work.
b.
Did they have any knowledge of any ins)ector's work being impaired
due to the labor / management dispute?
- or example, were they aware
of any inspector destroying quality documents or failing to identify
deficiencies on the appropriate nonconforming documents?
None of the inspectors heard of or knew of any inspectors that had
their inspections impaired due to the labor / management dispute.
Also, none of the inspectors interviewed knew of any inspector who
destroyed quality documents or failed to identify nonconformances,
c.
At the present time, are there any inspectors who are still upset to
a point where their inspections are being impaired?
All of the inspectors stated they had no knowledge or knew of any
inspector whose inspections were being impaired. They had no
knowledge that any inspectors were presently destroying quality
documents or failing to document deficiencies.
In addition to interviewing randomly selected QC inspectors, the SRI
(Construction) interviewed the two QC inspectors who contacted the
SRI (Operations) on December 22, 1986. The following is a summary
of the interviews:
Individual "A"
This individual stated that he had torn up four draft Inspection
Correction Reports (ICRs) as a result of being frustrated with
the work policies pertaining to overtime and absenteeism.
Specifically, tearing up these draft ICRs was a result of
learning that an individual had been terminated for missing one
hour of work due to illness. This individual stated that the
ICRs that were torn up had not yet been given an ICR number
(logged). These draft ICRs were written as a result of a
partial Area Turnover Inspection he had performed, but had
later been replaced by another inspector. The area he had
inspected was reinspected by the other inspector. The
individual stated the four draft ICRs he had written were given
to the inspector who was assigned to complete the area turnover
inspection. The individual stated that the four draft ICRs he
had written identified the following deficiencies:
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Two ICRs for segregation labels.
One ICR for torquing on a strap.
- The SRI (Construction) interviewed the QC inspector who had
been assigned to finish the individual's area turnover
inspection. This individual stated that he had been given
the four ICRs from individual "A" and that these ICRs were
rewritten and issued during his area turnover inspection. The
SRI (Construction) reviewed a computer printout of all the ICRs
issued during the area turnover inspection and determined that
the deficiencies identified by individual "A" were adequately
addressed. The individual further stated that the tearing up
of the four draft ICRs was the only time he did anything of
that nature.
Individual "B"
This individual stated that he had been frustrated with the
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labor / management dispute. When an individual was terminated
for missing one hour of work, he became very frustrated. This
frustration caused him to not write an ICR for a deficiency
that he noted in a MOV during a December 19, 1986 inspection.
The deficiency was a nick in a wire.
The individual stated
that after the December 23, 1986 meeting with the LKC QC
Manager, he initiated an ICR for the nicked wire. He could not
remember the ICR Number, but believed it was written for a MOV
associated with a Unit 2 essential service water valve.
The SRI (Construction) reviewed the ICR log at LKC and could
not find that an ICR was issued for a nicked wire in a MOV in
the time frame consistent with facts described by individual
"B".
The SRI (Construction) contacted individual "B" by
telephone on February 10 and 11, 1987 to discuss this anomaly.
Individual "B" stated he would review the LKC ICR log and
determine which ICR was issued by him to document the nicked
wire in the M0V. On February 10, 1987, individual "B"
contacted the Resident Inspector office. He gave the ICR
number in question to the SRI (Construction). The SRI
(Construction) contacted the LKC Assistant QC Manager and
asked him what was the subject of the ICR and when was it
issued. The LKC Assistant QC Manager stated that the ICR was
dated on December 19, 1986, and it pertained to a damaged
conductor and a loose flex connection with no identification.
This ICR pertained to valve ICV 084 located in the auxiliary
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building on the 364' elevation. The individual stated his
failure to document the deficiency for the MOV was the only
time he had acted in that manner.
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Based on the interviews of other QC inspectors, the SRI (Construction)
could not substantiate that the labor management dispute between the
QC inspectors and GE-MCIS impaired the inspection process.
The SRI
(Construction) evaluated the actions of the two QC inspectors pertaining
to the tearing up of ICRs and the failure to identify a deficiency in
the MOV. The SRI (Construction) confirmed the following:
a.
The ICRs that were torn up had not yet been logged.
In fact, these
" draft" ICRs were given to an inspector who had been reassigned to
the Area Turnover Inspection. The deficiencies noted on the " draft"
ICRs were subsequently addressed by ICRs issued during the Area
Turnover Inspection,
b.
The deficiency in the MOV was later identified on an ICR by the QC
inspector who failed to initially identify it on an ICR. This QC
inspector issued the ICR after the LKC QC Manager's meeting of
December 23, 1986.
The SRI (Construction) reviewed this ICR and
determined that this ICR addressed a nicked wire in an M0V. The SRI
(Construction) contacted the individual and asked him why the ICR
was dated December 19, 1986, when he earlier had stated that he
wrote the ICR after a December 23, 1986 meeting. The individual
stated that he performed the inspection on the M0V on December 19,
1986. Therefore, when he wrote the ICR on December 23, 1986, he
,
dated the ICR with the date of the inspection, December 19, 1986.
Based on the review of this ICR's content, the SRI (Construction)
has determined that this ICR was written for the deficiency in the
M0V, which had not been previously identified.
This allegation is closed.
No violations or deviations were identified.
4.
Regional Request
By memo, dated January 30, 1987, Region III, Division of Reactor
Projects, requested a summary of the licensee's actions on planned
action in response to IE Information Notice 86-106, "Feedwater Line
Break," relative to the catastrophic failure at Surry Power Station.
The questions asked and responses were:
a.
Question
For PWRs and DWRs, identify which licensees have and which
licensees have not prepe. red programs to determine whether their
large-diameter steam, feedwater, condensate, and connected system
piping is subject to thinning of the piping wall.
Response
Presently there are two procedures in place to test for thinning
in cross-under steam lines and MSR reheat lines:
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'BwVS 210-1, Rev. O, " Ultrasonic Thickness Measurement of
Extraction Steam Piping"
BwVS 210-2, Rev. O, " Examination of Cross-under Piping"
These procedures will be carried out during the refueling outages.
Prior to the Surry event, and in addition to the above procedures,
the licensee was in the process of developing a list of potential
pipe erosion points in the feedwater, heater drains, extraction
steam, and the steam generator blowdown piping. Since the Surry
event, the licensee has added consideration of main steam,
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condensate booster, and auxiliary feedwater piping.
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b.
Question
Piping for those systems is subject to wall thinnirg if the
following conditions exist:
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(1) Carbon steel fittings and spools.
(2) Fittings less than 10 pipe diameters apart.
(3) Bulk flow greater than 10 feet per second.
(4) Fluid temperature between 195'F and 440*F.
(5) 0xygen concentration less than 600 ppb.
Determine whether licensee's have included these factors in their
programs.
If other factors have been included, identify them.
Response
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The conditions for considering wall thinning are all of those in
the memo and in addition, the licensee considers high moisture
steam and 90* bends.
. c.
Question
Determine how many measurements will be taken and where they will
be taken.
Response
Presently specific locations for examinations are stated in the
two procedures listed in item above.
BwVS 250-1 identifies four extraction steam pipes with three
components on each pipe. The types of components are cold
reheat extraction tees, 90' elbows, reducers, and a reducer /
tee.
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BwVS 250-2' identified 25 locations for internal visual and UT -
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examinations of piping between the HP. turbine and the MSR,
and consists of 90' bends, diaphrages, 90* bend / ext, and
extractions.
Prior to the Surry event, the licasses wat.considering.three
additionalfeedwaterpipes.Ageheaterdrains,oneextraction
steam line, the steam genWatd
blowdown lines. These vary from
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one to eight locations for examination on each line.
Commonwealth Edison has developed a corporate (SNED) task force to
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review this issue.
Dresden, Quad, and Zion are being appropriately
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reviewed first due to their age. LaSalle, Byron, and Braidwood will
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follow.
The licensee is using at least the following as a guide line:
EPRI, NP3944, Project 2231-2, Final Report April 1985
" Erosion / Corrosion in Nuclear Plant Steam Piping: Causes
and Inspection Program Guidelines"
INP0 list using Industry Procedures July 10, 1986. An INPO
.
Good Practice is expected in the future.
Westinghouse - guidelines,
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d.
Question
If measurements have been taken, provide the results.
Response
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No measurements have been taken at Braidwood.
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It is recognized that Braidwood is a preoperational PWR and pipe wall
thinning has not had the urgency associated with older plants.
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This information was forwarded to Region III on February 2,1987, and
will be further forwarded to NRC Headquarters for a " National Survey
on the Licensee's Actions Regarding Thinning of Secondary Piping."
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No violations or deviations were identified.
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S.
Licensee Event Reports (LER) Review
Through direct observations, discussions with licensee personnel, and
review of records, the following event report was reviewed to determine
that reportability requirements were fulfilled, immediate corrective
action was accomplished, and corrective action to prevent recurrence had
been accomplished in accordance with technical specifications:
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(Closed) 456/86002-00-L: All Action Requirements of Tech Specs
Not Met While Both Source Range Channels Were Inoperable.
Upon
discovery of missed valve position verification, two SRO's were
dispatched to verify the position. Long tenn followup will be to
have all associated personnel review this LER.
(Closed)456/86003-00-L:
Inoperable Hard Rubber Fire Hoses Due to
Couplings Separating from the Hoses.
Several of the hoses failed
the hydrostatic test. The licensee is replacing the hoses with
properly certified hoses. The personnel involved are being trained
to prevent recurrence.
(Closed) 456/86004-00-L:
Steam Generator Low-Low Level Results in
Reactor Trip Due to Misadjusted Valve. The trip signal resulted
from secondary leakage through leaking blowdown valves while the
steam generator was in wet layup recirculation. A foreman walked
down the system for verification and a locking device was placed on
the throttle valve.
(Closed) 456/86005-00-L:
Containment Purge Isolation From a Voltage
Transient Caused by Construction Activity. The licensee has
provided instruction to construction personnel working in sensitive
areas.
(Closed) 456/86006-00-L:
Failure to Follow Action Statement
3.3.3.10 on Failure of 1RT-PR028. The licensee provided training
to operating department personnel on this subject.
(Closed) 456/86007-00-L: Temporary Procedure Change Not Reviewed
Within 14 Days Due to Unclear Steps in Procedure. The licensee has
issued a temporary procedure change to BwAP 1300-3, " Preparation
and Approval of Temporary Procedures and' Temporary Changes to the
Permanent Procedures," and has initiated a permanent procedure
change to follow.
(Closed) 456/86008-00-L:
Failure to Comply With Technical
Specification Action Statement Due to Cognitive Personnel Errors.
The individuals involved in this event were counselled and the event
will be covered in routine training of other individuals.
(Closed) 456/86009-00-L: Diesel Generators Declared Inoperable Due
to Incorrect Design of Emergency Mode Circuitry. This was
identified to the licensee through the 10 CFR 21 reporting system.
A temporary modification was put in place to bypass the unqualified
switches.
In addition, a permanent modification will be processed
to bypass the electrical over-speed trip on an emergency start
signal.
(Closed) 456/86010-00-L: Reactor Trip Caused by Damage to Source
Range Cable Due to Personnel Error.
This reactor trip was caused
when a contractor working in an instrument cabinet bumped an
electric cable against a support strut causing the cable to be
nicked and shorted. The cable has been repaired and the sharp edge
on the strut removed.
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(Closed) 456/86011-00 and 01-L:
Containment Ventilation Isolation
Due to Spurious Signal.
Although it could not be. proven, the
licensee suspects this was caused by construction activities in
the vicinity of the radiation monitoring cabinet.
A valid isolation
signal could not be identified and all related valves were in closed
positions at the time of the signal.
A letter was issued to the
contractor management personnel re-emphasizing the need for greater
awareness during this phase of the plant life.
(Closed) 456/87001-00-L:
Failure to Take Required Gas Sample
With the H2 and 02 Analyzers Inoperable Due to Personnel Error.
The licensee installed a moisture separator filter in the
hydrogen-oxygen gas sampler line, recalibrated the analyzers and
held discussions with operating personnel.
In addition, the event
significance was small in that no hydrogen was in the Volume Control
Tank (VCT) at the time of the event.
(Closed) 456/87002-00-L:
Containment Ventilation Isolation Train A
Due to Loss of Power to a Radiation Monitor.
ESF Bus 141 breaker
tripped which resulted in loss of power to the area radiation
monitor 1RE-AR011.
Loss of power to the radiation monitor initiates
the containment ventilation isolation.
The breaker trip probably
was caused by contractor personnel working in area.
A letter
addressing the issue was written to contract personnel concerning
work activity problems.
Licensee failed to report this item within
the required time specified by 10 CFR 50.72.
This violation
(456/86065-02(DRP))isaddressedinParagraph6.
(Closed) 456/87003-00-L:
Containment Ventilation Isolation Signal
Due to Loss of Pulses from 1RE-AR012.
The licensee replaced the
detector which appears to have corrected the problem.
(Closed) 456/87004-00-L:
Train A Containment Ventilation Isolation
- Trainee Actuated.
During a training walk through, the trainee
tripped the 480 volt feed breaker to MCC 131X2.
The tripped breaker
interrupted power to 1RT-AR011, which initiated a Train A
Containment Isolation. The personnel involved were briefed to
prevent recurrence.
The preceding LERs have been reviewed against the criteria of 10 CFR 2,
Appendix C, and when the incidents described meet all of the following
requirements, no Notice of Violation is normally issued for that item.
a.
The event was identified by the licensee.
b.
The event was an incident that, according to the current enforcement
policy, met the criteria for Severity levels IV or V violations,
c.
The event was appropriately reported.
The event was or will be corrected (including) measures to prevent
d.
recurrence within a reasonable amount of time .
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The event was not a violation that could have been prevented by the
e.
licensee's corrective actions for a previous violation.
In addition to the foregoing, the inspector reviewed all of the
licensee's Deviation Reports (DVRs) generated during the inspection
period. This was done in an effort to monitor the conditions related
to plant or personnel performance, potential trends, etc.
It was also
reviewed for assurance that DVRs were generated appropriately and
dispositioned in a manner consistent with the applicable procedures
and the QA manual.
No violations or deviations were identified.
6.
Onsite Followup of Events Occurring During the Inspection
On two separate occasions, the licensee failed to promptly notify the
NRC Operations Center via the Emergency Notification System (ENS)
within the required time limit specified by 10 CFR 50.72(b)(2)(ii),
10 CFR 50.72(b)(2)(ii) requires an ENS notification within four hours
of the occurrence of any event or condition that results in manual or
automatic of any Engineered Safety Feature (ESF) including the Reactor
Protection System (RPS). Details of the two occasions in which the ENS
notifications were not within the required time limit are as follows:
On January 29, 1987, at approximately 2245, operations personnel
de-energized the Nuclear Instrumentation (NI) panel which resulted
in receipt of several RPS actuations. The NIs were de-energized to
preclude damaging the NI circuits due to a water leaking into the
control room from the upper cable spreading room. The water source
originated from painting activities which resulted in water backing
up into an area directly above the control room which drained thru
floor penetrations into the control room due to floor drainage
system blockage. The RPS actuation was not reported until 1611 on
January 30, 1987, which is beyond the four hours reporting require-
ments. The other occurrence happened on January 7, 1987 at 1127
at which time an ESF actuation (containment ventilation isolation)
occurred. The ESF actuation was due to a loss of power to the
1RE-AR011 radiation monitor which resulted from ESF Bus 141
de-energization. The actuation was not reported until approximately
1600 on January 7, 1987. These failures to make the required ENS
notifications within the required time limit is considered a
violation of 10 CFR 50.72 (b)(2)(ii) (456/86065-02(DRP)).
The Region III Branch Chief responsible for Braidwood and the resident
inspectors met with Braidwood management on February 5, 1987 to discuss
the immediate notification requirements of 10 CFR 50.72(b)(2)(ii).
It
was emphasized that NRC expects that the makeup of RPS or ESF initiation
logics be promptly reported even if reactor trip breakers did not actuate
because the breakers were already open, and even if the ESF equipment
did not actually start.
It was also emphasized that for the ESF or RPS
actuation to be part of a preplanned sequence, the sequence of events
resulting in the actuation must be preplanned from beginning to end.
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Loss of Component Cooling
On January 21, 1986, at 1745, the licensee experienced a complete loss
of Unit I component cooling (CC). The event occurred during a "B" train
outage and resulted from attempting to drain the CC side of the "B"
residual heat removal (RHR) heat exchanger (Hx) and the inlet. valve
leaking sufficiently to cause the event. The sequence of events led to a
lowering of the CC surge tank level to cause the "A" CC pump to trip and
the "B" CC pump to start and trip resulting in the loss of CC. At the
time, the plant was in cold shutdown (Mode 5) and the licensee was
conducting work activities that would lead to higher modes and initial
criticality. The inspectors have performed a preliminary review of the
licer.see's followup investigation and preliminary corrective actions and
found them to be acceptable.
This matter will be given a final review when the Licensee Event Report
is submitted to the NRC.
7.
Operational Safety Verification
The inspectors conducted routine plant tours during the inspection
period to make an independent assessment of equipment conditions, plant
conditions, construction activities, security, fire protection, general /
personnel safety, housekeeping, and adherence to applicable regulatory
requirements. During the tours, the inspectors reviewed various logs,
daily orders, interviewed personnel, attended shift briefings and plan
of the day meetings, witnessed various construction work activities, and
independently determined equipment status.
During the shift changes,
the inspector observed operator and shift engiaeer turnovers and panel
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walkdowns.
Throughout the entire inspection period, Unit I remained in Mode 5 (cold
shutdown). The inspectors verified that all applicable requirements for
Unit I were met during this period which included periodic checks of the
locked valves for boron dilution prevention as required by the licensee.
These reviews and observations were conducted to verify that facility
operations were in conformance with the requirements established under
technical specifications, 10 CFR, and administrative procedures.
While on a routine tour of the plant the inspectors discovered that the
mini flow manual isolation valve for the IB charging pump was not secure.
This valve is one that has been designated to be locked open.
The
condition found was that the handwheel nut was missing which allowed
removal of the handwheel with the lock and chain attached. This left
the valve in an unlocked condition.
In addition, the handwheel nut for
the 1A mini flow valve for the charging pump was loose, such that the
handwheel could be removed rendering the valve into an unlocked
condition. Subsequent to the above inspection activity the inspectors
identified two five protection (FP) header isolation valves that were of
concern. On one of the valves, the chain and lock was placed such that
at a distance it appeared locked, but closer inspection (moving of the
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tee) showed it was unlocked. The other valve had the chain and lock'
attached to the tee bar. The tee.bar was not attached to the valve,
which allowed removal of the tee bar rendering the valve unlocked.
This is considered an unresolved item (456/86065-03(DRP)).
,
No violations or deviations were identified.
f
8.
Unit 2 Plant Tour
The inspector observed work activities in progress, completed work and
plant conditions during general inspections in Unit 2 work areas.
Observation of work included cable trays, junction boxes, pipe support
welding and mechanical equipment. Particular. attention was given to
material identification, nonconforming material identification and
housekeeping. The inspector reviewed work activities by reviewing
travelers while touring the plant. These travelers pertained to
electrical, piping, and hanger installation.
No violations or deviations were identified.
9.
Report Review
During the inspection period, the inspector reviewed the licensee's
Monthly Operating Reports for November and December 1986. The inspector
confirmed that the information provided met the requirements of Technical Specification 6.9.1.8 and Regulatory Guide 1.16.
In addition, the
inspector' reviewed the licensee initiated changes to the Offsite Dose
Calculation Manual (ODCM) submitted as Attachments 1 and 2 to the December
report that was provided in accordance with Technical Specification 6.14.
The inspector also reviewed the licensee's Monthly Plant Status Reports
for October, November, and December 1986.
No violations or deviations were identified.
10. Meetings, Training, and Other Activities
Plant Status Meeting
A meeting was held on December 16, 1986 and January 23, 1987, between the
CECO Project Manager, the Region III Project Director, and members of
each of their staffs. The purpose of the meetings was for the licensee
to provide an update on the status of Units 1 and 2.
The meetings were
also an opportunity to discuss the licensee's list of items that must be
dispositioned prior to a mode change in Unit 1.
11. Open Items
Open items are matters which have been discussed with the licensee, which
will be reviewed by the inspector and which involve some action on the
part of the NRC or licensee or both. An open item disclosed during the
inspection is discussed in Paragraph 2(a).
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12. Unresolved Items
Unresolved items are matters about which more information is required
in order to ascertain whether they are acceptable items, violations,
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or deviations. An unresolved item disclosed during the inspection is
discussed in Paragraph 6.
13. Exit Interview
The inspector met with licensee and contractor representatives denoted
in Paragraph I during and at the conclusion of the inspection on
February 13, 1987. The inspector summarized the scope and results of
the inspection and discussed the likely content of this inspection
report. The licensee acknowledged the information and did not indicate
that any of the information disclosed during the inspection could be
considered proprietary in nature.
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