ML20132E116
| ML20132E116 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 09/24/1985 |
| From: | Mizuno G NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | Bloch P, Grossman H, Jordan W Atomic Safety and Licensing Board Panel |
| References | |
| CON-#385-622 OL, NUDOCS 8509300395 | |
| Download: ML20132E116 (28) | |
See also: IR 05000445/1984045
Text
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D. C. 20555
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September 24, 1985
'85 SEP 26 R2:39
Peter B. Bloch, Esq., Chairman
Dr. Kenneth A. McCollom
$CNE AN b
Administrative Judge
Administrative Judge
ERMCH
Atomic Safety and Licensing Board
Dean, Division of Engineering,
U.S. Nuclear Regulatory Commission
Architecture and Technology
Washington, DC 20555
Oklahoma State University
Stillwater, OK 74078
Herbert Grossman, Alternate Chairman
Elizabeth B. Johnson
Administrative Judge
Administrative Judge
Atomic Safety and Licensing Board
Oak Ridge National Laboratory
U.S. Nuclear Regulatory Commission
P.O. Box X, Building 3500
Washington, DC 20555
Oak Ridge, TN 37830
Dr. Walter H. Jordan
Administrative Judge
881 W. Outer Drive
Oak Ridge, TN 37830
In the Matter of
Texas Utilities Generating Electric, et al.
(Comanche Peak Steam Electric Station, Units T and 2)
Docket Nos. 50-445 and 50-4460t-
Dear Administrative Judges:
By letter of March 6,1985, NRC Staff counsel transmitted copies of
NRC Inspection Report 84-45 (March 5, 1985) to the Board. Subsequently,
the Staff determined that the discussions for items 6.b.(3), 6.c. and
d. were incorrect or omitted in pages 14-16 of the inspection report.
The most significant item was the inadvertent omission of item 6.c.,
which is a discussion on Deviation 445/8415-01. Accordingly, the Staff
has issued new pages 14-17 to replace original pages 14-16. See Enclo-
sure 1.
Copies of Inspection Report 84-45 which include new pages 14-17
are enclosed for the information of the Board as Enclosure 2.
Hence,
the original version of Inspection Report 84-45 should be discarded.
Sincerely,
8509300395 850924
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ADOCK050004j5
p, - _
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Gea y S
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Counsel for NRC Staff
Enclosures:
As stated
cc w/encis.:
Service List
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ENCLOSURE 1
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_ ENCLOSURE 1
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In Reply Refer To:
SEP 18 25
Dockets:
50-445/84-45
Texas Utilities Generating Company
ATTN: Mr. W. G. Counsil
Executive Vice President
400 North Olive, L.B. 81
Dallas, Texas 75201
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Gentlemen:
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This refers to the NRC Inspection Report 50-445/84-45.
Enclosed are new
pages 14-17 to provide corrections to the report details, Section 6, Applicant
Action on Previous Inspection Findings (items 6.a.(3), b.c., and 6.d).
Please
replace the original pages 14-16 with new pages 14-17.
Should you have any questions, please contact us.
Sincerely,
,
Oricinal signed Byi
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incne.rd P. Denise
R. P. Denise, Director
Division of Reactor Safety
and Projects
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Enclosure:
As stated
cc:
Texas Utilities Electric Company
ATTN:
J. W. Beck, Manager,
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Licensing
Skyway Tower
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400 North Olive Street
Lock Box 81
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Dallas, Texas
75201
Texas Radiation Control Program Director
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b.
(Closed)OpenItem 445/8415-02:
Minor discrepancies found during NRC
inspection of station administrative procedures.
During a previous
inspection (445/84-15) of station administrative procedures, the RR1
found several minor discrepancies and made some suggestions to
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preclude future problems.
The applicant took action on those items
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that the applicant considered justification for implementing a
procedure change.
F,or example:
(1) STA-401, " Station Operation Review Committee," Revision 5,
Section 4.4 did.not fully implement the responsibilities of the
committee as stated in the CPSES Unit 1 Technical Specifications
(final draft).
This was corrected in Revision 8 of STA-401.
.(2) STA-203, " Control of Station Manuals," Revision 7. Section 4.3.3
required a notification memo to be sent to each onsite holder of
controlled station manuals to alert recipients of a revision or
new procedure.
This was not being done for holders of the
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manual who incorporate their own changes because they sign a
receipt for the changes or new procedures anyway.
Revision-9
clarified this such that the applicant is in compliance with the
procedure.
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(3) STA-307, " Forms Control," Revision 2, allowed minor changes to
forms without revising the parent procedure containing a sample
of the form as an attachment.
However, instead of changing the
revision number of the form itself, the office services staff
misinterpreted Section 4.2.6 of STA-307 and changed the revision
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of the parent procedure attachment page, which caused a conflict
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with the rest of the parent procedure pages.
This was corrected
by the applicant and STA-307 was revised to preclude
misinterpretation.
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This item is closed.
c.
(Clos.ed) Deviation 445/8415-01:
Failure of the applicant to use 50RC
approved instructions to perform work on the emergency diesel
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generators.
The CPSES FSAR commits to Regulatory Guide (RG) 1.22,
Revision 2, February 1978 with no exceptions. RG 1.33 and ANSI
N18.7-1976Property "ANSI code" (as page type) with input value "ANSI</br></br>N18.7-1976" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process. to which it refers, requires maintenance to be performed
using procedures / instructions receiving the same review and approval
as operating instructions, i.e., review and approval by the SORC.
During two previous inspections (50-445/84-07 and 50-445/84-15), the
NRC inspectors noted that the applicant had defined " instructions" as
procedures wnich do not require SORC approval, and had issued
" instructions" to perform work on safety-related equipment such as
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the emergency diesel generator (EDG). The apparent basis was that
EDG work performed by the maintenance department had no significant
impact on other departments, and/or was work unique to the
maintenance department.
Since the above NRC inspections, the issue
has been resolved as evidenced in Station Administrative Procedure
STA-707, " Safety Evaluations," (Revision 2) STA-202, " Preparation,
Review, Approval, and Revision of Station Procedures" (Revision 10),
and the final draft'of the CPSES Unit 1 Technical Specifications
(TS).
In. essence., all. safety-related procedures and instructions
will receive.a SORC review by virtue of the requirement that the SORC
review the related safety evaluations, as stated in the TS and
STA-401, which both list the responsibilities of the 50RC.
This deviation is closed.
d.
(Closed) Violation 445/8421-02:
Failure of preoperational test
procedures to provide adequate prerequisites.
During a previous
inspection (445/84-21), the RRI noted that during conduct of
preoperational test ICP-PT-29-02,RT1, " Diesel Generator (DG) Control
Circuit Functional and Start Test" the DG barring device was
connected to a portable air comprassor instead of the Service Air
System. There was no' prerequisite step in the test procedure to
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provide either temporary or permanent air for the barring device, yet
it needed air to be tested.
Also, during testing of the Service
Water System in accordance with ICP-PT-04-01, RT 1, " Station Service
Water (SSW)," a Barton D/P gage did not function due to air binding.
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There was no prerequisite in the test procedure to ensure the gage
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was recently filled and vented to assure accurate test data, nor did
the Startup Administrative Procedures for writing the test require
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it.
This wa:: e notentially generic problem. The applicant has since
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revised CP-SAP-7, " Format and Content of Test Instruction / Procedures"
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to require the appropriate prerequisites.
Each organization
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responsible for review of preoperational test procedures has been
instructed to ensure that test prerequisites receive a comprehensive
review to ensure system readiness and correct component configuration
to assure validity of the test results.
This item is closed.
7.
Plant Tours
During this reporting period, the SRRI and RRI conducted several
inspection tours of Unit 1.
In addition to the general housekeeping
activities and general cleanliness of the facility, specific attention was
given to areas where safety-related equipment was installed and where
activities were in progress involving safety-related equipment.
These
areas were inspected to ensure that:
Work in progress was being accomplished using approved procedures.
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Special precautions for protection of equipment were implemented, and
additional cleanliness requirements were being adhered to for
maintenance, flushing, and welding activities.
Installed safety-related equipment and components were being
protected and maintained to prevent damage and deterioration.
Also during these tours,'the SRRI and RRI reviewed the control room and
shift supervisors' log books.
Key items in the log review were:
plant status
changes in plant status
tests in progress
documentation of problems which arise during operating shifts
No deviations or violations were found.
S.
plant Status as of December 31, 1984
a.
The applicant was at the end of the Thermal Expansion Test sequence
and making preparations to roll the main turbine-generator.
Details
of the testing sequence and problems encountered are discussed in
paragraph 2 of this report.
b.
Unit No. 1 is 99% complete with 403 of 422 areas and 323 and 332
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subsystems turned over to operations custody.
" Custody" means having
immediate authority and responsibility for operational control of
system or equipment.
The applicant has accepted 260 of 332 subsystems for final
acceptance.
c.
Of the 199 preoperational tests, one is not yet completed on field
testing, and 21 are pending review and approval of completed data.
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Eighteen are pending NRC completed data inspections,
d.
The following items related to NRC resident operations office
findings are open pending applicant action and NRC followup
inspection to confirm completion of closure:
Violations
10
Deviations
0
Open items
100
Unresolved
7
Total
117
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Action is underway to complete these items.
Closure will be
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documented in future inspection reports.
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e.
Unit No. 2 is 65% complete. The preoperational test program on
systems associated with NRC.. inspections has not yet started.
9.
Exit laterview
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An exit interv.iew was conducted January 4, 1985, with applicant
represeatatives identified in paragraph 1.
During this interview, the RRI
and Mr. D. M. Hunnicutt of the Region IV NRC office reviewed the scope and
discussed the inspection findings. The applicant acknowledged the
findings.
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ENCLOSURE 2
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ENCLOSURE 2
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Texas Utilities Electric Company
kR 2 8 %
In Reply Refer To:
Docket:
50-445/84-45
Texas Utilities Electric Company
ATTN:
M. D. Spence, President ,TUGC0
Skyway Tower
400 North Olive Street
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Lock Box 81
Dallas, Texas
75201
Gentlemen:
This refers to the inspection conducted by Messrs. D. L. Kelley and W. F. Smith
of this office during the period November 1 through December 31, 1984, of
activities authorized by NRC Construction Permit CPPR-126 for the Comanche Peak
Facility, Unit 1, and to the discussion of our findings with Messrs. B. R.
Clements and J. C. Kuykendall and other members of your staff at the conclusion
of the inspection.
Areas examined during the inspection included:
(1)witnessingofthethermal
expansion test conducted in November and December 1984, (2) review of initial
startup test procedures (3) verification of completion of human engineering
deficiencies (4) Review of completed preoperational test data (5) applicant
actions on previous inspection findings (6) plant tours, and (7) plant status.
Within these areas, the inspection consisted of selective examination of
procedures and representative records, interviews with personnel, and
observations by the inspectors.
These findings are documented in the enclosed
inspection report.
During this inspection, it was found that certain of your activities were in
violation of NRC requirements.
Consequently, you are required to respond to
these violations, in writing, in accordance with the provisions of
Section 2.201 of the NRC's " Rules of Practice," Part 2, Title 10, Code of
Federal Regulations.
Your response should be based on the specifics contained
in the Notice cf Violation enclosed with this letter.
This violation maybe related to findings identified by the NRC Technical
Review Team (TRT).
If the issue is considered to be similar, you may respend
to this item separately or as part of the Comanche Peak Response Team Action
Plan.
One open item is identified in paragraph 2 and one in paragraph 5 of the
enclosed inspection report, which will require closure by the NRC inspectors at
a later date once the actions are completed by the applicant and a followup
inspection has been completed.
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Texas Utilities Electric Company
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The response directed by this letter and the accompanying notice is not
subject to the clearance procedures of the Office of Management and Budget as
required by the Paperwork Reduction Act of 1980 PL 96-511.
Should you have any questions concerning this inspection, we will be pleased to
discuss them with you.
Sincerely,
" Original $igned By:
D.R. HUNTER"
Dorwin R. Hunter, Chief
Reactor Project Branch 2
Enclosure:
Appendix A - Notice of Violation
Appencix B - NRC Inspection Report
50-445/84-45
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cc w/ enclosure:
Texas Utilities Electric company
ATTN:
J. W. Beck, Manager
Licensing
Skyway Tower
400 North Olive Street
Lock Box El
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Dallas, Texas
75201
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Texas Utilities Electric Company
ATTN:
B. R. Clements, Vice President, Nuclear
Skyway Tcwer
400 North Olive Street
Lock Box 81
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Dallas, Texas
75201
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APPENDIX A
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Texas Utilities Electric Company
Docket:
50-445/84-45
Comanche Peak Steam Electric
Construction Permit: CPPR-126
Station, Unit 1
Based on the results of an NRC inspection conducted during the period of
November 1, 1984, through December 31, 1984, and in accordance with the NRC
Enforcement Policy (10 CFR Part 2 Appendix C), 49 FR 8583, dated March 8,
1984, the following violation was identified:
Failure to provide adeouate procedures appropriate to
circumstances
-10 CFR 50, Appendix "B", Criterion V requires that, " activities affecting
quality shall be prescribed by documented instructions, procedures, or
drawings, of a type appropriate to the circumstances and shall be
accomplished in accordance with these instructions, procedures, or
drawings."
Contrary to the above, an Instrument and Control (I&C) technician
received a first degree thermal burn on his forearm while attempting to
fill the reference leg on a pressurizer level detector (1-LT-0460)'during
hot plant conditions using a procedure that did not contain sufficient
detail to accomplish the task.
The I&C technician was using Instruction
No. 1C1-2007, " Filling and Venting Level Transmitters and Level Indicating
Switches (Wet Leg)" which is a generic procedure that provides general
guidelines for filling and venting level instruments.
This use of a
generic procec'ure is inappropriate for the circumstances, and appears to
have directly contributed to the technician receiving thermal burns
because he connected the low pressure fill equipment incorrectly and
manipulated the wrong valves.
This action resulted in the low pressure
fill equipment being blown off and releasing hot reactor coolant to the
containment atmosphere.
The I&C technician received thermal burns to his
arm from the hot reactor coolant.
This is a Severity Level IV Violation.
(Supplement II.E) (445/8445-02)
Pursuant to the provisions of 10 CFR 2.201. Texas Utilities Electric Company is
hereby required to submit to this office, within 30 days of the dates of this
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Notice, a written statement or explanation in reply, including: (1) the
corrective steps which have been taken and the results achieved; (2) corrective
steps which will be taken to avoid further violations; and (3) the date when
full compliance will be achieved.
Consideration may be given to extending your
response time for good cause shown.
Dated:
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APPENDIX B
U. S. NUCLEAR RE3ULATORY COMMISSION
REGION IV
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NRC Inspection Report: 50-445/84-45
Construction Permit CPPR-126
Docket: 50-445
Category: A2
Applicant:
Texas Utilities Electric Ccmpany (TVEC)
Skyway Tower
400 North Olive Street
Lock Box 81
Dallas, Texas
75201
Facility Name:
Comanche Peak Steam Electric Station (CPSES)
Unit 1
Inspection At:
Glen Rose, Texas
Inspection Conducted:
November I through December 31, 1984
Inspectors:
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D. L. Kelley, Senior Resident Reactor
'Date
Inspector (SRRI)
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(paragraphs 1, 2, 7, and 8)
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- W. - F. Smith, Resident Reactor Inspector
(Tats
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(RRI)
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(paragraphs 1, 2, 3, 4, 5, 6, 7, 8, and 9)
Approved:
k [bwad
A/5/55'
D. M. Hunnicutt, Section Chief,
Oafe
Reactor Project Section B
Inscection Summary
Insoection Conducted:
November 1 through December 31, 1984 (Report
50-445/84-45)
Areas Inspected:
Routine, unannounced inspection of (1) the Thermal Expansion
Test conducted during November and December 1984, (2) Initial Startup
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Procedures (3) corrected Human Engineering Deficiencies, (4) completed
preoperational ttst data, (5) applicant actions on previous inspection
findings, (6) plant tours, and (7) plant status.
The inspection involved 109
inspector-hours by two NRC inspectors.
Results:
Within the 7 areas inspected, one violation was identified (failure
to provide adequate procedures, paragraph 2).
In addition, two open items
exist; one in paragraph 2 and one in paragraph 5 pending applicant action.
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DETAILS
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1.
Persons Contacted
Acolicant Personnel
- B. R. Clements, Vice Pres'ident, Nuclear Operations
- J. C. Kuykendall, Manager, Nuclear Operations
- C. H. Welch, Quality Assurance Supervisor
"J. C. Smith, Quality Assurance
- R. B. Seidel, Operations Superintendent
- H. A. Lancaster, Startup Quality Assurance Specialist
- J. M. Ward, Startup Quality Assurance Specialist
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- R. E. Camp, Startup Manager
"S. M. Franks, Special Project and Technical Support Lead
R. R. Wistrand, Administration Superintendent
J. J. Allen, Operations Engineer
- R. A. Jones, Manager, Plant Operations
- J. T.
Merritt, Assistant Project General Manager
- L. G. Barnes, Operations Supervisor
- T. Gosdin, Coordinator, Public Information
D. W. Braswell, Engineering Superintendent
J. C. Zimmerman, ISU Coordinator
D. B. Allen, ISU Test Coordinator
B. J. Browning, Thermal Expansion Test Engineer
M. R. Blevins, Maintenance Superintendent
B. Taylor, I&C Supervisor
M. D. Deen, Shift Supervisor
A. W. Rosette, Operations Engineer
- Denotes those present at exit interview.
The NRC inspectors also interviewed other applicant employees during this
inspection period.
2.
Witnessina of Thermal Expansion Test
During the period of this inspection the applicant conducted a series of
pre-fuel load initial startup tests at reactor system temperatures and
pressures ranging from ambient to hot standby.
The principle test
conducted in the 1984 was Thermal Expansion Testing (and Retesting) that
was not completed during the 1983 hot functional test (HFT).
In addition,
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other retests requiring hot standby conditions were completed thereby
reducing the extent of hot plant testing that would be deferred until
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after initial fueling.
The sequence was of approximately 54 days duration,
as planned by the applicant.
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The objectives of this inspection were to establish through observations,
records rev'fews, and independent checks that the testing was conducted in
accordance with approved procedures, and to evaluate the performance of
the applicant's personnel involved in test performance.
The final NRC
inspection of test results will be conducted during a subsequent period
after the Station Operations Review Committee (SORC) has completed its
review of the data.
The performance of these objectives were accomplished
on a sampling basis.
The SRRI and RRI determined that testing appeared
to be conducted.in a careful and controlled manner, with minimal problems
as noted in paragraph 2.b below.
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a.
The following tests were conducted:
(1)
ISU-300A, " Pre-Fuel Load Initial Startup Test Sequence." The
objectives of this test were to provide an overall sequencing of
all the other tests to be conducted, to establish and maintain
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the plant conditions for testing, and to verify acceptable
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reactor coolant pump seal flow.
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(2)
ISU-008A, " Thermal Expansion".
The purpose of this test was to
verify that the ASME Code Class 1, 2, and 3 systems and other
nonsafety class systems which operate at temperatures greater
than 200*F were not restrained during heatup to normal operating
temperature or during cooldown to ambient conditions. This
procedure included verification that loads and clearance gaps
of selected piping system snubbers, spring hangers and pipe
rupture restraints were properly set for free pipe movement.
Component checks consisted of items requiring retest after the
preoperational test conducted during the 1983 HFT (ICP-PT-55-11,
" Thermal Expansion") and items which were not covered by that
test. Measurements will be taken at an initial ambient temperature,
and plateaus of 250'F, 350 F, 450 F and at normal operating
temperature.
A final set of readings was taken after cooldown
to ambient temperature.
(3)
ISU-206A, " Auxiliary Feedwater Performance" The purpose of this
test was to verify that five consecutive cold quick starts of
Turbine Driven Auxiliary Feedwater (TDAFW) pump could be
performed and that 1-LV-2383 (condensate drain valve) functioned
properly during each of these starts to drain condensate from
the steam supply lines.
This test also verified that the time delay from receiving a
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start signal until the TDAFW pump delivered rated flow at rated
pressure was less than 60 seconds.
The time delay was
determined for both trains and will be a summation of all
system delays including channel sensor and actuation logic.
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(4)
ISU-220A, " Turbine Generator Initial Synchronization and
O'verspeed Test".
The objectives during this testing sequence
were to obtain turbine baseline data and to verify the proper
operation and adjustment of the turbine generator system and its
associated auxiliary and support systems to the extent
practicable during noncritical hot plant conditions.
(5)
ISU-234A, " Main Steam Isolation Valves Operability and Response
Times." The purpose of this test was to verify that the full
stroke closure times of the Main Steam Isolation Valves and Main
Steam Isolation Bypass Valves were within the limits specified
in the Comanche Peak Steam Electric Station (CPSES) Final Safety
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Analysis Report and Technical Specifications for Comanche Peak -
Unit 1.
This test also was to demonstrate the operability of
each Main Steam Isolation Valve and each Main Steam Isolation
Bypass Valve.
(6)
ISU-282A, " Containment & Feed Water (FW) Penetration Room
Temperature Survey" With the RCS at the normal operating
temperature and pressure, the objective of this test was to
demonstrate that the various cooling systems were maintaining
temperatures at or below their design limits in the following
areas:
(a) reactor coolant pipe penetrations; (b) containment
average air temperature; (c) neutron detector wells; (d) each
steam generator compartment; (e) the pressurizer room at the
905 foot elevation; and (f) supply air to each reactor vessel
support.
(7) EGT-712A, " Reactor Coolant System Pressure Isolation Valve
Leakage Testing." This was a retest of repaired or replaced
Safety Injection System check valves which did not meet the
acceptance criteria while being tested during the original HFT
of 1983.
In addition to the specific tests above, the applicant took the
opportunity to exercise several integrated plant operation and
standard operating procedures to confirm or correct their accuracy
and adequacy. Also a few dry runs were conducted on pending initial
Startup test procedures to help minimize procedure problems after the
fuel is loaded.
b.
The applicant conducted weekly status and problem review meetings
between the NRC resident inspectors and key managers including the Manager,
Plant Operation, Engineering Superintendent, Maintenance Superintendent,
Operations Superintendent, and Operations Supervisor.
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This meeting provided an opportunity for the RRI and SRRI to assess
applic' ant management involvement in the test sequence, and to address
NRC inspection concerns and actions taken by the applicant, and to
keep abreast of management decisions which affect testing plans for
the week that followed.
The following problems were encountered:
(1) Failed Reactor Coolant Pump Motor:
During the early phases of
reactor coolant' system (RCS) fill in preparation for this
testing sequence, No. 4 reactor coolant pump motor tripped due
to arcing in the' stator. This was apparently caused by a
foreign piece of metal resembling a washer which may have
damaged the stator insula. tion.
This appears to be an unusual,
isolated occurrence provided an electrical path to ground. The
motor has an open type enclosure.
The motor was replaced and
retested.
(2) The No. 1 Residual Heat Removal (RHR) pump tripped upon starting
due to an apparent upper wear ring failure. The applicant is
reviewing the problem as to cause and will report it as required
by the regulations.
(3) There were two cases of failure to maintain adequate procedural
control of plant conditions:
ISU-008, " Thermal Expansion" did not address the required
charging / letdown path and as such the paths were selected in
accordance with the plant operating procedures. As a result,
the lineup had to be changed to accommodate the test.
Plant
temperature stability, as defined in the test, was lost.
The
only consequence was about a 4-hour delay in reestablishing
stable temperature conditions which are prerequisites to the
test.
ISU-008 also failed to address the fact that RHR cross
connection valve 8716A was to be open for the test, because the
integrated plant operating procedure used to establish
conditions required the valve to be shut. When it became
apparent that the valve should be opened, verbal
miscommunications between test and operating personnel
resulted in a second delay in establishing stable temperatures
for the expansion test.
These problems were discussed with the ISU Coordinator, as well
,
as TUEC management.
TUEC committed to ensure that all test
procedures will be checked and revised as necessary to identify
any valve or breaker positions required that are not normally
provided by the operating procedures referenced by the test
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p,rocedures.
The NRC inspectors did not observe this problem any
further and thus considered the corrective action of the
applicant to be' adequate in this area.
(4) Main Steam Isolation Valve (MSIV) cycling met the acceptance
criteria, after correcting minor mechanical problems pending
final review and. approval of the data, but the bypass valves
(MSIBV) did not.
The' applicant is evaluating and has made
informal comments to the RRI that a design change to manual
valves will be implemented.
(5) During thermal expansion ; testing, at the various temperature
plateaus, of ambient, 250*F, 350 F, 450*F, and normal operating
temperature, numerous support snubbers and restraints required
some rework and retesting.
By the end of the sequence most had
been corrected and retested, except in some cases where it was
impractical or unsafe (due to hot surfaces) to make adjustments.
The latter cases have been identified and will become part of
what now appears to be a potentially small test deferral package
for postfueling hot functional testing.
(6) During starting of the Turbine Driven Auxiliary Feedwater
(TDAFW) pump, in accordance with ISU-206A, the pump manual
discharge valve IAF-041 was discovered locked shut. This valve
was recorded by the operators as " locked open" on the valve
lineup sheet provided by the system operating procedure. The
RRI noted this in the shift supervisor's logs about one 8-hour
shift later,and cuestioned the shift supervisor in charge of the
subsequent shift whether or not a deviation report (DR) had been
initiated, as required by Administrative Procedure STA-404,
" Control of Deficiencies." The shift supervisor indicated that
he would check into it and if necessary, initiate the
appropriate reports.
He further stated that the cause appeared
to be confusion over which way to turn the valve handwheel due
to the reach rod linkage, and the valve being overhead, rather
than a violation of the system operating procedure.
IAF-041 is
an 8" rising stem overhead valve with several reach rod links
with universal joints to get the handwheel within easy access.
Tne RRI inspected the valve. Although it may be difficult to
check whether the valve is shut, the operator should not have a
problem checking whether the valve is open by looking for inward
stem movement because he can always attempt to shut the valve
and see some stem movement.
There is no apparent reason for
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this problem, other than failure on the part of the operator
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to check the valve position in a positive manner.
Discussions
between the RRI and the applicant brought out a need for the
applicant to take definitive corrective action to preclude
future valvt lineup problems and to ensure that all such
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problems are documented in a timely manner by shift supervisors
with first-hand knowledge of the problems. At the discussion it was
pointed out by the RRI that the shift supervisor appears to be
burdened with an analysis of the problem and possible corrective
actions for the purpose of deciding in what format the problem
must be reported, i.e., Deficiency Report, Nonconformance
Report, or Problem Report.
These reports are controlled by
three different administrative procedures. The applicant
indicated that action will be taken to provide the shift supervisors
with simpler reporting instructions. Tne applicant has committed
to the above corrective actions. This is an Open Item
(445/8445-01).
.
(7) Pressurizer level indicator 1-LT-0460 did not compare favorably
with the redundant level channels as RCS pressure increased to
near normal operating pressure. Troubleshooting the piping for
the detector revealed a leaking drain valve which was tightened
thereby stopping the leak, but the reference leg needed to be
filled.
Upon attempting to fill the reference leg in accordance
with a generic " basic guidelines" procedure, an Instrument and
Control (I&C) Technician connected low pressure fill equipment
incorrectly to the detector piping and then operated the wrong
instrument valves.
This action resulted in the low pressure
fill tubing being blown off and the I&C technician received
thermal burns to his arm from hot reactor coolant.
The
personnel safety and postulated radiological implications of
this type of problem after initial criticality was discussed
with the applicant's representatives. As a result of this
discussion, Deficiency Report 84-127 was written.
Instruction
No. ICI-2007, " Filling and Venting Level Transmitters and Level
Indicating Switches (Wet Leg)" is not adequate to assure proper
controls over quality and radiological safety, and using such a
procedure is in violation of 10 CFR 50, Appendix B, Criterion V.
This is a Violation (445/8445-02).
3.
Review of Initial Startup Test Procedures
During the month of October, 1984, the RRI conducted a review of test and
administrative procedures to be used in the control of the Thermal
Expansion Test and other hot plant tests.
The results are listed in NRC
Inspection Report 445/84-39.
The RRI inspected the following procedures
during November 1984 to complete the review:
ISU-206A, " Auxiliary Feedwater performance" (Revisicn 2)
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ISU-282A, " Containment & F.W. Penetration Room Temperature
Survey" (Revision 1)
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Attributes checked included assurance that:
(1) the procedures were
consistent with regulatory requirements, (2) the procedures contained the
necessary administrative controls, (3) the test objectives would be met
and properly documented, (4) adequate Quality Assurance provisions were
incorporated as committed in the FSAR, and (5) there were no major
technical or editorial errors._
No violations or deviations were identified.
4
Verification of Completion of Human Engineering Deficiencies
The Human Factors Control Room Design Review of CPSES, conducted by the
Human.Facter Engineering Branch of the NRC, identified many Human
Engineering Discrepancies (HEDs).
NRC Inspection Report 445/84-31 reported that as of August 31, 1984, all
but 23 prelicensing HEDs had been closed by the Human Factor Engineering
Branch, and that the remaining HEDs will be verified by the resident
inspectors and documented in future inspection reports.
As of December 31, 1984, 20 of the 23 remaining HEDs have been verified by
the RRI as satisfactorily completed by personal observation of the
installed hardware.
There are now 3 HEDs remaining to be closed.
The
following is a listing of the HEDs remaining to be verified:
88.
HED DESCRIPTION
Trend recorder scale differs from chart paper scale.
ACTION
Confirmatory on recorders having paper matching recorder scales (all
recorders should have paper), including Hot Shutdown Panel (HSP).
Note: HED 122 was closed with exception of " proper paper in
recorders" which will be verified as a part of this HED.
181.
HED CESCRIPTION
The nuclear instrumentation system recorder lacks a scale for
differential power.
ACTION
Confirmatory on installation of a scale for differential power.
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184. HED DESCRIPTION
Counters require calculations by the operator when displayed values
run past 60 minutes.
Other counters require the operator to convert
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displayed values by multiplication factors other than a multiple of
ten.
ACTION
Confirmatory on full scale counters replacing 0.5 scale counters on
CPS-01.
The following is a listing of the HEDs that have been completed and then
verified by the NRC resident inspector:
3.
HED DESCRIPTION
Annunciator alarms are not visually prioritized.
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ACTICN
Verified completion of annunciator prioritization.
68.
HED DESCRIPTION
No storage space has been allocated for essential material.
ACTION
l
Verified installation of portable storage unit and storage of
equipment at the HSP.
80.
HED DESCRIPTION
Pointers on "J" handle / star / handle switches contrast poorly with
handle color.
ACTION
Verified "J"
handle / star handle pointers being painted white.
93.
HED DESCRIPTION
No control coding is currently being used for:
o
Mechanical valves, pumps, b.eakers, motors, etc.
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o
Throttle valves
o
Emergency or critical controls
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ACTION
Verified installation of "T" handles on transfer switches at the HSP
(14 handles).
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106.
HED DESCRIPTION
.
Labels are missing.
ACTION
Verified labels on record.ers on CV-Oi, incore panel, and for lights
on CV-03.
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120.
HED DESCRIPTION
Sound powered jack communications are incomplete.
ACTION
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Verified storage of sound powered headset at the HSP (see no. 68
above).
122.
HED DESCRIPTION
The HSP is in the process of complete redesign.
ACTION
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Verified completion of Hierarchical labeling at HSP and transfer
panels, labeling of light box, proper paper in recorders (sea no. 88
above), and sound powered headsets at HSP (see no. 68 above), and
transfer panel.
Note:
" Proper paper in recorders" has not been completed.
This
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action was moved to HED no. 88 so that item 122 could be
closed.
130.
HED DESCRIPTION
Controls have unlabeled switch positions.
.
ACTION
Verified new escutcheon plates for 1-HS-2491 through 1-HS-2494 on
CB-09.
214.
HED DESCRIPTION
A rotary control with clockwise-counter clockwise movement is used to -
control a " lower" and " raise" function.
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ACTION
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Verified permanent escutcheon plates on CB-11 (90-1EG2 and 65-1EG2).
225.
HED DESCRIPTION
Thelockingpositionorfbnctionoftheverniercontrollersisnot'
clearly indicated.
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ACTION
Verified " LOCK" position labels on Hagan controllers.
226.
HED DESCRIPTION
Setpoint adjustment knob covers on process controllers can be easily
removed.
. ACTION
Verified more secure attachment of setpoint adjustment knob covers on
controllers.
267.
HED DESCRIPTION
Trend re: orders used frosted glass.
ACTION
Verified replacement of frosted glass with clear glass on recorders
on CB-10.
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HED DESCRIPTION
Annunciator character sizes are inconsistent.
ACTION
Verified re-engraving of annunciator tiles
1-ALB-2:
3.7
1-ALB-3B
2.6
1-ALB-4A
4.4
1-ALB-4B
1.5, 2.6, 3.6
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1-ALB-5B
2.1, 3.4
1-ALB-5C.
3.1, 4.2
1-ALB-6C
1.2, 1.3, 2.1, 2.2, 2.7, 3.2, 3.3, 3.7,
4.2
1-ALB-60
1.4, 1.10, 1.14, 2.4, 2.13, 2.14, 3.13,
3.14, 4.13
1-ALB-8
1.13, 2.13, 2.14, 3.14, 4.14
1-ALB-9
1.4, 1.8, 1.11, 4.1, 7.6
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345.
HED DESCRIPTION
Abbreviations in computer displays do not conform to those in the
Comanche Peak Steam Electric Station (CPSES) " Directory of Acronyms
and Abbreviations."
ACTION
,
Verified revision of point descriptions in P2500 to use CPSES
abbreviations.
5.
Preoperational Test Results Evaluation
The RRI reviewed the followir.g completed test package, ICP-PT-66-01,
" Nuclear Instrumentation System," which has been approved by the Joint
Test Group (JTG). Attributes inspected included: (1) adequacy of the
evaluation of test results, (2) assurance that test data met acceptance
criteria, (3) assurance that deviations were properly identified and
resolved, and (4) the applicant's administrative practices with respect to
test execution and data evaluation were adequate.
The test package met the attributes above, with one apparent exception. A
Test Procedure Deviation (TPD-03) was written to delete the requirement to
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take data in paragraphs 7.1.7.5 and 7.2.7.5 of the procedure, because the
source range meters that were installed in the Hot Shutdown Panel did not
function properly.
The meters did not have the proper signal input
ratings.
This was documented on a Test Deficiency Report (TDR 3014).
TDR-3014 stated that the retest would be per TDR-3547.
TDR-3547 was
written because when the proper meters finally were installed, they would
load down the circuit and cause erroneous readings. The retest specified
on TDR-3547 was lined out, leaving an open-ended paper trail.
The
applicant's representative has committed to take action to correct this
problem. This is an Open Item (445/8445-03).
No violations or deviations are apparent at this time.
6.
Acplicant Action on Previous Inspection Findings
a.
(Closed) Unresolved Item 445/8424-02: Apparent conflict between FSAR
Figure 6.3-5 and the Safety Injection Pump 01 performance curve.
During a previous inspection (445/84-24) of the completed test data
of preoperational test procedure ICP-PT-57-01, " Safety Injection Pump
Performance," the NRC inspector noted that the Safety Injection Pump
.
01 performance curve in the completed test data did not meet the
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minimum acceptable performance curve of Figure 6.3-5 of the FSAR.
Since the time of the inspection, Amendment 53 of the FSAR (dated
November 5, 1984) changed the pump curve. The performance data and
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curves in the completed test package for ICP-PT-57-01 now meet the
requirements of the FSAR.
This item is closed.
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b.
(Closed) Open Item 445/8415-02: Minor discrepancies found during NRC
inspection of station administrative procedures.
During a previous
inspection (445/84-15) of station administrative procedures, the RRI
found several minor discrepancies and made some suggestions to
preclude future problems.
The applicant took action on those items
that the applicant consi.dered justification for implementing a
procedure change. .For example:
(1) STA-401, " Station.0peration Review Committee," Revision 5,
Section 4.4 did:not fully implement the responsibilities of the
committee as stated in the CPSES Unit 1 Technical Specifications
(final draft).
This was corrected in Revision 8 of STA-401.
(2) STA-203, " Control of Station Manuals," Revision 7, Section 4.3.3
required a notification memo to be sent to each onsite holder of
controlled station manuals to alert recipients of a revision or
new procedure.
This was not being done for holders of the
manual who incorporate their own changes because they sign a
receipt for the changes or new procedures anyway.
Revision 9
clarified this such that the applicant is in compliance with the
procedure.
,,
(3)
STA-307, " Forms Control," Revision 2, allowed minor changes to
forms without revising the parent procedure containing a sample
of the form as an attachment.
However, instead of changing the
revision number of the form itself, the office services staff
misinterpreted Section 4.2.6 of STA-307 and changed the revision
of the parent procedure attachment page, which caused a conflict
with the rest of the parent procedure pages.
This was corrected
by the applicant and STA-307 was revised to preclude
misinterpretation.
This item is closed,
c.
(Closed) Deviation 445/8415-01:
Failure of the applicant to use SORC
approved instructions to perform work on the emergency diesel
generators.
The CPSES FSAR commits to Regulatory Guide (RG) 1.22,
Revision 2, February 1978 with no exceptions. RG 1.33 and ANSI
N18.7-1976Property "ANSI code" (as page type) with input value "ANSI</br></br>N18.7-1976" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process. to which it refers, requires maintenance to be performed
using procedures / instructions receiving the same review and approval
as operating instructions, i.e., review and approval by the 50RC.
During two previous inspections (50-445/84-07 and 50-445/84-15), the
NRC inspectors noted that the applicant had defined " instructions" as
procedures which do not require 50RC approval, and had issued
" instructions" to perform work on safety-related equipment such as
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the emergency diesel generator (EDG). The apparent basis was that
EDG work performed by the maintenance department had no significant
impact on other departments, and/or was work unique to the
maintenance department.
Since the above NRC inspections, the issue
has been resolved as evidenced in Station Administrative Procedure
STA-707, " Safety Evaluations," (Revision 2), STA-202, " Preparation,
Review, h,pproval, and Revision of Station Procedures" (Revision 10),
,
and the final draft'of the CPSES Unit 1 Technical Specifications
(TS).
In essence, all safety-related procedures and instructions
will receive a SORC review by virtue of the requirement that the SORC
review the related safety evaluations, as stated in the TS and
STA-401, which both list the responsibilities of the SORC.
This deviation is closed,
d.
(Closed) Violation 445/8421-02:
Failure of preoperational test
procedures to provide adequate prerequisites.
During a previous
inspection (445/84-21), the RRI noted that during conduct of
preoperational test 1CP-PT-29-02,RT1, " Diesel Generator (DG) Control
Circuit Functional and Start Test" the DG barring device was
connected to a portable air compressor instead of the Service Air
System.
There was no' prerequisite step in the test procedure to
'
provide either temporary or permanent air for the barring device, yet
it needed air to be tested.
Also, during testing of the Service
Water System in accordance with ICP-PT-04-01, RT 1, " Station Service
Water (SSW)," a Barton D/P gage did not function due to air binding.
There was no prerequisite in the test procedure to ensure the gage
was recently filled and vented to assure accurate test data, nor did
the Startup Administrative Procedures for writing the test require
it. This was a potentially generic problem.
The applicant has since
revised CP-SAP-7, " Format and Content of Test Instruction / Procedures"
to require the appropriate prerequisites.
Each organization
responsible for review of preoperational test procedures has been
instructed to ensure that test prerequisites receive a comprehensive
review to ensure system readiness and correct component configuration
to assure validity of the test results.
This item is closed.
7.
Plant Tours
During this reporting period, the SRRI and RRI conducted several
inspection tours of Unit 1.
In addition to the general housekeeping
activities and general cleanliness of the facility, specific attention was
given to areas where safety-related equipment was installed and where
activities were in progress involving safety-related equipment.
These
areas were inspected to ensure that:
Work in progress was being accomplished using approved procedures.
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Special precautions for protection of equipment were imolemented, and
additional cleanliness requirements were being adhered to for
maintenance, flushing, and welding activities.
Installed safety-related equipment and components were being
protected and maintained to prevent damage and deterioration.
. Also during these tours,'the SRRI and RRI reviewed the control room and
shift supervisors' log books.
Key items in the log review were:
-
plant status
changes in plant status
tests in progress
documentation of problems which arise during operating shifts
No deviations or violations were found.
8.
plant Status as of December 31, 1984
a.
The applicant was at the end of the Thermal Expansion Test sequence
and making preparations to roll the main turbine-generator.
Details
of the testing sequence and problems encountered are discussed in
paragraph 2 of this report.
b.
Unit No. 1 is 99% complete with 403 of 422 areas and 323 and 332
subsystems turned over to operations custody.
" Custody" means having
immediate authority and responsibility for operational control of
system or equipment.
The applicant has accepted 260 of 332 subsystems for final
acceptance.
c.
Of the 199 preoperational tests, one is not yet completed on field
testing, and 21 are pending review and approval of completed data.
Eighteen are pending NRC completed data inspections,
d.
Tne following items related to NRC resident operations office
findings are open pending applicant action and NRC followup
inspection to confirm completion of closure:
Violations
10
Deviations
0
Open items
100
Unresolved
7
Total
117
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Action is underway to complete these items.
Closure will be
documented in future inspection reports.
e.
Unit No. 2 is 65% complete. The preoperational test program on
systems associated with NRC inspections has not yet started.
9.
Exit Interview
.
An exit interview was conducted January 4, 1985, with applicant
representatives identified in paragraph 1.
During this interview, the RRI
and Mr. D. M. Hunnicutt of the Region IV NRC office reviewed the scope and
discussed the inspection findings.
The applicant acknowledged the
findings.
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