ML20214M120
| ML20214M120 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 05/26/1987 |
| From: | Craig Harbuck, Hunter D, Johnson W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20214M107 | List: |
| References | |
| 50-313-87-13, 50-368-87-13, NUDOCS 8706010181 | |
| Download: ML20214M120 (14) | |
See also: IR 05000313/1987013
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APPENDIX B
U. S. NUCLEAR REGULATORY COMMISSION
REGION IV
Inspection Report:
50-313/87-13
Licenses: DPR-51
50-368/87-13
Dockets: 50-313
50-368
Licensee:
Arkansas Power & Light Company
P. O. Box 551
Little Rock, Arkansas 72203
Facility Name:
Arkansas Nuclear One (ANO), Units 1 and 2
Inspection At:
AN0 Site, Russellville, Arkansas
Inspection Conducted:
April 1-30, 1987
Inspectors
lAllD
w
5/s/8'7
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T. D. Johp56n, Senior Resident Reactor
Date '
Inspecto V
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C. C. Harbuck, Resident Reactor Inspector
Date
Accompanying
Personnel:
T. O. Martin and J. M. Sharkey
Approved:
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. Hunter, Chief, Reactor Project
Date
Section B, Reactor Projects Branch
is'A8MAE8886
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Inspection Summary
Inspection Conducted April 1-30, 1987 (Re ort 50-313/87-13)
Areas Inspected: Routine, unannounced inspection including operational safety
verification, maintenance, surveillance, emergency operating procedure,
allegation followup, Safety Review Committee, and followup on previously
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identified items.
Results: Within the seven areas inspected, one violation was identified
(installation of a temporary modification without performing engineering and
unreviewed safety question evaluations, paragraph 3).
Inspection Conducted April 1-30, 1987 (Report 50-368/87-13)
Areas Inspected: Routine, unannounced inspection of operational safety
verification, maintenance, surveillance, allegation followup, and Safety Review
Connittee.
Results: Within the five areas inspected, one violation was identified
(failure to maintain a seismic support in its design configuration,
paragraph 3).
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DETAILS
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1.
Persons Contacted
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J. Levine, Executive Director of Site Nuclear Operations
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R. Ashcraft, Electrical Maintenance Supervisor
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- B. Baker, Operations Manager
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- R. Barnes, Engineering Supervisor
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D. Bennett, Mechanical Engineer
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- B. Bhardwaj, Associate Engineer
- G. Campbell,'Vice President, Nuclear Operations
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- W. Cawthon, Electrical Engineering Manager
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A. Cox, Unit 1 Operations Superintendent
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- G. Dobbs, Senior Engineer
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- W. Eaton, Engineer
- T. Enos, Nuclear Engineering and Licensing Manager
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- E. Ewing, General Manager Technical Support
B. Garrison, Operations Technical Support
- J. Grisham, Administrative Manager
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L. Gulick, Unit 2 Operations Superintendent
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D. Harrison, Mechanical Engineer
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A. Hatley, Mechanical liaintenance Supervisor
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S. Hendrix, Safety Review Conraittee Secretary
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- R. Howerton, Civil Engineering Manager
H. Hollis, Security Superintendent
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- D. Howard, Special Projects Manager
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- M. Huff, Engineer
- L. Humphrey, General llanager, Nuclear Quality
J. Jacks, Licensing Engineer
- D. James, Licensing Supervisor
H. Jones, Plant Modifications Manager
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R. Lane, Engineering Manager
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- D. Lomax, Licensing Supervisor
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- J. McWilliams, Maintenance Manager
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J. Merryfield, Mechanical Maintenance Supervisor
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B. Michalk, Mechanical Engineer
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- P. Michalk, Licensing Engineer
S. Payne, Plant Services Contractor
- M. Pendergrass, Vice President, Engineering
V. Pettus, Mechanical Maintenance Superintendent
- S. Quennoz, General Manager, Plant Operations
M. Snow, Associate Engineer
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- J. Taylor-Brown, Quality Control Superintendent
B. Thibodeaux, Electrical Engineer
J. Tucker, Fire Watch Supervisor
- C. Turk, Engineer
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C. Taylor, Operations Technical Support Supervisor
- R. Wewers, Work Control Center Manager
- Present at exit interview on April 30, 1987.
- Present at exit interview on April 1, 1987.
The NRC inspectors also contacted other plant personnel, including
operators, technicians, and administrative personnel.
2.
Followup Cn Previously Identified Items (Unit 1)
a.
Violation 313/86-01(II.A):
Failure To Assure That Design and Design
Controls Were Properly Performed.
Item II.A.I.a (Unresolved Item 313/8601-04): The consequences
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of high-energy line breaks on EFW components were not addressed.
The licensee's response did not address the impact of a main
steam line break or critical crack on the safety-related
components added.during the EFW upgrade, including valves which
are located approximately four feet closer to a main line than
was assumed in the FSAR analysis. The licensee's response also
did not include consideration for steam jet impingement on the
main steam isolation system (MSIS) instrument lines, which
could lead to a reactor trip and MSIV closure.
Inconsistencies
were noted between technical information made available by the
licensee at the time of this review and licensee submittals
dated May 23, 1986, and December 26, 1986, on this topic. This
item will remain open.
ItemII.A.1.b(UnresolvedItem 313/8601-04): The determination
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as to whether safety-related room cooling was needed when both
EFW pumps were operating was not performed.
The licensee's response to this issue stated that the design
basis for the EFW pump room per the FSAR did not require
safety-related room cooling. The licensee also noted that the
NRC staff had not required safety-grade room cooling for the EFW
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system as originally installed and that the original design
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basis calculations for the EFW pump room were still valid since
no new heat sources had been added to the room.
On that basis,
the licensee considered the existing room cooling capability to
be adequate. However, because of the upgrade of the EFW system
to be safety-related, and because of the addition of
safety-related EFW initiation and control and flow indication
per NUREG-0737, Item II.E.1.2, the licensee must confirm that
this equipment will remain functional following a loss of the
existing non-safety-related room cooling to ensure the
operability of this equipment during design basis events. This
item will remain open.
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Items II.A.2.a and II.A.2.b (Unresolved Item 313/8601-05): The
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temperature correction factor, meter inaccuracy, and increased
loads during a design basis event were not considered in
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Battery 007 sizing Calculation GE-83D-1032-01.
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Calculation GE-83D-1032-06, performed during inspection
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50-313/86-01, provided adequate consideration for these issues.
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This item is closed.
ItemII.A.2.c(UnresolvedItem 313/8601-05): The valve actuator
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sizing analysis for motor starting torque did not adequately
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consider the design requirements.
Information provided to the licensee by the valve vendor was
incorporated into the required torque calculation for
Valve CV-2870 and yielded acceptable results. This item is
closed.
II.A.2.d(UnresolvedItem 313/8601-05): The compatibility of
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the dc distribution system components with the new, larger
batteries was not determined.
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The characteristic curve for the fuses that would protect the
bus in question was reviewed and found to be consistent with the
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licensee's conclusion that the bus would be adequately
protected. This item is closed.
ItemII.A.2.e(UnresolvedItem 313/8601-05):
The protective
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relay study for Breaker A311 for the EFW pump motor failed to
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document the source of the critical parameters.
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The licensee produced acceptable documentation of the sources of
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safe stall time, starting time, and locked rotor current. This
item is closed.
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b.
Violation 313/86-01(II.8): The Licensee's Program For Test Control
Did Not Demonstrate That Components Would Perform Satisfactorily In
Service.
Item II.B.1.a (Unresolved Item 313/8601-03):
The
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post-modification testing for Battery 007 did not include
certain design requirements.
The licensee stated that an acceptable performance test was
performed by the manufacturer on this battery before
installation.
This testing is considered acceptable to ensure
that the battery will meet its specifications or manufacturer's
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rating. However, it is essential that adequate
post-modification testing be performed to provide assurance that
the battery is properly installed and capable of meeting the
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design requirements of the de system.
Certain weaknesses were
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noted in the post-modification testing that were not addressed
in the licensee's response includin
(1) lack of consideration
for the minimum temperature (60 F) g: permitted by the licensee's
procedures for battery operation; (2) lack of monitoring the
acceptable minimum voltage during the critical period (one
minute into test) identified in licensee calculations; (3) and
lack of documentation for the actual test currents used. This
item will remain open.
Item II.B.1.b (Unresolved Item 313/8601-07):
Testing of
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Battery 007 in 1984 used an improperly applied temperature
correction factor.
The results of this test were reevaluated with the temperature
correction factor correctly applied. The results of this test
remained acceptable. This item is closed.
c.
Violation 313/86-01 (II.D 1.a-d) (Unresolved Item 313/8601-06):
Failure To Assure That Instructions and Drawings Were Appropriate To
The Circumstances
This violation concerned numerous examples of piping and
instrumentationdiagram(P&ID) errors,pipedesignspecification
drawing errors, and controlled drawing drafting errors. The
licensee's corrective action for the specific discrepancies were
reviewed and found to adequately resolve this concern, except for the
portion relating to indication of valve position and locked status on
P& ids.
Depiction of erroneous information on P& ids is unacceptable
even if a drawing note indicates that the information is "for
P
information only." This item will remain open.
3.
Operational Safety Verification (Units 1 and 2)
The NRC inspectors observed control room operations, reviewed applicable
logs, and conducted discussions with control room operators. The NRC
inspectors verified the operobility of selected emergency systems,
reviewed tag-out records, and ensured that maintenance requests had been
initiated for equipment in need of maintenance.
The inspectors made spot
checks to verify that the physical security plan was being implemented.
The inspectors verified implementation of radiation protection controls
during observation of plant activities.
The NRC inspectors toured accessible areas of the units to observe plant
equipment conditions, including potential fire hazards, fluid leaks, and
excessive vibration. The inspectors also observed plant housekeeping and
cleanliness conditions during the tours.
The NRC inspectors walked down the accessible portions of the Unit 1
emergency feedwater system to verify operability.
The walkdown was
conducted using Procedure 1102.01, Attachment F; Procedure 1106.06,
Attachment A; and Drawings M-202, M-204, M-206, and M-212.
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The NRC inspector made the following observations during the walkdown.
These items were discussed with licensee representatives so that
dppropriate action Could be taken as needed.
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Valves SV-2613 and SV-2663 were not on the control room control valve
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check list portion of Procedure 1106.06, Attachment A.
Valve MS-6894 had an old brass tag labeling it as MS-6873.
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Valve MS-6896 had an old brass tag labeling it as MS-6875.
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Valve CS-19 was locked open as required by Procedure 1102.01,but
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Procedure 1106.06 did not indicate the required locked status of this
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valve.
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Procedure 1106.06, Attachment A, used three different notations for
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indicating locked status of valves.
The order of listing valves in Procedure 1106.06, Attachment A, was
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not organized by area or room.
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Procedure 1102.01, Attachment F, listed outdated valve numbers for
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Manual Valves MS-6893,
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-5, and -6.
This was known to operations
personnel and scheduled to be corrected during the next revision of
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this procedure.
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Unit 2 Level Transmitter 2LT-0727 had a new label on it indicating
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that it was Unit 1 Level Transmitter LT-4205.
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The flexible conduit for the electrical leads to Flow
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Transmitter FT-4626 was broken.
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Procedure 1106.06, Attachment A, requires checking that
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Handswitches HS-2802 and HS-2800 are " Positioned to Condensate," but
these handswitches have a spring return to neutral feature.
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Also during the walkdown on April 15, 1987, the NRC inspector observed
that a fire hose was connected to the )ipe containing Valve CS-284. This
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valve is on a spare penetration throug1 the lower wall of the seismic
condensatestoragetank(T41B). The fire hose and associated fittings
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were connected to provide a temporary source of makeup water to T418.
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From discussions with licensee personnel, the NRC inspector learned that
this connection had probably been made in December 1986. The connection
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was removed on April 16, 1987, after the NRC inspector found that there
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was no temporary modification evaluation and authorization for the
connection. Section 6.1.6 of Procedure 1000.28, " Jumper and Lifted Lead
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Control," Revision 6, required that an engineering review of temporary
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modifications be conducted to determine compliance with applicable codes,
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standards, and design requirements. Section 6.1.2 of this procedure
required that temporary modifications shall not be made until the required
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engineering review has been performed and it has been determined that the
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temporary modification will not create an unreviewed safety question.
The
licensee's failure to perform the required evaluations prior to
installation of this temporary modification is an apparent violation.
(313/8713-01)
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During a plant tour on April 9,1987, the NRC inspector observed that the
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seismic support for Valve 2CV-4921-1 was missing. Hanger Drawing H25-911
shows the design configuration of this Seismic Category I support. Upon
determination that a required seismic support was missing, the licensee
declared Boric Acid Storage Tank 2T68, Boric Acid Addition Pump 2P39B, and
Control Valves 2CV-4920-1 and 2CV-4921-1 inoperable. A report of abnormal
conditions was prepared (RAC-2-87-047) and the support was fabricated and
installed on April 10, 1987, under Job Order 732234. The licensee's
failure to maintain Seismic Support UC-9-H11 in its design configuration
is an apparent violation.
(368/8713-01)
These reviews and observations were conducted to verify that facility
operations were in conformance with the requirements established under
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Technical Specifications,10 CFR, and administrative procedures.
4.
Monthly Surveillance Observation (Units 1 and 2)
The NRC inspector observed the Technical Specification required
surveillance testing on the turbine-driven Emergency Feedwater Pump (P7A,
Procedure 1106.06), and verified that testing was performed in accordance
with adequate procedures, that test instrumentation was calibrated, that
limiting conditions for operation were met, that removal and restoration
of the affected components were accomplished, that test results conformed
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with Technical Specification and procedure requirements, that test results
were reviewed by personnel other than the individual directing the test,
and that any deficiencies identified during the testing were properly
reviewed and resolved by appropriate management personnel.
The inspector also witnessed portions of the following test activities:
Monthly check on EFIC Channel C (Job Order 731892,
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Procedure 1304.147)
Monthly functional test of Containment High Range Radiation
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Moni' ors 2RITS-8925-1 and -2 (Job Order 732170, Procedure 2304.148)
Test of excore instrumentation Channel D (Job Order 732169,
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Pracedure 2304.103).
Monthly check on reactor protection system Channel D (Job
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Order 732158, Procedure 1304.40)
Monthly test of Unit 1 Emergency) Diesel Generator No. 2
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(Procedure 1104.26, Supplement 2
No violations or deviations were identified.
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5.
Monthly Maintenance Observation (Units 1 and 2)
Station maintenance activities of safety-related systems and components
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listed below were observed / reviewed to ascertain that they were conducted
in accordance with approved procedures, Regulatory Guides, and industry
codes or standards; and in conformance with Technical Specifications.
The following items were considered during this review:
the limiting
conditions for operation were met while components or systems were removed
from service; approvals were obtained prior to initiating the work;
activities were accomplished using approved procedures and were inspected
as applicable; functional testing and/or calibrations were performed prior
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to returning components or systems to service; quality control records
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were maintained; activities were accomplished by qualified personnel;
parts and materials used were properly certified; radiological controls
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were implemented; and fire prevention controls were implemented.
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Work requests were reviewed to determine status of outstanding jobs and to
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ensure that priority is assigned to safety-related equipment maintenance
which may affect system performance.
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The following maintenance activities were observed / reviewed:
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Fabrication and installation of support for 2CV-4921-1 (Job
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Order 732234)
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Calibrationcheckson2LIS-5643and2LIS-5644(JobOrders 530984 and
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530986)
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Repair leak on 2P89A seal water line (Job Order 719350)
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Installation of Fire Damper 2104-0089 (Job Order 726345-3)
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DrilldowelholesinP36Bcasing(JobOrder 732523)
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Install mechanical seal on P36B (Job Order 732523, Procedure 1402.10)
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Calibration of pressurizer low range pressure transmitter Channel 2,
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2PT-4623-2 (Job Order 732989)
Corrective maintenance on SV-2663, the P7A steam supply bypass valve
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(Procedure 1403.02, Job Order 732718)
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The NRC inspector identified a potential generic problem with certain
applications of Target Rock solenoid actuated isolation valves while
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following up the corrective maintenance on Solenoid Valve SV-2663, a
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bypass valve in the steam admission valve header to the turbine driven EFW
pump. The valve had failed to reseat following surveillance testing
apparently due to steam cutting /crosion of the internal pilot valve disc.
Valves of similar design are used extensively in the reactor coolant
system high point vents in both units. These valves have a history of
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leakage.
If during accident mitigation these valves were opened, the
failure mechanism noted above could result in these valves being unable to
reseat. This could adversely affect the ability of the operators to
protect the core. After discussing this issue with the licensee, an
evaluation of this issue was initiated. This item will be followed up as
anopenitem(313/8713-02).
No violations or deviations were identified.
6.
Followup on Allegation 4-86-A-104 (Units 1 and 2)
It was alleged by an anonymous caller to the NRC that a certain fire watch
slept on the job and was not disciplined. At the time of this followup,
the person named in the allegation was no longer a fire watch, but had a
good record as a fire watch and was working in a janitorial position.
The
NRC inspector interviewed the person named in the allegation, the fire
watch supervisor, the plant services contractor, and several fire watches.
The supervisors stated that sleeping fire watches had been a problem in
the past, but that no recent instances had occurred. Their policy is
immediate dismissal for a fire watch caught sleeping. This policy has
been provided in writing to all fire watches. The position of fire watch
rover was established in October 1986. One rover is assigned for each
shift to make continuous rounds to check on the posted fire watches.
The NRC inspector concluded that the allegation could not be substantiated
and that the licensee and its contractor appear to have a system in place
to assure that fire watch personnel are alert and attentive to their
duties.
No violations or deviations were identified.
7.
Safety Review Conraittee (Units 1 and 2)
The NRC inspector reviewed the minutes of the Safety Review Conmittee (SRC)
meetings held from April 1986 through March 1987. The SRC normally has
regularly scheduled meetings monthly, with one-third of these meetings
held at the ANO site.
Twenty-four meetings were held during the review
period of 1 year.
During this review, the NRC inspector determined that
the SRC meeting minutes indicated compliance with the following Technical
Specification requirements:
Composition
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Alternates
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Meeting frequency
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Quorum
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Review requirements
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Records
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No violations or deviations were identified.
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9.
Emergency Operating Procedure Review (Unit 1)
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The purpose of this review was to determine whether the emergency
operating procedure (EOP) (Procedure 1202.01) was prepared in accordance
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with the Procedures Generation Package (PGP) and is adequate to control
safety-related functions in the event of system or component malfunction.
This review was a continuation of a previous inspection documented in NRC
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Inspection Reports 50-313/87-05 and 50-313/87-08.
a.
Much of the basis for this review was predicated upon approval of the
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PGP by the Office of Nuclear Reactor Regulation. Review and approval
of the ANO-1 PGP has not been completed.
The possibility exists that
some items of concern may come out of this review process, and that
further inspection related to those items will be necessary.
For
tracking purposes, a post PGP approval review of the E0P is
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designatedasanopenitem(313/8713-03).
b.
The NRC inspector verified that the E0P incorporated all of the
various topics as stated in the PGP,
c.
The NRC inspector verified the technical adequacy of the E0P by
reviewing the plant-specific technical guidelines, normally referred
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to as the AT0G (Abnormal Transient Operating Guidelines), Part 2, the
basis document, and the entire E0P, ad through detailed discussions
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with the principle licensee representative responsible for
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maintaining the E0P. Several minor items were noted during this
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review and are being considered by the licensee for incorporation
into the E0P. These items were:
The thermocouples in the incore nuclear instrumentation should
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be referred to as either "incore thermocouples" or " core exit
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thermocouples," but not both, as is presently done.
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In the " overheating" section on Page 59, a caution statement
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contains an action statement. Also on this page, Step 4.c
should specify that after regaining the subcooling margin (SCM)
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and starting a reactor coolant pump, if the SCM is then lost
again, the pump should be tripped. Additional instructions for
regaining the SCM should then be added,
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In the " inadequate core cooling" section on Page 72, Step 2 says
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to " Bump RC pumps if available." This instruction should be
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more descriptive of what is trying to be accomplished; i.e.,
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" Attempt to restore primary to secondary heat transfer through
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natural circulation by bumping RC pumps, if available."
In the " inadequate core cooling" section on Page 74, the caution
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statenent at the top of the page should include an additional
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phrase to state that RCS hot leg and vessel level indications
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are not reliable when the RCS is superheated as indicated by the
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In the " inadequate core cooling" section, the caution statement
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on Page 75 should state the minimum steam pressure needed to
operate the turbine driven emergency feedwater pump.
In the " tube rupture" section on Pages 93 and 94, instructions
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regarding gland sealing steam to the main turbine should point
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out which of the two main steam headers supplies gland sealing,
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to ensure that isolation of the affected steam generator does
not remove the gland sealing steam supply and result in a loss
of condenser vacuum inadvertently.
The term " emergency cooldown" is important in the event of a
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tube rupture, and implies that several operational limits (such
as cooldown rates, fuel compression limits, and steam generator
tube to shell differential temperatures) are either modified or
not applicable. All such limits should be listed together in
the discussion part of the " tube rupture" section.
To assist the operator in determining the appropriate section of
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the E0P to use, a listing at the beginning of each section
describing all the places in the other sections of the E0P that
directs the operator to that section would be beneficial.
No instances were identified where steps in the E0P were inconsistent
with guidance provided by the AT0G or other technical references
listed in the basis document. The basis document was reviewed and
will be discussed below. Based on this review, the NRC inspector
determined that the E0P appeared to be technically adequate and
consistent with the AT0G and the basis document.
d.
The NRC inspector compared the E0P to the plant-specific writer's
guide (P-SWG).
(The P-SWG is part of the PGP and may therefore be
subject to change as a result of the NRR review.) The NRC inspector
identified no examples in which the E0P was not consistent with the
P-SWG.
e.
The NRC inspector reviewed licensee records of the verification and
validation of the E0P for Revisions 0, 4, 5, 6, 7, and 8.
In all
cases, the process appears to have complied with program requirements
specified in the PGP. The program for verification and validation of
the E0P as described in the PGP consists of the following elements:
Verification - ensures consistency of the drait E0P revision
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with the P-SWG, accuracy of equipment designations, and includes
a comparison with previous procedures or revisions which it
replaces to provide a basis for a 10 CFR 50.59 review of the new
revision. Coments generated during this process are evaluated
and incorporated where appropriate.
Validation - consists of walkthroughs and/or simulator runs of
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selected scenarios designed to utilize the parts of the E0P
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affected by the proposed revision. Walkthroughs are
accomplished both at the site and at the simulator. An
independent evaluator monitors the walkthrough and makes
comments in response to his own observations and in response to
questions provided on an " emergency procedure validation
checklist." Additionally, the operator (s) involved also may
provide comments. Once again comments generated are evaluated
and incorporated where appropriate.
In some cases, comments
have resulted in plant modifications.
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The final draft of the revision is then prepared and is reviewed and
approved by the Plant Safety Committee (PSC).
f.
Part of this inspection of the E0P requires an evaluation of the
operator training program for the E0P. Most of this part remains to
be done and will be accomplished during a subsequent inspection
period.
g.
The NRC inspector made the following general coments to the licensee
regarding the present system for maintaining and revising the E0P.
The plant-specific writers guide, including the basis document, and
the verification and validation program are not formalized by being
formatted and controlled as regular licensee procedures are, but are
maintained as a responsibility of a few licensee individuals in the
Unit 1 operations technical support group. These documents are, in
practice, used like procedures and should be formalized to meet the
licensee's established requirements for plant procedures. This would
ensure adequate revision and copy control and that changes to these
documents receive the proper level of review.
In practice past
changes to the writers guide and the basis document were made
concurrently with a revision of the E0P itself. The review and
approval of these changes were documented on licensee Form 1000.06G,
" Emergency Procedure Change Verification and Validation Checklist."
Recently this form was redesignated 1015.11A when it was moved to
Procedure 1015.11 " Operations Department Procedure Revision Control."
This procedure requires changes to the writer's guide, the basis
document, and the AT0G to be made if required when the E0P is
changed.
But, changes to these without changing the E0P are not
covered by procedure. The NRC inspector also found Procedure 1015.11
to be unclear about how changes to the Unit 1 AT0G were to be made
with respect to the review and approval process. This item will be
reviewed further in a later inspection.
Based upon the above reviews and discussions, the NRC inspector
determined that the Unit 1 E0P is adequate to control safety-related
functions in the event of system or component malfunction.
No violations or deviations were identified.
. . . .
14
10. Exit Interview
The NRC inspectors met with Mr. E. Ewing, General Manager, Plant Support,
and other members of the AP&L staff at the end of inspection. The NRC
inspectors also met with Mr. T. G. Campbell and other members of the AP&L
staff in Little Rock on April 1, 1987.
At these meetings, the inspectors
summarized the scope of the inspection and the findings.
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