IR 05000313/1977005
| ML19320A137 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 03/10/1977 |
| From: | Dickerson M, Madsen G, Spangler R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML19320A129 | List: |
| References | |
| 50-313-77-05, 50-313-77-5, NUDOCS 8004140734 | |
| Download: ML19320A137 (14) | |
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U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT
REGION IV
IE Inspection Report No. 50-313/77-05 Docket No. 50-313 Licensee:
Arkansas Power & Light Company License No. DPR-51 Sixth & Pine Streets Pine Bluff, Arkansas 71601 Category C Facility:
Arkansas Nuclear One, Unit 1 Location:
Russellville, Arkansas
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Type of Licensee:
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Type of Inspection:
Routine, Unannounced Dates of Inspection:
February 28 - March 4, 1977 l
Dates of Previous Inspection:
February 10-11 ' 15-18, 1977 h)
Principal Inspector: h. M O
.5/ o/77
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U M. W. Dickerson, Reactor Inspector Date r/
Accompanying
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Inspector:
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R. G. Spangler, Reactor Inspector (Intern)
Date Reviewed By:
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odcov 3//0/77 v
'G. L. Madsen, Chief, Reactor Operations and Date Maclear Support Branch
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O-2-V SUMMARY'0F' FINDINGS I.
Enforcement Action A.
Violation, None identified during this inspection.
B.
Infraction Contrary to 10 CFR 50, Appendix B, Criterion V and the licensee's QC
' Procedure 1004.19 " Hold, Caution and QC Tagging Procedure," Rev. 2, Temporary Change #2, paragraph 4.1.1, Hold Cards were not verified as having been installed and were not verified as having been removed.
(DETAILS, paragraph 2.b(10))
C.
Deficiencies None identified during this inspection.
II.
Licensee's Action on Previously Identified Enforcement Matters 77-02 18/1(a) Maintenance Procedure 1405.01 - Acceptance Criteria Maintenance procedure-1405.01 has been revised to include acceptance criteria for the station battery daily surveillance checks.
This matter is considered resolved.
1B/l(b)
Computer Printouts Not Signed Computer printouts have been reviewed and signed.
Additionally, plant personnel have been instructed by letter dated February 15, 1977 to comply
with the requirements of paragraph 7.1 of procedure 1304.32.
This matter is considered resolved.
III.
Design Changes
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.Not inspected.
IV.
Unusual Occurrences-
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None identified during this inspection.
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Other Significant Findings A.. Current Findings 1.
Plant Status The' plant is shutdown for refueling and maintenance and is expected to remain shutdown until approximately March 25.
Refueling has been completed except for insertion of one new
fuel element which has been damaged during fuel handling.
This element was returned to B&W for inspection and repair and was expected to be available at the site for insertion on March 6, 1977.
The major maintenance item remaining is rework of the "A" low pressure turbine.
As a result of blading. inspection, cracks were discovered in the Christmas tree on the spindle for the third stage blading.
At the time of the inspection, the blading had been removed and preparations were in progress for machining down the spindle at this location prior to reassembly.
2.
Unresolved Items No new unresolved items were identified during this inspection.
B.
Status of Previously Reported Unresolved Items 77-02 V.A.2a Procedural Requirements Not Clear Section 5.1.34 of Maintenance Procedure 1401.08 has been modified by pennanent change #1 to remove the conflicting requirements with section 4.5.5.
However, other required changes to the procedure have not been made.
This matter remains unresolved.
IV.
Management Meeting A.
Entrance Meeting A preinspection meeting was held with Messrs J. W. Anderson and L. Alexander on February 28, 1977. Mr. Anderson was informed that the following' items would be reviewed during the inspection.
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Surveillance of a complex nature.
2.
Plant Operations.
3.
Preparations for startup from refueling.
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Survey of Diesel Generator load sequencing.
5.
Previous items of noncompliance.
6.
Previous unresolved items.
i B.
Exit Meeting At the conclusion of the. inspection on March 4,1977 an exit meeting was held with representatives of the ANO-1 plant staff.
The following individuals were present:
J. W. Anderson, Jr., Plant Superintendent B. A. Terwilliger, Supervisor Plant Operations P. Jones, Instrumentation & Control Supervisor T. M. Martin, Supervisor Plant Maintenance T. H. Cogburn, Nuclear Engineer S. J. McWilliams, Planning & Scheduling Coordinator L. Alexander, Quality Control Engineer D. R. Hamblen, Quality Control Inspector The inspectors discussed the following items:
O 1.
The noncompliance as a result of failure to follow procedures regarding Hold Card Control.
(DETAILS, paragraph 2.b(10))
2.
The resolution of previous items of noncompliance.
(Section II of the Summary of Findings)
3.
The status of the previously unresolved item regarding maintenance procedure 1401.08.
(Section V.B of the Sumary of Findings)
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DETAILS 1.
Persons Contacted The following individual, i'
.. tion to those listed under the Exit Meeting section of this repo.., were contacted during this inspection.
Arkansas Power & Light Company'(AP&L)
I. J. Butler, Assistant Plant Supervisor Maintenance R. T. Elder, Assistant Instrumentation & Control Supervisor F. B. Foster, Production Engineer P. W. Jacobs, Production Engineer J. E. Maxwell,' Shift Supervisor S. S. Strasner, Quality Control Inspector
'I 2.
Review of Plant Operations During the inspection, the conduct of plant operations was reviewed to
. determine that selected phases of facility operations conform to require-ments of the facility license and the licensee's administrative procedures.
(N, a.
The inspector reviewed Shift Logs and Operating Records and verified
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the following:
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(1) Control room log sheet entries for the period 12/9/76 to 2/28/77 were filled out, initialed and reviewed.
(2) Auxiliary log sheet entries for the period 12/9/76 to 2/28/77
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were filled out, initialed and reviewed.
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(3) ANO-1 Station Log: entries made in this log from 12/9/76 to
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3/9/77 were observed to be filled out, initialed and reviewed.
i (4) Log book reviews were being conducted by the staff.
(5) Standing orders and special orders did not conflict with the intent of Technical Specification requirements.
(6) Bypass and Jumper Log entries from 12/9/76 to 2/28/77 did not contain bypassing that conflicts with Technical Specification requirements.
(7) Trouble reports from 12/9/76 to 2/28/77 to confirm there were no violations of Technical Specification reporting or LC0 requirements.
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tXJ b.
The inspector conducted a tour of accessible areas and included the following observations:
(1) Monitoring instrumentation was recording as required.
(2) Radiation controls were properly established.
(3) Plant housekeeping conditions were adequate, t
(4) There were no significant fluid leaks.
(5) There was no excessive piping vibrations.
(However,plantwas in a refueling status.) Pipehangers/ seismic restraint settings and oil levels were satisfactory.
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(6) Selected valves were properly positioned.
(7)
Discussion with a control room operator pertaining to the reasons for selected lighted annunciators.
(8) Plant tours are conducted by the plant superintendent and the shift supervisor on duty consistent with QA program / administrative procedures.
O (9) The control room manning was in conformance with the requirements of 10 CFR 50.54(k) and the facility Technical Specifications.
(10) Equipment caution and lockout tag information was reviewed.
A review of the Hold Card Index established that between January 29, 1977 and February 28, 1977, 31 hold cards had been placed without proper verification and three had been removed without proper verification as required by Quality Control procedure 1004.19, " Hold,' Caution and QC Tagging Procedure," Rev. 2, Temporary Change #2, dated April 26, 1976.
Examples of hold cards not verified as having been placed are numbers 7860, 7862, 7863, 7864, 7865, 7867 and 7668.
Those hold cards which
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were not verified as retroyed were 7806, 7862 and 7866.
Section 4.1.1 of procedure 1004.19 requires that "An operator will
remove the equipment from service, :lign components as required, and install the card.
The person obtaining tim clearance and the operator installing the card should then document the verification of proper installation in the Hold Card Index."
It also requires that. "the person who removes the Hold Card and the person obtaining the clearance should then document the verification of proper removal of the card in the Hold Card Index." Moreover,~ Exhibit II of the procedure (Hold Card Index) states that verification must be made by the person to whom the clearance is issued.
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Section 4.1.3 states that "the sign-off by the person verifying the installation on the Hold Card Index documents the installation and verification of cards listed on a particular Hold Card sheet" and that
"the sign-off of removal and verification of removal of cards in the Hold Card Index documents the completion of the particular Hold Card Sheet."
The inspector was informed that the licensee's Quality Control
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Inspector had also noted this problem, however, at the time of the inspection no corrective action had been taken to correct the situation.
Moreover, a similar problem with regard to control of Hold Cards was called to the attention of the licensee by an internal audit conducted on December 3 and 6, 1976.
Similar findings were established during the I&E inspection of December 6-10, 1976 and are noted in I&E report
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76-17.
l As a result of findings noted previously, the licensee was informed that this was considered a nonconformance for failure to follow QC Procedure 1004.19.
The inspector had no further questions in this area.
3.
Preparations for Startup Testing and' Power Operation The object of the inspection was to verify that:
a.
Plans exist to test the systems and components which have under-gone maintenance, or were disturbed during the refueling outage.
During the oQtage Master Work Lists are maintained which indicate the work required and what organizations are involved in the work.
Additionally, work lists are maintained for each organization (Operations, Maintenance, etc.) as a cross check. These lists are updated daily and daily work lists are issued and updated accordingly as work is completed.
Separate job orders are issued for tests of each system or component.
When work is complete on a system or component, it is turned over to Operations. At this point Operations procedures are utilized for check out.
Those Design Changes which have been made during the outage will be reviewed to determine changes required to procedures.
The program was checked by the inspector for consideration of the following systems or components:
(1)
Primary Coolant (2) Nuclear Instrumentation /In-Core Monitor
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Feedwater (4) Control Rod Dr.ve (5) Emergency Safeguards Actuation System b.
Plant startup procedure requires adherence to the licensee's Technical Specifications and commitments as they pertain to startup testing and power operation prerequisites.
The following procedures were reviewed:
1102.01 Plant Preheatup and Precritical Check, Rev. 4, T/C #3, 5/26/76 1102.02 Plant Startup, Rev. 5, 8/3/76 1105.01 Nuclear Instrumentation and Reactor Protective System, Rev.1, T/C #2,10/10/76 1105.09 Control Rod System, Rev.1, T/C #1, 6/17/74 1106.16 Condensate, Feedwater, and Steam System Operation, Rev. 3,10/23/74 O
In addition, the following procedures relative to physics tests are either in preparation or review for modification:
1302.04 Power Imbalance Detector Correlation Test and Calibration 1302.05 Core Power Distribution 1302.06 Reactivity Coefficients 1302.07 Determination of All Rods Out Critical Baron Concentration 1302.08 Control Rod Reactivity Worth Measurements 1302.10 Measurement of Ejected Rod Worth 1302.13 Sequence for Physics Testing Following Refueling 1302.35 Control Rod Drive Trip Test The-inspector had no further questions in this area.
4.
Survey of Diesel Generator ~ Load' Sequencing The objective of this inspection effort is to determine the present status (from an equipment and procedural standpoint) of the licensee's electrical installation and its ability to sustain operation of the emergency D/G p)
after a loss of off-site power following a LOCA, and subsequent to a Safety 3'"
Injection Actuation System (SIAS) reset.
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a. : Introduction Certain PWR designs permit the operator to reset the Safety Injection System (SIS) signal approximately two minutes after SIS initiation due to a LOCA.- This reset feature enables the operator to assume
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control of the ECCS equipment for realignment into the long-term
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recirculation configuration'or to secure ~ ECCS equipment. if the SIS initiation signal is found to have been inadvertently actuated.
If a loss of off-site power were to occur after the SIS signal is
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reset, the logic circuit for the emergency loads would initiate an automatic sequence to energize the normal shutdown cooling loads and not the LOCA-sequenced loads because the LOCA induced SIS signal is no longer present.
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In recent reviews, it was found that because of the presence of a LOCA-induced containment isolation signal, the preliminary logic design
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Since the pumps which
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supply cooling water to the diesels are sequenced loads, this design l
would result in failure of the diesels and lack of power for other i
safety-related components. Although a preliminary review of operating facilities indicated that the existence of this design defect is not
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i likely, the NRC elected to verify that this unacceptable design feature
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s is not present in any operating facility.
b.
Scope of Inspection
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The inspector interviewed licensee representatives, reviewed the licensee's emergency procedures and condu-ted a detailed in-depth review of the licensee's electrical drawings.
The objective of this detailed review was to:
(1)
Determine whether the Engineered Safeguards Actuation System
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includes an SIS reset feature by reviewing the appropriate logic diagrams.
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If.an SIS reset feature exists, determine whether the design includes a time dealy before reset can be perfomed, and the
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magnitude of the delay.
(3) Review existing procedures which describe actions to be taken by
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the reactor operator after the occurrence of a LOCA (i.e., after SIS actuation) to determine whether they require or prohibit an SIS reset,
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If'an SIS reset action is permitted, what is the time frame allowed
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for this action?
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(5) Determine the licensee's basis for the procedural prohibition of or permiss!on for reset action identified in items (3) and (4)above.
(6) Determine the licensee's concern, if any, regarding the adverse affects on plant safety of a NRC request to institute a procedural requirement that would prohibit SIS reset within 10 minutes after an SIS actuation.
(7) Verify by review of logic diagram /P&ID that either an LOP or SIS signal will automatically start the emergency diesel generators.
(8)
Identify any other signal or combination of signals that start the emergency diesel generators.
(9)
Identify by review of logic diagrams /P&ID those conditions that initiate the D/G load sequencing associated with an LOP, with and without concurrent LOCA.
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(10) Verify by review of logic diagrams /P&ID that for both LOP alone and for LOP concurrent with a LOCA, the loads that are auto-matically sequenced onto the emergency buses include:
O-(a)
Service water systems that are required for short-term and long-term cooling of the diesel engines, and (b)
All systems required to provide combustion and t:ntilation air to the D/G's.
(11) Determine the tests (preoperational and/or surveillance tests)
that have been performed to verify those combinations of load sequencing actuation signals identified in Items (9) and (10) above.
(12) Verify by review of logic diagrams /P&ID that the load sequencing for an LOP which occurs after an SIS reset includes the loads identified in Item (?s) above.
(13) Determine the tests which have been performed to verify this logic identified above.
c.
Inspection Findings The inspector reviewed the following drawings and wiring diagrams:
(1) Bechtel Drawings E-102, Rev.12, " Diesel Generator Engine Control" O
E-356, Rev. 12 " Auxiliary Building and Emergency Diesel Generator
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Room Exhaust Fans" (continued)
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V E-181, Rev. 8. " Decay Heat Removal Pumps" E-211, Rev. 7, " Primary Makeup Pump A & C" E-241, Rev. 7, " Reactor Building Spray Pumps" E-275, Rev. 12 " Service Water Pumps A & C" E-276, Rev. 12 " Service Water Pump B" E-283, Rev. 2, " Service Water System Movers" M-406, Rev. 5, " Functional Description & Logic Diagram, Service Water" (2) Vendor Drawings B&W 01-0029, Rev.1, " Instruction Book for ESAS - Vol. 3" M-12-2, Rev. 8, " Engine D. C. Schematic, Diesel Electric Generator Unit" The licensee's installation does not have a SIS reset feature (licensee's terminology is ESAS signal - Engineered Safeguards Actuation System
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signal) as described in 'a' above. The ESAS signal is initiated by low reactor coolant pressure or high containment pressure. The only reset feature of the reactor protective system-(RPS) is included in the digital subsystera of the RPS, and it is functional only after the analog inputs have returned to their normal range. The operator has the ability to place all ESAS equipment in manual control at any time following an ESAS actuation.
All ESAS equipment in manual mode will, c.t the time of an RPS reset, be returned to the automatic mode by RPS circuitry.
The diesel generators are started on an ESAS signal or on a loss of power (LOP) to the 4.16 kev Engineered Safeguards buses 'A3' and 'A4'.
A LOP on a bus causes all loads on that bus to be sLJ sith the exception of the 480 V load control center.
Whenever an ESAS signal exists in conjunction with a LOP, the buses 'A3' and 'A4' are reloaded sequentially.
A primary concern of this inspection was to determine the operation of the auxiliary equipment required by the diesel generators.
A service water pump is powered from each bus 'A3' and 'A4'.
On a LOP /LOCA these pumps are shed and reloaded to the bus 15 seconds after diesel generator breaker closure.
The diesel generator room exhaust fans and dampers are powered through the 480 V load control center.
A start signal is supplied to these fans on a diesel generator start and they receive power immediately on diesel generator breaker closure.
Thus the licensee's electrical installation does not reflect the design defect described in
'a' above.
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-12-O The licensee last tested'the load sequencing feature on May 12, 1976 using their surveillance procedure 1304.08. This procedure simulated a LOCA/ LOP condition.
The inspector reviewed the records of this test.
No discrepancies were noted.
5.
Surveillance of Pipe Support and ' Restraint System a.
Objective of the Inspection The object of this portion of the inspection was to:
(1) Review changes to the program, schedule and procedures.
(2) Examine the technical content of procedures, observe dynamic pipe supports, fixed pipe supports and component support structures to verify that they are in satisfactory condition.
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(3) Review records of the licensee's most recent inspection and determine if the required inspection frequency was met and the inspection was consistent with established procedures.
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b.
Procedure Review The inspector reviewed procedure 1304.84, " Hydraulic Shock Suppression,"
Rev. 2, dated February 2,1977, and Job Order #4007 which provided for
the inspection of hydraulic shock suppressors.
The procedure categorizes the seppressors by system and is directed to conformance with the Technical Specification requirements of inspection for proper operation
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during each refueling shutdown.
The procedure requires examination for fluid leaks, damage and evidence of proper operation.
No special tests are presently included for lock-up velocity, bleed rate or sampling of the hydraulic fluid for viscosity or contamination.
f A procedure is not utilized for examination of fixed pipe supports.
Instead, inspection was performed according to Job Order #4711.
c.
Observation (1) Dynamic Pipe Supports a
The inspector observed hydraulic snubbers HS-5, HS-7, HS-85, HS-86 and HS-87 (Main Steam Lines) for the following conditions:
(a)
Deterioration and corrosion (b)
Support plates, extension rods and connecting jointr, bent, deformed or loose (c)
Bolts, nuts, washers tight and secure
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(d);
Bleed holes open and free from foreign material
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Lubricants applied where required (f)
Fasteners not loose or removed (g)
Equipment not locked-up or frozen (h)
Seals not deteriorated (1)
Physical damage or deformation The snubbers appeared to be satisfactory, however, it was noted that DCR #508 had been issued for HS-85, HS-86 and HS-87 to provide additional stiffening, retorquing and realignment.
Failure of the support plant for HS-86 had occurred and the bolts on HS-85 and HS-87 were found to be loose.
(2)FixedPipeSupports The inspector observed EFW-33, EFW-35, EFW-36 (Emergency Feedwater),
MS-169 and MK-5-3 (Main Steam) fixed pipe supports for the following conditions:
(a)
Deterioration and corrosion (b)
Deformation or forced bending v
(c)
Bolts. a.ut.s, washers and fasteners, tight and secure (d)
Contact or rubbing of pipe or supports with other pipes, supports, other equipment or components (e)
Sliding supports and brackets are provided with lubricants at the point of contact (f)
Spring hangers show the appropriate " Hot" or " Cold" position No problems were noted in the above observations.
(3)ComponentSupportStructures The inspector observed the support structures for those dynamic and
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fixed supports noted above and in addition, those utilized for support of hydraulic snubbers l A, 2A,1C and 2C (Reactor Coolant Pumps) for the following conditions:
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Deformation or forced bending (b)
Cracks, or other detrimental indications A)
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-14-No problems were noted with the exceptions of those discussed v
previously for HS-85, HS-86 and HS-87 support plates in the main steam system.
d.
Review of Records A review of the records resulting from the licensee's inspection of pipe supports and restraints indicated no problems relative to the fixed pipe supports. However, the records for the hydraulic snubbers confirmed the problems discussed previously in 5.c(1) above and one hydraulic snubber HS-14 (pressurizer relief piping) which had a damaged (leaking) seal. The seal was replaced utilizing Job Order
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The inspector had no additional questions in this area.
6.
Retrieval of Loose Parts - Surveillance Specimen Holder During this outage the fuel was completely removed from the core and an inspection of the fuel and reactor vessel was performed. This resulted in the location of debris, both in the reactor vessel and in one fuel element.
The pieces found and removed are enumerated as follows:
Piece #
Core Location
3"X4" flattened piece of pipe P-4
1/2" piece A-8
1/2" piece E-9
1/2" Curl C-12
1" flat piece N-13
1-1/4"X5" machine chip N-13
5/8"X2" Lalt 2nd level flow dist.
Additionally a 2-1/2"X2" piece of metal was removed from the bottom of fuel element lA31 (not to be returned to the core).
Pieces designated as 7 and 8 may not be pieces of metal and the location is not readily reached through the core support plate and flow distribution plates.
A piece estimated to be 1/2"X1" long and cylindrical in shape remains on the bottom of the reactor vessel.
The debris was recovered using procedure 1701.19, " Debris Recovery and Vacuuming Procedure," Rev. 1, dated February 20, 1977.
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