IR 05000313/1977021
| ML19326B437 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 12/06/1977 |
| From: | Dickerson M, Madsen G, Westerman T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML19326B415 | List: |
| References | |
| 50-313-77-21, NUDOCS 8004150792 | |
| Download: ML19326B437 (10) | |
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U. S. NUCLEAR REGULATORY COMMISSION i
0FFICE OF INSPECTION AND ENFORCEMENT
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REGION IV
Report No. 50-313/77-21 Docket No. 50-313 License No. DPR-51 i
Licensee:
Arkansas Power and Light Company P. O. Box 551 Little Rock, Arkansas 72203 Facility Name: Arkansas Nuclear One, Unit 1 Inspection At:
Arkansas Nuclear One Site, Russellville, Arkansas Inspection Conducted: November 14-18, 1977 Inspector:
%.co.
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/2//./7E T. F. Westerman, Reactor Inspector Date
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M. W. Dickerson, Reactor Inspector Date
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/s/4/77 Approved By:
_e G. L. Madsen, Chief, Reactor Operations and Date Nuclear Support Branch
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s Inspection Summary
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Inspection on November 14-18, 1977 (Report No. 50-313/77-21)
Areas Inspected: Routine, unannounced inspection involving the review of IE Bulletins and Circulars; Licensee Event Reports; reactor shutdown margins; core thermal power; plant procedures; plant records; and followup on previously identified matters. The inspection involved 62
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inspector-hours on-site by two (2) NRC inspectors.
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Results: Of the seven (7) areas inspected, no items of noncompliance or deviations were identified in six (6) areas.
One infraction was
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identified (paragraph 7.b) in one area.
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-2-DETAILS 1.
Persons Contacted
- J. Anderson, Jr., Plant Superintendent
- M. Bishop, Records Supervisor
- J. Lowman, Assistnat I&C Supervisor P. Rogers, Reactor Engineer
- B. Terwilligner, Supervisor Plant Operations
- T. Cogburn, Nuclear Engineer
-*L. Humphrey, Quality Assurance Engineer
- L. Alexander, Quality Control Engineer S. Strasner, Quality Control Inspector D. Hamblen, Quality Control Engineer
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B. Austin, Clerk M. Jennette, Clerk
- Attended exit interview..
2.
Follow up on Previously Identified Findings (Closed) Unresolved Item (313/7713-9)
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Amendment No. 28 to the Facility Operating License has been issued.
This amendment updates the facility heatup and cooldown curves.
(Closed Unresolved Item (313/7709-11)
Corrections to the Nova Computer hardware inputs associated with 5-rod position indication and incore detectors have reduced the variation between the Nova and plant computer (linear heat rate and radial peaking) to within 1/4%.
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(Closed) Open Item (313/7710-13)
QCP 1004.13 has been revised (revision 4) and now conforms to the conditional release provisions of the licensee's approved QA Plan,
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Section 7.3.2.
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3.
Determination of Reactor Shutdown Margin
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The inspector reviewed the licensee's procedure for determining reactor shutdown margin (1103.15, Rev. 4, " Reactivity Balance Calculation") for technical adequacy. Calculations performed v
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O-3-May 14,1977 (other than hot shutdown conditions), May 13,1977 (hst shutdown), and October 9,1977 (reactor at power), were also reviewed.
The inspector's review included:
procedure Content
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Shutdown Calculations
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Consideration of Reactivity Variables
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Cr:11ance with Technical Specifications
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Within the areas inspected, there were no items of noncompliance. The establishment of a plant procedure designating operator actions (as prescribed in the Technical Specifications Section 3.5.2) in the event of an inoperative control rod is considered an open item. The licensee was in the process of changing the rod worth curves for the end of core life (all rods out) ~ condition. The variation between the total rod worth on attachments C-3 and C-4 of plant procedure 1103.15 was also brought to the licensee's attention. These latter two items were tc be resolved with the issuance of the new rod worth curves.
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4.
Core Thermal Power Evaluation The inspector reviewed the licensee's procedure for evaluating core thermal power (1103.16 " Heat Balance Calculation") for technical adequacy. Heat balance calculations for November 15,1977 (Nova and plant computer) were also reviewed. The licensee normally relies on the plant computer to perform heat balance calculations with the Nova
. computer as a backup mechanism. plant procedure 1103.16 has provisions for performing hand calculations. The inspector's review included:
Establishment of Initial Conditions
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Calibration of Associated Instrumentation
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Comparison of Results to Short Form Primary Calorimetric
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Heat Balance Equations
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Frequency of Calculation (September 1-30, 1977)
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Within the areas inspected, there were no items of noncompliance. The calibration of associated instrumentation is considered an open item.
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This includes the formulization of calibration procedures and follow up of calibrations to be performed during the upcoming refueling outage now scheduleo for mid-January 1978.
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Reportable Occurrences The following Reportable Occurrences (R0's) were reviewed by the inspector:
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RO 50-313/77-15 - Closed, based on previous in-office review.
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RO 50-313/77-16 - Closed, based on previous in-office review.
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RO 50-313/77-17 - Open, pending installation of permanent drain
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line.
Subsequent repair to drain line required cutting and capping the line.
RO 50-313/77-18 - Open, petiding documentation of training
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conducted.
R0 50-313/77-19 - Open, pending final corrective action.
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6.
IE Bulletin and Circulars The inspector reviewed the status of the following IE Bulletins and Circulars:
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Testing of pneumatic tire delay relays is to be conducted during the upcoming refueling outage (Ref: AP&L September 2,1977 letter to RIV).
The licensee's action to close out this Bulletin remains incomplete.
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The plant does not utilize AR latching relays.
The Bulletin is considered closed.
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The licensee reports that they have experienced no containment valve leakage test problems due to resilient valve seats.
Valves of this type are located outside of the containment building and
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are subject to a changing environment. However, based on the licensee's present experience with these valves, no further action is planned.
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This Circular is considered closed.
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-5-i IE Circular 77-13
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I The inspector verifi the licensee's review of the Circular as
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related to reactor sataty signals negated during testing.
The conclusions drawn by the licensee are as follows:
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Pressurizer levels are not inputs to safety systems in
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a B&W plant.
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Procedures specify the limitations, restrictions,
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precautions, etc., required to assure that safety-related protection systems are operable when required by Technical Specifications.
Procedures are written such that critical decisions
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i are ' built in' and not left to plant personnel.
Established operator training program.
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The inspector did conclude that further action by the licensee is not necessary. The ESAS and Reactor Protective System (RPS) are required by procedures to be operable during heatup, startup, normal operation and cooldown. One group of i
control rods are withdrawn during heatup and cooldown operations.
Only one channel of the RPS may be bypassed at a time.
7.
Procedures
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Areas Inspected The inspector reviewed plant procedures to verify that they had been reviewed and approved, that control of changes was in accord-
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ance with the Technical Specifications, that Technical Specification
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revisions have been incorporated into procedural changes, and that the content of the procedures confom to TS requirements.
In addition, the content of several procedures were reviewed for
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their technical adequacy to control the particular safety-related operation.
l The following areas were inspected by selective review of repre-
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sentative procedures in each area:
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(1)
Administrative
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1005.01 Administrative Controls Manual, Rev. 3, PC-3, 10/21/77 1005.04 Control and Use of Bypasses and Jumpers, Rev. 3, 7/27/77
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General Plant Operations
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1102.01 Plant Preheatup and Precritical Checklist, Rev. 4, PC-2,7/18/77 1102.02 Plant Startup, Rev. 5, PC-4, 10/28/77 1102.04 Power Operation, Rev. 1, PC-5, 8/3/77 (3)
Startup, Operation and Shutdown of Safety Related Systems 1103.02 Filling and Venting the R.C System, Rev. 2, PC-1, 8/18/77 1103.01 Soluble Poison Concentration Control, Rev. 4, 9/12/77 1103.05 Pressurizer Operation, Rev. 2,8/18/75 liO3.06 Reactor Coolant Pump Operation, Rev. 2, PC-1, 9/19/77 1103.11 Draining and N2 Blanketing of Reactor Coolant (N System, Rev. 1. PC-1, 9/19/77
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1104.12 Operational Test Control, Rev. 4,1/25/77 1104.28 ICW System Operating Procedure Rev. 2, PC-3, 7/18/77
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1104.29 Service Water and Auxiliary Cooling System, Rev. 3,
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PC-2, 9/29/77
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1104.35 Fuel Handling and Radwaste Ventilation, Rev.1, 8/18/75 1105.01 NI & RPS Operating Procedure, Rev.1, PC-1, 8/18/77 1106.16 Condensate, FW and Steam Systems Operation, Rev. 3,
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PC-2, /28/77 (4)
Abnormal Condition
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1203.01 ICS Abnonnal Operation, Rev. O, 5/26/73
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1203.04 Reactor High Startup Rate, Rev. 1,4/16/74
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Emergency and Other Significant Events 1202.02 Blackout-Loss of All Auxiliary Power Except Batteries, Rev. 2,7/22/75 1202.05 Degraded Power, Rev. 1, PC-1, 9/27/77 1202.06 Loss of Reactor Coolant /RC Pressore, Rev. 3, j
9/3/75 1202.09 Loss of Condenser Vacuum, Rev. 3,7/22/75 1202.14 Loss of Reactor Coolant Flow - RCP Trip, Rev. 3, 7/22/75 1202.23 Steam Generator Tube Rupture, Rev. 2,7/29/75 (6)
Maintcaance I
1401.01 Replacement of Strainers and Filters, Rev. O, 4/17/74 1401.04 OTSG Tube Replacement Maintenance Procedure, Rev.1, 2/16/74 1401.06 Repair of Pressurizer Code Relief Valves, Rev. O,
3/19/74 1401.13 Reactor Coolant System Temperature Detector, Removal and Replacement, Rev. 1,2/16/74
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Fj,ndings (1)
The inspector's review of the above listed procedures
established that two procedures 1102.04 " Power Operation,"
Rev.1, PC-5, dated August 3,1977, and 1104.12 " Operational Test Control," Rev. 4, dated January 25, 1977, were not being maintained up-t1-date.
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Review of procedure 1102.04 established that it contained Attachment L, " Control Rod Withdrawal Limits for Four (4)
,i Pump Operation for 0-115 EFPD + 10," while the plant was
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then operating at approximately 214 EFPD. The appropriate control rod withdrawal limits for the period 115 + 10 to
' i 225 + 10 should have been contained in the procediire. A change to the procedure correcting this deficiency was issued during the inspection on November 17, 1977.
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A review of revised Technical Specifications and procedure 1004.12 established that the procedure had not been revised since November 22, 1976 to provide for revised surveillance
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tests and frequencies. The Operctional Test Control procedure is utilized to establish the test control program
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i which assures that all testing required by the TS are identified and perfonned in accordance with written
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procedures.
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The licensee was infonned that failure to maintain up-to-date
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procedures was considered an infraction and is contrary to the Technical Specification requirements of Section 6.8.1 and to Section 5.1 of procedure 1005.01, Administrative Controls
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Manual. Section 6.8.1 of the TS states in part that,
" Written procedures shall be established, implemented and maintained covering the activities referenced below:
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The applicable procedures recomended in appendix A-of Regulatory Guide 1.33, November 1972.
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"c.
Surveillance and Test Activities of safety related
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equipment."
This is amplified by Section 5.1 of procedure 1005.01 which
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states in part that, "... the manual is written for the purpose of having an up-to-date compilation of procedures necessary for the safe operation of the plant...."
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Two additional Technical Specification changes were also noted which have not been reflected in current procedures.
i These are Amendments 17 dated December 17,1976, and 21 dated March 31, 1977. Seventeen pertains to fuel handling i
restrictions and 21 pertains to surveillance test require-ments of the Reactor Vessel Internals Vent Valve. Since neither of thest are required to be implemented at this
time, the procedural changes required will remain as open
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items.
(3)
. Several procedures reviewed contain incorrect references in various steps of the procedures, as follows:
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1102.02
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Section 6.2.13 states, " Verify that the Emergency D-G sets are in an operable conf tion per OP 1104.36, section 6.4."
However, section 6.4 or 1F 1104.36 states, " Verify oil
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level one inch above fW l with engine stopped." This is
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- apparently an incorrect reference as section 6 and 7 are
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placing D-G #1 and #2 in operation, respectively.
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_9 1104.13 Section 4.1 references section 4.0 (Limits and Precautions)
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of procedure 1103.11 which in turn references procedure i
1101.02, Plant Set Points. This is apparently the wrong
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reference (shculd have been referenced to Limits and i
Precautions) and should directly reference the applicable section of 1101.01, Limits and Precautions.
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Section 5.0 also references 1103.11, section 5, which references 1101.02.
Should directly reference the appropriate section of 1101.02.
Section 6.1.1.b references 1103.11, section 6.6.
However, there is no section 6.6 in procedure 1103.11.
Section 6.1.14, 6.1.17, 6.2.la, and 6.2.5.
The procedure
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in these sections does not appear to clearly indicate the sequential steps to be taken.
The revision of procedures 1102.02 and 1104.13 will remain an open item.
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Records The inspector reviewed the licensee's program for the control, storage, retention and retrieval of records and documents to determine if it is in conformance with the Technical Specifications and procedures.
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The following records were reviewed:
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Core flood check valve test, Supplement I to 1104.01,1/28/77
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RC pump #1 seal flow recorder, 10/1-28/77
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DG #1 Monthly Surveillance record per 1104.36, Supplement 1, 5/23 and 6/23/77 d.
NSS and Safeguards Log, 7/31-9/1/77 i
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Daily Power History Log, 12/1/76 - 1/28/77
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Tendon Surveillance 1304.91, 2/7/77
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DCR 484 Replacement of Operators - Decay Heat Isolation Valves, completed 2/6/77 and associated drawings:
E-182, Sheet 3, Rev. 4
' E-517, Sheet 4, Rev.16
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E-517 Sheet 9, Rev. 3
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DCR 412 Add Provisions for Batch Make-up to Core Flood Tank -
not yet approved..
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The following DCR's with intent changes i
DCR 281 Addition of Check Valve on Discharge Line of Emergency DG Lube Oil Soak Back Pumps. Safety Determination Report of 2/24/75.
DCR 424M Level Indicator and Alarm of Improper Process Lineup in Clean and Dirty Liquid Radwaste System. Safety Determination Report of 12/3/76.
DCR 566 Temporary Installation of Emergency Air Conditioning Units in Battery / Charger Areas. Safety Determination Report of 10/26/77.
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No items of noncompliance or deviations were identified in this area.
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Exit Interview The inspectors met with licensee personnel at the conclusion of the inspection on November 18, 1977. The inspectors summari2ed the scope of the inspection and the findings.
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