IR 05000313/1974002
| ML19319E466 | |
| Person / Time | |
|---|---|
| Site: | Oconee, Arkansas Nuclear, Crane  |
| Issue date: | 02/28/1974 |
| From: | Burke D, Kidd M, Murphy C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19319E447 | List: |
| References | |
| 50-313-74-02, 50-313-74-2, NUDOCS 8004100739 | |
| Download: ML19319E466 (15) | |
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R0 Inspection Report No. 50-313/74-2 Licensee:
Arkansas Power r.nd Light Company Sixth and Pine Streets Pine Bluff, Arkansas 71601 Facility me: Arkansas Nuclear One, Unit 1 Docket No.:
50-313 License No.:
CPPR-57 Category:
B1 Iocation: Russellville, Arkansas Type of License: B&W, PWR, 2568 Mwt Type of Inspection: Routine, Announced Dates of Inspection: February 5-8, 1974
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Dates of Previous Inspection: January 15-18, 1974 Principal Inspector:
M. S. Kidd, Reactor Inspector Facilities Test and Startup Branch Accompanying Inspectors:
D. J. Burke, Reactor Inspector Facilities Test and Startup Branch
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F. S. Cantrell, Reactor Inspector Facilities Test and Startup Branch Other Accompanying Personnel: None Principal Inspector:
4-J[-M
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M. S. Kidd,' React ar Inspector Date Facilit es Test and Startup Branch
.i Reviewed by:,
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C. E. Murphyo Qtief / \\
' Date Facilities Test and Startup Branch
'IRIPT Sil2rf 70: PDR LPDR /
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O RO Rpt. No. 50-313/74-2-2-J
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SUMMARY OF FINDINGS
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I.
Enforcement Action i
A.
Violations 1.
The following violations of Appendix B to 10 CFR 50 and your Final Safety Analysis Report (FSAR) commitments are considered to be of Category II severity.
a.
Step 5.8 of test procedure 200.08, " Pressurizer Relief Valve Test," was not adhered to during the conduct of the test, resulting in a test limit being exceeded, in violation of Criterion XI.
(Details II, para' raph 3.a)
g b.
Contrary to the requirements of Criterion XIII:
1) Stainless steel tubing was not stored as required by procedure 1004.11.
2) Stainless steel and bronze parts were not separated fsg (
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as required by Procedure 1004.11.
These items were corrected during the inspection. No additional response is required.
(Details III, paragraph 2.a)
c.
Contrary to the requirements of Criterion VI:
1) A copy of procedure 1004.11. " Handling, Storage, and Shipping of Q Listed Materials" was not provided in the store room or in the same building.
(Details III, paragraph 2.a)
2) Authority for approval of major changes to.preoperational test procedures was delegated to the Shift Supervisors.
0) eta 11s III, paragraph 2.c)
d.
Contrary to the requirements of Criterion III; The power supply boards for the pressurizer level and reactor coolant flow transmitters were modified without proper approval.
(Details III, paragraph 2.d)
a.
Contrary to the requiremen*s of Criterion XII, a program to maintain the radiation ronitor in the spent fuel area
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has not been implemented, (Details III, paragraph 3)
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RO Rpt. No. 50-313/74-2-3-
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f.
Contrary to commitments in paragraph 12.5 of the FSAR, procedure 1005.01 does not require that the Ope' rations Supervisor review the Station Log daily.
(Details III, paragraph 2.e)
2.
The following violatbus are considered to be Category III severity.
a.
Contrary to the requirements of 10 CFR 50.55(e),
1) The AEC was not notified of a breakdown in the quality assurance program that could have rendered the reactor building spray pumps inoperable.
(Details III, paragraph 2.d)
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2) The AEC was not notified of a design deficiency in the power supply to the pressurizer level and the reactor ceclant flow transmitters.
(Details III, paragraph 2.d)
III. New Unresolved Items b
74-2/1 Vibration / Noise in Reactor Vessel
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The cause of vibration / noises in the reactor vessel observed during hot functional testing has not been determined.
(Details I, paragraph 2)
74-2/2 control of Temporary Circuit Modifications
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The procedure for control of jumpers (ANO procedure 1005.04)
does not require jumpers, bypasses or lifted leads to be
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entered in the " Jumper Log" if the component or system is
" Hold Carded".
(Details III, paragraph 2.f)
74-2/3 controls For Maintenance Activities AN0's " level 4" maintenance activities do not require a procedure or job order.
" Level 4" includes torque settings and any minor adjustment without considering the requirements of Criteria II and XI of Appendix B to 10 CFR 50.
(Details III, paragraph 2.e)
IV.
Status of Previously Reported Unresolved Items 72-9/1 Incorporation of Safety Related Equipment in the FSAR Q-List s
Not inspected.
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R0 Rpt. No. 50-313/74-2-4-
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72-12/2 Valve Wall Thickness Verificat.sn Program
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Not inspected.
73-3/1 completion of Radiological Waste Disposal System No change in status.
The FSAR is to be revised in
Amendment 43, to be submitted in February of 1974, to identify the solid waste system as a Unit 2 system. This item remains open.
(Details I, paragraph 3)
73-5/2 Core Flood System Flow Rate Test No change in status.
See RO Report No. 50-313/73-19, Details I, paragraph 10.
73-8/1 Procedural Coverage per Safety Guide 33 Procedure development continues; however, considerable effort will be required to generate those not yet written and approved. This item remains open.
(Details I, paragraph 4)
73-10/5 Clean Radwaste System Test Procedure
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Test procedure 230.68 has been revised, but still may not conduct sufficient tasting to demonstrate system capability 5equired'.by acceptance criterion 7.5 of the procedure.
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This item remains open.
0) eta 11s I, paragraph 5)
73-10/6 Respiratory Protection Program and Procedures
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Not inspected.
73-10/7 Representative Sampling of Gaseous Wastes Not inspected.
- 73-12/2 Diesel Generator Trips Corrective action resulting from this problem have been completed on three of the five Unit 1 inverters. This item remains open.
(Details I, paragraph 6)
73-12/3 control Rod Trip Test Comments on TP 330.05 have not yet been resolved. This item remains open.
(Details I, paragraph 7)
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73-14/2 Initial Core Load Procedure
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The original draft of procedure 1502.04 has been rewritten and is ready for review for approval. This item remains open.
(Details I, paragraph 8)
73-14/3 Leak Testing of the Personnel Hatch No change in status.
The Unit 1 Technical Specifications are to be changed to agree with Appendix J to 10 CFR 50.
This item remains open.
(Details I, paragraph 9)
73-16/1 Radiography Review i
Not inspected.
73-17/1 Pressurizer Electromatic Relief Valve No change in status. This item remains open.
(Details I, paragraph 10)
73-17/2 Emergency Operating Procedures The licensee is approximately 75% complete with his review and rework of the emergency procedures and appears to be incorporating ce
- nsidering the inspector's comments.
(petails III, par graph 2 of 73-17). The majority of work a
to be completed consists of adding valve, indicator, or
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procedure numbers, retyping and approving the procedures involved.
(Details II, paragraph 2).
73-17/3 operational Test Program Not-inspected.
73-18/1 Emergency Planning j
Not inspected.
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73-18/3 Calibration of Radiation Monitors Not inspected.
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73-19/1 Inverter Malfunction Not inspected.
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R0 Rpt. No.,50-313/74-2-6-
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73-19/2 Reactor Building Ventilation System.Ductwork
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Testing of a mock-up of the ductuork is to be conducted to determine the magnitude of reinforcement needed. This item remains open.
(Details I, paragraph 11)
74-1/1 Valve Deficiencies (Regulatory Operations Bulletin 74-Q
No valves of the types discussed in ROB 74-1 are installed in Unit 1.
This item is considered closed.
(Details I,
paragraph 12)
i V.
Unusual Occurrences
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Noise in Reactor Vessel / Reactor Coolant System During the conduct of hot functional testing on February 4,1974, noises were heard in the upper and lower sections of the reactor vessel on the loose parts and vibration monitor instrument. The
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noises seemed to disipate as the reactor coolant system (RCS)
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temperature was lowered. The licensee had not determined the exact f
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causes of the noises at the end of the inspection.
(Details I, l
l paragraph 2)
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VI.
Other Significant Findings
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A.
Project Status The start of Unit 1 hot functional testing was delayed ten days due to problems in the cleanup of condensate and feedwater.
It was also interrupted for approximately ten days for repairs t.o a reactor coolant pump seal, the startup boiler, and a control rod drive mechanism stator. The inspector was informed by telephone February 13, 1974, that the licensee's fuel load estimate had been changed from March 25, 1974 to April 13, 1974.
B.
Personnel Changes
Effective February 14, 1974, N. A. Moore, Chief Quality Assurance Coordinator will be promoted to Manager of Quality Assurance.
This will involve relocation from the Arkansas Nuclear One site to AP&L corporate offices in Little Rock. Moore will report directly to J. D. Phillips, Senior Vice President. This change is to be discussed in Amendment 43 to the FSAR, due for submittal in February.
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RO Apt. No. 50-313/74-2-7-
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VII. Management Interview
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A management interview was held at the conclusion of the inspection February 8, 1974. The following licensee. representatives attended:
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Arkansas Power and Light Company (AP&L)
J. W. Anderson - Plant Superintendent T. Baker - Chemical and Radiation Protection Engineer T. H. Cogburn - Nuclear Engineer R. R. Culp - Test Administrator C. A. Halbert - Technical Support Engineer C. H. Miller - Assistant Plant Superintendent
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N. A. Moore - Chief Quality Assurance Coordinator J. A. Orliek - Quality Control Engineer M. H. Shanbhag - Procedure Administrator
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B. A. Terwilliger - Operations Supervisor i
The apparent violations listed in Section I were discussed.
Information on these items is given in Details II, paragraph 3.a and Details III, paragraphs 2 and 3.
The new unresolved items in Section III were discussed. Detailed information en these items is given in Details I, paragraph 2 and
Details III, paragraphs 2.c and 2.f.
i The inspector stated that the failure of the preoperational test program to require a review of the jumper log prior to initiating the specific test, appeared to be a weakness in the program.
(Letails III, paragraph 2b)
The inspector also stated that the use of a " loose leaf" jumper log appeared to be a weakness in the jumper control system in that a
page of the log could be lost without it being missed.
(Details III,
. paragraph 2b)
The status of previously identified unresolved items (Section IV)
was discussed..The inspector stated that the item regarding valve deficiencies (ROB 74-1) was considered closed and that all others listed in Section IV would remain open.
Information on AP&L's response to ROB 74-1 is given in Details I,. paragraph 12.
He also stated that he would request a meeting with AP&L management to discuss the status of Unit 1 unresolved items, procedure adherence (Details II, paragraph 3a.), and procedure development (Details I, paragraph 4).
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RO Inspection Report No. 50-313/74-2 I-1
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DETAILS I Prepared by:
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JN'M M. S. Kidd,' Reactor Inspector Date Facilities Test and Startup Branch Dates of Inspection: Feb ry 5-8, 1974
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Reviewed by: [
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C.' E. Murpl(y,, Chic
' Dade Facilities Test and Startup Branch 1.
Persons Contacted
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In addition to those listed in the Management Interview section, the following persons were contacted.
Arkansas Power and Light Company (AP&L)
P. Almond - Reactor Technician s
W. Cayanaugh - Production Project Manager Babcock and Wilcox Company (B&W)
J. N. Kaelin - Site Operations Manager 2.
Noise in Reactor Vessel / Reactor Cooling System During the conduct of hot functional testing on February 4,1974, vibrations / noises were heard in the upper and lower sections of
,1 the reactor vessel on the loose parts and vibration monitor. The noise, possibly caused by water hammer, was more noticeable at system temperatures close to normal operating values (532 F).
Also, the noise appeared to be associated most with the operation j
of only reactor coolant pump (RCP). "C".
Licensee personnel were
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j of the opinion that little or no noise was observed when pumps other than "C" or combination of pumps were running.
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AP&L conducted several tests involving different pump combinations of the three operable pumps, RCP "D" being shut down due to'a seal staging problem. Also testing was done at several temperature levels. H e noise seemed to dissipate when the system temperature was lowered. Almost no noise was heard at a temperature of 3000F.
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B&W site personnel and Lynchburg, Virginia, personnel with experience
'in the use of the vibration and noise monitor listened to tape
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recordings (Lynchburg) of the monitor' output and advised AP6L that sissilar noises were heard at Oconee I and Three Mile Island, Unit I, and that at those facilities the decision had'been made that the acises were of little concern.
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The inspector' stated that he would request a report on the' subject be
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jpada aya11able for his review such that RO could be informed of the exact natuye of the problem and gain assurance that the decision to continue
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testing was justified. He further stated tamo 1 utter would be carried as an unresolved item.
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3.
Completion of Radiological Waste Disposal System O
This unresolved item was initially discussed in RO Report No.
50-313/73-3, Details I, paragraph 4.
The inspector was informed
.that the status of the solid radwaste system would be discussed
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in Aneendment 43 to the' Unit 1 FSAR, to be submitted later in
s. February. This system is to be defined as a Unit 2 system. This j
item remains open.
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-4.
Procedural Coverage Per Safatv Cnido 33 i
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1 A detailed' status of the procedures needed for operation and not
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yet' written"was given'in RO Report No. 50-313/74-1, Details I,
. paragraph 4.
In the three weeks between inspections, several
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procedures had been written, but many remain to be written. The
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following information was given to the inspector on February 7, 1974:
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RO Rpt. No. 50-313/74-2 I-3
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Identified Written Approved Quality Control
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Administrative
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79 Operating
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Emergency, Abnormal,
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and Security
46 Alarm Response
~500
0 Surveillance Tests 123
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Maintenance
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Refueling
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Chemical and Radiation Protection
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The inspector stated that development of the procedures was of concern and that this unresolved item would remain open.
5.
Clean Radwaste System Test Procedures
Comments on approved test procedure 230.68, " Clean Radwaste System,"
l formed the basis for this unresolved item and were documented in RO
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Report No. 50-313/73-10, Details II, paragraph 6.c.
At that time, it was commented that the procedure as written would not conduct
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testing necessary to verify the. requirements of acceptance criteria 7.4 and 7.5.
During the current inspection, the inspector reviewed changes made to the procedure which would conduct testing needed to verify 7.4 and had no questions. Regarding criterion 7.5, the inspector informed licensee personnel by telephone February 14, 1974,
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E0 Rpt. No. 50-313/74-2 I-4
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that his review of changes made to the proeddure' revealed that they had not increased the scope of testing such that it would demonstrate that each of the clean vaste receiver tank transfer pumps was capable of transferring water from any one of the clean waste receiver tanks to the treated waste monitor tanks through the filter and
'demineralizar train at 100 gym and a discharge pressure of a65 psig. He further stated that the item would require further
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discussion and would remain open.
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6.
Diesel Generator Trips
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This' unresolved it:em was last discussed in RO Report No. 50-313/73-19,
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Details I, paragraph.12. The modifications to inverters described M.
in a li.censee report per 10 CFR 50.55Gil dated October 314 1973, entitled; " Loss of Power To Vital Busses," have been completed on I
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three of tha flye Unit 1 inverters. The. inspector stated that
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this item would remain open pending completion of modifications l
and subsequent ratesting.
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control Rod Trip Test
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Connants on' TP 330.05 were documented in RO Report No. 50-313/73-12, Detai,la I, paragraph. 5.
Additional licensee' questions on the
. coinments were discussed during the current inspection. No changes had yet been made to the procedure. The inspector stated that this
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item would remain open.
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8.
Initial Core Ioading Procedure
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ThisunresolveditemOonsistsofcommentsonprocedure 1502.04 documented in RO Report No's. 50-313/73-14, Details I, paragraph 4
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50-313/73-17, Details I, paragraph 18, and 50-313/74-1, Details I, paragraph 6.
Licensee personnel stated that the original draft had been revised substantially, retyped, and was ready for AP&L's
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RO Rpt. No. 50-313/74-2 l installed sourcs
. Regarding response checking of the perranent y
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range instruments (2) prior to the' start of fuel loading, licensee personnel stated that it was impossibl
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f ling
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This canal filled with water, as it will be for fuel loading.
l and statement was verified by examination of the reactor vesse*ih between the reactor vessel outside wall and the biological biological shield areas.
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Licensee repre-the area in which. the instruments are located.se
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first two fuel assemblies are inserted.The inspector stated that this m
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will contain a source.
should'be described in the procedure.
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for
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In discussing the two auxiliary detectors which will be used liability of the initial loading, the inspector questioned the reHe was inform the power sources for them.
powered by vital buses (non-interruptable).
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in a The inspector stated that approval of the fuel loading procedu O
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form which resolves previous comments would remain open.
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Personnel Hatch Leak Testing <
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Licensee personnel stated that the requirements for surveillance ill testing of the hatch in the Unit 1 Technical Specifications w be revised in Amendment 43 to the FSAR to be sub h Appendix J to 10 CFR 50.
, February to make them compatible wit
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This item remains open.
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discussed'in RO Report No. Licensee personnel stated that t
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This item remains open.
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Reactor Building Ventilation System Euctwork
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1974, 11.
_A licensee report per 10 CFR 50.SS(e) dated January 21,
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entitled, " Reactor Building Ductwork," has been recLicensee personnel inf 15, 1974.
t d at the site on a A final report is due March d
the inspector that testing would be con uc e of additional
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mock-up of the ductwork to determine the extentbra
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d the 2 paid
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expected during a design basis accident.
d subsequent modifications was sufficient time to perform the testing anThe inspector stated that this d modifications.
prior to fuel loading. remain open pending evaluation of the testing an 74-11 12. Valve Deficiencies (Regulatory Operations Bulletin
i garding l
R08 74-1, dated January 3, 1974, requested informat on reA license certain Walworth and Darling valves.
february 1,1974, in response to the ROB, states that no valv l
of the types described in the R0B have beenThe inspector stated that this matte nor are any to be installed.
i was considered resolved for Unit 1.
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k 11. Reactor Building Ventilation System Ductwork
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A licensee report per 10 CFR 50.55(e) dated January 21, 1974, entitled, " Reactor Building Ductwork," has been received by RO:II.
A final report is due March 15, 1974. Licensee personnel informed the inspector that testing would be conducted at the site on a
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mock-up of the ductwork to determine the extent of additional bracing / stiffening that will be required to withstand the 2 psid expected during a design basis accident. It was felt that there
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was sufficient time to perform the testing and subsequent modifications prior to fuel loading. The inspector stated that this item would remain open pending evaluation of the testing and modifications.
12. Valve Deficiencies (Regulatory Operations Bulletin 74-1)
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ROB 74-1, dated January ~3, 1974, requested information regarding certain Walworth and Darling valves. A licensee letter dated february 1, 1974, in response to the ROB, states that no valves of the. types described in the ROJ have been installed in Unit 1, nor are any to be installed. The inspector stated that this matter was considered resolved for Unit 1.
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DETAILS II Prepared by:.8I d--
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.D. J.5 Burke, Reactor Inspector
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Facilities Test and Startup Branch Dates of Inspection: Febru ry 5-8, 1974
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Reviewed by:
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C. E. Murphyk.Cgie f
/ Dafe Facilities Test and Startup Branch
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1. ' Individuals contacted
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Arkarisas Power and Light Company (AP&L)
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J. W. Anderson - Plant Superintendent B. A. Terwilliger - Operacians Supervisor C. A. Halbert - Technical Espport Engineer and Test Supervisor 2'.
Emergency Operating Procedures p
Tha licensee is approximately 75% complete with his review of the 5,
emergency procedures and appears to be incorporating or considering the inspector's general and specific comments (R0 Inspection Report No. 50-313/73-17, Details III, paragraph 2).
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. The inspectors have discussed the majority of the emergency operating procedures with the licensee and both'have agreed upon the revisions and/or changes required. The majority of work to be completed appears to be adding valva numbers or referenc1ng more procedures, retyping, and reapproving the procedures. Specifically, the inspector discussed-the following procedures with licensee personnel:
1202.0lR1, Load Rejection: No further comments.
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1202.02, Blackout: Rewrite is nearly complete following consideration of degraded power sources and the inspector's comments.
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