IR 05000313/1974015
| ML19320A088 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 01/06/1975 |
| From: | Herdt A, Kidd M, Robert Lewis NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19320A074 | List: |
| References | |
| 50-313-74-15, NUDOCS 8004140697 | |
| Download: ML19320A088 (14) | |
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UNITED STATES
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ATOMIC ENERGY COMMISSION DIRECT 0 HAT 4 0F REGULATORY OPERATIONS I
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RO Inspection Report No. 50-313/74-15 Licensee:
Arkansas Power and Light Company Sixth and Pine Streets Pine Bluff, Arkansas 71601 Facility Name: Arkansas Nuclear One, Unit 1 Docket No.:
50-313 License No.:
DPR-51 Category:
B2 Location: Russellville, Arkansas
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Type of License: B&W, PWR, 2568 MWe Type of Inspection: Routine, Announced j
Dates of Inspection: November 20-22 and December 10-13, 1974 Dates of Previous Inspection: November 5-8 and 14, 1974
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Principal Inspector:
M. S. Kidd, Reactor Inspector Facilities Section, Facilities Test and Startup Branch Accompanying Inspectors:
A. R. Herdt, Metallurgical Engineer
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Engineering Section, Facilities Construction Branch i
Other Accompanying Personnel: None Principal Inspector:
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M. S. Kidd, 'Reacto'r Inspector, Facilities Date Section Facilities Test and Startup Branch
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Reviewed by:
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or Reactor Inspector, Fac111 tics Date R.C. Lewis,Sen(itiesTestandStartupBranch Section, Facil PD3-l
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20 Rpt. No. 50-313/74-15-2-
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SUMMARY OF FINDINCS l
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I.
Enforcement Action
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A.
Violations
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1.
The following repeat violation is considered to be of
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i Category II severity:
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Contrary to Technical Specification 6.7.1, certain requirements of procedure 1005.04, " Control and Use of Bypasses and Jumpers,"
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were not met for an electrical jumper applied November 22, 1974, as documented on log sheet number 68.
(Details I, paragraph 3)
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2.
The following violation is also considered to be of
Category II severity:
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On November 23, 1974, the reactor coolant pump underpower
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monitoring relays failed tc sense the loss of pumps, violating
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a limiting condition for operations given in Technical
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Specification 3.5.1.
Corrective actions discussed in the
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licensee's abnormal occurrence report on this matter were verified to be complete during the inspection.
(Summary, V.C)
B.
Safety Items
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None
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II.
Licensee Action on Previously Identified Enforcement Matters
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A.
Violations
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1.
Conduct of Test Without an Approved Procedure (R0 Report No.
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A procedure has been written and approved for ratesting the control room "Halon" fire control system. This iten remains
open.
(Details I, paragraph 2)
2.
Control of Jumpers and Bypasses (R0 Report No. 50-313/74-13, I.A.1.a)
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Preventive and corrective measures discussed in the licensee response-on this item were verified to be complete and all old records were in order; however, a new violation, repetitious of certain of the original ones, was discovered.
(Details I,-paragraph 3)
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3.
Control of Portable Survey Instruments (R0 Report N. 50-313/74-14.
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Not inspected.
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RO Rpt. No. 50-313/74-15-3-
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Maintenance of Form AEC-5 (R0 Report No. 50-313/74-14. I.A.2)
Not inspected.
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Safety Items
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Not applicable.
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III. New Unresolved Items 74-15/1 Generator Trip Test
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The licensee does not wish to perform a generator trip or separation L
test at 100 percenc power.
CDetails I, paragraph 7.c)
74-15/2 Regulatory OperationsBulletin 74-15 l
l This item will be carried as unresolved pending receipt of the
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information requested in the bulletin, dated December 6,1974.
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j IV.
Status of Previously Identified Unresolved Items 74-11/1 Control Room Ventilation System Modifications to the system are necessary before retesting can be i
accomplished. This itemiremains open.
CDetails I, paragraph 41 l
74-14/1 Reactor / Turbine Trip Test
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Explanation as to why certain acceptance criteria for this test
were not met have been added in the test record. This item is resolved.
CDetails I, paragraph 5)
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l 74-14/2 Rod Reactivity Worth Mecsurements TP 800.20, " Rod Reactivity Worth Measurements," was conducted at the 75 percent power level. This item is resolved.
(Details I, paragraph 6)
74-14/3 Radiation Levels Inside Reactor Containment I*
Not inspected.
74-14/4 Contamination and Airborne Radioactivity Survey Forms i
Not inspected.
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74-14/3 Regulatory OperationsBulletin 74-12
?A response to this bulletin has been received and this item is considered resolved.
(Summary, VI, B)
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d) Rpt. No. 50-313/74-15-4 -
74-14/6 Regulatory OperationsBulletin 74-13
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A response to this bulletin is to be submitted within the time
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frame requested. This item remains open.
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74-14/? Reactor Building Spray Line Cracks
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The licensee is evaluating the cause of the cracks and determining whether other cracks might exist in other plant systems which utilize
similar stainless steel piping. Metallurgical examinations of the cracks are being performed by the licensee as well as an independent metallurgical examination for the AEC by Battelle Columbus Laboratories This item remains open.
(petails II, paragraph 3)
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V.
Unusual Occurrences
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Fish Impingement
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The licensee reported that fish impingement on the plant intake screens had exceeded the limit of 3,0)0 pounds per day in Environmental Technical Specification 4.1.2 (A) 2 for 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> periods on November 14-15,
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i November 17-18, November 20-21, December 2-3, December 3-4, December 5-6,
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j December 10-11, December 11-12, and December 12-13, 1974. Written reports for the first six sampling periods (ending December 5-6) were i
submitted by December 12, 1974, as non-routine environmental reports
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74-1 through 74-5 for docket number 50-313. Similar reports are to be
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submitted for the other occurrences.
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B.
Instrumentation Sensing Line Leak On November 10, 1974, the licensee reported an abnormal occurrence
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involving a leak through a socket weld on a small flow sensing
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instrument line.
(Nbnormal Occurrence Report No. 50-313/74-12,
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dated November 19, 1974)
C. ' Reactor Coolant Pumps Underpower/ Overpower Monitoring Relays
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On November 24, 1974, the licensee reported an abnormal occurrence
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involving the failure of the Reactor Protection System to function
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as designed when reactor coolant pump underpower/ overpower monitoring
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relays failed to sense an underpower condition due to maladjustment.
.(Abnormal Occurrence Report No. 50 ~13/74-13, dated December 3, 1974)
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VI.
Other Significant Findings A.
Decay Heat System Piping Crack - Supplemental Report
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The licensee submitted a supplementary report to Abnormal Occurrence
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Report No. 50-313/74-2, involving a crack in the socket weld for the c.
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vent valve DH-1010 nipple, on November 26, 1974.
(Abnormal Occurrence
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Report No. 50-313/74-2A)-
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Regulatory OperationsBulletin 74-12 A licensee response to this bulletin, dated November 27, 1974, sectes that no Westinghouse type "SG" relays are used at ANO, Unit 1.
C.
Witness of Power Ascension Testing l
The inspector witnessed the performance of TP 800.05, " Reactivity i
Coefficients at Power," at 100 percent power and had no comments.
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(Details I, paragraph 7.a)
D.
Review of Test Data
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The inspector performed a general review of 75 percent power ascension testing results and had no comments.
(Details I, paragraph 7.b)
VII. Management Interviews
-A.
The ihspection findings of November 21-22, 1974, listed below, were discussed with J. W. Anderson, Plant Superintendent, and members of his staff at the conclusion of the inspection.
1.
Review of radiographs taken of the reactor building spray and
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decay heat piping systems.
(Details II, paragraph 3)
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Investigation to determine whether other heats of material were furnished by Swepco Tube Corporation and the heat treating procedure used.
(Details II, paragraph 3)
3.
Status of metallographic report from Betchel Corporation.
(Details II, paragraph 3)
B.
A managosent interview was conducted December 12, 1974, with J. W. Anderson, Plane Superintendent, and members of his staff
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to discuss other findings of the inspection.
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The violations ir.Section I were discussed.
(Details I, paragraph 3)
The status of previously identified violations in Section II was discussed.
(Details I, paragraphs 2 and 3)
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The new unresolved items in Section III were discussed.
(Details I, paragraph 7.c)
Certain of the previously identified unresolved items in Section IV were discussed.
(Details I, paragraphs 4-6)
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O-RO Rpt. No. 50-313/74-15 I1 DETAILS I Prepared by:
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M. S. Kidd,' Reacto'r Inspector Date Facilities Section Facilities Test and Startup Branch
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Dates of Inspection: November 20-22 and December 10-12, 1974 Reviewed by:
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R.C. Lewis,@eniorReactor/ Inspector Da'te
Facilities Settion i
Facilities Test and Startup Branch 1.
Persons Contacted The following persons were contacted during this inspection:
Arkansas Power and Light Company (AP&L)
J. W. Anderson - Plant Superintendent
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T. C. Baker - Chemical and Radiation Protection Engineer J
T. H. Cogburn - Nuclear Engineer V
R. R. Culp - Test Administrator,
G. H. Miller - Assistant Plant Superintendent J. L. Orlicek - Quality Control Engineer C. N. Shively - Procedure Administrator D. R. Sikes - Results Engineer Two Shift Supervisors
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Two Plant Operators
'Two Assistant Plant Operators
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Babcock and Wilcox Company (B&W)
P. Griffith - Test Coordinator F. Walters - Test Program Coordinator D. Stutzman - Test Coordinator
2.
Halon Fire Control System This previously identified e'i -ceEtnt matter was last discussed in RO Report No. 50-313/74-14, cat ile I, paragraph 2.
A procedure had been written, and was rA ic'.e f ; ecember 4,1974. Licensee personnel stated that retesting,t ci.t
's.lon" system using the new procedure was scheduled for the nek uf DecamL r 16, 1974. This matter remains open.
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Control of Jumpers and Bypasses This matter was initially discussed in RO Report No. 50-313/74-13, i
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Details I, paragraph 2.d. (2), and involved the failure to maintain the station jumper and bypass log in accordance with procedure. A
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licensee response on the subject, dated December 4,1974, has been received and re7iewed. During the current inspection, the inspector reviewed licensee records to verify that corrective and preventive
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measures discussed in the referenced response had been completed.
The review revealed that all previous problems had been corrected; however, one new log sheet exhibited deficiencies identical to certain of those discussed in the referenced RO Report. Log sheet number 68, dated November 22, 1974, did not contain records of authorization for installation of a jumper by the shift supervisor or of verification of proper installation as required by steps 6.12
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and 6.1.4 of procedure 1005.04, " Control and Use of Bypasses and Jumpers." (Revision 2, dated July 6,1974) The jumper was installed by an instrument technician in a control room recorder which monitors turbine governor valve position.
The inspector stated that the omissions appeared to be in violation of 1005.04, and represented a repeat violation. He further stated
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that AP&L should discuss corrective and preventiva measures for all
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groups who are involved in the use of the jumper and bypass log in
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the response to the notice of violation, noting that the referenced
response only discussed the operations group. This matter was dis-cussed during the management interview. Licensee representatives
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defined those corrective actions which would probably be taken. The
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inspector stated that the original enforcement matter would be closed, but that the new violation would remain open pending inspection of j
corrective and preventive measures following the licensee's response.
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4.
Control Room Ventilation System
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This unresolved item was last discussed in RO Report No. 50-313/74 -14, Details I, paragraph 5.
Testing on the normal control room ventilation j
system could not be completed because of leakage through backdraft dampers for each of the supply fans VSF-8A and VSF-8B. During the current inspection, the inspector was informed that the decision had
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been made to replace the dampers rather than repair them. New dampers
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had been ordered, but would not be delivered for several weeks, after which testing would be completed. This item remains open.
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Turbine / Reactor Trip Test The documentation regarding inability to meet certain acceptance criteria in TP 800.14, '*rurbine/ Reactor Trip Test," at forty percent
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power did not fully justify acceptance of the results.
(R0 Report
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No. 50-313/74-14, Details'II, paragraph 2.a) On November 11, 1974, further explanation as to why the test results were acceptable was entered into the test summary. This additional justification was reviewed and accepted by the Test Working Group on November 20, 1974.
Regarding Criterion 8.1.01.002 which required that high pressure j
injection (HPI) not be initiated, it was explained that this referred i
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to automatic initiation of HPI due to low reactor coolant system pressure. The fact that HPI was manually initiated (starting of a makeup pump) due to low pressurizer level did not detract from the test results.
Regarding Criterion 8.1.01.003, which required that pressurizer level remain above 40 inches indicated, the entry noted that the reactor coolant system inventory contracted more than expected due to the fact that main steam reliefs actuated prematurely and blew down longer i
than desired, causing an excessive drop in reactor coolant temperature.
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(See R0 Report No. 50-313/74-12, Details I, paragraph 5) The inspector was informed that the bottom level tap for the pressurizer at ANO,1 is higher than at other B&W plants and that the acceptance criteria for this test at 100% power had been changed to require only that indicated level remain above 0 inches. The inspector confirmed that this procedure change had been made. Licensee personnel were informed that the inspector had no further questions and that the item was considered resolved.
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Rod Reactivity Worth Measurements
.t This unresolved item was initially discussed in R0 Report No.
50-313/74-14, Details II, paragraph 2.b.
It concerned the fact that TP 800.20, " Rod Reactivity Worth Measurements," had not been scheduled to be run in the controlling procedure for power ascension testing (TP 800.01) as required in that the FSAR committed AP&L to perform the test. During the current inspection, the inspector determined that the test has been scheduled in step 10.01.07.11 of TP 800.01 by addendum on November 18, 1974, and was run on December 3,1974, at
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75% power. Rod worths were within acceptance limits. Licensee personnel were informed that the inspector had no further questions and that the item was considered resolved.
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7.
Power Ascension Testing I
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Witness'of Reactivity Coefficients Test The inspector witnessed portions of the testing conducted per
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TP 800.05, " Reactivity Coefficients at Power," at 40% power on November 22, 1974. The test was not cow leted on that date due
i to an unplanned reactor trip.. The test was witnessed on December 10, 1974, while being run at 100% power. These obser-
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vations were made:
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(1) The latest procedure revision (1), including addenda were used.
(2) Minimum staffing requirements of the Technical Specifications (6.2 and 6.3) were met.
t (3) Prerequisites and initial plant conditions specified by
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TP 800.05 and TP 800.01 were met and documented.
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(4) Special test equipment required were in service and calibration
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records were in order. -
'd (5) The test was performed in accordance with the procedure. When it was found that the RCS average temperature (TAVG) could not be reduced 5*F from 579'F as called for by step 7.5.2.6
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because of reaching " BTU Limit" on one of the steam generators,
-the procedure was modified per administrative procedure to i-allow resumption of the test. TAVG was first increased to 581*F and then decreased to 576*F while taking the necessary data.
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(6) Reactimeter data was delogged on December 11, 1974, and detailed coefficient calculations made. Review of the calculations and resultant values revealed that the power doppler coefficient and moderator temperature coefficient were both within acceptance limits of FSAR Section 3.2.2.1.5 (doppler) and Technical Specification 3.1.7 (moderator temperature).
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The inspector stated that he had no comments regarding this test.
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b.
Review of 75% Test Data
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A cursory review of-the various tests conducted at the 75% power plateau was performed to verify that all had been completed and
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RO Rpt. No. 50-313/74-15 I-5 reviewed. Also, it was determined that approval for escalation to 100% power had been given by station management in accordance with PSAR Section 13.4.2.
No comments or questions resulted from this
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revietr.
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c.
Generator Trip Test -
The inspector was informed by licensee representatives that AP&L did not plan to conduct a generator trip or separation test at
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100% power due to the fact that this test would be identical to the turbine trip test, which was completed at 100% power on December 11, 1974. Licensee personnel stated that a generator trip immediately produces a turbine trip and vice versa. This was verified by review of electrical diagrams of the turbine-generator control circuits.
It was noted that a trip of the generator output breakers as recommended by Regulatory Guide 1.68,
"Preoperational and Initial Startup Test Programs for Water-Cooled Power Reactors," (Appendix A, item D.l.1) might produce results different from the turbine trip test, in that the turbine is not tripped directly in this case. AP&L is committed to performing power ascension testing in accordance with the AEC " Guide for the Planning of Initial Startup Programs," dated December 7,1970.
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This guide specifies a " generator trip" in item J.1.j.
The inspector stated that AP&L must perform the test in accordance with its constituent unless relief from the commitment could be obtained through the Directorate of Licensing. He further stated that the item would be carried as unresolved pending completion of the test or other resolution of the commitment.
This position was
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reiterated during the management interview and in subsequent telephone discussions with Region II on December 13, 1974.
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R0 Rpt. No.-50-313,
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DETAILS II Prepared By:
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A. R. Herdt, Hetallurgical Engineer Date Engineering Section
Facilities Construction Branch
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Dates of Inspection: ' November 21-22, 1974 and December 13, 1974
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Reviewed
c L. L. Beratan, Senior Inspector Dhte
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Engineering Section Facilities Construction Branch 1.
Persons Contacted i
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Arkansas Power and Light Company (AP&L)
J. Anderson - Plant Superintendent R. Miller - Assistant Plant Superintendent J. L. Orlicek - QC Engineer C. L. Bean - QA Inspector W. Cavanaugh - Manager Nuclear Services i
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Contractor Organizations (1) Bechtel Corporation (Bechtel)
D. D. Young - Field Welding Engineer (2) Battelle Columbus Laboratories (Battelle)
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W. K. Boyd - Manager, Corrosion Research Section
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W. E. Berry - Associate Manager, Corrosion Research Section 2.
Scope of Inspection The purpose of this inspection was to investigate the actions and measures already taken and proposed because of cracks which developed in the reactor building spray system in the 10-inch schedule 10 stainless steel pipe on the suction side of the pump.
This report covers the inspection made at the reactor site in Russellville, Arkansas, on November 21-22, 1974, and the visit to Battelle-Columbus Laboratories, Columbus, Ohio, on December 13, 1974, to view the
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metallurgical and chemical examinations on a sample of this 10-inch schedule 10 stainless steel pipe.
3.
Reactor Building Spray Line C3 The inspector reviewed the actions taken on this previous unresolved item (74-14/7), concerning the cracks which developed in the 10-inch,
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schedule 10 stainless steel pipe in the reactor building spray i-line..
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50-313/74-14, three cracks, -
As previously reported in RO Report'No.
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two located on the suction side of the "B" pump and one in a
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cross-over leg between two valves, were found in the reactor building spray line. The spool piece test reports identify L
the pipe in all cases as P10 or P15 which it-the same heat l
(No. 800201) of stainless steel material. The-pipe was
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associated with spool pieces -fabricated by M. W. Kellogg Company, J
Williamsport, Pennsylvania. The stainless steel pipe was obtained
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i from Swepco Tube Corporation, Clifton, New Jersey. The licensee
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initiated a program of examination of welds including the heat
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affected zone of the pipe of the heat number above. In addition,
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an increased visual surveillance program was initiated to search for additional leaks. Radiographic examination was performed on a
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random number of field and shop fabricated welds on the reactor building spray and decay heat lines where the pipe from this l
l particular heat is located. Metallurgical examinations cf these
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cracks are being performed by Bechtel and Battelle.
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Radiographic and Weld Record Review
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The inspector reviewed approximately 60 weld radiographs of the Decay Heat Line A (ISO 7DH 5, 6 and 10), Decay Heat Line B I
(ISO 7DH 9 and 11), Reactor Building Spray Line B (ISO 5BS-6)
I and the Reactor Building Spray Line cross-over (ISO SES-7).
-j The welds were both field and shop fabricated. Furthermore,
'j the inspector reviewed the original weld radiographs for the
weld joints where cracks were found as well as the replacement (repaired) weld radiographs.
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Il The radiographs were reviewed for film density, sensitivity, penetrameter requirements, weld quality specifically in the hea*
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affected zones since that is where the cracks have occurred.
The radiographs were shot using IR 192 utilizing a No. 10
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penetrameter and generally in accordance with ANSI B31.7.
Prior to the inspector's arrival at the site, three more welds all j
in the cross-over line were also found visually to be cracked.
i The inspector also reviewed the weld history data cards for all
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I welds found to have cracks to determine whether any correlation
j exists. The material certification review confirmed that the pipe i
was of the same heat. The inspector also examined a sample of
_ Crack No. 2 and noted that the crack was circumferential
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adjacent to the weld in the heat affected zone.
It was also
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noted that the pipe appeared to have been pickled / etched which
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may have occurred during its fabrication.
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RO Rpt. No. 50-313/74-15 II-3
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Within these inspection areas, the inspector found that:
(1) Two field welds, one in each decay heat line, had suspicious areas as viewed in the radiographs in the heat affected zone within the areas that cracking has occurred in other welds.
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(2) There appeared to be a relationship of high heat input and repair areas to the locations of the cracks in five of the six weld failures. The cracks appeared to be within these high heat / repaired areas and the cracks are occurring only in field welds.
(3) Swepco's heat treating procedure stated an air quench as
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the cooling medium for stainless steel which will be pursued by the licensee.
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For item No. 1, _ae licensee agreed to visually examine both welds in question and report their findings to RO.
Metallurgical Examinations - Battelle Columbus Laboratories c.
Battelle is performing for the AEC an independent chemical and metallurgical examination on a sample of a 10-inch schedule 10 (
stainless steel pipe to determine the fracture mode and probable cause of the observed cracking. The chemical examinations include radiochemical analyses of deposits to determine the principal constituents and radioisotopes present and base metal analysis for
comparison vith the mill certificates furnished.
The inspector visited Battelle on December 13, 1974, to review the metallographic examinations performed and discuss the possible cause of the cracking as well as the generic aspects. The crack
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being metallographically and chemically examined is designated as Crack No. 2 which was velded on November 10, 1974, repaired on
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November 11 and visually noticed to fail around November 15..This was the 'second time this area was welded; therefore, welding stresses
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would be present.
Preliminary examinations and results show that thn/ material is
partially sensitized in the center of the base metal, but not on
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the ID or OD surface.
It has also been observed that the base metal i
'
was pickled to a depth of approximately 0.0005 inches. The crack appears to have initiated from the inside and is intergranular
,
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in nature.
Examination by the scanning electron microscope l_
of the crack surfcce shows a stress' corrosion cracking mode and notLfatigue.
Stress most likely was induced by the welding l
l operation.
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, RD Rpt. No. 50-313/74-15 II-4
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During this visit, the inspector examined the metallography
already performed and noted the microstructure. Etching for sigma phase was performed during the visit with essentially none found in the base metal or welded structure. Chemical analysis was being performed as well as microprobe analysis to determine
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whether residual elements are present. Battelle will issue a complete report of their findings and conclusions in the near
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future.
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