Text

Insp Rept 50-313/74-11 on 740625-27 & 0730-0806. Noncompliance Noted:Steam Generator Hydro Test Modified W/O Prior Procedure Mod.Sample Analyses & Test & Calibr Data Documentation Do Not Meet Tech Spec Requirements
	ML19309D846
Person / Time
Site:	Arkansas Nuclear
Issue date:	09/06/1974
From:	Kidd M, Robert Lewis; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML19309D838	List: IR 05000313/1974011; ML19309D837; ML19309D841; ML19309D844;
References
	50-313-74-11, NUDOCS 8004110705
	Download: ML19309D846 (30)
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RO Inspection Report No. 50-313/74-11 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

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Location: Russellville, Arkansas Type of License: B&W, PWR, 2568 Mwt

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Type of Inspection: Routine, Announced Dates of Inspection: June 25-27, July 30 - August 6, 1974

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Dates of Previous Inspection: May 30 - June 1, 1974 Principal Inspector:

M. S. Kidd, Reactor Inspector

Facilities Section Facilities Test and Startup branch

Accompanying Inspectors:

W. D. Kelley, Reactor Inspector Engineering Section Facilities Construction Branch D. J. Burke, Reactor Inspector Nuclear Engineering Section Facilities Test and Startup Branch N. E. Bender, Reactor Inspector Nuclear Engineering Section Facil ties Test and Startup Branch b88 7M Principal Inspector:

ud M. S. Kidd, Reactor Inspector Date

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Facilities Section Facilities Test and Startup Branch Reviewed By:

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M R. C. Lewis, Senior Inspector Date

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. Facilities Section l

Facilities Test and Startup Branch p

8004110

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s RO Rpt. No. 50-313/74-11-2-

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SUMMARY OF FINDINGS I.

Enforcement Action A.

Violations

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1.

The following violation is considered to be of Category II severity:

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Contrary to Technical Specification 6.7.1, the requirements of Caction 7 of 1004.09, " Plan for Preoperational Testing," were not met during the conduct of a hydro test per TP 200.09, " Steam Generator Hydro Test," in that the method of testing was

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modified without first modifying the test procedure.

(Details I, paragraph 9.a)

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An official change to TP 200.09 was approved August 1, 1974,

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endorsing the actual test method used.

2.

The following violations are considered to,be of Category III severity:

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(a) Contrary to the requirements of Table 4.1-3 of the

, Technical Specifications:

(1) A radiochemical analysis for strontium (Sr) 89

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and 90 was not performed on the reactor coolant

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for the month of May 21 - June 21, 1974, as required

by item 1.b.

(Details I, paragraph 4.a. (1))

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(2) An analysis for dissolved oxygen (0 ) was not

performed 5 times per week on the reactor coolant for the weeks of May 20 - June 10, 1974, as required by item 1.c.

(Details I, paragraph 4.a.(2))

(3) Gross beta and gamma activities were not run weekly on the secondary coolant as required by item 5.a prior to the week of June 16, 1974.

(Details I, paragraph 4.a.(3))

(b) Contrary to the requirements of Table 2-2 of the Environmental Technical Specifications:

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(1) Radiochemical analyses for Sr 89 and 90 were not performed on the filtered waste monitor tank, treated g

waste nonitor tank, and laundry drain tank during

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the month of May 21 - June 21, 1974, per item 1.b.

(Details I, paragraph 4.b.(1))

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(2) Iodine spectrum analysis and other analyses of the unit vent particulates, as required by items 3.a and <+.b, were not performed between May 21 and the week of June 16, 1974.

(Details I, paragraphs 4.b.(2) and 4.b.(3))

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(3) No sampling or analysis was made of the reactor building purge vent during the use of this system prior to and during initial fuel loading as required by item 5.

(Details I, paragraph 4.b. (5))

(4) Gross gamma and tTitium analyses were not run on unit vent gas samples as required by item 4.c prior to the week of July 28, 1974.

(Details I, paragraph 4.b.(4))

.(c) Contrary to Criterion V of Appendix B to 10 CFR 50, the following documentation requirements were not fulfilled as required by Section 17 of the ANO Quality Control Program document:

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(1) The data sheets used to record results of test relief valve setpoint checks were not signed off as having been reviewed in the package for TP 200.09, -

(Details I, paragraph 9.c)

(2) The test sununary sheet for TP 266.11, " Service Water System Preop Test," had not been signed by the station test coordinator (STC),

(Details I, para-graph 9.d), and (3) The calibration data sheets for pressure gages used during the testing of a presaurizer relief valve per

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140.'. 03, " Pressurizer Relief Valve Test," had not been signed off as having been reviewed by the STC.

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(Details I, paragraph 9.c)

These records were appropriately signed prior to the conclusion of the inspection.

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B.

Safety Items

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None I

II.

Licensee Action on Previously Identified Enforcement Matters A.

Violations 1.

Calibration of Test Equipment (R0 Report No. 50-313/74-7. I.A)_

A licensee response on this item has been received which

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discusses corrective and preventive measures. This item is closed.

(Details I, paragraph 3)

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R0 Rpt. No. 50-313/74-11-4-

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2.

Conduct of Test Without an Approved Procedure (R0 Report

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No. 50-313/74-9 I.A)

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Not inspected.

3.

Review of' Procedure 1701.01 (R0 Report No. 50-313/74-10 I.A.l.a)

s Not inspected.

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4.

Entries in the Station Log (R0 Report No. 50-313/74-10 I.A.l.c)

Not inspacted.

5.

Documentation of Flushing Results (R0' Report No. 50-313/74-10

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I.A.2.b)

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Not inspected.

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Safety Items

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Not applicable.

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III. New Unresolved Items

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74-11/1 Control Room Ventilation System l

Preoperational test results of the control room ventilation system do not demonstrate whether the system meets the design basis discussed in the FSAR.

(Details I, paragraph

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9.1.(10))

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J 74-11/2 Hydrogen Purge System

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The discharge filter bank for train 3 of the hydrogen purge system was flooded during testing and is conse-

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quantly not qualified for operational purposes. A new filter must be installed and qualified or the present one requalified prior to August 23, 1974. *(Details I, paragraph 9.1.(11))

IV.

Status of Previously Identified Unresolved Items 72-12/2 Valve Wall Thickness Verification Program O

The licensee's valve wall thickness verification program

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has been documented and submitted to Region II for review.

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This item is closed.

(Details II, paragraph 2)

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RO Rpt. No. 50-313/74-11-5-

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74-4/1 Leak in P36B Recirculation Line

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Orifices of a new design have been installed in the recirculation lines for all three makeup pumps. This j

item.is closed.

(Details I, paragraph 5)

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74-6/2 Sensitivity of Plant Leak Detection System

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A ratest of the ability,to determine the magnitude of system leakage has verified the sensitivity of plant

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equipment. This item is resolved.

(Details I, paragraph 6)

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74-7/1 Preoperational' (Baseline) Inspection Data

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A revised Baseline Inspection Report by the licensee was

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submitted to Region II for review. All problems previously discussed were corrected in that report.

This item is closed.

(Details I, paragraph 7)

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74-7/2 Mechanized Ultrasonic Data Verification and Repeatability i

The final Baseline Inspection Report resolved all questions

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in this area. This item is closed.

(Details I, paragraph 8)

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74-9/1 System Generator Hydrotest

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j This hydro has been rerun for both steam generator

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secondary sides. This item is closed.

(Details I,

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paragraph 9.a)

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74-9/2 Checkout of Reactor Coolant Pumps (RCP)

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The RCP's were tested subsequent to fuel loading to

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assure all instrumentation was functional. This item is closed.

(Details I, paragraph 9.b)

74-9/3 Pressurizer Code Safety Valve a

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The setpoint of the safety valve whose seat was replaced r

had been rechecked.

This item is closed.

(Details I,

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paragraph 9.c)

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RO Rpt. No. 50-313/74-11-6-

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74-9/4 Service Water Isolation valves l

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l Air accumulators have been installed on the service water l

isolation valves for the reactor building coolers and

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i successfully tested. This item is closed.

(Details I,

paragraph 9.d)

74-9/5 Main Steam Piping Restraints l

Shims have been instalred on the main steam headers to prevent undue deflections. This item is closed.

(Details j

I, paragraph 9.e)

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l 74-9/6 Gracina over Steam Generators

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Seismic Class I tie-downs and storage racks have been

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provided for the grating.

This item is closed.

(Details I, paragraph 9.f)

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74-9/7 Quench Tank Rupture Disk Deflector A deflector has been installed over the pressurizer quench tank rupture disk to protect adjacent instru-mentation in the event of rupture. This item is closed.

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(Details I, paragraph 9.g)

74-9/8 DC Panel Transfer Switches

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The manual transfer switches for panels D11 and D12 have been replaced with a more reliable type switch. The new j

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r switches have been successfully tested. This item is closed.

(Deta".ls I, paragraph 9.h)

74-9/9 Preoperational Test Deficiencies All deficiencies affecting initial criticality in the tests given in RO Report No. 50-313/74-9, Details I, paragraph 2.b.(9), have been resolved with the exception of the Hydrogen Purge System Test, TP 370.01.

This item will be handled as a new unresolved item. All other aspects of this item are considered resolved. (Details I, paragraph 9.1)

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RQ Rpt. No. 50-313/74-11-7-(

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74-10/1 Control Room Staffing

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The Administrative Controls Manual was revised to clarify the portion of the control room which is to be occupied at all times by a licensed operator when fuel is in the reactor vessel.

The control room was defined as the area in front of the vertical panels. This item is closed.

(Details I, paragraph 10)

V.

Unusual Occurrences

A.

Crack in Decay Heat Piping

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On June 26, 1974, the licensee reported that a leak had been

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i discovered in the decay heat pump P34A discharge line. The line had cracked adjacent to the coupling weld for a vent valve, DH-1010. An abnormal cecurrence report dated July 5, i

1974, attributes the cause of cracking to excessive heating during welding. No visible signs of overheating were found in similar welds in the system. The section of pipe containing the crack was removed and another section welded in.

B.

Reactor Building Purge Isolation Valve Failure On July 8, 1974, the licensee reported an abnormal occurrence involving a reactor building purge system isolation valve which failed in mid-position during a surveillance stroking test. A written report dated July 16, 1974, states that one of the bearings in the valve operator had frozen due to rusting and lack of lubrication,. caused by damage to an 0-ring seal and subsequent moisture intrusion.

The valve operator was disassembled, cleaned, reassembled, lubricated and stroke i

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tested. The redundant valve was successfully stroked at the t-time of the occurrence.

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C.

Makeup Pump Bearing Failure On July 7, 1974, the licensee reported an abnormal occurrence involving the failure of the-radial bearing on makeup pump P-

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l 36A. Mating pins on the bearing had become dislodged, allowing j

the bearing to slide on shaft, cutting off the oil supply

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to the bearing. A written report on the matter, dated July 16, i

1974, gives corrective actions on all three makeup pumps.

D.

Valve-Leaks On July.12, 1974, AP&L reported that leaks had been observed-

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in,several 3-inch, stainless steel cast valves. Apparently r"%

the leaks were caused by casting flaws. A written abnormal

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RO Rpt. No. 50-313/74-11-8-

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states that 6 valves with defects had been found. An investi-gation of all of some 65 valves ordered on the same reqdisition as these ones was initiated to determine the status of all other valves. The six with known defects were being repaired or replaced. A seventh valve was identified as being defective following the July 12, 1974, report. This valve and additional information regarding findings of the investigation will be discussed in a supplemental report.

E.

Source Range Neutron Detector

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On July 14, 1974, source range (SR) detector NI-1 failed due to excessive heating of the cable between the detector and preamplifier. NI-l cabling is run through the same junction box as NI-5, one of the power range (PR) detectors. The

cables for NI-5 showed some deterioration upon inspection.

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All other detector cables showed no adverse effects. Air flow was increased up around these junction boxes located in the

reactor vessel annulus cavity by placing asbestos material

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over the openings around the vessel cold leg penetrations in

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the primary shield and lifting the seal ring which rests on i

the top of the cavity. This matter will be reported by August 14, I

1974, as an unusual event per Technical Specifications.

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VI.

Other Significant Findings A.

Initial Criticality

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The approach to initial criticality was observed August 5-6, 1974. Criticality was achieved at 0602 hours0.00697 days 0.167 hours 9.953704e-4 weeks 2.29061e-4 months (CDT) August 6, 1974, within the rod position and boron concentration ranges

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predicted.. The inspectors had no questions or comments on this activity.

(Details I, paragraph 2 and Details III, paragraph 2)

B.

Regulatory Operation Bulletin (ROB) 74-6

In response to this R0B which requested information regarding

the use of Westinghouse Type W-2 control switchen, a licensee

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letter of July 1, 1974, stated that none of these switches are instclied in safety systems in ANO, Unit 1.

VII. Management Interview A management interview was conducted with N. A. Moore, Manager, Quality Assurance, on June 26, 1974, to discuss the findings of the

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thin-walled valve program inspection.

(Details II, paragraph 2)

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RO Rpt. No. 50-313/74-11-9-

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Management interviews were conducted June 27, 1974, and August 6, 1974, to discuss other findings of the inspection. The following licensee representatives attended:

Arkansas Power and Light Company (AP&L)

J. W. Anderson * - Plant Superintendent T. C. Baker - Chemical and Radiation Protection Engineer C. A. Bean - QA Inspector, Welding T. H. Cogburn - Nuclear Engineer R. R. Culp - Test Administrator C. A. Halbert - Technical Support Engineer G. H. Miller - Assistant Plant Superintendent J. L. OrJicek - Quality Control Engineer B. A. Terwilliger - Operations Supervisor The violations in Section I were discussed.

(Details I, paragraphs 4, 9.a 9.c and 9.d)

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The status of previously identified enforcement matters in Section l

g II were discussed.

(Details I, paragraph 3)

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The new unresolved items in Section III were discussed.

(Details I, paragraphs 9.1. (10) and 9.1. (11))

The status of old unresolved items in Section IV was discussed.

The inspector stated that all of these items were considered resolved.

(Details I, paragraphs 5 - 10 and Details II, paragraph 2)

Attended the August 6, 1974, meeting only.

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V RO Rpt. No. 50-313/74-11 I-1

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~1 NO DETAILS I Prepared By:

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M. S. Kidd,' Reactor Inspector Date Eacilities Section Facilities Test and Startup Branch Dates of Inspection: June 25-27 and July 30 - August 6,1974

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Reviewed By: [ de

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R. C. Lewis, Senior Inspector Date Facilities Section Eacilities Test and Startup Branch 1.

Persons Contacted In addition to those listed in the 'anagement Interview section, these M

individuals were contacted during the inspection:

Arkansas Power and Light Company (AP&L)

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W. Cavanaugh, III - Manager of Nuclear Services

's, P. A. Almond - Reactor Technician

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C. N. Shively - Procedure Administrator D. R. Sikes - Results Engineer Three Shift Supervisors Three Plant Operators

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Four Assistant Plant Operators 2.

Witness of Initial Criticality The approach to criticality was conducted on August 5-6, 1974, and was witnessed by the inspector.

Instructions for this activity were provided by TP 710.01, "Zero Power Physics Test," which had been

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previously reviewed and discussed with licensee personnel. (See paragraph 10.a) The RCS boron concentration at. the start of the approach to criticality was 1846 parts per million (ppm). At 1840 hours0.0213 days 0.511 hours 0.00304 weeks 7.0012e-4 months on August 5,1974, the operator in charge started pulling safety control rod group 1.

At 2107 hours0.0244 days 0.585 hours 0.00348 weeks 8.017135e-4 months , groups 1-6, and 8 had been com-pletely withdrawn and group 7 had been 75 percent withdrawn per procedure. At 2130 hours0.0247 days 0.592 hours 0.00352 weeks 8.10465e-4 months , boron dilution was started with a demineralized water flow to the makeup tank of 20 GPM. This resulted in a dilution rate of approximately 30 ppm boron per hour. Lt 0602 hours0.00697 days 0.167 hours 9.953704e-4 weeks 2.29061e-4 months (CDT) criticality was achieved at a boron concentration of 1603 ppm. The predicted value was 1635 plus or minus 50 ppm.

The following aspects of the evolution were witnessed by this inspector. Other

'~'g inspection activities are discussed in Details III, paragraph 2.

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RO Rpt. No. 50-313/74-11 I-2

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The latest revision of TP 710.01, including approved addenda, was used in the control room during the approach.

It was followed and signoffs were made at proper times.

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b.

Staffing in the control room and for each entire shift was in conformance with Technical Specifications 6.2.1, 6.2.2.1, 6.2.2.2, 6.2.2.3, and 6.2.2.5.

These specifications cover normal crew staffing (including health physics coverage) and special staffing

required for startup.

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c.

Crew working times were held to a maximum of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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Two shift turnovers were witnessed. These were performed in

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accordance with. the Administrative Controls Manual (1005.01),

Section 6.1.4.

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Data was analyzed in a timely fashion, including inverse multiplication plots versus boron concentration and deboration

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time. These plots were quite linear and predicted criticality well.

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Criticality was achieved at 1603 ppm boron versus predictions of i

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1625 plus or minus 50 ppm in TP 710.01 and 1635 ppm per OP 1103.15,

'.' Reactivity Balance Calculation."

l The inspector had no adverse consnents regarding the operation and noted that it was conducted according to the procedure and that re-suits were well predicted.

3.

Calibration' of Test Equipment As discussed in RO Report No. 50-313/74-7, Details I, paragraph 3, a quality assurance calibration procedure for test equipment on the mechanized ultrasonic inspection apparatus was not followed. The

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licensee response to the notice of violation on this matter, dated May 30, 1974, discussed this question and ascertained that although

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calibration dates were not placed on calibration stickers as required, the instruments in question were qualified.

Certification of all

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instrumentation in question was included in the final Baseline Inspect-i-

ion Report. The inspector informed licensee personnel that this matter was considered resolved.

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i 4. ' Technical Specification Surveillance Requirements

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On June 18,.1974, licensee personnel informed the inspector by

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telephone that certain surveillance requirements had not been com-

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RO Rpt. No. 50-313/74-11 I-3

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pleted when they apparently should have been. These involved certain

chemical and radiochemical analyses required by Table 4.1-3 of the Technical Specifications (Appendix A to the operating license) and Table 2.2 of.the Environmental Technical Specifications (Appendix B to the operating license). Discussions with plant personnel at the site revealed the following information:

a.

Table 4.1-3 - Minimum Sampling Frequency (1) An analysis for Sr 89 and 90 was not performed on a reactor coolant sample during the month after the issuance of the operating license (May 21, 1974) as required by item 1.b.

A sample was taken, but no analysis was made because plant personnel had been unable to obtain the necessary Sr 89 and

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90 sources to calibrate their. equipment.

Licensee. personnel stated that the sources had been on order for several months but had not been obtained. The analyses for subsequent samples were run in the AP&L offices in Little Rock and will be run there until plant equipment is available.

(2) Analyses for 02 were not run on RCS samples five times per week as required by item 1.e from May 21, 1974, through the

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week of June 9, 1974. Licensee personnel felt that this analysis was not required since the RCS was open to atmosphere during most of the time period, with no ability to control dissolved 0. This analysis was started the week of

June 16, 1974, after the RCS had been closed up after fuel

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loading and has been completed as required since.

(3) Gross beta and gamma activities were not run weekly on the

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secondary coolant ac required by item 5.a prior to the week

of June 16, 1974. Licensee personnel felt that this was unnecessary in that the steam generators were in a wet layup condition and no activity was present in that the reactor had not been taken critical. The analysis was started the week of June 16, 1974, and run on schedule since.

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b.

Table 2.2 - Minimum Sampling Frequency (1) No analysis for Sr 89 and 90 was performed on the filtered waste monitor tank, treated waste monitor tank, and laundry drain tank during the month following May 21, 1974, as re-quired by item 1.b.

The reason for this is the same as discussed in 4.a. (1) above. Analyses for samples taken in June and July were run in Little Rock.

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RO Rpt. No. 50-313/74-11 I-4

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(2) An iodine spectrum analysis was not completed weekly on the unit vent sample per item 3.a from May 21 through the week of July 21, 1974. Up until the latter date, no capability existed to gather a sample suitable for iodine analysis.

(3) The analyses required of unit vent particulates by item 4.b

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were not run prior to the week of June 16, 1974.

(4) Gross gamma and tritium analyses were not run on unit vent gas samples as required by item 4.c prior to the week of July 28, 1974, because of problems in equipment.

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(5) Samples of the reactor building purge vent were not taken each time the system was used as required by item 5 during initial fuel loading. At that time, no capability existed to gather a sample. The system had not been used since initial fuel loading according to licensee personnel, but that sampling capability has been added such that it can be sampled during i

each purge in the future.

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The inspector stated that all of these matters appeared to be violations (j of the Technical Specification requirements. He stated that the licensee

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has the responsibility and opportunity to request changes in requirements which he feel are not meaningful and/or which cannot be met through the Directorate of Licensing.

5.

Leak in P36B Recirculation Line This unresolved item was the subject of a licensee report per 10 CFR 50.55(e) dated April 13, 1974, titled " Makeup Pump Orifice Leak."

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The wall of the orifice was eroded due to high exit flow from the

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last stage of the orifice. Also, the last stage exit was located eccentrically in the orifice shell such that its discharge was near one side of the orifice wall.

Corrective action included modifying the orifice to reduce the exit velocity and to locate the last stage exit concentrically.

During the current inspection, hydrostatic test records for the new orifices for all three of the makeup pumps were reviewed. Testing was accomplished through use of a modified hydro procedure for the system. Test records revealed all orifices and welds were success-fully hydroed at 3815 psig (1.25 times design) for a minimum of ten minutes per code requirements. Calibration records for pressure

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test gauges used for the tests were in order. The test results were endorsed by AP&L June' 25,1974. This item is considered closed.

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RO-Rpt..No. 50-313/74-11 I-5

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6.

Sensitivity of Leak Detection System This unresolved item was initially discussed in R0 Report' No.

50-313/74-6, Details I, paragraph 3.f.

The results of HFI test TP 600.10, "RCS Hot Leakage Test," did not appear to confirm the sensitivity of the inventory balance method to detect leakage as described in Section 4.2.8.3.b of the Unit 1 FSAR. The test was run to determine the accuracy of equipment in measuring a known leak of 1 gallon per minute (GPM) and results after 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months showed an accur-acy of only about 25 to 30 percent.

On July 14, 1974, a similar known leak test was conducted using i

OP 1103.13, to calculate leak rates before and during the test.

Re-suits of that test showed that for stable system conditions, a leak

'

of 1 GPM can be determined in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months with an accuracy of about 5 to 10 percent. This item is considered closed.

.

,, 7.

Preoperational (Baseline) Inspection Data This unresolved item was initially discussed in RO. Report No.

50-313/74-7, Details I, paragraph 2.b.

On June 21, 1974, RO:II b

received a copy of the revised Baseline Inspection Report from AP&L.

-

A review of this report by RO:II personnel revealed that the problem areas in the referenced RO report had been resolved. This item is

.

considered closed.

8. ' Mechanized Utrasonic Data Verification and Repeatability The question of whether the data generated by the mechanized remote

utrasonic tool was repeatable was discussed in RO Report No. 50-313/74-7, t

Details I, paragraph 2.c.

The licensee included data which had been rerun for those weld areas which had ultrasonic indications along with the original data in the Baseline report. This item is considered

'

resolved.

,

9.

Items to be Completed Prior to Initial Criticality RO Report No. 50-313/74-9, Details I, paragraph 2.b, discussed nine

,

areas of concern which had not been completed or where deficiencies t

existed in previous testing at the time of the issuance of an operat-ing license for Unit 1.

A letter of commitment to resolve these and

'

certain other problems prior to initial criticality was submitted by

AP&L (Phillips to Giambusso) to DL on May 20, 1974.

"~ne resolutions

of these items are discussed here in the same sequence as listed in

'

the referenced letter of conunitment.

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RO Rpt. No. 50-313/74~11 I-6

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a.

Steam Generator Hydro l

The additica nf isolation valves upstream of the atmo' spheric l

dump valves ror each main steam header necessitated rehydroing l

the secondary side of each SG.

This testing was accomplished using TP 200.09, the original test procedure. During the conduct

'

of each rehydro, the atmospheric dump valves CV-2618 and CV-2668 both leaked. The new isolation valves, MS-44 and MS-45 were closed such that the test could proceed. The piping between l

MS-44 and CV-2618.ad 115-45 and CV-2668 were tested separately

'

on a later date.

l The inspector noted that the testing accomplished at the later l

date was described in a writeup attached to the procedure. This writeup had not been approved before the change in testing nor I

addended to the procedure using normal methods. The inspector stated that this appeared to be in violation of section 7 of 1004.09, " Plan for Preoperational Testing," in that the change to

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the proced tre had not been approved prior to use. Licensee representatives agreed that the change should have been approved

.

prior to use and the subject writeup was officially addended to

!

the procedure on August 1,1974, by approval of the plant super-

-

intendent.

The test package contained two data sheets used to record the data from checkouts of the two relief valves used during the test.

The data sheets had not been signed off as having been reviewed by the station test coordinator (STC). The inspector stated that this omission appeared to be in violation of section 17 of the ANO Quality Control Program Document.

i

l The data sheets were signed on August 1,1974, with a note that the sheets had been reviewed on the proper date (June 15, 1974).

Unresolved item 74-9/1 is considered closed.

b.

Checkout of.RCP's l

Numerous instrumentation problems were encountered on the RCP's during HFT. Also, quick disconnects for cables were replaced with terminal boxes to improve leak-tightness of the connections.

The instrumentation problems were verified to be resolved per l

TP 200.06, "RC Pump and Motor Initial Operation," af ter fuel-loading. The test results were endorsed August 1, 1974. This

item is considered resolved.

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Pressurizer Code Safety Valve The setpoint of valve PSV-1001 was rechecked because'the seat had been replaced due to erosion by steam when it failed to reseat completely during the original setpoint check during hot funct-

,

ional testing (HFT). The test was conducted per surveillance teet procedure 1401.03, " Pressurizer Relief Valve Test."

The setpoint was found to be within the tolerance allowed by ASNE Section III (design 1%) on two successive lifts. This item is considered resolved.

In reviewing the calibration records for the test gauges used for

.

this test, the inspector noted that the data sheets had not been

'

signed off by the supervisor responsible for reviewing them.

Licensee personnel were informed that this appeared to be in violation of the ANO Quality, Control Program document.

i The date sheets were signed prior to the conclusion of the

-'

'

inspection.

d.

Service Water Isolation Valves

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~ Air accumulators were installed for service water system valves

-

CV-3812, 3813, 3814, and 3815, such that they would remain c.1 3 sed in the event of loss of instrument air concurrent with a high

-

radiation signal. A test involving loss-of-air on these valves l

'

was accomplished per TP 266.11, " Service Water System Preop Test,"

'

on July 3, 1974. The test was endorsed on July 11, 1974. This item is resolved.

.

, -

In reviewing this test package, the inspector noted that the station test coordinator (STC) had not signed the test summary sheet as required. Licensee personnel were informed that this

'

appeared to be.in violation of the ANO Quality Control Program document.

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The summary sheet was signed prior to the conclusion of the in-

,

spection.

e.

Main Steam Piping Restraints Shims were installed on the main steam piping restraints in the auxiliary. building due to deflections observed during HFT. The

-

shims were inspected in the field and a Field Inspection Report (FIR) by Bechtel personnel who witnessed the installation

[

(July 30, 1974) was reviewed, with no comments re sulting.

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This item is considered resolved.

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f.

Crating over Steam Generators An FIR dated June 24, 1974, for this job was reviewed. The FIR indicated that Bechtel had performed visual as well as liquid Obser'ation in penetrant examination of these storage racks.

v the reactor building revealed that the grating had been properly stored. This item is considered resolved.

g.

Quench Tank Rupture Disk Deflector This deflector was installed to protect instrumentation lines overhead in the event of a rupture. An FIR written June 13, 1974, stated that the deflector was installed per drawing with one minor exception necessitated by insufficient clearance. Observation in the field verified the installation. This item is resolved.

h.

DC Panel Transfer Switches The new manual transfer switches for panels Dll and D12 were

successfully tested per TP 400.02, "D-C Power Systems,"

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June 20, 1974, and the test results were endorsed June 25, 1974.

This item is-c1csed.

1.

Preoperational Test Deficiencies

'

.The preoperational test packages listed in RO Report No.

50-313/74-9, Details I, paragraph 2.b.(9), were not fully endorsed prior to licensing of Unit 1.. The deficiencies in the results which precluded endorsement were to be resolved prior to initial criticality where applicable. The following represents the status of these items as of August 3,1974:

.

(1) TP 165.01

" Filter Tests" The deficiencies for this test involved receipt of the final filter test report from the subcont actor and the fact thac certain plant filters were efficiency tested at flows below rated values. The final test report was received June 4,1974, and was included in the test package. A letter of justification for the test results at lower flow rates was also included.

In the case of the gaseous waste discharge header filter, the flow obtained was 3 standard cubic feet per minute (SCFM) versus a design valve of 11 SCFM given in Table 11-6 of the FSAR. The

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R0 Rpt. No. 50-313/74-11 I-9

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inspector stated that even though the lower flow was justified and adequate, the FSAR should be revised to reflect the as-built

system. A licensee representative stated that the FSAR would

,

be revised.

The test results were fully endorsed by AP&L on July 13, 1974.

,

(2) TP 172.01

" Penetration Room Ventilation Test"

.

(a) The filter bleed flows for each filter train were less than called for in the FSAR.

(40 CFM and 45 CFM versus 130 CFM)

The inspector asked why this lower value was acceptable. A j

letter of explanation from Bechtel was received August 3, 1974, which stated that the heat load generated by a maximum filter leading of iodine (34 BTU per hour) could be easily

'

removed with a flow of 40 CFM.

It was noted that the applicable piping and instrument drawing (P&ID) had been revised, but the old value of 130 CFM was still reflected in

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Figure 6-10 of the FSAR. Licensee personnel stated that the FSAR would be revised.

(b) Individual penetration room air flows and flows for the

'

two train headers were slightly different than FSAR values.

The total measured flow for both trains was 1850 CM versus the expected value of 2000 CFM. The P&ID for the system was modified to show as-built values, in that adequate vacuums can be maintained. The inspector noted that Figure 6-10 of the FSAR still showed individual room flows and total flow of 2000 CE.

Licensee personnel stated that the FSAR would be revised.

(c) Floor drains leaked air into the rooms, preventing the necessary vacuum drawing capability. Traps were installed in the drains and a retest successfully accomplished July 25, 1974.

!,

(d) The final filter test report review was completed

,

May 30, 1974.

-

l The results of this test were endorsed August 1,1974.

(3) TP 201.03

" Core Flood System Functional Test"

.

Flow paths to the gaseous waste header and clean liquid radwaste tanks were verified May 30, 1974. This resolved

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all deficiencies in this test, which was endorsed June 22, 1974.

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(4) TP 231.64

" Dirty Radwaste System Acceptance Test" The problems with this test dealt with the compl$ tion of certain portions of the procedure af ter modifications and the evaluation of tank level alarm setpoint changes. Retest-ing was-accomplished June 4,1974, and a letter of justificat-ion for changes in the level alarms was provided by Bechtel on August 2,1974, after questions were raised by the inspector regarding the changes. The test results were endorsed July 23,1974.

,

(5) TP 234.01

" Resin Sluicing Test" Test results showed that level instrumentation for the spent resin storage tank was inoperable and that contents of the tank could not be recirculated due to pluggage of a screen in the transfer pump suction line. Licensee personnel stated that these items were currently being corrected.

The inspector i

stated that the spent resin system must be operable by October 1, 1974, per FSAR Section 11.

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(6) TP 240.14

" Intermediate Cooling Water System Preop Test"

.

(a) The ICW pump shutoff heads were tested to be 96 psig versus the design value of 85 psig given in the procedure.

-

A Bechtel letter of May 15, 1974, stated that the design value in the procedure should have been 95 plus or minus 5 instead of 85 plus or minus 5.

.

(b) Modifications to certain valve logics required retesting.

The ratesting was completed June 18, 1974.

(c) The cooling water flow to the instrument air compressors was insufficient. The associated piping was replaced

>

and this test was successfully rerun July 22, 1974.

(d) Two booster pumps were added to this system to provide additional cooling for the RCP seals. The pump capacities are satisfactory for their intended function but pump j

curves are not yet available to compare test data to

'

verify that they meet design flow. This item remains outstanding but is not considered a restraint to criticality.

.

The inspector noted that the addition of these pumps (P 114A and B) had not been reflected in the FSAR. Licensee

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personnel stated that a change would be made to show this N

addition.

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(7) TP 273.36

" Auxiliary and Emergency Feedwater Test" (a) During the first test per this procedure, ceitain alarm

'

functions were not operable. These were successfully tested May 28, 1974.

(b) It was found that for the turbine driven emergency feed-water pump P7A, that if the turbine speed is set to give the design flow, the low flow pressure is excessive. An engineering evaluation (Bechtel) of May 8,1974, agreed with AP&L's recommendation to set the speed at a higher value.

At design conditions, the heat removal capacity will be reduced from 5 percent to 4 percent, whereas, the FSAR re-quires only a 3.5 percent decay heat capacity. The test results were endorsed August 4, 1974.

(8) TP 276.44

" Condensate System Acceptance Test"

(a) Deficiencies relating to apparent discrepancies between measured and predicted dead-head pressures for the

'g condensate pumps were resolved when it was recognized that J

the condenser vacuum level had to be taken into account

'

in computing the pressure.

(b) Certain sections of the test had to be rerun after instrumentation problems were corrected. Retesting was completed June 6,1974.

All deficiencies imposing a restraint to criticality were resolved as of June 11, 1974.

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(9) TP 330.04

"CRD Integrated Test" Several in-limit switches for the CRDM's had malfunctioned during HFT. These were repaired per Job Order 207 on June 26, 1974. The test was fully endorsed June 29, 1974.

(10) TP 351.30

" Computer and Control Room Air Conditioning" (a) The emergency att reservoirs which hold the normal ventilation system dampers closed for isolation purposes had not been installed as of May 17, 1974.

These reservoirs were installed and successfully tested to hold the dampers

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closed for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a loss of instrument air subse-

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quent to fuel loading.

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RO Rpt. No. 313/74-11 I-12

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(b) Backup fan 2VUC-9 had not been released from construction as of May 17, 1974.

It was released and tested the week of May 20, 1974.

(c) The air balance data for the various fans did not meet predicted values in certain cases. This problem had not been fully resolved as of the current inspection. AP&L was awaiting evaluation by Bechtel Engineering. The inspector noted that the results of this test did not demonstrate whether the normal and emergency ventilation systems met the qualifications and design bases stated in Section 9.7 of the FSAR. The FSAR speaks in terms of maintaining maximum temperature and humidity limits dur-ing normal and emergency conditions, whereas the test, in acceptance criteria 7.10, emphasizes air flows.

The inspector asked licensee representatives how they would show that the systems do indeed meet the FSAR de-

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sign bases. They stated that either the data already gathered would be correlated to values.given in the FSAR

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or the test would be rerun during power ascension testing A _,/

to verify the FSAR descriptions. The inspector stated s

-

that this item would be carried as unresolved.

(d) The inspector asked if a test had been run to demonstrate that the standby emergency fan (VSF-9, 2-VSF-9) will start automatically when the ficw on the lead fan drops to -90 percent of normal as discussed on page 9-30a of the FSAR.

Licensee personnel stated that the design had been changed and the auto start eliminated. They further stated that

,

the FSAR description regarding this function would be changed.

(11) TP 370.01

" Reactor Building Hydrogen Removal System" (a) The alarm setpoints for differential pressures across the system filters were incorrect during the original test. After recalibration, these were satisfactorily retested on July 30, 1974.

(b) The design flow rate of 50 CFM for each train was not met during the original test.

Internals were removed from check valves downstream of the exhaust fans via design change and the proper flows were demonstrated

during retesting on July 30, 1974.

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(c) The exhaust filter bank for train B was inadvertently flooded with water during further testing after efficiency testing, thus rendering it unqualified.

In that the licensee would not be abla to purchase a replacement set of filters prior to the scheduled date for criticality, a request was made to RO and DL to proceed with criticality and power ascension tecting with the one qualified train.

R0 and DL concurred with this approach and a letter of August 2,1974, from Moore (DL) to Phillips states that AP&L may proceed with criticality and subsequent testing provided that train A remain operable and the replacement filter bank or the flooded bank is qualified for service by August 23, 1974. The qualification of either set of filters by that date will be carried as an unresolved item.

(12') TP 500.03

" Initial Radiochemistry Test"

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The only exception to this test as of May 17. 1974, was sampling and analyzing the fuel transfer canal during initial fuel loading. These samples were taken May 25-27, 1974, and results demonstrated typical background radioactivity levels.

-

(13) TP 600.13

" Pressurizer Operation and Spray Test" (a) Portions of the procedure were rerun, with the last rerun being completed August 3,1974.

.

(b) During HFT, it was found that the heater banks used to offset ambient heat losses were insufficient due to the excessive losses through the insulation of the pressurizer.

,

The insulation was adjusted, but the heaters still were insufficient. A heater bank out of a backup group was rewired such that it would be available for use in off-setting ambient losses and normal pressure control if needed by the operator.

The test results were endorsed August 4, 1974.

l (14) TP 600.14

" Pipe / Components Hanger Hot Deflection Test" This test had not been endorsed as of May 17,1974, due to variances in measured and predicted pipe movements. All these variances were explained in a letter from Bechtel

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monitored when they reach normal operating temperature during low power physics testing. The test was interim endorsed on August 1, 1974.

(15) TP 600.17

"CRD System Operation Test" Several problems were encountered with the relative and absolute position indicators and zero percent withdrawn switches for the CRDM's during HFT.

(See RO Report No.

50-313/74-4, Details I, paragraph 5). Additional cooling air was supplied to the vessel head service structure for the CRDM's.

A significant improvement was noted during the conduct of TP 330.05, "CRD Trip Test," following initial fuel loading. The test received final endorsement July 23, 1974.

10.

Control Room Staffing This unresolved item was initially discussed in R0 Report No.

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50-313/74-10, Details I, paragraph 2.e, and involved clarification of the portion of the control room which must be occupied at all times by a licensed operator when fuel is in the reactor vessel as required N.

by Technical Specification 6.2.2.1.

Section 6.1.3. A of the ANO Administrative Controls Manual was revised August 5, ~1974, to define the control room to be in area in front of the vertical panels for this purpose (see FSAR Figure 7-25). The inspector observed that this area was manned by at least one licensed operator at all times

' during inspection activities in the control room for this inspection.

This item is considered resolved.

11.

Review of Zero Powec Physics and Power Ascensicn Test Procedure The zero power physics and power ascension test programs for ANO-1 are described in Section 13 of the FSAR. A commitment is made on page 13-1 to preparing these programs in accordance with the AEC publication " Guide' for the 3'anning of Initial Startup Programs,"

with exceptions listed in Tabi 13-4.

This guide was superseded by Regulatory Guide 1.68, "Precperational and Initial Startup Test Programs for Water-Cooled Power Reactors," which is almost identical in its description of the low power physics and startup test programs.

These documents, along with ANSI 18.7, " Standard for Administrative Controls for Nuclear Power Plants," (Draf t,1972), the FSAR and Technical Specifications, and applicable station procedures were used as a basis for review of the following test procedures. The review

was made to determine that general content of the procedures was in accordance with RG 1.68 and ANSI 18.7 and to determine that specific testing commitments in the FSAR would be met.

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The following tests, representing 50% of the Category I (more important)

tests were reviewed and discussed with licensee representatives:

a.

TP 710.01, "Zero Power Physics Test" This procedure provides instructions for attaining initial criticality and for the performance of zero power physics testing. The inspector had no comments relating to the general content and format of the procedure, s

Initial Criticality (a) The inspector noted that the procedure did not speak to the source range instrumentation' signal to noise ratio as recommend-

,

ed by RG 1.68, Appendix C, paragraph C.

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(b) The inspector asked if the manual scram shouldn't be tested just before startup as required by the normal startup procedure

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1102.02.

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These comments were resolved by the addition of provisions to check

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the signal-to-noise ratio and test the manual scram in TP 710.01 on July 31, 1974.

b.

TP 800.11. " Core Power Distribution" This procedure provides for the collection of various core parameter data such as power distributions, peaking factors, linear heat rates, l

and departure from nucleate boiling ratio (DNBR), implied by item i

D.1.g. of Appendix A of RG 1.68.

The inspector had no comments on

-

this procedure.

c.

TP 402.01, " Remote Shutdown"

'

f This test will demonstrate that the reactor can be safely brcunh'.

to a shutdown condition using controls outside the main control room. Emergency operating procedure 1202.33, of the same title, is used as the body of the test procedure in order to verify its

,

i feasibility. The test will be run at about 15% power in accordance with item D.1.j. of Appendix A of RG 1.68.

(1) The inspector noted that no spaces had been provided for initials on the data sheet in Enclosure 1.

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A licensee representative stated that provisions would be

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made for documenting the identity of the data recorder.

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(2) The inspector asked why the reactor was not tripped from the control room during this test.

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A licensee representative stated that this would be done whenever possible during a true emergency, but that for this test, the ability to trip the reactor from the computer room by tripping the CRD breakers was to be verified.

The inspector agreed with this approach.

(3) The inspector noted that the procedure did not specify that

the calibration of all applicable instrumentation for the test be checked.

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A licensee representative indicated that provisions for

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assuring that a3'. calibrations were up-to-date would be added to the procedure,

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d.

TP 800.05, " Reactivity Coefficients at Power"

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This procedure provides instructions for measuring the power

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reactivity coefficient per item D.l.b. ^f Appendix A of RG 1.68.

The inspector had no questions regarding the content and format

,

or specific test details. He noted that it did not provide for measurement of the coefficient at 25% as recommended by'RG 1.68

and as committed to in Section 13 of the FSAR.

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A licensee representative stated that the test would be run at

'

the lower testing level as well as the others called for by i

RG 1.68 or that relief would be sought from the FSAR commitment to the AEC Guide regarding this test.

a.

TP 800.28, "Psuedo Control Rod Ejection Test" The purpose of this test is to verify the safety analysis for the psuedo ' rod ejection accident by measuring-the reactivity of

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the highest worth ejected rod and comparing that worth to Tech-i nical Specification limits. The test is to be run at 40% power

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in accordance with item D.1.q of Appendix A of RG'l.68 (greater than or equal to 10% power).

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(1) The inspector asked if the procedure should not address' the potential problems of abnormal flux profiles and, rod in-sertions exceeding limits.

A licensee representative stated that these matters would be spoken to in the limits and precautions r.ection of the procedure.

(2) The inspector asked why control rod group 6 would be used to compensate for reactivity changes rather than boration and deboration when the control rod being tested was pulled and i

reinserted.

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A licensee representative stated that tests at similar B&W plants using both methods of compensation have shown them to

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be equally accurate for calculation purposes and that the use of control rods was easier and quicker.

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RO Report No. 50-313/74-11 II-1

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'

? ? ".' /,g DETAILS II Prepared by:

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W. D. Kelley,'Reactorf nspector.

Date i

Engineering Sec. tion /

Facilities Construction Branch Dates of Inspection: Ju 74 f

,

Reviewed by:

M [f r,vs 2/*,'9'/]

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B. J. Coc

, Acting Senior

' Date'

Inspec -

Engineering Section Facilities Construction Branch 1.

Individuals Contacted l

Arkansas Power and Light Company (AP&L)

N. A. Moore, Manager, Quality Assurance Contracting Organization Bechtel Engineering Corporation (Bechtel)

A. S. Meyers, Specialist, San Francisco

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E. Patel, Piping Engineer, San Francisco 2.

Valve Wall Thickness Verification Progran

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A meeting was held in the Region II office on April 18, 1974, with the licensee and his contractors' representatives to discuss Region II comments on their valve wall thickness veri-fication program submitted with their letter of December 7, 1973.

In answer to these comments, the licensee submitted their Valve l

Wall Thickness Verification Program, Revision 3, May 14, 1974.

On June 12, 1974, in a telephone conversation, the licensee was

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inf,rmed there were four items'that required action by them. The

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f su.: items were as follows:

a.

The AP&L audit did not contain resolution to all their findings.

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The weld end of valve DH-13A did not meet the requirements

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of the figure they referenced from the ASME Pump and Valve Code (Draft).

c.

Data was not available for thickness measurements after repair of valves DH-14A and B.

d.

There were insufficient measurements to verify the wall i

thickness of the large gate valves.

During the inspection, the licensee's corrective action on the above

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items and the documentation transmitted to Region II office by their letter June 25, 1974, were reviewed. The inspector has no comment and unresolved item 72-12/2 is closed.

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RO Report No. 50-313/74-11 III-1

DETAILS III Preparedby:/)

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D..y!)hrke, Reactor Inspector

/Dat'e Facilities Test and Startup Branch Dates of Inspection: August 4-6, 1974 Reviewed by: 8.(,

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[jf8d 74 R. C. Lewis, Senior Reactor Date Inspector Facilities Test and Startup Branch 1.

Individuals Contacted Arkansas Power and Light Company (AP&L)

P. Almond - Reactor Technician

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T. Baker - Chemical and' Radiation Protection Engineer Ox T. H. Cogburn - Nuclear Engir eer J. L. Orlicek - Quality Control Engineer D. R. Sikes - Results Engineer

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B. A. Terwilliger - Operations Supervisor 2.

Initial Criticality

' Initial criticality at ANO-1 was achieved at 6:02 a.m. (CDT) on August 6, 1974. TP 710.01 R1 was used to position the control

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rod groups and then to direct the deboration to criticality.

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Criticality occurred as predicted when the RCS was diluted to

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i approximately 1600 ppm boron. The items which the inspector witnessed as part of initial criticality are summarized below.

No deficiencies were identified.

The RCS water was sampled as required (5 times / week) and a.

was within the Technical Specification requirements. The

l gaseous and liquid effluents were monitored as required.

b.

The plant and procedure prerequisites for initial criticality were met.

TP 710.01 and OP 1102.01 were used to coordinate l

these activities. RB integrity was established, the various l

pre-operational tests and calibrations required were completed, j

and the RCS leak test was performed and within limits.

The inspector verified that the plant systems were in service p}

as required. The boric acid addition tank and equipment was

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operating at temperature, etc., the radiation monitors were l

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calibrated and in service as required, and the nuclear and process instrumentation was calibrated and checked as re-quired. The systems checklists were complete.

c.

The nuclear instrumentation was checked as required; cali-bration records existed, the precritical checks were performed as required, the signal to noise ratio was above two, and the minimum count rate requirement was satisfied (>2 cps).

The reactor trip setpoints on the power range RPS channels were lowered to approximately h % rated power. The source and intermediate range channels have SUR rod blocks, but no trips.

d.

The special instrumenta ion was available or in service as required by the procedures. The boronometer and source range count meters were used and the B&W reactimeter was available. _

The approach to criticality was orderly; the count rate and e.

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dilution data were properly recorded and presented for caalysis, and the resultant 1/m curves developed as expected.

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v t e Inspection Report - Arkansas Nuclear - 1974011
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \| 25 \|
1974 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 7(0) 9(0) 10(0) 11(0) 12(0) 13(0) 14(0) 15(0) Security Operator Exams Emergency Planning Other

IR 05000313/1974011

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