IR 05000313/1993023

Insp Repts 50-313/93-23 & 50-368/93-23 on 930628-0702. Violations Noted But Not Cited.Major Areas Inspected: 10CFR50.59 Safety Evaluation Program

ML20056D004
Person / Time
Site:	Arkansas Nuclear
Issue date:	07/19/1993
From:	Westerman T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20056D003	List: IR 05000313/1993023 ML20056D002
References
50-313-93-23, 50-368-93-23, NUDOCS 9308020089
Download: ML20056D004 (10)

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APPENDIX U.S. NUCLEAR REGULATORY COMM'SSION l

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REGION IV

Inspection Report: 50-313/93-23; 50-368/93-23 Operating Licenses: DPR-51; NPF-6

Licensee: Entergy Operations, In Route 3, Box 137G Russellville, Arkansas 72801 *

Facility Name: Arkansas Nuclear One (AN0), Unit 1 and Unit 2 i

Inspection At: ANO, Russellville, Arkansas Inspection Conducted: June 28 through July 2, 1993 ,

i Inspector: W. McNeill, Reactor Inspector, Engineering Section Division of Reactot Safety Accompanying Personnel: T. W. Alexion, Project Manager, Nuclear Reactor Regulation ,

Approved: - ( #

N 38-//8 Thomas F. Westerman, Chief, Engineering Section Date ;

Division of Reactor Safety

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Inspection Summary Areas Inspected (Unit I and Unit 2): Routine, announced inspection of the '

10 CFR 50.59 safety evaluation progra Results (Unit I and Unit 2):  !

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• The 10 CFR 50.59 determinations and evaluations were excellently prepared. The logic used was clearly defined. In general, the ;

determinations were of high quality (Section 2.3). j

= Review and approval of 10 CFR 50.59 determinations and evaluations were [

by the plant safety committee. It was noted although'not a procedural' ;

requirement, only about one-half of this committee was trained as !

certified reviewers (Section-2.1). i

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• Evaluations and the associated change packages were not always clear as .,

to "why" a change was being made. One determination was found that was i not representative of the generally high quality of work (Section 2.2). .

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• The licensee identified certain 10 CFR 50.59 evaluations of changes to  !

safety analysis report figures that were not identified in annual >

10 CFR 50.59 reports or annual safety analysis report updates. This was identified as a noncited violation (Section 2.2).

Summary of Inspection Findings: l

A noncited violation was identified (Section 2.2). >

Attachments:

• Attachment 1 - Persons Contacted and Exit Meeting ,

• Attachment 2 - Documents Reviewed r

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DETAILS 1 PLANT STATUS During this inspection period the plants were operating at 100 percent power in Mode !

2 SAFETY EVALUATION PROGRAM, 10 CFR 50.59 (37001)

2.1 Program

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The inspectors found Arkansas Nuclear One had established a program (1000.131)

for reviewing changes, tests, and experiments for changes to the facility as described by the safety analysis reports. The program for safety evaluations included consideration of other licensing basis documents such as an

" Environmental Protection Plan," as required by Appendix B to the Facility *

Operating License. There was a different procedure (1012.015) in response to Inspection and Enforcement Circular No. 80-18 for radiological safety evaluations . .

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The licensee considered the program conformed to Nuclear Safety Analysis Center (NSAC)-125, " Guideline for 10 CFR 50.59 Safety Evaluations" with one :

exception. The licensee was using only the margin of safety as defined in the '

bases section of the Technical Specifications as apposed to the margin found in the design documents. This was considered to be verbatim compliance with !

the 10 CFR 50.59 rul l The inspectors found the process for evaluations was a three step proces '

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The first step was a screening documented on a "10 CFR 50.59 Determination" form. For those changes, experiments, or tests not resolved at the first ,

step, then the second step, a safety evaluation was performed. This was- i documented on "10 CFR 50.59 Safety Evaluation" for If resolution could not be obtained by the evaluation, then the third step was taken. The third step ,

was called an unreviewed safety question. The purpose of the third step was to assure the change, experiment, or test was not done without prior NRC ;

approva All determinations and evaluations were to be prepared by a " certified-reviewer." At other sites, determinations and evaluations were reviewed and -

approved by an independent reviewer much like other types of design document The final approval of determinations and evaluations at Arkansas Nuclear One i was performed by the Plant Safety Committee, the onsite review committee. As an observation, the inspectors noted that the Plant Safety Committee members ;

were not all trained as " certified reviewers." Although not a procedural r requirement, about one-half of the members were trained as " certified reviewers."

The inspectors found from review of the procedure, training plans, and training material that the licensing basis used was the safety analysis '

reports and the operating licenses. The safety analysis reports were defined

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-4-to be the safety analysis reports (one for etch unit) including the emergency plan, quality assurance manual, fire hazards analysis, Technical Specifications bases, and safety evaluation reports. The operating licenses included all license conditions, such as Technical Specifications and orders issued by the NRC. This was considered by the inspectors to be a good definition of the scope of documents to be used as the basis to determine a change. Some minor editorial changes of the procedure were identified by the .

inspectar .2 Implementation The inspectors reviewed the "1991 Annual 10 CFR 50.59 Report." During that period, the licensee had made 458 changes of which 297 were design changes,14 were temporary modifications, 87 were procedure changes, and 60 were miscellaneous changes. It was found by the inspectors that the report contained some 178 changes that had been omitted from past reports which was a violation of 10 CFR 50.59(b)(2) and also 10 CFR 50.71(e). The licensee .

identified this violation in 1991 during the process of turnover of the J responsibility of compilation of reportable changes from one individual to anothe The problem was identified on Condition Report C-91-073, dated July 17,1991. Certain changes to the safety analysis report drawings were not reported in the annual 10 CFR 50.59 report nor was the safety analysis report updated. These changes, although not reported, were evaluated and implemented. This problem was mostly limited to safety analysis report drawings. The corrective action was to review all safety analysis drawing changes and identify the unreported changes. Most of these changes, except for 58, were finally reported in the 1991 annual report. The balance will be reported in the 1992 annual report. The failure to provide certain 10 CFR 50.59 evaluations of changes in the annual 10 CFR 50.59 reports and also to have updated the safety analysis report annually in regard to these changes l was identified as a noncited violation because it was self-identified and I corrected as provided for by the NRC Enforcement Polic . Evaluations The inspectors reviewed the evaluations associated with four procedure changes, nine design change packages (including one limited change package, and one plant change), three temporary modifications, one material technical evaluation, one plant engineering action request, one condition report and two safety analysis report changes. In this review the inspectors verified that

" certified reviewers" were used to prepare evaluations. The evaluations reviewed are listed in Attachment 2. The following are the inspectors comments on specific evaluation Design Change Package 80-1172 This modification was initiated to add a new pump and piping to ensure the ;

continuous oil replenishment of the oil cooler and filter regardless of oil ,

temperature and viscosity. Following discussions with the licensee, the l

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-5-inspectors learned that this change was completed in 1986, however, the Safety Analysis Report, Figure 8-3, was only recently updated to reflect the chang The evaluation documentation was of poor qualit It stated the purpose of the modification and then restated the evaluation question negatively. This type of writeup was not in accordance with the licensee's current procedure (1000.131). The licensee informed the inspectors that the evaluation would not be revised. This was because it was done according to the procedures in place at the time of preparatio In addition, the licensee informed the inspectors they had received a violation in 1986 for the lack of proper documentation of evaluations and they revised their program sufficiently for NRC to close the violation in a followup inspection in 1988. Based on the above, the inspectors concluded this was just another example of a problem that had already been identified and corrected. The licensee had engineering and other evaluations on file in support of the change. This was not a reflection of the licensee's current performanc Temporary Modification 93-2-008 This temporary modification provided an alternate power supply to a service water pump. However, the pump was not considered operational from a Technical Specifications viewpoint. This pump was the third service water pump and two were necessary to satisfy Technical Specifications requirements. The safety analysis report described all three pumps and their functions, power cables, loads, etc. The evaluation was detailed and concise. This modification did not clearly identify that the reason for the modification. The "why" of this modification was found in the Limited Change Package 93-6011. This pump's cables were grounded and Condition Report 2-93-0122 was generate .2.2 Determinations The inspectors reviewed the determinations associated with two design change '

packages, two limited change packages, two plant changes, and two temporary modifications. The determinations reviewed are listed in Attachment 2. In i this review, the inspectors verified that " certified reviewers" were used to prepare determinations. The following is the inspectors comments on one specific determinatio Limited Change Package 92-5021 As a result of the " Iso Update Project" it was discovered that a nozzle on each emergency feed water ring had been removed. As a result, there was not agreement with drawings and calculations. A Condition Report 1-91-0126 '

documented this condition. During the repair, it was discovered that thermal lockup could occur because of insufficient gaps required by ASME Code. This modification addressed the thermal lockup problem. Some supports were removed and others had additional gaps provided. This modification did not change information in the safety analysis report and the licensing basis document .

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-6-The answer:; on .i.w determination form did not appear to fully relate to the questions. After discussion with licensee personnel, it was established that the determination was satisfactor ,

2.3 Conclusions In general, the inspectors found safety evaluations and determinations were excellently prepared and detailed with clearly defined logic use !

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ATTACHMENT 1  ;

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1 PERSONNEL CONTACTED l.1 Licensee Personnel

• S._ Bennet, Licensing Specialist

• H. Cooper, Licensing Specialist

• B. Day, System Engineering Manager Unit 1 J. Dorsa, Licensing Specialist

• Eaton, Director Design Engineering R. Edington, Unit 2 Plant Manager  ;

• J. Fisicaro, Licensing Director

• D. James, Licensing Supervisor

• D. Hims, System Engineering Manager Unit 2

• J. Taylor-Brown, Acting Quality Director J. Yelverton, Vice President Operations 1.2 NRC Personnel

• L. Smith, Senior Resident Inspector L

In addition to the personnel listed above, the inspectors contacted other personnel during this inspection perio P

• Denotes personnel that attended the exit meetin EXIT MEETING An exit meeting was conducted on July 2, 1992. During this meeting, the inspectors reviewed the scope and findings of the report. The licensee did -

not identify as proprietary any information provided to, or reviewed by,. the inspector !

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ATTACHMENT 2 DOCUMENTS REVIEWED Procedures Procedure 1000.131, "10 CFR 50.59 Review Program," Revision 1, March 15, 1993 Training Lesson TUDI AS10800-050, "10 CFR 50.59 Changes, Tests and Experiments," Revision 4, June 16, 1993 Training Lesson TUDI AS10800-052, "10 CFR 50.59 Changes, Tests and Experiments, Recertification Training," Revision 3, February 15, 1993 ,

l Evaluations Associated with Procedure Chances  ;

Temporary Changes 2 and 3 to 1103.005, " Pressurizer Operation," Revision 21, July 2, 1991 b Temporary Change 1 to 1103.011 " Draining and Nitrogen. Blanketing the Reactor Coolant System," Revision 12, June 14, 1990 Revision 23 to 2104.040, " Low Pressure Safety Injection Operations,"

December 14, 1990 Permanent Change 4 to 2105.016, " Radiation Monitoring and Evacuation Alarm System," Revision 13, November 15, 1991 Determinations and Evaluations Associated with Modifications Design Change Package 80-1172, " Diesel Generator Lube Oil System Modification (K4A&B)," January 25, 1983 (Evaluation)

Design Change Package 84-1014, "Reactnr Vessel Level . Monitoring System,"

September 10,1986 (Evaluation)

Design Change Package 87-1097, "High Pressure Injection Motor Operator Valve Replacement (CV-1219, CV-1220, CV-1227 and CV-1228)," January 28, 1992, and Revision 1, April 30, 1993 (FFN-93-020) (Evaluation)

Design Change Package 90-2053, " Control Room Expansion Facility," March 28, 1993 (FFN 93-017) (Evaluation) ,

Design Change Package 91-1013, " Decay Heat Removal Vortex Monitoring," May 13, l 1993 (FFN-93-035) (Evaluation)

Design Change Package 91-1016, " Gamma Metrics Cable Supports," March 12, 1993 (Determination)

Design Change Package 91-2003, "ANO-2 Hydrogen Monitoring Modification,"

August 19, 1992, and Revision 1, March 30 (FFN 93-008) (Evaluation)

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-2-Design Change Package 92-1022, " Service Water Truncated Branch Line Code Qualification," May 4,1993 (Determination) j Design Change Package 92-2007, " Nitrogen Supply System for the ANO-2 Reactor Building Penetrations," April 15, 1992, Revision 1, August 11, 1992 and 1 Revision 2, May 28, 1993 (FFN 93-007) (Evaluation)

Determinations and Evaluations Associated with limited Chance Packaces Limited Change Package 92-5021, " Emergency Feed Water Ring Header Pipe Support Modifications," May 7, 1993 (Determination)

Limited Change Package 92-5040, " Iso Update Project Engineering Analysis Code Compliance," May 14, 1993 (Determination)

Limited Change Package 93-6011, " Service Water Pump 2P4A Cable Replacement,"

June 7, 1993 (FFN-93-042) (Evaluation)

Determinations and Evaluations Associated with Plant Chances

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Plant Change 92-8068, "2P4B Engineered Safety Feature Actuation System Annunciator," November 25, 1992 (Determination)

Plant Change 93-7007, " Traveling Screen Header Replacement." May 6, 1993 '

(FFN 93-027) (Evaluation)

Plant Change 93-8007, " Addition of Compression Fittings to Instrument Tubing on Off-side B Recirculation Steam Generator Hand Hold," February 19, 1993 (Determination)

Determinations and Evaluations Associated with Temocrary Modifications Temporary Modification 91-2-010, " Margin to Saturation Monitor Input Modification," February 25,1991 (Evaluation)

Temporary Modification 93-2-003, " Oil Fill Line for Reactor Coolant Pump 'B'

Motor (2P-328)," March 10, 1993 (Determination)

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Temporary Modification 93-2-004, " Install RTDs Control Element Drive Mechanisms Cooling System Thermowells and Control Element Drive Mechanism Shrouds," May 3, 1993 (FFN-93-029) (Evaluation)

Temporary Modification 93-2-006, "Nomex Packing Installation in 2P36C,"

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March 31, 1993 (Determination)

Temporary Modification 93-2-008, " Temporary Power feed to 2P4A," May 13, 1993 (FFN-93-031) (Evaluation)

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Evaluations Associated with Safety Analysis Report Changes Section 4.2.4.7, " Correction of High-point Vent Description," June 25, 1992 Table 6.2-18, " Containment Spray Pumps (2P35) Material of Construction,"

April 9, 1992 Miscellaneous Evaluations ,

i Material Technical Evaluation 91-0197, " Disc Material Change," May 4, 1993 '

(FFN-93-023)

i Plant Engineering Action Request 91-7253, " Ensure NRC Barriers are Included in '

the Penetration Log," May 13, 1993 (FFN-93-033)

i Condition Report 2-91-0353 Action Item 4, " Changes to Unit 2 Safety Analysis i Report Figure 9.7-12 (Table 9.2-1)," March 4, 1993:(FFN-93-024) ~  ;

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