ML20056C967

Insp Rept 50-458/93-17 on 930614-18.Violations Noted.Major Areas inspected:10CFR50.59 Safety Evaluation Program

ML20056C967
Person / Time
Site:	River Bend
Issue date:	07/19/1993
From:	Westerman T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20056C965	List: ML20056C963 ML20056C966 ML20056C967
References
50-458-93-17, NUDOCS 9307300166
Download: ML20056C967 (12)

See also: IR 05000458/1993017

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APPENDIX B

U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Inspection Report: 50-458/93-17

Operating License: NPF-47

Licensee: Gulf States Utilities

P.O. Box 220

St. Francisville, Louisiana

Facility Name: River Bend Station (RBS)

Inspection At: RBS, St. Francisville, Louisiana

Inspection Conducted: June 14-28, 1993 _,

inspectors: W. M. McNeill, Reactor Inspector, Engineering Section

Division of Reactor Safety

Accompanying Personnel: E. T. Baker, Senior Project Manager, Nuclear Reactor

Regulation

Approved: )+ .

Thomas F. Westerman, Criief, Enginehing Section

7-/9-f_?

Date

Division of Reactor Safety

Inspection Summary

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Areas Inspected: Routine, announced inspection of the 10 CFR 50.59 safety

evaluation program.

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Results:

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. There was a good definition of scope of the documents to be used as the

basis to determine if a change effected the licensing bases and, -

therefore, required a safety evaluation (Section 2.1).

. The consolidation of the safety evaluation process into a single

procedure has been slow in development, in that, Procedure RBNP-57 has

been in a draft status since August 1991 (Section 2.1).

• Comprehensive training has been recently implemented (September 1992). l

Based on a limited observation of recent evaluations, the inspectors

noted an improvement in both the level of detail and quality of '

evaluations (Section 2.2).  ;

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• It was difficult to locate all the evaluations and screenings because of

the use of different data bases and forms. Self-assessment activities  :

by the licensee have identified this as a problem and the licensee had ,

taken steps to assure that all evaluations and screenings are accounted

for and reviewed (Section 2.2). ,

• The effort to prepare and review safety evaluations and screenings was '

good, however, one exception was noted. A violation was identified i

concerning the failure to comply with Technical Specifications while

establishing the minimum critical power ratio. In addition, one

deficient evaluation and one deficient screening were identified. The

impact of these deficiencies on plant safety were not significant ,

(Section 2.3).

Summar_y of Inspection Findings:

• Violation 458/9317-01 was opened (Section 2.1)

Attachments:

• Attachment 1 - Persons Contacted and Exit Meeting

• Attachment 2 - Documents Reviewed

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DETAILS

1 PLANT STATUS

During this inspection period the plant was shutdown and in a forced outage.

2 SAFETY EVALUATION PROGRAM, 10 CFR 50.59 (37001)

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2.1 Program

The inspectors found River Bend Station had established a program for

reviewing changes, tests, and experiments to the facility as described by the

safety analysis report. The program for safety evaluations included

consideration of other documents such as an " Environmental Protection Plan,"

Appendix B to the facility operating license. The program referenced and

followed the guidance of the Nuclear Safety Analysis Center 125, " Guideline

for 10 CFR 50.59 Safety Evaluations," which has not yet been endorsed by the

NRC staff.

The inspectors found the process for evaluations was a three step process.

The first step was to determine if the change, experiment, or test required a

full safety evaluation. If a hardware type change was made then it was

documented on an initial safety and environmental form. If a software type

change was made then it was documented on a safety and environmental

evaluation form.

For those changes, experiments, or tests that were not resolved at the first

step, the second step, an evaluation as an unreviewed safety question .

determination was performed and documented on an unreviewed safety question

determination form or an unreviewed environmental question determination form.

The third step was to determine if NRC approv'. of the change, experiment, or

test was to be obtained. The third step was called an unreviewed safety

question or unreviewed environmental question.

Currently, design engineering follows Procedure ENG-3-004 for preparation and

review of hardware type changes. The training requirements for both preparers

and reviewers were contained in Procedure EDP-AA-62. Procedure EDP-AA-62 was

referenced in Eng-3-004. The activities of plant personnel .(system

engineering) in making software changes, such as procedure changes, was  ;

described in Procedure ADM-0003, including the training requirements for the  !

reviewers only. This procedure could also be used by design engineering i

personnel. Procedures SSP-1-001, QAP 1.1, and QAP 1.0 required procedures

such as training, security, licensing, quality, etc., to be evaluated in terms

of 10 CFR 50.59 when they were changed. The requirements for the evaluation

of condition reports were described in Procedure RBNP-030. There was under

preparation, since August 1991, a Procedure RBNP-057 that consolidated the

safety evaluation process. Interest and support was observed to be slow in

developing.

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The inspectors found by reviewing the procedures and training information

that the licensing basis was used for evaluation of changes. The licensing  !

basis was defined as the safety analysis report including the quality

assurance manual, security plan and safeguards plan, environmental report,

safety evaluation report, Technical Specifications, operating license, final

environment statement, emergency plan, radiological environment effluent

monito-ing manual, offsite dose calculation manual, process control program ,

and con-!tments in NRC correspondence. 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. It was also noted that there were provisions in *

procedures and training for discretionary margins. Discretionary margins are

that which can be changed without prior NRC approval.  ;

2.2 Implementation

The inspectors reviewed the " Gulf States Utilities Company Annual Report

Pursuant to 10 CFR 50.59 (b)(2) March 1, 1991 through March 1, 1992." During

that period, the licensee had made 28 changes, 12 of which were hardware-type

changes (design changes). Fifteen changes were software type changes

(procedure changes) and one was a test.

The inspectors requested a list of modifications completed since March 1992

and found that 90 modifications had an evaluation and 104 had a screening.

In the same period, the inspectors found that 25 prompt modifications

(temporary modifications) were processed,11 of which had evaluations and 14 ,

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had screenings; and 424 condition reports (nonconforming condition reports)

had been issued, 8 of which had evaluations and 416 had screenings.

Comprehensive training had been a recent effort since September 1992.

Observation of some recent evaluations did note an improvement in both the  :

level of detail and the quality of evaluations.

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It was difficult to locate all the evaluations and screenings because of the

use of different data bases and forms. The licensee's self-assessment

activities had identified this as a problem. The independent safety

engineering group had taken steps to assure that all screenings and i

evaluations were reviewed as required by the offsite review committee. .

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The inspector noted there was an unreviewed safety question established ,

recently at River Bend Station. This unreviewed safety question had been the ;

subject of a violation in NRC Inspection Report 50-458/9311. This was in

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regard to an inoperable interlock on the containment ~ personnel hatch. There ,

will be no further discussion in this report of this subject.  !

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2.2.1 Evaluations

The inspectors reviewed the evaluations associated with seven modification ,

requests, four prompt modification requests (temporary modifications), three .

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document change notices, two condition reports, and the core operating. limits

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report. The evaluations reviewed are listed in Attachment 2. The following '

are the inspectors comments on specific evaluations:

Prompt Modification Request 90-0005

This temporary modification added a trailer mounted chiller package,

associated piping and valves to the ultimate heat sink in order to maintain ,

the water in the standby cooling tower basin below Technical Specifications'

maximum temperature limits. The evaluation included a seismic analysis of the

-piping, valves, and supports attached to the standby cooling tower. Also

analyzed was a break in the piping outside the tower and the possibility of

syphoning water out of the basin. Included in Prompt Modification

Request 90-0005 was a drawing, EP-19H-6, showing the suction and discharge

piping terminating at elevation 112 ft. 4 in., which was 6 in. above the

Technical Specifications required minimum water level. Also noted on the

drawing was the minimum water level required for operating the chiller at 114

ft. 2 in. to insure adequate net-positive suction head for the chiller pumps. ,

A maximum water elevation was not specified.

Chapter 9.2.5 of the Updated Safety Analysis Report states that the standby

cooling tower basin has a minimum water elevation of 111 ft.10 in., a normal

water elevation of 113 ft. 4 in., and a maximum water depth of 52 ft. 6 in.

Figure 9.2-11 in the Updated Safety Analysis Report shows the lowest elevation

in the basin as 60 ft. 10 in. Adding the maximum water depth of 52 ft. 6 in.

yields a maximum water elevation of 113 ft. 4 in., the normal water elevation.

The inspectors noted that the safety evaluation did not analyze for raising

the water level to an elevation of 114 ft. 2 in. to provide net-positive

suction head for the chillers. The concern was if the standby cooling tower

basin was seismically qualified when filled to that elevation. Based on

further review and discussion with the licensee, the inspectors determined

that the basin was seismically qualified to an elevation of 117 ft.10 in. In

addition, the control room received an alarm when water reached an elevation

of 116 ft.

While the safety elevation was deficient, additional review indicated there

was no unresolved safety question associated with the plant change. The

licensee will revise the Updated Safety Analysis Report to reflect current

limitations on maximum water depth in the cooling tower basin.

Core Operating Limits Report  !

On June 19, 1992, General Electric submitted a Part 21 report notifying that '

an error existed in the General Electric Standard Application for Reactor Fuel

Report (GESTAR-II/NEDE-240ll-P-A). The original approved version of GESTAR-II ,

concluded there was no need to perform cycle-specific evaluations of a rotated '

fuel assembly for a C-lattice plant. The change in critical power ratio was

analyzed at that time to be insignificant because of the fuel design

configuration. This conclusion was based on a generic analysis performed in ,

1982. However, the rotated bundle analysis remained. a requirement of the NRC j

approved methodology documented in GESTAR-II.  :

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The Part 21 report stated that recent studies had shown that the conclusion of ,

the generic analysis was sometimes inappropriate for modern-typ fuel designs. i

For BWR-6 S-lattice plants, the delta critical power ratio for ie rotated i

bundle was larger than the delta critical power ratio calculated for the

identified limiting transient. Additionally, if a BWR-6 S-lattice plant was

operating at the operating limit minimum critical power ratio, the postulated

rotated bundle condition could exceed the Technical Specifications' safety

limit minimum critical power ratio. The Part 21 also stated that General

Electric would perform cycle-specific rotated bundle analyses for future C-

and S-lattice fuel designs.

General Electric performed the reload analysis for Reload 4, Cycle 5, for

River Bend Station using the methods documented in Revision 10 of GESTAR-II.

General Electric submitted the results of the reload analysis to the licensee  :

in the Supplemental Reload Licensing Report, 23A7181, Revision 0, dated August

1992. General Electric performed a cycle-specific rotated bundle analysis as

part of the reload analysis and concluded that the operating limit minimum

critical power ratio for this cycle was 1.22.

In preparing the core operating limits report, the licensee performed an

evaluation which analyzed the impact of adding 200 bundles of Fuel

Type GE8-P8SQB334-10GZ-120M-4WR-150-T in the core during Reload 4. In that

evaluation, the licensee analyzed the probability of having rotated fuel and

considered the method of verifying placement of fuel bundles. The licensee

concluded that a rotated bundle did not exist and that the .04 minimum

critical power ratio penalty for rotated bundles was not applicable to River

Bend Station. The licensee documented in the core operating limits report

that the operating limit minimum critical power ratio applicable to River Bend

Station for Reload 4, Cycle 5, was 1.18.

However, neither the screening nor evaluation addressed Technical '

Specification 6.9.3.2, which required that the analytical methods used to

calculate the core operating limits be those described in the latest approved

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version of GESTAR. In establishing the operating limit minimum critical power

ratio as 1.18, rather than the 1.22 calculated by General Electric, the

licensee violated Technical Specification 6.9.3.2. This was identified as a

violation (458/9317-01).

The licensee reviewed the operating history for River Bend Station from

September 8,1992, when the plant started up from the refueling outage, to

April 18, 1993, the last day the plant operated with an operating limit

minimum critical power ratio of 1.18. The licensee identified one instance on .

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January 8,1993, when the plant operated below an operating limit minimum

critical power ratio of 1.22. However, the plant was below 1.22 for less than

2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, which met the action statement for Technical Specification 3.2.3.

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2.2.2 Screenings

The inspectors reviewed the screenings of five modification requests and two

prompt modification requests. The screenings reviewed are listed in '

Attachment 2. The following are the inspectors' comments on specific

screenings reviewed.

Modification Request 91-0013 ,

This modification added fencing and a detection system to the radwaste and

auxiliary control buildings that had been removed previously. The security

plan which was reportedly used as part of the bases for answers to the

screening was not identified in the documentation. Procedures require the

bases for answers to the screenings be documented. The inspectors reviewed

the security plan and found it did not detail the fencing and detection system

and, therefore, an unreviewed safety question did not exist. This screening

was considered deficient.

2.3 Conclusions

In general, the inspectors found the effort to prepare and review safety

evaluations and screenings was good, however, one exception resulted in the

identification of a violation. In addition, one deficient evaluation and one

d2ficient screening were identified. The impact of these deficiencies on -

plant safety were not considered significant. -

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ATTACHMENT 1

1 PERSONNEL CONTACTED i

1.1 Licensee Personnel

• K. Barrow, Board of Directors

• J. Booker, Manager Safety Assessment & Quality Verification

J. Brescher, Senior Electrical Engineer

• G. Bysfield, Electrical & Special Projects Supervisor

• T. Crouse, Director Engineering Support Services

• W. Curran, Cajun Site Representative

• D. Dorbonna, Assistant Plant Manager Operations

• L. Dietrich, Supervisor Nuclear Licensing

J. Egan, Senior Nuclear Safety Engineer

• D. Galle, Nuclear Safety Assessment Comittee Member

• K. Gardner, Licensing Engineer

• K. Gladrusich, Director Quality Assurance

• P. Graham, Vice President - River Bend

• J. Hamilton, Manager Engineering

W. Herman, Senior Technical Specialist

• T. Hoffman, Supervisor Civil / Structural Plant Engineering

• P. Hughes, Senior Licensing Engineer

A. Kahan, Senior Electrical Engineer

• G. Kimell, General Maintenance Supervisor

K. Klamert, Senior Mechanical Engineer

• L. Leatherwood, Supervisor Core Engineering

J. Leavines, Supervisor Nuclear Safety Assessment Group

J. Maher, Licensing Engineer

D. Melear, Senior Design Engineer

N. Pressler, Equipment Qualification Specification Engineer

• L. Rougeux, Senior Independent Safety Group Engineer

G. Scronce, Senior Nuclear Fuels Engineer

• W. Simmons, Senior Licensing Engineer

• B. Smith, Mechanical Maintenance Supervisor

• J. Spivey, Senior Quality Assurance Engineer

• M. Stein, Director Plant Engineering

• K. Suhrke, Manager Site Support

• J. Thompson, Supervisor Balance of Plant

• H. Woodson, Nuclear Safety Assessment Comittee ,

S. Woody, Director Nuclear Station Security

• J. Vachon, Senior Compliance Annalist

P. Vo, Senior Nuclear Fuels Engineer

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1.2 NRC Personnel

• W. Smith, Senior Reactor Inspector

In addition to the personnel listed above, the inspectors contacted other

personnel during the inspection.

• Denotes personnel that attended the exit meeting.

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2 EXIT MEETING

An exit meeting was conducted on June 18, 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

inspectors.

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ATTACHMENT 2

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DOCUMENTS REVIEWED

PROCEDURES

River Bend Nuclear Procedures Manual, RBNP-057, " Safety and Environmental

Evaluations," Revision draft, August 12, 1991  ;

River Bend Nuclear Procedure Manual, RBNP-030, " Initiation and Processing of

Condition Reports," Revision 1, April 10, 1992 with Change Notices 1-4

River Bend Nuclear Procedure Manual, RBNP-027, " initiation of a Change to a

Licensing Document," Revision 3, March 15, 1991

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River Bend Nuclear Procedure Manual, RBNP-026, " Evaluating Potentially

Reportable Conditions Pursuant to 10 CFR 21," Revision 2, August 11, 1992

Station Support Manual Procedures, ENG-3-004, " Safety and Environmental

Evaluations," Revision 1, November 21, 1989, with Change Notices 1-3

Station Support Manual Procedures, ENG-3-026, " Document Change Notices (DCN)," '

Revision 0, Nuvember 11, 1991

Station Support Manual, TTP-7-031, "10 CFR 50.59 Reviewer Training Program,"

Revision 0, April 14, 1993

Engineering Department Procedures Manual, EDP-AA-62, " Training Requirements

for Preparers, Reviewers and Approvers of Unreviewed Safety / Environmental

Question Determinations," Revision 3, March 17, 1990 with Interim Procedure

Change 1

Station Operating Manual, ADM-0003, " Development, Control and Use of

Procedures," Revision 18C, September 9, 1992 with Change Notices 93-0370 and

92-1347

Training Program Lesson Plan Number SET-721-01, "10 CFR 50.59 Reviewer"

EVALUATIONS AND SCREENINGS ASSOCIATED WITH MODIFICATION REQUESTS (MR) AND DATE

HR-88-0088, Modification of the Leak Off of Valve Packing, September 7,1988 i

(Evaluation)

MR 91-0013, Add Fence and Detection System to West Walls of Rad Waste and Aux  !

Centrol Buildings, October 2,1991 (Screenings)

MR 91-0044, Converts Temporary Piping Installed under Prompt Modification  !

Requests 90-0005 for Cooling the Standby Cooling Tower to a Permanent

Installation, January 29,1992 (Evaluation)

HR 91-0059, Service Water Piping Replacement Install Flanges on Alternate

Cooler, July 21, 1991 (Evaluation)  ;

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MR 91-0079, Provides a Source of Demineralized Make Up Water to the Closed

Loop Service Water System, January 28,1992 (Evaluation)

MR 91-0087, Install Permanent Taps in the Reactor Water Clean Up System for

Decontamination, February 27, 1992 (Evaluation)

HR 92-0010, Revise Documents to Spare Cables and Junction Box that Had Been a  ;

Temporary Modification, February 28, 1992 (Screenings)

MR 92-0011, Provide Alternate Model and Associated Mounting and Termination

Details for Replacement Differential Pressure Switches, January 23, 1992  !

(Screenings)

MR 92-0036, Revises Setting for Pressure Regulators, March 29, 1992

(Screenings)

HR 92-0047, Place a New Isolation Valve on the Condensate Demineralizer,-

July 16,1992 (Evaluation)

HR 92-0048, Replace Motor on Feed Water Containment Isolation Valves,

April 30,1992 (Evaluatic, -) .

MR 92-0053, Upgrade Pressure Switch, July 18, 1992 (Screenings)

EVALUATIONS AND SCREENINGS ASSOCIATED WITH PROMPT MODIFICATION REQUESTS (PMR)

AND DATE

PMR 90-0005, Installation of Temporary Chillers for Standby Cooling Tower,

April 30, 1990 (Evaluation)

PMR 92-0014, Provide Temporary Bulk Temperature of the Reactor Vessel Water in

Mode 5, July 18, 1992 (Screenings)

PMR 92-0025, Install Different Model Chemical Feed Pump in the Normal Cooling i

Tower Make Up Water Treatment System, December 16,1992 (Evaluation)

PMR 93-0008, Terminate All Flow to the Condensate Storage Tank from the

Control Rod Drive System, March 16, 1993 (Evaluation)

PMR 93-0012, Installation of Furmanite Clamp Around a Flange, March 20, 1993

(Screenings)

PMR 93-0015, Install Temporary Pressure Transmitter, April 23, 1993

(Evaluation)

EVALUATIONS ASSOCIATED WITH DOCUMENT CHANGE NOTICES (DCNs) AND DATE

DCN 91-00189, Revise P&ID to Show Extant of ASME Boundary of Hydraulic Control

Units, undated ,

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DCN 92-00413, Revise P&ID 6-1B and FSK-6-IC to Refit .t As-Built Configuration,

November 24, 1992

DCN 93-0082, Revisc System Design Requirements Document P-21 "Feedwater and

Auxiliaries," February 14, 1993

EVALUATIONS ASSOCIATED WITH CONDITION REPORTS (CR) AND DATE

CR 92-0572, Feed Water Check Valve IB21*A0VF032A, Surface Indications on Valve

Seat, August 8, 1992

CR 92-0675, Reactor Pressure Vessel Closure Stud Ultrasonic Indication, August

18, 1992