Text

Enforcement Conference Rept 50-331/93-21 on 930824-1014. Areas Discussed:One Apparent Violation Along W/Corrective Actions Taken or Planned by Licensee
	ML20059B313
Person / Time
Site:	Duane Arnold
Issue date:	12/16/1993
From:	Lanksbury R; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:	;
Shared Package
ML20059B306	List: ML20059B303; ML20059B313;
References
	50-331-93-21-EC, EA-93-255, NUDOCS 9401040077
	Download: ML20059B313 (36)
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U.-S. NUCLEAR REGULATORY COMMISSION

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REGION III

Report No.

50-331/93021(DRP)

EA-93-255

Docket No.

50-331

License No. DPR-49

Licensee:

Iowa Electric Light and Power Company

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IE Towers

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P. O. Box 351

,

Cedar Rapids, IA 52406

Meeting Conducted: November 19, 1993

Meeting Location:

Region III Office

799 Roosevelt Road

Glen Ellyn, IL 60137

Type of Meeting:

Enforcement Conference

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Inspection Conducted:

Duane Arnold Energy Center

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August 24 through October 14, 1993

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Inspectors:

J. Hopkins

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C. Miller

Approved By:

'd

(2\\Nh1

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R. D. Lanksbu'ry, Chief',

Date

Reactor Projects Sedt4dn 3B

Meetino Summary

Enforcement Conference on November 19. 1993 (Report No. 50-331/93021(DRP)

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Areas Discussed:

One apparent violation with two examples identified during-

the inspection was discussed along with the corrective actions taken or

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planned by the-licensee. This. apparent violation involved the inability of

the *B" standby diesel generator (SBDG) to automatically supply.~ power to

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essential bus IA4 had offsite power been lost.

In the first example, the "B"

SBDG was inoperable for a time in excess of that. allowed by technical

specifications.

In the second example, core alterations were in progress

while both the "A"

and "B" SBDGs were inoperable contrary to technical.

specifications.

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9401040077 931219 .

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DETAILS

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Persons Present at the Enforcement Conference

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Iowa Electric Liaht and Power Company

J. Franz, Jr., Vice President-Nuclear

D. Wilson, Plant Superintendent, Nuclear

M. McDermott, Manager, Engineering

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K. Peveler, Manager, Corporate Quality Assurance

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J. Bjorseth, Maintenance Superintendent

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T. Gordon, Supervisor, Electrical Maintenance

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T. Allen, Senior Discipline Engineer, Electrical-

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L. Heckert, Principle Licensing Specialists

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U. S. Nuclear Reaulatory Commission

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H. Miller, Deputy Regional Administrator, Region III

E. Greenman, Director, Division of Reactor Projects, Region Ill

R. DeFayette, Director, Enforcement and Investigation Coordination

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Staff, Region III

L. Greger, Chief, Reactor Projects Branch 3, Region III

B. Berson, Regional Counsel, Region III

R. Pulsifer, Project Manager, Nuclear Reactor Regulation (NRR)

.

J. Beall, Office of Enforcement, NRR (via telephone)

J. Hopkins, Senior Resident Inspector, RIII

Z. Falevits, Electrical Inspector, Region III

,

M. Khanna, Reactor Engineer, Region III

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2.

Enforcement Conference

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An enforcement conference was held in the NRC Region III Office cn

November 19, 1993. This conference was conducted as a result of the

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findings of inspection conducted from August 24 through October 14,

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1993, in which two examples of an apparent violation of NRC regulations

were identified.

Inspection report (IR) 50-331/93015(DRP), dated

November 5, 1993, documented the results of the inspection.

The purpose of. this conference was to discuss the apparent violation,

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root causes, contributing factors, and the licensee's corrective

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actions. During the enforcement conference, the licensee presented the

event investigation, safety significance, causes, and corrective.

actions.

.

The licensee's presentation contained no significant additions to the

description of the event documented'in IR 50-331/93015(DRP).

In addition to the corrective actions documented in IR 50-

331/93015(DRP), the licensee's presentation indicated that the following

additional corrective actions had been taken:

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A description of the event was placed on the " Notepad" network for

the Institute of Nuclear Power Operations and on the nuclear plant

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reliability data system.

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The licensee attended a workshop sponsored by the Nuclear

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Maintenance Application Center, a subgroup of Electrical Power

Research Institute, to review the event and draft a revision to

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EPRI's 4160 Vac circuit breaker maintenance procedure

recommendations.

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Event specific train'ig was completed for all plant operating

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crews.

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All maintenance electricians were trained on the details of the

event and the corrective actions in place to prevent recurrence.

The licensee determined that the issue was not reportable under

10 CFR Part 21. However, the licensee committed to review its

initial evaluation.

Labels were attached to the 4160 Vac circuit breakers stating that

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the correct breaker plunger was to be verified when a circuit

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breaker was " racked in."

Licensee event report 50-331/93-008 was issued on October 18,

1993, which described the event.

The licensee presented the results of its risk assessment of the

"B" SBDG being inoperable during power operations and while core

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alterations were in progress. The licensee concluded that even

though the core damage frequency was slightly increased, prior

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training, outage risk management, and existing plant procedures

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minimized the overall risk to the plant and the public.

In addition to the corrective actions documented in IR 50-

331/93015(DRP), the licensee's presentation indicated that the following

additional corrective actions were planned or under consideration

,

following the "B" SBDG being inoperable for a time in excess of that

allowed by technical specifications:

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The circuit breaker vendor, General Electric (GE), planned to

issue a Service Advisory Letter (SAL) describing the issue'.

The licensee contracted with GE to review the revised 4160 Vac

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circuit breaker maintenance procedure.

It planned to wait until

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GC completed the evaluation before the revision was issued.

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A copy of the licensee's and NRC's presentations are attached to this

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report.

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Attachments: As stated >

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U.S. NRC REGION lli

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DUANE ARNOLD

ENFORCEMENT CONFERENCE

November 19,1993

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10:00 A.M. (CST)

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E A 93-255

REPOPT NUMBER 50-331/93015

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REGION 11! OFFICE

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799 ROOSEVELT ROAD, BUILDING 4

GLEN ELLYN, ILLINOIS'

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DUANE ARNOLD

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ENFORCEMENT CONFERENCE

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Agenda

INTRODUCTION AND OPENING REMARKS:

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Edward G. Greenman, Director, Division of Reactor Projects

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NRC OVERVIEW:

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Edward G. Greenman, Director, Division of Reactor Projects

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SUMMARY OF EVENTS:

Jay A. Hopkins, Senior Resident inspector, Duane Arnold

Diesel Generator inoperability

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SUMMARY OF APPARENT VIOLATIONS:

Jay A. Hopkins, Senior Resident inspector, Duane Arnold

LICENSEE PRESENTATION AND DISCUSSION:

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Iowa Electric Light and Power Company

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NRC FOLLOWUP QUESTIONS

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CLOSING RE% ARKS:

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Hubert J. Millet; :Egion 1 Deputy Administrator

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APPARENT VIOLATION

Duane Arnold Energy Center technical specification 3.5.G.1

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states that with one standby diesel generator (SBDG)

inoperable, continued reactor operation is permissible for the

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next 7 days unless the SBDG is made operable. If that

condition is not met, an orderly shutdown shall be conducted

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and the reactor shall be taken to hot shutdown within the next.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and taken to cold shutdown within the following 24

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hours.

Technical specification 3.9.D.1. requires that with core

alterations in progress, one SBDG be operable with its

associated standby gas system train and its main control room

ventilation standby filter unit subsystem. If that condition can

not be met, core alterations are not permitted.

CONTRARY TO THE ABOVE

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a.

On July 21,1993, with the reactor operating at approximatel

75 percent power, the "B" SBDG became inoperable. The "B"

SBDG was not restored to operable status within 7 days and

the reactor was not taken to hot shutdown within the next 12 '

hours, or cold shutdown within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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b.

From August 7 to 11,1993, with core alterations in progress,

the "A" and "B" SBDGs were inoperable.

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The apparent violatio.ns discussed in this enforcement conference are

subject to further review and may be subject to change prior to any

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resulting enforcement action.

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"B" SBDG INOPERABILITY TIMELINE

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04/16/92 Completion of " LOOP /LOCA" STP on "B" SBDG.

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07/21/93 Routine maintenance on breaker 1 A401.

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07/29/93 Reactor in Hot Shutdown.

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.07/31/93 Reactor in Cold Shutdown

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08/04/93 "A" SBDG tagged out for maintenance.

08/07/93 Core Offload Started.

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08/11/93 Core Offload Completed.

08/22/93 "A" SBDG operable.

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09/16/93 1 A411 failed to close during "B" side LOOP /LOCA.

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09/22/93 1 A411 failed to close during "B" side LOOP /LOCA

09/23/93 1 A401 plunger gap adjusted.

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09/24/93 "B" side Loop-LOCA successfully completed 3rd attempt.

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09/25/93 "B" SBDG declared operable

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09/27/93 All auxiliary contacts verified operable "as-found".

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09/28/93 Measurements taken on all 4160 breakers. 26 breakers

outside acceptance criteria.

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10/02/93 10 breakers were adjusted.

Retests were cc.mpleted satisfactori,1y.

10/02-04 5 non-essential breakers adjusted.

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10/04/93 Reactor S/U.

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ENFORCEMENT CONFERENCE

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DUAXE ARNOLD EXERGY CESTER

FRIDAY, NOVEMBER 19,1993

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DUANE ARNOLD ENERGY CE:5TER

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AGE:SDA

OPENING REMARKS

- JOHN FRANZ

VICE PRESIDENT-NUCLEAR

DISCUSSION OF EVENT

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TOM GORDON

SUPERVISOR,

ELECTRICAL MAINTENANCE

&

SAFETY SIGNIFICANCE

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TIM ALLEN

SENIOR DISCIPLINE ENGINEER,

ELECTRICAL

CLOSING REMARKS

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DAVE WILSON

PLANT SUPERIN'ENDENT

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DAEC

20/ EMBER 19,1PF3

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DUANE ARNOLD E:SERGY CE3TER

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DISCUSSION OF EVENT

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TOM GORDON

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SUPERVISOR, ELECTRICAL MAINrENANCE-

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DUANE ARNOLD ENERGY CENTER

DISCUSSION OF EVENT

e SEQUENCE OF EVENTS

- Initial LOOP /LOCA Test Identification of Failure

LOOP /LOCA Surveillance (Figure 1)

Problem Summary

- ProblemIsolation

1A411 Circuit Description (Figure 2)

Troubleshooting Sequence

Root Cause Identification

- Successful LOOP /LOCA Test

e CORRECTIVE ACTIONS

- Short Term

Maintenance History Review

VendorInput

Existing Breaker Conditions

Breaker Restrictions

Training

Industry Notifications

- Long Term

Procedure Revision

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Industry Notification

Training

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NCNEMBER19,19P3

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"B" SBDG

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Essential Buse

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Figure 1-

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FIGURE 2:

DIESEL GENERATOR OUTPUT BREAKER

CONTROL SCHEMATIC

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Magne Blast

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4.16 W Breaker

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Breaker Lifting Rail-to-Plunger Bolt Measurement

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FIGURE 3

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4160 VAC Magneblast Breaker Plunger Measurements

-(Acceptance Value of 117/32 to 11 11/32 inches)

Plunger toteft Lifting Rail

Plunger to Right Lifting Rail

Team A -

11 12/32

11 2/32

Team B .

11 6/32

11 8/32

Team C -

11 6/32

11 8/32

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Breaker Plunger Bolt

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Magne-Blast

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Breaker Lifting Rail-to-Plunger Bolt Measurement

FIGURE 4

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FIGURE 5: BREAKER DIMENSIONS

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FIGURE 6

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Breaker Open Position

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Essential 4160 V Magne-blast Circuit Breakers

Breaker

Description

Adj. Needed Adjusted

Comments

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1A301

Standby Transformer Feeder

X

1A302

Startup Transformer Feeder

1A303

480 V load center transformer 1X-31

1A304

Core Spray Pump 1P-211 A

X

1A30S

RHR Pump 1P-229A

1A306

RHR Pump 1P-229C

X

1A307

RHR Service Water Pump 1P-22A

1A308

RHR Service Water Pump 1P-22C

1A309

General Service Water Pump 1P-89A

1A310

CRD Feed Pump IP-209A

1A311

SBDG 1G-31 Feeder

X

1A312

480 V Load Center Traasformer 1X-91

X

Contacts feed computer point only.

1A401

Standby Transformer Feeder

1A402

Startup Transformer Feeder

1A403

480 V load center transformer IX-41

1A404

Core Spray Pump 1P-2118

X

1A405

RHR Pump IP-229B

X

1A406

RHR Pump 1P-229D

X

1A407

RHR Service Water Pump 1P-228

X

1A408

RHR Service Water Pump 1P-22D

X

1A409

General Service Water Pump 1P-898

X

No Safety related functions (GSW)

1A410

CRD Feed Pump 1P-209B

1A411

SBDG-1G-21 Feeder

X

l1A412

480 V Load Center Transformer 1X-20

X

Contacts feed computer alarm only.

Recirculation Pump Trip Breakers

1A501

Recire. Pump 1P-201 A Trip Breaker A

1A502

Recire. Pump 1P-201B Trip Breaker A

X

Contacts have no function.

1AS01

Recirc. Pump 1P-201 A Trip Breaker B

1A602

Recire. Pump 1P-201 B Trip Breaker B

FIGURE 8: EVALUATION MATRIX FOR

AFFECTED BREAKERS

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Non-essential 4160 V Magne-blast Circuit Breakers

Breaker

Description

Adj. Needed Adjusted

Comments

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1A101

Auxiliary Transformer Feeder

1A102

Startup Transformer Feeder

X

Evaluated by Systems Engineering

1A103

Reactor Feed Pump 1P-1 A

X

1A104

Reactor Recire. MG set 1G-201 A

X

Non Critical Plant impact

1A105

Circulating Water Pump 1P-4A

1A106

Condensate Pump 1P-8A

X

1A107

480 V load center transformer IX-11

1A108

480 V load center transformer 1X 71

1A109

480 V load center transformer 1X-51

1A110

480 V switchyard transformer

1A201

Auxiliary Transformer Feeder

X

Evaluated by Systems Engineering

3202

Startup Trsnsformer Feeder

X

Non Critical Plant impact

IA203

Reactor Feed Pump 1P-1B

1A204

Reactor Recire. MG set 1G-2010

X

Non Critical Plant impact

1A205

Circulating Water Pump 1P-48

1A206

Condensate Pump 1P 88

X

1A207

480 V load center transformer 1X-21

X

Non Critical Plant impact

1A208

480 V load center transformer IX-81

X

Non Critical Plant impact

1A209

480 V load center transformer 1X-61

1A210

General Service Water Pump 1P-89C

1A211

Well Water Pump 1P-58D

X

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FIGURE 8: EVALUATION MATRIX FOR

AFFECTED BREAKERS (Continued)

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Breaker Inspection Section

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TASK

CKTBKR-G080-002 Rev.10

GEK-7320F <5000 cycles

EPRI NP-7410 Aug. 93

step #

allowance

step #

allowance

step #

allowance

____

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1 Record initial breaker Operation Counter reading.

5.1.1

x

2 Remove breaker, verify safe and remove barrier

5.1.2-3

a.1

3 Inspect interruptors:

5.1.5

sat /unsat

a.3_

det.

a

a.3

det.

damage to the arc chute sides

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det.

contamination to the throat area

b

a.3

_

det.

breaks in the plastisal covering for pole pieces

c

a.3

4 Check condition of arcing and main contacts

5.1.6

sat /unsat

b.2 b.3

dot.

5 Check all bolts are tight

5.1.7

6 Clean breaker

5.1.8

gen.

7 Examine moving joint on load carrying members

5.1.9

galling (y/n)

!

8 Examine fiberglass sheets

5.1.10

sat /unsat

5 Lubricate as required

5.1.11

b.7

lub.

10 inspect control wiring and terminations

5.1.12

b.8

~ ii Megger form the studs on Sec. Cpir. to frame 250v

5.1.13

x

elec.

> = 1.25Mohm

12 Arcing contact wipe

5.1.14

> = 5/16"

b.4.1.a

> = 5/16"

mech.

> = 5/16"

13 Primayy,, contact wipe ___

_ . _

__

5.1.15_ 1/4";5/16"_ _ __ b 4.1.b_ 1/_4"-5/16"_

mech

1/4"-5/16"

14 Buffer block clearance to contact arm

5.1.15.1

> = 1/16"

b.5

inspect buffer blocks

_

.]j

ndry joliac[ gap

__

5.1.16

3-5/8"-3-15/16"

b.4.1.c

3-5/8"-3-15/16"

mech.

3-3/5"-3-15/16"

16 Trip latch wipe

5.1.17

3/16"-1/4"

adj

3/16"-1/4"

mech.

3/16"-1/4"

mech.

25 lbs.

16.1 Trip latch tension

5.1.18

1/16"-3/16"

b.4.1.d

1/16*-3/16" (1/32" otvil mech.

1/16"-3/16"(1/32o1)

,17 Trip armature travel _

_

5.1.10

3 /16"-1/4"

b.4.1.f

3/16"-1/4"

{ mech,

no tol. given

_

18 Release latch wipo

19 Release latch monitoring switch

5.1.20

< = 1/3 2 "

b.4.1.g

< = 1/3 2 "

mech.

< = 1/3 2"

20 Motor and relay switches

5.1.21

< = 1/32"

b.4.1.h

< = 1/32"

mech.

< = 1/32"

21 Interlock switch

5.1.22

< = 1/32"

b.4.1.i

< = 1/32"

mech.

< = 1/32"

b.6

21.1 C,ontro_t relay contact inspection

22 Oriving a_ndhtching pawl adjustment

5.1.23

> = 0.015"

b.4.1.)

> = 0.015"

mech.

> = 0.015"

5.1.24

< = 0.015"

adj.

< = 0.015"

moch.

< = 0.015"

23 Crankshaf t endplay

_

24 Prop pin cicarance

5.1.24.b > = 0.025"

adj.

> = 0.02 5"

mech.

> = 0.025

25 Latch checking switch

5.1.25

b.4.1.k

mech.

contact check

5.1. 25.e < - 1/16"

_

1/16"

mech.

app.1/16"

switch arm gap

5.1.25.g > = 1/64"

adj.

> = 1/64"

mech.

> = 1/16"

26 Plunger interlock-measured lif ting rail to plunger

5.1.26

11-7/32"-11-11/32"

b.4.1.1

11-7/32"-11-11/32"

mech.

11-7/32"-11-11/32"

27 Operating mechanism inspection

5.1.27

b.5

g/det,

w/ interlocks

28 Trip coil and armature inspection

5.1.28

sat /unsat

b.4.7

29 Low resistance (ductor) readings on each phase

5.1.29

x

eloc.

< = 54microchm

_ 30 70v trip test

5.1.31

70v

Ctrl Pwr

70v trip, 90v close

31 Charging motor check

5.1.32

lyrush 1/4" min.

6.19

brush 1/4" min.

FIGURE 9: PROCEDURE REVIEW MATRIX

-

- .

_

32 Arc-qnch cylinder, counter, aux.sw. plung:r check

5.1.34

set /unsat

g/dzt.

32.1 Manual / electrical stroke test

b.9

gen.

w/intsrlocks

.

3') St iker p atn inspection

5.1.35

sat / unset

34 Aux. switch contact check / inspection

5.1.36

g/det.

_

,

35 Reassemble

5.1.39

~

36 Hypot Sky phase-frame, phase-phase

5.1.41

x

elec.

> = 1 Mohm/kV

36.1 Opening / closing speed test

op.cl. test 14 fps cl.,15 fps op.

5.1.43

x

'37 Ending counter reading _

,

38 Secondary coupler check

5.1.44

0.038" gap

det.

39 Primary stab inspection

det.

~ dU insulation inspection

det.

li

Prop spring damage check

Cubicle inspection Section

- - - .

TASK

cKT8KR-G080 002 Rev.10

GEH-1802X

EPRI NP 7410 Aug. 93

. . _ . .

_.

step #

allowance

step #

allowance

step #

allowance

___

_ _ _ _

...__ _____._._.__ . . . _ _ . _

..1 Clean cubicle

5.2.1

m.1

1.1 Resistance to ground and phase-phase

m.2

? Fuse holder contact clip gap

5.2.3

grense method

3 Aux, switch mech. vert. and secure

5.2.5

3.1 CI'ock anchor bolts and control wiring

m.5

4 Wire connections on aux contacts

5.2.6

sat /unsat

_

_ S. Breaker position switch operator check

5.2.7

sat /unsat

G Breaker lif ting moch. inspect and lube

5.2.8

m.3

dot.

anti-croop pingr?

7 High voltage bushing check

5.2.10

8 Load terminal inspection

5.2.12

m.4

9 CT secondary termination check

5.2.14

1

10 Wire edge gaurd check

5.2.15

sat /unsat

11 Protective boot check

5.2.16

12 Local handswitch check

5.2.17

13 Interference block to interlock mechanism check

oper.

1/1 G"- 1/8"

mech.

1/16~-1/8"

14 Breaker lif ting rail to upper stop clearance

oper.

3/32"-1/8"

15 Primary stab wipe

oper.

mech.

start

< = 1/8" from top of ball mech.

< = 1/8" from top

_

Iength

3/4"-7/8"

mech.

3/4"-1 -1/32"

16 Spring discharge cam interlock

oper.

locked w/bkr < 1/4" up

17 Positive interlock motor clutch gap check

oper.

> = 1/16'

..

Lif ting bracket play

mech.

1/16"-1/8"

18

,19 Secondary disconnect seated

mech.

20 Lifting cradle width check bt notch)

21-1/4"-21-3/4"

FIGURE 9:

PROCEDURE REVIEW MATRIX

(Continued)

.

-

._______-_-_-__ ___- -

.

..

.. .

..

...

, . .

Acceptance Requirements

TASK

cxTaxR-o080-002 Rav.10

GEH-1802X

EPRI NP-7410 Aug. 93

step #

allowance

step #

allowance

step #

a!!owance r

Breaker and cubicle clean and tools removed

6.0.1

_1

Positive interlock roller check

G.O.3

in V Ond 1/16" gap

oper.

centered in *V"

2 Plunger / aux. switch gap

6.0.6

some to 1/8"

oper.

0*-1/8*

Note: all specs given for 1200 A breaker model AM-4.16-350-2H only

i

l

FIGURE 9: PROCEDURE REVIEW MATRIX

'

(Continued)

l

_

i

'

_._m__

____ _

..

,

-

.

DUANE ARNOLD ENERGY CE:NTER

SAFETY SIGNIFICANCE

1

i

TIM ALLEN

SENIOR DISCIPLINE ENGINEER, ELECTRICAL

i

t

i

.

!

I

DAEC

NCNEMBER19,1993

-

.-

.

-

._

.

. .

.

l

>

j;

-

.

,

DUANE ARNOLD ENERGY. CENTER.

i

SAFETY SIGNIFICANCE

EVENT TIMEFRAMES

i

L

ON-LINE EVENT (JULY 21,1993 TO JULY 31,1993)

,

'

- B' Diesel Generator Inoperable

l

- Plant was not placed in cold shutdown within

i

required time limits.

<

l

CORE OFF-LOAD EVENT (AUG 7,1993 TO AUG 11,1993)

'

I

- Both Diesel Generators Inoperable

- Fuel Movement (Core Off-Load) In Progress

l

.

!

DAEC

1

NCn/ EMBER 19,1993

!

I

l

_

.

ONLINE DIESEL AUTOMATIC TRANSFER INOPERABILITY

.

-

-.

.

l

ll' ll

'

--

-

,.

.

- ACTUAL

-

- -

-

. . euwr..

"

-

.

.:

.

'

S

I

-

-

.

'

.

!

-

!

-

l unas ;

,

-

.

'

=

-

TECH SPEC

'

^ Yr$

.'

.

' -

-W-

LIMIT-

/ /

,

i ?""""-

,

l

i

.

-

{

.

-

,

-

.

^.

i

,

.

-

i

.

-

.

.,

DATE:7/21 -

7/22

7/23

7/24

7/25

7/26

7/27

7/28

7/29

- 7/30

7/31

= 8/1

.-m r

---3w+.

-s,wmw%

- +

-

ws_,

-,-

si,-e-+

-r

r,.s

s,

,* _

-~e*+-

-er,w-=--w*s---==..--

w

ew,w.,-e,--

w-==--ww.+4en--e,-1...-%-

m-

---m-a

-+.-w

,o

-

-m

m. -+-m- - m

w---

-

.

.-

.

i

- .

,

i

DUANE ARNOLD ENERGY CENTER

-

l

ONLINE EVENT

1

!l

e Cold shutdown achieved on 7/31/93 exceeded

1

Tech Spec requirements by 1.2 days.

e The diesel generators are required durmg power:

operations for a Loss of Offsite Power (LOOP). Diesel

,

generator will automatically supply power to its

j

associated essential bus.

e Inoperable diesel as allowed by Tech Specs increases

j

the core: damage frequency by approximately 8 fold. .

~

j

i

e Exceeding the 7 day Tech Spec limit by 1.2 days increases

.

the probability of core damage by an additional

j

factor of17%.

i

,

e Probabilityis within the bounds of NUREG 880.

1

e Conservatively assumes the 'B' diesel generator cannot

!

be returned to service.

-

'

,

DAEC

NOVEMBER 19.1993

'

i

.

. . - . -

.

-

,

- .

,

1

!

DUANE ARNOLD ENERGY CENTER

l

SAFETY SIGNIFICANCE

ON-LINE EVENT (JULY 21,1993 TO JULY 31,1993)

i

- The diesel was capable of an automatic start

(rated speed and voltage within the allowed time).

'

- Diesel generator inoperability was due to. inability to

automatically reenergize the essential bus on a Loss of

offsite power.

'

,

- Output breaker could be manually closed from the

i

,

control room.

i

- A loss of the 1A4 bus would have been readily

}

recognizable by the operators.

t

- Existing procedures provide the necessary guidance to

reenergize the essential bus.

- A blind test was run on two operating crews. Both crews

identified the failure and reenergized the bus within 5-

minutes.

- Event-specific training has been completed for all

operating crews.

- 1992 licensed operator continuing training speci5cally

addressed 4160v distribution and the diesel generator

j

systems. This included the transfer permissives between

,

offsite power and the diesels.

,

- The 'A' diesel generator was capable of supplying 100%

of the emergencyloads required under design basis

1

accident conditions.

sov SEEt9,im

.

_ . .

.

_.

,

-

3

,

l

DUANE ARNOLD ENERGY CENTER

-

ONLINE EVENT

I

.'

e The calculated effect on plant safety even without

operator action was small.

e Operators had been trained on how to identify and handle

i

the loss of an essential bus prior to the event.

,

e Simulator tests demonstrated the operating crews'

ability to promptly restore power using existing , lant

,

procedures.

'

e The 'A' diesel generator was operable and capable of-

!

supplying 100% of the emergency loads required under

!

design basis accident conditions.

e

,

i

!

.

-

NOVEMBER 19,1995

.

..

.

DUANE ARNOLD ENERGY CENTER

CORE OFFLOAD EVENT

e Fuel offload commenced on 8/7/93 and was completed on

8/11/93: a total of 4.25 days.

e Tech Specs require one offsite power source, and one

diesel generator to be operable during fuel movement.

e 'A' diesel was out of service for planned maintenance.

e 'B' diesel was available with the exception of the capability .

of the diesel generator's output breaker to automatically

close in on the essential bus.

e The probability of a fuel handling accident concurrent

with a LOOP during the 4.25 days of fuel movement

(while assuming both diesels inoperable) is 6 x 10E-8.

This assumes no operator action.

e This probabilityis remote and does not significantly add

to the overall probability of a release.

e Secondary containment was intact and would isolate

on a LOOP.

The essential bus could have been reenergized from the

Control Room with minimal operator action.

DAEC

l

NOVEMBER 19.1PP3

i

$

4

-

.

?

DUANE ARNOLD ENERGY CENTER

CORE OFFLOAD EVENT

e Prior to RFO12 an independant shutdown risk review

of the outage schedule was performed.

e The risk factor during fuel move was only 4.5% of the

Phase 1 risk

planned main (even with the 'A' diesel inoperable for

tenance).

e With the 'B' diesel inoperable and no operator actions,

the relative risk during this time frame is increased to

only 4.83% of the highest risk phase of the outage (Phase 1).

e The increase is minimal due to the aredetermined plant

configuration during this phase of the outage:

- all 'B' side safety systems were considered protected

- the reactor cavity was flooded

- two fuel pool cooling systems were availaable

- both SBGT trains were operable

- both S/U and S/B transformers were available

- work in the switchyard was prohibited

e Due to preplanning, the LOOP scenerio only accounts

for approximately 2% of the risk during this phase.

DAEC

NOVEMBER 19,1993

.

i

-..

,

!

DUANE ARNOLD ENERGY CENTER

CORE OFFLOAD EVENT

.

Operators were well trained and kept wellinformed of-

current plant status.

- Briefed on the outage risk analysis study prior

to the outage.

- A daily safety assessment analysis was complete.

- Electrical systems were maintained in the " acceptable

risk" category.

- Daily time to boil calculations after a totalloss of

l

decay heat removal were performed (average tim.e

to boil 35-40 hours).

o

- The 'B' diesel was not operable, but was available and

could have been closed onto the bus from the Control Ro

l

P

' -

J

'

I

DAEC

~

NOVEMBER 19,19PS -

'!

1

..

,

t.

DUANE ARNOLD ENERGY CENTER

CORE OFFLOAD EVENT

'

s The probability of a concurrent LOOP and fuel handling

accident is low.

e The second inoperable dieselincreases the probability

of boiling to less than 5% of other acceptable phases

of the outage.

Operators were trained on the outage risk study prior

'

to RFO12.

e Operations was kept informed of daily outage risks including

equipment out of service and time to boil.

Outage preplanning and defense in depth minimized the

effect of the 'B' diesel inoperability.

e The diesel can be considered available and operator

actions taken into account.

)

,

wac

NOVEMBER 19,1995

,

.

-

.

. .

.

-

..

t

,

!

DUANE ARNOLD ENERGY CENTER

.

SAFETY SIGNIFICANCE CONCLUSIONS

'

!

Iowa Electric fully recognizes the importance of

maintaining our diesels in an . operable status.

o

e The inability of the 'B' Diesel to automatically

supply power to its essential bus had a small-

i

effect on risk even without operator actions.

l

e This failure was analyzed prior to the event and is

included in both the online and RFO12 risk

analysis.

1

u

e The operators had been trained on how to

recover from the event and.were well informe'd

i

of the current status of plant safety systems and .

time to boil during RFO12.

RFO12 risk assessment and preplanning

minimized the effect of both diesels being inop

during core offload using the defense in depth

philosophy.

l

4;

I

Y

'

ji

!)

-

NOVEMBER 19,1995

i

_.

..

.

o

-.

,

3

DUANE ARNOLD ENERGY CEMER

>

i

I

f

&

CLOSING REMARKS-

.

i

!

>

p

DAVE WILSON

PLANT SUPERISTENDENT

!

,

f

1

'

,

t

.l

.

DAEC

7

NOVEMBER 19.1993 .

,

h

1

. .

-

. .

.

..

.

_

..

-

_.-

. ..

,

DUANE ARNOLD ENERGY CENTER-

MI

G TI

G FACTOR

!

.

1

e SELFIDENTIFIED

e NO PRIOR OPPORTUNITY TO IDENTIFY

j

o PROMPTREPORTING

i

e COMPREHENSIVE SHORT-TERM

'

CORRECnVE ACTIONS

e ISOLATED OCCURRENCE

-

1

e STRONG SUPPORT OFINDUSTRY

NOTIFICATION

e PRIOR TRAINING, RISK MANAGEMENT:

AND EXISTING PROCEDURES MINIMIZED

1

EFFECTS ON PLANTSAFETY

.!

i

.

j

tue

-1u ,

.l

'l

..

.

- - .

. - - -

.

-

.

..

- :. .

DUANE ARNOLD EXERGY CEMER

ATmNDEES:

JOHN FRANZ

-VICE PRESIDENT-NUCLEAR

DAVE WILSON

-PLANT SUPERINTENDENT

MIKE MCDERMOTT-MANAGER, ENGINEERING

KEN PEVELER

-MANAGER, QUALITY ASSURANCE

JOHN BJORSETH

-MAINTENANCE SUPERINTENDENT

TOM GORDON

-SUPERVISOR,

ELECTRICAL MAINTENANCE

TIM ALLEN

-SENIOR DISCIPLINE ENGINEER,

ELECTRICAL

LARRY HECKERT

-PRINCIPLE LICENSING SPECIAIlST

,

NOVEMBER 19,1993

ML20059B313

Text

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