IR 05000458/1990007

EOP Insp Rept 50-458/90-07 on 900416-27.Weaknesses Noted. Major Areas Inspected:Adequacy of Eops,Including Implementation of Vendor Owners Group Emergency Procedure Guidelines in Developing plant-specific Guidelines

ML20043F095
Person / Time
Site:	River Bend
Issue date:	06/05/1990
From:	Cummins J, Gagliardo J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20043F085	List: IR 05000458/1990007 ML20043F084
References
50-458-90-07, 50-458-90-7, NUDOCS 9006140158
Download: ML20043F095 (31)

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I i-U.S. NUCLEAR REGULATORY COMMISSION

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REGION IV

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Inspection Report:

50-458/90-07 Operating License: NPF-47 i-

'E Docket:

50-458

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t Licensee: Gulf States Utilities Company P.O. Box 220

St. Francisv111e, Louisiana 70775 i

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Facility Name: River Bend Station

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i Inspection At:

River Bend Station, St. Francisv111e, Louisiana 70775

Inspection Conducted: April 16-27, 1990

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Team Leade -

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T. E. Cummins, Reactor Inspector, Operational Date

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i Programs Section, Division of Reactor Safety Team Members:. M. I. Good, Consultant

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I D. N. Graves, Examiner, Operator Licensing Section, Division of Reactor Safety W. B. Jones, Senior Project Engo..

Project Section D i

Division of Reactor Projects J. M. Keeton, Examiner, Operator Licensing Section, Division of Reactor Safety

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T. O. McKernon, Reactor Inspector, Operational Programs Section Division of Reactor Safety

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B. A. Paramore, Consultant F

$!f 76 Approved:

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w J. E. Gagliard,' Chief, Operational Programs DRe '

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Section, Di / sion of Reactor Safety, Region IV

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9006140158 900607

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PDR ADOCK 05000458-Q PDC

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l Inspection Summary: Special, announced emergency operating procedures inspection conducted April 16-27, 1990 (NRC Inspection Report 50-458/90-07).

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Area Inspected: This inspection was conducted to determine the adequacy of the emergency operating procedures (EOPs) and included the implementation of the i

vendor owners' group emergency procedure guidelines in developing plant-specific

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guidelines, technical and human factors evaluation of the E0Ps, validation of

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the plant-specific E0Ps by plant walkdowns, evaluation of the E0Ps using simulator exercises, and reviews of the E0P training program and engineering calculations.

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,Results:

The inspecting team concluded that the River Bend Station licensee had developed a set of emergency operating procedures that were generally in g

accordance with the Boiling Water Reactor Owners' Group Emergency Procedures Guidelines were technically correct and could be used by trained operators

utilizing existing equipment, instruments, and controls to mitigate the consequences of an accident.

However, the team identified a number of weaknesses in the development and implementation of the E0Ps, which were discussed with the licensee.

These included:

Some of the E0P entry conditions and decision blocks did not clearly state when the action was to be initiated or how to obtain the information to make the decision.

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.The E0Ps contained some erroneous information which resulted from errors in the calculation development process and the transfer of data to the E0Ps.

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Some of the methods used in developing the E0Ps, while they enhanced the E0Ps and made them user friendly, were not addressed in the E0P writer's guide.

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In some cases, tools and equipment necessary to perform E0P actions were not staged and dedicated for E0P use.

• Control room instrumentation design in some cases did not support quick and clear identification of E0P entry conditions and decision making.

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Training of some of the personnel that would participate in E0P response activities was weak,

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SUMMARY A Nuclear Regulatory Commission (NRC) inspection team evaluated the River Bend Station (RBS) emergency operating procedures (EOPs) from April 16-27, 1990. The inspection was conducted to verify the E0Ps; the capability to physically carry

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out the specified actions of the E0Ps using existing equipment, instrumentation, and controls; and the ability of plant staff to correctly perform the

procedures.

The inspection was conducted in accordance with the guidelines in

Temporary Instruction:2515/92, " Emergency Operating Procedures Team Inspection,"

Revision 1 dated July 5, 1989.

CONCLUSIONS The inspectors concluded that the licensee had developed a set of E0Ps that were generally in accordance with the BWROG EPGs; that were technically correct; and that-could be used by trained operators, utilizing existing equipment, instruments, and controls to mitigate the consequences of an accident.

The licensee documented all the items identified by the inspectors on E0P discrepancy sheets and stated that these items would be considered for any appropriate incorporation into the E0Ps at the next revision scheduled for the summer of 1990.

STRENGTHS

The operations staff was well trained and knowledgeable of the plant and the E0Ps and had confidence in the E0Ps.

The E0Ps, which were in flowchart format, were user friendly and had been

mounted on a stiff backing material and covered with a clear plastic that could be marked with a grease pencil so that place keeping was not a problem.

• Labeling of components in the' control room and in the plant was good.

Sound and comprehensive procedural guidance had been " tablished for

i ongoing evaluation and maintenance of the E0Ps.

WEAKNESSES

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Some of the E0P entry conditions and decision blocks lacked definitiveness in that they were not clear as to where to obtain the information necessary to make the decision or when the specified action was to be mitigated.

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The E0Ps contained some erroneous information which resulted from errors in

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the calculation development process and the transfer of data to the E0Ps.

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The team concluded that the licensee had effectively developed the flowchart

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format E0Ps. However, the writer's guide needed to include additional guidance to ensure continuing quality since some of the methods used in developinD'the E0Ps were not addressed in the writer's guide.

In addition, the writer s guide did contain sufficient guidance to ensure that the i

preparation and issuance of E0P revisions were appropriately controlled.

i The writer's guide did not contain guidance for the preparation and control of enclosures.

In. some cases, tools and equipment necessary to perform E0P actions were

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not dedicated, staged, or readily available.

• In some cases, control room instrumentation design did not support quick and clear identification of E0P entry conditions and decision making.

• Training of nuclear equipment operators and shift technical advisors was weak.

The nuclear equipment operators did not receive E0P training, and the shift technical advisor E0P training was not integrated with the operator training to assure that everyone involved in accident mitigation was trained as a team.

The licensee documented all the concerns identified by the team on E0P discrepancy sheets and stated that these items would be reviewed for appropriate corrective action.

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INSPECTION DETAILS 1.

INTRODUCTION The purpose of the announced team inspection was to evaluate the E0Ps. The team reviewed the E0Ps and the documents used to develop them, compared the E0Ps with the Boiling Water Reactor Owners' Group (BWROG) Emergency Procedure Guidelines (EPGs) and reviewed the licensee's justification for deviations from the guidelines, performed in plant walkdowns of the E0Ps, evaluated the E0Ps during the performance of accident scenarios on the site-specific simulator, and considered human factors aspects of the E0Ps during all phases of the inspection.

Findings from other inspection tasks were augmented and clarified through interviews with procedure users, developers, and other appropriate plant staff.

The tasks referred to in the report are the inspection tssks deteribed in TI 2515/92, Revision 1,- dated July 5,1989.

They are identined below.

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Task 1 - Basic E0P/ vendor generic technical guideline comparison Task 2 - Independent technical adequacy review of the emergency response guidelines Task 3 - Review of the E0Ps by control room and plant walkdowns Task 4 - Evaluation of E0Ps on the plant-specific simulator Task 5 - Ongoing evaluation of E0Ps Task 6 - Interviews with E0P users, developers, and other appropriate plant staff.

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FINDINGS 2.1 Comparison of E0Ps and the Owners' Group Emergency Procedure Guidelines (Task 1)

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t The inspection team compared the plant-specific E0Ps with the BWROG EPGs, Revision 4, to ensure that the licensee had generated procedures in accordance with the BWROG recommendations.

When deviations between these documents were identified, the team verified that the deviations were appropriately documented.

The team concluded that the BWROG EPGs, the PSTGs, and the E0Ps appeared consistent. The licensee provided sufficient justification for those deviations that were noted.

Furthermore, the licensee had appropriately established and implemented procedures for the development of the E0Ps from the BWROG EPGs, which included procedures for verification and validation and for an E0P writer's guide.

However, a number of weaknesses in the verification and validation program were identified. These weaknesses are discussed in Section 2.3 of this report and specific deficiencies are discussed below, i

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Plant-specific technical guidelines (PSTGs) did not have references to the plant-specific procedures incorporated for some steps as the BWROG EPGs had intended.

Examples included SP/T-2, page 26; DW/T-3, page 27;, PC/P, page 29;

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and SP/L, page 30 of PSTG Attachment 1.

The EPG step for SP/L stated, "...

enter [ procedure developed from Contingency #6]."

The PSTG for this step just i

restated the EPG step with the brackets removed rather than adding specific reference to a River Bend Station (RBS) procedure.

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l Although no documented justification for E0P deviation existed, E0P-3 contained an action that was not addressed in the BWROB EPGs or the PSTGs.

E0P-3, Step 13, stated, " Operate the following as necessary: Auxiliary building crescent area suppression pool pump-back lineup in accordance with System

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Operating Procedure (SOP)-0104." A review of $0P-0104 indicated that the

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prerequisite section to accomplish this task was titled, " Post-LOCA Operation of DFR Sump TK-5A and TK-5B With Discharge to Suppression Pool." The task assigned

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and the inconsistent procedure name could have led to the operators being

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confused about the actual procedure to perform.

E0P-1A, Step 80 required maintaining the containment water level between 62 feet and that shown Figure 2 (i.e., 86 feet with a containment pressure of 0-37 psig). However, the E0P did not address aspects of boron dilution that would result from the additional water sources used to flood the RPV and containment.

The E0P did not discuss or reference guidance for sampling boron concentrations, assessing core flux profiles, or precluding positive reactivity increases. The BWROG EPGs also failed to address these issues.

The licensee stated that this concern was being evaluated and, if appropriate, additional guidance would be provided.

2.2 Technical Adequacy of E0Ps and the Writer's Guide (Task 2)

2.2.1 Emergency Operating Procedures The team reviewed selected E0Ps to determine if they were technically adequate and appropriately incorporated the BWROG EPGs. The team concluded that the PSTG were technically adequate and accurately incorporated the BWROG EPGs.

Deviations from the BWROG EPGs were documented and justified. Although the team's findings were essentially positive in this area, certain deficiencies and discrepancies are discussed below.

Entry Conditions One of the seven entry conditions specified for E0P-3 was differential pressure

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above 0 inches of water. The EPSTG addressed this entry condition by referencing RBS Technical Specification (TS) 3/4.6.5, " Secondary Containment,"

which required that the shield building annulus, the auxiliary building, and the fuel building pressures be less than or equal to 3.0, 0.00, and 0.00 inches of i

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vacuum water gauge, respectively. The team reviewed the indications available

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to the control room operators.to determine if the E0P-3 entry condition had been

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met. A licensed operator stated that there were three indications that would

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alert an operator to the need to enter E0P-3 for differential pressure above t

0 inch of water.

These indications were altrms for containment-to-annulus differential pressure, auxiliary building-to-atmosphere differential pressure, and fuel building-to-atmosphere differential pressure.

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The auxiliary building-to-atmosphere differential pressure was annunciated in i

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the main control room. The auxiliary building differential pressure switch was

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to alarm at a pressure greater than -0.6 inch water gauge (i.e., -0.5 to 0 inch water gauge).

However, the licensee stated in Condition Report 90-0303,(dated

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April 6, 1990) that the delta p switch setting of -0.6 inch water gauge was too low for the normal and accident differential pressure (-0.25 inch water gauge)

maintained within the auxiliary building.

The licensee further stated that the alarm was not alarming during normal plant operation because the auxiliary building variable pressure leg was connected to the differential pressure

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transmitter high side. This meant that the alarm would not have occurred until 0.6 inch water gauge, af ter the entry condition for E0P-3 had been exceeded.

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The alarm setting for the fuel building differential pressure was set at-0.1 inch water gauge.

High winds and changing meteorological conditions

sometimes caused this alarm to activate spuriously.

The licensee had previously installed a 63-second time delay to dampen out the spurious alarms; however, the alarm still annunciated periodically.

The operators may become conditioned to invalid spurious alarms and not respond in a timely manner to valid fuel building differential pressure alarms, which must be responded to and validated

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locally.

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The team concluded that the E0P-3 entry condition for the 0 inch water gauge differential pressure was not identified by readily accessible indications i

within the main control room; the associated alarm response procedures did not alert the operators to the possible E0P entry conditions; the operators had not i

received consistent training on identifying this specific E0P-3 entry condition; i

and the PSTGs in some areas lacked definitive details in identifying proper parameters for the E0P entry or decision making condition.

Transition Step Steps 19 and 26 of E0P-2 stated, "... manually scram the reactor emergency RPV depressurization is required (EOP-1)"; however, there were no instructions which indicated at what step the E0P-1 procedure should be entered.

During the walkdowns, the operators showed some uncertainty as to what step the procedure should be entered.

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The E0P-3 initiating conditions for the alert and general emergency l

classifications were not specific enouch to allow the operator to easily determine when one of these conditions was reached.

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E0P-3 had an entr/ condition that stated, "offsite radioactivity release rate

abo"e the offsite release rate which required an alert (EIP-2-001)." Step 1 of E0P4 required the control operations foreman (C0F) to calculate if the offsite

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radioactivity release rate was above the offsite release rate that required an alert classification in accordance with Emergency Implementing Procedure

(EIP)-2-001, " Classification of. Emergencies," Revision 5.

This same type of calculation was required for a general emergency classification before the C0F could proceed through-Step 5 of the E0P.

I The conditions necessary-to meet the requirements for an alert cicssification i

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were confusing to the operators. The first condition was a high alarm on one of the three plant. ventilation exhaust monitors (i.e., the radwaste building

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ventilation exhaust [1RMS-REGA), the fuel building ventilation exhaust j

[1RMS-RESA], or the main plant exhaust duct [1RMS-RE125] and the summation of

releases exceeding 10 times the TS limit.

The wording regarding the summation

of releases was not exclusive of radioactive liquid releases that were in

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progress-at the time of the summation. The general entry condition for the j

alert classification was radioactive releases greatcr than 10 times the TS limit

aad referenced TS hetions 3.11.1 and 3.11.2.

TS Sectica 3.11.1 and

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e JSubsections 3.11.1.1 through 3.11.1.4 dealt with liquid effluents. TS 3.11.2 and Subsections 3.11.2.1 through 3.11.2.7 dealt with gaseous effluents.

The senior reactor operator questioned was not sure if any TS sections of liquid effluents should be considered in the summation.

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The EIP was not specific about which sections of the TS were being referenced.

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Because of the vague wording in the E0P steps, the operators may not have used the correct TS limit to determine when 10 times the TS limit was reached.

Step 5 of E0P-3 required the C0F to make a determination as to whether or not

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the offsite radioactive release rate was at the level of a general emergency classification.

Initiating Condition 1 in EIP-2-001 for a general emergency

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classification was met if the post-accident effluent radiation monitor confirmed i

noble gas and iodine release rates corresponding to i rem /hr whole body or 5 rem /hr thyroid at the site boundary for existing meteorological conditic?s.

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It was not clear which instruments were included in " post-accident effluent radiation monitor" because they were not identified by instrument number.

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Step 13 of E0P-3 instructed a licensed operator to initiate the suppression pool-pump-back system, as necessary in accordance with System Operating Procedure (50P)-0104, " Floor and Equipment Drain System." This system was designed to pump water from the auxiliary building crescent area to the

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suppression pool if a significant leak occurred from the suppression pool to tne auxiliary building. To perform the task in Step 13 of E0P-3 only Steps 5.1 through 5.1.2.1.b in SOP-0104 were required; however, these specific steps were not identified in the E0P.

Continuation of activities beyond Step 5.1.2.1.b of the procedure would have realigned the suppression pool pump-back system to the.

. liquid radiological waste holding tank instead of the suppression pool. The licensee said it would review the 50P for any needed clarification.

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The team reviewed selected plant-specific calculations.

In each case the licensee had properly reviewed the calculations and the conclusions were based on plant-specific data.

The team identified one instance in which the licensee had.not transferred the calculation results to the applicable E0P table. The licensee had previously identified ar.

'1 stance in which the effects on the E0P calculations-were not considered fer a plant modification. These instances are discussed below.

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Calculation-13.8.14*30-1, which determined the minimum core flooding interval (MCFI) relativ9 to the number of open safety relief valves did not co; espond to the MCFI given in Table 1 of E0P-1.

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However, the nonconservative time was for MCFI given.in the table that-would be used during reactor pressure vessel (RPV) flooding.

Step 7 to E0P-1 required the C0F to find out if RPV level could be-determined.

If the answer was no, the C0F would enter into RPV flooding at Step 57 of the E0P. The RPV would be subsequently depressurized and necessary water sources aligned to the RPV to establish an RPV pressure of at least 88 psig above the containment pressure. This would ensure that the -steam mass flow rate from the RPV was si.fficient to maintain the cladding temperature below

1500 F.

With sufficient pressure established and RPV water i

instrumentation available, RPV pressure would be maintained at least 88 psig.for the time specified in the MCFI table for the number of SRVs s

open.. This activity would ensure that, when all injection was terminated to the RPV and vessel level was lowered, the maximum cladding temperature

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would not exceed 15006 F for the times specified in Figure 3, " Maximum Core

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Uncovery Time Limit."

The. licensee initiated Condition Report 90-0361 to evaluate the' reason'the

" sample" plant-numbers were used in the MCFI table instead of the calculated values..Tnz licensee nad added an operator aid to Table 1 of E0P-1 to indicate the correct MCFI times.

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.The licensee had revised Table 2 of E0P-1A to be consistent with the changes made to the core loading plan during the December 1988 refueling outage. The change involved reloading the core with some higher enriched

fuel (i.e., a design for linear heat generation rate of 14.4 kW/ft in lieu.

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of 13.4 kW/ft).

However, the licensee had failed to recognize the need to revise the minimum alternate reactor flooding pressure for the RPV in

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Table 2 of E0P-1A before startup from the outage.

Subsequent analysis and evaluation by the licensee' determined that the two affected parameters were j

MARFP and maximum core uncovery time limit (MCVTL).

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Report No. 90-0322 and Calculation G13.18.14'.4*30-0 showed that the resultant revisions to'E0P-1, Figure 3, and E0P-1A, Table 2, did not differ to such a degree that a significant safety concern existed.

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The licensee took corrective actions to revise Table 2 values by issuing a

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temporary change through a memorandum dated April 12, 1990.

The licensee also noted.that E0P-1, Figure 3 did not change appreciably and its revision was not warranted..The licensee stated that further evaluation, review, and, where appropriate, incorporation into the next E0P revision would be

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Although it appeared that significant core damage would not be ' sustained,

core coolable geometry would be maintained, challenge to the ECCS criteria

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of 1 percent hydrogen generation would be avoided and the likelihood of

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occurrence would be minimal, the team concluded that further evaluation may

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be. warranted for using-four SRVs in. lieu of three SRVs. The licensee L

stated-that evaluation to determine long-term corrective action was ongoing.

t The licensee used the guidance of BWROG EPG (NED0-31331) Appendix C, dated December 7, 1987, for calculating the minimum number of safety relief valves (SRVs) to bei specified for emergency depressurization. A discrepancy was noted between BWROG EPG Appendix A (dated March 1987) and Appendix C.

- Appendix C. suggested-that the licensee calculate the minimum number of SRVs

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required to ensure that adequate makeup coolant from the low pressure core i

spray. (LPCS) system would be released to remove decay heat from the core and

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sustain peak cladding temperature (PCT) below a bounding limit of 1950 degree F.

However,= Appendix-A defined the basis for'this requirement as the capability of

.the lowest capacity emergency core cooling system (ECCS) to reduce the reactor pressure vessel (RPV) pressure enough, with a minimum number of SRVs, to remove decay heat from the core. However, the applicable ECCS at RBS is the

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low pressure coolant injection (LPCI) system, not the LPCS-system.

The licensee's analysis to determine the minimum number of SRVs required for f

emergency depressurigation (Condition Report No. 90-0335) showed that the most

e limiting case was one in which the operator attempted to depressurize the RPV to

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the-minimum alternate reactor flooding pressure (MARFP) while only one LPCI pump was available.

In this case, there would be a time lag during.which pressure dropped the MARFP level because the steam flow would be less than with the four

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SRVs open. However, when RPV pressure was maintained above the MARFP the steam

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flow was sufficient to sustain PCT below 1500 F.

The 1500' F criteria was the'

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. point at which some cladding perforation begins and provided some safety margin to the 1950 F criteria associated with limiting hydrogen generation.

In the above analysis, if the operator did not take the required rapid depressurization-6-

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Also, the licensee's probabilistlecrisk assessment studies, indicated that the probability of transitioning to the E0P step with only three SRVs open and only one LPCI pump available was about 10-E12-

. per reactor year, This scenario had a very low probability of occurrence.

Certain E0P procedure steps contained deficiencies that could result in delay or difficulty in implementing actions during emeroencies. These deficiencies are discussed below.

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The indication for Step 8-of E0P-2 could not be read directly and required

a conversion calculation.

The following problems with the implementation of Step 8 were noted:

No conversion table was provided to obtain differential pressure by

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subtracting annulus pressure in inches-water (vacuum or pressure on

, Instrument 1LMS-PT126) from containment pressure in pounds per square inch obsolute.

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The units of pressure, inches-water, were not identified on the

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instrument'and one operator had to look up the units in the system operating procedure.

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When the annulus mixing system and standby gas treatment systems were in operation, the annulus pressure instrument would be downscale and pressure would be indeterminate.

The indicator wa: not safety related and no backup indication was

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available if the instrument was out of service.

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The licensee said-it would address these discrepancies on an E0P

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_ discrepancy. sheet and installed an-approved operator aid during the

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inspection as an interim measure.

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Step 7 of EOP-1 referenced Enclosure 1, "RPV Water Level Trend Verification (Temperature / Temperature and Radiation Effects)," to the procedure to

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determine actual RPV water level if adversely high containment or drywell i

oli temperature levels were present.

The inspector noted that the RPV level instrumentation error correction _ graphs corresponded to containment jq-temperatures from 100 F to 350 F in 50 F increments.

However, the

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instrumentation provided in the main control area only indicated up to 200 F.

The licensee reviewed this finding and determined that the appropriate relay temperature units were displayed on instrumentation used to measure L

. temperature in areas such as the reactor water cleanup domineralizer room

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and that they could be used to measure containment temperatures up to

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350' F.< The licensee said it would review the need for additional-operator

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training on the appropriate locations to take containment temperature readings and on'the need for any additional guidance in the enclosure.

• Step 47 of E0P-2 required the operator to obtain information to determine if the site radioactive release level was below the limiting condition.for'

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operation level.

However, training material on how to obtain this.

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information was not included in the appropriate lesson plan (HLO-511-2,

" Emergency Operating Procedures," dated November.10,'1988). Operators had a range of methods to obtain the information,: including TS, the digital l

radiation monitoring system, and dose projection calculations.

Standard methods for determining this information should be developed and included r

in the training program.

Steps 27 and 65 of E0P-1 referenced Enclosure 8, " Injection into RPV With

SLC [ Standby Liquid Control). Test Tank," to the procedures to restore RPV level or if an additional water source was needed during RPV flooding.

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enclosure required an operator to open the breakers to the SLC tank suction valves (IC41*MOVF001A and B) and then.to fill the SLC test tank through the-test tank' discharge valve (1C41*F031) using the condensate system fill

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line. During the test tank fill operation,.the operator would be required-l to be in containment. Once the test tank was filled, the operator was

required to close the condensate fill line before starting the SLC pumps, s

The interlock between valves IC41*MOVF001A-and B and valve IC41*F031

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prevented opening valves IC41*M0VF001A and B when valve 1C41*F031 was open.

Therefore, opening the.1C41*MOVF001A and B breakers was an unnecessary action-with an additional time delay.

It also appeared that the condensate

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fill 1ine could be 1 eft open to the test tank-to supply a continuous source

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of water so that.an operator would not be required-to re-enter the containment to fill the SLC test tanks. The licensee initiated an E0P

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discrepancy report to review this: enclosure during the summer-1990 E0P review.

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Step 14 of E0P-1A required the reactor operators to stabilize RPV pressure

at a pressure below 1064.7 psig.

However, the control room instrument that

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would be read to obtain this information showed values in increments'of

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20 psig. The E0Ps also referenced other plant parameter values that were

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beyond the' accuracy of.the instruments in the main control room.

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However, the licensee stated that the use of the specific values in the E0Ps was appropriate because the values correspond to reactor scram setpoints and the operators had been trained extensively on these setpoints..The team concluded that the use of the specific values was

appropriate if the significance of the value was readily apparent to personnel using the E0Ps.

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I 2.2.2 Writer's Guide for E0P Development The team performed a desktop review of RBS Operations Station Procedure (OSP)-0009, " Author's Guide / Control and Use of Emergency Operating Procedures," Revision 4, and of selected E0Ps to assess their consistency with

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OSP-0009 and other sources of guidance for effective procedural information i

presentation (NUREG-0899,-1358,' and NUREG/CR 5228).

Discrepancies identified during the desktop review were evaluated during walkdowns and' simulator exercises.

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OSP-0009 guidare to authors (writers of E0Ps) was clearly written, consistent with establishbd principles of procedural information presentation using a big picture flowchart format, and well organized.

However, some details were omitted from the guidance.

In some of these cases, good practices were in use even though they had not been described in the OSP.

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' The-0SP-0009. guidance, for the most part, had been consistently implemented in the E0Ps.

Recurring deviations were found in only two areas: wording of logic statements and use of heavy lines around some blocks to indicate hold points.

The guidance on logic statements was correct,_although it omitted the distinction between "and" and "or" as conjunctions versus logic terms.

However, most logic statements in the E0Ps were not properly written. This was not a major problem area because logic statements were not extensively used in the RBS E0Ps, No confusion about the meaning of the instructions was observed, although the wording of steps in which logic terms were misused was found to be awkward and more difficult to follow. The practice of using different line weights

.(heavier lines) for coding was not in conformance with Section 8.1.2.of the OSP and_ operators were not all aware of the line weight difference and its -intended meaning. The licensee stated that the use of different line weights would be eliminated in the next flowchart' revision (summer of 1990).

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.The team concluded that the licensee had effectively prepared the E0Ps,-but that Procedure OSP-0009 needed to include additional guidance to ensure continuing quality. The_following items, which the' team perceived to be weaknesses in the

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guidance, were' discussed with the licensee.

The licensee stated that the items

_ would be~ reviewed and resolved during the next revision of-Procedure OSP-0009 and. considered for. the E0Ps scheduled to be completed in the summer of 1990.

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L rocedure OSP-0009 did not describe the total process required to develop P

or to revise an E0P although Section 6 addressed E0P revisions and updates.

For example: The OSP did not specify the responsibilities of the E0P coordinator and the other personnel involved in the development process.

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-It did not explain the relationships of the BWROG EPGs and the RBS PSTGs to'

the E0Ps or how the E0P preparer was to use the generic and plar.t-specific-technical guidelines. The OSP did not specify that tha procedure preparer was. responsible for completing the safety-evaluation (10 CFR 50.59)

checklist or any other forms required to be used in the development-9-

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It did'not indicate how verification and validation (V&V) fit

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into the overall process-or clearly refer to Procedure OSP-0008, J

" Verification.and-Validation'of Emergency Operating Procedures,"

l Revision 2,: for-V&V. requirements.

Finally, it did not identify the

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documentation required to be' retained for an E0P. revision or for a new E0P.

i The need for' requirements-regarding the retention-of documentation had been noted in the licensee's E0P audit report (P-90-03-012, March 1990). To i

resolve this last concern, the licensee issued Temporary Change-Notice (TCN) 90-0353 on April 23, 1990. The E0P development and revision process as to be provided in the summer-1990 revision of Procedure OSP-0009, t

Procedure OSP-0009 stated that revised E0Ps would be reviewed and issued in accordance with Administrative Procedure ADM-0003, " Development Control and Use of Procedures," Revision 17A, and Procedure OSP-0008.

However, cross j

. referencing among these three= documents was insufficient to ensure correct integrated-implementation of the guidance from all three.

For example, Procedure ADM-0003 did not indicate when there were special requirements

- for E0Ps (except for the periodic review) and it did not refer to

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Procedures OSP-0009 or 0$P-0008.

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Procedure OSP-0009 did not give guidance and there was no other source of

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J guidance for the preparation of enclosures to operating procedures. There was no guidance as'to when an enclosure was needed, how to identify and

paginate-an' enclosure. The required content and sections, wording of enclosure step instructions, format and style, and particular information requirements (e.g., component identification and location'information) were

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not described.

It was not clear if enclosures were to be considered part-of the. flowchart E0P regarding the requirements for development, review,-

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'V&V, approval, and issuance.

Although existing enclosures generally were found to be clearly and

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r, consistently written there was nothing to ensure that this would continue C

to.be-the case. The licensee's E0P audit in March 1990 had identified the

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L need for guidance concerning E0P enclosures in Procedure OSP-0009 and the licensee issued TCN 90-353 in response to that audit during the inspection,

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~Although the. intent of the TCN was correct,'it did not provide sufficient guidance. The licensee stated that the needs for guidance on enclosures would be re-evaluated for the summer 1990 revision.

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Neither Procedure OSP-0009 nor Procedure ADM-0003 addressed temporary

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Jchanges to the E0Ps and enclosures. The practice was that temporary

changes to E0Ps were not allowed; but this practice was not stated in these procedures. Although it was very important to carefully control changes to g

E0Ps a method was needed to make an essential change outside a major-

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revisionicycle.

This need occurred twice during the inspection, and the

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licensee made E0P' changes by means of approved operator-aids.

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Several conventions used in the E0Ps were not defined in

Procedure OSP-0009.

For example, the use of a color other than red to-delineate override decision. steps and the practice of listing all

enclosures in the E0P at the bottom of each flowchart.

The 1 administrative / control'pages (approval sheet, table of contents, etc.) that were included with each E0P were not mentioned in Procedure OSP-0009, Also, it was.not made clear that E0P-1A was administrative 1y part of.EOP-1;-

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there were no administrative / control pages for E0P-1A.

The. licensee stated that these omissions had been recognized and would be corrected in the summer-1990 revision of Procedure OSP-0009.

During the inspection, the licensee issued TCN 90-0353 to clarify the relation between E0P-1 and

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E0P-1A.

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Procedure OSP-0009 did not address using notes as a specific format for

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providing information.

Note-type information, when presented, was included

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in a caution statement (e.g., Cautions 5 and 6) or an action step

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Linstruction (e.g., EOP.2, Steps 19 and 26).

Combining different types of information could make both the note information and the step or-caution information ambiguous, as observed during the walkdown of E0P-2, Steps 19 and 20.

The inclusion of note information concerning emergency

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depressurization caused some uncertainty among the operators with regard to where to. enter E0P-1.

(These steps also' contained reference instructions, r

which added to the problem.)

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' Procedure 0SP-0009, Enclosure 1, included a symbol to be used for an

"information step" in the E0Ps to present note information.

However, the E0P authors did not use this symbol.

The licensee stated that the information symbol. would be incorporated during the next: revision of the

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E0F and that the: guidance in Procedure OSP-0009 would be augmented to explain the use of information steps.-

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Procedure OSP-0009 did not require the use of a special symbol for

-reference. instructions, i.e., instructions to go to another procedure and

perform it concurrently (in whole or part) with the present procedure.

Reference instructions were mixed in with other action steps.

The licensee stated that use of a separate step, indicated by a special symbol, would be implemented in the next E0P revision and that Procedure OSP-0009 would be Lrevised to provide guidance to the E0P author on this topic.

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Procedure OSP-0009 did not require identification of a specific point of entry in referencing and branchfng instructions and did not provide any

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general rules such as, "for cited E0Ps, go to the symptoms / entry conditions

unless otherwise speriVied." During the walkdowns, operators showed some

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-uncertainty about where a' referenced procedure should be entered (e.g.,

-E0P-2,'StepsL19 and 26).

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.In some cases, when.another procedure was to be performed concurrently with.

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the present. procedure, the entire referenced procedure was.to be performed.

.In other cases, only a section of the referenced procedure was to be

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performed..The licensee stated that the practice was to list within the-E0P step both the referenced procedure number and the section title of the section to be used.. This practice was found to be followed in most cases, but:it was not explained in Procedure'0SP-0009.

In E0P-3, Step 13, this practice was not followed, which resulted in confusion about where to enter

'the referenced procedure.

There also was a question about how the-operators would know whether or not the entire referenced procedure was supposed to be performed.

The licensee stated that specific step or section numbers were omitted so S

that-an. E0P would not have to be revised just because of a change in a cited procedure. The team believed that this type-of change would not occur so frequently that:it would -justify trading clarity of the-instructions. The licensee also commented that a change affecting the step or'section number in a cited procedure could be overlooked, resulting in creation of'an error in the E0P.

However, there should be an established process to ensu're that any change to a cited procedure would be evaluated l

for effect on the E0P in which it was cited.

Procedure OSP-0009 provided no guidance to review referenced procedures to

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. ensure that_the applicable steps were clear, accurate, and' appropriate for-the-emergency. conditions. 'The need for this kind of guidance was indicated by an E0P discrepancy sheet concerning'the relationship between E0P-1, c

Enclosure 24, and SOP-0059, which was referenced in Enclosure 24.

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example was SOP-0065 which was referenced in E0P-1, Enclosure 23, but did not provide' appropriate' guidance for the E0P situation.

Procedure OSP-0009 guidance on caution statements did not specify that-there.was a standardized set of cautions and that the numbers for the L

caution statements did not change.

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The guidance on caution statements did not clearly define the_ practice for-

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placement of caution statements. All caution = statements appeared at the bottom of the E0Ps, flagged by a circled caution number, wh_ich was color-i-coded to the step to which the caution applied.

This method of o'

incorporating caution statements into the E0Ps appeared to be an

enhancement;.however, Section 4.3.7.2 of Procedure OSP-0009 stated that caution statements "where convenient, will be included in the body of thel flowchart." It would be undesirable to mix the placement of caution

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Procedure OSP-0009, Section 4.4.5, which provided the guidance on logic-l

terms, did:not discuss the difference between the use of "and" and "or" as

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-logic terms or their use as ordinary conjunctions. A need for this

' guidance was indicated in the E0Ps. There were a number of steps in which

"and" and "or" were treated improperly as lo'gic terms (e.g., E0P-1, I-Steps 8,.13, 22, and 29; E0P 1A, Steps 5, 45, and 47). This was not observed to cause any confusion during the walkdowns and simulator exercises, but it detracts from the consistency of usage which would have

helped to make logic statements easier to understand.

Procedure OSP 0009-Section 8.0, contained incomplete guidance on flowchart-

• production.

For example, it did not include specifications of type, size, and style,

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j-2.3. Evaluation of E0Ps by Control Room and Plant Walkdowns (Task 3)

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.The team conducted detailed walkdowns of all selected E0Ps, E0P enclosures, and applicable sections.of supporting procedures referenced by the E0Ps, The team asked operators to locate and demonstrate or simulate the demonstration of equipment necessary to. carry out the E0P. Additional walkdowns were conducted

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to resolve questions and to evaluate the ability of a cross section of operators to' implement the.EOPs.

Shift supervisors, shift foremen, reactor operators, and nuclear equipment operators participated and the team shared questions and concerns with the licensee during walkdowns and dailv briefings. The human factors specialist evaluated the human factors asm.ts of the E0Ps during selected walkdowns, a-The team concluded that with.noted exceptions the operators were able to execute l-or simulate executing:the E0Ps and supporting procedure steps.

Most steps and

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actions were. carried out with little difficulty or assistance.

The

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implementation'offkeylock switches to accomplish jumpering activities required

'by some enclosures-was.a strength.

The color coding on the flowchart E0Ps was a strength.and provided visual clarity for the operators. The inspectors noted j

the relay and jumpering locations were well specified by the procedure and with

few exceptions the operators were knowledgeable of the execution of enclosure steps. The exceptions generally involved deficiencies that should be corrected

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1but would not prevent implementing the E0Ps in an emergency.

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'Asstatidearlier,someE0Psreferencedlengthyproceduresratherthanthe appropriate.section necessary to implement the E0P.

Referencing the entire-procedure would place an additional burden on operators during emergencies.

For

' example, E0P-2, Steps 17 and 23, referenced the entire 100 pages of

Procedure SOP-0031, " Residual Heat Removal," for raising and lowering suppression pool level instead of the appropriate section of the procedure.

Only a few steps of the procedure would be used to implement the E0P actions.

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'i The. inspectors identified inconsistencies'in labeling between equipment and procedures? Some equipment was labeled differently than indicated in the

procedure and some indicators had labeling deficiencies.

For example:

E0P-2, Enclosure 28, " Bypassing Primary Containment Purge Isolation Interlocks," Step 2.2.8, identified the timer as 1HVR*A0V123 while the-timer was labeled 45-1HVR20.

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.EOP-2, Enclosure'21, Step 2.2.6, identified as " Hydrogen Purge Outlet to Annulus ICPP*MOV105" while it was labeled as " Hydrogen Purge Fan. Discharge Valve'1CPP*MOV105."

Hydrogen concentration instrument ICMS-AR25A & B on Control Room Panel P808'

.had two low-range scales for reading hydrogen concentration, which could

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cause unnecessary confusion during an emergency.

The.. instrument was used to r

determine E0P-2, entry condition, " Hydrogen in Containment or Drywell above

0.5%."

A single pointer was used for a 0-10 percent and a 0-30 percent scale.- During wudowns, operators were not positive which scale was in use. The team subsequently found-that the simulator. instrument for the same indicator only had a single scale of 0-10 percent.

The licensee-stated that labeling deficiencies.would be documented on E0P deficiency sheets and corrected by maintenance work orders or modifications.

i The team identified discrepancies with E0P enclosures that could result ln-delayed or difficult implementation of action by operators, The discrepu cies included training, staging of equipment, and inadequate equipment labelini.

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example:

During walkdowns of Enclosure 11, " Venting Scram Air Header," the nuclear i

equipment operator had difficulty finding both the C11-VF095: instrument air supply pilot valve and the 1-C11-N052-V2 pilot air header pressure transmitter root valve.

The difficulty was partly the result of training and partly the result of the small labels identifying the valves.

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Additionally, there was no emergency lighting in the area, which would have made the' performance of even routine tasks difficult during adverse ennditions.

During walkdowns of Enclosure 15, " Alternate SLC Injection and SLC TK Gal

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to LBS Conversion," the required dedicated equipment was not prestaged and -

readily available for use.

Step 2.1 directed' mechanical maintenance personnel to obtain equipment for alternate SLC injection.

The equipment included a hydrostatic test pump, an empty 55 gallon drum, hydro pump suction hose, hydro pump discharge hose, adapter fittings, and a sufficient

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quantity of sodium penta-borate solution. Mechanical maintenance personnel stated that dedicated prestaged equipment was not available and that it-would take some time to assemble the necessary equipment for the task.

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Enclosure 15 specified injection via the alternate path,'wh'ich would be

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much slower than the SLC pumps.(high head'about 50 gpm) because.of.the line

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size and pump head.,The alternate injection path was through 1/2-inch tubing as opposed to 2-inch line for normal injection.

Enclosure 15 also

did not provide guidance to determine how much boron had been injected by the alternate method.

The amount of boron injected was necessary to satisfy E0P 1A, Step 4, "Have 111 lbs (enc 1 15) of Boron Been Injected Into t

The RPV."-

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_During walkdowns of Enclosure 17, " Venting CRD Overpiston Volumes," did not -

identify a special T-handled wrench required to support the task.

There-was no' ladder staged in or near the area even thoughLone would be needed, flatform supports were installed in the overhead but there was ne grating installed for,the platform.

In addition, there was no emergency lighting in,the-area-to support performance of the tasks.

• Electrical. jumpers needed to perform action in Enclosures 2, 3, 4,-5, 21, 22, 23, 24, 25, and'28 had wire clips that would be difficult to attach to

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relay terminal connections without. bending the installed wiring.

The clips were " mini grabber" type clips designed for use on bare wires rather than

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terminal lugs.

River Bend staff indicated that E0P verification i

activities, which included attachment of clips, had not been conducted.-

Attempts to-attach the clips to a shop mock-up indicated terminal lugs may have to be-bent to attach the clips properly.

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Enclosure,22, " Bypassing Primary Containment Chilled Water Isolation i

' Interlocks" specified positioning the lifted lead out of the way of

'p'ossible contact with other components.

However, n'o instructions were provided for' insulating the lifted lead or otherwise ensuring it'did noti

short other energized circuits..Several other enclosures-contained thet

same-wording.

Bare leads'within safety-related electrical enclosures could i

!cause inadvertent. system or component realignment.

i Enclosure 22, Step 2.1.2, required lead removal and jumpering on

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RelayL3A-3-11SCA01, the relay was obscured by other panel-components making the task very dif ficult and unsafe. The action would have been'further ccomplicated by the use of the mini grabber clips _ explained above.

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addition, relay locations for other enclosures requiring the bottom row of relays to be jumpered would be difficult.

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' Enclosure 22 called for a No. 3 regular blade, stubby screwdriver.to be s

included in the jumpering kit. Both the blade size and length were.

incorrect for the small relay terminal screws _it was intended to be used-on. The licensee-placed an additional screwdriver of the proper size in the kit.

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2.4-E0P Evaluation Usina the Simulator (Task 4)

- The licensee provided two operating shif t crews to assist in the plant-specific -

simulator' exercise responses to accident scenarios.

The team was able to evaluate the capability of the operators to use the E0Ps to-successfully

mitigate the consequences of an accident. The placekeeping method used by the

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operaters while making transitions between procedures and for maintaining an awareness of the steps completed within procedures was assessed, as was the adequacy of training on procedures, procedure flow, and actions taken.

Furthermore, the team was able to observe physical limitations or hindrances in.

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completing procedures and the adequacy of the information provided in the E0P for completing the correct procedural step.

The team concluded that the flowchart E0Ps could be used to mitigate the

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accident scenarios demonstrated on the simulator. Occasional errors in

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executing the flowchart EOPs -were made by the crew during the scenarios; however, the required results were attained in all cases.

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Although accident mitigation required implementation of multiple enclosures to the E0Ps, the team could not evaluate the operators' execution of enclosures

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during the scenarios because the time involved in performing the actions delineated in the enclosures could not be determined.

The table space available to lay out the flowcharts was limited in the simulator

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room. When entering more than one E0P concurrently, the control operations foreman (C0F), who'was directing the E0P response, found it difficult to lay out the flowcharts.

The control room had less space for E0P layout than the

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simulator room.

The role of the shift technical advisor was not adequately defined in the E0P.

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'The oversight function maintained by each of the three STAS appeared to be t

clearly inconsistent. The STAS' actions varied from providing minimal input to the C0F on E0P decisions to that'of being involved with the details of the E0Ps and the E0P enclosures. The specific role of the STA during emergencies should

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be evaluated and defined.

Integrated training involving all personnel who would

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participate in control room accident response would enhance the control room Lemergency response effort'during a real. emergency.

The C0F was inconsistent with regard to the point at which the referenced E0P was' entered.

In one instance, E0P-1 was entered at the beginning and in another l

instance'the E0P was entered in an override step.

There was no clear guidance

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to indicate the specific place within an E0P that should be entered when transitioning from another E0P.

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L During.a scenario involving a blackout, the Division III diesel generator was to run for the duration of the scenario although it was not being used to supply cooling water.

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. preclude this from happening during accident response.

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The team noted.several human. factors deficiencies in.the control room and t

-_ simulator room that could have a negative effect on-training or emergency-

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. response.'

.The:residualLheat removal (RHR) service water valves and meter indication

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of RHR' service water flow were on different panels; When the operators adjusted flow through Heat Exchangers A and B by controlling Valves E12F6068A or B, RHR HX SVCE WTR RTN, on Panel 1H13*P870, the flow s

had to= be read on Panel 1H13*P601, which was on the opposite side of the -

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- control room. Operators said that they had difficulty reading the meter while adjusting the flow.

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The plant monitoring system (PMS) had suppression pool level indication, but the control room PMS did not. During exercise scenarios, the reactor operator used the suppression pool level on the plant computer, which provided a reading that would not be available in the control room.

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The actual control room design as well as the. simulated design of the annunciator system did not provide an indication as to where the auditory signals were located (i.e., there was only one horn and a stereo effect was produced). - During simulator _ exercises, the operators had to search the control room to identify incoming alarms._ In addition, it was not'possible-to silence the auditory signal from any set of annunciator response d

controls in'either the control room or the simulator room.

These design deficiencies caused distraction from tasks, unnecessary movement from panel-to panel, unnecessary noise, and greater difficulty in identifying alarm conditions.

Furthermore, the control room auditory signal (ringback) 'for a cleared alarm condition had to be manually silenced just like an incoming

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alarm., Ringback should be a sound clearly distinguishable from the auditory signal for an incoming alarm, and ringback should stop automatically after a fixed time interval. The visual signal for cleared

alarm conditions should remain in place until the alarm was reset, not the

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auditory signal.

To correct these alarm problems, the licensee had developed Modification Request (MR) 86-1132 in 1986, but it was subsequently cancelled. Although Procedure MR 86-1132 provided for an indication of the location of auditory-signals and the silencing from any annunciator response control set, it did r

not address the problems with the auditory signal for a cleared alarm which also needed to be evaluated. The licensee stated that it would consider

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reactivation of MR 86-1132.

Another difficulty with discriminating between incoming alarm was that all annunciator alarms did not print out in the control room.

Since the-17-

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operators needed a record of which alarms were activated, they could not reset alarms-during a transient.. The licensee had requested that'the

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computer group determine what change would be required to provide a

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printout of annunciator alarms and had planned to evaluate this problem..

2.5 E0P Training

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i Initial licensee operator training was conducted using Lesson Plan HLO-511-2,

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" Hot License Operator Training." This training was primarily in a classroom

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setting and was approximately 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> in length. Practice using the E0Ps was obtained during.the simulator portion of operator training.

The simulator-training was approximately 6 weeks long und was divided evenly between classroom fand simulator exercises.

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While'this provided training on document development, bases, figures, steps, and

use, it did not provide for specific E0P training on the simulator or for specific. training on performance of the E0P enclosures. The licensee stated that a major revision to HLO-511-1 was planned for the summer of 1990.

However, the revision was not intended to expand the scope of the lesson plan to include more in-depth training on the E0P enclosures, Simulator scenarios used for E0P training were developed based on the scenarios used for E0P validation. The scope of the scenarios was expanded to include

. training on normal'and abnormal procedures in addition to emergency operating procedures.

The reviewed scenarios were comprehensive and provided training on all: E0P paths.

Procedure OSP-0009, stated that nuclear equipment operators (NE0s), maintenance, or. technical personnel may be required to perform enclosures to the E0Ps outside the control-room or on "back panels." Training for NE0s on enclosures out:,ide the control room was in progress and was scheduled to be completed by June 2E,

1990.

However, NEOs.did not receive training on enclo:ures performed inside the

control room, and maintenance and technical. personnel did not receive training

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on the performance of any E0P enclosures.

The. licensee needed to ensure that j

sufficient' training was provided on E0P enclosures so that any individual

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allowed to perform an enclosure.had the necessary knowledge and ability to perform the task.

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In addition, operators indicated that they had not received-training on where to lobtain specific information (e.g., indicators, meters, or values) needed for

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certain E0P steps.

Procedure HLO-511-2 did not contain information or objectives for determining which indicators or meters should be used for making EOP decisions based on parameter values.

i Frocedure OSP-0009, Step 6.2.2, stated, that, "If the revision is significant,' a special training session should be requested." The person responsible for-18-

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determining whether a revision was significant and the objective criteria for

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determining what was-significant was not identified.

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The STAS were trained separately frun the operating crews, which may have

. i contributed to the inconsistent perfonmance of the STAS during the E0P_ simulator

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evaluation.

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2.6 Licensee's Ongoina Evaluation of E0Ps (Task 5)

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Section 6.2.3 of NUREG-0899, " Guidelines for the Preparation of Emergency l

Operating. Procedures," requires that licensees establish a program for the ongoing evaluation of E0Ps including the evaluation of the technical adequacy of

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the E0Ps on the basis of operational experience and use, training experience,

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simulator _ exercises, and control room walkdowns.

Section 6.2.4 of NUREG-0899

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specifies that processes'should be established to ensure timely revision of E0Ps based on input from these evaluations and from assessment of the effects on the E0Ps of. design modifications and changes to technical specifications, technical

guidelines, and the writer's guide.

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The team assessed the licensee's program for the ongoing evaluation of the E0Ps

based on a review of the procedures that defined program elements, review of i

other documentation, and interviews. The inspectors found that RBS-had established a program that included the elements necessary to maintain j

technically adequate and usable E0Ps over time.

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The team's principal. concern about the program for ongoing E0P evaluation was-a

'the deficiencies indicated.in the transfer of information related-to E0P

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calculations and changes in_the plant..The following three problems of this

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kind.were identified just before.and dur_ing this inspection:

The. licensee failed to recognize that a change in fuel bundles affected the

xEOP calculations, and the'information in the E0Ps was erroneous (Condition j

Report No 90-0322 and Section 2.2.1 of this repo.t).

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Emergency procedure guideline methodology interpretation were not j

reevaluated after issuance of updated EPG guidance (Condition Report l'

No. 90-0335 and Section 2.1 of this report).

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The licensee failed to incorporate calculation results into an E0P

flowchart (Condition-Report No. 90-0361 and Section 2.2.1 of this report).

The team agreed with the licensee's analysis that the root causes of these problems was inadequate utilization of change documents (e.g., modification request and technical specification change request)~and inadequate procedural i

guidance to ensure that calculation results become E0P inputs, j

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There was no applicable guidance when the two of these calculation-related:

problems arose.

The licensee subsequently issued several procedures with appropriate-provisions. They include Engineering Procedures EDP-AA-19,

" Preparation of Engineering Input-for Emergency and Abnormal Plant Operating-Procedures," Revision 3; ENG-3-020, " Control and Use of Plant Specific Technical l

Guidelines," Revision 3; and Procedure OSP-0008.

The-licensee stated that it-Would review and revise Procedures EDP-AA-19 and ENG-3-020, as necessary, to-ensure that the effects on the E0Ps of changed documents were. evaluated.

This

~ ill be completed in conjunction with the next E0P revision.

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The:following= issues needed to be considered by the licensee in the ongoing

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program for EOP evaluation:

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The position of E0P coordinator and the coordinator's responsibilities were L'

not identified in any program guidance.

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The method for operations, and other personnel,.to raise questons and request changes to the E0Ps was not spelled out in any of the guidance documents.

The practice was for personnel to use the comment control form (

that was the method used for other types of procedures; the form was L:

available _in the control room. However, there was another practice by L

which a proposed change to an E0P would be documented on an E0P discrepancy

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-sheet. This _ form was identified only in Procedure OSP-0008, concerning E0P verification and validation, and the form was not readily available to I-operations and training personnel. The E0P-coordinator prepared E0P discrepancy sheets from the information provided on comment control forms.

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. Procedure OSP-0009 did not mention either form or the need.for the-EOP preparer to address these forms during a revision.

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Procedure OSP-0008 did not clearly require verification and validation of enclosures, and the verification checklist in the: procedure did not provide guidance:on special issues that need to be addressed in' verification of s*

' enclosures (such as staging of tools / materials; provision of component

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lt identification information and consideration of lighting, component t

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accessibility, and hazard issues that may be associated with performance of

tasks:in the plant). The inspection walkdowns indicated that there were

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j, deficiencies in the verification of enclosures.

Procedure OSP-0008 did not require verification of the ease of-accurate

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. entry into supporting procedures (such as A0Ps and SOPS), the

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appropriateness of supporting procedure steps for the emergency conditions, or provision for the staging needs associated with supporting procedure p

steps. The verification checklist did not address these issues.

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. Procedure OSP-0008 provided no guidance on the process that should be used

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to check calculations and setpoints, and there was no training for-l personnel responsible for E0P verification.-

Problems related to

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verification were identified by th'e licensee's E0P audit conducted in March 1990'and reported in Quality Assurance finding Report QAFR P-90-03-018, Procedure OSP-0008 did not provide guidelines for determining when

validation of an E0P revision was required or-for the. type of validation that should be conducted (e.g., simulator, walkthrough, tabletop).

The modifications request procedure provided for a review of modifications to determine their effects on E0Ps; however, there were some types of-changes made outside the modification process (e.g., fuel changes) that were not addressed. The licensee stated that an analysis was underway to-

determine how to ensure that all changes related to plant design and

' operating criteria would be evaluated for impact on the E0Ps.

The licensee-had used spot checks on E0P items through operations quality assurance' surveillance: actions to ensure the quality of the E0P program until this year.

In March 1990, a full-scale E0P audit was conducted by the quality assurance systems-group, which resulted in identification of a number of

deficiencies.that were documented in QAFRs.

Corrective action recommer.dations

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had been developed and the licensee was in the process of reviewing and.

approving.these corrective actions. One of the QAFRs resulting from this audit

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identified the need for more frequent and thorough quality assurance review of

.the E0P program.

Therefore, the licensee decided to include an annual audit of the E0P program in the master audit plan with a followup audit' scheduled for 6 months after the-initial-audit of March 1990.

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3.

EXIT MEETING AND PERSONS CONTACTED

On Apr11.27, 1990, the team held an exit meeting with Mr. J. Deddens and other members of.the licensee's staff and discussed the scope and findings of the t

inspection.

Persons contacted by the team and attendees at the exit meeting are

l identified in Attachment A.

The' licensee did not identify as proprietary any of the material provided to the-team during this inspection.

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ATTACHMENT A:

I PERSONS CONTACTED AND EXIT MEETING ATTENDEES s.

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'Name'

Title:

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- "D. : Andrews-Director Nuclear Training

'T.:Autrey Training-Instructor

• R. Backen Supervisor Quality Assurance Systems

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• J.. Booker Manager Nuclear. Industry Relations

• J. Bowlby Operations Shift Supervisor

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• J. Burton Supervisor Probalistic Risk Assessment-

• J.

Cook Technical Assistant

• B. Curran Cajun Electric Site Representative t

• T. Crouse Managar Administration

• J; Deddens Senior Vice President

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G; Degraw'

Training Instructor

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• L. England Director Nuclear Licensing i

• A. Frelieu=

Operations Supervisor S. Fiore Reactor Operator

P. Graham-Plant Manager D. Grimes Quality Assurance

• R.~~ Jackson Training Coordinator

'*J. Miller Director Engineering Analysis

• J. Peters

' Contractor, River Bend Station

• W.-Odell-Manager Oversight

• M..Sankovich Manager Engineering Department

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• J. Schippert Assistant Plant Manager, Operations, Radwaste & Chemistry

• K Suhrke General Manager Engineering and Administration

.R. Westley Simulator.0perator

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"The. inspectors also contacted other members of the licensee's staff'during the I

inspection. period to discuss identified issues.

• Denotes those personnel in attendance at the exit meeting held on April.27, P

'1990

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ATTACHMENT B DOCUMENTS REVIEWED I

EMERGENCY OPERATING PROCEDURES (EOPs)

E0P-1, "RPV Contro1~,'.' Revi sion 9

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E0P-1A,." Anticipated Transient Without Scram," Revision.9

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E0P-2, " Emergency Procedure-Primary Containment Control," Revision 7 -

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E0P-3, " Emergency Procedure-Secondary Containment and Radioactive Release Contro1~," Revision 7 CALCULATIONS o

t EAPPC*0002-NE,:" Primary System Data for Rev. 4 EPG App C, 9-10-87,", Revision 0

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EAPPC*0003-NE,." Mass of RPV & Recire Loops for Revision.4 EPG, Appendix C,.

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9-10-87," Revision 0

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'EAPPC*007-NE, "Flowrate for Revision 4, EPG Appendix C, 8-31-87," Revision 0

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FAPPC*0008-NE, "SRV Pressure for Rev.4 EPG Appendix C, 9-10-87," Revision 0 EAPPC*0010-NE,'" Pool Water Volume for Revision 4, Appendix C, 8-31-87,"

' Revision 0

EAPPC*0011-NE, " Containment Imput Data-for Rev.4 EPG Appendix C, 9-10-87,"

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Revision 0

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EAPPC*0013-NF, "RPV ESP Instrumentation for Appendix-C Calculation Input, Rev 4, EPG," Revision 0

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EAPPC*0015-Ne, " Water Level A & Top of Active Fuel-(TAF) per App C cales Rev.4, EPG," Revision 0

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'EAPPC*0017-NE,'"ECCS Data for Appendix C Calculation Rev.4," Revision 0 EAPPC*0019-NE, " Vortex Worksheet for Vortex Limit, Revision-4, EPG Appendix C,.

.9-9-87," Revision 0

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EAPPC*0024-BP,'" Heat Capacity Temperature Limit-2, 3-29-90," Revision 0 EAPPC*0030-NE, "RPU Variables, 3-25-90," Revision 0

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' CALCULATIONS (continued).

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WPPX*0031-BP, " Pressure-Ranges For Injection Into the RPV using.CRD,.FPW, SSW, Land NSW, 3-27-Also for CondensateLTransfer Pumpstand.ECCS

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Keepfi11 & SLC, 3-27-90," Revision 0

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LEAPPC*0032-NE,," Mass of Zircolloy In Reload Core 2," Revision 0

LE 10THER PROCEDURES;AND DOCUMENTS

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LA0P-0001,=" Reactor Scram,". Revision 5

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AOP-0002,;" Main: Turbine and Generator Trips," Revision 5'

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' ADM-0022, " Conduct of Operations," Revision 12B

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.A0P-0050, " Station Blackout," Revision 3

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l COP.-1022, " Post Accid Hydrogen Analysis of the Containment Atmosphere," Revision 1_

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Condition Report No, 190-0335, " Minimum Number of SRVs--Required for Emergency-

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-Depressurization," dated April 18, 1990.

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EDP-AAe19, " Preparation of-Engineering Input For Emergency and~ Abnormal Plant

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Operating: Procedures," Revision 3

EIP-2-001, " Classification'of. Emergencies,". Revision 5

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i ENG-3-006, " River Bend Station: Design and Modification-Request Control Plan,'!

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. Revision 4, j'

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ENG-3-020, " Control and Use-of Plant. Specific-Technical. Guidelines," Revision 3.

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EPSTG*0002-0," River Bend Station PSTG to E0P Flowcharts Comparison," Revision 0.

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EPSTG*0001-0'& Attachments I & II, " River Bend Station Plant. Specific-Technical j

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Guidelines'!,.. Revision 0 j

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GOP-002," Power Decrease / Plant Shutdown," Revision 5

GOP-003, " Scram Recovery," Revision 6 NED0-31331',' Class 1, "BWR Owner's Group-Emergency Procedures Guidelines,"

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Revision 4

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0SP-0009, " Author's Guide / Control and Use of Emergency- 0perating Procedures,"

Revision 4 c4 j

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3 CALCULATIONS-:(continued).

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, l.. hL i SOP-004, 'l Hydrogen Mixing,- Purge, Recombiners, and Igniters," Revision' 6.

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' SOP-0031, " Residual Heat Removal," Revision 6C-

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da SOP-104, " Floor and Equipment Drains System,"' Revision 8A'

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OSP-0008, " Verification and validation of Emergency Operating Procedures,"'

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Revision 2-

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S0P-0008,," Condensate, Storage,1 Makeup land Transfer,": Revision 6

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.. SOP-0031,:"High Pressure Core. Spray," Revision-6A i

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50P-0035, " React'or Core Isolation Cooling System," RevisionuSA1

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o Condition Report'No 90-0322. " Change in Technical Specification LHGR not-

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. Applied.to, Revise'Affected E0P Values," April 12, 1990 g

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LADM-0003;)' Development, Control and Use of Procedures," Revision 17A

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Condition Report.90-0361, " Discrepancy Between Calculated Values:For the Minimum

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CoreLFlooding Interval and E0P-0001,' Table'1," April 24, 1990 4

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10SP-0005, " Operations-Procedure Review and Revision," Revision 5B-

-0SP-007, "PreparationJof Operations Sections Procedures," Revision 6

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-ATTACHMENT C

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ABBREVIATIONS AND ACRONYMS

'ADM Administrative:

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A0Pa Abnormal Operating Procedure BWR-

-Boil.ing~ Water Reactor C0F

. Control Operations Foreman.

COP'

Chemistry Operating Procedure

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0/P pressure differential l

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DRMS Digital Radiation Monitoring System

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ECCS Emergency Core Cooling System-EIP Emergency Implementing Procedure

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ENG Engineering-E0P Emergency Operating Procedure

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ERIS Emergency Response Information. System GAL Gallons HPCSL High Pressure Core Spray LBS-Pounds

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'LPCI

' Low Pressure Coolant Injection

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-LPCS Low Pressure Core Spray i

MARFP Minimum Alternate RPV Flooding Pressure

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MCUTL-Maximum Core Uncovery Time

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MCFI Minimum Core Flooding Interval

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MCR Main Control Room

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MOV Motor.0perated Valve

MR.

Modification Request

'MSR-

' Moist'ure Separator Reheater l

'MWO-Maintenance Work Order i

NE0-Nuclear Equipment Operator

-0G Owners Group-

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0SP Operations l

PCT Peak Clad Temperature l

PMS-Plant Monitor System-1 PSTG Plant Specific Technical Guidelines-

'QA~

Quality Assurance-QAFR Quality Assurance Finding Report

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RBS

. River Bend Station-

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RHR-Residual Heat. Removal

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R0 Reactor Operator

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RPV Reactor Pressure. Vessel

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SLC Standby Liquid Control S0P System Operating Procedure o

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'SRV Safety Relief Valve L

SS Shift Supervisor q

.STA Shift Technical' Advisor i

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TCN Temporary Change Notice TS Technical' Specification LV & V Verification and Validation

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