ML20140J316

Insp Rept 50-382/97-10 on 970428-0502.Violations Noted.Major Areas Inspected:Licensee Operations,Maint & Engineering

ML20140J316
Person / Time
Site:	Waterford
Issue date:	06/16/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20140J301	List: ML20140J299 ML20140J304 ML20140J316
References
50-382-97-10, NUDOCS 9706190362
Download: ML20140J316 (31)

See also: IR 05000382/1997010

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ENCLOSURE 2

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l U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

l Docket No.: 50-382

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License No.: NPF-38

Report No.: 50-382/97-10

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Licensee: Entergy Operations, Inc.

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Facility: Waterford Steam Electric Station, Unit 3

Location: Hwy.18

Killona, Louisiana

Dates: April 28, through May 2,1997

Inspectors: Linda Joy Smith, Reactor inspector, Engineering Branch

James Melfi, Resident inspector, Projects Branch C

Approved By: Chris VanDenburgh, Chief, Engineering Branch

Division of Reactor Safety

ATTACHMENT: Supplemental information

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9706190362 970616

PDR ADOCK 05000382

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TABLE OF CONTENTS

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EXECUTIVE SUMMARY ....................................... .... iii

R e p o rt D e t a il s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

1. Operations .................................................... 1  ;

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08 Miscellaneous Operations issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 l

08.1 (Closed) Unresolved item 50-382/96202-04 ............... 1

11. M a i n t e n a n c e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 j

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M8 Miscellaneous Maintenance issues ........................... 1

M8.1 (Closed) Unresolved item 50-382/96202-09 ............... 1 J

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lit. Engineering ................................................... 3 j

E8 Miscellaneous Engineering Issues . . . . . . . . . . . . . .......,...... 3

E8.1 (Closed) Unresolved item 50-382/96202-05 ............... 3 s

E8.2 (Closed) Unresolved item 50-382/96202-06 ............... 9  !

E8.3 (Closed) Unresolved item 50-382/96202-07 .............. 10 )

E8.4 (Closed) Unresolved item 50-382/96202-11 .............. 12

E8.5 (Closed) Unresolved item 50-382/96202-12 .............. 15

E8.6 (Closed) Unresolved item 50-382/96202-13 .............. 18

E8.7 (Closed) Unresolved item 50-382/96202-17 .............. 20

V. Ma nageme nt Meeting s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2

X1 Exit Meeting Summary ..................................22

ATTACHMENT: Supplemental Information

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EXECUTIVE SUMMARY

Waterford Steam Electric Station, Unit 3

NRC Inspection Report 50-382/97-10 t

insoection Secoe

On April 28 through May 2,1997, two NRC inspectors conducted a followup inspection to

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determino ti:e appropriate enforcement for issues identified in NRC Inspection

Report 50-382/96-202.

Operations

• The licensee did not incorporate a vendor requirement to restrict operation of the

wet cooling tower during cold weather into the system r,perating procedure. This

was cited as one example of a violation of 10 CFR Part 50, Appendix B, Criterion 111,

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" Design Control," (Section 08.1).

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Maintenance

• The licensee did not incorporate requirements to install the reactor shield building

door ir. a seismically qualified configuration into the applicable maintenance i

procedure. This was cited as one example of a violation of 10 CFR Part 50,

Appendix B, Criterion lit, " Design Control," (Section M8.1).

Enaineerina

• Since original licensing, the licensee had not consistently maintained their

cornmitment to include a 10 percent fuel oil margin in the Technical Specification

limiting condition for operation for fuel oil. In addition, the licensee did not perform

an effective safety evaluation when they removed the 10 percent fuel oil margin

commitment from the Final Safety Analysis Report. In addition, they did not

recognize they were removing a commitment, which was repeeted in a pending

license amendment request. The f ailure to provide an adequate basis that the Final

Safety Analysis Report change did not involve an unreviewed safety question was l

cited as a violation of 10 CFR 50.59, " Changes, Tests, and Experiments,"

(Section E8.1). j

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• The NRC identified three examples wherein the revisions to the mechanical design j

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of the plant had not been adM;M,tely coordinated with the electrical design. The

failure to adequately control aesign interfaces between design organizations was  ;

cited as one example of a violation of 10 CFR Part 50, Appendix B, Criterion Ill, i

" Design Control," (Section E8.2).

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The 18-month diesel generator load shed surveillance tests were not detailed

(1) all nonsafety-related electrical breakers opened as

enough to confirm that:

required; or (2) electrical loads, which received process variable starts (i.e., high

reactor temperature or pressure), were load shed and resequenced with the

appropriate load block. This was cited as a violation of Technical Specification

Surveillance Requirement 4.8.1.1.2.e (Section E8.3).

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The licensee did not aggressively upgrade the emergency feedwater flow calculation

and associated license bases documents when they recognized that accident

conditions were not conservatively modeled (Section E8.4). Furthermore, the

inspector concluded that the licensee had not correctly evaluated emergency

feedwater requirements, in that Calculation EC-M96-004, " Design Basis

Reconstitution for EFW Flow Rate," Revision A, did not evaluate emergency

feedwater requirements when offsite power was available. This was cited as one

example of a violation of 10 CFR Part 50, Appendix B, Criterion lil, " Design

Control," (Section E8.5).

• The failure of the licensee to tale action when they recognized a potential

significant condition adverse to quality (i.e., the ultimate heat

10 CFR Part 50, Appendix B, Criterion XVI, " Corrective Action," (Section E8.6).

The licensee did not effectively correct the relief valve nozzle ring setting procedure

errors which were identified in NRC Information Notice 92-64, " Nozzle Ring Settings

on Low Pressure Water Relief Valves," which resulted in erroneous nozzle ring

settings for installed valves. This was cited as a second example of a violation of

10 CFR Part 50, Appendix B, Criterion XVI, " Corrective Action," (Section E8.7).

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Report Details

Summary of Plant Status

The Waterford 3 Steam Electric Station was in a refueling outage during this inspection.

1. Operations

08 Miscellaneous Operations issues

08.1 (Closed) Unresolved item 50-382/96202-04: Adequacy of cold weather operating

instruction for auxiliary component cooling water system.

Backaround -In NRC Inspection Report 50-382/96-202, the NRC team had

identified that the licensee had not incorporated a caution statement regarding cold

weather operation into System Operating Procedure OP-002-001, " Auxiliary

Component Cooling Water," Revision 10. The vendor technical manual stated the

wet cooling tower should not be started or operated without at least 30 percent

heat load when the inlet wet bulb temperature is below 35 F.

Inspector Followuo - The licensee wrote Condition Report 96-1506 to address this

concern, and informed the NRC team that the issue concerning cold weather

operation of the wet cooling tower would be resolved before the onset of winter.

The licensee subsequently determined that the wet cooling tower fans would not

automatically start on high temperature without at least a 30 percent heat load.

The licensee agreed that System Operating Procedure OP-002-001, " Auxiliary

Component Cooling Water System," Revision 10, did not include adequate

instructions to limit manual operation during cold weather. To correct this

deficiency the licensee installed caution cards on the system controls and revised

Procedure OP-002-001 to include appropriate cold weather controls.

10 CFR Part 50, Appendix B, Criterion Ill, " Design Control," requires that applicable

design basis requirements are correctly translated into procedures. The failure to

translate the design requirements for cold weather operation of the wet cooling

tower into System Operating Procedure OP-002-001, " *yxiliary Component Cooling

Water," Revision 10, is one example of a violation of 1-

"R Part 50, Appendix B,

Criterion ill (50-382/9710-01).

II. _ Maintenance

M8 Miscellaneous Maintenance issues

M8.1 LClosed) Unresolved item 50-382/96202-09: Adequacy of maintenance hatch

shield door installation instructions.

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Backaround - In NRC Inspection Report 50-382/96-202, the NRC team questioned ,

the reactor shield building door installation. They had discovered the shield building {

door secured by rails, an inflatable seal and a chain fall, i

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Insoector Followuo - The licensee subsequently determined that the reactor l

building shield door was not installed according to'the design drawings. Design  ;

Drawings 5817.7129 and 5817.7130 required installation of four 1-1/4" bolts to ,

hold the door in place. These bolts were credited as seismic restraints in i

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Calculation EC-C-90-038, " Design Calculation for Door Maintenance Hatch,"

Revision O. The licensee took immediate action to install the missing bolts so

that the field configuration matched the design drawings.

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The team noted that the door would be required to be operable for Technical

Specification Action Statements 3.6.6.1, 3.6.6.2, and 3.6.6.3. The licensee

performed an evaluation and determined that if the inflatable seal pressure

was greater than 27 psig, the door would be prevented from sliding along the

rails due to frictional resistance. The licensee promptly initiated measures .

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to maintain the seal pressures greater than 27 psig.

Subsequent to the team inspection, the licensee completed

Calculation EC-C90-038, " Design Calculation for Door Maintenance i

Hatch," Revision 1, which showed that in the as-found configuration, ,

the reactor building shield door would have met the intended seismic l

requirements. The licensee concluded there were no past operability concerns. I

Nevertheless, the licensee determined that Maintenance ProceFure MM-008 01, _

"Unside) Maintenance Access Hatch and (Outside) Maintenance Access Hatch i

Shield Door Opening, inspection and Closing," Revision 5, did not include. {

instructions whicu would assure that the reactor shield building door would l

be installed in seismically qualified configuration. The licensee revised  !

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Section 8.5 of Maintenance Procedure MM-OO8-01 to include installatica

instructions which would assure that the door is installed in accordance with design  :

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drawing requirements. (

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10 CFR Part 50, Appendix B, Criterion ill, requires that applicable design '

basis requirements are correctly translated into procedures. The failure to '

translate the design requirements for seismic qualification into Maintenance

Procedure MM-OO8-01, "(Inside) Maintenance Access Hatch and (Outside)

Maintenance Access Hatch Shield Door Opening, inspection and Closing," l

Revision 5, the installation instruction for the reactor shield building door,  ;

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is a second example of a violation of 10 CFR Part 50, Appendix B, Criterion 111

(50-382/9710-01).  ;

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E8 Miscellaneous Engineering issues i

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E8.1 (Closed) Unresolved item 50-382/96202-05: The adequacy of emergency diesel  ;

l generator fuel oil volume including 10 CFR 50.59, " Changes, Tests, and  ;

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Backaround - Two issues had been identified: (1) the licensee did not include

the 10 percent minimum margin in their time-dependent calculation for arriving at

the Technical Specification value of 38,760 gallons; and (2) the adequacy of the

10 CFR 50.59 evaluation for Licensing Document Change Request 93-0091.  !

Insoector Followuo - The previous NRC inspection had determined that the licensee

l was not currently in compliance with their commitment to include the 10 percent

l minimum margin in their time-dependent calculation for arriving at the Technical j

Specification value of 38,760 gallons. The licensee agreed with this observation, ,

but believed that the 10 percent margin was an initial sizing reouirement, which l

l was not intended to be maintained throughout the life the facility. The licensee

argued that the margin was included in the initial sizing to provide for minor load l

growth.-

The inspector reviewed documentation provided by the licensee related to their

interpretation of the applicability of the margin and to determine the original license

basis. The inspector also reviewed the licensing documentation and NRC inspection

reports to determine if there was a record that the NRC had previously accepted the

licensee's position that they were not currently required to maintain the 10 percent I

margin.

l The inspector also evaluated the adequacy of the 10 CFR 50.59 evaluation

associated with Licensing Document Change Request 93-0091 in the light of this i

additional information. The safety evaluation was approved during Plant Operations ,

Review Committee Meeting Number 93-038 as item lli-A. l

Purpose of ANS-59.51/ ANSI N195-1976, " Fuel Oil Systems for Emergency Diesel I

Generators"

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The inspector found that the 10 percent margin requirement was discussed in the

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system performance requirements section of ANS-59.51/ ANSI N1951976, " Fuel

Oil Systems for Emergency Diesel Generators." This section provides system

performance requirements for the amount of oil stored onsite. The standard

requires sufficient onsite stored oil for either 7 days or the time required to

replenish the oil, whichever is less. In the system performance section, the  ;

standard provides two calculation methods for determining how much oilis needed

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method (the time-dependent load method), then the standard requires that a

minimum margin of 10 percent be added to the calculated storage requirements.

The inspector concluded that the 10 percent margin was intended to ensure the

system performed acceptably in service.

On November 12,1993, Commonwealth Edison requested a formalinterpretation of

ANS 59.51/ ANSI N195-1976 from the American Nuclear Society Standards

committee. The committee determined that the 10 percent margin was specified to

account for uncertainties that might be present at the time of initial system design,

such as the potential for small future changes in loadings, specific design features

of the fuel oil storage and delivery systems, changes in engine performance, fuel oil

quality variations, vortexing margins, and instrument error. The committee stated

that it was recognized that final design features and loadings could effectively alter

the value of this design margin, but it was not intended that the storage capacity be

reduced below the limits allowed by the plant Technical Specifications. They

further stated that licensees must evaluate changes and if significant loads are

added or a load duration is changed, it might be necessary to increase the minimum ,

volume in the Technical Specifications.

While this clarification partially supported the licensee's position, the inspector

noted that: (1) the actual standard had not been revised, and (2) the NRC had not

endorsed the clarification. The inspector also noted that even if you accepted the ,

standards committee's informal clarification, more than tank sizing was involved. In

addition to load growth, the standards committee said the margin was also intended

to address other uncertainties associated with doing the less conservativo

time-dependent load calculation (i.e., potential changes in engine performance, fuel

oit quality variations, vortexing margins and instrument error).

The inspector concluded the 10 percent margin was intended, in part, to address

uncertainties assnciated with the less conservative time-dependent load calculation.

ucensee's Commitment to ANSI N195-1976, " Fuel Oil System for Standby Diesel

Generators"

The inspector found that at the time of initial licensing, in Final Safety

Analysis Report Question 40-20, the NRC staff had asked the licensee to either

confirm compliance with ANSI N195-1976, " Fuel Oil System for Standby Diesel

Generators," or provide justification for non-compliance. The licensee's

response referred to revised Final Safety Analysis Report Subsection 9.5.4.1.

In Amendment 2 (dated March 1979), the licensee revised Subsection 9.5.4.1,

" Design Bases," to state that all safety-related portions of the diesel engine fuel oil

storage and transfer system are designed to ANSI Standard N195-1976," Fuel Oil

System for Standby Diesel Generators."

As stated above, when a time-dependent analysis is performed to estimate the

amount of fuel required, then ANSI Standard N1951976 requires that an additional

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10 percent margin be included in the fuel oil capacity calculation. The calculation

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for fuel oil storage capacity at the time of the issuance of Amendment 2 of the Final

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Safety Analysis Report, Calculation MN(Q)-9-11, " Fuel Oil Storage for Diesel

Generators," Revision 0, dated March 20,1979, was a time-dependent analysis.

Further, after including a 10 percent margin, the predicted fuel oil consumption

requirement was stillless than 40,000 gallons, the value in the draft Technical

Specifications at the time.

The inspectcr noted that in NRC inspection Report 50-382/90-23, the team

reviewed a preliminary diesel generator fuel oil consumption calculation and

fuel oil controls at the time. The team identified weaknesses with the

calculations related to including enough fuel for testing, but stated that updates

to address the weaknesses should not affect the results of the calculation Th

inspection report states that the licensee's preliminary calculation concluded

that fuel oil consumption, including a 10 percent margin, was less than the

Technical Specification value. However, the licensee stated that during NRC

Inspection 50-382/90-23, Waterford 3 personnel indicated that the 10 percent

design margin was not available. The NRC documented in NRC Inspection

Report 50-328/93-01, that the licensee stated the 10 percent margin was available

if the licensee restricted the emergency diesel generator iotding during an accident

to only equipment needed for safe shutdown. In either case, the licensee

committed to clarify the basis for the fuel oil stnrage requirement in the Technical

Specification with respect to the 7-day requirement.

The licensee submitted and discussed several Technical Specification amendment

requests related to this issue. On March 16,1994, the NRC staff ultimately

approved Amendment 92, which included the licensee's request to update the

bases section of the Technical Specification to indicate use of the time-dependent

method in conformance with Regulatory Guide 1.137 (October 1979). Regulatory

Guide 1.137 endorses the use of ANSI N195-1976, as an acceptable method for

complying with General Design Criteria 17, " Electric Power Systems." The

inspector also found that at the time of the licensee's request for Amendment 92,

the licensee had not taken exception to the requirements of ANSI N195-1976 in the

Updated Final Safety Analysis Report.

The inspector reviewed the NRC's safety evaluation report for Amendment 92 to

the Technical Specifications, which modified the fuel storage requirements to allow

a lower fuel level for a short time to support additional testing. The inspector

found that the staff's conclusion to approve the final version of the amendment was

based on the retention of the minimum 7-day requirement in accordance with

Regulatory Guide 1.137 (October 1979), " Fuel-Oil Systems For Standby Diesel l

Generator." J

The inspector determined that the licensee was consistently committed to

ANSI N195-1976 from the time of initial licensing. Further, the NRC staff had not

approved deviation from this commitment.

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Determination of Emergency Diesel Generator Fuel Oil Margin at initial Licensing

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In Calculation MN(O)-0-42, " Emergency Diesel Generator Oil Consumption,"

Revision 1,~ dated August 18,1983, Ebasco Services, Incorporated, determined  ;

that 38,758 gallons were needed for the time-dependent load analysis, without a

10 percent margin. On September 14,1983, Ebasco advised the licensee that the

Waterford fuel storage capacity no longer met the ANSI N195-1976-recommended

design margin of 10 percent. They also stated that they believed the 7-day criteria

was conservative and that no additional margin was riecessary and that it was not

necessary to include the 10 percent design margin as a Technical Specification

minimum safety requirement. Ebasco stated that they would amend Final Safety

Analysis Report, Subsection 9.5.4, to reflect the slight reduction in tank sizing

margin from 10 percent to 6.2 percent. Although the licensee stated this 1

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amendment was never processed, the licensee implemented the recommendation

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and changed the Technical Specification-required value to the current number, 1

38,760 gallons.

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in Final Safety Analysis Report Question 40-98, the NRC staff specifically asked

the licensee to clarify an apparent discrepancy between the draft Technical i

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Specification 3.8.1.2 and Final Safety Analysis Report, Subsection 9.4.5.2, to  !

state precisely how much fuel is necessary for a 7-day sgply. The licensee

responded to Question 40-98 by stating,

" Calculations were performed using the time-dependent load

method . . . These calculations show that 38,760 gallons are

required as the minimum onsite inventory of fuel oil for each

redundant diesel generator. This requirement is reflected in the

plant Technical Specifications." t

The am noted that this response did not clearly indicate the change in

cor atment with respect to inclusion of the 10 percent margin required by

ANbi N195-1976, if the total fuel storage is calculated based on the less *

- conservative time-dependent load method.

The inspector determined that the original Technical Specifications were not

prepared in accordance with the commitments the licensee made regarding full

compliance to ANSI N195-1976 in response to Final Safety Analysis Report,

Question 40-20, which was subsequently included in the text of Final Safety

Analysis Report. Subsection 9.5.4.1.

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Emergency Diesel Generator Fuel Oil Margin at the time of NRC Inspection

Report 50-382/96 202

Technical Specification 3.8.1.1 requires, except for testing, the licensee maintain a I

minimum of 38,760 gallons of fuel oil in each storage tank. At the time of NRC 1

Inspection 50-382/96 202, the licensee could not demonstrate that they had a

10 percent margin between the results of their time-dependent load analysis and the

Technical Specification value of 38,760 gallons and they did not apply for a license

amendment.

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10 CFR 50.59 Safety Evaluation for Licensing Document Change Request 93-0091 1

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In NRC Inspection Report 50-382/96-202, the team noted Licensing Document '

! Change Request 93-0091 included a Final Safety Analysis Report change which ,

deleted commitment to ANSI N195-1976, with respect to sizing of the fuel oil

storage tank.

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in NRC Inspection Report 50 382/96-202, the NRC noted the licensee's

10 CFR 50.59 safety evaluation stated that the change did not affect the

design as specified in Technical Specification bases. However, the bases

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Section stated that the minimum required volume of emergency diesel generator l

fuel was based on conformance with Regulatory Guide 1.137, October 1979, that l

endorsed the ANSI standard. Implementation of Licensing Document Change

Request 93-0091 resulted in a 10 percent reduction in the required fuel oil storage

capacity as specified in Technical Specification bases. The team considered this

deletion to be a 10 percent reduction in the required fuel storage capacity.

Therefore, the licensee's 10 CFR 50.59 evaluation was inadequate to demonstrate

that an unreviewed safety question did not exist.

During this inspection, the inspector noted that Licensing Document Change

Request 93-0091 was approved on December 10,1993. On January 28,1993,

the licensee had submitted the request for Technical Specification, Amendment 92,

which clarified that the limiting condition for operation for diesel generator

storage requirements is based on load dependent calculations in conformance

with Regulatory Guide 1.137, October 1979. However, Technical Specification,

Amendment 92, was not issued until March 16,1994, after the safety evaluation

was approved. The inspector concluded that a violation of 10 CFR 50.59 did

not occur for the reasons which were presented in NRC Inspection

Report 50-382/96-202. However, a violation of 10 CFR 50.59 did occur for the

reasons described below.

10 CFR 50.59(a)(2) states that a proposed change shall be deemed to involve an

unreviewed safety question if the probability of a malfunction of equipment

important to safety previously evaluated in the safety analysis report may be

i increased.10 CFR 50.59(b)(1) requires that the licensee maintain records of

! changes in the facility and of changes in procedures made pursuant to this section,

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to the extent that these changes constitute changes in the facility as described in l

the safety analysis report or to the extent that they constitute changes in the ,

L procedures as described in the safety analysis report. Further, these records must

j- include a written safety evaluation which'provides the bases fof the determination l

that the change does not involve an unreviewed safety question.

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In the safety evaluation for Licensing Document Change Request 93-0091, the ,

i licensee determined that the proposed changes did not increase the probability of .

l occurrence of a malfunction of equipment important to safety previously evaluated l

e in the safety analysis report. The inspector reviewed the basis for the licensee's  !

determination and noted that the licensee did not address the reduction in required {

fuel oil storage margin and the associated increase in probability that the emergency '!

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diesel generator would run'out of fuel before 7-days because of uncertainties '

i associated with the time-dependent load calculation. The inspector determined the

licensee's 10 CFR 50.59 written safety evaluation did not provide an adequate -

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basis that the change does not involve an unreviewed safety question. The failure  ;

to provide this basis is a violation of 10 CFR 50.59 (50-382/9710-02). ,

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. Conclusion - The inspector did not agree with the licensee's interpretation of

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ANSI N195-1976, in that the 10 percent margin related to the time-dependent load

calculation was only intended for initial tank sizing. While the informal i'

clarification developed by the standards committee partially supported the

licensee's position, the inspector noted that: -(1) the actual standard had not- ,

been revised; (2) the NRC had not endorsed the clarification; and (3) in addition

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to load growth, the standards committee stated that the margin was also intended i

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to address other uncertainties associated with doing the less conservative

time-dependent load calculation, which the licensee had not addressed (i.e.,  :

potential changes in engine performance, fuel oil quality variations, I

vortexing margins and instrument error).

The inspector concluded that the original Technical Specifications were not prepared )'

in accordance with the ccmmitments the licensee made regarding full compliance to

ANSI N195-1976, in that the limiting condition for operation for fuel oil storage

- system was not based on a time-dependent calculation, which included the required

10 percent margin, j

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The inspector determined that the licensee was consistently committed to )

ANSI N195-1976 from the time of initial licensing. Further, the NRC staff had i

not approved deviation from this commitment.

Further, at the time of NRC Inspection Report 50-382/96-202, the licensee was not

in compliance with their commitment to include the 10 percent minimum margin in

their time-dependent calculation for arriving at the Technical Specification value of

38,760 gallons.

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~ The inspector determined the licensee's 10 CFR 50.59 written safety evaluation for j

the change to the Final Safety Analysis Report, which removed the commitment to '

maintain the 10 percent margin, was inadequate. The safety evaluation did not

address the reduction in required fuel oil storage margin and the associated increase i'

t in probability that the emergency diesel generator would run out of fuel before

l 7-days because of uncertainties associated with the time-dependent load

calculation. As a result, the safety evaluation did not provide an adequate basis

that the change did not involve an unreviewed safety question. The failure to  !

provide this basis is a violation of 10 CFR 50.59.

E8.2 (Closed) Unresolved item 50-382/96202-06: The adequacy of administrative i

controls to ensure the emergency diesel generator loading and fuel oil calculations j

and the related Final Safety Analysis Report tables are updated when the

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mechanical design is changed.

Backaround -In NRC Inspection Report 50-382/96-202, the NRC identified three

examples of the revisions to the mechanical design of the plant not being ,

adequately coordinated with the electrical design. In each case, the licensee did not

update Calculation EC-E90-OO6, " Emergency Diesel Generator Loading and Fuel Oil i

Consumption," or Final Safety Analysis Report Table 8.3-1, " Emergency Diesel

Generator A and B Loading Sequence," when mechanical aspects of the design

were changed.

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The first example involved Licensing Document Change Request 96-0161, which l

added the manual start of a fuel pool cooling pump 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after a loss-of-offsite <

power with a safety injection actuation signal. The team identified that the Final j

Safety Analysis Report and the emergency diesel generator load calculation should  :

also have been updated to add this load for a loss-of-offsite power without a safety

injection actuation signal.  !

The second example involved Mechanical Calculation MN(Q)-9-9, " Wet Cooling i

Tower During a LOCA," Revision 3, Change 1, paragraph 5.2, which showed that ,

'

half of the fans operated for 5 days and the other half operated for 27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br />. The

Final Safety Analysis Report and the diesel generator load calculation still showed all

of the wet cooling tower fans operating only 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />.

The third example involved the 10 CI150.59 evaluation, Plant Operations <

Review Committee Meeting 96-037, item lil-D (for: License Document Change i

Request 96-0161) changed Final Safety Analysis Report, page 9.2-16, to indicate >

that the auxiliary component cooling water system was not required after 5 days

(rather than the previous 7-day requirement) following a large break loss-of-coolant

accident. The Final Safety Analysis Report and the diesel generator load calculation  ;

still showed auxiliary component cooling water pumps operating 7 days. j

inspector Followu_n - The inspector reviewed NRC Inspection Report 50-382/962-02

and Licensee Condition Report 96-1586 with respect to the control of emergency

dierel generator loading and fuel oil calculations. The inspector also interviewed ,

licensee personnel.

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The licensee noted that the electrical section had been informally notified about

these design changes but no formal mechanism had been employed to ensure the

design of the support systems was adequately evaluated when the mechanical

design assumptions were changed.

l

10 CFR Part 50, Appendix B, Criterion Ill, requires that measures be established

for the identification and control of design interfaces and for coordination among

participating design organizations. The failure to adequately control design

,

interfaces between the mechanical design organization and the electrical design ]'

organization is a third example of violation of 10 CFR Part 50, Appendix B,

Criterion lli (50-382/9710-01).

I

In NRC Inspection Rcport 50-382/96-202, the NRC team determined emergency

dies : generator loading changes for the calculated equipment operating times

were not properly considered in the 10 CFR 50.59 evaluations for the above

changes. Therefore, the team concluded that these 10 CFR 50.59 reviews were

incomplete and inadequate to demonstrate that an unreviewed safety question did

not exist. While the failure to provide an adequate basis that the change does

not involve an unreviewed safety question is also a violation of NRC requirements,

the inspector concluded the primary error was a failure to coordinate between

design organizations. Therefore, this issue is most appropriately characterized as

a 10 CFR Part 50, Appendix B, Criterion ill, violation, as described above.

Conclusions - The inspector concluded that the licensee did not adequately

( control design interfaces between the design organizations which was a violation

of 10 CFR Part 50, Appendix B, Criterion Ill.

l

E8.3 (Closed) Unreselved item 50-382/96202-07: The item involved the adequacy of

18-month Technical Specification integrated tests to verify deenergization of the

emergency buses and load shedding from the emergency buses, and the adequacy

of dieselloading and fuel oil consumption calculations to ensure that untested loads

were properly considered in the calculations.

Backaround - In NRC Inspection Report 50-382/96-202, the NRC team had

identified that the licensee was not verifying all nonemergency 480 Vac loads on

safety-related buses shed and sequenced as required. The licensee asserted that

i they were only required to verify that devices shed by the loss-of-offsite power or

! loss-of-offsite power / safety injection actuation signal sequencer relays, were in f act

deenergized. The NRC team did not agree with this position.

l Inspector Followup - The inspector reviewed NRC Inspection Report 50-382/96-202

and Licensee Condition Report 96-1594 with respect to the adequacy of the

18-month Technical Specification integrated tests to verify deenergization of the

! emergency buses and load shedding from the emergency buses. The inspector

interviewed licensee personnel and watched the licensee demonstrate the

functioning of the type of electrical contactors, which were installed at Waterford 3.

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Technical Specification 4.8.1.1.2.e.3 requires that each diesel generator shall be

demonstrated operable at least once per 18 months during shutdown by simulating

a loss-of-offsite power by itself, and (1) verifying deenergization of the emergency

busses and load shedding from the emergency busses, and (2) verifying the diesel  !

energizes the auto-connected shutdown loads through the load sequencer.

I

Technical Specification 4.8.1.1.2.e.5 requires that each diesel generator shall be

demonstrated operable at least once per 18 months during shutdown by simulating

a loss-of-offsite power in conjunction with a safety injection actuation signal test

signal, and (1) verifying deenergization of the emergency busses and load shedding

from the emergency busses, and (2) verifying the diesel energizes the auto-

connected shutdown loads through the load sequencer.

During this inspection, the inspector determined that the loads in question fellinto

three general categories: (1) loads shed through the use of electrical contactors,

(2) loads shed by opening a nonsafety-related circuit breaker, and (3) loads which

received process variable signals (i.e., high reactor temperature or low pressure) to

start and stop.

The inspector was not able to resolve whether or not loads shed through the use

of electrical contactors need to be independently verified. The licensee

stated that the design of electrical contactors was to open upon loss of

holding voltage and, hence, required no verification past confirmation that

the switchgear had deenergized. The inspector observed a demonstration of

contactor operation and confirmed the contactors were designed to open upon

loss of holding voltage. Whether or not the licensee can assume contactors

open upon loss of holding voltage and still meet the verification requirements of

Technical Specifications 4.8.1.1.2.e will be followed as an unresolved item pending

further evaluation by the Office of Nuclear Reactor Regulation to determine

the intent of the license requirements (50-382/9710-03).

In NRC Inspection Report 50-382/96-202, the NRC team had observed that

Procedure OP-903-028, " Pressurizer Heater Emergency Power Supply Functional

Test," Revision 3, deenergized Switchboard 3A32 and stated that all loads would

be deenergized, but did not require observation or verification that the circuit

breakers for the pressurizer heater groups opened and deenergize the loads. The

licensee stated that this procedure was credited to ensure that the nonsafety

loads shed from the emergency powered busses during a loss-of-offsite power or a

loss-of-offsite power with a safety injection actuation signal.

During this inspection, the licensee stated that they planned to revise their

18-month tests to require verification that all nonsafety-related circuit breakers shed ,

and sequence properly prior to restart from the current outage. The inspector

c

concluded, that at the time of NRC Inspection 50-382/96-202, the licensee had not

l verified that the pressurizer heater group circuit breakers opened as required and

l load shed during the simulated loss-of-offsite power test or the loss-of-offsite power

in conjunction with the safety injection actuation signal test. This is one example of

! a violation of Technical Specification 4.8.1.1.2.e (50-382/9710-04).

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In NRC Inspection Report 50-382/96-202, the NRC team had observed that

Shutdown Heat Exchanger A and B room coolers and the Component Cooling Water

Heat Exchanger A and B room coolers were not being periodically functionally

tested to verify that they correctly started on a high temperature on their

corresponding load sequencer block. Control Room Heater EHC-34 and Switchgear

Room Heater EHC-36, which were shed and reenergized by a loss-of-offsite power

or loss-of-offsite power / safety injection actuation signal, were also not part of the

integrated testing.

During this inspection, the licensee stated that they plan to revise their 18-month

tests to ensure allloads, controlled by a process variables, shed and sequence

properly, prior to restart from the current outage. The inspector concluded that at

the time of NRC Inspection 50-382/96-202, the licensee had not verified that the

following loads correctly shed and started on their corresponding load sequencer

block: Shutdown Heat Exchanger A and B room coolers, the Component Cooling

Water Heat Exchanger A and B room cooters, and Control Room Heater EHC-34 and

Switchgear Room Heater EHC-36. This is the second example of a violation of

Technical Specification 4.8.1.1.2.e (50-382/9710-04).

In NRC Inspection Report 50-382/96-202, the NRC team informed the licensee that

any loads that were not verified as being shed during a loss-of-offsite power or loss-

of-offsite power / safety injection actuation signal test should be considered as part

of the diesel loading. In addition, applicable calculations (including the diesel fuel oil

calculation) needed to be updated because the existing calculations assumed that

these loads were shed.

As stated above, the licensee plans to update their testing practices to ensure

proper load shedding for: (1) loads powered through nonsafety-related circuit

breakers and (2)Ioads controlled by a process variable. The licensee stated that

their testing practices will be consistent with the analysis assumptions for loads of

these +ypes. The licensee did not plan to update the analysis to assume allloads

supplied by contactors remain connected. As stated above, this issue remains

enresolved.

Conclusions - The inspector concluded that the licensee had not properly

established surveillance tests to meet the requirements of Technical Specifications 4.8.1.1.2.e for loads which were shed by opening nonsafety-related c,ircuit

breakers, and loads which received process variable signals (such as high

temperature or low pressure) to start and stop. This is a violation of the Technical

Specifications.

E8.4 (Closed) Unresolved item 50-382/96202-11: The adequacy of the licenseo's

corrective actions for an error in the bases section of the Technical Specification.

Backaround - The inspector had identified two issues: (1) the licensee did not

perform an operability confirmation review as required by licensee's procedure, and

(2) the licensee's resolution of the nonconforming condition was not timely.

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Insoector F_p!!owuc - The inspector reviewed Condition Report 95-0656 which l

- identified that the emergency feedwater pumps could not perform as specified in

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l' the Technical Specification bases section. The inspector also reviewed the '

following documents and interviewed NRC and licensee personnel. J

l * - Administrative Procedure UNT 006-011, " Condition Report," Revision 3

Site Directive W4.101, " Operability / Qualification Confirmation Process,"

Revision 1

,

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Final Safety Analysis Report, Section 10.4.9, " Emergency Feedwater

l

- System," Revision 8

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• Final Safety Analysis Report, Table 10.4.9A-1, ' Evaluation of the Waterford

SES [ Steam Electric Station] Unit 3 Emergency Feedwater System Versus the

Requirements of Standard Review Plan (SRP) 10.4.9 and Branch Technical

Position (BTP) ASB 10-1," Amendment 33

• NRC Inspection Reports 50-382/96-03 and 50-382/96-05

I . . .

On July 17-21 and July 31-August 4,1995, the licensee performed a safety

system functional inspection of the emergency feedwater system. The licensee

identified that the emergency feedwater pumps were not capable'of meeting i

the flow / pressure requirements provided in the Technical Specification

Bases, Section 3/4.7.1.2, which stated that each electric-driven emergency

feedwater pump was capable of delivering a total feedwater flow of 350 gpm at  !

a pressure of 1163 psig to the entrance of the steam generators. The licensee

had determined that the basis section should have read that each electric-driven

emergency feedwater pump was capable of delivering a total flow of 350 gpm at a

pressure of 1163 psig at the discharge of the pump.

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The NRC reviewed the technical aspects of this issue in NRC Inspection

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Reports 50-382/96-03 and 50-382/96-05. The inspectors verified that, considering

this corrected Technical Specification Basis, Calculation EC-S89-3, "EFW

Requirements," a vendor-generated mass and energy balance calculation, still  !

! confirmed that the emergency feedwater pump capability was sufficient to meet the

Final Safety Analysis Report requirement to provide 450 gpm emergency feedwater  !

flow to the steam generator at the lowest safety relief valve setpoint,1070 psig,

plus 3 percent margin for the setpoint tolerance to prevent dryout of the steam j

generators.  !

h The licensee determined that this error.was caused by unclear calculations,

l

which did not completely model expected conditions. The licensee developed

[ Calculation EC-M96-004, " Design Basis Reconstitution for EFW Flow Rate,"

j Revision A, to address these calculation weaknesses. The licensee planned to use

l the results of this improved analysis to correct inconsistencies in the Technical

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Specificatic is, the Final Safety Analysis Report, design basis documents, technical

manuals, design specifications, and procedures, if necessary. During this inspecta

the licensee indicated that they planned to docket their plan and schedule for

updating the appropriate license basis documents.

in NRC Inspection Report 50-382/96-202, the NRC noted that the new calculation

determined that when reasonable decay heat uncertainties and steam generator

back-pressure were considered, while maintaining a reasonable steam generator

inventory margin, the original calculation was nonconservative. The new calculation

determined that minimum required emergency feedwater flow rate was 575 gpm.

This flow rate was needed to provide a 4980 lbm steam generator margin while

using a 10 percent decay heat uncertainty value.

In NRC Inspection Report 50-382/96-202, the NRC team documented concerns that

the licensee did not perform an operability confirmation review as described in

Procedure W4.101, " Operability / Qualification Confirmation Process," when they

discovered that the emergency feedwater pumps were not capable of meeting the

flow requirements at the pressure specified in the Technical Specification bases

section.

During this inspection, the NRC reviewed in detail the requirements of

Procedure W4.101 and determined that the licensee was not procedurally

required to perform an operability confirmation review. The licensee had used

an existing approved calculation to determine that the discrepancy in the

Technical Specification basis section was not an operability concern.

However, the inspector noted that the weaknesses in the calculation which existed

at the time were not identified as a separate adverse condition requiring an

operability review. The licensee stated that the updated calculation still resulted in

flow requirements which were within the capability of the pumps. They did not

view the calculation weaknesses as a separate adverse condition. The general

concern of incompleteness of information contained in design documentation was

identified in a separate condition report.

The licensee's corrective action to upgrade the calculation was approved on

September 22,1995. The initial corrective action due date for the new calculation

was December 29,1995. However, the engineering service request to procure the

calculation from Asea Brown Boveri-Combustion Engineering Nuclear Power was not

approved until February 1,1996. The new Calculation EC-M96-004, " Design Basis

Reconstitution for EFW Flow Rate," Revision A, was approved by Waterford

engineering on June 28,1996, approximately 10 months after initial discovery of

the nonconforming condition.

l Regardless of the calculation outcome, the inspector determined that the calculation

and license bases document update was not timely considering the licensee  !

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recognized that the new essumptions more accurately modeled expected conditions

and would likely result in requirements, which were nonconservative with respect to

the original analysis. ,

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Conclusions - The licensee did not aggressively upgrade the emergency feedwater

flow calculation and associated license bases documents when they recognized that

accident conditions were not conservatively modeled. The licensee has completed

the calculation update and concluded the required flows are still within the

capability of the pumps. During this inspection, the licensee indicated their intent to

docket their plan and schedule for submitting the necessary revisions to license

basis documents.

E8.5 (Closed) Unresolved item 50-382/96202-12: The heat removal capability of the

emergency feedwater system.

Backaround -In NRC Inspection Report 50-382/96-202, the NRC team identified

that Calculation EC-M96-004, " Design Basis Reconstitution for EFW Flow Rate,"

Revision A, had not been analyzed for a feedwater break accident where a loss-of-

offsite power did not occur as required by 10 CFR Part 50, Appendix A, General

Design Criteria 34, " Residual Heat Removal."

Inspector Followu.n - The inspector interviewed licensee personnel and reviewed the

following:

• Standard Review Plan 15.2.8, "Feedwater System Pipe Breaks inside and

Outside Containment (PWR)," Revision 1

• Root Cause Andysis Report for Condition Report 96-1593, "EFW Flow

Calculation Did Not Ir dude RCP Heat Load," dated April 30,1997

• ABB-CE (Gardner) to Entergy (Holman), "Results of Feedwater Line Break

initial Analyses for EFW Requirements with Reactor Coolant Pump Heat,"

dated February 10,1997

• ABB-CE (Gardner) to Entergy (Holman), " Evaluation of the Acceptability of

575 gpm EFW Flow for the Feedwater Line Break and Loss of Condenser

Vacuum Events with inclusion of RCP Heat," dated April 30,1997 i

• A sensitivity study related to primary and secondary safety valve tolerances ]

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The licensee stated that they did not agree that a violation occurred because they I

were only required to evaluate bounding cases. They stated that with respect to l

reactor pressure, a feedwater line break coincident with a loss-of-offsite power was {

the bounding case. During a loss-of-offsite power, the reactor coolant pumps trip i

,

and are not a source of heat. They concluded they were not required to include j

reactor coolant pump heat in Calculation EC-M96-OO4.  !

The licensee stated that a sensitivity study performed in 1994 related to primary

and secondary safety valve tolerances supported the conclusion that the loss-of-

offsite power case was the most limiting. The inspector reviewed the study and

determined that it did not support the licensee's position.

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Additional inspection is planned to determine how the licensee used the results of

this study. Additional inspection will also be performed to determine whether the

Updated Final Safety Analysis Report contains a description of the bounding

analysis for a feedwater line break. This inspection will be tracked as inspection

followup item (50-382/9710-05).

The inspector noted that reactor pressure was only one of the variables reviewed

during initial licensing. The standard review plan also indicated that pressurizer

level is an important accident variable. As a result of the NRC team's concerns

documented in NRC Inspection Report 50-382/96-202, the licensee contracted with

ABB-CE to perform a second analysis to determine the required emergency

feedwater flow rate considering reactor coolant pump heat. ABB-CE ran several

preliminary cases using Final Safety Analysis Report assumptions and including

reactor coolant pump heat. They noted that for a feedwater line break inside

containment, letdown isolates and all three charging pumps start. ABB-CE

concluded that if three charging pumps continued to operate, the charging pumps

would fill the pressurizer solid regardless of the emergency feedwater flow rate

assumptions. To address the issue of pressurizer over-fill, ABB-CE assumed that

operators would take action to secure the charging flow either 7 minutes after

reactor trip or 2 minutes after a high pressurizer level alarm.

The inspector noted that the standard review plan, which was used to evaluate

license application related to feedwater line breaks, required the NRC staff to

evaluate operator actions which were necessary to secure and maintain the reactor

in a safe shutdown condition. The standard review plan further stated, "During the

initial 10 minutes of the transient, should credit for operator action be required

(i.e., reactor coolant pump trip), an assessment for the limiting consequences must

be performed in order to account for operator delay and/or error." The analysis

described in Updated Final Safety Analysis Report, Section 15.2.3.1, indicated that

no operator actions were assumed for the first 30 minutes. The inspector did not

review any documentation which indicated that the NRC staff had licensed the

facility assuming that the operators would secure charging flow either 7 minutes

after a reactor trip or 2 minutes after a high pressurizer level alarm.

Subsequent to the inspection, the licensee stated that their preliminary analysis now

shows that operator action is not required until approximately 15 minutes. As

stated above, additional inspection will also be performed to determine whether the

Updated Final Safety Analysis Report contains a description of the bounding

analysis for a feedwater line break. This inspection will be tracked as an inspection

followup item (50-382/9710-05).

They noted that the most severe reactor coolant system pressure spike occurred

before emergency feedwater was delivered to the steam generator. They also

stated that they had performed a preliminary analysis, which indicated that even if

emergency feedwater is lost, all subsequent reactor coolant system pressure spikes

were less severe than the initial pressure spike.

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The inspector noted some inconsistencies in timing for emergency feedwater

initiation, however, in general the inspector confirmed that the most severe reactor

pressure peak occurred prior to emergency feedwater initiation. Nevertheless, the

inspector noted that emergency feedwater is credited as a system, which will be

used for removing residual heat. While the licensee's point was true, the inspector

determined it did not invalidate the violation.

The licensee also changed the thermal hydraulic modeling assumptions for the

feedwater line break to be more rea:istic and credit the cooling effect of steam from

generator with the broken feedwater line. As the steam generator level drops from

the feedwater ring to the top of the steam generator tube sheet (approximately

30 feet) due to steaming, heat will be removed from the reactor coolant system.

The licensee stated that the previous analyses had not taken credit for this cooling

effect. They stated that previous analyses had assumed that all of the steam

generator inventory went out the break as a saturated liquid without cooling the

reactor coolant system.

The inspector determined the modeling assumption change was physically

reasonable, but less conservative than the assumption in the original

analysis. After assuming that charging was secured as stated above and using

more realistic modeling assumptions, the licensee was able to conclude that

Calculation EC-M96-004, Revision A, had acceptably determined minimum required

emergency feedwater pump flow. However, the inspector noted that the successful

outcome was predicated on the use of new less conservative assumptions.

10 CFR Part 50, Appendix B, Criterion ill, states that measures shall be established

to assure that applicable regulatory requirements are correctly translated into

specifications, drawings, procedures and instructions. 10 CFR Part 50, Appendix

A, Criterion 34, requires that suitable redundancy in components and features shall

be provided to assure that, for onsite electric power system operation (assuming

offsite power is not available) and for offsite electric power system operation

(assuming onsite power is not available), the system safety function can be

accomplished assuming a single failure. The failure of the licensee to fully evaluate

the feedwater line break accident, assuming offsite power was available as required

by 10 CFR Part 50, Appendix A, Criterion 34, is the fourth example of a violation of

10 CFR Part 50, Appendix B, Criterion ill (50-382/9710-01).

Conclusions - The inspector concluded that, prior to NRC questioning, the licensee

had not fully evaluated the feedwater line break accident, assuming offsite power

was available as required by 10 CFR Part 50, Appendix A, Criterion 34. This

inadequate design control is a violation of 10 CFR Part 50, Appendix B, Criterion 111.

After assuming prompt operator action to secure charging and more realistic heat

removal factors, the licensee was able to demonstrate that the facilities response to

a feedwater line break with offsite power available was acceptable, with respect to

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both reactor pressure, pressurizer level and steam generator inventory. The

inspector was not able to determine whether these changes in accident analysis

assumptions were consistent with the current license basis. Further inspection to

evaluate the license basis in this area will be tracked as an inspection followup item.

E8.6 1 Closed) Unresolved item 50-382/96202-13: Adequacy of tornado protection for

ultimate heat sink.

Backaround -In NRC Inspection Report 50-382/96-202, the NRC team had

identified that there were some ultimate heat sink components that may not be

protected from tornado missiles. The team was concerned that the lice 1see had

previously identified this issue, but the corrective actions did not adequately address

the identified conditions. Specifically, some electric power conduits were above the

cooling tower walls of the cooling towers and were not protected from horizontal

missiles; the dry cooling tower fans and fan motors, which were credited in the

accident analysis, were not protected from high-trajectory (i.e., almost vertical)

tornado missiles; and the licensee had not developed a clear design basis for

tornado protection.

Insoector Followuo - The inspector toured the ultimate heat sink to determine the

adequacy of design modifications, and reviewed the licensee's design basis

document for tornado missile protection, and historical documents about the

tornado design basis to determine the original licensing basis for the ultimate heat

sink. The inspector also reviewed previous condition reports, which documented

problems with the tornado design basis for the ultimate heat sink, to assess the

licensee's corrective actions.

Partially in response to inspection team questions, the licensee identified cables

and conduits that were potential targets of missiles and initiated Condition

Report 96-1591 to document this issue. The licensee performed modifications to

the plant to limit possible tornado damage. The inspector determined that the

previously exposed electrical conduits from both trains were either routed lower

or through protected areas.

The inspector was concerned that some "B" train cables had not been routed

through protected areas and were still susceptible to high trajectory or near vertical

missiles, in a related issue, the licensee did not take corrective action to protect the

dry cooling tower f ans and fan motors, which were also vulnerable to high-

trajectory (i.e., almost vertical) tornado missiles. The licensee stated that their

design basis allowed for these vulnerabilities.

Subsequent to NRC Inspection 50-382/96-202, the licensee developed a design

basis document for tornado missiles (Entergy letter to the NRC, dated January 31,

1997), which was consistent with the Final Safety Analysis Report. As stated

above, some "B" train electrical conduits and the dry cooling tower fans and fan

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motors are still susceptible to a high trajectory missiles. The licensee pointed out

that the original Final Safety Analysis Report, Section 9.2.5.3.2, stated that design

features help ensure that the wet and dry cooling towers have a low probability of

damage from tornado missiles.

The licensee believed that the probability of a missile hitting a conduit or dry cooling

tower motor is very remote, below 10 4/ reactor year. The licensee was in the

process of formally proving this probability analysis for which plant structure can be

hit by tornado missiles, to provide a basis for the 10 4/ reactor year number. The

licensee intended to use a computer code to project where missiles could strike

structures on the plant site, and run this computer code enough to provide the basis

that they had adequately protected the ultimate heat sink. The licensee projected

that this analysis would be completed by June 30,1997.

The inspector reviewed the licensee's design basis document to determine

conformance with Regulatory Guide 1.76, " Design Basis Tornado for Nuclear Power

Plants," Revision O. The design basis document noted the assumed wind speeds

and projected effects of missiles generated by a tornado near the plant. These

characteristics for the licensee's design basis tornado met the characteristics in  !

Regulatory Guide 1.76, " Design Basis Tornado for Nuclear Power Plants," with two

exceptions. The licensee used a maximum translational tornado speed of 60 mph

versus 70 mph in Regulatory Guide 1.76. The licensee used 1.0 psi /second as the

rate of pressure drop versus 2.0 psi /second in Regulatory Guide 1.76.

As stated above, the NRC team was concerned that the licensee had previously

identified this issue, but the corrective actions did not adequately address the

identified conditions, in Condition Report 95-1345, dated December 18,1995, the i

licensee noted that they could not find adequate documentation to prove the

ultimate heat sink was adequately protected. The licensee determined that the

issue was only a documentation issue, and not an operability issue since:

1. A loss-of-coolant accident is not considered concurrent with a tornado.

2. The NRC accepted Waterford 3 ultimate heat sink design in a safety

evaluation.

3. Each cell of a dry tower (a cellis 20 percent of the dry cooling tower) is

protected from other cells by concrete walls that would resist the design

basis missile to one cell.

4. The tornado protection for the dry cooling tower motors is not valid because

the protection concern for the dry towers is loss-of-camponent cooling water

system flow, and the loss of fans does not prevent a loss of this safety

function. .

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In subsequent Condition Report 96-1563, dated October 3,1996, the licensee

again noted that they could not find documentation to support the basis of Updated

Final Safety Analysis Report commitments to protect the ultimate heat sink from

tornadoes. Neither condition report specified any specific corrective action.

10 CFR Part 50, Appendix B, Criterion XVI, " Corrective Action," states measures

shall be established to assure that conditions adverse to quality, such as failures,

malfunctions, deficiencies, deviations, defective material and equipment are

promptly identified and corrected. In the case of significant conditions adverse to

quality, the measures shall assure that the cause of the condition is determined and

the corrective action taken to preclude repetition. The identification of the

significant condition adverse to quality, the cause of the condition, and the

corrective actions taken shall be documented and reported to appropriate levels of

management.

The f ailure of the licensee to take action when they recognized a potential

significant condition adverse to quality (i.e., that the ultimate heat sink may

not be adequately protected from a tornado) is one example of a violation of

10 CFR Part 50, Appendix B, Criterion XVI (50 382/9710-06).

Conclusions - The licensee did not have a design basis document for tornado

missiles at the time of the previous inspection, but has subsequently written one.

The licensee and the inspection team identified several pieces of unprotected

ultimate heat sink equipment, which the licensee modified to improve protection.

The design modification was substantially complete at the time of the inspection.

The licensee questioned, in previous condition reports, the adequacy of the tornado

design basis for the ultimate heat sink, but did not take action until questioned by

the NRC. The inspector concluded that the licensec missed two opportunities to

identify a significant condition adverse to quality when they recognized that the

ultimate heat sink may not be adequately protected from tornados. The failure to

identify and correct a significant condition adverse to quality is a violation of

10 CFR Part 50, Appendix B, Criterion XVI, " Corrective Action."

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E8.7 (Closed) Unresolved item 50-382/96202-17: Evaluations for nonconforming Crosby

relief valve installations.

Backoround -In NRC Inspection Report 50-382/96-202, the NRC team had

identified that: (1) the licensee performed an inadequate review of NRC Information

Notice 92-64, " Nozzle Ring Settings on Low Pressure Water-Relief Valves"; (2) as a

result maintenance personnel incorrectly set some safety-related relief valve guide

rings; and (3) the licensee left six safety related relief valves in service with

incorrect guide ring settings without performing and an operability evaluation or a

10 CFR 50.59 safety evaluation,

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Inspector Followuo - The inspector assessed licensee actions following their review '

of NRC Information Notice 92-64, " Nozzle Ring Settings on Low Pressure Water- l

Relief Valves." The inspector reviewed previous condition reports which

documented problems with the incorrect setting of relief valve nozzle rings, to .

assess the licensee's corrective actions. l

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As noted in NRC report 50-382/96-202, the licensee did a review of NRC

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Information Notice 92-64 and implemented several corrective actions. However, I

the licensee did not implement the recommendation to correct their maintenance '

instruction for determining the zero reference point for nozzle ring settings. The

notice alerted licensee's to the fact that the method for determining the zero l

reference point varies depending on the model of relief valve. Nozzle rings affect

the porting area immediately around the valve disc. If the nozzle rings are set  ;

improperly, the valve reseat pressure after lifting will be different than desired.

After discussions with the NRC team, the licensee initiated Condition

Report 96-0463, which documented the inadequate control of safety and relief

valve nozzle settings, caused by the inadequate review of Information Notice 92-64. ,

in addition, the licensee changed Procedure MM-007-01, " Safety and Relief Valve

Bench Testing and Maintenance," to include the proper method to determine the

nozzle ring zero reference point on a relief valve. The inspector reviewed the

procedure change and concluded it would prevent future problems with setting of

relief valves.

In response to the NRC teams questions, the licensee also initiated Condition

Report 96-0429, which documented severalinstances where nozzle rings had been

set incorrectly. The licensee performed an operability evaluation for the six

valves, which were in service with incorrect settings, which determined that it was

acceptable to leave these valves inservice until the outage. The inspector

determined the operability evaluation was acceptable.

As noted in the root cause evaluation for Condition Report 96-0429, there were

several prior condition reports (95-1115,951263), which had documented

inaccurate blowdown ring settings. The inspector determined these errors could

have been prevented if the information in the notice had been implemented properly.

10 CFR Part 50, Appendix B, Criterion XVI, states measures shall be established to

assure that conditions adverse to quality, such as failures, malfunctions,

deficiencies, deviations, defective material and equipment are promptly identified

and corrected. In the case of significant conditions adverse to quality, the

measures shall assure that the cause of the condition is determined and the

corrective action taken to preclude repetition. The identification of the significant

condition adverse to quality, the cause of the condition, and the corrective actions

taken shall be documented and reported to appropriate levels of management.

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i The failure to promptly identify and correct nozzle ring setting deficiencies, a

significant condition adverse to quality, is a second example of a violation of

r 10 CFR Part 50, Appendix B, Criterion XVI (50-382/9710-06).

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Conclusion - The licensee inadequately reviewed NRC Information Notice 92-64,

" Nozzle Ring Settings on Low Pressure Water-Relief Valves," and continued to

utilize incorrect instructions for setting certain relief valve nozzle ring settings.

The f ailure to do an adequatu review of the information notice resulted in erroneous

nozzle ring settings for installed valves. The inspector concluded that the licensee

missed an opportunity to identify and correct nozzle ring setting deficiencies,

similar to those described in Information Notice 96-24, which was a violation of

l 1C CFR Part 50, Appendix B, Criterion XVI, " Corrective Action."

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V. Manaaement Meetinas

X1 Exit Meeting Summary

The inspectors presented the results of the inspection to members of licensee

management at the conclusion of the inspection on May 2,1997. In addition,

the inspector contacted the director of engineering to discuss Example 4 of

which had not been fully characterized at the exit.

Violation 50-382/9710-01,

The licensee's comments were evaluated and incorporated into the inspection

report.

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ATTACHMENT

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SUPPLEMENTAL INFORMATION

PARTIAL LIST OF PERSONS CONTACTED

Licensee

R. Burski, Director, Plant Modifications and Construction

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M. Byram, Safety Analysis

T. Gaudet, Manager, Licensing

P. Gropp, Mechanical Supervisor, Design Engineering

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J. Howard, Manager, Procurement Programs Engineering

P. Jackson, Electrical /l&C Supervisor, Design Engineering

T. Leonard, General Manager, Plant Operations

J. O'Hearn, Director, Training

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J. Holman, Manager, Safety Analysis

l D. Matthews, Specialist, Licensing

l B. Proctor, Superintendent, Systems Engineering - NSSS

l P. Prasankumar, Manager, Design Electrical / Instrumentation and Control

A. Rustaey, Safety and Engineering Analysis

L. Thomas, Licensing Supervisor

l D. Vinci, Manager, Plant Engineering,

l G. Wilson, Acting Quality Assurance Manager

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P. Wagner, Consultant

A. Wrape, Director, Design Engineering

INSPECTION PROCEDURES USED

92903 Engineering - Followup

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ITEMS OPENED, CLOSED, AND DISCUSSED

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50-382/9710-01 VIO Design control violation involving: (1) failure to translate

cold weather requirements for the wet cooling tower into

operating procedures, (2) failure to translate seismic

requirements for the reactor shield building door into the

installation instruction, (3) failure to provide adequate

coordinat. ion between the mechanical and the electrical

decign organizations, and (4) failure to evaluate emergency

i feedwater flow rates considering offsite power is available.

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VIO Failure to provide adequate basis that a reduction in diesel

50-382/9710-02

generator fuel oil margin did not involve an unresolved

safety question.

50 382/9710-03 URI (1) Whether or not the licensee can assume

contactors open upon loss of holding voltage and still

meet the verification requirements of Technical

Specification 4.8.1.1.2.e, and (2) whether or not the

licensee is required to update the applicable calculations

(including the diesel fuel oil calculation) to assume

contactor loads do not shed.

VIO Inadequate implementation of Technical

50-382/9710-04

Specification 4.8.1.1.2.e involving: (1) the failure

to document verification that the pressurizer heater

breakers opened as required, and (2) the failure to

verify that loads controlled by process variables

(e.g., Shutdown Heat Exchanger A and B room coolers, the

Component Cooling Water Heat Exchanger A and B room

coolers, and Control Room Heater EHC-34 akt Switchgear

Room Heater EHC-36) shed and started on their

corresponding load sequencer block.

IFl Determine if the Final Safety Analysis Report contains a

50-382/9710-05 description of the bounding analysis for a feedwater line

break.

VIO Corrective action violation involving: (1) the failure to

50-382/9710-06 ensure ultimate heat sink was adequately protected from a

tornado, and (2) the failure to promptly correct problems

with the relief valve razzle rirs settings identified in NRC

Information Notice 92-64, " Nozzle Ring Settings on Low

Pressure Water-Relief Valves."

Closed

URI Adequacy of cold weather opercing instruction for

50-382/96202-04

auxiliary component cooling water system (Section 08.1).

URI The adequacy of emergency diesel generator fuel oil

50-382/96202-05 volume including 10 CFR 50.59, " Changes, Tests, and

Experiments," (Section E8.1).

URI The adequacy of administrative controls to ensure the

50-382/96202-06

emergency diesel generator loading and fuel oil calculations

and the related Final Safety Analysis Report tables are

updated when the mechanical design is changed

(Section E8.2).

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50-382/96202-07 URI (1) The adequacy of 18-month Technical Specification

integrated tests to verify deenergization of the emergency

buses and load shedding from the emergency buses, and

(2) the adequacy of dieselloading and fuel oil consumption

calculations to ensure that untested loads were properly

considered in the calculations (Section E8.3).

50-382/96202-09 URI Adequacy of maintenance hatch shield door installation

instructions (Section M8.1).

50 382/96202-11 URI The adequacy of the licensee's corrective actions for an

error in the bases section of the Technical Specification

(Section E8.4).

50-382/96202-12 URI The heat removal capability of the emergency feedwater

system (Section E8.5).

50-382/06202-13 URI Adequacy of tornado protection for ultimate heat sink

(Section E8.6).

50-382/96202-17 URI Evaluations for nonconforming Crosby relief valve

installations (Section E8.8).

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Entergy Operations, Inc. -4-

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E-Mail report to T. Boyce (THB)

E-Mail report to NRR Event Tracking System (IPAS)

E Mail report to Document Control Desk (DOCDESK)

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bec to.DCD (IE01)

bec distrib. by RIV:

Regional Administrator Resident inspector i

DRP Director DRS-PSB

Branch Chief (DRP/D) MIS System

Project Engineer (DRP/D) RIV File

Branch Chief (DRP/TSS)

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DOCUMENT NAME: R:\_WAT\WT9710RP.LJS

To receive copy of document, Indicate in box: "C" = Copy wthout enclosuK "E" = Copy with enclosures "N" = No copy

RIV:Rl:EB C:EB D:DRf ), g D:DRSj})/>  !

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LJSmith/Imb N _ . CAVanDenbqty TPGdy )(@'/,/ _ ATHo $ 19 ]

06/T/97 VW 06/(-/97 NW 06/GgP4 fd 066/9T

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OFFICIAL RECORD COPY

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