Text

Discusses 970415 Predecisional Enforcement Conference Re Apparent Failures to Maintain Licensing Bases,As Discussed in Insp Repts 50-298/96-24 & 50-298/96-31.W/draft Apparent Violations,Agenda & Errata Sheets for Insp 50-298/96-24
	ML20141E139
Person / Time
Site:	Cooper
Issue date:	05/14/1997
From:	Howell A; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:	Horn G; NEBRASKA PUBLIC POWER DISTRICT
Shared Package
ML20141E145	List: ML20141E139; ML20141E961;
References
	50-298-96-24, 50-298-96-31, EA-97-017, EA-97-17, NUDOCS 9705200344
	Download: ML20141E139 (58)
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G. R. Horn, Senior Vice President

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Nebraska Public Power District

141415th Street

Columbus, Nebraska 68601

SUBJECT:

APRIL 15,1997, PREDECISIONAL ENFORCEMENT CONFERENCE

REGARDING APPARENT VIOLATIONS IN NRC INSPECTION

REPORTS 50-298/96-24 AND 50-298/96-31

AND ERRATA SHEETS FOR NRC INSPECTION REPORT 50-298/96-24

Dear Mr.

mrn:

This refers to the meeting with you and your staff conducted in the Region IV office

on April 15,1997. This Predecisional Enforcement Conference was convened to discuss

apparent failures to maintain the licensing bases as discussed in NRC Inspection

Reports 50-298/96-24 and 50-298/96-31. The meeting was conducted in accordance

with the enclosed agenda. During the meeting, the NRC described the enforcement

process and discussed the draft apparent violations which are also attached. NRC

personnel also requested that your staff discuss three additional issues.

We were concerned about the apparent failure of your staff to maintain the margins

described in a license amendment submittal for the standby liquid control system

Technical Specification relief valve surveillance requirement.

We wanted to better understand the basis for your response to

Violation 50-298/96030-02. We noted that you expressed the view that the

violation would have more appropriately been described as a 10 CFR 50.71(e)

violation, rather than a 10 CFR 50.59 violation.

We wanted to understand whether or not Licensee Event Report 96-14 identified a

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loss of control of the licensing basis, which was similar to the issues identified in

the apparent violations.

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Your staff presented their views regarding these issues. The handout which was provide

as the basis for the discussion is also enclosed.

9705200344 970514

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During the predecisional enforcement conference, Mr. Phil Graham noted two potential

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errors in NRC Inspection Report 50-298/96-24. He believed that the team's conclusion in

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Section E1.3 and the executive summary that the " licensee modified their standby liquid

control system to make it less reliable," was inaccurate. He stated that the modification

was performed for the purpose of adding an additional computer monitoring point and that

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- the engineering staff initially overlooked the single failure implications because they viewed

the system as nonessential. The NRC agrees with Mr. Graham's clarification, in that there

was no evidence that the engineering staff set out with the intent of reducing the reliability

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of the system. However, as a result of the modification, the final design was less reliable

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because of increased vulnerability to single failure. This clarification is included in the

attached errata sheets (pages iv and 5) for NRC Inspection Report 50-298/96-24.

The second error related to the application of quality assurance requiremerns to the

standby liquid control system. Mr. Graham stated that his staff had incorrectly told the

NRC that the standby liquid control system was not subject to any quality assurance

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program requirements. He committed to provide an accurate description of the quality

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assurance requirements for the standby liquid control system and to evaluate Section E7.4

of the inspection report to ensure all of the identified quality assurance program

weaknesses are addressed. The NRC plans additional inspection followup of this

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information which will be tracked as inspection Followup item 50-298/9624-12.

Nevertheless, the inspection report accurately describes the licensee statements and NRC

conclusions made at the end of the inspection. Therefore, no clarifications are needed.

During the conference, Mr. Pelletier also made a second commitment. During discussion of

the apparent reduction in margin related to the standby liquid control system relief

valve Technical Specification surveillance requirement, he stated that he would determine

whether or not a 10 CFR 50.59 evaluation had been performed when the nonconformance

was first identified. During a subsequent conversation, he confirmed that a 10 CFR 50.59

evaluation had not been performed. Mr. Pelletier also stated during the predecisional

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enforcement conference that the District planned to provide the NRC staff with

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the technical basis for the continued acceptability of the current Technical

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Specification surveillance requirement, after considering the apparent reduction in

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margin. The NRC was concerned that the reduction in margin should have been identified

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during a 10 CFR 50.59 evaluation as a potential unreviewed safety question. As discussed

in NRC Inspection Report 50-298/96-24, no current operability issues exist because the

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current relief valve setting includes the margin credited in the license amendment

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submittal. During a teleconference on May 1,1997, we concluded that the current

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Technical Specifications were inadequate, in that the setpoint of the standby liquid control

system relief valve did not support safe plant operation as required by 10 CFR 50.36. As

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such, we requested that you amend the Technical Specifications to correct this

discrepancy. The NRC plans additional inspection followup of this issue, which will be

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tracked as Unresolved item 50-298/9624-13.

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We request that you confirm the two regulatory commitments made during this meeting

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regarding (1) providing an accurate description of the Quality Assurance Requirements for

the standby liquid control system and evaluating Section E7.4 of the Inspection Report to

address program weaknesses; and (2) providing an evaluation for the acceptability of the

former Technical Specification surveillance requirement for the standby liquid control

system relief valve setpoint that existed at the time of the inspection. Although, we

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recognize that this setpoint has been changed at our request, we still need to understand

your basis for the previous value. Please confirm these regulatory commitments in writing

within 30 days of the date of this letter.

You will be advised by separate correspondence of the results of our deliberations

regarding the apparent violations discussed during the conference. No response regarding

these apparent violations is required at this time.

In accordance with Section 2.790 of the NRC's " Rules of Practice," Part 2, Title 10, Code

of Federal Regulations, a copy of this letter will be placed in the NRC's Public Document

Room.

Should you have any questions concerning this matter, we will be pleased to discuss them

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with you.

Sincerely,

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f Arthu T. Howell Ill, Director

Division of Reactor Safety

Docket No.: 50-298

License No.: DPR-46

Enclosures:

1. Attendance List

2. Agenda

3. Apparent Violations

4. Licensee Presentation

5. Errata Sheets for NRC Inspection Report 50-298/96-24 (pages iv and 5)

cc w/ enclosures:

John R. McPhail, General Counsel

Nebraska Public Power District

P.O. Box 499

Columbus, Nebraska 68602-0499

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P. D. Graham, Vice President of

Nuclear Energy

Nebraska Public Power District

P.O. Box 98

Brownville, Nebraska 68321

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B. L. Houston, Nuclear Licensing

and Safety Manager

Nebraska Public Power District

P.O. Box 98

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Brownville, Nebraska 68321

R. J. Singer, Manager-Nuclear

Midwest Power

907 Walnut Street

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P.O. Box 657

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Des Moines, Iowa 50303

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Mr. Ron Stoddard

Lincoln Electric System -

11th and O Streets

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Lincoln, Nebraska 68508

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Randolph Wood, Director

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Nebraska Department of Environmental

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Quality

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P.O. Box 98922

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Lincoln, Nebraska 68509-8922

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Chairman

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Nemaha County Board of Commissioners

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Nemaha County Courthouse

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1824 N Street

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Auburn, Nebraska 68305

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Cheryl Rogers, LLRW Program Manager

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Environmental Protection Section

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Nebraska Department of Health

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301 Centennial Mall, South

P.O. Box 95007

Lincoln, Nebraska 68509-5007

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Dr. Mark B. Horton, M.S.P.H.

Director

Nebraska Department of Health.

P.O. Box 950070

Lincoln, Nebraska 68509-5007

R. A. Kucera, Department Director

of Intergovernmental Cooperation

Department of Natural Resources

P.O. Box 176

Jefferson City, Missouri 65102

Kansas Radiation Control Program Director

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ENCLOSURE 1

ATTENDANCE APRIL 15,1997

PREDECISIONAL ENFORCEMENT CONFERENCE

Nebraska Public Power District Attendees

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G. Horn, Senior Vice President

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P. Graham, Vice President, Nuclear

B. Houston, Licensing Manager

J. Pelletier, Senior Engineering Manager

M. Boyce, Engineering Project Manager

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R. Wenzel, Engineering Project Manager

R. Sessoms, Senior Quality Assurance Manager

B. Victor, Sanior Licensing Specialist

M. Holmes, Control Room Supervisor

R. Stoddard, Lincoln Electric

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B. Turnbull, Mid America Energy

_NRC Attendees

E. Merschoff, Regional Administrator, Region IV

,

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A. Howell, Director, Division of Reactor Safety

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K. Brockman, Deputy Director, Division of Reactor Projects

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C. VanDenburgh, Chief, Engineering Branch, Division of Reactor Safety

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E. Collins, Chief, Projects Branch D, Division of Reactor Projects

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C. Skinner, Resident inspector, Division cf Reactor Projects

L. Smith, Reactor inspector, Division of Reactor Safety

W. Brown, Office of General Counsel

T. Reis, Enforcement Specialist, Office of Enforcement

F. Akstulewicz (by phone), Section Chief, Generic Issues and Environmental Project

Branch, Office of Nuclear Reactor Regulation

E. McKenna, (by phone), Senior Reactor inspector, Generic issues and Environmental

Project Branch, Office of Nuclear Reactor Regulation

J. Hall, Project Manager, Project Directorate IV-1, Office of Nuclear Reactor Regulatio

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ENCLOSURE 2

PREDECISIONAL ENFORCEMENT CONFERENCE AGENDA

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CONFERENCE WITH NEBRASKA PUBLIC POWER DISTRICT

APRIL 15,1997.

NRC REGION IV, ARLINGTON, TEXAS

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1.

INTRODUCTIONS / OPENING REMARKS - ELLIS MERSCHOFF, RA

2.

ENFORCEMENT PROCESS - MICHAEL VASQUEZ, RIV ENFORCEMENT SPECIALIST

3.

APPARENT VIOLATIONS & REGULATORY CONCERNS - ART HOWELL, D/DRS

4.

LICENSEE PRESENTATION -

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BREAK (10-MINUTE NRC CAUCUS IF NECESSARY)

6.

RESUMPTION OF CONFERENCE

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CLOSING REMARKS - LICENSEE

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8.

CLOSING REMARKS - ELLIS MERSCHOFF, RA

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ENCLOSURE 3

APPARENT VIOLATIONS *

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PREDECISIONAL ENFORCEMENT CONFERENCE

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NEBRASKA PUBLIC POWER DISTRICT

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COOPER NUCLEAR STATION

APRIL 15,1997

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NOTE: THE APPARENT VIOLA TIONS DISCUSSED A T THIS PREDECISIONAL

ENFORCEMENT CONFERENCE ARE SUBJECT TO FURTHER REVIEW AND MA Y BE

REVISED PRIOR TO ANY RESULTING ENFORCEMENT ACTION.

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APPARENT VIOLATION

A.

10 CFR 50.71(e) requires that the Updated Safety Analysis Report be updated

periodically to assure that the information in the Updated Safety Analysis Report

contains the latest material developed. Subsequent revisions are required to reflect

all changes up to a maximum of 6 months prior to the date of filing.

Contrary to the above,

1.

Since initial construction of the facility, Updated Safety Analysis Report,

Section Xil-2.3.5.2.2, "[ Seismic Analysis] Piping," and Updated Safety

Analysis Report, Appendix C, " Structural Loading Criteria," Section 3.3.3.2,

" Piping Seismic Analysis," (and the Final Safety Analysis Report) describe, in

detail, the procedure for dynamically analyzing Class-l seismic piping systems

without restricting the requdement for dynamic analysis to large bore piping.

However, as of November 1,1996, (and since initial construction), the

dynamic seismic analysis described in the Updated Safety Analysis Report

was not performed for 2-inch and under piping systems and the Updated

Safety Analysis Report was not updated to reflect the change.

2.

Updated Safety Analysis Report, Section Vll-2.3.5.1, " Alternate Rod

Insertion (ARI)," indicated that the time delay for initiating alternate rod

insertion should be applied only to the low reactor water level initiation signal

and not for the high reactor pressure initiation signal. This was consistent

with Drawing 945E632, "ARl/ATWS Recirc Pump Trip," Sheets 3 and 7,

dated January 15,1995. However, as of November 1,1996, Updated

Safety Analysis Report, Section lll-5.5.3.4, " Alternate Rod Insertion," was

incorrectly updated to state that the time delay for initiating alternate rod

insertion should be applied to both the low reactor water level initiating

signal and ibe high reactor pressure initiating signal.

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3.

Updated Safety s'nalysis Report, Section X-10.3.2 "[ Heating, Ventilation and

Air Conditioning Systems] Station Heating System," states that winter

design temperatures for the system are given in Table X-10-1. Table X-10-1,

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"[ Heating, Ventilation and Air Conditioning Systems] Station Heating System

Design Temperatures (Winter)," states that the normal minimum indoor

temperature for the reactor building is 50 F. The equipment containing the

solution is installed in a room in the reactor building. However, as of

November 1,1996, (and since initial construction) Updated Safety Analysis

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Report, Section 111-9.3 "[ Standby Liquid Control System] Description,"

inaccurately claims that "The equipment containing the solution is installed in

a room in which the air temperature is to be maintained within the range of

65 F to 100

F."

Updated Safety Analysis Report, Section 111-9.3 describes

the contingency plans for heater failure down to 65 F. On ten different days

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between December 5,1995, and February 4,1996, the room temperature

was below 65 F.

THIS APPARENT VIOLA TION IS SUBJECT TO FURTHER REVIEW AND MA Y

BE REVISED

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APPARENT VIOLATION

4.

Design Change 86-34A, Revision 0, dated March 4,1988, "SLC/ATWS

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Modifications," indicated that the Standby Liquid Control System relief valve -

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settings were intended to be based on assuring injection into the reactor

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above the anticipated transient without scram reactor pressure conditions,

which would equal the reactor safety / relief valves'setpoints plus the

accumulation at the maximum anticipated transient withcut scram steam

flow, (i.e.', approximately 1100 psig plus the static head in the reactor

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vessel). However, as of November 1,1996, Updated Safety Analysis

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Report, Section 111-9.4, "[ Standby Liquid Control System] Safety Evaluation,"

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had not been accurately updated. Specifically, Section 111-9.4 continued to

state that "The SLC system and pumps have sufficient pressure margin, up

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to the allowed system relief valve setting range of 1450 to 1680, to assure

solution injection into the reactor above +he normal [ emphasis added]

pressure of approximately 1030 psig in : lie bottom of the reactor." The old

SLC system design was based on normal operation.

5.

Updated Safety Analysis Report, Section lll.9.3, "[ Standby Liquid

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Control] Description," states that at the minimum room temperature of

65 F, the maximum permitted solution concentration is 12.5 weight percent.

Section 111.9.3 also states that a concentration of 11.5 percent corresponds

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to an adjusted saturation temperature of 61 F. The Updated Safety Analysis

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Report adjusted saturation temperature includes a 10 F margin over

saturation and corresponds to the Technical Specification minimum allowable

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temperature. However, Technical Specification Figure 3.4.2, " Percent

Sodium Pentaborate by Weight of Solution versus Temperature," indicates

that at 65 F the maximum permitted concentration was 12.1 percent. At

11.5 percent concentration, the minimum allowable temperature was 62 F.

6.

Updated Safety Analysis Report, Section IV-9.3, "[ Reactor Water Cleanup

System] Description," states that "In the inlet piping to the cleanup

recirculation pumps, two motor operated isolation valves, one on either

side of the primary containment, are automatically closed . . . as shown

in Figure IV-9-4 [by the] standby liquid control system actuation."

Design Change 86-34A, Revision 0, dated March 4,1988, "[ Standby Liquid

Control / Anticipated Transient Without Scram] SLC/ATWS Modifications,"

changed the reactor water cleanup systems isolation valves' control logic

such that two inputs from the standby liquid control system are required

to isolate both valves. However, as of November 1,1996, Updated Safety

Analysis Report, Figure IV-9-4, "[ Reactor Water Cleanup System] RWCU

Leak Detection and Isolation," was not updated to describe Design

Change 86 34A. As of November 1,1996, Figure IV-9-4 still showed that

a single input from the standby liquid control system willinitiate isolation

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of both motor-operated isolation valves.

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THIS APPARENT VIOLA TION IS SUBJECT TO FURTHER REVIEW AND MA Y

BE REVISED

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APPARENT VIOLATION

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7.

Technical Specification Figure 3.4.1 and Calculation NEDC 93-142,

Revision 0, dated August 2,1993, "SLC Storage Tank Setpoints and

Concentration Requirements," required the minimum volume of sodium

pentaborate solution in the standby liquid control system storage tank

to be between 3132 and 4414 gallons for concentrations between 16 and

11.5 percent, respectively. Updated Safety Analysis Report, Section 111.9.4,

"[ Standby Liquid Control System] Safety Evaluation," and Technical Specifications 4.4.A.1 and 4.4.A.2.b, " Standby Liquid Control System,"

stated that the minimum required flow rate for the standby liquid control

system pumps was 38.2 gpm. Based on these values, the minimum time for

one pump to inject the required volume of solution would have been 82 to

116 minutes, and 41 to 58 minutes for two pumps. However, the following

Updated Safety Analysis Report sections were not correctly updated when

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the standby liquid control injection criteria was changed to meet the ATWS

rule:

a.

Updated Safety Analysis Report, Section 111.9.3, "[ Standby Liquid

Control System] Description," stated that "Each positive displacement

pump is capable of injecting the required weight of solution into the

reactor in 53 to 120 minutes, independent of the amount of solution

in the tank (within the required volume), and the pump rate (within

the specified Technical Specification limits)."

b.

Updated Safety Analysis Report, Section 111.9.4, "[ Standby Liquid

Control System] Safety Evaluation - Compliance with the NRC ATWS

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Rule," stated, "With the simultaneous operation of both positiva

displacement pumps, the solution can be injected into the reactor in

27 to 60 minutes, independent of the amount of solution in the tank

(within the required volume), and the pump rates (within the specified

Technical Specification limits)."

8.

Alarm Procedure 2.3.2.28, Revision 26, dated March 15,1996, " Panel

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9-5-2, Window F-8, [ Standby Liquid Control] SLC Tank Hi/ Low Level,"

specified the high and low alarm setpoints for the standby liquid control

system sodium pentaborate storage tank as 84 and 74 percent respectively.

Surveillance Procedure 6.SLC.601, Revision 0, dated November 17,1995,

"[ Standby Liquid Controll SLC Tank Sampling," indicated that these setpoints

corresponded to 3835 and 3378 gallons respectively. However, Updated

Safety Analysis Report, Section 111.9.3, "[ Standby Liquid Control System]

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Description," inaccurately stated that these alarms were set at 3850 and

3350 gallons respectively.

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THIS APPARENT VIOLA TION IS SUBJECT TO FURTHER REVIEW AND MA Y

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BE REVISED

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APPARENT VIOLATION

9.

In 1994, during the Surveillance Test Validation Program Status review, the

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licensee identified at least two discrepancies in the Updated Safety Analysis

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Report, which were not corrected in the July 22,1996, update to the

Updated Safety' Analysis Report.

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Updated Safety Analysis Report, Table Vll-3-1, " Pipeline Penetrating

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Containment," Note 4, incorrectly stated that the control' rod drive

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system solenaid valves open during a reactor scram. On reactor

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SCRAM the solenoid valves remain closed and the air-operated

SCRAM valves open to insert the control rods and to exhaust water

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to the SCRAM discharge volume,

b)

Updated Safety Analysis Report, Section Vil-4.5.44, "[ Core Spray .

System Control and instrumentation) Core Spray Valve Control,"

incorrectly stated that two pressure switches monitor system

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pressure (for the low pressure permissive). -In addition, it indicates

that either switch can initiate opening of the discharge valves for core

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spray. There actually are four pressure switches du,bned in~a 1-out-

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of-2 twice logic and a minimum of two switches are required to

actuate to initiate opening of the core spray valves.

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10.

As of March 16,1996, Updated Safety Analysis Report,' Table V-2-2,

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" Penetration Schedule," Pages V-2 9 to V-2-12, failed to list all the

penetrations; incorrectly listed the quantity of lines in three penetrations; and

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incorrectly listed line descriptions in five penetrations.

11.

As of May 4,1996, Updated Safety Analysis Report, Table V-2-7, " Testable

Primary Containment isolation' Valves," Pages V-2-44 to V-2-46, did not list

23 penetrations (X20, X-30E and -30F, X 33E and -33F, X-35A through E,

X-45D, and X-229A through L) and their associated valves.

B.

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, to the_ extent

that these changes constitute changes in the facility as described in the safety

analysis report or to the extent that they constitute changes in the procedures as

described in the safety analysis report. Ftrther these records must include a written

safety evaluation which provides the bases for the determination that the change

does not involve an unreviewed safety question.

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Contrary to the above,

1.

Updated Safety Analysis Report,Section X.8.2.8.C, " Common Mode Failure

Analysis - Fire," required that combust!bles not be located in the service

water booster pump room area since both trains of the service water system

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THIS APPARENT VIOLA TION IS SUBJECT TO FURTHER REVIEW AND MA Y

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BE REVISED

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APPARENT VIOLATION

were located in close proximity. However, Procedure 0.7.1, " Control of

Combustibles," Revision 6, allowed up to 90 pounds of wood or 5 gallons of

flammable liquid for this area of the plant. In addition, on December 2,

1996, combustible materials (rags, papers, and flammable chemicals) were

located in the service water system booster pump room. The safety

evaluation for this change in procedure was inadequate, in that a common

mode failure analysis had not been performed to justify the presence of

combustible materials in the service water booster pump area.

2.

Updated Safety Analysis Report,Section XII.2.2.7.1, " Intake Structure,"

states, in part, that in order to keep ice away from the intake structure

during cold weather, an ice deflector is installed during the winter months.

On November 26,1996, the ice deflector had not been installed because

river level was 6 feet higher than normal for this time of year, which

,

interfered with the attachment points for the deflector. On December 18,

1996, a portion of the ice deflector was subsequently installed. However,

i

the failure to install the ice deflector prior to the occurrence of ice in the

intake structure and the partialinstallation of the ice deflector were

l

configuration changes that had not been evaluated as changes to the facility.

3.

Updated Safety Analysis Report,Section IV.10.3, " Nuclear System Leakage

Rate Limits - Description," states, in part, that each containment drywell

sump has an alarm system and automatic starting sequence on rising water

level. Both containment drywell sumps are equipped with a fill rate timer

and alarm. This alarm can be set at or below ine Technical Specification

limits and would provide immediate indication when this preselected rate is

reached or exceeded. However, on December 20,1996, a safety evaluation

to address the failure of the automatic pump starting system and the failure

of the sump fill rate timer and high level alarm did not address the lack of

control room alarm, nor did a separate safety evaluation exist for this change

to the facility.

i

l

THIS APPARENT VIOLA TION IS SUBJECT TO FURTHER REVIEW AND MA Y

BE REVISED

l

{

l

!

.

4

0

ENCLOSURE 4

LICENSEE HANDOUT

l

1

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4

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.

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.

..

.

-

rammmmprmmammwmemm. . ~

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.

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.

,

1 ~~ebraska Public Power

District

,

Enforcement Conference

,

April 15,1997

!

l

,

I

i

I

Nebraska Public Power District

1

!

I

,

__

_ _.___.___ _ _

_ _ _ , - - . _

-

_- I

-- ._

_

- - - _

--

- -

. . .

.-. -

.

-

.

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mamm?Jmamlygmayawgypys a :yecme m . ' -. - .- .

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t

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f

Phil Graham

.

T

Cooper Nuclear Station

Site Vice-President

-

.

1

Nebraska Public Power District

2

.

-.

-

-.

-

. - -

-

- . . .

- - .

-

.

- - . -

- -

.

. . -

-

.

- - . - -

- -

-

- - - -

-

- _ _ _ _ _ _ _ _ _ _ _ _ _

_ _ _ _

.

e

i

AGEXDA

.

>

'

d2li%*TEf?ENT1M2fEfMEsf#2WSP

e Presentation Methodology

.

o Broad Issues

'

8

e Specific Enforcement Issues

l

,

o Regulatory issues

,

.

e Closing Remarks

i

!

l

i

l

Nebraska Public Power District

3

!

,

-

- -

-

- - - -

.;

.

i

1

,

Presentation Methodology

___,,,,,,,_____

_

,

i

e Focus on broader issues resulting from

specific findings

}

Safety conscience

FSAR philosophy

10 C.F.R. 50.59

l

I

Nebraska Public Power District

4

I

.

Safety Conscience

nn,-

-

CNS Site

o Safety is paramount at CNS

working to improve procedural adherence

improving PIR/ corrective action implementation

e NPPD management working to improve all

l

personnels' communication skills

~

e Disagreement with NRC is not indicative of

inadequate safety conscience

j

e Safety conscience and USAR/50.59 problems are

different concepts

[

Nebraska Public Power District

5

'

.

s

a

t

e

- - ,,- - - - _

.

Safety Conscience

Engineering

,

James Pelletier

Engineering Manager

!

l

!

,

I

Nebraska Public Power District

6

l

i

. . .

.

_ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _

-

p

l

.

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i

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.

i

.

Safety Conscience

_

.

i

.

Enaineerina Department

i

!

e Noted improvement of timeliness and quality of

1

engineering response

e Monthly meetings with Engineering Staff

e After SSFI, memo issued renewing management

expectations

e Provide feedback to staff through regular reviews

e Broad ongoing training

-

e Rigorous implementation of systems training course

!

Nebraska Public Power District

7

!

t

-

. -

- - - - - -

-

.

Safety Conscience

Engineerina Department

e CNS Vital Signs - Conservative Decision Making

e Initiation of Significant Conditions Adverse to Quality

-

(SCAQ) and Conditions Adverse to Quality (CAQ)

e Resolution time of SCAQ and CAQ

e Self-identification rate of Problem Identification

Reports (PIRs)

e 50.59 SORC rejection rate

e My aersonal safety conscience performance

indicators

Nebraska Public Power District

8

.

- ..

_

. - - - - _ _

-

- - -

-

.

i

I

!

ENGINEERING INITIATED SCAO AND CAQ

ITEMS

!

,

50

a

45

-

l

40

- - -

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l

35

-

30

-

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8/1/96

9/1/96

10/186

11/1S6

12/1/96

1/1/97

2/1/97

3/187

l M ENGINEERING IDENTIFICATION OF SCAQ/CAO ITEMS

Unear (ENGINEERING IDENTIFICATION OF SCAQ/CAO ITEMS) l

4/12/972:32 PM

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.

Page 1

i

!

- -

-

- - - - - - -

.

- - _ _ _ _ .________--_

______

.

t

,

.

___.______.______,=m___,s_

,

!

FSAR Philosophy

,

r

Brad Houston

,

Licensing Manager

James Pelletier

1

t

l

Engineering Manager

.

l

f

i

r

,

Nebraska Public Power District

11

i

-

- -

- - - - - - - -

-

- -

- - - - - -

- -

- - - -

.-;

.

!

i

.

FSAR Philosophy

-e-rweme-

l

'

Historical Perspective

.

e FSAR initially issued in 1971

e FSAR updates not required in industry until

'

1980 (50.71(e) issued)

e Detailin FSAR and update philosophy

consistent with industry in this time period

FSAR originally used as reference

document

!

i

l

Nebraska Public Power District

12

!

I

. _

_

_ _ _ _ _ _ _ _ _ _ - _ _ _ _ _

.

.;

i

.

FSAR Philosophy

j

--,n=~

.-

!

-!

!

Historical Perspective

l

r

e Technical specifications, design modification

i

!

packages, configuration control processes / programs,

primary tools to control plant design and safety

i

e Updates submitted for over decade with the same

-

level of detail, scope, and clarity

e Material added from 1980 to present consistent with

industry practices

e NRC increased focus beginning in 1995 timeframe

!

4

Nebraska Public Power District

13

---

-

_._---_

.

}

.

FSAR Philosophy

_ -__e__

,

l

,

1

Historical Perspective

!

'

e While 50.71(e) always required accurate USAR, level

!

of detail, scope of USAR, language clarity are

j

unsettled issues

Deputy Director NRR recently stated at

Commission meeting that NRC has not applied a

consistent standard for, nor always clearly

j

articulated the scope of, 50.71(e) or the USAR

!

NRC considering additional regulatory guidance in

'

this area

i

!

Nebraska Public Power District

14

.

-

--

!

_

--

_

--

.

Violation Examples (USAR)

Admit

.

_ __,=____ _ _

.

+

e Seismic qualification of small bore piping

o

Reactor Water Cleanup System valve logic

e Standby Liquid Control System (SLCS) room teinperature

control

'

e

SLCS solution temperature / concentration

[

e

Control Rod Drive solenoid indication

o

Core Spray System pressure switch descriptio~n

!

I

!

Nebraska Public Power District

15

!

-

- - -

- - - - - - - - - - -

- -

- - -

-

- -

-

_ - - _ _

..

.

'

Violation Examples (USAR)

Disagree

.

_ _ ,,,,,_ _ , _

.

o

SLC relief / pressure description

!

e

SLCS tank level setpoint round-off

'

e

SLCS injection time clarity

e

Rod insertion time delay

,

General Bases For Disaareement

!

e

50.71(e) requires that the USAR be accurate and contain latest.

information within the scope of the USAR

findings involve level of detail or clarity of accurate

j

information already described in USAR --- do not involve

inaccuracies or failure to update

!

Nebraska Public Power District

16

.

-

.

- _ _ _ _ _ _ _ _ _ _ _ _ _ - _ , _

-

.

_ _

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _

- _

.

.,

Violation Examples (USAR)

!

.

Disagree

.

_ _

_ , , ,

.

_

!

SLC Relief / Pressure Description

i

e USAR accurately reflects CNS design basis

t

requirements

!

e issue is level of detail needed in USAR to describe

t

certain aspects of ATWS licensing basis

e USAR Rebaselining Project will add details about

j

licensing basis as appropriate

i

l

!

,

'

Nebraska Public Power District

17

1

.

Violation Examples (USAR)

.

Disagree

- , - , , - - _ -

.

+

i

t

l

SLC Tank Level Setpoint Roundoff

!

i

'

e

USAR accurately provides nominal tank level alarm values in

gallons

j

'

e

Setpoint Change process translates nominal values into actual

settings

e

Instrument and procedure must be in whole number

i

percentages in this instance so USAR levels translated and

!

rounded in conservative ~ direction

e

USAR Rebaselining project will clarify that USAR values are

j

nominal

'

Nebraska Public Power District

18

3

i

--

-

--

-

- - -

- ---

-

---

- -

-

--

-

--

-

____

.;

.

-

!

Violation Examples (USAR)

.

Disagree

---=

SLC Injection Time Clarity

l

e USAR text accurately reflects system design basis,

'

not meant to describe system capability

e issue of clarity; NRC interpreted text to be system

1

capability not design basis requirements

j

e CNS has always understood that this text to be

design basis requirements

I

Nebraska Public Power District

19

l

-_

.___

-

__-___

. - - -

-

.

-

_

_ _ _ _ _

_ _

_ .

.

.;

.

Violation Examples (USAR)

.

Disagree

,

__._.___________

!

'

Rod Insertion Time Delay

l

!

e USAR incorrecily revised in 1993

o CNS identified and corrected in July 1996

,

l

I

i

e Corrected information not incorporated into

'

controlled USARs at time of inspection

!

l

.

!

f

Nebraska Public Pcwer District

20

I

.

-

.

-

__ ________

_ _ _ _ _ _

_ _ _

.I

.

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i

FSAR Lpdating Violations

j

_ __,s_s__ _

!

l

i

l

Causes

!

l

!

-

L

e Discrepancies originating prior to 1980

,

(prior to issuance of 50.71(e))

FSAR not updated or required to be updated

.

When updated in 1983, based on readily available

information regarding plant' changes which may

,

have lacked clarity and/or completeness

!

.

i

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Nebraska Public Power District

21

i

!

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,

t

4

=

i

.

FSAR Updating

_________________.___._

i

!

Causes

t

e Discrepancies originating from 1986 to 1996

resulted from a combination of an evolving USAR

control process, inadequate rigor / oversight

Lack of generic industry / regulatory focus

l

e Although cited examples did not occur between 1983

and 1986, during this period there was no single

update control process and minimal focus on USAR

Nebraska Public Power District

22

-

.]

.

FSAR Updating

i

_________

_

!

Corrective Actions - Specific

'

!

e USAR changes made for admitted violations

i

r

o USAR clarified for denied violations

i

e 50.71(e) interim update

i

e Audited all on-site controlled USARs to

.

t

ensure they contained latest updates

l

Nebraska Public Power District

23

_ _ _ - - _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ , _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ .

_ _ _ . - - - - - _ _ _ _

_ - _ - _ _

_

,

e

i

!

.

FSAR Updating

-

, _ , ,, _ , _ _

i

Corrective Actions - Broad

r

!

I

o USAR Rebaselining Project

i

o Increased frequency of USAR controlled copy

i

updating

i

e Increased focus on PIR initiation with

l

.

identification of USAR inconsistency

i

!

?

l

Nebraska Public Power District

24

!

,

. -

-

- -

-

- - - - -

_ - .

. _ _ _ _ . _ _ _ - -

--

- - - - . _ - -

-

_ . _ _ _ _ _ .. _ ___ _ _ _ . _ . _ _.._ _ . _ _. _ _ . . .

~ ... _

!

.

i

!

USAR DISCREPANCIES PER QUARTER (w/Six Month Rolling Average)

20

20 -

- z.

.

!

l

!

18 1

18

l

17

l

L

16 +

16

14 -

14

'12.5

12

l

i

,

10

!

,

10 -

10

- to

l

8

l

8+

'8

l

7

!

6-

-6

(

,

4

-4

2-

2

-2

-

- - - -+ -

I

-

1

I

i

I

O

0-

i

'

3/94

4/94

ISS

2/95

3SS

4/95

IS6

2S6

3S6

4/96

1/97

lmNUMBER PER QUARTER -+--SIX MONTH ROLLING AVERAGE l

.

,

'

I

t

i

t

I

,

_ _ _ _ _ _ _ _ _ _ _

.

USAR Rebaselining Project

- -

- - _ , - -

_

e Verify accuracy of current USAR

e Update USAR using 10 CFR 50.71(e)

criteria

e Resolve discrepant conditions identified

e Develop Linkage Database for

i

improved 50.59s

Nebraska Public Power District

26

.

-.-

- . - - - . - -

- .

.

-

_ _ _ _ _ _ _ _ -

. .

.

Equipment -

License /USAR

!

Station

_

Engineering

M

SpecialTest

Special

Evaluations

Change

."" ""

Specification

Modifications

Procedures

Instructions

(3.4.3)

Changes

Requests

1r

'1r

1r

l

1r

l

Safety Analysis

Equip. Safety Classification

Temporary Shielding

Procedure

setpoint changes

Change

1r

MWR SpecialInstructions

j

calculadons

(0.4)

50.59 Applicability

No

Add Compensatory

Activity Safe?

>

Measures or Do Not

t

Screen

1r

Perform

,

50.59

Applicability

37

Screen

i

T.S. Change or

'P

Other Codified

Process?

No

'

Within 50.59

I

Scope?

1r

!

Yes

50.59 Scope?

-

'

1r

implement

i

37

1r

'

u

Change

-

I

USQE

USQ Exista

y

Prior NRC Approval

1r

Technical

SORC

Implement

,

>

Change

Review

i

-

- _ _ - - _

.

I

!

10 C.F.R. 50.59

_ - e ,m eme - -

o Pre-1995 program / process marginal

--

implementation was primary problem

j

e Post-1995 program / process improved -- reworked

and clarified in 1996

i

Consistent implementation of program / process

>

remains a concern

t

e 50.59 in a state of flux; NRC proposing new guidance

h

!

l

. Nebraska Public Power District

28

,

-

. - - -

-

_ _ _ _ _ . _ - . _ _ _ _ _ _ _ - . _ _ _ . - _ - _ - - _ _ _ _ - - _ _ _ . . _ _ _ _ - _ _ _ . . - _ _ _ _ . _ . - , _ _ _ - - _ _ - - - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ . _ _ _ _ _ _ _ _ - - . - - _ _

_ _ _ _ _ . _ _ . _ _ _ _

. .

.

.,

t

Facility Changes / Tests Reported under 10 CFR 50.59

l

,

l

350

-

'

l

1

i

'

i

soo

-

,

l

!

.

!

'

!

f

250

!

209

l

t,

200

-

i

I

!

150

-

f

r

s

90_

_

,00

69

70

!

57

54

48

!

38 -

-

!

so _

27

26

'

18

11

12

11

8

4

8

7

6

4

,E

":

1"

+

"+

l

+

"

0

.-

-

r

-

1974

1976

1978

1980

1982

1984

1986

1988

1990

1992

1994

1996-97

h

pending

[

l

lENumberIdentified l

l

!

i

.

!

-

. -

-

- -

-

- -

-

- -

-

- - -

-

--

_. .

.

10 C.F.R. 50.59 Initiatives

-

I

ramm5staesw;crtja9qggmaarmarm

!

e issues still being identified

Training continuing

j

Additional management oversight, review

of 50.59 products

.

Nov.1996 upgrade to 50.59 procedure

j

Efforts to enhance training program

l

!

-

!

,

Nebraska Public Power District

30

!

_

.

t

l

IR 96-31:50.59 Violations

Admit

-

i

MWM y=?%%gy @ '?,*" ""^;yRT*

!

.

e Non-safety related ice deflector

.

No installation, partial installation

!

i

e Fill rate timer and alarm not evaluated as part

of system evaluation for condition being

corrected

j

.

l

Nebraska Public Power District

31

"

,

.

-

- -

.

- - . .

- -

-

- - - - - -

. -

- -

. . -

-

-__ _ ______ _ _ _

._.-

.

_.

.

$

!

IR 96-31:10 C.F.R. 50.59

-

==e a m a m es e

.

!

Causes

e ice deflector

'

failure to proceduralize deflector installation

to prompt 5C.59 and/or USAR activity

incorrect designation of activity as

,

1

maintenance

!

'

e fill rate timer

Inadequate 50.59 evaluation

Ncbraska Public Power District

32

i

j

!

-

- -

..

_ _ _ _ _ _ _ _ _ _ - - _ ..

_ .

.

50.59 Violation - Disagree

. 50.71(e) Violation - Agree

__n--__

-

o Transient combustibles used for a

'

maintenance activity

condition already evaluated in FHA and controlled

j

by approved procedure

j

'

evaluated maintenance activity does not require

50.59 - no change to plant

was evaluated; therefore is not an unevaluated

{

plant change

.

l

"no combustibles" does not accurately reflect FHA

~

and procedure -- is a 50.71(e) violation

!

Nebraska Public Power District

33

l

,

.

.;

.

1

[

.i

10 C.F.R. 50.59

.

Corrective Actions - Specific

e Ice Deflector

l

completed 50.59 - no USQ

proceduralized instal!ation

,

l

coached individual involved

'

updated USAR description

l

e Fill rate timer

l

revised system 50.59 to include timer - no USQ

!

l

coached 50.59 preparer on 50.59 sufficiency

l

e USAR language revised for combustible loading

Nebraska Public Power District

34

l

.

l

i

)

-

- - - -- - -

-

--

-.

_

.-

-

- - _

-

.

10 C.F.R. 50.59

,

mmenesexpptew=gmapanemm

t

!

l

Corrective Actions - Broad

i

!

e USAR rebaselining will compare USAR to

implementing instructions and resolve discrepancies

as applicable

,

i

e Unauthorized Modification Action Plan

i

!

e Continued management oversight and 50.59 training

for CNS staff

4

!

'

Nebraska Public Power District

35

l

- -

-

- -

.

-

--

.l

.

'

Safety Significance

o

USAR 50.59

_ _ _ _ _ _ , _ _ _ _ _ _ _ _ _ _ _ _ _ . _

_ .

e No actual safety consequences

e Minimal potential safety consequences

multiple sources used to make safety and

regulatory decisions

safety evaluations of activity do not rely on USQ

determinations (50.59 is NRC approval issue)

,

fill rate timer is not a safety issue; ice deflector not

needed for safety; transient combustibles

evaluated and controlled

!

t

!

Nebraska Public Power District

36

[

t

- _ _ _ _ _ _ _ _ - _ _

-

.

-

- -

-

_ _ _ _

_ _

-

.

.t

Regulatory Considerations

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _

l

e

Regulatory significance due to potential number of USAR

l

discrepancies but incentives provided to rest of industry also

'

appropriate for NPPD

Enforcement at this time appears to be inconsistent with

l

spirit of Enforcement Policy for FSAR issues

'

> CNS QA first identified USAR accuracy issue in May 1996

i

CNS is performing a comprehensive USAR review, a goal of

l

the revised Policy

e

50.59 deficiencies involve implementation of an improving

program in a time of changing regulator expectations

j

50.59s are regulatory review issues -- not indicative of

i

inadequate safety conscience

i

Nebraska Public Power District

37

l

!

-

. . - -

- -

-

- - -

-

- - - -

. .

-

- - -

- - - .

-

O

.

t

Regulatory Considerations

_ m.mgm,,ggg,g ggg_,

,

o Violations would have been captured by current USAR

rebaselining program

Program findings qualify for discretion per Enforcement

Policy

'

- Voluntary initiative

- Comprehensive corrective actions taken

- Scope of initial findings expanded

'

- Described in writing and publicly available

.

k

(

!

Nebraska Public Power District

38

-

- -

-

- -

-

- - - - -

-

____________

.

Xon-Violation Issue

.

SLCS Tech Spec

--wszw=sz?==~~~

i

!

e issues are:

effect of 3% setpoint tolerance

adequacy of tech spec surveillance requirement

l

e Setpoint tolerance appropriately accounted for in

valve nominal setpoint

j

tech spec valve setpoint minimum is an allowable

value (like other instruments)

valve set so that will not lift below minimum

e No need to change tech spec surveillance

Analysis supports adequacy of tech spec minimum

f

Nebraska Public Power District

39

i

.

-

- -

-

- - --

- -

-

- -

-

.

LER 96-14-01

,=u==-----

e Oct.1996 determined that fuel prep

machine upper stops set too high

o Design change control process

e No safety significance

l

i

[

!

Nebraska Public Power District

40

-

.

e

o

-e===eee-------

Closing Remarks

Phil Graham

Site Vice-President

Nebraska Public Power District

41

ML20141E139

Text

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