United States Nuclear Regulatory Commission - Protecting People and the Environment

EA-98-050 - Robinson 2 (Carolina Power & Light Co.)

March 04, 1998

EA 98-043
EA 98-050

Carolina Power & Light Company
ATTN: Mr. J. S. Keenan
Vice President
H. B. Robinson Steam Electric Plant Unit 2
3581 West Entrance Road
Hartsville, South Carolina 29550

 

SUBJECT: NOTICE OF VIOLATION
(NRC INSPECTION REPORT NO. 50-261/98-03)

Dear Mr. Keenan:

This refers to the inspection conducted on January 5-9, 1998, at the Robinson Steam Electric Plant. The purpose of the inspection was to follow up on the inspection findings resulting from the Nuclear Regulatory Commission's (NRC) Design Inspection which was conducted between April 7 and May 23, 1997. The results of the inspection were discussed with your staff in an exit interview on January 30, 1998 and were formally transmitted to you by letter dated February 6, 1998. An open predecisional enforcement conference was conducted in the Region II office on February 19, 1998, to discuss the apparent violations, the root causes, and your corrective actions to preclude recurrence. A list of conference attendees and copies of the NRC's presentation material, and materials you presented at the conference are enclosed.

Based on the information developed during the inspection and the information you provided during the conference, the NRC has determined that violations of NRC requirements occurred. The violations are cited in the enclosed Notice of Violation (Notice), and the circumstances surrounding them are described in detail in the subject inspection report.

Violation A involves the failure to verify the adequacy of a design change to the safety injection (SI) system. The design change, Modification M-951, which was implemented in March 1988, disabled the automatic start feature for one of the three SI pumps. The calculation that supported the design change did not include verification that SI pumps B and C had sufficient net positive suction head (NPSH) in this configuration for an event involving a large break loss of coolant accident (LOCA) and loss of one of the two remaining SI pumps due to a single failure. As a result, the SI system design failed to meet the operability requirements of Technical Specification (TS) 3.3.1.1.c (the TS in effect in June 1997 when the violation was discovered).

The NPSH calculation for Modification M-951 included an incorrect assumption that the piping configuration for SI pump A represented the most limiting condition with regard to NPSH. As a result, the NPSH available for pumps B and C under the conditions established by the 1988 modification was not reviewed at the time of the modification nor when NPSH issues were raised in subsequent reviews. The potential consequences of the design deficiencies were limited, during the time of occurrence, due to the high, actual availability and reliability of the A SI pump and the limited LOCA break size and pump configurations necessary to create an inadequate NPSH condition. However, given the regulatory significance and duration of this problem and the potential impact on plant safety, Violation A has been categorized in accordance with the "General Statement of Policy and Procedures for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600, at Severity Level III.

In accordance with the Enforcement Policy, a base civil penalty in the amount of $55,000 is considered for a Severity Level III violation. Because your facility has been the subject of escalated enforcement actions within the last two years (1), the NRC considered whether credit was warranted for Identification and Corrective Action in accordance with the civil penalty assessment process in Section VI.B.2 of the Enforcement Policy. With regard to the factor of Identification, the NRC considered the following: (1) you had prior opportunities to identify the violation including an April 1988 LOCA reanalysis, a March 1989 safety system functional inspection, a June 1989 review based on NRC questions on SI pump runout concerns and SI flow tests in September and November 1990; and (2) NRC Information Notice (IN) 96-55, dated October 22, 1996, alerted licensees to problems with assuring NPSH for similar system configurations. However, you performed an extensive evaluation to verify NPSH for the SI pumps after receiving requests for calculations related to the SI system from NRC inspectors who were preparing for the Design Inspection. As a result of the new calculation, you identified this subtle design error. Absent your decision to conduct a reanalysis, the issue may not have been identified. Therefore, on balance, the NRC determined that credit should be granted for the factor of Identification.

Your corrective actions included: (1) modifications to the refueling water storage tank to increase the tank water level; (2) development of an emergency core cooling system hydraulic model for the SI and residual heat removal systems; and (3) staff training on appropriate methods to ensure the adequacy of calculations. In addition, you intend to modify the suction piping for the B and C SI system piping and conduct a design verification inspection of the component cooling water system in 1998. Based on these facts, the NRC determined that credit was warranted for the factor of Corrective Action.

Therefore, to encourage prompt identification and comprehensive correction of violations, I have been authorized, not to propose a civil penalty in this case. However, significant violations in the future could result in a civil penalty.

NRC Inspection Report No. 50-261/98-03, dated February 6, 1998, included an apparent violation for a number of calculation deficiencies identified during the NRC Design Inspection. The deficiencies included: (1) failures to include appropriate design parameters or utilize appropriate design inputs in calculations; (2) failure to translate design requirements into other design documents such as drawings or procedures; and, (3) failures to perform adequate design verification. After review of the information provided at the predecisional enforcement conference, the NRC concludes that the deficiencies do not appear to have affected design assumptions to such an extent that safety-related equipment was inoperable. In addition, although the NRC believes that the deficiencies indicate weaknesses in your control of design-related calculations, the NRC has determined that the violations do not constitute a breakdown in your design control program. Therefore, three violations, each categorized at Severity Level IV, are cited in the enclosed Notice for these design deficiencies.

You are required to respond to this letter and should follow the instructions specified in the enclosed Notice when preparing your response. In your response, you should document the specific actions taken and any additional actions you plan to prevent recurrence. After reviewing your response to this Notice, including your proposed corrective actions and the results of future inspections, the NRC will determine whether further NRC enforcement action is necessary to ensure compliance with NRC regulatory requirements.

In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter, its enclosures, and any response will be placed in the NRC Public Document Room (PDR).

 

Sincerely,



Luis A. Reyes
Regional Administrator

Docket No. 50-261
License No. DPR-23

Enclosures: 1. Notice of Violation
2. List of Attendees
3. NRC Slides
4. Licensee Material

cc w/encls:
Dale E. Young
Director, Site Operations
Carolina Power & Light Company
H. B. Robinson Steam Electric Plant
3581 West Entrance Road
Hartsville, SC 29550

J. W. Moyer
Plant General Manager
Carolina Power & Light Company
H. B. Robinson Steam Electric Plant
3581 West Entrance Road
Hartsville, SC 29550

D. B. Alexander, Manager
Performance Evaluation and
Regulatory Affairs OHS7
Carolina Power & Light Company
412 S. Wilmington Street
Raleigh, NC 27601

H. K. Chernoff, Supervisor
Licensing/Regulatory Programs
Carolina Power & Light Company
H. B. Robinson Steam Electric Plant
3581 West Entrance Road
Hartsville, SC 29550

T. M. Wilkerson, Manager
Regulatory Affairs
Carolina Power & Light Company
H. B. Robinson Steam Electric Plant
3581 W. Entrance Road
Hartsville, SC 29550

Max Batavia, Chief
Bureau of Radiological Health
Dept. of Health and Environmental Control
2600 Bull Street
Columbia, SC 29201

Mel Fry, Acting Director
Division of Radiation Protection
N. C. Department of Environment,
Health and Natural Resources
3825 Barrett Drive
Raleigh, NC 27609-7721

William D. Johnson
Vice President & Senior Counsel
Carolina Power & Light Company
P. O. Box 1551
Raleigh, NC 27602

Karen E. Long
Assistant Attorney General
State of North Carolina
P. O. Box 629
Raleigh, NC 27602

Robert P. Gruber
Executive Director
Public Staff - NCUC
P. O. Box 29520
Raleigh, NC 27626-0520

Public Service Commission
State of South Carolina
P. O. Box 11649
Columbia, SC 29211

Hartsville Memorial Library
147 W. College Avenue
Hartsville, SC 29550


NOTICE OF VIOLATION

Carolina Power & Light Company
H. B. Robinson Steam Electric Plant
Unit 2
Docket No. 50-261
License No. DPR-23
EAs 98-043 and 98-050

During an NRC inspection conducted on January 5-9, 1998, violations of NRC requirements were identified. In accordance with the "General Statement of Policy and Procedures for NRC Enforcement Actions," NUREG-1600, the violations are listed below:

A. 10 CFR 50, Appendix B, Criterion III, in part, requires that "design control measures shall provide for verifying the adequacy of design, such as by the performance of design review, by use of alternate or simplified calculational methods, or by performance of a suitable testing program." Design changes shall be subject to design control measures commensurate with those applied to the original design."

Technical Specification 3.3.1.1.c requires two safety injection (SI) pumps to be operable.

Section 3.4.3 of CP&L Corporate Quality Assurance Manual, Revisions 11 through 18, dated January 29, 1988 through September 29, 1995, states that "sufficient design verification shall be performed by one or more methods to substantiate that final design documents meet the appropriate design inputs." It further states that a design verification should confirm that "the design is technically adequate with respect to the design basis."

Contrary to the above, between March 24, 1988 and June 27, 1997, the licensee failed to verify the adequacy of design to substantiate that final design documents met the appropriate design inputs and were technically adequate for a design change affecting SI pumps B and C. Specifically, the licensee implemented Modification M-951, which disabled the automatic start feature for one of the three SI pumps but failed to verify that SI pumps B and C had sufficient net positive suction head in the event that a large break loss of coolant accident occurred and one of the two remaining SI pumps failed to operate due to a single failure. As a result, the SI system failed to meet the operability requirements of Technical Specification 3.3.1.1.c, the TS in effect in June 1997 when the violation was discovered. (01013)

This is a Severity Level III violation (Supplement I).

B. 10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that "design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design review, by use of alternate or simplified calculational methods, or by performance of a suitable testing program."

Section 3.4.3 of CP&L Corporate Quality Assurance Manual, Revisions 14 through 18, dated August 1, 1990, through September 29, 1995, states that "sufficient design verification shall be performed by one or more methods to substantiate that final design documents meet the appropriate design inputs." It further states that a design verification should confirm that "the design is technically adequate with respect to the design basis."

Contrary to the above, as of April 7, 1997, the licensee failed to verify the adequacy of design in certain calculations. Specifically, the licensee failed to consider appropriate design parameters or failed to utilize appropriate design inputs to ensure the design was technically adequate in the calculations listed below:

1. Calculation number RNP-I/INST-1023, "Refueling Water Storage Tank Level Indication Accuracies", Revision 0, dated June 28, 1991, did not consider potential vortexing in the Refueling Water Storage Tank above the drain nozzle.

2. Calculation number RNP-I/INST 1109, "Containment EOP Setpoint Parameters", Revision 0, dated November 29, 1994, did not determine the correct containment water level required for post accident residual heat removal (RHR) pump recirculation operation (EOP setpoint No. 20).

3. Calculation number RNP-I/INST-1058, "Containment Water Level Instrument Uncertainty", Revision 0, April 4, 1994, used an incorrect value for the containment water level above the containment floor.

4. Calculation number RNP-I/INST-1040, "Main Steam Flow Accuracy and Scaling Calculation", Revision 0, dated May 16, 1994, and RNP-I/INST-1043, "Main Steam Pressure Channel Accuracy and Scaling Calculation", Revision 1, dated April 15, 1994, did not include seismic uncertainty factors specified in Section 10 of Design Guide DG-VIII.50, Engineering Instrument Setpoints.

5. Calculation number RNP-M/MECH-1620, "Evaluation of Effects of High Energy Pipe Rupture on the CCWS", Revision 0, dated July 18, 1996, excluded the design inputs for high energy line breaks in Reactor Coolant System piping and their jet impingement effect on adjacent component cooling water (CCW) piping and supports.

6. Calculation number RNP-M/MECH-1362, "SW Screen Wash Piping Flow Analysis", Revision 1, dated September 5, 1991, did not include rupture of the non-seismic piping that supply the instrument and station air compressors.

7. Calculation number RNP-E-6.020,"Load Profile and Battery Sizing Calculation for Battery B", Revision 2, dated November 24, 1993, incorrectly referenced a time period of "2 minutes to 59 minutes", instead of "1 minute to 59 minutes. The calculation did not consider some of the connected non-safety related loads and referenced an incorrect battery cell type (MCT instead of MCX) in Attachment U to the calculation.

8. Calculation number RNP-E-6.23, "Minimum Inverter Voltage Verification", Revision 2, dated December 1, 1993, did not consider the increased inverter current at reduced battery voltage.

9. Calculation number RNP-E-6.004, "DC Short Circuit Study", Revision 2, dated May 19, 1993, did not consider a small DC motor that was connected to the system. The battery open circuit voltage used in the calculation was less than the voltage measured during testing. This calculation along with RNP-E-018, "Ampacity Evaluation of Safety Related 125VDC and 120V AC Power Cables", Revision 4, dated March 16, 1994, analyzed cables rated at 75C, whereas cables rated at 60C were installed.

10. Calculation number RNP-E-6.018, "DC Control Circuit Loop Analysis", Revision 0, dated April 19, 1994, used incorrect solenoid valve power values for design input.

11. Calculation number RNP-E-8.016, "Emergency Diesel Generator Static and Dynamic Analysis", Revision 5, dated September 19, 1994, used an incorrect reference and only modeled SI pump motor B.

12. Calculation number RNP-M/MECH-1460, "NPSH vs. CST Level for SDAFW Pump", Revision 0, dated June 19, 1992, a value for the condensate storage tank (CST) water temperature of 100F was used, instead of the 115F temperature value listed in the Plant Parameter Document for Cycle 18.

13. Calculation number RNP-M/MECH-1394, "AFW Pump Recirculation Flowrates for RNP-2", Revision 2, dated August 21, 1995, used an incorrect specific gravity for the CST water.

14. Discrepancies were identified in calculation numbers RNP-I/INST-1015, Revision 0, dated December 22, 1990, and 84065-M-06-F, Revision 3, dated January 14, 1991, for the condensate storage tank level at which to change the auxiliary feedwater (AFW) pump suction supply to the service water system. Calculation RNP-I/INST-1015 shows a 10 percent level, whereas calculation 84065-M-06-F shows a 15 percent level. (02014)

This is a Severity Level IV Violation (Supplement I).

C. 10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that "design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design review, by use of alternate or simplified calculational methods, or by performance of a suitable testing program."

Section 3.4.2 of CP&L Corporate Quality Assurance Manual, Revisions 12 through 18, dated June 1, 1989, through September 29, 1995, states that "Applicable design input requirements shall be developed and documented. The design inputs shall be specified to a level of detail sufficient to allow translation into other design documents such as specifications, drawings, analyses, procedures, etc."

Contrary to the above, as of April 7, 1997, the licensee failed to verify the adequacy of design in that design inputs were not correctly translated into other design documents such as drawings or procedures for the examples listed below:

1. A design change was implemented in 1990 to provide for isolation of the RHR pumps by closure of valve numbers SW-906, SW-907, CC-927, and CC-928 as discussed in LER 89-008-01. The licensee failed to incorporate the effects of this design change in the ASME Section XI inservice testing (IST) program. These valves were incorrectly classified as passive valves in the IST program when in fact they should have been classified as active valves as a result of the design change.

2. The design basis for CCW thermal relief valve numbers CC-747 A and B, CC-774, and CC-791G was incorrectly translated into the installation drawings. Consequently, the valves were installed in locations which resulted in the 10 psig back pressure values specified in Westinghouse E-spec No. 676257 being exceeded by 5 psig.

3. The design basis for performance of testing on station batteries (IEEE Standard 450-1980, Recommended Practice for Maintenance, Testing, and Replacement of Large Storage Batteries for Generating Stations and Substations) was incorrectly translated into MST-920, Station Battery Performance Capacity Test, Revision 6, dated September 28, 1995, and MST-921, Station Battery Service Test, Revision 7, dated April 20, 1995. Step 7.5.10 of procedure MST-921 accepted voltages less than the minimum value of 1.0 volt DC specified in IEEE Standard 450-1980. The duration of capacity testing of station battery B specified in MST-920 was different from that used by the battery manufacturer. The minimum acceptance criteria of 107 volts specified in MST-921 for the station battery load profile test was less than the value of 109.8 volts evaluated in Calculation RNP-E-6.018.

4. The design basis for performance of maintenance on station batteries was incorrectly translated into procedures PM-410, Installation of Battery Bank and Cell Connections, Revision 6, dated November 2, 1995, and PM-411, Disassembly, Cleaning, Assembly, and Testing of A and B Station Battery Cell Connections, Revision 6, dated October 6, 1995. The procedures stated the acceptance criteria was 50 microohms, whereas the vendor calculations specified that the B station battery may not exceed 50 microohms and the A station battery may not exceed 34 microohms.  The requirements for installation (torque) of intercell connections and mechanical cable connections and the thickness of the intercell connectors specified in PM-411 conflicted with requirements specified in vendor technical manuals. (03014)

This is a Severity Level IV Violation (Supplement I).

D. 10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that "design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design review, by use of alternate or simplified calculational methods, or by performance of a suitable testing program."

Section 3.4.3.9 of CP&L Corporate Quality Assurance Manual, Revision 18, dated September 29, 1995, states that "A design verification shall be performed to verify an appropriate design verification has been performed for applicable documents contained in the package."

Section 3.4.5 of CP&L Corporate Quality Assurance Manual, Revisions 12 through 16, dated June 1, 1989, through December 17, 1992, states that "design change documents shall provide for identification of necessary revisions to existing design documents."

Contrary to the above, the licensee failed to verify the adequacy of design in that:

1. A calculation in ESR 96-00474, Revision 0, dated August 19, 1996, which evaluated whether failure of a non-seismic pipe would affect water supply to the SI pumps, was not design verified.

2. Calculation numbers 789M-M-02, Revision 0, dated December 15, 1989, 789M-M-05, Revision 0, dated December 18, 1989, and RNP-E-6.002, Revision 0, dated December 1, 1987, were not identified as voided or superseded when replaced by other design calculations. (04014)

This is a Severity Level IV Violation (Supplement I).

Pursuant to the provisions of 10 CFR 2.201, Carolina Power and Light Company is hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555, with a copy to the Regional Administrator, Region II, and a copy to the NRC Resident Inspector at the Robinson facility, within 30 days of the date of the letter transmitting this Notice of Violation (Notice). This reply should be clearly marked as a "Reply to a Notice of Violation" and should include for each violation: (1) the reason for the violation, or, if contested, the basis for disputing the violation or severity level, (2) the corrective steps that have been taken and the results achieved, (3) the corrective steps that will be taken to avoid further violations, and (4) the date when full compliance will be achieved. Your response may reference or include previously docketed correspondence, if the correspondence adequately addresses the required response. If an adequate reply is not received within the time specified in this Notice, an order or a Demand for Information may be issued as to why the license should not be modified, suspended, or revoked, or why such other action as may be proper should not be taken. Where good cause is shown, consideration will be given to extending the response time.

If you contest this enforcement action, you should also provide a copy of your response to the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001.

Under the authority of Section 182 of the Act, 42 U.S.C. 2232, this response shall be submitted under oath or affirmation.

Because your response will be placed in the NRC Public Document Room (PDR), to the extent possible, it should not include any personal privacy, proprietary, or safeguards information so that it can be placed in the PDR without redaction. If personal privacy or proprietary information is necessary to provide an acceptable response, then please provide a bracketed copy of your response that identifies the information that should be protected and a redacted copy of your response that deletes such information. If you request withholding of such material, you must specifically identify the portions of your response that you seek to have withheld and provide in detail the bases for your claim of withholding (e.g., explain why the disclosure of information will create an unwarranted invasion of personal privacy or provide the information required by 10 CFR 2.790(b) to support a request for withholding confidential commercial or financial information). If safeguards information is necessary to provide an acceptable response, please provide the level of protection described in 10 CFR 73.21.

Dated at Atlanta, Georgia
this 4th day of March 1998


1. A Severity Level III violation was issued on May 16, 1996 for failure to control safeguards information properly (EA 96-120). A Severity Level III problem with a $55,000 civil penalty was issued on December 12, 1997 for violations related to a mispositioned EDG output breaker control switch (EA 97-490).

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