EA-96-442 - Brunswick 2 (Carolina Power and Light Company)

December 13, 1996

EA 96-442

Carolina Power & Light Company
ATTN: Mr. W. R. Campbell
Vice President
Brunswick Steam Electric Plant
Post Office Box 10429
Southport, North Carolina 28461

SUBJECT:

NOTICE OF VIOLATION (NRC INSPECTION REPORT NOS. 50-325/96-16 AND 50-324/96-16)

Dear Mr. Campbell:

This refers to the integrated inspection completed on October 26, 1996 at your Brunswick facility. The inspection included a review of your failure to provide temperature compensation for the Plant Process Computer (PPC) feedwater flow algorithm which resulted in operation of Brunswick Unit 2 in excess of (1) the maximum thermal power authorized by the license and (2) thermal limits required by Technical Specification (TS) 3.2.1. The inspection report was sent to you by letter dated November 22, 1996. A closed, predecisional enforcement conference was conducted in the Region II office on December 9, 1996, with you and members of your staff to discuss the apparent violations, the root causes, and your corrective actions to preclude recurrence. A list of conference attendees, NRC slides, and a copy of your presentation materials 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 maintain the Unit 2 thermal power within the operating license limits. On August 28, 1996, a reactor engineer reviewing core thermal power calculations associated with the Power Uprate Project determined that the Unit 2 PPC point value for feedwater flow was not properly compensated for feedwater temperature deviations from the normal operating temperature. The condition had existed on Unit 2 since the unit restarted after a refueling outage ending in July 1994 when Unit 2 was modified to add a new PPC and associated software. Due to the failure to temperature compensate the feedwater flow process point value, the core thermal power calculated and indicated by the PPC was less than the actual core thermal power. During periods when feedwater temperature was lower than the normal operating value, Unit 2 was operated at indicated power levels of up to 100 percent power or 2436 megawatts (MW) thermal which was equivalent to actual power levels of up to 102.4 percent power or 2494 MW thermal. This is a violation of License Condition 2.C.1 of Facility Operating License Number DPR-62 which requires that Unit 2 be operated at or less than 2436 MW thermal.

Violation B involves the failure to maintain the calculated Average Planar Linear Heat Generation Rate (APLHGR) within the limits of TS 3.2.1. The APLHGR limits vary based on power level and feedwater flow to assure that the fuel thermal-mechanical design criteria are preserved during abnormal transients. TS 3.2.1 specifies the approved methodology for determining the limits placed on ALPHGR for a given power level and feedwater flow rate. Due to the failure to appropriately compensate for feedwater temperature, the calculated reactor power level inputs to the APLHGR calculation were incorrect and the resulting APLHGR value was non-conservative. The APLHGR values, as calculated using the actual power levels, exceeded the limits specified by TS 3.2.1 between December 10 and December 20, 1995. During the predecisional enforcement conference, your staff noted that the approved methodology for calculating the APLHGR limits specified by TS 3.2.1 was based on generic APLHGR adjustment factors. Your re-analysis of the APLHGR limits using cycle specific adjustment factors indicated that the APLHGR values, based on actual power levels between December 10 and December 20, 1995, were within cycle specific design limits. Although the re-analysis indicates that the actual safety consequence of Violation B was low; the NRC considers any change in reactor parameters that cause unanticipated reductions in the margin of safety to be a significant regulatory concern.

The root causes of the violations included the failure of your design team to properly link the Unit 2 feedwater flow process points to the appropriate compensation formula in the Unit 2 compensation database. The computer index labels for the Unit 2 feedwater flow process points were changed when additional process points were loaded into the database. Due to the inappropriate index labels, the computer linked a compensation value of one to the points instead of the correct compensation value. Your post-modification acceptance testing for the new PPC did not verify that process point numbering was the same in both units and did not verify that the correct relationships between process points and compensation values were preserved when your design team, in an effort to reduce differences between the two units, copied the existing PPC database configuration from the Unit 1 PPC to the Unit 2 PPC.

These violations represent a significant failure to control design parameters that affected the integrity of reactor core protection systems. The NRC expects licensees to provide meticulous oversight of vendor changes to plant process computer software and to conduct comprehensive post-modification testing of new software used to assure operation within specified acceptable fuel design limits. In this case, the NRC is particularly concerned that specified core operating limits were exceeded due to the inadequate design control and testing. Therefore, these violations are classified in the aggregate in accordance with the "General Statement of Policy and Procedures for NRC Enforcement Actions" (Enforcement Policy) NUREG-1600, as a Severity Level III problem.

In accordance with the Enforcement Policy, a base civil penalty in the amount of $50,000 is considered for a Severity Level III problem. 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 described in Section VI.B.2 of the Enforcement Policy. The NRC concluded that credit was warranted for Identification because your staff identified both violations. With regard to consideration for Corrective Action, your corrective actions included: (1) effective, immediate corrective actions to reduce power and correct the software deficiency; (2) revalidation of critical plant process computer functions and confirmation of appropriate testing of the software; (3) review of other computer applications; (4) enhancements to the control of design and testing of computer products; and, (5) training on lessons learned and software configuration control. Based on the above, the NRC determined that credit was warranted for Corrective Action.

The application of the factors considered in the civil penalty assessment process, absent the exercise of discretion, resulted in no civil penalty. However, you should be aware that the NRC considered imposing a civil penalty, under Section VII.A of the Enforcement Policy, because of the potential impact of weak vendor oversight and inadequate design control and testing of software affecting core operating parameters. However, because your reactor engineer demonstrated a safety conscious attitude which resulted in the identification of the software deficiency and to encourage prompt identification and comprehensive correction of violations, I have been authorized, after consultation with the Office of Enforcement, not to propose a civil penalty in this case. Significant violations in this area in the future could result in a civil penalty.

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 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 your response will be placed in the NRC Public Document Room (PDR).

		Sincerely,
		Original Signed by Luis Reyes for
		Stewart D. Ebneter Regional Administrator

Docket Nos. 50-324
License Nos. DPR-62

Enclosures:
Notice of Violation

cc w/encl:
W. Levis, Director
Site Operations
Brunswick Steam Electric Plant
P. O. Box 10429
Southport, NC 28461

R. P. Lopriore
Plant Manager
Brunswick Steam Electric Plant
Carolina Power & Light Company
P. O. Box 10429
Southport, NC 28461

J. Cowan, Vice President
Operations & Environmental
Support MS OHS7
Carolina Power & Light Company
P. O. Box 1551
Raleigh, NC 27602

Gerald D. Hicks
Manager - Regulatory Affairs
Carolina Power & Light Company
P. O. Box 10429
Southport, NC 28461

W. D. Johnson, Vice President
and Senior Counsel
Carolina Power & Light Company
P. O. Box 1551
Raleigh, NC 27602
Dayne H. Brown, Director
Division of Radiation Protection
N. C. Department of Environmental
Health & Natural Resources
P. O. Box 27687
Raleigh, NC 27611-7687

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

Jerry W. Jones, Chairman
Brunswick County Board of
Commissioners
P. O. Box 249
Bolvia, NC 28422

Dan E. Summers
Emergency Management Coordinator
New Hanover County Department of
Emergency Management
P. O. Box 1525
Wilmington, NC 28402

William H. Crowe, Mayor
City of Southport
201 East Moore Street
Southport, NC 28461

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NOTICE OF VIOLATION

Carolina Power and Light Company		Docket Nos. 50-324
Brunswick Steam Electric Plant		License Nos. DPR-62
Unit 2		EA 96-442

During an NRC inspection completed on October 26, 1996, 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. Facility Operating License Number DPR-62, Section 2,C,1, Maximum Power Level, authorizes the licensee to operate the facility at steady state reactor core power levels not in excess of 2436 megawatts (Mw) (thermal).

Contrary to the above, during the time periods listed below, the licensee failed to operate the facility within steady state reactor core power level limit of 2436 (Mw) (thermal):

Dates	Power Level (MW)	Percent Power
July 5, 1994 thru September 6, 1995	2446Mw	100.4% Power
(including February 26, 1995 at	2460 Mw	101.0% Power)
March 26 thru August 28, 1996	2441 Mw	100.2% Power
(including April 17 thru 26, 1996 at	2492 Mw	102.3% Power
and July 19 thru 26, 1996 at	2494 Mw	102.4% Power)(01013)

B. Technical Specification 3.2.1 requires in part, that during power operation, the AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR), for each type of fuel as a function of axial location and AVERAGE PLANAR EXPOSURE shall not exceed limits based on applicable APLHGR limit values that have been approved for the respective fuel and lattice type and determined by the approved methodology described in GESTAR-II.

Contrary to the above, between December 10 and December 20, 1995, during power operation, the licensee failed to maintain the APLHGR within the applicable approved APLHGR limit values specified in Technical Specification 3.2.1. (01023)

These violations represent a Severity Level III problem (Supplement I).

Pursuant to the provisions of 10 CFR 2.201, Carolina Power & 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 Brunswick Steam Electric Plant, 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, (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 previous 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.

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. However, if you find it necessary to include such information, you should clearly indicate the specific information that you desire not to be placed in the PDR, and provide the legal basis to support your request for withholding the information from the public.

Dated at Atlanta, Georgia
this 13th day of December 1996

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¹A Severity Level III violation was issued on November 19, 1996, (EA 96-354) related to environmental qualification program deficiencies. A Severity Level III violation was issued on July 12, 1996. (EA 96-181) related to design control measures for service water system modifications. A Severity Level III violation was issued on April 4, 1996, (EA 96-054) for failure to meet fitness-for-duty requirements. A Severity Level III violation was issued on November 20, 1995, (EA 95-228) related to suitability of materials used in valves in the residual heat removal system. A Severity Level III problem was issued on September 8, 1995, (EA 95-166) related to design control, modification and testing of the high pressure injection system and reactor core isolation cooling system.