IR 05000346/2004010

IR 05000346-04-010(DRS) on 6/28/2004 - 7/2/2004 for Davis-Besse Nuclear Power Station; Evaluations of Changes, Experiments or Tests, and Permanent Plant Modifications

ML042160297
Person / Time
Site:	Davis Besse
Issue date:	08/03/2004
From:	Grobe J NRC/RGN-III
To:	Bezilla M FirstEnergy Nuclear Operating Co
References
IR-04-010
Download: ML042160297 (23)

Text

ust 3, 2004

SUBJECT:

DAVIS-BESSE NUCLEAR POWER STATION NRC EVALUATIONS OF CHANGES, EXPERIMENTS, OR TESTS AND PERMANENT PLANT MODIFICATIONS INSPECTION REPORT 05000346/2004010(DRS)

Dear Mr. Bezilla:

On July 2, 2004, the U.S. Nuclear Regulatory Commission (NRC) completed a routine baseline inspection at your Davis-Besse Nuclear Power Station. The enclosed report documents the inspection findings, which were discussed on July 2, 2004 and on July 23, 2004, with you and other members of your staff.

The inspection examined activities conducted under your license as they relate to safety and to compliance with the Commissions rules and regulations and with the conditions of your license.

Specifically, this inspection focused on the baseline biennial inspections for evaluations of changes, tests, or experiments (10 CFR 50.59) and permanent plant modifications. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel. Because of the large number of plant modifications and change evaluations performed during the recent Davis-Besse shutdown, the team reviewed a larger than normal sample size of plant changes and modifications.

Based on this inspection, the team identified two Severity Level IV violations of NRC requirements associated with the failure to perform an adequate safety evaluation review as required by 10 CFR 50.59. Because the violations were non-willful and non-repetitive and because they have been entered into your corrective action program, the NRC is treating the issues as Non-Cited Violations in accordance with Section VI.A.1 of the NRC's Enforcement Policy.

If you contest these Non-Cited Violations, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission - Region III, 2443 Warrenville Road, Lisle, IL 60532-4352; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector Office at the Davis-Besse Nuclear Power Station. In accordance with 10 CFR 2.390 of the NRCs "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publically Available Records (PARS) component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

John A. Grobe, Chairman Davis-Besse Oversight Panel Docket Nos. 50-346 License Nos. NPF-3 Enclosure: Inspection Report 0500346/2004010(DRS)

w/Attachment: Supplemental Information cc w/encl: The Honorable Dennis Kucinich G. Leidich, President - FENOC J. Hagan, Senior Vice President Engineering and Services, FENOC L. Myers, Chief Operating Officer, FENOC Plant Manager Manager - Regulatory Affairs M. OReilly, Attorney, FirstEnergy Ohio State Liaison Officer R. Owen, Administrator, Ohio Department of Health Public Utilities Commission of Ohio President, Board of County Commissioners of Lucas County C. Koebel, President, Ottawa County Board of Commissioners D. Lochbaum, Union Of Concerned Scientists J. Riccio, Greenpeace P. Gunter, N.I.R.S.

SUMMARY OF FINDINGS

IR 05000346/2004010(DRS); 6/28/2004 - 7/2/2004; Davis-Besse Nuclear Power Station;

Evaluations of Changes, Experiments or Tests, and Permanent Plant Modifications.

This report covers a five day period of announced baseline inspection on evaluations of changes, tests, or experiments and permanent plant modifications. The inspection was conducted by Region III inspectors. Two Green Severity Level IV Non-Cited Violations were identified. The significance of most findings is indicated by their color (Green, White, Yellow,

Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process,

Revision 3, dated July 2000.

A. Inspector-Identified and Self-Revealed Findings

Cornerstone: Initiating Events

No findings of significance were identified.

Cornerstone: Mitigating Systems

Green.

The inspectors identified a Severity Level IV Non-Cited Violation associated with the failure to perform an adequate safety evaluation review as required by 10 CFR 50.59 for changes made to the facility as described in the Updated Safety Analysis Report. Specifically, the licensee failed to perform a safety evaluation in accordance with 10 CFR 50.59 for changes made to Section 9.2.7.3.c of the Updated Safety Analysis Report concerning the low-low pressure interlock for the auxiliary feedwater pumps. The changes made by the licensee adversely affected an Updated Safety Analysis Report-described function in that a previously described automatic feature of the steam inlet valve to the auxiliary feedwater pump was changed to clarify that this automatic feature was not available under certain conditions.

Because the Significance Determination Process is not designed to assess the significance of violations that potentially impact or impede the regulatory process, this issue was dispositioned using the traditional enforcement process in accordance with Section IV of the NRC Enforcement Policy. However, the results of the violation, that is, the failure to evaluate the changes made to Section 9.2.7.3.c of the USAR, were assessed using the Significance Determination Process.

This finding was determined to be more than minor because the inspectors could not determine reasonably that the change would not ultimately require NRC approval. The inspectors determined that this issue was of very low safety significance, because the design basis safety-related function of the auxiliary feedwater system, to remove reactor decay heat following a loss of normal feedwater, was not adversely affected, and because the team determined from the mitigating systems evaluation in the Phase 1 Screening Worksheet that all the questions were answered No. Therefore, the results of the violation were determined to be of very low safety significance and the violation was classified as a Severity Level IV Violation. (Section 1R02)

Green.

The inspectors identified a Severity Level IV Non-Cited Violation of 10 CFR 50.59, Changes, Tests, and Experiments, based on the licensee performing an inadequate evaluation of a proposed change to the plant, regarding tornado missile protection of the diesel generator exhaust stacks and plant doors. Specifically, the licensee's response to the question posed in 10 CFR 50.59(c)(2)(vi) did not demonstrate that the proposed change did not create the possibility of a malfunction of equipment important to safety with a different result than any previously evaluated in the Final Safety Analysis Report (as updated).

Because the Significance Determination Process is not designed to assess the significance of violations that potentially impact or impede the regulatory process, this issue was dispositioned using the traditional enforcement process in accordance with Section IV of the NRC Enforcement Policy. However, the results of the violation, that is, the failure to demonstrate that the proposed change did not create the possibility of a malfunction of equipment important to safety with a different result, were assessed using the Significance Determination Process.

This finding was determined to be more than minor because the inspectors could not determine reasonably that the change would not ultimately require NRC approval. The finding was determined to be of very low safety significance based on a significance determination process analysis of a loss of offsite power concurrent with loss of one emergency diesel generator and the violation was classified as a Severity Level IV Violation. (Section 1R02)

Cornerstone: Barrier Integrity

No findings of significance were identified.

Licensee-Identified Violations

No findings of significance were identified.

REPORT DETAILS

Summary of Plant Status

Davis-Besse operated at or near full power throughout the inspection period.

1R02 Evaluations of Changes, Tests, or Experiments

.1 Review of 50.59 Evaluations and Screenings

a. Inspection Scope

From June 28, 2004 through July 2, 2004, the inspectors reviewed eight evaluations performed pursuant to 10 CFR 50.59 requirements. The evaluations related to permanent plant modifications, setpoint changes, procedure changes, conditions adverse to quality, and changes to the Updated Safety Analysis Report (USAR). The inspectors reviewed the evaluations for thoroughness and to determine if prior NRC approval was obtained as appropriate. The inspectors also reviewed 16 screenings where the licensee had determined that a 10 CFR 50.59 evaluation was not necessary.

In regard to the changes reviewed where no 10 CFR 50.59 evaluation was performed, the inspectors reviewed the changes to determine if they met the threshold to require a 10 CFR 50.59 evaluation. These evaluations and screenings were chosen based upon a consideration of the risk significance of samples from the different cornerstones. The list of documents reviewed by the inspectors is included as an attachment to this report.

b. Findings

b.1 Inadequate Evaluation of USAR Change for Steam Inlet Valve to Auxiliary Feedwater Pumps

Introduction:

The inspectors identified that the licensee failed to perform an adequate safety evaluation in accordance with 10 CFR 50.59 before making changes to the USAR associated with the low-low pressure interlock for the steam inlet valve to the auxiliary feedwater (AFW) pumps. The issue was considered to be of very low safety significance (Green) and was dispositioned as a Severity Level IV Non-Cited Violation (NCV).

Description:

On April 24, 2004, the licensee completed USAR Change Notice (UCN)04-021, Clarification to the Operation of the Interlocks Associated with the Auxiliary Feedwater Pump Turbine Steam Supply Valves. UCN 04-021 changed statements in the USAR that described the operation of the low-low pressure interlock for the AFW Pumps while a SFRCS signal is also present. Specifically, the original USAR wording in Section 9.2.7.3.c stated, If suction pressure remains low (1psig for 60 seconds) the steam supply valves will close to protect the AFP [Auxiliary Feedwater Pump] from cavitation and the SFRCS signal to the valves will be locked out to prevent valve motor damage. The steam supply valves to the AFPT [Auxiliary Feedwater Pump Turbine] will be opened automatically when a suction pressure is re-established and an SFRCS signal remains present. This wording was changed to the following: The steam supply valves to the AFPT will be opened automatically when adequate suction pressure is re-established and an SFRCS signal remains present provided the steam pressure remained above the AFPT inlet steam pressure interlock setpoint. If steam pressure dropped below the AFPT inlet steam pressure interlock setpoint, manual action will be required to open the steam supply valves.

In Regulatory Applicability Determination (RAD) No. 04-00747, the licensee evaluated UCN 04-021, and determined that the changes made in UCN 04-021 were already evaluated and accepted by a previous safety evaluation. When the inspectors reviewed the referenced safety evaluation, they were not able to find a specific evaluation of this change to this USAR described function. The inspectors determined that this change adversely affected a USAR-described function in that, a previously described automatic feature of the steam inlet valve to the AFW pump was changed, to clarify that this automatic feature was not available under certain conditions. Because of these adverse effects to the USAR, the change should have been evaluated in accordance with the requirements contained in 10 CFR 50.59. Based upon the inspectors concerns, the licensee initiated CR 04-04338.

Analysis:

Because violations of 10 CFR 50.59 are considered to be violations that potentially impede or impact the regulatory process, they are dispositioned using the traditional enforcement process instead of the Significance Determination Process (SDP). In this case, the licensee failed to perform a safety evaluation for changes made to the USAR concerning the low-low pressure interlock for the AFW pumps in accordance with 10 CFR 50.59.

This finding was determined to be more than minor because the inspectors could not reasonably determine that the change would not ultimately require NRC approval. The inspectors determined that even though the change was not adequately evaluated in accordance with 10 CFR 50.59, this violation of the regulatory requirements was of very low safety significance, because the design basis safety-related function of the AFW system, to remove reactor decay heat following a loss of normal feedwater, was not adversely affected. The inspectors completed a significance determination of this finding using IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. The inspectors determined from the mitigating systems evaluation in the Phase 1 Screening Worksheet that all the questions were answered No, therefore the finding was determined to be of very low safety significance (Green).

Enforcement:

On July 2, 2004, while performing the baseline procedure 71111.02, the inspectors identified a NCV of 10 CFR 50.59(d)(1). Title 10 CFR 50.59(d)(1) states, in part, that the licensee shall maintain records of changes in the facility, of changes in procedures, and of tests and experiments. These records must include a written evaluation which provides the bases for the determination that the change, test, or experiment does not require a license amendment.

Contrary to the above, the licensee failed to provide a basis in RAD 04-00747 for the determination that changing requirements in USAR Section 9.2.7.3.c associated with the low-low pressure interlock for the AFW pumps (as documented in UCN 04-021 dated April 24, 2004), was acceptable without a license amendment. The results of this violation were determined to be of very low safety significance. Therefore, this violation was classified as a Severity Level IV Violation. The licensee entered this issue into the corrective action program (CR 04-04338). This Severity Level IV violation is being treated as a NCV consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 05000346/2004010-01(DRS)).

b.2 Inadequate Evaluation of Missile Protection for Diesel Generator Exhaust Stacks and Plant Doors

Introduction:

The inspectors identified an inadequate evaluation performed pursuant to 10 CFR 50.59 associated with the vulnerability of plant doors and the emergency diesel generator (EDG) exhaust stacks to tornado driven missiles. Specifically, the licensee did not provide an adequate response to the question posed in 10 CFR 50.59(c)(2)(vi)and did not demonstrate that the proposed change did not create the possibility of a malfunction of equipment important to safety with a different result than any previously evaluated in the final safety analysis report (as updated). This issue was considered to be of very low safety significance (Green) and was dispositioned as a Severity Level IV NCV.

Description:

The inspectors reviewed 10 CFR 50.59 evaluation, 02-1740, Revision 1. In this evaluation, the licensee accepted an as-found plant condition that the EDG exhaust stacks, as well as certain plant doors, were not physically protected from the effects of tornado missiles as stated in the USAR. On page 8 of evaluation 02-1740 in Section 4.6, the licensee responded to the question posed in 10 CFR 50.59(c)(2)(vi). This question asked, "Does the proposed activity create a possibility for a malfunction of a system, structure, or component important to safety with a different result than any previously evaluated in the final safety analysis report (as updated)?" The licensee identified that a non-USAR-described malfunction was introduced. However, the licensees evaluation attempted to justify why there was not a credible effect on any system, structure or component such that any potential malfunction was eliminated.

The licensee referred to a "threshold of credibility" on the order of 1E-6 per year, which the licensee stated was in accordance with Nuclear Energy Institute (NEI) 96-07, "Guidelines for 10 CFR 50.59 Implementation," Revision 1, as supported by the FENOC 10 CFR 50.59 Guidance Document. To support the credibility threshold determination, the licensee performed a calculation which used a computer code to calculate the probability of tornado missile damage to safety related systems, structures and components without physical protection.

The inspectors were unable to locate a definition of a "threshold of credibility" within NEI 96-07, which the NRC endorsed in Regulatory Guide 1.187. Furthermore, the inspectors determined that NEI 96-07 did not discuss any numerical value below which potential malfunctions were eliminated. Instead, the NEI document stated that possible malfunctions were limited to those as likely to happen as those described in the USAR and that a proposed change or activity which increased the likelihood of a malfunction previously thought to be incredible to the point where it becomes as likely as the malfunctions assumed in the USAR could create a possible malfunction with a different result.

The inspectors noted that tornado generated missiles were evaluated as a potential failure mechanism in the USAR. Therefore, the inspectors determined that the possibility of the EDG exhaust stacks being hit by a tornado generated missile was as likely (on approximately the same order of magnitude) as any other plant feature being struck by a tornado generated missile. The inspectors also noted that failure of the EDGs due to a tornado generated missile was previously considered to be incredible, as the USAR stated that they were physically protected from the effects of tornadoes.

Consequently, the inspectors concluded that the change being evaluated appeared to have increased the likelihood of a malfunction previously thought to be incredible to the point where it becomes as likely as the malfunctions assumed in the USAR. The inspectors consulted with the staff in the Office of Nuclear Reactor Regulation (NRR).

The NRR staff determined that the inspectors conclusions were in accordance with the staffs understanding of both 10 CFR 50.59 and NEI 96-07, Revision 1.

Based on the above, the inspectors determined that the licensee had not provided sufficient justification to answer 10 CFR 50.59(c)(2)(vi) as "No. Specifically, the inspectors concluded that the possibility of a malfunction with a different result appeared to have been created due to the change, because the possible malfunction was at least as likely to happen as those described in the USAR and because the proposed change increased the likelihood of a malfunction previously thought to be incredible to the point where it becomes as likely as the malfunctions assumed in the USAR.

Analysis:

This issue was determined to involve a performance deficiency because the licensee misapplied the criteria of 10 CFR 50.59 and consequently concluded that prior NRC approval was not required when such a conclusion could not be supported by the documented 50.59 evaluation. Because violations of 10 CFR 50.59 are considered to be violations that potentially impede or impact the regulatory process, they are dispositioned using the traditional enforcement process instead of the significance determination process (SDP) described in Inspection Manual Chapter (IMC) 0609, "Significance Determination Process.

The inspectors determined that the finding was more than minor because physical barriers were degraded or missing and because those barriers being degraded could result in one or more of the diesel generators failing to fulfill their design function during a tornado. This was a design issue which affected the Mitigating Systems Cornerstone objective of equipment reliability. Additionally, this finding was determined to be more than minor because the inspectors could not reasonably determine that the change would not ultimately require NRC approval.

The inspectors then assessed the issue through Phase 1 of the SDP. The inspectors answered the question, "Does this issue involve an actual loss of safety function," as "Yes," because under a design basis tornado, the EDG exhaust stacks were not physically protected. Based on this premise, the inspectors entered Phase 2 of the SDP.

The inspectors determined that the only event tree affected was a loss of offsite power concurrent with a loss of one EDG. This was based on the assumption that a tornado missile hitting both EDG exhaust stacks would be an incredible event. The team decreased the initiating event frequency from a "5" (once in 100,000 years) to a "3" (once in 1,000 years) based on the fact that the Davis-Besse switchyard was struck by a tornado in 1998. Based on these credible assumptions, this issue was determined to have very low safety significance or Green. Because the issue was of very low safety significance, the 10 CFR 50.59 violation was categorized as Severity Level IV.

Enforcement:

On July 2, 2004, while performing the baseline procedure 71111.02, the inspectors identified a NCV of 10 CFR 50.59 (d)(1). Title 10 CFR 50.59(d)(1) requires the licensee to maintain records of changes in the facility, of changes in procedures, and of tests and experiments made pursuant to 10 CFR 50.59(c). It further requires that these records include a written evaluation which provides the bases for the determination that the change, test or experiment does not require a license amendment pursuant to 10 CFR 50.59(c)(2).

Contrary to the above, on October 16, 2003, the licensee approved a 10 CFR 50.59 evaluation (02-1740) incorporating a change in the design basis to accept the lack of physical protection for the EDG exhaust stacks from tornado missiles. However, this evaluation did not provide an adequate basis for why a possibility for a malfunction of the diesel generators due to impact on the diesel generator exhaust stacks by a tornado driven missile did not produce a different result than any previously evaluated in the USAR as required by 10 CFR 50.59(c)(2).

The failure to provide a written evaluation which described the basis for concluding a license amendment was not needed was a violation of 10 CFR 50.59(d)(1). This issue had been entered into the licensees corrective action program as CR 04-04685. This Severity Level IV violation is being treated as an NCV consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 05000346/2004010-02).

1R17 Permanent Plant Modifications

.1 Review of Recent Modifications

a. Inspection Scope

From June 28, 2004 through July 2, 2004, the inspectors reviewed 16 permanent plant modifications. The modifications were chosen based upon a consideration of probabilistic risk analysis (PRA) significance in the licensees Individual Plant Evaluation (IPE). The inspectors reviewed these modifications to verify that the completed design changes were in accordance with the specified design requirements and the licensing bases and to confirm that the changes did not affect any systems safety function. Design and post-modification testing aspects were verified to ensure the functionality of the modification, its associated system, and any support systems.

The inspectors also verified that the modifications performed did not place the plant in an increased risk configuration. The list of documents reviewed by the inspectors is included as an attachment to this report.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

.1 Routine Review of Condition Reports

a. Inspection Scope

From June 28, 2004 through July 2, 2004, the inspectors reviewed a selected sample of condition reports associated with Davis-Besse Nuclear Power Stations permanent plant modifications and concerning 10 CFR 50.59 evaluations and screenings. The inspectors reviewed these issues to verify an appropriate threshold for identifying issues and to evaluate the effectiveness of corrective actions. In addition, condition reports written on issues identified during the inspection were reviewed to verify adequate problem identification and incorporation of the problem into the corrective action system.

The specific corrective action documents that were reviewed by the inspectors are listed in the attachment to this report.

b. Findings

No findings of significance were identified.

4OA6 Meetings

.1 Exit Meeting

The inspectors presented the inspection results to Mr. Bezilla and other members of licensee management on July 2, 2004, and by a telephone exit on July 23, 2004. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

D. Blakely, Design Engineer

J. Grabner, Design Engineering Manager

P. Jacobsen, Design Engineer

S. Osting, Design Engineer

G. Wolf, Regulatory Affairs

K. Zellers, Design Engineer

Nuclear Regulatory Commission

J. Grobe, 0350 Chairman

D. Hills, Region III Materials Engineering Branch Chief

J. Lara, Region III Electrical Engineering Branch Chief

C. Lipa, Reactor Projects Branch 4

C. Thomas, Davis-Besse Senior Resident Inspector

Attachment

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000436/2004010-01 NCV Inadequate Safety Evaluation for Changes to the Plant made as Described in the USAR Concerning the low-low pressure interlock for the AFW Pumps

05000436/2004010-02 NCV Inadequate 10 CFR 50.59 Evaluation Regarding Tornado Missile Protection for EDG Exhaust Stacks

Closed

05000436/2004010-01 NCV Inadequate Safety Evaluation for Changes to the Plant made as Described in the USAR Concerning the low-low pressure interlock for the AFW Pumps

05000436/2004010-02 NCV Inadequate 10 CFR 50.59 Evaluation Regarding Tornado Missile Protection for EDG Exhaust Stacks

Discussed

None.

Attachment