IR 05000454/2001016

IR 05000454/2001-016(DRP), IR 05000455/2001-016(DRP), on 11/13-12/31/2001; Exelon Generation Company, LLC; Byron Station, Units 1 & 2. Other Activities

ML020300374
Person / Time
Site:	Byron
Issue date:	01/30/2002
From:	Ann Marie Stone Division of Nuclear Materials Safety III
To:	Kingsley O Exelon Generation Co
References
EA-02-016 IR-01-016
Download: ML020300374 (25)

Text

ary 30, 2002

SUBJECT:

BYRON STATION, UNITS 1 AND 2 INSPECTION REPORT 50-454/01-16(DRP); 50-455/01-16(DRP)

Dear Mr. Kingsley:

On December 31, 2001, the NRC completed an inspection at the Byron Station, Units 1 and 2.

The enclosed report documents the inspection findings which were discussed on January 7, 2002, with Mr. R. Lopriore and other members of your staff. A followup discussion was held with Mr. S. Kuczynski on January 29, 2002.

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

The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

Based on the results of this inspection, the inspectors identified a Severity Level IV violation of NRC requirements. Specifically, in July 1998, your staff implemented a change to the diesel generator (DG) ventilation system that involved an unreviewed safety question and failed to obtain prior NRC approval in accordance with the 10 CFR 50.59 requirements in effect at the time. The change involved defeating the automatic start function of a diesel generator room ventilation fan and covering the outside air damper with a prefabricated cover. The change also substituted operator manual actions in place of automatic system actuation described in the Updated Final Safety Analysis Report (UFSAR). We also evaluated this issue against the current and revised 10 CFR 50.59 requirements. We determined that this issue would have been a violation of the revised 10 CFR 50.59 rule because the change would represent more than a minimal increase in the likelihood of occurrence of a malfunction of a system previously evaluated in the UFSAR. However, because the violation was non-willful and non-repetitive and because it has been entered into your corrective action program, the NRC is treating this issue as a Non-Cited Violation in accordance with Section VI.A.1 of the NRCs Enforcement Policy. If you deny this Non-Cited Violation, you should provide a response with the basis for your denial, within 30 days of the date of this inspection report, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator, Region III; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington DC 20555-0001; and the NRC Resident Inspector at the Byron facility. The current Enforcement Policy is included on the NRCs website at www.nrc.gov/OE. In addition, immediately following the terrorist attacks on the World Trade Center and the Pentagon, the NRC issued an advisory recommending that nuclear power plant licensees go to the highest level of security, and all promptly did so. With continued uncertainty about the possibility of additional terrorist activities, the Nation's nuclear power plants remain at the highest level of security and the NRC continues to monitor the situation. This advisory was followed by additional advisories and although the specific actions are not releasable to the public, they generally include increased patrols, augmented security forces and capabilities, additional security posts, heightened coordination with law enforcement and military authorities, and more limited access of personnel and vehicles to the sites. The NRC has conducted various audits of your response to these advisories and your ability to respond to terrorist attacks with the capabilities of the current design basis threat. From these audits, the NRC has concluded that your security program is adequate at this time.

In accordance with 10 CFR 2.790 of the NRCs Rules of Practice, a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly 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 current link to the Public Electronic Reading Room).

We will gladly discuss any questions you have concerning this inspection.

Sincerely,

/RA/

Ann Marie Stone, Chief Branch 3 Division of Reactor Projects Docket Nos. 50-454; 50-455 License Nos. NPF-37; NPF-66

Enclosures:

Inspection Report 50-454/01-16(DRP);

50-455/01-16(DRP)

See Attached Distribution DOCUMENT NAME: G:\byro\byr2001016 drp.wpd *See previous concurrence To receive a copy of this document, indicate in the box:"C" = Copy without enclosure "E"= Copy with enclosure"N"= No copy OFFICE RIII RIII RIII NAME TTongue/trn AMStone HBClayton DATE 01/29/02 01/30/02 01/30/02 OFFICIAL RECORD COPY

REGION III==

Docket Nos: 50-454; 50-455 License Nos: NPF-37; NPF-66 Report No: 50-454/01-16(DRP); 50-455/01-16(DRP)

Licensee: Exelon Generation Company, LLC Facility: Byron Station, Units 1 and 2 Location: 4450 N. German Church Road Byron, IL 61010 Dates: November 13 through December 31, 2001 Inspectors: R. Skokowski, Senior Resident Inspector B. Kemker, Resident Inspector P. Snyder, Resident Inspector H. Peterson, Senior Operations Engineer C. Thompson, Illinois Department of Nuclear Safety Approved by: Ann Marie Stone, Chief Branch 3 Division of Reactor Projects

SUMMARY OF FINDINGS IR 05000454-01-16(DRP), IR 05000455-01-16(DRP), on 11/13-12/31/2001; Exelon Generation Company, LLC; Byron Station, Units 1 & 2. Other Activities.

The baseline inspection was conducted by resident and region based inspectors. The inspectors identified one Severity Level IV Non-Cited Violation. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter 0609, Significance Determination Process (SDP). The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described at its Reactor Oversight Process website at http://www.nrc.gov/NRR/OVERSIGHT/index.html. Findings for which the SDP does not apply are indicated by No Color or by the severity level of the applicable violation.

A. Inspector Identified Findings Cornerstone: Mitigating Systems The inspectors identified a Severity Level IV Non-Cited Violation. In July 1998, the licensee implemented a change to the diesel generator (DG) ventilation system that involved an unreviewed safety question and failed to obtain prior NRC approval in accordance with the 10 CFR 50.59 requirements in effect at the time. Specifically, the licensee failed to adequately evaluate the defeating of the automatic actuation of the DG ventilation system and replacing it with operator manual actions to recover the systems function. This change increased the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the safety analysis report.

Because the SDP was not designed to assess the significance of violations that potentially impact or impede the regulatory process, this issue is being dispositioned using the traditional enforcement process in accordance with Section IV of the NRC Enforcement Policy. The result of this violation (when a DG was inoperable due to the implementation of the procedure) was assessed significance through the SDP and the severity level of the violation was based on the significance determination. This issue was considered to have more than minor significance, in that, it had a credible impact on safety by affecting the operability, availability, reliability, or function of the DGs.

Because the licensee caused one DG to be inoperable for about 21 days which was longer than the outage time allowed by Technical Specification, the inspectors performed a bounding Phase II SDP analysis. The inspectors evaluated the loss of offsite power and a loss of offsite power coincident with a loss of one division of AC power accident sequences using the following assumptions: (1) minimal credit for operator recovery actions, (2) the DG was inoperable at the start of the event; and (3)

the exposure time for this type of failure occurred for an entire year instead of just during the winters months. The result of these analyses determined that this issue was of very low safety significance (i.e., Green).

The regional senior reactor analyst also performed a qualitative Phase III SDP analysis and determined that external conditions would not be sufficient to increase the safety significance of the issue. Therefore, the issue was classified as a Severity Level IV violation of 10 CFR 50.59. However, because this issue is of very low safety significance and it was captured in the licensees corrective action program, this issue is being treated as a Non-Cited Violation, consistent with Section VI.A.1 of the NRC Enforcement Policy. (Section 4OA5).

Report Details Summary of Plant Status The licensee operated Unit 1 and Unit 2 at or near full power for the duration of the inspection period.

1. REACTOR SAFETY Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity 1R05 Fire Protection (71111.05)

a. Inspection Scope The inspectors examined the plant areas listed below to observe conditions related to fire protection:

+ Unit 1 Turbine Building General Area (Zone 8.5-1), and

+ Unit 2 Turbine Building General Area (Zone 8.5-2).

These areas were selected for inspection because systems, structures, and components that could potentially cause plant transients were located in the areas. The inspectors reviewed applicable portions of the Byron Station Fire Protection Report and assessed the licensees control of transient combustibles and ignition sources, material condition, and operational status of fire barriers and fire protection equipment. During this inspection, the inspectors also interviewed the stations fire marshal. The documents listed at the end of this report were also used by the inspectors to evaluate this area.

In addition, the inspectors assessed fire brigade performance and the drill evaluators critique during a fire brigade drill conducted in the 2A diesel generator (DG) room on December 15, 2001. The drill simulated a lube oil fire associated with a lube oil leak on the 2A DG. The inspectors focused on command and control of fire brigade activities, fire fighting and communication practices, material condition and use of fire fighting equipment, and implementation of pre-fire plan strategies. The inspectors also reviewed the shift managers emergency classification of the simulated event.

b. Findings No findings of significance were identified.

1R11 Licensed Operator Requalification (71111.11)

a. Inspection Scope The inspectors assessed licensed operator performance and the training evaluators critique during a licensed operator training session in the Byron Station operations training simulator on November 19, 2001. The inspectors focused on alarm response, command and control of crew activities, communication practices, procedural adherence, and implementation of emergency plan requirements.

b. Findings No findings of significance were identified.

1R12 Maintenance Rule Implementation (71111.12)

a. Inspection Scope The inspectors evaluated the licensees implementation of the maintenance rule, 10 CFR 50.65, as it pertained to identified performance problems with the following equipment:

+ Miscellaneous Electric Equipment Rooms and Engineered Safety Features Battery Rooms Ventilation During this inspection, the inspectors evaluated the licensees monitoring and trending of performance data, verified that performance criteria were established commensurate with safety, and verified that equipment failures were appropriately evaluated in accordance with the maintenance rule. The documents listed at the end of this report were also used by the inspectors to evaluate this area. The inspectors interviewed system engineers and the stations maintenance rule coordinator.

In addition, the inspectors reviewed the issues that the licensee entered into its corrective action program to verify that identified problems were being entered into the program with the appropriate characterization and significance. The inspectors also reviewed the licensees corrective actions for maintenance rule related issues documented in selected condition reports.

b. Findings No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Evaluation (71111.13)

a. Inspection Scope The inspectors reviewed the licensees evaluation of plant risk for maintenance activities on the following equipment:

+ 2A DG, and

+ 2B Containment Spray System Train.

The inspectors selected these maintenance activities because they involved systems which were risk significant in the licensees risk analysis. The maintenance activity associated with the 2A DG was considered emergent work to repair a damaged voltage regulator. During this inspection, the inspectors assessed the operability of redundant train equipment and verified that the licensees planning of the maintenance activities minimized the length of time that the plant was subject to increased risk. The inspectors interviewed operations, engineering, maintenance, and work control department personnel. The documents listed at the end of this report were also used by the inspectors to evaluate this area.

b. Findings No findings of significance were identified.

1R15 Operability Evaluations (71111.15)

a. Inspection Scope The inspectors evaluated the licensees basis that the issues identified in the following operability evaluation did not render the involved equipment inoperable or result in an unrecognized increase in plant risk:

+ Operability Evaluation 01-017, Potential Distortion of Stuffing Box Extension Wear Ring During Thermal Transients on the Residual Heat Removal Pumps, Revision 0.

The inspectors interviewed operations, engineering, and regulatory assurance department personnel and reviewed applicable portions of the Updated Final Safety Analysis Report (UFSAR) and Technical Specifications (TS). The documents listed at the end of this report were also used by the inspectors to evaluate this area.

b. Findings No findings of significance were identified.

1R17 Permanent Plant Modifications (71111.17)

a. Inspection Scope The inspectors reviewed the permanent plant modifications associated with the Design Change Packages (DCPs) listed below to verify that the modification did not adversely affect the availability, reliability, and functional capability of the systems:

+ DCP 9900292 Provide a 3/16 Inch Diameter Venting Hole in the Upstream (RH Pump) Side of the Valve Disc for Valve 2CV8804A; and

+ DCP 9900387 Revise Trip Logic for Digital Electro-Hydraulic Controller for a Loss of Direct Current Power.

The first modification was installed to address possible thermally and pressure induced pressure locking of the power-operated gate valve 2CV8804A. The second modification rewired the turbine trip logic so that a single card failure would not initiate a turbine trip.

During this inspection, the inspectors evaluated the implementation of these designs to verify that:

• the compatibility, functional properties, environmental qualifications, seismic qualification, and classification of materials and replacement components were acceptable;

• the affected operating procedures and training have been identified and necessary changes were complete;

• the pressure boundary integrity was not compromised;

• the implementation of the modifications did not impair key safety functions;

• no unintended system interactions occurred;

• the system performance characteristics affected by the modification continued to met the design basis; and

• the modification design assumptions were appropriate.

The documents listed at the end of this report were also used by the inspectors to evaluate this area.

b. Findings No findings of significance were identified.

1R19 Post Maintenance Testing (71111.19)

a. Inspection Scope The inspectors evaluated the licensees post maintenance testing for maintenance conducted on the following equipment:

+ 2A DG The inspectors selected this post maintenance activity because the DGs were identified as risk significant in the licensees risk analysis. The inspectors reviewed the scope of the work performed and evaluated the adequacy of the specified post maintenance testing. The inspectors verified that the post maintenance testing was performed in accordance with approved procedures, that the procedures clearly stated acceptance criteria, and that the acceptance criteria were met. During this inspection, the inspectors interviewed maintenance and engineering department personnel and reviewed the completed post maintenance testing documentation. The documents listed at the end of this report were also used by the inspectors to evaluate this area.

b. Findings No findings of significance were identified.

1R22 Surveillance Testing (71111.22)

a. Inspection Scope The inspectors evaluated the surveillance testing activity listed below to verify that the testing demonstrated that the equipment was capable of performing its intended function:

+ Unit Two - 2B DG Operability Surveillance.

The inspectors selected this surveillance test activity because the DGs were identified as risk significant in the licensees risk assessment and the engines were credited as operable in the licensees safety analysis to mitigate the consequences of a potential accident. The inspectors interviewed operations and engineering department personnel, reviewed the completed test documentation, and observed the performance of all or portions of the surveillance testing activity. The documents listed at the end of this report were also used by the inspectors to evaluate this area.

b. Findings No findings of significance were identified.

4. OTHER ACTIVITIES (OA)

4OA1 Performance Indicator Verification (71151)

a. Inspection Scope The inspectors verified the following performance indicators:

+ Safety System Unavailability - Auxiliary Feedwater, and

+ Safety System Unavailability - Emergency Alternating Current (AC) Power.

The inspectors reviewed operating logs, maintenance rule data base entries, maintenance history and surveillance test history for unavailability information for these systems from October 2000 to September 2001. The inspectors also verified the licensees calculation of required hours for both units and evaluated applicable safety system equipment unavailability against the performance indicator definition.

b. Findings No findings of significance were identified. See Section 4OA5.2 for the resolution of a previous performance indicator reporting unresolved item (URI) involving the emergency AC power system.

4OA3 Event Follow-up (71153)

.1 (Closed) Licensee Event Report (LER) 50-454/2001-002-00: Main Steam Isolation Valves (MSIVs) Surveillance Not Performed in Mode 3 as Required by TS Bases Due to Improper Procedure Revision. On September 26, 2001, the licensee identified that both units MSIVs had not been tested in Mode 3 as required by the TS. Subsequently, the licensee requested a Notice of Enforcement Discretion (NOED) to allow both units to continue operating without the immediate completion of this surveillance test requirement. The NRC approved this NOED on September 27, 2001. The inspectors reviewed the LER and concluded that it accurately described the event and that the root cause determination and corrective actions appeared to be adequate. Therefore the LER is closed. However, the regulatory aspects and risk significance of the issue (Unresolved Item (URI) 50-454/455-01-10-02) remains open pending actual testing in Mode 3 and additional NRC review.

4OA5 Other

.1 (Closed) URI 50-454/455-01-06-01(DRP): Review of the Licensees Change to the DG Ventilation System. The inspectors initiated a Task Interface Agreement which requested additional assistance from the Office of Nuclear Reactor Regulation (NRR).

The inspectors identified a Severity Level IV Non-Cited Violation. The licensee failed to obtain prior NRC approval for a change to the DG ventilation system which required a license amendment in accordance with 10 CFR 50.59.

In December 2000, the inspectors identified a URI associated with the 10 CFR 50.59 evaluation for a change the licensee made to the DG ventilation system. Specifically, the change involved defeating the automatic start function of a diesel generator room ventilation fan and covering the outside air damper with a prefabricated cover. The licensee also substituted operator manual actions to recover the ventilation system in place of automatic system actuation described in the UFSAR. The inspectors noted that the licensee had used this modification for several years during the winter months because the air dampers did not seal tightly and allowed excessive leakage of cold air from outside into the DG rooms. The cold air affected DG operability due to minimum temperature requirements for safety related components, lube oil system, and jacket water system. The inspectors reviewed the station operating history and noted that the licensee disabled the DG ventilation system for one DG room at a time with the longest duration of 21 days. The inspectors also noted that while the DG room ventilation system was disabled, the opposite train DG was available as well as the capability to cross-tie emergency power from the other operating unit.

In March 2001, the 10 CFR 50.59 requirements were revised. Because the licensee made this change in July 1998, the NRR staff reviewed the issue against the previous 10 CFR 50.59 requirements. The NRR staff concluded that the licensees substitution of operator manual actions in place of automatic system actuation as described in the UFSAR resulted in an unreviewed safety question and required prior NRC review and approval before implementation. Specifically, the staff determined that the change required prior NRC approval because: (1) the plant subsequently relied on operator intervention for the effective performance of systems that are important to safety and

(2) this reliance on human intervention potentially introduced unanalyzed failure modes caused by operator errors of omission or commission. In accordance with NRC Enforcement Manual Section 8.13, the NRC staff also reviewed the issue against the current 10 CFR 50.59 requirements and determined that the change resulted in more than a minimal increase in the likelihood of a malfunction of equipment that is important to safety. The NRR staff noted that the licensee did not conduct a comprehensive 10 CFR 50.59 evaluation to support the change, failed to perform either a task analysis or walk-through, and did not consider the possibility of operator errors or the likelihood of recovering from such errors. The NRR staff also concluded that the licensee did not provide adequate evidence to support its contention that 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> was sufficient operator response time to: (1) ensure accurate diagnosis of a transient that requires a DG to start, (2) perform the required manual actions on up to four separate DGs, and (3) recover from potential operator errors.

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).

Although the SDP is not designed to assess the significance of violations that potentially impact or impede the regulatory process, the result of a 10 CFR 50.59 violation is assessed significance through the SDP and the severity level of the violation is based on the significance determination. In this case, the licensee modified the 1B DG ventilation in 1999 for a 21 day period. Therefore, the inspectors assessed the significance of having the DG inoperable for 21 days using the SDP.

The inspectors concluded that this issue had a credible impact on safety because the change to the DG ventilation system resulted in an increased likelihood of a malfunction and could have affected the operability, availability, reliability, or function of the DGs.

Because this issue only affected the mitigating systems cornerstone, the inspectors performed a Phase I analysis using the SDP. The inspectors answered yes to Question 3, specifically, that a single DG train was unavailable for greater than the 14-day outage time allowed by TS 3.8.1. The inspectors and regional senior reactor analyst performed a bounding Phase II analysis for the loss of offsite power and the loss of offsite power coincident with a loss of one division of AC power accident sequences using the following assumptions: (1) minimal credit for operator recovery actions, (2) the DG was inoperable at the start of the event; and (3) the exposure time for this type of failure occurred for an entire year instead of just during the winters months. The result of these analyses determined that this issue was of very low safety significance (i.e., Green). The regional senior reactor analyst also performed a qualitative Phase III SDP analysis and determined that external conditions would not be sufficient to increase the safety significance of the issue.

Because this issue was identified prior to March 2001, the issue was evaluated against the previous 10 CFR 50.59 requirements. Specifically, 10 CFR 50.59(a)(1) stated, in part, that the holder of a license authorizing operation of a utilization facility may make changes in the facility as described in the safety analysis report without prior Commission approval, unless the proposed change involved an unreviewed safety question. 10 CFR 50.59(a)(2) stated, in part, that a proposed change shall be deemed to involve an unreviewed safety question if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously

evaluated in the safety analysis report may be increased. 10 CFR 50.59(b)(1) required, in part, that the licensee shall maintain records of changes in the facility to the extent that these changes constitute changes in the facility as described in the safety analysis report. 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.

The Byron/Braidwood Stations UFSAR, Section 9.4.5.2, Diesel Generator Facilities Ventilation System, states, in part, that each diesel generator room ventilation system is interlocked to start when its associated diesel generator starts.

Contrary to the above, on July 8, 1998, the licensee failed to perform an adequate written safety evaluation which provided the bases that a change in the facility did not involve an unreviewed safety question. Specifically, the written safety evaluation for Byron Operating Procedure VD-5, Diesel Generator Room Ventilation System Operation, Revision 4, failed to adequately evaluate the licensee's defeating of the automatic actuation of the diesel generator ventilation system and replacing with operator manual actions to recover the systems function. This change in the facility increased the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the safety analysis report. Consequently, the change involved an unreviewed safety question and was made without prior NRC approval. The result of the violation was determined to be of very low safety significance; therefore, this violation of 10 CFR 50.59 was classified as a Severity Level IV violation. However, because this non-willful violation was non-repetitive, and was captured in the licensees corrective action program (CR 00084634), it is considered a Non-Cited Violation (NCV 50-454/455-01-16-01(DRP)) consistent with Section VI.A.1 of the NRC Enforcement Policy. This URI is closed.

.2 (Closed) URI 50-454/455-00-14-01(DRP): Review of the Licensees Reporting of Unavailability Time for the Emergency AC Power System.

In September 2000, the inspectors identified three discrepancies with respect to the performance indicator for the emergency AC power unavailability time. The inspectors noted that the licensee did not consider the affected DG inoperable when the DG room ventilation system was not capable of performing its safety function. Therefore, the licensee did not include these occurrences against the unavailability time for the emergency AC power system. The licensee submitted a Frequently Asked Question form to the NRC, requesting clarification of this performance indicator. The NRC staff reviewed each of the reporting discrepancies and concluded that the licensee should have accounted for unavailability time for each case. The inspectors determined that if the licensee included unavailability time for these occasions, the performance indicator would not have changed color. Therefore, these reporting discrepancies are considered minor. The licensee entered this issue into its corrective action program as CR B2000-03275, CR 00084936, and CR 00087455. This URI is closed.

.3 (Closed) URI 50-454/455-00-301-01(DRS): This unresolved item involved a potential emergency procedure deficiency.

Procedure 1BEP-0, Reactor Trip or Safety Injection, required operators to assess reactor coolant system (RCS) pressure and if RCS pressure was decreasing with abnormal auxiliary building radiation levels, to eventually transition to BCA-1.2, LOCA

Outside Containment. An operator licensing examination scenario was written expecting that the applicants would perform BCA-1.2 and isolate the inter-system loss of coolant accident (LOCA). However, due to multiple, independent, and unrelated malfunctions, the plant conditions were such that reactor pressure was increasing, and therefore, the applicants were not directed by the emergency procedure to address the leak outside containment when such a condition existed. Following the review of the facilitys condition report B2000-01829 and action item 00031721, the inspectors determined that the current procedure was in compliance with Emergency Response Guidelines developed by Westinghouse and the Westinghouse Owners Group guidance; and the facilitys probabilistic risk analysis results did not meet the threshold to require a revision to the emergency procedure.

The inspectors determined that the scenario with multiple, independent, and unrelated malfunctions occurring in a short time frame placed the operators outside the procedure.

This practice was not uncommon during training and examination scenarios when multiple events are occurring. In addition, procedures are not, and can not be written to cover every conceivable scenario situation. During the scenario the operators correctly followed procedures and addressed the plant conditions that would have steered them to the inter-system LOCA procedure. However, the plant conditions did not warrant a transition out of the existing emergency procedure being implemented at that time. In general, the inspectors concluded that the operators adequately followed the emergency procedures. Although the transition into the inter-system LOCA procedure did not occur as expected, the operators appropriately followed procedures and satisfactorily addressed the LOCA condition and placed the plant in a safe condition. The inspectors concluded that the emergency procedure, based on multiple scenario conditions, was adequate. The licensees actions were considered reasonable, and no findings of significance were identified. This URI is closed.

4OA6 Meetings Resident Inspector Exit Meeting The inspectors presented the inspection results to Mr. R. Lopriore and other members of licensee management at the conclusion of the inspection on January 7, 2002. The licensee acknowledged the findings presented. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary.

Proprietary information was examined during this inspection but is not specifically discussed in this report.

On January 29, 2002, Mrs. A. M. Stone contacted Mr. S. Kuczynski and discussed the changes in the characterization of the findings as originally presented during the January 7 exit meeting.

KEY POINTS OF CONTACT Licensee R. Lopriore, Site Vice President B. Altman, Maintenance Manager R. Blaine, Radiation Protection Director D. Combs, Site Security Manager D. Drawbaugh, Regulatory Assurance B. Grundmann, Regulatory Assurance Manager K. Hansing, Site Nuclear Oversight Manager J. Heaton, Lead License Requalification Specialist M. Heinzer, Nuclear Oversight Assessment Manager D. Hoots, Operations Manager W. Kolo, Work Management Director S. Kuczynski, Station Manager T. Roberts, Engineering Director T. Schuster, Executive Assistant D. Spoerry, Training Manager S. Stimac, Shift Operations Superintendent Nuclear Regulatory Commission A. Stone, Chief, Branch 3, Division of Reactor Projects S. Burgess, Senior Reactor Analyst, Division of Reactor Safety

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-454/455-01-16-01 NCV Failure to obtain prior NRC approval for a change to the diesel generator ventilation system that resulted in an unreviewed safety question Closed 50-454/2001-002-00 LER Main Steam Isolation Valves Surveillance not Performed in Mode 3 as Required by Technical Specification Bases Due to Improper Procedure Revision 50-454/455-00-14-01 URI Review of the licensees reporting of unavailability time for the emergency ac power system 50-454/455-00-301-01 URI Potential emergency procedure deficiency.

50-454/455-01-06-01 URI Review of the licensees change to the diesel generator ventilation system 50-454/455-01-16-01 NCV Failure to obtain prior NRC approval for a change to the diesel generator ventilation system that resulted in an unreviewed safety question Discussed None

LIST OF ACRONYMS USED AC Alternating Current BAP Byron Administrative Procedure BOP Byron Operating Procedure BOSR Byron Operating Surveillance Requirement Procedure CFR Code of Federal Regulations CR Condition Report DCP Design Change Package DG Diesel Generator DP Differential Pressure DRP Division of Reactor Projects ESF Engineered Safety Features FW Feedwater LCOAR Limiting Condition for Operation Action Requirement LER Licensee Event Report LOCA Loss of Coolant Accident MSIV Main Steam Isolation Valve NCV Non-Cited Violation NEI Nuclear Energy Institute NOED Notice of Enforcement Discretion NRC Nuclear Regulatory Commission NSP Nuclear Station Procedure OOS Out-of-Service RCS Reactor Coolant System RH Residual Heat Removal SDP Significance Determination Process SER Safety Evaluation Report SPP Special Plant Procedure TS Technical Specification UFSAR Updated Final Safety Analysis Report URI Unresolved Item WR Work Request

LIST OF DOCUMENTS REVIEWED 1R05 Fire Protection Byron/Braidwood Stations Fire Protection Revision 19 Report Byron Station Pre-Fire Plans and Drawings Byron Administrative Byron Station Pre-Fire Plan Revision 0 Procedure (BAP)

1100-17T1 BAP 1100-7 Fire Prevention for Transient Combustibles Revision 10 BAP 1100-7A1 Minor Transient Combustibles Revision 1 Nuclear Station Fire Drill Performance Revision 3 Procedure (NSP)

OP-AA-201-003 NSP OP-AA-201-003 Fire Drill Record December 15, 2001 Attachment 1 NSP OP-AA-201-003 Fire Drill Scenario October 12, 2001 Attachment 3 Condition Report (CR) Plant Page Near 401' Fire Brigade Cage December 17, 2001 00087056 1 Inaudible 1R11 Licensed Operator Requalification Program Licensed Operator Simulator Training Scenario for Scenario completed November 19, 2001 1R12 Maintenance Rule Implementation NSP ER-3010 Maintenance Rule Revision 0 NUMARC 93-01 Industry Guideline for Monitoring the Revision 2 Effectiveness of Maintenance at Nuclear Power Plants

Maintenance Rule Performance Monitoring October 1, 1999 Data for Criteria VE-1, Provide Ventilation through for the Miscellaneous Electric Equipment October 19, 2001 Rooms and Engineered Safety Features (ESF) Battery Rooms CR B2000-00100 2TS-VE003 Out of Tolerance, Expanded January 11, 2000 Tolerance Exceeded CR B2000-00324 Out-of-Service (OOS) Card Appears to Be January 30, 2000 Left on Equipment When OOS Was Reset to Approved Status CR B2000-00429 Unplanned Limiting Condition for February 8, 2000 Operation Action Requirement (LCOAR)

Entry for Battery Room 111 Exhaust Fan Trip CR B2000-01448 Unplanned LCOAR Entry for ESF Battery May 19, 2000 111 Room Ventilation CR B2000-01694 Instrument Out of Tolerance, Expanded June 14, 2000 Tolerance Exceeded CR B2000-03692 Unplanned LCOAR Entry 1BOL [Unit 1 December 5, 2000 Byron Operating Limits Procedure] VE1 Due to 111 Battery Room Fan Tripping CR B2000-03989 Nuisance Alarm on 2VE05C High December 29, 2000 Differential Pressure (DP) Trip Alarm CR B2001-00251 Trend on Diesel Fuel Oil Pump Relief January 18, 2001 Valve Replacement CR B2001-00252 Inappropriate Corrective Action to Prevent January 18, 2001 Recurrence for Trend 97-014 CR B2001-00300 Results from Common Cause Analysis on January 22, 2001 the Process Radiation Monitoring System CR B2001-00374 Maintenance Rule Peer Group January 26, 2001 Containment Closure Industry Event Review CR B2001-00600 Housekeeping / Foreign Materials February 8, 2001 Exclusion Concern CR B2001-00646 Unplanned Reduction in Circulating Water February 11, 2001 Blowdown Flow Due to River Screen House Traveling Screen Plugging

CR B2001-00750 Unit 1 Division 12 Miscellaneous Electrical February 17, 2001 Equipment Room Temperature Cold CR B2001-00831 Failure of Unit 2 Turbine Vibration February 23, 2001 Supervisory Module CR B2001-00875 0B Primary Water Makeup Pump February 27, 2001 Excessive Amperes/Trip CR B2001-01446 Unplanned Administrative LCOAR Entry April 6, 2001 2BOL VE1 Due to Fan 2VE02C Tripped on High DP CR B2001-02206 Unplanned Administrative LCOAR Due to May 12, 2001 Fan Trip (1VE02C)

CR B2001-02622 Unplanned Technical Requirements June 8, 2001 Manual LCOAR Entry - 112 Battery Room Exhaust Fan 1VE02C Tripped CR B2001-02577 Unplanned LCOAR Entry (1BOL VE1) Due June 5, 2001 to ESF Battery Room Exhaust Fan Trip CR B2001-02833 Unplanned Administrative LCOAR Entry June 23, 2001 (1BOL VE1) - Trip of ESF Battery Room 112 Exhaust Fan CR B2001-03438 Battery 112 Room Ventilation Fan Tripped August 8, 2001 CR 00074710 112 Battery Room Ventilation Fan 1VE02C September 9, 2001 Trip - Unplanned LCOAR 1BOL VE1 CR 00077989 Battery Room Fan Division 11 Tripped October 6, 2001 CR 00078056 Battery Room 111 Exhaust Fan Trip on October 7, 2001 High DP CR 00078092 Unplanned LCOAR Entry on 1VE03C, October 8, 2001 Division 11 Battery Ventilation Fan Trip CR 00078498 Battery Room 111 Exhaust Fan Trip October 11, 2001 CR 00080602 Unplanned Administrative LCOAR Entry, October 27, 2001 111 Battery Room Ventilation Fan Trips CR 00080833 Unplanned LCOAR for 111 Battery Room October 29, 2001 Exhaust Fan (1VE03C)

CR 00076596 2A Main Steam Power Operated Relief September 26, 2001 Valve Will Not Stroke With Manual Pump

1R13 Maintenance Risk Assessments and Emergent Work Evaluation Byron Station Technical Specifications (TS)

Byron/Braidwood Stations Updated Final Safety Analysis Report (UFSAR)

Byron Operating On-Line Risk/Protected Equipment Revision 2 Department Policy No. 400-47 NSP WC-AA-103 On-Line Maintenance Revision 4 NSP WC-AA-104 Review and Screening for Production Risk Revision 4 CR 00087525 1 Failure to comply with Operations December 19, 2001 Department Policy No. 400-47 1R15 Operability Evaluations Byron Station TS Byron/Braidwood Stations UFSAR NSP CC-3001 Operability Determination Process Revision 0 NSP LS-AA-105-1000 Operability Determination Guidance Revision 0 Manual NRC Generic Letter Information to Licensees Regarding NRC Revision 1 91-18 Inspection Manual Section on Resolution of Degraded and Nonconforming Conditions NRC Inspection Operable/Operability: Ensuring the October 8, 1997 Manual, Part 9900 Functional Capability of a System or Component Operability Evaluation Potential Distortion of Stuffing Box Revision 0 01-017 Extension Wear Ring During Thermal Transients on the Residual Heat Removal (RH) Pumps Byron Operating Placing the RH System in Shutdown Revision 21 Procedure (BOP) RH-6 Cooling 50.59 Screening Revision 21 to BOP RH-6, Placing the RH Revision 0 6D-01-0336 System in Shutdown Cooling

F-95N3D-136186-000 Finite Element Analysis of Casing Cover September 18, 1995 Deformation Due to Thermal Transients, Pump Model 8X20WDF Ingersoll-Dresser Pump Company, Prepared for Watts Bar Nuclear Power Plant Westinghouse RH Pump Operating Recommendations Revision 0 Technical Bulletin ESBU-TB-96-03 Watts Bar Incident RH Pump 1B-B Failures (With Pump 1A-A Revision 2 Investigation Event Supplemental Results)

Report II-W-94-014 RH in Service at Temperature < 350 November 19, 2001 Degrees Fahrenheit, email Annie Wong to John Panici, et al CR 00082603 Byron Review of Braidwood 2B RH Pump November 12, 2001 Failure CR 00087847 1 BOP RH-6 Revision 21 50.59 Screening December 21, 2001 Needs Improvement 1R17 Permanent Plant Modifications Byron/Braidwood Stations UFSAR Byron Station TS Design Change Provide a 3/16 Inch Diameter Venting Hole January 26, 2000 Package (DCP) in the Upstream (RH Pump) Side of the 9900292 Valve Disc for Valve 2CV8804A DCP 9900387 Revise Trip Logic for Digital Electro- December 29, 1999 Hydraulic Controller for a Loss of Direct Current Power Drawing M-62 Diagram of Residual Heat Removal Revision AY Drawing M-64 Diagram of Chemical and Volume Control Revision H Sheet 4B and Boron Thermal Regeneration NRC Generic Letter Pressure Locking and Thermal Binding of August 17, 1995 95-07 Safety-Related Power-Operated Gate Valves

1R19 Post Maintenance Testing Byron/Braidwood Stations UFSAR Work Request (WR) Contingency Package Troubleshooting on August 6, 2001 00347069-00 2A Diesel Generator (DG)

WR 00347069-01 Electrical Maintenance Troubleshooting on November 19, 2001 2A DG WR 00347069-02 Operations Post Maintenance Test November 19, 2001 Special Plant 2A DG Voltage Regulator Special Revision 1 Procedure (SPP) 01- Procedure 026 Byron Plant Review Post Maintenance Test Plan for 2A DG November 19, 2001 Report 01-062 Voltage Regulator Byron Plant Operating SPP 01-026, 2A DG Voltage Regulator November 19, 2001 Review Committee and Plant Review Report 01-062, 2A DG Minutes01-090 Post Maintenance Test 2BOL 8.1 LCOAR AC [Alternating Current] Sources - Revision 5 Operating CR 00082913 Unplanned LCOAR Entry - 2A DG Failed to November 15, 2001 Reach Rated Voltage CR 00082931 Chart Recorder Jumper Lead Caused November 15, 2001 Short on 2A DG Circuit CR 00083302 Isolation Transformer Installed on 2A DG November 17, 2001 on Hold CR 00083474 2A DG SPP 01-026 Procedure and November 20, 2001 Performance Problems 1R22 Surveillance Testing Byron Station TS Byron/Braidwood Stations UFSAR Unit 2 Byron Operating Unit Two - 2B Diesel Generator Operability Revision 8 Surveillance Surveillance Requirements Procedure (BOSR)

8.1.2-2

4OA1 Performance Indicator Verification NEI [Nuclear Energy Regulatory Assessment Performance Revision 1 Institute] 99-02 Indicator Guideline NSP RS-AA-122-104 Performance Indicator - Safety System Revisions 2 and 3 Unavailability (High Pressure Safety Injection/High Pressure Core Injection, Residual Heat Removal, Reactor Core Isolation Cooling/Auxiliary Feedwater, Emergency Diesel Generator)

NSP LS-AA-2040 Monthly Performance Indicator Data Revision 06/25/2001 Elements for Safety System Unavailability -

Emergency AC Power NSP LS-AA-2060 Monthly Performance Indicator Data Revision 06/25/2001 Elements for Safety System Unavailability -

Reactor Core Isolation Cooling (BWRs) or Auxiliary Feedwater (PWRs) Systems Byron Shift Managers Logs October 1, 2000 through September 30, 2001

CR B2000-03275 Carbon Dioxide Puff Test Impact on NRC October 30, 2000 Indicator on DG Safety System Unavailability CR B2000-03441 Additional Unavailability for 1A DG November 14, 2000 CR B2000-03632 NRC Information Notice Implementation November 30, 2000 Warrants Review

CR B2000-03952 1A DG Lube Oil Temperature Affected By December 26, 2000 Cold Outside Air Temperatures CR B2001-00296 1B DG Jacket Water Pump Seal Repair January 22, 2001 Impact on Performance Indicators CR B2001-01802 NEI Performance Indicator Database April 18, 2001 Problem CR B2001-02748 Work Process Improvement June 18, 2001

CR 00084936 DG Unavailability Due to Ventilation December 4, 2001 Damper Covers

CR 00087455 Unfavorable Response to Frequently December 19, 2001 Asked Question on Safety System Unavailability

OA3 Event Follow-up Licensee Event Report Main Steam Isolation Valves Surveillance November 26, 2001 50-454/2001-002-00 not Performed in Mode 3 as Required by Technical Specification Bases Due to Improper Procedure Revision 4OA5 Other Byron/Braidwood Stations Updated Final Safety Analysis Report (UFSAR), Section 9.4.5.2, Diesel Generator Facilities Ventilation System NUREG-0876 Safety Evaluation Report (SER) Related to the Operation of Byron Station, Units 1 and 2, Section 9.4.5, Engineered Safety Features Ventilation and Cooling Systems NRC Generic Letter Guidance on Resolution of Degraded and November 7, 1991 91-18 Nonconforming Conditions and on Operability NRC Information Crediting of Operator Action in Place of October 23, 1997 Notice 97-78 Automatic Actions and Modifications of Operator Action, Including Response Times BOP VD-5 DG Room Ventilation System Operation Revision 4 Onsite Review 97-003 DG Ventilation System Impact on Diesel January 23, 1997 Generator Operability Safety Evaluation Safety Evaluation Supporting the Findings January 24, 1997 TI-97-0008 and Recommendations Made in Onsite Review 97-003 Memorandum from Task Interface Agreement - TIA 2001-008, November 19, 2001 Ledyard B. Marsh to Evaluation of a Change to the Byron Geoffrey E. Grant Station DG Ventilation System Per 10 CFR 50.59" NEI 99-02 Regulatory Assessment Performance Revision 1 Indicator Guideline CR B2000-01829 Condition Report: Potential Emergency June 21, 2000 Procedure Problem Identified During ILT NRC Exam Administration

AR 00031721 Action Request: Action Item to Evaluate December 20, 2000 Revision of Emergency Procedures and Search WOG as Needed BOP VD-5 DG Room Ventilation System Operation Revision 4

CR B2000-03275 Carbon Dioxide Puff Test Impact on NRC October 30, 2000 Indicator on DG Safety System Unavailability

CR B2000-03952 1A DG Lube Oil Temperature Affected By December 26, 2000 Cold Outside Air Temperatures

CR 00084936 DG Unavailability Due to Ventilation December 4, 2001 Damper Covers

CR 00087455 Unfavorable Response to Frequently December 19, 2001 Asked Question on Safety System Unavailability

• 1 Condition Report written as a result of the inspection.

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