IR 05000219/2009007

IR 05000219-09-007; 05/04/2009 - 05/15/2009; Oyster Creek Generating Station; Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications

ML091800174
Person / Time
Site:	Oyster Creek
Issue date:	06/29/2009
From:	Doerflein L Engineering Region 1 Branch 2
To:	Pardee C Exelon Generation Co, Exelon Nuclear
Pindale S
References
IR-09-007
Download: ML091800174 (28)

Text

une 29, 2009

SUBJECT:

OYSTER CREEK GENERATING STATION - NRC EVALUATION OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT MODIFICATIONS TEAM INSPECTION REPORT 05000219/2009007

Dear Mr. Pardee:

On May 15, 2009, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Oyster Creek Generating Station. The enclosed inspection report documents the inspection results, which were discussed on May 15, 2009, with Mr. T. Rausch, Site Vice President, and other members of your staff.

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.

In conducting the inspection, the team reviewed selected procedures, calculations and records, observed activities, and interviewed station personnel.

The report documents two NRC-identified findings; one Severity Level IV violation and one finding of very low safety significance (Green). Both of these findings were determined to involve violations of NRC requirements. However, because of the very low safety significance and because they were entered into your corrective action program, the NRC is treating these findings as non-cited violations (NCV) consistent with Section VI.A.1 of the NRCs Enforcement Policy. If you contest any NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspectors at Oyster Creek.

In addition, if you disagree with the characterization of any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region I, and the NRC Resident Inspectors at Oyster Creek. The information you provide will be considered in accordance with Inspection Manual Chapter 0305.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the

Mr. NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRC's 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/

Lawrence T. Doerflein, Chief Engineering Branch 2 Division of Reactor Safety Docket No: 50-219 License No: DPR-16 Enclosure: Inspection Report 05000219/2009007 w/Attachment: Supplemental Information

M

SUMMARY OF FINDINGS

IR 05000219/2009007; 05/04/2009 - 05/15/2009; Oyster Creek Generating Station; Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications.

The report covers a two week inspection of the evaluations of changes, tests, or experiments and permanent plant modifications. The inspection was conducted by three region based engineering inspectors. One Severity Level IV violation and one finding of very low safety significance (Green) were identified, both of which were also considered to be non-cited violations. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.

Cornerstone: Mitigating Systems

• Severity Level IV. The team identified a Severity Level IV non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments, in that, Exelon did not obtain a license amendment for a change in the facility that involved a change to the technical specifications (TS). Specifically, Exelon implemented a temporary modification that changed the secondary containment boundary, but was prohibited by TS requirements, without first obtaining the necessary license amendment. In response, Exelon entered the issue into the corrective action program for evaluation. Current compliance with TS was not challenged since the temporary modification was restored as of November 15, 2008.

The violation is more than minor because the change that required the 10 CFR 50.59 evaluation would have required NRC review and approval prior to implementation.

Because this was a violation of 10 CFR 50.59, it was considered to be a violation that potentially impedes or impacts the regulatory process. Therefore, this violation was evaluated using the traditional enforcement process. Comparing this item to the examples in NUREG 1600 (Enforcement Policy), Supplement I, this finding is similar to Item D.5, Violations of 10 CFR 50.59 that result in conditions evaluated as having very low safety significance (i.e., Green) by the SDP. This is an example of a Severity Level IV violation. The team determined the violation to be of very low safety significance (Green) because it did not adversely impact shutdown mitigation capabilities and did not result in a loss of control.

This finding has a cross-cutting aspect in the area of Human Performance, Decision-Making Component, because Exelon did not use conservative assumptions in decision making during the safety evaluation performance and review. Specifically, Exelon did not consider the TS requirements and UFSAR and TS bases when performing and reviewing a safety evaluation that permitted a configuration that was not authorized by TSs. (IMC 0305, Aspect H.1(b)) (1R17.1)ii

Cornerstone: Barrier Integrity

Green.

The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, in that Exelon did not ensure the adequacy of a reactor building closed cooling water system containment isolation check valve design. Specifically, Exelon modified the check valve but did not ensure that the replacement valve could meet the existing design basis temperature value. In response, Exelon entered the issue in their corrective action program and evaluated the design temperature of the check valve to assure the valve would function properly during postulated events.

The finding is more than minor because it is associated with the design control attribute of the Barrier Integrity Cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The team determined the finding screened as very low safety significance (Green) because it did not represent a degradation of the radiological barrier function provided for the control room, auxiliary building, or spent fuel pool, did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere, did not represent an actual open pathway in the physical integrity of reactor containment, and did not involve an actual reduction in function of hydrogen igniters in the reactor containment.

This finding has a cross-cutting aspect in the area of Human Performance, Work Practices Component, because Exelon did not define and effectively communicate expectations regarding procedural compliance and personnel did not follow procedures.

Specifically, Exelon did not comply with procedure CC-AA-102, Design Input and Configuration Change Impact Screening, to evaluate the design temperature of the newly installed check valve to ensure that all affected systems can perform their design basis functions. (IMC 0305, Aspect H.4(b)) (1R17.2.1)

Licensee-Identified Violations

None.

iii

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R17 Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications

(IP 71111.17)

.1 Evaluations of Changes, Tests, or Experiments (22 samples)

a. Inspection Scope

The team reviewed eight safety evaluations to determine whether the changes to the facility or procedures, as described in the Updated Final Safety Analysis Report (UFSAR), had been reviewed and documented in accordance with 10 CFR 50.59 requirements. In addition, the team evaluated whether Exelon had been required to obtain NRC approval prior to implementing the change. The team interviewed plant staff and reviewed supporting information including calculations, analyses, design change documentation, procedures, the UFSAR, technical specifications (TS), and plant drawings, to assess the adequacy of the safety evaluations. The team compared the safety evaluations and supporting documents to the guidance and methods provided in Nuclear Energy Institute (NEI) 96-07, Guidelines for 10 CFR 50.59 Evaluations, as endorsed by NRC Regulatory Guide 1.187, "Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments," to determine the adequacy of the safety evaluations.

The team also reviewed a sample of fourteen 10 CFR 50.59 screenings and applicability determinations for which Exelon had concluded that no safety evaluation was required.

These reviews were performed to assess whether Exelon's threshold for performing safety evaluations was consistent with 10 CFR 50.59. The sample included design changes, calculations, procedure changes, and temporary alterations.

The team reviewed safety evaluations Exelon had performed during the time period covered by this inspection (i.e., since the last modifications inspection). The screenings and applicability determinations were selected based on the risk significance of the associated structures, systems, and components.

In addition, the team compared Exelon's administrative procedures used to control the screening, preparation, review, and approval of safety evaluations to the guidance in NEI 96-07 to determine whether those procedures adequately implemented the requirements of 10 CFR 50.59. The reviewed safety evaluations, screenings, and applicability determinations are listed in the attachment.

b. Findings

Introduction:

The team identified a Severity Level IV non-cited violation (NCV) of 10 CFR 50.59, Changes, Tests, and Experiments, in that, Exelon did not obtain a license amendment for a change in the facility that involved a change to the TSs.

Description:

The team reviewed safety evaluation 2008-E-0004, PM Task PM00311M -

TCC to Move Secondary Containment Barrier in Trunnion Room. This safety evaluation was performed to support a temporary modification that permitted plant staff to keep the trunnion room door open in the Refuel and Shutdown modes of operation. Specifically, the temporary modification permitted Exelon to seal the trunnion room penetrations (i.e.,

the hub drain, deck drain, vent return duct, and floor vent supply hole), while keeping the trunnion room door open, to maintain secondary containment. In effect, the temporary modification re-defined the secondary containment boundary, moving the boundary from the trunnion room door and associated external wall to the internal walls, ceiling and floor of the trunnion room.

The trunnion room has a single door to the area, which houses the two outboard main steam isolation valves. As per the Updated Final Safety Analysis Report (UFSAR),

Section 6.2.3.2, the trunnion room is an integral part of the reactor building (Secondary Containment). It further states that the door design has been considered within the capability of the standby gas treatment system and the reactor building ventilation system to maintain reactor building leakage within the designed maximum.

TS 3.5.B, Secondary Containment, requires secondary containment integrity be maintained at all times unless certain conditions are met (i.e., reactor subcritical and in the cold shutdown condition, reactor or drywell head in place, and no work on reactor or connected systems and no operations in, above or around the spent fuel pool that could result in a release of radioactive materials). Secondary Containment Integrity is defined (TS Definition 1.14) as having the reactor building closed and at least one door at each access opening closed. The definition further states that momentary opening and closing of the trunnion room door does not constitute a loss of secondary containment integrity.

Based on the above, the team determined that Exelon approved a 10 CFR 50.59 safety evaluation (2008-E-0004) on October 25, 2008, which authorized the implementation of a temporary modification that changed the facility as described in the UFSAR, but was prohibited by TS 3.5.B. This is contrary to the requirements of 10 CFR 50.59(c)(1)(i),which states in part, that a licensee may make changes in the facility as described in the final safety analysis report without obtaining a license amendment pursuant to Section 50.90 only if a change to the TSs incorporated in the license is not required. The team found that the temporary modification was implemented during the period October 28, 2008, to November 15, 2008, during the refueling outage; and as a result, the requirements of TS 3.5.B were not satisfied.

In evaluating the significance of this event, the team assessed the details associated with implementing the temporary modification. The penetrations in the trunnion room were identified in the temporary modification and were, in fact, sealed to represent a secondary containment boundary. Specifically, ventilation penetrations were covered and taped closed, and drains were plugged. While these boundaries were not qualified as secondary containment boundaries, their integrity was verified via test in that the standby gas treatment system demonstrated its TS required capability to maintain a 1/4 inch of water vacuum in the reactor building.

Exelon entered this issue into their corrective action program as IR 920109 for evaluation and correction of this issue and the associated cause(s). Because this temporary modification had been restored as of November 15, 2008, current compliance with TSs was not challenged, and additional immediate actions were not required. Also, based upon the results of NRC review of additional safety evaluations during this inspection, the inspectors did not identify extent-of-condition concerns.

Analysis:

The team determined that the failure to obtain NRC approval prior to implementing the temporary modification associated with the trunnion room door was a performance deficiency that was reasonably within Exelons ability to foresee and prevent. Because this was a violation of 10 CFR 50.59, it was considered to be a violation that potentially impedes or impacts the regulatory process. Therefore, this violation was characterized using the traditional enforcement process.

This violation was more than minor because the change that required the 10 CFR 50.59 evaluation would have required NRC review and approval prior to implementation.

Although the SDP is not designed to assess traditional enforcement violations, the NRC assesses the significance of 10 CFR 50.59 violations through the SDP. Accordingly, the team completed an SDP Review in accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and Appendix G, Shutdown Operations Significance Determination Process. In accordance with Appendix G, this finding did not adversely impact shutdown mitigation capabilities and did not result in a loss of control. Accordingly, this finding did not require quantitative assessment and screened to Green (very low safety significance). This violation involved a change that required a license amendment before its implementation. Comparing this item to the examples in NUREG 1600, Supplement I, Reactor Operations, this finding is similar to Item D.5, Violations of 10 CFR 50.59 that result in conditions evaluated as having very low safety significance (i.e., Green) by the SDP. This is a Severity Level IV Violation.

This finding has a cross-cutting aspect in the area of Human Performance, Decision-Making Component, because Exelon did not use conservative assumptions in decision making during the safety evaluation performance and review. Specifically, Exelon did not consider the TS requirements, UFSAR and TS bases when performing and reviewing a safety evaluation that permitted a configuration that was not authorized by TS 3.5.B. (IMC 0305, Aspect H.1(b))

Enforcement:

10 CFR 50.59(c)(1)(i) states in part, that a licensee may make changes in the facility as described in the final safety analysis report without obtaining a license amendment pursuant to Section 50.90 only if a change to the TSs incorporated in the license is not required. Contrary to the above, on October 25, 2008, Exelon approved a 10 CFR 50.59 evaluation (2008-E-0004) that authorized the implementation of a temporary modification that changed the facility as described in the UFSAR, but was prohibited by TS requirements. Subsequently, the temporary modification was implemented during the period October 28, 2008, to November 15, 2008.

Because this Severity Level IV violation of very low safety significance 1) did not represent a condition where Exelon failed to restore compliance; 2) was entered into Exelons corrective action program (IR 920109); 3) did not have any willful aspects; and 4) was not repetitive, this violation is being treated as a non-cited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy.

(NCV 05000219/2009007-01, Inadequate 10 CFR 50.59 Evaluation for Trunnion Room Door/Secondary Containment Temporary Modification)

.2.1 Replacement of Containment Isolation Check Valve V-5-165

a. Inspection Scope

The team reviewed a modification (08-01004) that replaced the reactor building closed cooling water (RBCCW) system containment isolation check valve, V-5-165. The modification was implemented to address the local leak rate test (LLRT) failure history of the installed valve. The review was performed to verify that the design and licensing bases and performance capability of the RBCCW system and containment integrity had not been degraded by the modification. Additionally, the 10 CFR 50.59 screen associated with this modification was reviewed as described in section 1R17.1 of this report.

The team assessed whether the component safety classification and specific safety functions were maintained. The team reviewed calculations and other technical evaluations to assess whether the modification was consistent with assumptions in the design and licensing bases related to the operation of the RBCCW system and to assure containment integrity. Surveillance and post-modification testing were reviewed to verify whether the check valve would function in accordance with the design assumptions and to verify that test results appropriately supported system operability. Finally, the team conducted interviews with engineering staff to determine if the valve would function in accordance with technical and design assumptions. The documents reviewed are listed in the attachment.

b. Findings

Introduction:

The team identified a finding of very low safety significance (Green)involving a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion III, Design Control, in that, Exelon did not ensure the adequacy of RBCCW system containment isolation check valve, V-5-165, design. Specifically, the design temperature of the replacement valve was less than the design basis temperature at which the valve was designed to function.

Description:

The team reviewed modification 08-01004, which involved the November 2008 replacement of RBCCW check valve V-5-165. This valve is located inside primary containment and is designed to close on loss of RBCCW system flow to provide containment isolation during a postulated accident. The modification was performed as an equivalent change package as per procedure CC-AA-103, Configuration Change Control for Permanent Physical Plant Changes, due to the similarities between the originally installed valve and replacement valve. This modification was implemented to address the failure of the LLRT of the installed valve that occurred in November 2008.

This valve has historically failed the LLRT in the past and was last replaced with an identical valve during the October 2006 refueling outage.

The team identified an issue with the design temperature of the replacement valve regarding its ability to function in the full range of expected containment temperatures during a postulated accident. Specifically, the design temperature of the replacement valve was rated for 250°F, which was based on the polymer seat material. The section of RBCCW piping that contained (and included) the valve, which was located within the primary containment, was rated for 350°F. The team reviewed the primary containment accident composite temperature profile in the UFSAR, which indicated that the maximum primary containment temperature peaks at 317°F. The team determined that the modification evaluation and associated documentation failed to evaluate the acceptability of the valve to function at a primary containment temperature of 317°F. In addition, the team determined that the process used to perform the modification was not appropriate since the equivalent change process can only be used if the change does not result in a change to bounding technical requirements, which ensures performance of the design basis function. Rather, the design change process should have been used to evaluate that the replacement valves design basis function was not degraded by its maximum design temperature limit of 250°F.

In response to the teams concern, Exelon entered the above issue into the corrective action program (IR 917125), and implemented actions to verify operability of the check valve. Specifically, Exelon performed technical evaluation (A2184770) that concluded the temperature of the valve polymer seat material would not exceed its design temperature for the relatively short period of time that the primary containment temperature would be elevated during a postulated accident. In addition, the technical evaluation accounted for the affect of radiation damage to the polymer seat material, which was also omitted in the equivalency change evaluation. Exelon reclassified the modification to a design change package (09-00396) and conducted the appropriate 10 CFR 50.59 screen. The team reviewed Exelons bases for operability and independently evaluated valve operability. The team similarly concluded that the issue identified did not render the isolation valve inoperable.

Analysis:

The team determined that the failure to properly assess the design temperature of the modified RBCCW isolation check valve, V-5-165, was a performance deficiency that was reasonably within Exelons ability to foresee and prevent. The finding was more than minor because it was similar to NRC Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, Example 3j, in that the modification did not properly assess the adequacy of the design temperature of the RBCCW containment isolation check valve, which represented a reasonable doubt on the operability of the valve. The finding was associated with the design control attribute of the Barrier Integrity Cornerstone and affected the objective of providing reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function, and was not the result of any willful violation of NRC requirements. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding screened as very low safety significance (Green) because it did not represent a degradation of the radiological barrier function provided for the control room, auxiliary building, or spent fuel pool, did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere, did not represent an actual open pathway in the physical integrity of reactor containment, and did not involve an actual reduction in function of hydrogen igniters in the reactor containment.

This finding has a cross-cutting aspect in the area of Human Performance, Work Practices Component, because Exelon did not define and effectively communicate expectations regarding procedural compliance and personnel did not follow procedures.

Specifically, Exelon did not comply with procedure CC-AA-102, Design Input and Configuration Change Impact Screening, to evaluate the design temperature of the newly installed check valve to ensure that all affected systems could perform their design basis functions. (IMC 0305, Aspect H.4 (b))

Enforcement:

10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that measures be established for verifying or checking the adequacy of design such as by the performance of design reviews, by the use of alternate or simplified calculation methods, or by the performance of a suitable testing program. Contrary to the above, Exelon modified and replaced the RBCCW containment isolation check valve in November of 2008, and did not properly assess the adequacy of the design temperature of the new valve. Because this finding was of very low safety significance and was entered into the corrective action program as IR 917125, this violation is being treated as an NCV, consistent with section VI.A.1 of the NRC Enforcement Policy.

(NCV 05000219/2009007-02, Inadequate Design Control for RBCCW Containment Isolation Valve Modification)

.2.2 In-Service Inspection Reclassification of the Service Water System

a. Inspection Scope

The team reviewed a modification (08-00041) that reclassified the service water (SW)system from an ASME Section XI Class 3 system to a non-Class system. The modification was implemented to provide future relief from ASME Code leak monitoring requirements and to allow flexibility in resources.

The team assessed whether the modification was consistent with assumptions in the design and licensing bases. The review included verifying drawings and procedures were updated to incorporate the modification. The team discussed the reclassification modification and the applicability of NRC Regulatory Guide (RG) 1.26, Quality Group Classifications and Standards for Water, Steam, and Radioactive Waste Containing Components of Nuclear Power Plants, to the SW system with engineering staff.

Additionally, the 10 CFR 50.59 screen associated with this modification was reviewed as described in section 1R17.1 of this report. The documents reviewed are listed in the attachment.

b. Findings

Introduction:

An unresolved item (URI) was identified because additional NRC review and evaluation is needed to determine the applicability of NRC RG 1.26 to the SW system, and therefore, the acceptability of modification 08-00041. The team plans to request additional guidance and clarification from NRCs Office of Nuclear Reactor Regulation (NRR).

Description:

The SW system is classified as a non-safety related system at Oyster Creek. The SW system cools the reactor building closed cooling water (RBCCW)system, which in turn, cools the shutdown cooling (SDC) and spent fuel pool cooling systems. Exelon stated that the function to remove residual heat from the reactor, while normally performed with the SDC system, can be accomplished using safety related equipment. Specifically, the electromatic relief valves, the core spray system and the containment spray/emergency service water systems are all safety related systems that can perform this function, and the use of these systems is documented in a station procedure. The function to remove residual heat from the spent fuel storage pool is accomplished by the fuel pool and augmented fuel pool cooling systems. The RBCCW, SDC and spent fuel pool cooling systems are also classified as non-safety related systems.

Over the last several years, Exelon has experienced underground leakage events from SW system piping. In response, they replaced sections of the affected underground piping. To date, all repair and testing of the piping has been performed in accordance with the ASME Code due to the ASME Class 3 designation of the SW piping. As documented in Exelons ISI Program Plan, the SW system piping had been classified as an ASME Class 3 system using NRC RG 1.26 as guidance. The purpose of this classification effort was to identify those systems subject to In-service Inspection (ISI)and In-service Testing (IST) requirements. As stated in Section 1.8 of the UFSAR, a cross-reference with design criteria (including NRC RG 1.26), the Systematic Evaluation Program, and the NRC Safety Evaluation for Oyster Creek (NUREG 1382) was conducted. The safety evaluation documents Exelons conformance to NRC RG 1.26.

In January, 2008, Exelon implemented modification 08-00041 to reclassify the SW system from an ASME Section XI Class 3 system to a non-Class system, concluding that the SW system piping classification (early 1970s) was overly conservative. The declassification allows removal of the SW system from the ISI and IST programs and also allows non-ASME code repairs.

The team questioned Exelon on the declassification of the SW system piping due to the functions provided by the SW system, such as removal of residual heat from the reactor and from the spent fuel storage pool via the RBCCW system, with the SDC system and spent fuel pool cooling systems, respectively. The team referenced NRC RG 1.26 and the UFSAR to assess whether the SW system was an important to safety system and therefore, should not have been declassified to a non-Class designation. In particular, Section C.2, Quality Group C, of NRC RG 1.26 states, in part, that cooling water or portions of those systems important to safety that are designed for residual heat removal from the reactor and from the spent fuel storage pool correspond to ASME Class 3 components.

In response to the teams questions, Exelon stated, based on the introduction section of NRC RG 1.26, that the guide is only applicable to safety related structures, systems, and components; and therefore, does not apply to the SW system because it is not safety related. Specifically, the 10 CFR 50.59 Screen for modification 08-00041 stated that NRC RG 1.26 specifically addresses nuclear safety related components; and the SW system is not safety related and does not perform any functions that are important to safety, therefore it is not Quality Group A, B, or C; therefore, it is not necessary that the SW system be within the ASME Section XI ISI Program. In addition, the team noted that in the safety evaluation screen associated with modification 08-00041, Exelon stated that important to safety is analogous to nuclear safety related.

This issue will be opened as a URI in order to determine the applicability of NRC RG 1.26 for classifying the Oyster Creek SW system, and the acceptability of modification 08-00041. (URI 05000219/2009007-03, Declassification of the Service Water System from ASME Class 3 to a non-Class Designation)

.2.3 Isolation Condenser Level Indication for Fire

a. Inspection Scope

The team reviewed a modification (03-00670) that revised the circuitry for the B isolation condenser shell side level instrumentation. The purpose of the change was to prevent a loss of control room indication due to cable damage that could result from a fire in the A/B battery room or the A 480 Volt room. This issue was identified during an update of the fire safe shutdown analysis in 2002. Interim procedure changes had been implemented at the time the issue was identified to utilize an instrument outside of the control room to maintain safe shutdown capability until this modification was implemented.

The team reviewed the modification to ensure it was consistent with the design and licensing bases, including the 10 CFR 50, Appendix R (Fire Protection System) fire safe shutdown analysis. Post-modification testing results and updates to affected procedures were also reviewed. Additionally, the team reviewed the 10 CFR 50.59 screen associated with this modification. The documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified.

.2.4 Modification of Condensate Storage Tank Isolation Valve V-2-90

a. Inspection Scope

The team reviewed a modification (08-00505) that converted the manual operation of V-2-90, which isolates the condensate storage tank (CST) from the condenser hotwell, to an air-operated valve (AOV). The modification was implemented to automatically close V-2-90 upon loss of instrument air during a station blackout (SBO) or fire event to maintain CST inventory by isolating the drain path of the CST to the condenser hotwell.

The CST inventory is needed to support reactor shutdown and to recover from an SBO or fire event. The teams review was performed to verify that the design bases, licensing bases and performance capability of the condensate and transfer system had not been degraded by the modification. Additionally, the 10 CFR 50.59 screen associated with this modification was reviewed as described in section 1R17.1 of this report.

The team verified drawings and procedures were properly updated. Additionally, post-modification, AOV diagnostic, and stroke-time testing data was reviewed to verify the operability of the valve. The team performed a walkdown of the AOV to verify no abnormal conditions existed. Finally, the team discussed the modification and design basis with design engineers to assess the adequacy of the modification. The documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified.

.2.5 Emergency Diesel Generator Starting Battery Calculation

a. Inspection Scope

The team reviewed a modification (06-00010) that evaluated the minimum operating temperature for the emergency diesel generator (EDG) starting batteries. Exelon previously identified that calculation C-1302-741-5350-009, Oyster Creek EDG Battery Sizing Calculation, Rev. 0, was based on a minimum ambient temperature of 70°F. Data collected during weekly battery inspections was generally above 70°F. However, lower temperatures were noted on several occasions and the minimum recorded value was noted to be 55°F. An initial assessment by Exelon concluded that the battery would function as low as 40°F and the need to update the calculation was being tracked in corrective action program document CAP 2004-0315. In addition, the team reviewed the 10 CFR 50.59 screen associated with this modification.

The team assessed whether the modification was consistent with assumptions in the design and licensing bases. The team reviewed the associated revision to the calculation and discussed the issue with the responsible design engineer. The team also reviewed affected plant procedures to ensure they had been updated to reflect the revised minimum operating temperature for the batteries. A walkdown of the EDG enclosures was also performed. The documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified.

.2.6 Main Feedwater Pump Trip Logic Modification

a. Inspection Scope

The team reviewed the modification (07-00786) that revised the main feedwater pump trip circuitry. The original logic associated with the low bearing oil pressure and low pump suction pressure protection utilized a de-energize to trip design. This design increased the potential for an inadvertent trip due to an individual component or fuse failure. The inadvertent pump trip could result in an unnecessary plant transient and associated challenge to plant systems and operators.

The team reviewed the modification to verify that the design bases, licensing bases and performance capability of the plant systems or components had not been degraded by the logic changes. The team also reviewed the adequacy of the post-modification test and verified that affected drawings had been properly revised. Finally, the team reviewed the 10 CFR 50.59 screen associated with this modification. The documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified.

.2.7 Average Power Range Monitor Input Logic Modification

a. Inspection Scope

The team reviewed modification 04-00575 that removed the local power range monitor (LPRM) downscale condition from the average power range monitor (APRM) downscale circuit. Previously, an automatic rod block signal was initiated when any LPRM associated with an APRM was in a downscale condition (even if the APRM was above its downscale setpoint). The modification was implemented to eliminate the need to clear unnecessary rod blocks during a plant startup. The modification did not disable the ARPM downscale rod blocks, nor did it adversely affect accident analyses or nuclear instrumentation operability requirements.

The teams review was performed to verify that the design bases, licensing bases and performance capability of the nuclear instrumentation system had not been degraded by the modification. Additionally, the associated 10 CFR 50.59 screen and safety evaluation were reviewed as described in section 1R17.1 of this report. The team reviewed drawings and other documents to verify that they were properly updated. The team also interviewed engineering staff and reviewed a technical evaluation associated with the modification to determine if the APRMs would function in accordance with the design assumptions. The documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified.

.2.8 New Emergency Service Water and Service Water Underground Pipe

a. Inspection Scope

The team reviewed modification 05-00341 that installed approximately 150 feet of underground piping in both the emergency service water (ESW) and service water (SW)systems. The new piping replaced the existing piping, which had experienced generic degradation of both the internal and external surface coatings. The coating failures have allowed salt water to come in contact with the carbon steel pipe wall and result in pitting corrosion. Groundwater had come in contact with the exterior pipe wall and result in pitting and galvanic corrosion.

The teams review was performed to verify that the design bases, licensing bases and performance capability of the ESW and SW systems had not been degraded by the modification. Additionally, the team reviewed the 10 CFR 50.59 screen associated with this modification. The team reviewed drawings and other documents to verify that they were properly updated. The team also interviewed engineering staff and reviewed technical evaluations associated with the modification to determine if the ESW and SW systems would function in accordance with the design assumptions. The documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified.

.2.9 Replacement of Service Water Pump P-3-1A

a. Inspection Scope

The team reviewed modification 08-00914 that provided for the installation of a new service water (SW) pump that has a different bearing lubrication design. The modification was implemented after the existing pump had developed a seawater leak into the upper bearing housing. The new pump incorporated a pump shaft seal design without the need for packing or an upper bearing oiler. The new SW pump uses process flow for bearing lubrication.

The teams review was performed to verify that the design and licensing bases and performance capability of the service water system had not been degraded by the modification. Additionally, the team reviewed the 10 CFR 50.59 screen associated with this modification. The team reviewed pump performance parameters, such as net positive suction head, pump flowrate, and electrical needs, to verify that system performance requirements were not adversely affected by the modification. The team also interviewed engineering staff and conducted a walkdown of the installed pump to determine if the material condition and performance of the SW system was acceptable and in accordance with design assumptions. The documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified.

.2.10 Demineralized Water Make-Up to the Isolation Condensers

a. Inspection Scope

The team reviewed a modification (07-00947) that changed the demineralized water transfer system to provide a preferred make-up source used for maintaining the isolation condensers in the standby condition. The modification was implemented to minimize the use of the condensate storage tank for isolation condenser make-up. The demineralized water transfer system takes its clean water supply from an new demineralized water storage tank, and it connects upstream of the isolation condenser shell side inlet valves.

The teams review was performed to verify that the design bases, licensing bases, and performance capability of the isolation condenser system had not been degraded by the modification. Additionally, the 10 CFR 50.59 screen and safety evaluation associated with this modification were reviewed as described in section 1R17.1 of this report.

As part of this review, the team assessed the tie-in connection to the isolation condenser make-up to determine if the connection maintained the integrity of the isolation condenser piping code ASME Class 3 requirements. As the condensate storage tank remained the credited make-up water supply for the isolation condensers, the team reviewed the modified condensate system connection to ensure that, during events where the demineralized water system is unable to provide the make-up water source, the condensate system would still be able to perform its design function properly. The team interviewed system engineers and reviewed valve testing to ensure credited valves would operate as required. The documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified.

.2.11 Isolation Condenser Alternate Shell Level Indication

a. Inspection Scope

The team reviewed a modification (07-00154) that installed an alternate detector for each isolation condenser (IC) for shell water level indication. This modification supports long term IC shell level indication during a station blackout (SBO) event when control room IC level indication becomes unavailable. The modification provides operators the ability to control IC level to ensure adequate control of reactor parameters during long term SBO scenarios. The alternate level indicators are maintained in service during all modes of reactor operation.

The team conducted a review to verify that the design bases, licensing bases and performance capability of the IC and connected systems had not been degraded by the modification. The 10 CFR 50.59 screen associated with this modification was reviewed as described in section 1R17.1 of this report. The team verified that drawings and procedures were properly updated to incorporate the level indicators. Additionally, post-modification testing and calibration data was reviewed to verify proper operation of the level indicators. The team assessed the seismic classification of the detectors and associated piping to ensure operability after a seismic event. The team performed a walkdown of the IC level indicators to assess operation while in service and verify any abnormal conditions. Also during the walkdown, the team compared local IC level with control room IC level to ensure consistency. Finally, the team discussed the modification and design basis with reactor engineers to assess the adequacy of the modification. The documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified.

.2.12 Installation of Current Sensing Devices

a. Inspection Scope

The team reviewed a modification (05-00319) to install current sensing devices that provide indication of current flow through selected solenoid-operated valves (SOV). A pair of pilot SOVs is used to operate each of the main steam isolation valves (MSIV), the backup scram valves and the drywell instrument air isolation valves. If one of the pilot SOVs were to fail and was undetected, routine testing that operates the other SOV in the pair could result in operation of the associated main valve (i.e. MSIV) and result in an inadvertent plant trip or transient.

Since the pilot SOVs did not have any direct indication of valve position, the addition of the current sensing devices provided for a more positive verification of an SOV position prior to operation of the second SOV of a pair during testing. Current flow to an SOV is indicated by illumination of a light emitting diode (LED) installed for each of the SOVs.

The team conducted a review to verify that the design bases, licensing bases and performance capability of the affected components had not been degraded by the installation of the current sensing devices. The team reviewed the associated design documents and discussed the modification with the responsible engineers and plant operators. The team also reviewed the 10 CFR 50.59 screen associated with this modification. A walkdown of the associated indication for MSIV pilot SOVs was also performed. Updates to affected plant procedures were verified by the team. The documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems (IP 71152)

a. Inspection Scope

The team reviewed a sample of condition reports associated with 10 CFR 50.59 and plant modification issues to determine whether Exelon was appropriately identifying, characterizing, and correcting problems associated with these areas and whether the planned or completed corrective actions were appropriate. The condition reports reviewed are listed in the attachment.

b. Findings

No findings of significance were identified.

4OA6 Meetings, Including Exit

The team presented the inspection results to Mr. T. Rausch, Site Vice-President, and other members of Exelon's staff at an exit meeting on May 15, 2009. The team verified that this report does not contain proprietary information.

ATTACHMENT

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

M. Carlson Design Engineer

P. De Design Engineer

S. Dupont Regulatory Assurance

T. Fenton Design Engineer

M. Floyd Design Engineer

G. Harttraft ISI Program Manager

S. Markos Engineering Design

P. Procacci Design Engineer

R. Pruthi Design Engineer

H. Ray Senior Manager Design Engineering

T. Ruggiero Mechanical Engineering Manager

S. Schwartz System Manager

Others

R. Pinney, State of New Jersey, Bureau of Nuclear Engineering

LIST OF ITEMS

Opened and Closed

NCV

05000219/2009007-01 Inadequate 10 CFR 50.59 Evaluation for Trunnion Room Door/Secondary Containment Temporary Modification (Section 1R17.1)

NCV

05000219/2009007-02 Inadequate Design Control for RBCCW Containment Isolation Valve Modification (Section 1R17.2.1)

Opened

URI

05000219/2009007-03 Declassification of the Service Water System from ASME Class 3 to a non-Class Designation (Section 1R17.2.2)