IR 05000266/2003007
| ML042610132 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 02/04/2004 |
| From: | Caldwell J NRC/RGN-II |
| To: | Vanmiddlesworth G Nuclear Management Co |
| References | |
| EA-02-031, EA-03-057, EA-03-059, EA-03-181, FOIA/PA-2004-0277 IR-03-007 | |
| Download: ML042610132 (16) | |
Text
February 4, 2004 EA-02-031 EA-03-057 EA-03-059 EA-03-181 Mr. Gary Van Middlesworth Acting Site Vice-President Point Beach Nuclear Plant Nuclear Management Company, LLC 6610 Nuclear Road Two Rivers, WI 54241-9516 SUBJECT: POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 95003 SUPPLEMENTAL INSPECTION NRC INSPECTION REPORT 05000266/2003007; 05000301/2003007
Dear Mr. Van Middlesworth:
On December 16, 2003, the results of a three-phase supplemental inspection conducted at the Point Beach Nuclear Plant in accordance with NRC Inspection Procedure (IP) 95003,
"Supplemental Inspection for Repetitive Degraded Cornerstones, Multiple Degraded Cornerstones, Multiple Yellow Inputs, or One Red Input," were discussed with Messrs. John Paul Cowan, Douglas Cooper, and Fred Cayia, and members of the Point Beach staff at a public meeting at the Holiday Inn in Manitowoc, Wisconsin. A summary of the public meeting on December 16, 2003, was documented in a letter to Mr. Cayia, dated December 31, 2003.
The inspection was an examination of activities conducted under your license as they relate to safety and to compliance with the Commission's rules and regulations and with the conditions of your license. Within these areas, the inspection consisted of a selective review of procedures and representative records, observations of activities, and interviews with personnel.
In a letter dated May 9, 2003, we informed Mr. Cayia of our decision to conduct the IP 95003 supplemental inspection. The inspection was conducted to review your corrective actions for the Red inspection finding associated with the auxiliary feedwater and instrument air systems (AFW/IA) and for the inspection finding associated with the potential common mode failure of the AFW pumps because of plugging of the recirculation line pressure reduction orifices. This second finding associated with the AFW system was subsequently determined to be a Red finding for Unit 2, and a Yellow finding for Unit 1. Details of these findings are provided in Inspection Reports (IRs) 50-266/01-17(DRS); 50-301/01-17(DRS), dated April 3, 2002, and 50-266/02-15(DRP); 50-301/02-15(DRP), dated April 2, 2003, and in Final Significance Determination letters dated July 12, 2002, and December 11, 2003. The inspection also assessed your performance in the Reactor Safety Strategic Performance Area, which included detailed inspections of the effectiveness of your corrective action, emergency preparedness, and engineering programs. /f /\
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G. Van Middlesworth -2 -
From our inspection, we concluded that your evaluation of the causes of the significant AFW inspection findings was adequate and your proposed corrective actions were reasonable.
Though we concluded that generally your planned corrective actions were adequate to prevent problem recurrence, the implementation of some of the corrective actions was weak as reflected in repeated extension of due dates and the lack of quality in the implementation of identified corrective actions. For example, the IP 95003 inspection was extended for one week because of inconsistent quality of your implementation and documentation of corrective actions which resulted in several corrective actions involving the AFW system not being completed.
The NRC determined that additional review of corrective actions related to the AFW system was warranted to gain assurance that the system was operable. The NRC subsequently concluded that the AFW system was operable; however, extensive inspection effort was required to verify that previous corrective actions adequately addressed historic AFW system performance problems.
Regarding our assessment of your performance in the Reactor Safety Strategic Performance Area, the NRC determined that the plant is being operated in a manner that ensures public safety. However, the NRC also identified several performance issues which warrant increased attention to ensure continued plant safety. Specifically: (1) the quality of your implementation of programs and processes related to the identification and resolution of problems was inconsistent, resulting in inadequate or incomplete corrective action, (2) we identified multiple findings and violations related to emergency preparedness which indicated that Point Beach management and staff did not have a good understanding of license and regulatory requirements, (3) electrical design basis calculations were poorly controlled, and (4) ineffective communication between engineering and operations contributed to the lack of a common understanding of some system design basis and operational practices.
The NRC determined that your performance improvement plan (Excellence Plan) provides an adequate framework for the improvement of station performance. However, the success of this Plan is contingent on the adequate commitment of resources, the timely and quality implementation of the Plan, and the establishment of measures or indicators of successful completion at various stages during Plan implementation, including when all necessary action steps of the action plans in the Excellence Plan and effectiveness reviews for the actions have been completed. Additionally, we determined that the Excellence Plan did not completely address all problem areas. The Plan required changes to ensure that problems associated with implementation of the corrective action and emergency preparedness programs, engineering design basis calculation adequacy, and organizational effectiveness, are adequately addressed to affect and sustain long-term improvement in these areas.
You are requested to respond to this letter by February 13, 2004, and describe the actions that you will take to address the issues raised during this inspection, and your schedule for submission of your revised Excellence Plan. The NRC will review the adequacy of the revised Excellence Plan and its implementation. The NRC will continue to provide increased oversight of activities at Point Beach until you have demonstrated that your corrective actions are lasting and effective. Consistent with Inspection Manual Chapter (IMC) 0305 "Operating Reactor Assessment Program," guidance regarding the oversight of plants in the multiple/repetitive degraded cornerstone column of the Action Matrix, the NRC will continue to assess performance at Point Beach and will consider at each quarterly performance assessment
G. Van Middlesworth -3 -
review the following options: (1) declaring plant performance to be unacceptable in accordance with the guidance in IMC 0305; (2) transferring to the IMC 0350 "Oversight of Operating Reactor Facilities in a Shutdown Condition with Performance Problems" process; and (3) taking additional regulatory actions, as appropriate. Until you have demonstrated lasting and effective corrective actions, Point Beach will remain in column four of the Action Matrix.
During the inspection, an apparent violation of 10 CFR 50.54(q) and 50.47(b) was identified for changes Point Beach made between October 1998 and December 1999 to the previously NRC-approved Emergency Action Level scheme. This apparent violation is described in Section 3.6 of the enclosed inspection report and in a letter sent to Mr. Cayia, dated December 2, 2003. It was also discussed with Point Beach management during a technical debriefing by the emergency preparedness inspectors on August 8, 2003; at the conclusion of the onsite portion of the emergency preparedness phase of the inspection on August 27; during a preliminary exit meeting for all three phases of the IP 95003 inspection on November 17; during a telephone conference on December 1; and during the public final exit meeting for the IP 95003 inspection on December 16. As discussed in the December 2nd letter, this apparent violation is being considered for escalated enforcement action in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy),
NUREG-1 600. The current Enforcement Policy is included on the NRC's website at www.nrc.cov. A predecisional enforcement conference was held on January 13, 2004, in the Region IlIl office. From the information presented at the conference, it was apparent that corrective actions taken, thus far, have been less than fully effective. On January 16, after further discussions between the NRC and NMC representatives about the need to take corrective actions to return to compliance, NMC representatives informed the NRC that the EALs had been changed. A summary of the predecisional enforcement conference was provided to you in a letter dated January 27, 2004. You will be notified by separate correspondence of the results of the NRC's deliberations on the apparent violation and the adequacy of your corrective actions.
In addition to the apparent violation, the NRC identified that your Emergency Plan and implementing procedures did not provide a range of protective action recommendations as required by NRC regulations. The only protective action recommendation that would have been given to State and local officials by your staff in the event of an emergency at Point Beach was evacuation. This issue is being treated as an unresolved item while the NRC evaluates the potential generic implications. We confirmed by direct observation that the Emergency Plan and implementing procedures have been changed since the inspection was completed to provide an appropriate range of recommendations.
Based on the results of this inspection, ten NRC-identified violations of very low safety significance (Green) and one NRC-identified Severity Level IV violation were identified.
Additionally, a licensee-identified violation is listed in Section 5 of this report. These violations are being treated as Non-Cited Violations (NCVs) consistent with Section VL.A of the Enforcement Policy. These NCVs are described in the subject inspection report. If you contest the severity level or significance of these NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U. S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with a
- I G. Van Middlesworth -4 -
copy to the Regional Administrator, U.S. Nuclear Regulatory Commission - Region III, 801 Warrenville Road, Lisle, IL 60532-4351; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Point Beach Nuclear Plant facility.
In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and any responses you provide will be made available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records System (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at httD://www.nrc.cov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RN James L. Caldwell Regional Administrator Docket Nos. 50-266; 50-301 License Nos. DPR-24; DPR-27
Enclosure:
Inspection Report 05000266/2003007; 05000301/2003007 w/Attachment: Supplemental Information
REGION IlIl Docket Nos: 50-266; 50-301 License Nos: DPR-24; DPR-27 Report No: 05000266/2003007; 05000301/2003007 Licensee: Nuclear Management Company, LLC Facility: Point Beach Nuclear Plant Location: 6610 Nuclear Road Two Rivers, WI 54241 Dates: Corrective Action Inspection, July 28 - August 8, 2003 Emergency Preparedness lnspection, August 4-15, 2003 Engineering, Operations, and Maintenance Inspection,.
September 8 - October 3, 2003 Preliminary Exit Meeting, November 17, 2003 Public Final Exit Meeting, December 16, 2003 Personnel: A. Vegel, 95003 Team Leader; Chief, Branch 7, Division of Reactor Projects (DRP), Region IlIl M. Kunowski, 95003 Assistant Team Leader; Project Engineer, Branch 7, DRP, Region lil Corrective Action Inspection L. Kozak, Inspection Leader, Project Engineer, DRP, Branch 6, Region III J. Lenahan, Senior Engineering Inspector, Division of Reactor Safety (DRS), Region II R. M. Morris, Resident Inspector, Point Beach J. Chiloyan, NRC Contractor K. Elsea, NRC Contractor D. Baxley, Administrative Assistant, Office of Nuclear Reactor Regulation (NRR)
Enclosure
(Cover page cont)
Emergency Preparedness Inspection R. Lantz, Inspection Leader; Senior Emergency Preparedness (EP) Inspector, Region IV T. Blount, Senior EP Analyst, NRR R. Kahler, Senior EP Analyst, NRR T. Ploski, Senior EP Inspector, Region Ill R. M. Morris, Resident Inspector, Point Beach Engineering. Ogerations. and Maintenance Inspection A. Vegel, Inspection Leader S. Burgess, Senior Reactor Analyst, Region III L. Kozak, Project Engineer, Branch 6, Region III D. Pelton, Senior Resident Inspector, Vermont Yankee, Region I R. Daley, Reactor Inspector, DRS, Region Ill M. Maymi, Reactor Inspector, DRS, Region 11 C. Baron, NRC Contractor G. Skinner, NRC Contractor S. Billings, Administrative Assistant, NRR P. Krohn, Senior Resident Inspector, Point Beach R. M. Morris, Resident Inspector, Point Beach Approved by: Anton Vegel, Chief Branch 7 Division of Reactor Projects Enclosure
SUMMARY OF FINDINGS IR 05000266/2003-007, 05000301/2003-007; 7/28/2003 - 12/16/2003; Nuclear Management Company, LLC; Point Beach Nuclear Plant, Units 1 and 2. Supplemental inspection 95003 was performed to evaluate corrective actions for a Red inspection finding pertaining to the auxiliary feedwater and instrument air systems, and for a Red inspection finding pertaining to the potential for a common mode failure of the auxiliary feedwater pumps because of the plugging of the recirculation line pressure reduction orifices. The inspection also reviewed the corrective action, emergency preparedness, and engineering programs.
The Nuclear Regulatory Commission (NRC) began the three-phase supplemental inspection on July 28, 2003. At that time, the orifice plugging finding was still considered a preliminary Red finding. However, because the licensee had completed the root cause investigation and had developed and began implementation of corrective actions for the preliminary Red finding before the completion of the inspection, the IP 95003 inspectors reviewed the adequacy of these corrective actions. The Final Significance Determination of the orifice plugging issue was subsequently completed and transmitted to the licensee in a letter dated December 11, 2003.
The NRC concluded that this issue was appropriately characterized as Yellow for Unit land Red for Unit 2. The difference in significance is a result of the longer time that the orifices were installed in Unit 2.
This inspection was conducted in accordance with NRC Supplemental Inspection Procedure 95003, "Inspection for Repetitive Degraded Cornerstones, Multiple Degraded Cornerstones, Multiple Yellow Inputs, or One Red Input." In the first phase of the inspection, the licensee's corrective action program was reviewed, with a focus on problem identification, in general, and implementation of corrective actions for the two AFW issues, in particular.
On August 4, 2003, the second phase of the inspection, a review of the emergency preparedness program, began. An apparent violation was identified during this phase for changes made to the Emergency Action Level scheme that decreased the effectiveness of the Emergency Plan and did not receive prior NRC approval. The licensee was informed of this apparent violation in a letter dated December 2, 2003, and a predecisional enforcement conference to discuss this issue was conducted on January 13, 2004.
On September 8, 2003, the final phase of the inspection began. The focus of this phase was the licensee's engineering program, particularly design engineering, and additional review of implementation of corrective actions for the two AFW issues. Plant operations and maintenance, as they interact with engineering, were also reviewed. This phase of the inspection was extended an extra week because of problems identified by the inspectors with AFW system corrective actions.
The licensee's corrective action program was adequate. Examples of poor implementation continue to exist despite licensee efforts to improve overall implementation. Identified program weaknesses that could contribute to implementation problems included the potential for issues to be categorized and analyzed at too low a level and a weak trending process. Recently implemented program improvements were good initiatives but were not formalized.
1 Enclosure
The overall root and contributing causes for the two AFW Red findings were the lack of understanding of the design, corrective action program weaknesses, and poor operations/engineering interface. And while overall, corrective actions taken for the findings were adequate, several important corrective actions to prevent recurrence had not been adequately implemented.
In emergency preparedness (EP), the inspectors concluded that the licensee's program was adequate. Program challenges and areas needing improvement included EP staff experience level and training, maintenance of EP design bases and understanding of EP regulatory guidance documents. An apparent violation was identified for the failure to maintain a standard emergency action level scheme, and an unresolved item whose significance is greater than Green was identified for a lack of range of protective actions in the Emergency Plan and implementing procedures.
In engineering, the inspectors concluded that the CCW system design and licensing basis were understood and adequately supported by controlled testing and calculations. The 125-volt direct current (VDC) system design basis calculations, however, were poorly controlled and design basis calculations related to several alternating current (AC) systems were poorly understood by engineers. The operability of the electrical systems were verified through calculations by the inspectors. The inspectors determined that communications between operations and engineering staff regarding the understanding of system design and operating practices was not consistently effective.
This report covers a 5-month period of supplemental inspection by NRC contractors and NRC inspectors from all four NRC Regional offices and from Headquarters. Ten Green findings, which were associated with Non-Cited Violations, one Severity Level IV Non-Cited Violation, one apparent violation, and one unresolved item of significance to be determined were identified. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process" (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1 649, "Reactor Oversight Process,"
Revision 3, dated July 2000.
A. NRC-Identified and Self-Revealing Findings Cornerstone: Emergency Preparedness
- Green. The inspectors identified a Non-Cited Violation of emergency planning standard 10 CFR 50.47(b)(2) because the licensee failed to assign onshift responsibilities for reading facility seismic monitors, thereby affecting the ability to timely classify certain seismic emergency events.
This finding is greater than minor because it was associated with a cornerstone attribute and affected the emergency preparedness cornerstone objective to ensure the adequate protection of the public health and safety. This finding is of very low safety 2 Enclosure
significance because it was a degradation in the emergency response organization (ERO) onshift staffing and did not represent a planning standard function failure.
(Section 3.2.b.2)
Severity Level IV. The inspectors identified a Severity Level IV Non-Cited Violation of 10 CFR 50.9 because the licensee failed to provide complete and accurate information in the submittal of information for the emergency response organization (ERO)
performance indicator (PI). Twenty-three onshift communicators should have been tracked and reported in the ERO PI, but were not. The licensee has subsequently submitted corrected PI data to the NRC.
This issue is greater than minor because it caused the PI to cross the Green-to-White threshold for the 3r quarter of 2001. Because this issue affected the NRC's ability to perform its regulatory function, it was evaluated with the traditional enforcement process. (Section 3.2.b.3)
L' Green. The inspectors identified a Non-Cited Violation of emergency planning standard 10 CFR 50.47(b)(1 6) because the licensee failed to develop and implement an emergency planning staff training program to ensure that emergency planners were properly trained.
This finding is greater than minor because it was associated with a cornerstone attribute and affected the emergency preparedness cornerstone objective to ensure the adequate protection of the public health and safety. This finding is of very low safety significance because lack of a staff training program presented a potential degrading condition for the level of qualification and proficiency of the emergency preparedness staff, but did not represent a failure of the planning standard function. (Section 3.5)
/
e ,/ To Be Determined. The inspectors identified an unresolved item for the lack of a range of protective actions in the Emergency Plan and implementing procedures. This issue is being treated as an unresolved item while the NRC evaluates the industry-wide generic implications of this issue. Since the identification of the issue by the inspectors, the licensee has revised the Emergency Plan and implementing procedures to include the appropriate range of protective actions. (Section 3.6.b.1)
- .:."To Be Determined. The inspectors identified an apparent violation of 10 CFR 50.54(q),
associated with emergency planning standard 10 CFR 50.47(b)(4), which will be subject to the NRC traditional enforcement process not the revised Reactor Oversight Process.
Specifically, the licensee failed to maintain a standard scheme of emergency action levels (EALs). Eight EALs were changed in 1998 and 1999. The changes decreased the effectiveness of the Emergency Plan in that emergency conditions that would have resulted in classifications at the General Emergency (GE), Alert, and Notification of Unusual Event (NOUE) levels would result in a lesser classification under the current EAL scheme. Approval of the NRC was not obtained prior to the changes being made.
Since the identification of the issue by the inspectors, the licensee has revised the eight EALs to be equivalent with those approved by the NRC in 1984. (Section 3.6.b.2)
3 Enclosure
Green. The inspectors identified a Non-Cited Violation of emergency planning standard 10 CFR 50.47(b)(4) because the licensee failed to properly calibrate the facility seismic monitors to ensure they were capable of supporting implementation of a Notice of Unusual Event EAL.
This finding is greater than minor because it was associated with a cornerstone attribute and affected the emergency preparedness cornerstone objective to ensure the adequate protection of the public health and safety. This finding is of very low safety significance because a Notice of Unusual Event could still be declared based on ground shaking. (Section 3.6.b.3)
Cornerstone: Mitigating Systems
!* Green. The inspectors identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion IlIl, "Design Control," because Technical Specification Surveillance Requirement 3.8.4.6 for testing the safety-related battery chargers was non-conservative in relation to the design basis calculation for battery charger sizing.
This finding is greater than minor because it affected the mitigating systems cornerstone objective. This finding is of very low safety significance because it was a design deficiency that did not result in the loss of function. (Section 4.1.1.1 .b.1)
- Green. The inspectors identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion IlIl, "Design Control," because the licensee failed to maintain the 125-volt direct current (VDC) voltage drop calculations accurate and up-to-date.
This finding is greater than minor because it affected the mitigating systems cornerstone objective. This finding is of very low safety significance because it was a design deficiency that did not result in the loss of function. (Section 4.1.1.1.b.2)
- V Green. The inspectors identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." Specifically, the licensee failed to implement timely corrective action (for over 5 years) for safety-related electrical equipment in the primary auxiliary building (PAB) that was not environmentally qualified, a condition adverse to quality.
This finding is greater than minor because if left uncorrected, the finding would become a more significant safety concern and have adverse effects on the capability to prevent or mitigate the consequences of accidents. The finding is of very low safety significance because it was a design deficiency that did not result in the loss of function.
(Section 4.1.2.b.2.1)
- Green. The inspectors identified a Non-Cited Violation of 10 CFR 50.49(f). Specifically, v, the licensee identified equipment important to safety located in the primary auxiliary building that would be susceptible to a harsh environment during a postulated high-energy line break but failed to environmentally qualify that equipment.
4 Enclosure
This finding is greater than minor because if left uncorrected, the finding would become a more significant safety concern and have adverse effects on the capability to prevent or mitigate the consequences of accidents. The finding is of very low safety significance because it was a design deficiency that did not result in the loss of function.
(Section 4.1.2.b.2.2)
- Green. The inspectors identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." Specifically, the licensee failed to include appropriate quantitative setpoint values for the minimum low head safety injection "A" train flow in plant emergency operating procedures (EOPs).
This finding is greater than minor because it could have affected the mitigating cornerstone objective of ensuring the availability of the low head safety injection system when required to respond to the initiating event. The finding is of very low safety significance because it did not represent an actual loss of safety function.
(Section 4.1.5)
Green. The inspectors identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion Xl, "Test Control," because the licensee failed to include in the inservice testing program manual component cooling water (CCW) valves that were required to perform a safety function.
This finding is greater than minor because it could have affected the mitigating cornerstone objective of ensuring the availability of the COW or residual heat removal (RHR) systems when required to respond to the initiating event. The finding is of very low safety significance because it did not represent an actual loss of safety function.
(Section 4.1.3.2)
- Green. The inspectors identified a Non-Cited Violation of 10 CFR Part 50, Appendix R, Section III.L.1 .c. Specifically, the licensee failed to ensure, without the need for "hot standby repairs," adequate control air to the speed controllers for the charging pumps during a postulated fire requiring an alternative shutdown method.
This finding is greater than minor because the finding would become a more significant safety concern if left uncorrected. The finding is of very low safety significance because it is likely that the licensee would have been successful in completing the repairs and allowing the plant to be maintained in hot standby until cold shutdown could be achieved. (Section 4.1.7)
B. Licensee-identified Violation A violation of very low safety significance, which was identified by the licensee, has been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensee's corrective action program. The violation and the licensee's corrective action tracking number are listed in Section 5 of this report.
5 Enclosure
however, a plant-specific LBB analysis (WCAP-14439, Revision 0) was subsequently performed for PBNP Units 1 and 2 by Westinghouse. This analysis included parameters associated with the Unit 2 steam generator replacement, partial power uprate conditions, and a 40-year operating period.
The plant specific analysis was not submitted to the NRC for review and approval as required by 10 CFR Part 50, Appendix A, Criterion 4, "Environmental and Dynamic Effects Design Basis." This plant specific analysis was credited in SER 96-084-02,
"PBNP Unit 2 Replacement Steam Generator Design,' and SE 2001-0007, "Component Cooling System Closed Loop Inside Containment."
CAP033580 evaluated the condition and, on June 17, 2003, incorrectly determined that NRC review and approval was not required for the plant-specific LBB analysis. On August 1, 2003, the licensee reevaluated the condition and determined that NRC review was required. As a result, OD OPR000072 was performed, and concluded that the reactor coolant systems of both Units were operable, but non-conforming. CAP034513 was initiated on August 1 to address the failure to recognize the NRC submittal requirement during the June 17, 2003, screening. CAP034513 concluded that this was a knowledge-based error, apparently due to reliance on a dated piece of correspondence rather than on the current regulatory requirements.
Based on discussions with PBNP personnel, the inspectors determined that the plant-specific LBB analysis (WCAP-14439, Revision 0) had not been submitted to the NRC at the time of the inspection. Instead, the licensee intended to submit Revision 1. The revised analysis incorporated parameters associated with the proposed full power uprate and a 60-year operating period. The licensee stated that it was working with Westinghouse to incorporate corrections into WCAP-14439, Revision 1, prior to the submittal. These activities were being tracked by OPR000072.
The licensee-identified failure to submit the plant-specific LBB analysis to the NRC for review and approval as required by 10 CFR Part 50, Appendix A, Criterion 4, constitutes a violation of minor significance, not subject to enforcement action in accordance with Section IV of NRC's Enforcement Policy. The licensee documented the failure to submit the analysis in CAP034513.
4.1.7 Miscellaneous Issue - Appendix R Concern for Speed Controllers for the Charging Pumps a. Inspection Scope While reviewing corrective actions for the AFW/IA Red finding, the inspectors identified a finding related to Appendix R and the speed controllers for the three per Unit positive displacement charging pumps.
b. Observations and Findings Introduction: The inspectors identified a Green, Non-Cited Violation of 10 CFR Part 50, Appendix R, lIl.L.1.c, for not ensuring adequate control air to the speed controllers for 70 Enclosure
the charging pumps during a postulated Appendix R fire event requiring an alternative shutdown method.
Description: For certain postulated fire scenarios that required the use of an alternative shutdown method, instrument air to the charging pump speed controllers would be lost and the controllers would then fail to slow speed. For certain alternative shutdown scenarios, where only one charging pump was available, the pump was required to be operating in fast speed to ensure adequate makeup to the RCS.
A 12-pack of nitrogen bottles that were hard-piped to the speed controller instrument air header served as a backup. However, the capacity of the backup was 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. As compensatory measures, the licensee staged a dedicated air compressor, electrical cables, hoses, and fittings so that if the postulated fire were to occur, operators could assemble the equipment and supply air to the controllers. These actions were necessary for the plant to maintain hot standby conditions.
The inspectors concluded that this compensatory measure was a repair activity required to maintain hot standby conditions, that is "hot standby repairs," and as such were not allowed by 10 CFR Part 50, Appendix R.
Analysis: This failure to ensure that control air to the speed controllers for the charging pumps would be maintained during a postulated Appendix R fire event is a violation of 10 CFR Part 50, Appendix R, Section IlI.L.1.c. This finding is more than minor because if left uncorrected, the finding would become a more significant safety concern. Without control air available for the charging pump speed controllers, operators would not have been able to maintain adequate makeup to the RCS. However, the inspectors determined that even though the operators would have been challenged by this type of hot shutdown repair, the operators most likely could have performed the necessary actions to assemble the staged equipment (air compressor, electrical cables, hoses, and fittings) so that the temporary air compressor could supply the required control air to the charging pump speed controllers.
This violation had no significant impact on the cornerstone because it did not involve the impairment or degradation of a fire protection feature. Therefore, this finding was considered to be a Green finding.
Enforcement: 10 CFR Part 50, Appendix R, Section lll.L.1.c states, in part, that alternative shutdown capability shall be able to achieve and maintain hot standby conditions for a pressurized water reactor (such as Point Beach).
Contrary to this, the licensee could not maintain hot standby conditions, because for an alternative shutdown scenario, control air to the charging pump speed controllers could have been lost. This could have resulted in insufficient makeup to the RbS. The results of this violation were determined to be of very low safety significance. Therefore, because this violation of 10 CFR Part 50, Appendix R, Section IlI.L.1.c, was captured in the licensee's corrective action program (CAP050456), it was considered a Non-Cited Violation (NCV 50-266/03-07-13; 50-301/03-07-13) consistent with Section VI.A.1 of the NRC Enforcement Policy.
71 Enclosure
CAP050173, NRC Required Programs May Not Fully Implement Commitments, September 15, 2003 CAP050192, EOP Setpoint Basis Document for V.14 and V.35 Is Not Accurate, September 15, 2003 CAP050229, Relief Set Point Changed on 1&2CC-736A&B Without.Considering MOV Design Basis, September 17, 2003 CAP050258, Multiple Drawing Errors Discovered During Creation of Calculation 2003-0006, September 17, 2003 CAP050276, Lot Number for Spare CCW Motor in Procedure RMP 9006-4 Incorrect, September 18, 2003 CAP050284, Appendix R Discussion in Section 2.2.3 of DBD-02 Is Not Accurate, September 18, 2003 CAP050340, Determine Safety Function of Component Cooling Water System Manual Valves, September 22, 2003 CAP050350, Perform New Review of AFW System to Support Recirc AOV Safety Function Upgrade, September 23, 2003 CAP050367, 50.59 Screening for CC Relief Valve Setpoint Change Inadequately Documented, September 23, 2003 CAP050388, EOPSTPT L.3 and L.13 Existing Values Are Non-Conservative, September 24, 2003 CAP050398, Remove Reference to Second Spare CCW Pump Motor for Appendix R Use, September 24, 2003 CAP050405, CCW Licensing Basis, September 24, 2003 CAP050420, Procedure Feedback Request Concern, September 25, 2003 CAP050429, EOP Setpoint Calculations Recommendation, September 25, 2003 CAP050456, Establishment of Appendix R Backup Air For Charging Pumps May Be Hot Shutdown Repair, September 26, 2003 CAP050499, Emergency Procedure Conflicts Not Yet Corrected, September 29, 2003 CAP050502, Local Manual Operation of MOVs and Manual Valves Not in IST Program, September 29, 2003 CAP050509, Incorrect Closure of CAP032355, September 29, 2003 37 Attachment