IR 05000369/1999005
| ML20211K891 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 08/26/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20211K887 | List: |
| References | |
| 50-369-99-05, 50-370-99-05, NUDOCS 9909080094 | |
| Download: ML20211K891 (18) | |
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U.S. NUCLEAR REGULATORY COMMISSION l
REGION 11 -
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j Dobket Nost 50-369,50-37'O License Nos:
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. Report No:
'50-369/99-05,50-370/99-05
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Licensee:
Duke Energy Corporation Facility:
. McGuire Nuclear Station, Units 1 and 2 Location:
12700 Hagers Ferry Road Huntersville, NC 28078 Dates:
June 20 - July 31,1999 inspectors:
S. Shaeffer, Senior Resident inspector M. Franovich, Resident inspector J. Blake, Senior Project Manager (Sections R2.1 and R2.2)
S. Vias, Senior Reactor inspector (Sections R2.1 and R2.2)
Approved by:
C. Ogle, Chief, Projects Branch 1 Division of Reactor Projects
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9909080094 990826
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PDR ADOCK 05000369
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EXECUTIVE SUMMARY McGuire Nuclear Station, Units 1 and 2 NRC Inspection Report 50-369/99-05,50-370/99-05
- This integrated inspection included aspects of licensee operations, maintenance, engineering,
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and plant support. The report covered a six-week period of resident inspections and also included regionalinspections in the area of the independent spent fuel storage facility, l
Operations _
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Operator response to an inadvertent Unit 2 reactor trip was in accordance with
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applicable response procedures. Operator oversight of unit parameters was well controlled during the reactor trip recovery and during the subsequent unit restart.
(Section O2.2)
l Specialized training for time critical operator actions was well researched, provided
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excellent focus on safety, and provided a redundant method to verify Updated Final
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Safety Analysis Report ass' mptions, risk significant operator actions, and procedural u
implementation. Overall, the augmented training was considered a strength. (Section 07.1)
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Maintenance l
A non-cited violation was identified conceming an inadequate reactor trip beaker
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l maintenance procedure which resulted in a Unit 2 reactor trip. Inappropriate personnel actions also contributed to the root cause of the event. (Section O2.2)
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Comprehensive corrective actions for flood protection deficiencies were being
implemented in a timely manner to reduce plant vulnerability to an internal flood. Flood protection equipment and instrumentation were operable and were maintained in good condition. Plant procedures for coping with an internal flood were being improved through the development of abnormal procedures for flooding. Operators demonstrated adequate knowledge of flood indications and associated flood protection.a_ctions.
(Section M2.1),
l Visible material condition of potential internal plant flood sources (i.e., fire protection
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piping, service water system piping)in the auxiliary building was generally adequate, with the exception of uninsulated portions of service water pipe in the Unit 1 auxiliary feedwater pump room which exhibited substantial exterior corrosion. Recent inservice l
inspection of large-bore, non-isolable service water pipes (in the auxiliary building) to the standby nuclear service water pond indicated deep pitting on the interior pipe wall.
Corrective actions were initiated to address the degraded pipe conditions. (Section M2.1)
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The licensee's performance of a specialinspection for ice accumulation on the Unit 1 ice condenser lower inlet doors was well planned and implemented. The conservative inspection was based on response to associated industry events and exceeded the requirements of routine Technical Specification surveillances. (Section M2.2)
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Enaineerina A post-trip review of a Unit 2 reactor trip event was well organized, accurately identified
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the apparent cause of the trip, and resolved other equipment anomalies prior to restart of the unit. (Section 02.2).
The licensee's response to NRC Generic Letter 89-13, Action V, was less than
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adequate, in that, no specific review of plant procedures was performed to i.) duce human errors in operation of the service water system. This contributed, in part, to the
. flood of the Unit 1 auxiliary feedwate.r pump room during draining of the Unit 2A service water system in March 1999. (Section M2.1)
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The licensee's review of NRC Generic Letter 93-06 was weak, in that, it did not identify or evaluate the potential risk of a non-seismic hydrogen line routed through the Unit 1
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safety-related refueling water storage tank trench. (Section F8.1)
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Plant Support A non-cited violation was identified by the NRC for failing to meet the requirements of 10
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CFR 50.48 regarding the routing of non-seismic hydrogen piping near the refueling water storage tank and associated piping. (Section F8.1)
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The licensee performed adequate installation of concrete rebar and formwork, as well as
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adequately controlled placement of the concrete for the independent spent fuel storage installation pads. The construction records and related documents were adequate.
(Section R2.1)
' The licensee performed adequate identification and evaluation of the transportation
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pathway for the specially designed fuel cask transporter over buried piping and components inside the protected area. (Section R2.2)
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Report Details Summary of Plant Status Unit 1 Unit 1 operated at approximately 100 percent of licensed thermal power throughout the inspection period.
Unit 2 Unit 2 began the inspection period at approximately 100 percent of licensed thermal power. On.
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July 15,1999, a reactor trip occurred due to an inadvertent turbine trip signal being generated during testing of reactor trip breaker circuitry following replacement of Unit 2 reactor trip breaker 2RTA.. Following review of the apparent cause of the trip and the response of other plant equipment, the unit was restarted on July 16,1999. The unit operated at approximately 100 l
percent power for the remainder of the inspection period.
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1. Operations
Conduct of Operations 01.1 General Comments (71707)
Using inspection Procedure 71707, the inspectors conducted frequent reviews of ongoing plant operations. In general, the conduct of operations was professional and safety-conscious. Specific events and noteworthy observations are detailed in the l
sections which follow, including response to a Unit 2 reactor trip (see Section 02.2).
01.2 10 CFR 50.72 and Other Reauired Notifications a.
Inspection Scope (71707)
During the inspection period, the licensee made the following notification tp the NRC as required by 10 CFR 50.72. The inspectors reviewed the event for impact on the operational status of the facility and equipment.
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Observations and Findinos On July 15,1999, the licensee made a report to the NRC in accordance with 10 CFR 50.72 (b)(2)(ii) concerning a Unit 2 reactor trip and emergency safeguard features actuation from 100 percent power. Along with the reactor trip, the report also indicated that the 2A motor
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driven auxiliary feedwater (AFW) pump did not automatically start and was manually started l
by the operators in accordance with applicable reactor trip response procedures. It was later j
determined that the 2A pump not starting automatically was expected due to ongoing solid
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l state protection system testing. This event is further discussed in Section O2.2.
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. Conclusions
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The inspectors concluded that the licensee reported the above event in accordance with the requirements of 10 CFR 50.72.
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O2 Operational Status of Facilities and Equipment O2.2 Unit 2 Reactor Trio Durina Reactor Trio Breaker Replacement Testina a.
Inspection Scope (71707. 61726. and 62707)
The inspectors reviewed the circumstances surrounding an inadvertent Unit 2 reactor trip which occurred on July 15,1999. The inspectors responded to the control room, observed the operators' response to the reactor trip, reviewed the performance of plant equipment
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during the event, and observed the subsequent restart of the unit. The inspectors also l
attended a Plant Operating Review Committee (PORC) meeting which discussed the subject
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event and approved restart of the unit.
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Observations and Findinas On July 15,1999, a Unit 2 automatic reactor trip from 100 percent power occurred due to an l
inadvertent initiation of a turbine trip signal. This occurred when a P4 signal was generated during the replacement of reactor trip breaker RTA and testing of associated circuitry.
The inspectors responded to the control room following the trip and observed that operator response was in accordance with applicable procedures. Oversight of unit parameters was maintained throughout the trip recovery and the identification and resolution of several minor equipment problems was well controlled. During the event, the Train A AFW Pump did not automatically start and the reactor trip breaker RTA did not automatically open. The operators recognized these conditions promptly after the trip and manually started the AFW i
pump and manually opened the reactor trip breaker. Subsequent evaluation identified that the failure of these components to actuate automatically was expected due to the Train A solid state protection system testing which was in progress at the time of the event. No other major equipment anomalies / problems were identified during the' reactor trip recovery.
The licensee performed a post-trip review per applicable procedures to identify the apparent cause of the reactor trip, review plant and system response, and provide justification for restart of the unit to the PORC The review team determined that the cause of the event was a mispositioned switch lever arm on the front auxiliary contacts of the refurbished reactor trip breaker RTA. The mispositioned switch lever arm (associated with the P4 reactor trip signal circuitry) resulted in an inadvertent reactor trip signal being generated following replacement of RTA and solid state protection system testing of the bypass breaker BYA. Based on observation of the dispositioned component following its removal from the cabinet and l
review of circuit wiring associated with the auxiliary contacts, the inspectors concluded that the licensee appropriately identified the apparent cause of the reactor trip.
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.The inspectors reviewed procedure SI/0/A05100/002, Westinghouse DS-416 Air Circuit Breaker inspection and Maintenance, Revision 6, used for pre-installation checkout of the refurbished RTA breaker and discussed with maintenance personnel how the switch lever arm was dispositioned. During performance of the breaker checkout in the shop facility, maintenance technicians had completed successful electrical verifications for a number of circuits within the front auxiliary contact block, including circuitry for the P4 contacts. During subsequent resistance verifications being performed on the front auxiliary contact blocks, a resistance check was performed that indicated an unacceptable value. The technicians then proceeded to troubleshoot this problem via partial disassembly of the front contact block to
_ allow intemal resistance measurements to be acquired..These activities resulted in the unacceptable resistance value clearing and the contact block was reassembled. A satisfactory resistance check was then obtained for the previcus unacceptable contact.
However, during the troubleshooting activity, the orientation of the switch lever arm affecting the P4 contact was unknowingly mispositioned by the maintenance technicians. No additional re-verification of electrical continuity for the P4 or other front auxiliary contact block circuitry was performed after the block was manipulated for the troubleshooting
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activities. Based on this review, the inspectors determined that the arm was inadvertently mispositioned when breaker maintenance personnel manipulated the front auxiliary contact blocks to perform additional work / troubleshooting activities. These additional manipulations performed outside of established procedures caused the switch arm to be mispositioned.
After reattaching the auxiliary switch to the breaker, no additional continuity checks were.
performed that would have revealed this problem. Upon further review of the procedure, the inspectors considered that it was inadequate, in that, it did not clearly specify the order in which key steps were to be performed.
In addition, the inspectors concluded that the involved maintenance technician did not adhere to the intent of the procedural steps. Procedure Sl/0/5100/002, Revision 6 required maintenance technicians to notify engineering or their supervisor when contact resistance measurements were out of range. Resistance measurements performed during the breaker maintenance were beyond acceptable resistance values specified in the procedure. The lead maintenance technician for the job activity indicated that he did not notify engineering or his supervisor because he believed he met the intent of the step through his temporary supervisor training. The inspectors considered that this decision circumvented the intent of the procedural step which was to provide for an independent assessment of abnormal conditions.
The inspectors concluded that the above procedural problems are in violation of Technical Specification 5.4.1.a. This violation is in the licensee's corrective action program as Problem investigation Process report (PIP) 2-M99-3276. Planned corrective actions included procedural enhancements and personnel training. Accordingly, this Severity Level IV violation is being treated as a Non-Cited Violation (NCV), consistent with /.ppendix C of the NRC Enforcement Policy. It will be identified as NCV 50-370/99-05-01: Failure toFollow Procedure and an inadequate Procedure Regarding Reactor Trip Breaker Maintenance and Testing.
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.For the restart of the unit, the licensee reinstalled and tested the original reactor trip breaker.
The inspectors observed restart (criticality) of the unit. No problems were noted for these evolutions, c.
Conclusions Operator response to an inadvertent Unit 2 reactor trip was in accordance with applicable response procedures. Operator oversight of unit parameters was well controlled during the reactor trip recovery and during the subsequent unit restart. A non-cited violation was identified conceming an inadequate reactor. trip beaker maintenance procedure which resulted in a Unit 2 reactor trip. Inappropriate personnel actions also contributed to the root cause of the event. A post-trip review of the event was well organized, accurately identified the apparent cause of the trip, and resolved other equipment anomalies prior to restart of the unit.
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Operations Organization and Administration 06.1 Institute Of Nuclear Power Operations (INPO) Report Review (71707)
During the inspection periot., the inspectors reviewed the final INPO report from October 1998. The final report was not available to the NRC until July 1999. The inspectors
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i determined that the results of the INPO evaluations were generally consistent with the results of inspections conducted by the NRC. No new items for followup were identified.
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Quality Assurance in Operations 07.1 Auomented Trainina for Time Critical Operator Actions (71707)
The inspectors reviewed the licensee's efforts to develop and implement special training for plant operators and support staff on time critical operator actions for abnormal and emergency situations. The inspectors discussed this initiative with operations staff personnel involved in the effort. Twenty-six time critical actions required fo.r design, design basis events, or economic reasons that have a defined time limit were included in the training. These operator actions were consolidated into one training session and were identified through a review of the Updated Final Safety Analysis Report (UFSAR),
emergency procedures (EPs), abnormal procedures (APs), and probabilistic risk assessment (PRA) risk significant operator actions. The training augmented original operator training which was performed by plant system or event scenario specific means.
The training cov~ered the actions, basis for actions, expectations for operator performance of the tasks, and references (PIPS, UFSAR, EPs, APs, etc.). In addition, the licensee has piloted this initiative in an industry working group with other nuclear power plants. The inspectors concluded that this initiative was well researched, provided excellent focus on safety, and provided a redundant method to verify UFSAR assumptions, PRA risk significant j
operator actions, and proceduralimplementation. Overall, the augmented training was considered a strength.
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. Miscellaneous Operations issues (92901,90712)
08.1 (Closed) Licensee Event Report (LER) 370/99-004-00: Reactor Trip During Solid State
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Protection System (SSPS) Testing l
Based on review of the subject reactor trip discussed in Section 02.2, this LER is closed.
I II. Maintenance M1 Conduct of Maintenance
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M1.1 General Comments a.
jnspection Scope (61726.62707)
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The inspectors reviewed a variety of maintenance and/or surveillance activities during the inspection period, focusing on outage-related testing and maintenance activities that included the following specific items:
Sl/0/A/5100/002, Revision 6, Westinghouse DS-416 Air Circuit Breaker Inspection
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and Maintenance IP/0/A/3215/004, Revision No. 8, Magnetrol Liquid Level Control Switch Calibration
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PT/0/A/4600/029, Revision No. 6, Doghouse Water Level Trip Actuating Device
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FunctionalTest b.
Observations and Findinos The inspectors witnessed selected surveillance tests to verify that approved procedures were available and in use; test equipment was calibrated; test prerequisites were met; system restoration was completed; and acceptance criteria were met. In addition, the inspectors reviewed or witnessed routine maintenance activities to verify, where applicable, that approved procedures were available and in use, prerequisites were m_et, equipment restoration was completed, and maintenance results were adequate.
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Conclusions i
The inspectors concluded that the reviewed maintenance and surveillance activities were adequately completed.
l M2 Maintenance and Material Condition of Facilities and Equipment M2.1 intemal Plant Flood Protection Readiness a.
Inspection Scope (62707.61726.40500.71707)
The inspectors reviewed the licensee's preparedness for coping with a potentialinternal plant flood event. The licensee's event investigation and proposed corrective actions to a
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recent flooding event were reviewed. A walkdown of plant areas was performed to assess visible material condition of potential flood sources that could affect safety-related and important-to-safety equipment. The inspectors also reviewed a corrosion related issue in large-bore service water system (SWS) piping in the auxiliary building. Flood detection, mitigation equipment, and control room indications were also inspected and selected maintenance records were inspected. Other plant information contained in plant procedures, TS, UFSAR, design basis documents (DBDs), and PRA was also consulted.
. The licensee's response to NRC Generic Letter (GL) 89-13, SWS Problems Affecting Safety-Related Equipment, was also reviewed. Operator knowledge, training materials, and
. plant procedures related to ficoding were assessed.
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Observations and Findinas The inspectors reviewed flood protection readiness for the auxiliary building, turbine building, interior / exterior doghouses (main steam and feed water valve vaults), and the emergency
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diesel generator buildings. This followup inspection was performed because the licensee experienced two floods in the auxiliary building within the last year. These floods were documented in inspection Report (IR) 50-369,370/98-09 (fire protection pipe leak in Unit 2 spent fuel pool cooling pump room) and IR 50-369,370/99-02 (Unit 1 AFW pumps room
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flood during draining of Unit 2 SWS).
l The inspectors evaluated the licensee's progress in resolving corrective actions from the licensee's event investigation team (EIT) and NRC findings from the SWS flood in the Unit 1 l
AFW pump room. The root causes of this flood were attributed to inadequate resolution of industry operating experience provided in 1985 and McGuire specific experience.
Numerous corrective actions were proposed by the licensee to improve flood prevention (equipment operation) and mitigation capabilities in PIPS 0-M99-1343,0-M99-0431, and 0-l M99-1596. Some of the proposed and implemented corrective actions included additional
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l levelindication in the AFW pump rooms; development of AP(s) for flooding; addition of
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groundwater sump level switches to the preventive maintenance program; enhancements to procedures for draining systems such as condenser cooling water and service water; incorporation of flooding information into the engineering support program;.and periodic inspection of the underdrain system portals in the groundwater sumps. The inspectors concluded that the proposed corrective actions were comprehensive measures to reduce
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In reviewing past operating experience, the inspectors identified a less than adequate
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response to GL 89-13, Action V, which stipulates, in part, that licensee's conduct a review of (
maintenance practices and operating procedures to ensure operators of the equipment will perform effectively. By letter dated January 26,1990, the licensee informed the NRC that no specific review was required because the licensee h'ad seven programs that covered plant procedures, training, and operating experience. The inspectors concluded that the licensee's response did not acknowledge the safety-issues in the GL and attached NUREG-l 1275, Operating Experience Feedback Report - SWS Failures and Degradations, Volume 3, in that, several risk-significant floods at nuclear power plants were directly attributable to operator errors and procedural problems that resulted in mispositioned valves. The licensee's incomplete response to GL 89-13 contributed, in part, to the March 1999 flood
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event. Ho' wever, the inspectors considered the current corrective actions to enhance system
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block tagout procedures; improve operator training to prevent inadvertent flooding; and -
develop APs to cope with a potential flood address the intent of the GL 89-13, Action V.~
The licensee's original PRA study of intemal flooding hazards contained some industry operating experience on flood events and noted particular plant vulnerability to flooding in the AFW pump room. Beyond design basis floods were analyzed with potential impact on -
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critical equipment such as the AFW pumps and the auxiliary shutdown panel. However, the
-inspectors noted that the PRA did_not account for a flood in one AFW pump room affecting
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the other unit's AFW pump room by.way of the underdrain system. The underdrain system
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engineering results indicated that the estimated flow rate into groundwater sump B (in the Unit 2 AFW pump room) was approximately 80 gallons per minute (gpm) during the March 1999 flood in the Unit 1 AFW pump room. For room water level above the level experienced in the March 1999 flood, the licensee estimated that flow rates would not significantly increase through the underdrain system and that ground water sump pumps in each sump
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have sufficient capacity (approximately 500 gpm per sump) to cope with the input rate. It should be noted that the ground water sump pumps were provided to cope with a 700 gpm
' design basis extemal plant leak and to prevent excess hydrostatic loading on the auxiliary building. The licensee was reassessing the PRA at the end of the inspection period as part of the corrective actions for flood protection deficiencies.
The licensee did not have APs for coping with a plant flood and relied mostly on annunciator z
response procedures (ARPs). The ARPs for the turbine building sump, auxiliary building ground _ water sumps, residual heat removal and containment spray pump areas, diesel
- generator building sumps, and interior / exterior doghouses were available in the control room. An operating procedure was in place for standby shutdown facility operation in case of a sabotage-induced flood. Operators were familiar with flood indications and were knowledgeable of ARP actions; however, some of the ARPs provided limited instructions.
For example, the ARP for turbine building sumps provided an instruction to isolate the source of the flood. No plant shutdown criteria were provided. The licensee indicated that additional guidance wouid be provided during the development of the AR(s) for flooding.
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During the review of the licensee's recent corrective actions, the inspectors noted that the licensee solicited and received feedback from other nuclear stations in the industry who have APs for coping with intemal floods.
Flood-related level alarms provided in the control room were not in alarm and were
- consistent with the inspectors field observations of sump and area levels. Many sumps and areas inspected had redundant level switches, and maintenance records indicated that level switches were periodically inspected through the preventive maintenance program.
Sampled surveillances for doghouse level switches indicated satisfactory instrument calibration and logic testing in accordance with TS surveillance requirements 3.3.2.1 and
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The inspectors observed that the visible material condition of SWS and fire protection (FP)
piping in the auxiliary building was generally adequate; however, exterior surface corrosion
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.was noted on uninsulated SWS piping in the Unit 1 AFW pump room. In addition, the
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licensee recently discovered piping degradation in 36-inch diameter SWS discharge piping.
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The affected pipes were SWS discharge lines to the standby nuclear service water pond l
(SNSWP) (past valves ORN149 and ORN152) where ultrasonic testing revealed increased
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l not be isolable and would result in the pond draining into the plant. The licensee concluded that the pipes were operable; documented the evaluation in PlP OM-99-0329; and initiated -
increased monitoring of pipe conditions. The licensee informed the inspectors that a l
conceptual modification to sleeve the affected segments of pipe was under review.
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- Ultrasonic testing in locations upstream of the above valves did not reveal a similar problem.
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Conclusions (
Comprehensive corrective actions for flood protection deficiencies were being implemented i
in a timely manner to reduce plant vulnerability to an internal flood. Flood protection l,
equipment and instrumentation were operable and were maintained in good condition. Plant i
procedures for coping with a flood were being improved through the development of APs for flooding. Operators demonstrated adequate knowledge of flood indications and associated flood protection actions. The licensee's response to GL 89-13. Action V, was less than j
adequate,in that, no specific review of plant procedures was performed to reduce human errors in operation of the SWS. This contributed, in part, to the flood of the Unit 1 AFW pump room during draining of the Unit 2A service water system in March 1999. Visible material condition of potential internal plant flood sources (i.e., FP piping, SWS piping) in the auxiliary building was generally adequate, with the exception of uninsulated portions of service water pipe in the Unit 1 AFW pump room which exhibited substantial exterior corrosion. Recent inservice inspection of large-bore, non-isolable service water pipes (in the
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auxiliary building) to the SNSWP indicated deep pitting on the interior pipe wall. Corrective actions were initiated to address the degraded pipe conditions.
M2.2 lee Condenser Lower inlet Door inspection a.
Insoection Scope (61726)
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During the inspection period, the inspectors reviewed licensee activities associated with a specialinspection of the Unit 1 ice condenser lower inlet doors.
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Observations and Findinas Due to elevated humidity levels in the Unit 1 containment and previous industry problems associated with adverse ice accumulations, the licensee determined that a specialinspection for ice accumulation which could affect the operability of the ice condenser. lower inlet doors was warranted. The elevated humidity levels were postulated as being from a non-reactor coolant source, such as chilled water systems, and was not posing any other problem for I
operation of the unit. The scope of the work involved lowering a video camera from the intermediate deck location in each bay to identify any ice buildup on the lower doors. The work was performed via work order 98157177 within the allowable action times of TS 3.6.13(A). The inspectors reviewed Revision 1 of the critical evolution plan developed for
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' the evolution and considered it very detailed and complete with compensatory measures for l a variety of potential situations. The results of the inspections did not identify any visible ice
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buildup on the lower door locations inspected.
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Conclusions e
The licensee's performance of a special inspection for ice accumulation on the Unit 1 ice condenser lower inlet doors was well planned and implemented. The conservative inspection was based on response to associated industry events and exceeded the
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requirements of routine TS surveillances.
IV. Plant Sunoort R1 Radiological Protection and Chemistry Controls -
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R1.1'. General Comments (71750)
. The inspectors made frequent tours of the controlled access area and reviewed radiological
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j postings. The inspectors observed that workers were adhering to protective clothing -
requirements. The inspectors also determined that locked high radiation doors were properly controlled, high radiation and contamination areas were properly posted, and radiological survey maps were updated to accurately reflect radiological conditions in the
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respective areas.
R2 Status of RP&C Facilities and Equipment R2.1 - Insoection of Construction Activities and Review of Documents for independent Spent Fuel Storaoe installation (ISFSI) Pad Construction a.
Insoection Scope (60853)
The inspectors examined forms and installed rebar for the ISFSI concrete pads to verify that they were installed in accordance with procedures, drawings, industrial codes, and standards. The inspectors witnessed concrete placement activities and reviewed concrete pad construction documents to verify that the installation and construction activities were adequately performed and documented.
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Observations a5d Findinos The desigb and construction specifications and instructions for the placement of rebar and construction of concrete forms for the initial ten ISFSI pads were reviewed and inspected.
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.The app!icable codes for rebar installation were American Concrete institute (ACl), American Institute for Steel Construction (AlSC), and American Society for Testing and Materials
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(ASTM). The procedure and drawing used for the rebar and concrete formwork installation
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, Specification Number MCS-1140.00-0010, " Specification for the McGuire Nuclear
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' Station independent Spent Fuel Storage installation," dated June 28,1998 f
Drawing Number MC-1030-10.04-01," Dry Cask Storage Project Cask Pads and-
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- ... Trenches Concrete and Reinforcing" Specification Number MCS-1140.00-0010 covered the technical requirements for placing reinforcing steel and concrete for the ISFSI and requirements for the construction contractor.
Drawing Number MC-1030-10.04-01 shows the layout details and design of the rebars to be
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The elements inspected were pad dimension, concrete formwork and bracing, and size,
spacing, clearance, support, wire tightening, and protective comer steel angle with Nelson
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'i The inspectors were satisfied with the installation of the rebar and the construction of the
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formwork for the concrete. The bottom layer of rebar was adequately supported by epoxy
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coated beam bolsters (high chairs), and the top layer of rebar was supported from the
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bottom layer of the rebar by the vertical rebar. The tightening wires did not protrude into the concrete protection zone, which were two and three inches from the surface.
On June 22,1999, the' inspectors witnessed the placement of the concrete for the ISFSI
. pads, l.icensee activities included inspection of each truck of concrete for transit time, mixing, and documentation; testing for slump and air entrainment; and fabrication of compressive strength samples. Contractor activities included placement, consolidation (vibration), and finishing of the concrete.
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Other documentation ' reviewed were MNS Variation Notice (VN) reports that pertained to the
' design, layout, and construction of the ISFSI pads. Variation Notices reviewed included: VN-42481/P2C, VN-42481/P2AK, VN-42481/P2AH and VN-42481/P2K. No issues were identified during this review.
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Conclusions I
l The licensee' performed adequate installation of concrete rebar and formwork, as well as adequately controlled placement of the concrete for the ISFSI pads. The construction records and related documents were adequate.
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R2.2 Review of Documents for ISFSI Pads Construction a.
Inspection Scope (60853)
The inspectors reviewed design calculation documents and observed the proposed transportation pathway for the spent fuel cask transporter over buried piping and components.
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Observations and Findinos
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McGuire Nuclear Station plans to transport spent fuel casks utilizing a specially designed transporter over bu'ried piping and components inside the prctected area. The transportation path was detailed in drawing " Survey of Drycask Haul Route Plan View" and detailed in engineering calculation MCC-1151.03-00-0004, " Spent Fuel Cask Transportation Path Evaluation of Buried Pipes and Components." This calculation identified and evaluated buried piping from various plant systems and drainage lines. 'Also included in the review
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were the radwaste facility pipe trench, other electrical trenches, ground water catch basins, security trenches and valve boxes. Three items were identified in the above calculation as needing additional review and/or modification to the existing field condition. The licensee is tracking these items until final resolution.
- Other documentation reviewed was MNS VN-42481/P2Z that pertained to the transportation pathway. No issues were identified during this review.
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, Conclusions The licensee performed adequate identification and evaluation of the transportation pathway for the specially designed fuel cask transporter over buried piping and components inside the protected area.
F8 Miscellaneous Fire Protection issues (64704,92904)
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F8.1 - (Closed) Inspector Followup item ( IFI) 50-369/98-09-03: Adequacy of Hydrogen Line impact on Refueling Water Storage Tank (RWST)
This issue involved the inspector's questioning whether a non-seismic hydrogen line routed through the Unit 1 RWST safety-related trench had been properly evaluated for compliance with applicable sections of 10 CFR 50 Appendix A, Criterion 3 and other FP system requirementsf Licensee initial operability reviews of the subject line were discussed in IR 50-369,370/98-09. Subsequent licensee operability reviews were documented in PIP 1-M98-2666, which concluded that the issue did not impose any immediate operability
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concerns for the RWST. The inspectors reviewed available documentation regarding the subject piping and discussed the as built configuration with engineering personnel.
Historicalinformation indicated that the original design routed a carbon steel, non-seismic hydrogen line from a satellite hydrogen supply tank through the RWST trench to the auxiliary
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. building. In 1986. after external corrosion problems were identified on the original piping, modification NSM MG-11922 was implemented which replaced most of the piping in the trench with stainless steel piping with welded connections. However, the stainless steel piping installed in 1986 was also non-seismically qualified. The RWST trench is a safety-related structure which encloses the RWST supply line and other support equipment for
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operation of the safety-related RWST. The trench is installed with missile protection on the top and submerged sump pumps to remove water in leakage from rain or runoff. The trench j
runs from the auxiliary building and has a 90 degree tum, which point it intersects the RWST. This area of the trench is beneath the RWST missile protection and the potential for a hydrogen leak accumulation in this area would result in the gas being in the immediate
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area of the RWST tank shell and associated piping. The licensee's evaluation of the subject I
hydrogen line concluded that it was not in conformance with the requirements of 10 CFR 50, Appendix A, Criterion 3. '
The inspectors reviewed the regulatory requirements regarding this issue and determinci the following. Appendix A to Branch Technical Position (BTP) APCSB 9.5-1, Guidelines for
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Fire Protection for Nuclear Power Plants Docketed Prior to July 1,1976, provided applicable, basic fire protection guidance for nuclear power plants such as McGuire to meet 10 CFR 50 Appendix A, Criterion 3. This guidance was similar to that contained in NUREG 0800, Standard Review Plan regarding the same subject. Specifically, Section 5.d of the BTP, Control of Combustibles, states that, " Hydrogen lines in safety-related areas should be either
designed to seismic Class I requirements, or sleeved such that the water pipe is directly
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vented to the outside, or should be equipped with excess flow valves so that in case of a line break, the hydrogen concentration in the affected areas will not exceed 2 percent." Based on the inspectors' and licensee's reviews, it was determined that the hydrogen line routed in the safety-related RWST trench structure was not seismically qu lified nor was the piping sleeved / vented, and the piping supply did not contain excess flow valves as described by i
Section 5.d of the BTP. No exception to the applicable BTP guideline was identified during the review.
10 CFR 50.48, Fire Protection, states, in part, that each operating nuclear power plant must have a fire protection plan that satisfies Criterion 3 of Appendix A of this-part (10 CFR 50).
The plan must describe specific features necessary to implement the fire protection program, including the means to limit fire damage to structures, systems, or components important to safety so that the capability to safely shut down the plant is ensured.10 CFR 50, Appendix A, Criterion 3, Fire Protection, requires, in part, that structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions. Contrary to 10 CFR 50.48, the licensee failed to ensure that their fire protection plan included necessary provisions to satisfy the requirements of Appendix A, Criterion 3. Specifically, the design and location of a hydrogen line routed in the Unit 1 RWST safety-related pipe trench was not adequately evaluated to ensure the probability and effect of fire and explosion was minimized. In addition to the above requirements, the subject piping did not meet the guidance provided in BTP APCSB 9.5-1, Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1,1976, provided as a means to satisfy the requirements of 10 CFR 50, Appendix A, Criterion 3. This NRC identified Severity Level IV violation is being treated as an NCV, consistent with Appendix C of the NRC Enforcement Policy. It is
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. identified as NCV_50-369/99-05-02: Failure to Meet 10 CFR 50.48 Conceming Non-Seismic l
Hydrogen Piping in a Safety-Related Area. This violation is in the licensee's corrective
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action program as PIP 1-M98-2666. The licensee evaluated the non-conforming condition in l
accordance with their corrective action program and initiated a modification to incorporate
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excess flow valves, as well as other improvements to address this issue.
in addition to the above, the inspectors also reviewed NRC GL 93-06, Research Results on
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Generic Safety issue 106, " Piping and the Use of Highly Combustible Gases and Vital i
Areas", dated October 25,1993. This GL detailed a variety of hydrogen storage and l
_ distribution system design vulnerabilities similar to the hydrogen line identified in the Unit 1
RWST trench. The GL also provided numerous design altematives which could be J
I incorporated to minimize the associated risk of non-seismically qualified hydrogen lines in or around safety-related structures. These included the incorporation of excess flow valves,
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l which the licensee plans to incorporate to resolve the subject issue. Based on the above, the inspectors considered that the licensee's review of GL 93-06 was weak in that it did not
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identify or evaluate the potential risk of a non-seismic hydrogen iine routed through the Unit
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i 1 safety-related RWST trench. This IFl is considered closed.
l V. Manaaement Meetinas
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i X1 Exit Meeting Summary
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The resident inspectors presented the inspection results to members of licensee I
management at the conclusion of the inspection on August 5,1999. The licensee l
acknowledged the findings presented. No proprietary information was identified.
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PARTIAL LIST OF PERSONS CONTACTED
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Licensee
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Barron, B., Vice President, McGuire Nuclear Station Bhatnagar, A., Superintendent, Plant Operations Boyle, J., Manager, Civil / Electrical / Nuclear Systems Engineering
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Byrum, W., Manager, Radiation Protection i
Cash, M., Manager, Regulatory Compliance Dolan, B., Manager, Safety Assurance Evans W., Security Manager Geddie, E., Manager, McGuire Nuclear Station Peele, J., Manager, Engineering Loucks, L, Chemistry Manager
- Thomas, K., Superintendent, Work Control Travis, B., Manager, Mechanical Systems Engineering
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. INSPECTION PROCEDURES USED p
IP 3' 7551:
'Onsite Engineering.
IP,40500:.
Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems
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IP 60853:
ISFSI Construction IP 61726:
Surveillance Observations IP 62707:
Maintenance Observations
- IP 64704:
Fire Protection
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. Conduct of Operations J f
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IP 90712:
In Office Review of Written Reports of Non-Routine Events ilP 71750:
Plant Support.
IP 92901:
Followup-Operations IP 92904:
Followup - Plant Support
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ITEMS OPENED AND CLOSED DMD9.51
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50-370/99-05-01 NCV~
Failure to Follow Procedure and an inadequate Procedure Regarding Reactor Trip E.reaker Maintenance and Testing (Section C2.2)
50-369/99-C5-02 NCV-
. Failure to Meet 10 CFR 50.48 Concerning Non-Seismic Hydrogen Piping in a Safety-Related Area (Section F8.1)
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'50-370/99-004-00 LER Reactor Trip During SSPS Testing (Section 08.1)
50-369/98-09-03 IFl Adequacy of Hydrogen Line impact on Refueling Water Storage Tank (Section F8.1)
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LIST OF ACRONYMS USED -
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'ACl
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American Concrete institute AISC -
American Institute for Steel Construction
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. 4uxiliary Feedwater
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AP.
Abnormal Procedure
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i ARP Annunciator Response Procedure
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. ASME American Society of Mechanical Engineers
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L ASTM.
American Society for Testing and Materials
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- Branch Technical Position
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- CFR Code of Federal Regulations
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Design Basis Document -
- ECCS_
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EIT Event investigation Team
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ESF Engineered Safety Features
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Emergency Procedure FP
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Fire Protection
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GL Generic Letter
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GPM G611ons Per Minute
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IFl.
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inspector Followup item
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IC Ice Condenser
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INPO Institute of Nuclear Power Operations
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IR --
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Inspection Report
ISI-
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Inservice inspection
Independent Spent Fuel Storage Instauation
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LER
' Licensee Event Report _
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Loss of Coolant Accident
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MN_S
McGuire Nuclear Station
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Non-Cited Violation
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NRC
Nuclear Regulatory Commission
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NRC Office of Nuclear Reactor Regulation
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NSM
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Nuclear Station Modifications
OAC'.
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Operator Aid Computer-
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P4
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Permissive 4
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Public Document Room
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Problem investigation Process
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Plant Operations. Review Committee
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Pounds Per Square Inch Gauge
Periodic Test
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Radiologically Controlled Area
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Residual Heal Removal
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Refueling Water Storage Tank
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Safety Evaluation Report
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SNSWP
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Standby Nuclear Service Water Pond
SR
Surveillance Requirement
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Senior Reactor Operator
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SSPS
Solid State Protection System
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Service Water System
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TS
Technical Specifications
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Opdated Final Safety Analysis Report
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VN
Variation Notice
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