IR 05000263/1998007
| ML20236F002 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 06/25/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20236F000 | List: |
| References | |
| 50-263-98-07, 50-263-98-7, NUDOCS 9807010330 | |
| Download: ML20236F002 (21) | |
Text
- _ _ - _ _ _ _ _ _ _ - _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ - _ _ _
____
.
.
I U.S. NUCLEAR REGULATORY COMMISSION REGION lli l
1:
Docket No:
50-263 License No:
DPR-22 Report No:
50-263/98007(DRF)
Licensee:
Northern States Power Company l
Facility:
Monticello Nuclear Generating Station
~
"
Location:
2807 West Highway 75 Monticello, MN 55362 Dates:
April 13 through May 27 1998
.
Inspectors:
A. M. Stone, Senior Resident inspector D. Wrona, Resident inspector D. Roth, Resident inspector, Dresden Approved by:
J. W. McCormick-Barger, Chief Reactor Projects Branch 7 9907010330 990625 PDR ADOCK 05000263
_PDR
,
,
,
-_____
_ _ _ _ _ _ _ _ _. _. _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _
_ _ _ _ _. _ _ _ _. _ _
_.. _ _ _
__.__..______m_____________
_ _ _. _ _ _ _ _ _ _ _ _ _ _. _
__
_ _
_
_ _ _.. _. _ _ _ _ _ _ _. _ _ _ _ _
. _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
.
EXECUTIVE SUMMARY Monticello Nuclear Generating Station, Unit 1 NRC Inspection Report 50-263/98007(DRP)
This inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a 6-week period of resident inspection.
Operations Overall, licensed and non-licensed operator performance inside and outside the control
.
room was satisfactory. Communications observed were clear. Performance of routine and reactive procedures, including entry into Technical Specification limiting conditions for operation, was proper. Pre-evolution briefings were thorough and informative.
However, some housekeeping issues identified by the inspectors indicated that non-licensed operators were not identifying some discrepant plant radiological, housekeeping, and equipment staging items. (Section 01.1)
The operations committee review of activities and outstanding issug to support startup
.
was thorough. The initial decision to delay startup because of a cracked control rod drive withdraw line was conservative and reflected safety conscious operation. Reactor startup was controlled in a deliberate manner with good communication and procedure adherence by operations personnel. (Section 01.2)
Operations personnel did not document a May 7,1998, reactor building differential
.
pressure indication problem as part of the corrective action process. Operations personnel performed compensatory actions for the indication problem instead of initiating repairs, which demonstrated their willingness to accept discrepant conditions. The root cause of the condition was not adequately corrected until the problem recurred on May 12,1998. A lack of documentation of the problem contributed to the delay in determining the root cause. (Section 01.3)
No operability concerns were identified during tours of normally inaccessible areas and
.
walkdowns of plant systems. (Sections O2.1 and O2.2)
General quality services (quality assurance) auditors were present in the control room
.
during the refueling outage. Good interaction between the auditors and operations personnel was noted. The auditors maintained an appropriate independent oversight function. (Section O7.1)
The identification of an unexpected valve closure resulted from a good questioning
.
attitude by operations personnel. Also, when the third and final control rod to be withdrawn to achieve criticality did not move with normal control rod drive pressure, operators displayed conservative decision-making by backing out of a procedure, inserting the three control rods, and performing control rod drive exercise tests. The shift supervisor demonstrated excellent attention-to-detailin the identification of the cracked control rod drive withdraw line. (Sections M1.1 and M2.1)
l
l
_ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
.
Maintenance During the refueling outage, maintenance personnelincorrectly insta!!ed the switches for
.
flow and pressure controllers in the recombiner system. It appeared that a post-maintenance test was not completed since the licensee was not able to locate a copy of the test. Operations personnel returned the recombiner trains to service without verifying a post-maintenance test was complete. (Section M1.2).
The installation of the new control rod drive withdraw lines was conducted in a controlled
.
manner as a result of good teamwork between operations, radiation protection, maintenance, and construction personnel. (Section M2.1)
General quality services (quality assurance) auditors and quality control inspectors
.
observed several maintenance and surveillance testing activities during the refueling outage. Issues identified by the auditors were promptly communicated to licensee management and were appropriately addressed through the condition reporting or employee observation reporting systems. Good interaction between the auditors and
maintenance personnel was observed. The auditors maintained an appropriate independent oversight function of maintenance activities. (Section g7.1)
,
Engineering The license's investigation into the cause of the cracked control rod drive line was
.
thorough. Defect indications on the line were removed or repaired. Engineering personnel provided a strong technicaljustification for continued operation. The initial decision to delay reactor startup was conservative and subsequently determined to be appropriate. (Section E2.1)
Plant Support Radiation protection support for the control rod drive withdraw line repairs was excellent.
.
Technicians were actively engaged in the job by assisting maintenance, construction, and engineering personnel. (Section R1.1)
+
l General quality services (quality assurance) auditors assessed plant personnel
.
performance with respect to radiological protection. Concems, such as maintenance personnel leaning over a contaminated area boundary and a protective clothing l
changeout area located over grating, were promptly and appropriately addressed. The l
auditors findings were similar to the inspectors' observations. (Section R7.1)
l
!
.
_. _ - - _.
-
_ _ _ _ _ _ _ __-_ _ __ - _ _ _
.
Report Details Summary of Plant Status On April 24,1998, the licensee initiated a unit startup from a refueling outage which began on
,
March 20,1998. At about 12:07 p.m., the unit was made critical. Startup was delayed by two
'
days because of discovery and repair of a cracked control rod drive withdraw line inside of the drywell. This issue is discussed in Sections M2.1 and E2.1. The unit operated at power levels up to 100 percent for the remainder of the period.
l. Operations
Conduct of Operations 01.1 General Performance Comments a.
Inspection Scope (Inspection Procedure (IP) 71707)
s
.1 The inspectors observed routine operations in the control room and in the plant. Tasks assessed included logkeeping, the identification of and response to abnormal conditions, and operator shift turnovers.
b.
Observations and Findinas The inspectors noted that operations personnel discussed significant issues and equipment status during individual shift turnovers. The control room staff generally used a repeat-back technique in communications; however, the inspectors noted one example where the lead reactor operator was unaware that another operator enterea an annunciator response procedure. The control room atmosphere was professional, quiet, and free from distractions. The inspectors observed that shift personnel entered the appropriate Technical Specification (TS) limiting conditions for operation for planned surveillance tests. The inspectors identified one example of inadequate corrective actions when a safety-related reactor building pressure indicator was discovered to not be reading accurately. This is discussed in Section 01.3.
The inspectors observed severalinfrequent evolution briefings conducted in the control room. Shift management and test directors clearly defined roles and responsibilities to those involved and discussed the sequence of activities as well as contingency actions should a problem occur. Parameter limitations such as vessel pressure and reactor water temperature were discussed during the infrequent evolution briefing for the vessel pressure test.
The inspectors accompanied non-licensed operators (NLO) on rounds and assessed the operators' identification of issues. The NLOs correctly performed the required rounds and recorded the results on the rounds sheets. However, the inspectors independently identified severalissues that the NLOs did not identify during plant rounds. For example, on May 11,1998, the inspectors noted that a continuous air monitor partially blocked an area labeled to be kept free of obstructions. On May 14,1998, the inspectors identified that the residual heat removal (RHR) B corner room was cluttered with cut sections of
__.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _
_
__
i
!
l pipe, debris, and tools apparently left over from the recently completed refueling outage.
l A blade for a band saw was hung from a small drain line such that part of the blade was l
inside and against a contaminated drain trough while the rest of the blade was in a clean
area. In both instances, the inspectors notified operations personnelimmediately of the concerns and the licensee subsequently addressed the issues.
l The inspectors recognized that the NLOs did not necessarily cause these housekeeping problems; however, the inspectors concluded that the NLOs had not been vigilant in identifying discrepant plant radiological, housekeeping, and equipment staging items.
c.
Conclusions Overall, licensed and non-licensed operator performance inside and outside the control room was satisfactory. Communications observed were clear. Performance of routine and reactive procedures, including entry into TS limiting conditions for operation, was l
proper. Pre-evolution briefings were thorough and informative. However, some housekeeping issues identified by the inspectors indicated that NLOs were not identifying some discrepant plant radiological, housekeeping, and equipment staging items.
s-r O1.2 Preparation for Startup and Startup Activities a.
Inspection Scope (IP 71707)
The inspectors observed preparation activities for startup and various portions of the April 24,1998, startup including approach to criticality and power ascension. The following procedure was reviewed:
Procedure C.1, "Startup Procedure," Revision 19
.
b.
Observations and Findinas The inspectors observed portions of the startup readiness operations committee meeting.
Committee members discussed completed maintenance and modification work, dispositioned outstanding work and surveillance testing issues, confirmed that regulatory commitments were addressed, and discussed outstanding operability issues and condition reports (CRs). The inspectors reviewed the CRs generated during the refueling outage and did not identify any conditions which conflicted with TSs or Updated Safety l
Analysis Report (USAR) requirements. The inspectors also reviewed the startup checklist to verify TS-required surveillance tests were completed or scheduled appropriately during the startup sequence. Also, the inspectors noted thet the licensee's initial decision to delay startup after a crack was identified in a control rod drive (CRD)
withdraw line reflected conservative operation and safety perspective.
On April 24,1998, operations personnel commenced the approach to criticality. The infrequent evolution briefing included a review of the caution statements concerning reactivity management and assignment of personnel roles and responsibilities. The shift managers had command and control of the activities as evidenced by maintaining an i
extra senior reactor operator to assist the lead operator and removing unnecessary personnelin the control room to minimize distractions. Operators conducted the startup in a slow, deliberate manner and executed the activity without error. Appropriate
l
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _
_ _ _ - _ - _ -. _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_ _ _ _ _ _ _ _ _ _ _ _
-
.___
.
'
procedures were used. Temporary procedure changes were made when necessary in accordance with established practices. The inspectors noted good communication of l
expected and unexpected alarms; an improvement compared to shutdown observations.
c.
Conclusions I
l The operations committee review of activities and outstanding issues to support startup l
was thorough. The initial decision to delay startup because of a cracked control rod drive i
withdraw line was conservative and reflected safety-conscious operation. Reactor startup was controlled in a deliberate manner with good communications and procedure l
adherence by operations personnel.
01.3 Reactor Buildino Differential Pressure a.
Inspection Scope (IPs 71707 and 61726)
The inspectors observed portions of preventive maintenance (cleaning and inspection) on secondary containment dampers. During the maintenance activity, an unexpected reactor building d. differential pressure low alarm was received. The igppectors assessed operations personnel response to the alarm, including the response to previous occurrences of the alarm. The assessmant was based on extensive discussions with the system engineer and the control room staff, and review of the following documents:
Control room log book entries for May 7-15,1998,
CR 98001299, Reactor building differential pressure indication in the control room
.
'
(DPI-4424) indicated near zero during Secondary Containment damper cleaning,"
Drawing NF-46108,
.
Operations Work Instruction 02.02, Revision 0, " Operations Written Logs," and
!
.
Administrative Work Instruction 10.01.03, Revision 7, " Condition Report Process."
.
b.
Observations and Findinos On May 12,1998, the licensee initiated preventive maintenance activities to inspect and clean the secondary containment dampers. During the performance of the work, the control room operators received Annunciator 3-A-27, "RX [ reactor] BLDG [ building] at or
above 0" water." The operators observed that the control room manometer, DPI-4424, j
which indicated reactor building differential pressure, had approached zero. A plant j
operator confirmed that other indicators, such as one located on the refuel floor, showed I
normal building pressure. The operators noted that the ventilation systems were functioning property to maintain pressure and concluded that the annunciator was caused by an indicator problem. Work related to the damper inspections was delayed until the l
cause could be corrected. The licensee later identified that a test connection fitting on the reference leg just above the differential pressure switch, SPS 8000E-8, was leaking.
l I
1
!
L__
_ _ _ _ _ _ _ _ _ _ _ _. _ _ _
_ _ _ _ _ _.. _ _ _ _ _ _ _ _ _ _
---
_ _ __
-
-
.
During discussions with the operators, the inspectors were informed that on May 7,1998, operations personnel contacted the system engineer to resolve a similarindicator problem with DPI-4424. Because it was raining at the time the indicator problem occurred, the engineer suspected that rain water may have entered the differential pressure instrument reference line. The system engineer suggested that the cap be removed from the control room manometer to vent off any trapped water. This corrected the condition temporarily as the operators had to vent three times during that shift.
Operators in the next shift also experienced this problem and informed the system engineer who suggested that maintenance personnel be contacted to resolve the problem. Operators contacted a maintenance supervisor, but, unknown to the system engineer at the time, no knowledgeable maintenance personnel were available to troubleshoot the condition. Since these events were not documented in the control room log books, work order, or CRs, the system engineer believed that the issue had been resolved. However, the problem reoccurred on May 12,1998.
Step 4.2.4 of Operations Work Instruction 02.02 stated that operators should log all significant operating actions or occurrences including unexpected annunciator alarms and the reason for the alarm when determined. The events of May 7,1998, including the receipt of alarms, contact with the system engineer, and venting thegnanometer were not
,
included in the log entries for that day. Also, Step 4.1.1 of Administrative Work Instruction 10.01.03 stated that site personnel should ensure that conditions such as degraded components are documented in a CR. After discussion with the inspectors, operations personnel made a late log book entry to document the events and initiated a CR. The inspectors concluded that this concern did not constitute a significant condition adverse to quality; therefore, a violation with respect to failure to take appropriate corrective actions was not warranted. However, the inspectors concluded that the operators did not aggressively pursue resolution of the problem and that the failure to document the activities associated with the indicator problem contributed to the recurrence.
c.
Conclusions Operations personnel did not document a May 7,1998, reactor building differential pressure indication problem as part of the corrective action process. Operators performed compensatory actions for the indication problem, instead of initiating repairs, which demonstrated their willingness to accept discrepant conditions. The root cause of the condition was not adequately corrected until the problem recurred on May 12,1998.
A lack of documentation of the problem contributed to the delay in determining the root cause.
O2 Operational Status of Facilities and Equipment O2.1 System Walkdowns a.
Inspection Scope (IP 71707)
The inspectors walked down selected portions of the following systems: instrument air, condensate and feedwater, main steam, standby gas treatment, high pressure coolant injection (HPCI), reactor core isolation cooling (RCIC), primary containment, reactor water cleanup, and emergency service water.
- _ _ - _
_ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
.
The inspectors used portions of the following procedures to aid in the determination if systems were available for reactor startup:
Surveillance Procedure 2154-03," Compressed Air System Prestart Valve
.
Checklist," Revision 29, Surveillance Procedure 2154-04, Condensate and Feedwater System Prestart
.
,
Valve Checklist," Revision 29, l
Surveillance Procedure 2154-05," Main Steam System Prestart Valve Checklist,"
.
Revision 23, Surveillance Procedure 2154-06, " Standby Gas Treatment System Prestart Valve
.
Checklist," Revision 8, Surveillance Procedure 2154-10, "High Pressure Coolant injection System
.
p Prestart Valve Checklist," Revision 21, Surveillance Procedure 2154-13,"RCIC System Prestart Vage Checklist,"
,
.
Revision 23, Surveillance Procedure 2154-15, * Primary Containment System Prestart Valve
.
Checklist," Revision 22, Surveillance Procedure 2154-19, " Reactor Water Cleanup System Prestart Valve
.
Checklist," Revision 21, and Surveillance Procedure 2154 34," Emergency Service Water System Prestart
.
Valve Checklist," Revision 18.
b.
Observations and Findinas Before reactor startup, the inspectors performed system walkdowns and valve lineup checks to ensure the selected systems were in a condition to support the startup. The material condition of the systems inspected was acceptable. Valves were lined up as required to support the reactor startup. Minor concerns were identified to the licensee and were promptly resolved. No operability concerns were noted.
02.2 Review of Normally inaccessible Areas (IP 71707)
The inspectors toured areas of the plant that were normally inaccessible during reactor operations. The tours included the following areas: drywell, steam chase, condenser room, and steam jet air ejectoi room. The tours were conducted shortly before the areas were closed up in preparation for the startup. Equipment and tools used in the areas during the outage were removed. Housekeeping was acceptable. Minor concerns were brought to the attention of the system engineer or area supervisor and were promptly resolved. No operability concerns were noted.
- - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _. _ _ _ _ _
. _ _ -. _ _
_
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _
.
.
Operator Training and Qualification l
l The inspectors observed an operator training session regarding a recent modification of I
the intermediate range monitors. The lecture was informative and included discussion on l
the reason for the modification and potential effects on control rod blocks. Operations personnel were afforded the opportunity to ask questions. The inspectors had no concerns with this training session.
O7 Quality Assurance in Operations
07.1 General Quality Services (GOS) Audits and Inspections (IP 71707)
The inspectors reviewed the activities of GQS (quality assurance) auditors during their observations of operational activities. The inspectors noted that GOS auditors were present in the control room during the refueling outage. Activities observed by the GQS
,
auditors included operations shift turnovers and shift briefings, infrequent evolution briefings, surveil!ance testing, and startup activities. The inspectors noted good interaction between the auditors and operations personnel and that the auditors maintained an appropriate independent oversight function. One fin @g involved a a.
concem with the use of secure cards instead of hold cards in some system isolations.
The auditors noted that although this practice was not contrary to procedures, the use of secure cards was considered a personnel safety risk. The licensee is evaluating this
finding.
Miscellaneous Operations issues 08.1 (Closed) Violation (VIO) 50-263/97002-01(DRP): Ventilation test not performed during suitable environmental conditions.
In January 1997, the licensee opened the turbine building railway doors during an emergency filtration system test. The cold air reduced the ambient temperature which affected nearby instrument lines and could have resulted in a reactor trip because of indicated low condenser vacuum conditions. The licensee's corrective actions included discussing this event with maintenance, operations, engineering, and security personnel and revising administrative procedures to wam site personnel of adverse environmental conditions. The inspectors had no further concerns.
11. Maintenance M1 Conduct of Maintenance M1.1 General Comments a.
Inspection Scope (IPs 62703 and 61726)
The inspectors observed all or portions of selected maintenance and surveillance testing activities. Included in the inspection was a review of the surveillance procedures or work orders (WOs) listed, as well as the appropriate USAR sections pertaining to activities.
__.
___ __ ______ _ - __
_ _ _ - _
________ - _______ - _ - _-.
__
!
b.
Observations and Findinas in general, the inspectors observed that the work associated with these activities was conducted in a professional and thorough manner. All work observed was performed with the work package present and in active use Technicians were experienced and l
knowledgeable of their assigned tasks. The inspectors frequently observed supervisors and system engineers monitoring job progress, and quality control personnel were present whenever required by procedure. When applicable, appropriate radiation control i
measures were in place.
The following work was observed. Specific concerns or observations are provided where appropriate.
Surveillance Test 0255-20-IIC-1, Revision 11, " Reactor Coolant Pressure
.
Boundary Leakage Test."
The inspectors verified that all surveillance tests which were required to be completed prior to increasing reactor water temperature above 212 degrees Fahrenheit were accomplished. This verification included a gview of the startup checklist and incomplete partial tests.
During a panel walkdown, a shift supervisor noticed that the inboard RHR head spray cooling valve, MO-2027, had changed position from open to closed. The licensee determined that the valve closed during Surveillance Test 0391, "RHR Shutdown Cooling Supply," which was performed by instrument and control technicians after the system lineup for the boundary leakage test was performed.
Steps within Surveillance Test 0391 required the technicians to increase indicated pressure on the reactor high pressure shutdown cooling isolation pressure switches. When indicated pressure was greater than 75 pounds per square inch-
!
gauge, MO-2027 closed as designed. The similar outboard valve, MO-2026, had j
been closed previously as part of the system lineup for the boundary leakage test.
After reviewing the event, the system engineer revised Surveillance Test 0255-20-IIC-1 to allow MO-2027 to remain closed. The licensee initiated a CR to document the cause and planned corrective actions. Steps within Surveillance Test 0391 addressed the RHR shutdown cooling system, but did not address the status of the head spray cooling valves. Since MO-2027 is normally closed except during a short period of time when it is opened to reduce temperatures on the vessel head at the start of an outage, the inspectors concluded that Surveillance Test 0391 was adequate.
In addition, the licensee noted that MO-2027 would have closed as reactor pressure increased during preparation for the boundary leakage test. During previous tests (conducted once a cycle), the valve closure was unnoticed. The inspectors verified that the results from previous tests remained valid since the requirements for pressure boundary system leakage tests as specified in Table IWB-2500-1 of the 1986 Edition of the American Society of Mechanical Engineers (ASME)Section XI Code were met. Specifically, the ASME Code allowed the licensee to place MO-2027 in its normally closed position. In summa;y, although the licensee intended MO-2027 to remain open during this test, the unexpected closure of the valve did not ir, validate the test results.
_ - _ - - _ - _ _ _ _ _ _ _ - _ _ _ _ - _
-
_ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ - _ - _ _ _
.
Surveillance Test 0153-02, Revision 2, " Group 2&3 Iso;ation Simulated Automatic
.
initiation Test."
j l
The inspectors noted good communication and procedure adherence by the I
control room operators during the conduct of this test. The inspectors verified that I
the acceptance criteria for valve stroke times were consistent with USAR I
Table 5.2-3b and that the test was conducted in accordance with TS requirements. The procedure step sequence required the operators to obtain
{
valve stroke times after the valves had been previously cycled. The system
)
engineer stated that stroke time testing to meet TS 4.15, ASMF Section XI testing requirements, was accomplished by another surveillance test performed earlier in the outage. In addition, the inspectors noted that the RHR piping and instrumentation diagrams, P&lD 120 and 121, did not indicate that injection valves MO-2014 and MO-2015 had a Group 2 isolation function. Thew valve; would normally receive an open permissive signal upon generation of a Group 2 isolation signal, however, during conditions with shutdown cooling in service, these valves were designed to isolate on receipt of a Group 2 isolation signal. This drawing discrepancy was not safety significant since established tests, which verify operability, were not affected. The system engineer wrote 4R to address this
,,
drawing discrepancy.
WO 9801597, " Replace Section of CRD Pipe for 34-27," WO 98011615, " Replace i
.
Section of CRD Pipe for 42-43," and WO 9801622, " Wipe Down of Northwest Bundle of CRD Pipes."
WO 9801618, " Tube Leak in E-12A or E-DC-12A (feedwater heaters)."
.
Surveillance Test 8079, "Beginning of Cycle Core Analysis Benchmark Critical,"
'
.
Revision 4 During performance of Surveillance Test 8079, the third and final control rod to be withdrawn to achieve criticality did not move with normal CRD pressure. The operators displayed conservative decision-making by backing out of the procedure, inserting three control rods, and performing CRD exercise tests. The operators chose not to increase drive pressure or use notch override due to the potential" double notching" at criticality. After CRD exercise testing was complete, Surveillance Test 8079 was completed with no further problems.
Surveillance Test 1070, Revision 8, "RCIC Flow Control System Dynamic Test
.
Procedure," and Surveillance Test 0255-08-IA-1, Revision 39, "RCIC System Tests with Reactor Pressure at Rated Conditions" Surveillance Test 0255-08-IA-1 was used to startup the RCIC system. The RCIC turbine steam supply valve, MO-2078, was initially opened and stroke timed during performance of this test. Once the system was running, Surveillance Test 1070 was used to determine if system response was adequate. During Surveillance Test 1070, MO-2078 was closed and opened; however, a closing time was not measured. The operators measured the valve's closing time later during completion of Surveillance Test 0255-08-lA-1. The inspectors questioned the system engineer concerning why Valve MO-2078 was not timed closed during l
i L_____________________
_ _ _ _ _ _ _ _ _
._ __
____ ____________ _ _-___ _ _ _ ______ ___-___________ ___ _ _
.
Surveillance Test 1070. The system engineer stated that this was an oversight and the licensee normally timed the valve on its initial closing. However, the system engineer also stated that valve MO-2078 was thied in Surveillance Test 0114, "RCIC System Tests with Reactor Pressure < 150 psig [ pounds per square inch-gauge)," which was performed prior to Surveillance Tests 1070 and 0255-08-IA-1. Following verification that MO-2078 was adequately tested in Surveillance Test 0114, the inspectors had no further concems.
Surveillance Test 1383, Revision 5, " Core Flow Measurement System Calibration"
.
c.
Conclusions Observed work and surveillance testing were conducted in a professional and thorough manner. The identification of an unexpected valve closure resulted from a good questioning attitude by operations personnel. Also, when the third and final control rod to be withdrawn to achieve criticality did not move with normal CRD pressure, the operators displayed conservative decision-making by backing out of the procedure, inserting three control rods, and performing CRD exercise tests.
M1.2 Recombiner Switch Replacement a.
Inspection Scope (IP 62707)
The inspectors reviewed the causes of unexpected responses from the recombiner system during a train swap. The review ir*cluded discussions with plant maintenance staff and review of the following documents:
WO 9801768, " Repair "A" Train Offgas Flow Control,"
.
CR 98001312, "11 Recombiner Trip While Placing OG [offgas) Controller to Auto,"
.
and I
Surveillance Procedure 1279-5,"Recombiner Train Switching and Hot Area l
.
inspections."
b.
Observations and Findina,
On May 13,1998, when operators attempted to switch from the 8 recombiner to the A recombiner, the off gas flow unexpectedly spiked which caused the A recombiner to trip. The licensee identified that maintenance personnelinstalled the wires for two switches for the recombiner pressure controllers and recombiner flow controllers (HS-0201 A and HS-0202A) incorrectly. A broken terminal post for one of the switches l
i was not documented in the comments section of the associated work order. The two I
Train A switches and two train B switches had been installed during the week of March 13,1998, during the refueling outage. The inspectors observed that a post-maintenance test (PMT) was referenced as an attachment to the WO, however, a record of the completed PMT was not present. The ins 9mors determined that the PMT would have enabled the licensee to detect the wiring error. At the end of the inspection period, the licensee was still attempting to find the completed PMT documentation. Since the switch was non-safety related and not included in the quality assurance program, no
..
_ _ _
_ - - -
.
'
I I
violation will be cited. However, the inspectors concluded that operations personnel had returned the equipment to service and accepted it as operable without an adequate PMT.
. c.
Conclusions I
During the refueling outage, maintenance personnel incorrectly installed the switches for flow and pressure controllers in the recombiner system. It appears that a PMT was not completed since the licensee was not ab!e to locate a copy of
the test. Operations personnel returned the recombiner trains to service without verifying a PMT was complete.
M2 Maintenance and Material Condition of Facilities and Equipment j
l M2.1 Crack in Control Rod Drive Withdraw Line identified Durina Leakaae Test a.
Inspection Scope (IPs 37551. 61726. 71707. and 71750)
!
On April 17,1998, during the leakage test of the reactor pressure vessel, a shift supervisor identified a small crack on the withdraw line for CRD 34-%7. The inspectors
,
observed repair activities and reviewed the following documents:
Licensee Event Report 98-003,"Transgranular Stress Corrosion Cracking
.
identified in Control Rod Dnve,"
'
WC 9801675, " Replace Section of CRD pipe for 42-43,"
.
CR 98001094, " Freeze Seal Failure on CRD 42-43 Withdraw Line,"
.
Procedure 816'7, " Freeze Sealing of Steel Pipe 4" (inches) and under " Revision 4,
.
WO 9801558, " Repair / Replace cracked CRD 34-27 pipe,"
.
WO 9801597, " Replace section of CRD pipe for 34-27,"
.
Alteration 9BA022, *CRD Withdrawal Line 34-27 Repair," Revisions 0 and 1, and
-*
Alteration 98A023, "CRD Withdrawal Line Repair (for CRD 42-43]," Revision 0.
.
b.
Observations and Findinas The CRD withdraw and insert lines for the 121 CRDs traverse through the reactor building l
to the drywell and penetrate the reactor vessel support pedes'al to the CRD housings l
beneath the vessel. The lines were arranged in four bundles, each consisting of about 60 lines. During the vessel pressure test, a shift supervisor saw water droplets on a withdraw line inside the drywell. This line is not considered part of the reactor coolant boundary. Upon detailed inspection, the licensee identified a small through-wall crack, l
near the bend prior to penetrating the reactor vessel support pedestal. Other inspections were conducted while the vessel boundary was at pressure and no additional leaks were
.
identified.
I
i
i
_ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ - _ _ - _ _ _ _ __-____- _ ___________________________ _ ______-___ ___-_______
.
.
The licensee initiated several correct ve actions which included removal of two withdraw lines and buffing a third line. (A more detailed discussion of the licensee's investigation activities is provided in Section E2.1 of this report.) The inspectors observed portions of the removal and installation of the new sections of piping. Authorization for the freeze seal was obtained by the general superintendent of engineering and the WOs were reviewed by the operations committee prior to the start of work. The inspectors reviewed the alteration packages, including the 10 CFR 50.59 safety evaluation screenings, and noted that sufficient detail was provided. Radiation protection technicians monitored dose and provided guidance on low dose locations. The licensee responded appropriately when one of the freeze seals failed. Contingency actions for a seal failure were discussed during the pre-evolution briefing and were executed as planned.
Dye-penetrant tests (pts) were conducted after the freeze seals were removed to verify the pipe was not affected by the freeze seal application. The inspectors had no concerns with the installation of the new piping.
c.
Conclusions The shift supervisor demonstrated excellent attention-to-detailin the identification of the cracked control rod drive withdraw line. The installation of the new r,pntrol rod drive withdraw lines was conducted in a controlled manner as a result of gond teamwork between operations, radiation protection, maintenance, and construction personnel.
M7 Quality Assurance in Maintenance Activities M7.1 Quality Services Audits Durina Refuelina Outaae (IPs 61726 and 62703)
The inspectors observed the activities of GOS auditors and quality controlinspectors during their observations of maintenance and surveillance testing activities. The inspectors noted that GQS auditors and quality control inspectors were present at the work sites. Activities observed by the GOS auditors included reactor water cleanup modification installation, CRD withdraw line replacements, and other maintenance jobs.
The auditors identified discrepancies with lifting and landing of ! cads, a missed hold point in a post-maintenance test, a conflict between two precedures with respect to foreign material control, several personnel safety issues, and a pressure gauge with an expired calibration date. The auditors discussed these issues with licensee management and they were properly addressed through the condition reporting or employee observation reporting systems. The inspectors noted good interaction between the auditors and maintenance personnel and that the auditors maintained an appropriate independent oversight function.
M8 Miscellaneous Maintenance issues M8.1 (Closed) VIO 50-263/96008-01(DRP): Three examples of inadequate test control were identified.
Two of the violation examples involved operations and engineering personnel not recognizing that the acceptance criteria in a standby liquid control tank level surveillance test and in a fire hydrant surveillance test were exceeded. During subsequent training sessions, the licensee discussed these events and the need for greater attention-to-detail. The third example concerned a standby liquid control routine
.
__-__ -____ _
___
_ _ _ _ _ _ _.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
__ _ - _
____
_ _ _ -
--
.
.
surveillance test which had not been revised to reflect an inservice testing relief request granted by the NRC. The licensee revised the inservice testing administrative procedure to require personnel to reference relief requests in affected plant procedures.
I M8.2 (Closed) Licensee Event Report (LE_R) 50-263/98003: Transgranular Stress Corrosion identified in Control Rod Drives. This event is discussed in Section M2.1.
Ill. Enaineerina E2 Engineering Support of Facilities and Equipment E2.1 Control Rod Drive Withdraw Line Crack Investigation a.
Inspection Scope (IPs 37551 and 71707)
On April 17,1998, during the leakage test of the reactor pressure vessel, a shift supervisor identified a small crack on the withdraw line for CRD 34-27. The repair and replacement of the lines is discussed in Section M2.1. The justifica6pn for operation and
_,.
engineering-related issues are discussed below. The inspectors also reviewed the following documents:
CR 98001023, "Possible Cracked CRD Withdrawal Line in DW [drywell) Found
.
During RPV [ reactor pressure vessel) Hydro," and Letter dated April 21,1998, from R. Bloch, Carolina Power & Light Company to
.
J. Ricker, NSP [ Northern States Power).
b.
Observations and Findinas The crack was identified while the reactor vessel was pressurized for the pressure boundary leakage test. Although the CRD withdraw lines were not considered part of the reactor coolant boundary, the lines were inspected during this evolution. No leaks were identified on the other withdraw or insert lines. Engineering personnel performed another inspection after reactor vessel pressure was reduced and did not identify additionalleaks.
The licensee initiated several other actions to evaluate the extent of the problem. These included:
Removed the cracked portion of the CRD 34-27 withdraw line and sent it to a
.
laboratory for analysis. The contractor performing the analysis indicated that
,
chlorides were present on the pipe sample and concluded that the crack resulted
{
from chloride-induced transgranular stress corrosion initiated from the outer diameter.
Performed a PT inspection of 14 withdraw and nine insert lines in the same
.
bundle at the bend location. No indications were identified.
Performed PT inspections on six withdraw lines in each of the other three
.
bundles. No indications were identified.
-__ _ _ _-_ _ _ __-_-_________ ____ __ - _ - _ _ _ _ _ _
.
.
Performed PT inspections on the verticallength of CRD 34-27 and found several
.
indications. This entire line was replaced. The licensee also conducted a PT examination of two withdraw lines for CRD 42-43 and 38-27. Indications found were repaired or removed.
Swiped several lines within each bundle to identify areas of high chloride
.
concentrations. A higher chloride concentrat;on was found in one of the bundles.
The accessible portions of the lines were hand-wiped and retested by the licensee.
The licensee's prompt and thorough investigation into this event reflected safety-conscious, operations. The licensee maintained focus on resolving the issue prior to starting up the reactor. Although the licensee could not identify the source of the chlorides, the work accomplished (PT examinations and repair of three lines) provided a strong base to justify startup and continued operation until the next refueling outage. The inspectors reviewed CR 9801023 which documented the justification for startup.
Concerns such as the ability to scram a rod with a cracked withdraw or insert line and the potential effect of multiple lines breaking were addressed in the report. The failure of multiple lines was bounded by the loss-of-coolant accident analysis,This issue was also
"
discussed in great detail with other NRC staff members including Region lli management and NRR technical support personnel. The licensee planned to evaluate the need for further inspections during the next refueling outage. Contingency actions included revising containment leakage procedures to alert operators to an additionalleakage source.
c.
Conclusions The license's investigation into the cause of the cracked CRD line was thorough. Defect indications on the lines were removed or repaired. Engineering personnel provided a strong technical justification for continued operation. The initi6l decision to delay reactor startup was conservative.
E2.2 Investigation into Hioh Pressure Iniection System Waterhammer Events As discussed in Section E1.1 of Inspection Report 50-263/98004, the licensee experienced a waterhammer in the HPCI piping on February 16 and February 27,1998.
l
'
At the time of these events, the licensee suspected that the discharge valve, MO-2068, was backleaking. The valve was inspected during the subsequent refueling outage, but the licensee identified only minor indications on the valve. On May 13,1998, as part of l
the followup actions for this issue, an engineer performed an ultrasonic examination of I
the water-filled piping near the valve and identified that a small void existed. The system engineer evaluated the size of the void and determined that the HPCI system remained operable, since the system was capable of withstanding a large waterhammer event.
The system engineer initiated work orders to measure the pressure across the testable check valve located downstream of MO-2068 and to fill the piping between these valves with demineralized water to collapse the void. However, another minor waterhammer l
occurred when MO-2068 was stroked after these maintenance activities. The system l
engineer and operations personnelinspected the piping supports and determined that no
'
damage had been done. The licensee continued to investigate the cause of the waterhammers at the end of this inspection period. The inspectors' review of the
L___________--_.---
. - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _
__- __ ____ _ _ _ _ _ - _ _ _ - _____-_______ - - -__ - _ - _ _
,
licensee's root cause determination and corrective actions is an Inspection Follow-up
,
item (50-263/98007-01(DRP)).
j
!
E8 Miscellaneous Engineering issues E8.1 (Closed) Escalated Enforcement item (EEI) 50-263/96009-11(DRS): Apparent violation of j
10 CFR 50.59 due to undetected unresolved safety question; I
(Closed) eel 50-263/97010-2(DRS): Use of containment overpressure; and
(Closed) LER 50-263/97001: " improper Review Fails to identify Unreviewed Safety
Question."
in a letter dated April 20,1998, to Mr. M, Wadley, Vice President Nuclear Generation, from A. B. Beach, Regional Administrator, the NRC informed the licensee that based on the information developed during inspections and the predecisional enforcement conference, two violations of NRC requirements were identified. The first violation pertained to a reduction in the number of RHR and RHR service water pumps required for containment cooling. Initially, the inspectors questioned the adequqgy of the
,
10 CFR 50.59 safety evaluation performed by the licensee of the reduction; however, upon further review, the NRC concluded that the violation involved poor implementation of the licensee's design change process. The second violation involved the use of containment overpressure in the emergency core cooling system pump net positive suction head calculations without previous approval by the NRC. The failure to identify
"
that this constituted an unresolved safety question is a violation of 10 CFR 50.59.
Although both violations were classified as Severity Level til, enforcement discretion was granted in accordance with Section Vll.B(6) of the Enforcement Policy. The licensee's corrective actions included revising the 10 CFR 50.59 safety evaluation implementation process, training engineering personnel on these changes, and submitting a license change for the use of containment overpressure.
E8.2
{ Closed) Unresolved item (URI) 50-263/97018-03(DRP): Questions concerning whether the low pressure instruments for HPCI and RCIC should be in TS Table 3.2.1.
The licensee has committed to address the HPCI and RCIC low steam line pressure isolation signals in the improved technical specification program (See NSP Letter dated February 3,1998, Subject: NSP Commitment to include RCIC System and HPCI System Low Steam Line Pressure isolation Signals in the Monticello TS). The inspectors noted that versions of 10 CFR 50.36 dated January 1,1994, and before did not include the criteria for establishing limiting conditions for operation as does the curreni 10 CFR 50.36(c)(2)(ii). The inspectors had no further concerns.
IV. Plant Support R1 Conduct of Radiological Protection and Chemistry (RP&C) Controls (IP 71750)
During normal resident inspection activities, routine observations were conducted in the area of radiation protection. The inspectors noted that radiation protection personnel provided excellent support to maintenance, engineering, and operations personnel during
_ _ - _ _ _ - _ _ - _ _ _ _ _ - _ _ -
___
_ _ _ _ _ _ _ _ - _
____-__
_--
.
.
i the refueling outage. Radiation protection support for the control rod drive withdraw line changeouts was excellent. Technicians were actively engaged in the job by assisting maintenance, construction, and engineering personnel.
R7 Quality Assurance in RP&C Activities R7.1 Quality Assurance Audits Durina the Refuelina Outaae (IP 71750)
The inspectors noted that general quality services auditors assessed plant personnel performance with respect to radiological protection. Concerns, such as maintenance i
personnel leaning over a contaminated area boundary and a protective clothing changeout area located over grating, were promptly and appropriately addressed. The auditors firidings were similar to the inspectors' observations.
S1 Conduct of Security and Safeguards Activities (IP 71750)
During normal resident inspection activities, routine observations were conducted in the areas of security and safeguards activities. No concerns were noted.
se
.>.
V. Manaaement Meetinas X1 Exit Meeting Summary On May 27,1998, the inspectors presented the inspection results to members of licensee management. The licensee acknowledged the findings presented. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary.
No proprietary information was identified.
i
!
l
_ ____ - ____ ___ _
.
_
.
_ _ _ _ _ _ _ _ _ - - _ - _ -
_ - _ _ - _ _ - -. _ _ _ - _ _ _ _ _ _ - - _ _ - - _ _ _ - - _ - _ _ - _ _ -
.
.
I PARTIAL LIST OF PERSONS CONTACTED l
Licensee
'
M. Wadley, Vice President Nuclear Generation
- M. Hammer, Plant Manager
- B. Day, General Superintendent of Operations K. Jepson, Superintendent, Chemistry & Environmental Protection L. Nolan, General Superintendent Safety Assessment E. Reilly, General Superintendent Maintenance
- C, Schibonski, General Superintendent Engineering A. Ward, Manager Quality Services L. Wilkerson, Superintendent Security J. Windschill, General Superintendent, Radiation Protection
.NRC
^
- J. McCormick-Barger, Chief, Reactor Projects Branch 7
- S. Thomas, Resident inspector, Prairie Island l
l
- Indicates those present during an exit meeting conducted on May 28,1998.
l l
l f
,
.
19 ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
I
!
INSPECTION PROCEDURES USED I
IP 37551:
Onsite Engineering
)
IP 61726:
Surveillance Observations i
IP 62703:
Maintenance Observations IP 71707:
Plant Operations IP 71750:
Plant Support IP 92700:
Onsite Followup of Written Reports of Nonroutine Events at Power Reactor Facilities IP 93702:
Prompt Onsite Response to Events at Operating Power Reactors
ITEMS OPENED, CLOSED, AND DISCUSSED Opened
,
50-263/98007-01(DRP)
IFl Continued investigation into the origin of the slight water-hammers in the HPCI injection piping,
.
Closed 50-263/96008-01(DRP)
VIO Three examples of inadequate test control were identified 50-263/96009-11(DRS)
eel Apparent violation of 50.59 due to undetected unresolved safety question 50-263/97002-01(DRP)
VIO Ventilation test not performed during suitable environmental i
conditions 50-263/97010-02(DRS)
eel Use of containment overpressure 50-263/97018-03(DRP)
URI Questions concerning whether the low pressure instruments for HPCI and RCIC should be in TS Table 3.2.1 50-263/97001 LER Improper Review Fails to identify Unreviewed Safety
,,
Question 50-263/98003 LER Transgranular Stress Corrosion Identified in Control Rod Drives l
l
. _ _ _ _
__.
. _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _
_
._
. _ _ _
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
.
i LIST OF ACRONYMS USED
_!
ASME American Society of Mechanical Engineers CFR Code of Federal Regulations CR Condition Report CRD Control Rod Drive DRP Division of Reactor Projects DRS Division of Reactor Safety eel Escalated Enforcement item GQS General Quality Services HPCI High Pressure Coolant injection IFl Inspection Follow-up Item
'IP Inspection. Procedure LER Licensee Event Report NCV.
Non-Cited Violation NDE Non-Destructive Examination NLO Non-Licensed Operator NRC Nuclear Regulatory Commission NRR Nuclear Reactor Regulation
,
_,.
~ NSP Northern States Power PMT.
Post-Maintenance Test PT Dye-Penetrant Testing RCIC Reactor Core Isolation Cooling RHR Residual Heat Removal RP&C Radiological Protection and Chemistry TS.
Technical Specification URI
. Unresolved item USAR Updated Safety Analysis Report VIO Violation WO -
Work Order
l l
I
!
!
i l
l
'
E