IR 05000237/1999011

Insp Repts 50-237/99-11 & 50-249/99-11 on 990522-0624. Non-cited Violations Noted.Major Areas Inspected:Operations, Maintenance & Plant Support

ML17191B402
Person / Time
Site:	Dresden
Issue date:	07/21/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML17191B401	List: IR 05000237/1999011 ML17191B400
References
50-237-99-11, 50-249-99-11, NUDOCS 9907280125
Download: ML17191B402 (21)

Text

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U.S. NUCLEAR REGULATORY COMMISSION Docket Nos:

License Nos:

Report No:

Licensee:

Facility:

Location:

Dates:

Inspectors:

Approved by:

9907280125 990721 1PDR ADOCK 05000237

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G PDR REGION 111 50-237; 50-249 DPR-19; DPR-25 50-237/99011 (DRP); 50-249/99011 (DRP)

Commonwealth Edison Dresden Nuclear Power Station, Units 2 and 3 6500 North Dresden Road Morris, IL 60450 May 22 through June 24, 1999.

K. Riemer, Senior Resident Inspector*

D. Roth, Resident Inspector B. Dickson, Resident Inspector M. Ring, Chief Reactor Projects Branch 1 Division of Reactor Projects

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EXECUTIVE SUMMARY Dresden Nuclear Power Station, Units 2 and 3 NRC Inspection Report 50-237/99011 (DRP); 50-249/99011 (DRP)

This inspection included routine resident inspection from May 22 through June 24, 199 Operations Leakage into the Unit 2 torus suggested that a potential secondary containment bypass existed. After questioning by the inspectors, the licensee performed an operability evaluation and concluded that the secondary containment was operable. The inspectors agreed with the licensee's conclusion. (Section 02.1)

Operator performance during plant evolutions and power operations was good. The operators performed the correct Technical Specification actions. The operations staff performed professionally. (Section 04.1)

Maintenance

The licensee correctly performed the maintenance activities directly observed by the inspectors. Additionally, the inspectors observed that the workers practiced good communication and good radiation worker practices. (Section M 1.1)

On May 22, 1999, while moving valves and pipe fittings from a lay down area to the job site for building intake canal cooling towers, a crane boom came into proximity to nonsafety-related Line 1263 (a 34-kV line) and caused the line to trip. This was an additional example of a problem with vehicle/heavy equipment control onsit (Section M 1.2)

The material condition of the Unit 3 emergency diesel generator's heat exchanger caused the licensee to enter an unplanned diesel outage. During the course of work to replace the heat exchangers, the licensee found cracks in one of the replacement heat exchanger's end caps. After the end of the inspection period, the licensee concluded that the cracks were the *result of overtorquing the end bell housing onto the heat exchanger. (Section M2.1).

The maintenance work on the Unit 2 emergency diesel generator was performed correctly and within the time allowed by the Technical Specifications. The licensee noted some problems in achieving cooling water flow and turbo oil pressures during post-maintenance testing and the problems were corrected. (Section M2.2)

Loose bolts penetrating the control room envelope permitted water from a heating coil to leak into the control room and onto the control room panels. The water damaged nonsafety-related chart recorders and displays. (Section M2.3)

Malfunctions of reactor feed pump ventilation and the 2A reactor feedwater regulating valve's actuatqr affected smooth operations and required, or will require, repai (Section M2.4)1

• Plant Support

Workers were following good radiological practices. The radiation protection personnel often spoke with and provided guidance to other radiation workers on keeping exposure low while performing tasks in the plant. (Section R4.1)

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Report Details Summary of Plant Status Unit 2 entered the inspection period at or near full power (2527 MWth).

On May 31, 1999, the operators reduced power to 790 MWe to perform circulating water pump bay cleanin On June 20, 1999, the operators reduced power to approximately 150 MWe to perform an entry into the drywell to add oil to the 2A and 28 recirculation pump Unit 3 entered the inspection period at or near full power (2527 MWth).

On May 23, 1999, the operators reduced power to about 20 percent to perform single loop operation to. repair the dust cover of the reactor redrculation MG se On May 31, 1999, the operators reduced power to 750 MWe due to high condensate-demineralizer inlet temperatur I. Operations

Conduct of Operations 01.1 * General Comments (71707)

Using Inspection Procedure 71707, the inspectors conducted frequent reviews of ongoing plant operations. Overall, the conduct of operations Was professional and safety conscious; the inspectors have detailed specific events and noteworthy observations in the sections belo During the inspection period, no events occurred that required prompt notification of the NRC per 10 CFR 50.72 or licensee event reports (LERs) per 10 CFR 50.7 Operational Status of Facilities and Equipment 02.1,Unit2 Torus and Primary Containment

, Inspection Scope (71707) The inspectors assessed an issue related to an increase in the water level in the Unit 2 torus caused by leakage from the condensate storage tank Observations and Findings In March of 1999, the licensee discovered that the Unit 2 torus water level was slowly increasing unexpectedly. The licensee determined that the source was not from the relief valves, but instead was from condensate system valve leakage. Operators also wrote Problem Identification Form # 01999-01899 on April 29, 1999, to document the increase of the Unit 2 torus water level caused by leakage through valve 2-1501-37 (from the condensate storage tank). The licensee listed this as an operator work aroun...

. ~

Following the discovery that the source of the water was the condensate storage tanks, *

the inspectors became concerned that reverse flow could bypass containment. In a post-accident environment, the pressure in the torus could force* contaminated water back into the condensate storage tank then into the environment. The issue appeared similar to the back flow from the control rod drive system discussed in Section E8.1 of this report. On April 30, 1999, the inspectors discussed the concerns about a containment bypass with the operators. The operators documented the discussions in Problem Identification Form # D1999-01929. The licensee initially assigned the investigation to engineering, with a 45-day due date. The licensee subsequently changed this item to a final due date of September 9, 199 On June 15, 1999, the licensee again documented leakage* into the. torus in the "Plan of the Day Risk Report." The licensee recorded that torus level increased after the 2-1501-37 valve was opened and subsequently closed to support emergency core cooling system jockey pump maintenance. The licensee closed (seated) the manual condensate storage tank valves to each low pressure coolant injection and core spray pump; however, this did not affect the in-leakage into the torus. The licensee believed that valve 2-1501-37 needed repair and elevated the issue to the "Plan of the Day" meetin An estimated 200 gallons per day were leaking into the toru The licensee initially believed that engineering might have addressed the issue of secondary containment integrity in a previously written evaluation (Engineering.

DOC ID 0005408107, "Potential Leakage Paths from Primary Containment Bypassing

.. Secondary Containment," issued May 9, 1997). On June 18, 1999, following discussions with the inspectors, the shift manager reviewed the previous evaluation and concluded that the evaluation did not specifically address the current issue. The operations and engineering staff then initiated Operability Evaluation 99-02 The operability e~aluation was completed on June 24, 1999. The evaluators concluded that secondary containment was operable in that the leakage remained within allowable dose limits. Therefore, the containment required no compensatory measures to assure that secondary containment worked properly. The inspectors compared the evaluation with LER 50-237/95021-02 "Maximum Thermal Power Exceeded Due to Inadequate Modification Safety Evaluation and Unanalyzed Secondary Containment* Bypass Pathway Discovered," and concluded that the results were consistent and acceptabl LER 95021 is discussed in Section E8.1 of this repor Conclusions Leakage into the Unit 2 torus suggested that a potential secondary containment bypass existed. After questioning by the inspectors, the licensee performed an operability evaluation and concluded that the secondary containment was operable. The inspectors agreed with the licensee's conclusion.

Operator Knowledge and Performance 0 Operator Performance Inspection Scope {71707)

During this inspection period th~re were several planned evolutions performed. The inspectors assessed the operators' performance during the execution of these evolutions. This assessment was complete*d through review of operators' logs, completed operating procedures, and, occasionally, direct observatio Observations and Findings During routine evolutions and throughout this inspectio.n period, the inspectors noted that conduct in the control room was professional and proper. The licensee used detailed and informative, written, turnover sheets to docur:nent limiting conditions for operation, abnormal equipment response, planned and recently completed maintenance, shift priorities, and required compensatory checks. The verbal turnovers were also professional and informative. Specific other comm~nts on the operators' performance are belo Single Loop Operations (Unit 3)

on* May-24, 1999, the licensee performed single loop operations to repair a loose dust screen on the Unit 3 recirculation pump motor-generator set. A review of operator logs*

and co.mpleted procedures indicated that all the appropriate actions required.by the*

Technical Specifications were completed. AJso, during single loop operations, th operators continuously monitored the neutron monitoring system and communicated with the station qualified' nuclear engineer to avoid exceeding any thermal limitation Emergency Diesel Planned Maintenance and Operability Surveillance (Unit 2/3).

On June 6, 1999, the licensee placed the Unit 2/3 emergency diesel generator out of*

service for planned maintenance. The inspectors confirmed that the.operators.

completed the action statements contained in Dresden Technical Specification 3. The inspectors also noted that the operators used proper three-way communication and self-checking techniques during the operability surveillance test Unit 2 Power Reduction On June 20, 1999, the licensee performed a power reduction to approximately 150 MWe to perform a. drywell entry to add oil to both.the A and B reactor recirculation pumps. The inspectors again noted good three-way communications and proper self-checking i techniques. Also, the inspectors noted that all control board manipulations during these evolutions were checked by peers. A review of the operator logs showed that all actions required by the Technical Specifications were me * Conclusions Operator performance during plant evolutions and power operations was good. The licensee performed the correct Technical Specification actions. The operations staff performed professionall *Miscellaneous Operations Issues (92901)

0 (Closed) LER 50-249/95014-02: Unexpected Operation of the Pressure Suppression Chamber to Drywell Vacuum Breakers Due to Inadequate Training and Procedure Deficiency. The licensee submitted Revision 2 to this LER on March 28, 1996. The LER described the event of September 9,.1995, when all 12 pressure suppression chamber 0 *(torus) to drywell va*cuum breakers opened as the operators were starting up Unit Operators were not expecting the opening when it occurred (at about 0.15 psid) because the operators were familiar with an upper limit of 0.25 psid from a routine Dresden Operations Surveillance procedure. The Technical Specifications required an upper limit of an equivalent of a 0.5 psid to open the vacuum breakers. The licensee concluded that the vacuum breakers functioned properly based on the settings from a Dresden Mechanical Surveillance procedure. The corrective actions included training the operators and revisiAg the operations-type procedures. There were no safety consequences to this event. This item is close *

(Closed) LER 50-249/96007-00: Failure to Perform Surveillance During Unit Shutdown*

Due to P,ersonnel Error. This LER documented an occurrence on June 21, 1996, when the licensee failed to perform two required neutron monitoring instrument tests during a shutdown. The licensee was performing the planned shutdown following failures in the safety-related 4-kV breakers (see Inspection Report 96006 for more information about the breaker issue). Technical Specifications Table 4.2.1, Note 2, required that source range monitor rod block functional tests and that intermediate range monitor rod block and scram calibrations be dc;me within the 7 days before a planned shutdown; the tests were last performed on June 5, 1996 (15 days before the shutdown). The LER documented that the event happened because the unit supervisor failed to communicate clearly with the instrument maintenance staff during the order to perform the require tests. The LER also documented contributing causes of the other licensed operators not subsequently questioning if the tests had been done. Corrective actions included training and counseling operators and revising shutdown procedures. As previously stated, Technical Specifications Table 4.2.1, Note 2, required that the tests be done within 7 days. Contrary to this, the licensee failed to perform the tests during or within 7 days prior to the controlled shutdown. This Severity Level IV violation is being treated as a Non-Cited Violation (NCV 50-249/99011-01(DRP)), consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as LER 50-249/9600.3 (Closed) LER 50-237/96013-00: Licensee Exceeds Technical Specification Time Clock During Calibration of 28 Main Steam Line Ra_diation Monitor due to Personnel Erro This LER documented the events of September 8, 1996, when operations and.*

maintenance staff failed to assure that a test of the 28 Main Steam Line Radiation Monitor was completed within the 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> allowed by Technical Specifications Table 3.1.1, Note 12. The LER noted that the test performance crossed shift turnover,

)

• other activities were going on in the control room, and that a maintenance worker was having some trouble with the calibration. The end result was that the radiation monitor was returned to service 15 minutes late. Operators failed to place the channel in "trip" after the expiration of the 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, thus violating Technical Specifications. This Severity Level IV violation is being treated as a Non-Cited Violation (NCV 50-237/99011-

. 02(DRP)), consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as LER 50-237/96001.4 (Closed) LER 50-249/96011-00: Unexpected Cycling of the Low Pressure Coolant Injection Minimum Flow Valve During Low Pressure Coolant Injection System Fill Due to Personnel Error. *This LER documented that while shutdown on August 23, 1996, during post-maintenance restoration (filling) of the 3A low pressure coolant injection heat exchanger, the low pressure coolant injection 3A and 38 minimum flow valves unexpectedly went closed after receipt of low system pressure signals as the heat exchanger was filling. The licensee declared the low pressure coolant injection and core spray systems inoperable, repositioned the valves, and vented the systems. The LER noted that the minimum flow valves would have automatically opened under accident conditions. Corrective actions included training and procedural improvements. This item is close II. Maintenance M1 Conduct of Maintenance *

M1.1 General Comments (62707).

The maintenance activities directly observed by-the inspectors were performed correctl Additionally, the worker~ were observed to practice good communication and good radiation worker practice M1.2 Crane for Cooling Towers Causes 34-kV line 1263 Trip During this period, the licensee assembled new cooling towers on the intake cana The work required use of a crane outside of the protected area. -On May 22, 1999, while moving valves and pipe fittings from a lay down area to the job site, the _crane boom came into close proximity to Line 1263 (34 kV). An arc jumped to the crane and was dissipated through the outriggers. Line 1263 tripped and reclosed. There

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were no injuries. The licensee documented the event in Problem Identification Form# D1999-02196, and issued Nuclear Operations Notification DR 99-022 to alert the other site At the end of the inspection, the root cause of the event was not complete. However, the apparent cause of the event was a lack of attention to detail on the part of the spotter that allowed the crane to get closer than 5 feet to the overhead line. Contributing to the event was the location of the lay down area under the 34-kV line. The contractor's lifting procedure was revised to increase the minium clearance to 20 fee Although line 1263 was not safety related, and the event occurred outside of the protected area, the inspectors were still concerned. Part of the inspectors' concern

• came from two other incidents involving fork trucks colliding with electrical lines and with a walkway. The licensee considered this an example of a problem with vehicle use at the station. At the end of the inspection period, the licensee was still developing better procedures and training for cranes, fork trucks, and other vehicles used on the sit M2 Maintenance and Material Condition of Facilities and Equipment M Unit 3 Emergency Diesel Generator Inspection Scope (62707)

The inspectors assessed the licensee's detection and response to a failure in the Unit 3 emergency diesel generator cooling water heat exchange Observations and Findings On June 3, 1999, the licensee declared the Unit 3 emergency diesel generator inoperable due to a failure (crack) in the cooling water heat exchanger. An alert non-licensed operator had noted the symptoms (low cooling water level) of the failure during routine round The licensee entered the limiting conditions for operation for the diesel generator and commenced replacing the cooling water heat exchangers. In the process of replacement, a maintenance worker discovered what appeared to be cracks in the end cap on one of the heat, exchanger * The licensee investigated how the heat exchangers could have been issued to maintenance staff with cracks present. In Nuclear Operations Notification DR 99-025, the licensee.documented that, in March of 1999, the heat exchangers had been quarantined with hold tags after initial receipt because of missing nameplates and documentation. When the heat exchangers were required for the emergent concern in June, the licensee performed an evaluation and accepted two heat exchanger Originally, the licensee-believed that an inadequate inspection may have been done on the heat exchangers. However, subsequent to the end of the inspection report, the licensee concluded that the cracks were the result of over-tightening the end bell housing onto the heat exchanger. The licensee sent the cracked end bell out for failure analysis, and the analysis was not complete at the end of the inspection period. Nevertheless, the licensee assigned corrective actions to check the other diesels' heat exchangers for the proper torque value Conclusions The material condition of the Unit 3 emergency diesel generator's heat exchanger caused the licensee to enter an unplanned diesel outage. During the course of work to replace the heat exchangers, the licensee found cracks in one of the replacement heat exchanger's end caps. After the end of the inspection period, the licensee concluded that the cracks were the result of overtorquing the end bell housing onto the heat exchanger.

M Unit 2 Emergency Diesel Generator Inspection Scope (62707)

The inspectors observed portions of the. work on the Unit 2 emergency diesel generator planned maintenance and discussed the progress with maintenance staf Observations and Findings The maintenance included pressure-testing buried cooling water piping and inspecting the engine. No significant material condition issues were discovered during the planned

• maintenance. The licensee accomplished the work during the time permitted by the Technical Specification On June 9, 1999, during testing, the operations staff noted that diesel generator cooling water flow was only 885 gpm, while the su.rveillance test required a "nominal 900 gpm."

A separate procedure, not executed at the time, required 900 gpm flow. The flow element was fl.ushed and recalibrated, the system then achieved 900 gp The first operability run also produced low turbo oil pressure. The licensee identified a sticking check valve and replaced it. The pressure then was correc Conclusions The maintenance work on the Unit 2_emergency diesel generator was performed correctly and within the time allowed by the Technical Specifications.* Some problems were noted in achieving cooling water flow and turbo oil pressures during post-maintenance testing. The licensee corrected these problem M2.3 Water Leak intb Unit 2 Main Control Room Inspection Scope (62707)

The inspectors reviewed the circumstances that led to a water leak into the Unit 2 main control room. *

• Observations and Findings The Leak and its Impact On June 2, 1999, water from a leak in the heating coils for. the Unit 2 East Turbine Building ventilation leaked into the main control room. The water damaged four nonsafety-related chart recorders and the power supply for two nonsafety-related digital display The water originated from a heating coil that was being restored to service after an extended period of being out-of-service. The water spilled on the floor above the control room.. The water then migrated into the control room by leaking around bolts holding

down the 125 VDC charger. The bolts were in the control room between the front and back panels, roughly at the 902-5 "B" section and were loose at the tim The water migrated along the cabling in the control room and eventually out of the annunciator tiles, along the upright control panel, and down the horizontal section of the 902-5 panel. The water caused a full 125 VDC positive ground, and impacted the source range monitor recorder, the core differential pressure recorder, and two operator-selected trend recorder Cause of loose bolts The licensee investigated to determine when the bolts were loosened. The 'last time that the chargers were recorded as being installed was 1985. No recent work could be located that could have led fo the bolts being left loos Corrective Actions The licensee used sealant around the bolts to fix the leaks. The sealant and the work package (Work Request # 990060330) were based on the work done to seal the control room envelope following discovery that the control room was not maintaining positive pressure (see Inspection Reports 96014 and 96016 for more information about control room sealing). The licensee also performed visual and smoke-assisted checks of the floor above the control room to determine if other leaks existed. None was found. The control room envelope was routinely tested to assure that it could perform its safety-related function of maintaining positive pressure, and the leaks into the control room had not been previously identifie No formal root cause report was assigned. The licensee's Events Screening Committee a_lso did not require that a Nuclear Operations Notification be written to notify the other Com Ed utilities of the event.* Subsequent to the inspection exit, the licensee issued a Nuclear Operations Notification on the subjec Conclusions Loose bolts penetrating the control room envelope permitted water leaking from a heating coil to leak into the control room and onto the control room panels. The water *

damaged nonsafety-related chart recorders and display M2.4 Other Material° Condition Issues Reactor Feed Pump Ventilation.

On June 9, 1999, the operators received a reactor feed pump ventilation alarm in the control room. This alarm was caused by the discharge damper for the reactor feed pump ventilation system closing unexpectedly. When the licensee opened the reactor feed pump ventilation access door, the damper re-opened. This was the third time this year that the nonsafety-related reactor feedwater pump system ventilation failed_. The licensee changed out the positioner on June 16, and also developed a plan to install a temporary modification by July 1 to prevent spurious damper closing. The licensee has

been working on critical ventilation system availability, and tracking the progress in the Shift Manager Tracking repor Reactor Feedwater Regulating Valve While recovering from the downpower on June 20, 1999, to add oil to the 2A and 28 recirculation pump, the operators noted the 2A feedwater regulating valve was oscillating in manual without any input signal. The licensee discovered a gross air leak on the valve actuator, which was causing local oscillations up to 20 percent. The operations staff has

• continued.to monitor the performance of the valve. In the field, the actuator pulses (to*

make minor adjustments to its position) about 40 times per minute. The licensee considered issues such as maximum permissible flow through a single feedwater regulating valve, signs of additional valve degradation, and parts availability, then developed a plan to change out the actuator. At the end of the inspection period, the licensee was planning a load drop to. perform the wor MS Miscellaneous Maintenance Issues (92902).

M (Closed) LER 50-237/96004-01: Main Steam Safety Valve 2-0203-4G As-found Lift Set Point Outside Tech Spec Limit Due to Set Point Drift. The LER documented that on October 5, 1995, the Main Steam Safety Valve 2-0303-4G opened at 1225 psig, which was outside of the values allowed by Technical Specification 4.6.E of 1228 psig to 1252 psi The licensee found no error or mechanical malfunction that caused the drift. The report concluded that the safety consequence was minimal. The inspectors also concluded that there was no safety consequenc The LER was submitted to the NRC late, following discovery on February 28, 1996, by engineering that the test performed October 5, 1995, showed the safety valve was not within the Technical Specifications. The maintenance staff and operations staff involved with the test at the time had not realized the need for an LER. The licensee reviewed *

additio.nal safety valve tests and identified an instance in 1993 when safety valve 2-0203-

. 4C opened at 1265 psig (13 psig above the maximum allowed) but an LER was not create The LER cited the Updated Final Safety Analysis Report to show that three (of eight)

safety valves were required to mitigate events and conform with requirements. The safety consequence of the 2-0203-4C valve opening slightly too high was, therefore, minima Section 72 of 10 CFR Part 50, required that licensees submit an LER for any operation

• or condition prohibited by the plant's Technical-Specifications, and that the LER be submitted within 30 days after the discovery of the event. Contrary to this, the licensee twice failed to submit LERs within 30 days of the original determinations in 1993 and 1995 that safety valves' were outside the Technical Specifications. This Severity Level IV violation is being treated as a Non-Cited Violation (NCV 50-237/99011-03(DRP}), co_nsistent with Appendi~ C of the NRC Enforcement Policy. Th.is violation is in the licensee's corrective action program as LER 50-23719600 r_;

M (Closed) LER 50-249/97007-00: Main Steam Safety Valve 3-0203-4G As-found Lift Setpoint Outside Tech Spec Limit Due to Set Point Drift. This LER described the discovery on May 21, 1997, that main steam safety valve 3-0203-4G opened at 1234 psig, whereas its required band was 1238 psig to 1262 psig. The licensee concluded that the cause was setpoint drift. The other three safety valves tested during this period passed. The LER noted that the Updated Final Safety Analysis Report requirec;l only three safety valves to conform to the American Society of Mechanical Engineer's Code ove~protection limits. The safety consequences were, therefore, minimal. The licensee identified no common mode of failure for the safety valve and previous "G" safety valve failures (see Section M8.1 ). This item is close M !Closed) LER 50-249/96012-00: Out of Tolerance Anticipated Transient Without Scram Time Delay Relay due to Inadequate "As-Found" Calibration Check Method. This LER documented the discovery on September 6, 1996, with Unit 3 in cold shutdown, that

. three of four low-low reactor water level Anticipated Transient Without Scram time delay relays were found outside of the Technical Specification tolerance. The licensee replaced the relays. The licensee then re-tested the removed relays and found them to be acceptable. The licensee concluded that the initial failures were attributable to human error in use of the stopwatch. The licensee switched to the use of a chart recorder to enhance time delay measurements. This item is close M (Closed) LER 50-249/97008-00: Inadvertent Unit 3 Scram While Shutdown During Performance of Instrument Surveillance due to Personnel Error. This LER documented that on May 26, 1997, a worker moved the controls for the wrong Average Power Range Monitor and caused a full scram. The unit was partially defueled at the time. The scram

• had no safety consequences. Corrective actions included training on a "Stop-Think-Act-Review" simulator to enhance self-checking and equipment manipulation. This item is close Ill. *Engineering ES Miscellaneous Engineering Issues (92903)

E (Closed) LER 50-237195021-02: Maximum. Thermal Power Exceeded Due to Inadequate Modification Safety Evaluation and Unanalyzed Secondary Containment Bypass Pathway Discovered. The LER, which the licensee submitted to the NRC on August 19, 1996; documented three major issues related to an inadequate review of a modification that installed seal purge lines from the control rod drive system to the reactor recirculation pump *

First, ah evaluation conducted in.197 4 for the modification did not correctly account for the control rod drive system flow to the reactor recirculation system pump. seal purge lines, resulting in a calculated core thermal power that was 0.8 megawatts (about 0.03%

of full power) too lo Second, the addition of the seal purge lines created potential unanalyzed secondary containment bypass pathways from the reactor recirculation pump seals through the seal purge lines to the condensate storage tanks.

Third, the control rod drive hydraulic control units could also be a source for back-leakage through the control rod drive system out to the condensate storage tank With regards to the first issue of exceeding maximum thermal power, the licensee could not identify any specific time when the reactor exceeded the maximum allowable powe However, the licensee hypothesized that the reactor may have exceeded the maximum

• power historically. The LER documented that a letter from General Electric to the Boiling Water Reactor Owners' Group concluded that the error in thermal power was not a safety issu Regarding the second issue of containment bypass, the LER stated that if the operating control rod drive pump was lost and the standby pump was not started, primary system fluid could potentially leak back through approximately one thousand feet of seal purge line and control rod drive hydraulic system piping, through the untested control rod drive pump discharge check valves or the sample panel, and to the outdoor contaminated condensate storage tank, the condenser, and the vented tank level standpipe located within the Unit 2 turbine building. The LER went on to state that if a loss-of-cooling accident was to occur with severe core damage, coincident with this event, 0.4 gallons per minute of primary system fluid could potentially bypass secondary containment and leak to the condensate storage tanks, condenser, and vented standpipe in the turbine building resulting in a potential airborne release to the environment. In this unlikely,

event, and without additional measure, the potential exists that the control room dose could exceed the General Design Criterion 19 limits. *

The licensee implemented procedure changes to isolate the control rod drive system via manual valves if a loss-of-cooling accident occurred and the control rod drive pumps*

were de-energized. The licensee eventually installed check valves as final corrective actio *

On March 29, 1996, the licensee estimated the dose due to reverse-flow through the

• excess flow check valves in the control rod drive hydraulic piping. In ComEd NuClear. *

Design Information Transmittal # S040-DH-0369, the licensee identified that both the estimated dose in the control room and the estimated dose at the low population zone exceeded the regulatory guideline *

The licensee subsequently revised the estimates on July 7, 1996, in Nuclear Design Information Transmittal #S040-DH-0377 by including an 8-hour delay for the reverse-flow to reach the contaminated condensate storage tank and to be released to the atmosphere. The previous estimate of March 29, 1996, used an iodi11e partition coefficient - the ratio of airborne-to-liquid chemical concentrations - of ten, and assumed a continuous release. The licensee used information in NUREG-series reports to determine partition coefficients based on the 8-hour delay: and selected a higher partition coefficient of one thousand. The new estimated doses were then within regulatory guidelines. The LER did not include the revised information from Nuclear Design Information Transmittal #S040-DH-0377; therefore, it did not show that the dose estimates were within guidelines.

The corrective actions for the second issue encompassed the third issue of the control rod drive hydraulic control units being a source for back-leakage through the control rod drive system out to the condensate storage tank Appendix B to 10 CFR Part 50, Criterion Ill, "Design Control," required that design control measures shall be provided for checking the adequacy of the design. Design control measures shall be applied to items such as thermal and accident analyse Contrary to the above, the licensee did not provide an adequate check of the design of the recirculation pump seal purge modification when the licensee performed an evaluation in 197 4. As a consequence, the licensee failed to identify that the design caused the thermal power calculations to be incorrect, and created a secondary containment bypass pathway. This Severity Level IV violation is being treated as a Non-Cited Violation (NCV 50-237/99011-04(DRP)), consistent with Appendix C of the NRG Enforcement Policy. This violation is in the licensee's corrective action program as LER 50-237/95021.-0.

E (Closed) LER 50-249/96015-00: High Pressure Coolant Injection System Inoperable Due to Instrumentation and Auxiliary Equipment Problems. This LER described how, on October 2, 1996, the Unit 3 high pressure coolant injection system failed a surveillance test due to high area tem*perature at the turbine inlet. Also, the high pressure coolant injection system's a1.ixiliary oil pump and the high pressure coolant injection system*~

emergency oil pump did not perform correctly during the surveillance. The licensee found that the temperature probe was not reflective of the high pressure coolant injection room temperature due to the probe's location. The licensee also revised the surveillance

.

I test to take an average of four temperatures, rather than declare the high pressure coolant injection system inoperable based on a single temperature. The auxiliary oil pump's problem (it unexpectedly tripped) was actually a correct response of the pump, based on pressure switch setpoints and normal fluctuations in oil pressure from the.

turbine-driven oil pump. However, the licensee identified in 1992 that the setpoint should be increased, but did not change the setpoint. The licensee could not identify why the emergency oil pump failed to start, nor could they repeat the failure. Subsequent to this LER, the high pressure coolant injection systems on Unit 2 and Unit 3 experienced a variety of failures. These failures, and the corrective actions, were discussed extensively in Inspection Reports 98009, 98014, 98021 and other reports, as well as the associated LERs. This item is close *

E (Closed) LER 50-237i97008-00: Unit Shutdown due to Degradation of Auxiliary Switches in 4-kV Circuit Breakers Caused by Design/Manufacturing Deficiency. This LER documented the shutdown of Unit 2 on April 10, 1997, following discovery that breakers for various systems, including the containment cooling service water system, were degraded. No failures had taken place on the breakers, but the licensee identi~ed *

that the mounting blocks for the breakers' auxiliary switches were cracked. The licensee then declared all the breakers of that type as "inoperablen and entered the various limiting conditions for operation. Repairs were made before either unit was restarte Supsequently the breakers' manufacturer issued a formal "Notification of Potential Defectsn under 10 CFR Part 21. The NRG documented review of this issue in Dresden Inspection Reports 97006, 97007, 97021, and in Quad Cities Inspection Report 97006, and 98009. This item is closed.

E (Closed) LER 50-237197001-01: Primary Containment Electrical Penetrations Never Subjected to Type B Local Leak Rate Test due to Break Down of the Modification Process. This LER described the discovery on January 9, 1997, that two primary containment boundaries on Unit 2 and two boundaries on Unit 3 had never been challenged by a Type B local leak rate test. The penetrations subsequently passed Type B local leak rate tests (no leakage was found). The penetrations were created by a modification that was completed March 3, 1983. The licensee found that the previous two local leak rate test coordinators were aware of the penetrations, but believed that the penetrations were not testable. The licensee committed in the LER to review and re-verify all the primary containment penetrations, and to revise the local leak rate test program. Because the penetrations had no leakage, the inspectors considered this to be of minor significance and not subject to formal enforcement actio E (Closed) LER 50-237/96021-00: Failure to Declare Refuel Floor Radiation Monitor Inoperable and Take Technical Specifications Required Action Due to Inadequate 10 CFR 50.59 Safety Evaluation. This LER documented the discovery on December 9, 1996, that on several occasions the refueling floor radiation monitor had been made inoperable by being placed on a nonsafety-related power supply. The licensee cited an inadequate historical (1993) 50.59 Safety Evaluation as the root cause, and the licensee noted that various process improvements to the evaluation process were S:ubsequently started. Recently, the NRC documented acceptable performance of 50.59 safety evaluations-in Inspection Report 9802 *

From July 9, 1995, to September 22, 1995, the Unit 2 "A" channel refuel floor radiation monitor was inoperable because it did not have safety-related power. The licensee identified that similar errors occurred in 1993 and 1994. In 1995, Technical Specification 3.2.D.2, required that standby gas be started if the monitor was not returned to operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Contrary to the above, the licensee did not start standby gas after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> because the licensee did not realize the monitor was inoperable. This Severity Level IV violation is being treated as a Non-Cited Violation (NCV 50-237/99011-05(DRP)), consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as LER

. 50-23711996-02 E (Closed) LER 50-237/97007-00: Low Pressure Coolant Injection Recirculation Loop Line Break Detection 900 psig R~actor Pressure Permissive Setpoint Set Outside Design Basis Limit Due to Personnel Error. On February 26, 1997, the licensee identified that the low pressure coolant injection loop select logic was nonconservative, and, therefore,

• inoperable. According to the licensee, this nonconservatism was caused when engineering personnel used the wrong design information to calculate the low pressure coolant injection loop selection initiation logic setpoint in 199 The inspectors verified that corrective actions for this issue were completed. The corrective actions for this event included redoing the calculation using the correct design limits and revising the calculation procedure to include the correct setpoint value. The licensee also changed administrative procedures to address the preparation of calculations and established a program to monitor and improve the quality of calculation *

Appendix B to 10 CFR Part 50, Criterion Ill, "Design Control," required that measures be established to assure that applicable regulatory requirements and the design basis are correctly translated into ~pecifications, procedures, and instructions. Contrary to the above, the licensee failed to translate the design basis of the recirculation loop selection into correct specifications, procedures, and instructions. This Severity Level IV violation is being treated as a Non-Cited Violation {NCV 50-237/99011-0G{DRP)), consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as LER 50-237/1997-00 IV. Plant Support R4 Staff Knowledge and Performance in Radiation Protection and Chemistry R Performance of Radiation Workers Inspection Scope (71750)

During plant tours, the inspectors assessed performance of radworkers and radiation protection staf Observations and Findings The inspectors observed the.maintenance workers following good radiological practices.

. Qften when the inspectors observed ongoing work in contaminated or high radiation, radiation protection personnel were obs.erved giving advice and guidance to other rad workers involve in the.tasked on how to keep exposure as low as reasonable achievabl Conclusions Workers were following good radiological practices. The radiation protection personnel often interfaced with and provided advice to other radiation workers on keeping exposure low while performing tasks in the plan *

'

V. Management Meetings X1 Exit Meeting Summary *

The inspectors presented the inspection results to members of license management at the conclusion of the inspection on June 24, 1999. 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.

PARTIAL LIST OF PERSONS CONTACTED.

Licensee K. Beverly, Licensing Engineer P. Boyle, Chemistry Manager L. Coyle, Shift Operations Superviso B. Hanson, Shift Operations Supervisor L. Jordan, LSS Training R. Kelly, Regular Assurance NRC Coordinator J. Kocek, Supply Chain Manager L. Licata, Engineering Administration Supervisor J. Mosier, Radiation Protection M. Pacilio, Work Control Management R. Peak, Design Manger *

P. Planing, Unit 1 Plant Manager

• G. Ponce, Maintenance Manager R. Rybak, NLA, Dresden Nuclear Power Station S. Stiles,* Assessment Management P. Swafford, Station Manager T. Yarbrough,.SVP Staff INSPECTION PROCEDURES USED IP 37551:

IP 61726:

IP 62707:

IP 64704:

IP 71707:

• 1p 71750:

IP 92901:

IP 92902:

IP 92903:

Onsife Engineering Surveillance Observations Maintenance Observations Fire Protection Program Plant Operations Plant Support Activities Followup - Operations

• Followup - Engineering Followup - Maintenance *

• Opened 50-249/99011-01 50-237/99011-02 50-237/99011-03 50-237/99011-04 50-237/99011-05 50-237/99011-06 Closed 50-237197008-00 50-249/96007-00 50-237196013-00 ITEMS OPENED, CLOSED, AND DISCUSSED NCV failure to perform surveillance during unit shutdown due to personnel error NCV * licensee exceeds Technical Specification time clock during calibration of 28 main steam line radiation monitor due to personnel error NCV licensee twice failed to submit LERs within 30 days NCV licensee did not provided an adequate check of the design of the recirculation pump seal purge modification

NCV failure to declare refuel floor radiation monitor inoperable and. take Technical Specifications required action due to inadequate 10 CFR 50.59 safety evaluation NCV low pressure coolant injection recirculation loop line break detection 900 psig reactor pressure permissive setpoint set outside design basis limit due to personnel error LER unit ~hutdown due to degradation of auxiliary switches in 4-kV circuit breakers caused by design/ manufacturing deficiency *

.LER failure to perform surveillance during unit shutdown due to personnel error LER licensee exceeds Technical Specification time clock during *

calibration of 28 main steam line radiation monitor due to personnel error 50-249/96011-00 LER unexpected cycling of the low pressure coolant injection minimum flow valve during low pressure coolant injection system fill due to personnel error 50-249/96015-00 LER high pressure coolant injection system inoperable due to instrumentation and auxiliary equipment problems 50-249/95014-0 LER unexpected operation of.the pressure suppression chamber to drywell vacuum breakers due to inadequate training and procedure deficiency

50-237195021-02 50-237196004-01 50-249/96012-00 50-237196021-00 50-237/97001-01 50-237197007-00 50-249/97007-00 50-249/97008-00 50-249/99011-01 50-237/99011-02 ITEMS OPENED, CLOSED, AND DISCUSSED (cont'd)

LER maximum thermal power exceeded due to inadequate modification safety evaluation and unanalyzed secondary containment bypass pathway discovered LER main steam safety valve 2-0203-4g as-found lift set point outside Technical Specification limit due to set point drif LER out-of-tolerance anticipated transient without scram time delay relay due to inadequate "as-found" calibration check method LER failure to declare refuel floor radiation monitor inoperable and take Technical Specifications required action due to inadequate 10 CFR 50.59 safety evaluation LER primary containment electrical penetrations never subjected to Type 8 local leak rate test due to break down of the modification process LER low pressure coolant injection recirculation loop line break detection 900 psig reactor pressure permissive setpoint set outside design basis limit due to personnel error LER main steam safety valve 3-0203-4g as-found lift set point outside Technical Specification limit due to set point drift LER inadvertent unit 3 scram while shutdown during performance of instrument surveillance due to personnel error NCV failure t9 perform surveillance during unit shutdown due to personnel error NCV licensee exceeds Technical Specification time clock during calibration of 28 main steam line radiation monitor due to personnel error 50-237/99011-03 NCV licensee twice failed to submit LERs within 30 days 50-237199011-04 NCV licensee did riot provided an adequate check of the design of the recirculation pump seal purge modification 50-237/99011-05 *

NCV failure to declare refuel floor radiation monitor inoperable and take Technical Specifications required action due to inadequate 10 CFR 50.59 safety evaluation

• ITEMS OPENED, CLOSED, AND DISCUSSED (cont'd)

50-249/99011-06 NCV low pressure coolant injection recirculation loop line break Discussed None detection 900 psig reactor pressure permissive setpoint set outside design basis limit due to personnel error 21