ML20138P395
| ML20138P395 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 02/26/1997 |
| From: | Jamila Perry COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| CAL, GL-82-12, GL-83-28, GL-90-03, GL-90-3, JSPLTR-97-0041, JSPLTR-97-41, NUDOCS 9703040282 | |
| Download: ML20138P395 (56) | |
Text
Commonu rahh I.dimn Company Drexlen Generating Nation bMN) North IJrevien Road Morrh, Il G)150 Tel Hi% 912 2920 February 26,1997 JSPLTR #97-0041 l
U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 i
Subject:
Dresden Station Units 1,2, and 3 Response to Dresden Independent Safety Inspection Report 50-237/96-201; 50-249/96-201 NRC Docket Nos.50-010. 50-237. and 50-249
Reference:
(a) Letter from J. M. Taylor, NRC Executive Director for Operations, ,
to J. J. O'Connor, Chairman and Chief Executive Officer, Comed, i dated December 24,1996, transmitting report ofIndependent Safety Inspection of Dresden Station.
(b) Letter from J. S. Perry, Site Vice President, Dresden Station, to H. L. Thompson, NRC Acting Executive Director for l
Operations, dated January 13,1997, Interim Response to i Independent Safety Inspection of Dresden Station.
This letter and its attachment constitute Dresden Station's response to the NRC Independent Safety Inspection Report.
Since our interim response was submitted on January 13,1997, we have had additional time to reflect on the Independent Safety Inspection Team's findings and conclusions. 4 We are very pleased that the Team identified improvement in all areas, and particularly pleased that this Team recognized operator performance as a noteworthy strength. We believe that these conclusions are a direct result of the improvement efforts implemented as part of the Dresden Plan which was completed in December 1996.
At the same time, based on our self assessments and the performance review conducted as part of the Dresden Plan closure process, as well as the findings of the Independent Safety [
Inspection, we recognize that the desired level of performance has not been achieved in all areas. To a large extent, the Independent Safety Inspection Team confirmed our ow 9703040282 970226 PDR ADOCK 05000010 G PDR u n e,m o,mnen, ll0 ilEhhh
h February 26,1997 USNRC/JSPLTR 97-0041 Page 2 of 5 assessment regarding the remaining areas of weakness and the actions necessary to impove performance. In at least one area, Engineering, the Inspection Team identified concems indicating the need for additional management attention to correct the performrace deficiencies to our satisfaction.
With the support of our Corporate Nuclear Operations Division, Dresden Station is systematically addressing each of the root causes, deficiencies and other issues identified in the Independent Safety Inspection Report.
Root Causes t
As described in our Interim Response to the Independent Safety Inspection Report -
(Reference (b)), we have already implemented or have underway actions to address the root causes identified by the Independent Safety Inspection. These actions include a i substantial upgrade, on a corporate basis, to our corrective action program, as well as site- I specific training to ensure that problems are identified, properly analyzed and effectively resolved. We are also taking corporate-wide action to address engineering support issues, as well as several Dresden-specific steps. Actions to address root causes are proceeding on the schedule indicated in our Interim Response.
Deficiencies and Unresolved items During the course of the Independent Safety Inspection, the NRC inspectors generated in excess of 2000 questions. As each question arose, the Dresden Independent Safety inspection Support Team provided the NRC with responsive information. In most cases, these responses resolved the inspector's question.
However, as noted in the Independent Safety Inspection Report, there were several cases for which, at the conclusion of the inspection, questions remained unresolved or in which performance deficiencies were identified. In each of these cases, totaling twenty-three deficiencies and six unresolved items, Dresden Station evaluated the inspector's concem and where appropriate, began to implement corrective action prior to the end of the inspection. As a result, corrective action for most of these deficiencies was implemented prior to or shortly after the conclusion of the inspection. When the inspection report was
4 February 26,1997 USNRC/JSPLTR 97-0041 Page 3 of 5 received, the specific deficiencies and unresolved items were reviewed and the adequacy of the corrective actions verified. The attachment contains a summary of these deficiencies and unresolved items, and a description of the corrective action taken in response, both for the specific deficiency and to address potential similar events.
Observations The Independent Safety Inspection also contained a number of performance
" observations". While these observations did not rise tc the level of a deficiency or unresolved item, Dresden Station believes that they are iriportant and is acting to address them. Therefore, to ensure that corrective action is tal.en for each of the observations, two independent reviews of the inspection report were performed and over 360 observations were identified and placed in a data base. Based on the process described above to address NRC questions during the actual inspection, the corrective action for these observations in most cases is complete. We are currently verifying the adequacy of the corrective actions and developing objective evidence of corrective action completion using a process similar to that reviewed by the Independent Safety Inspection team for closure of the Dresden Plan. Records of our actions to address each of the Independent Safety Inspection observations are available at the Dresden site for NRC review.
Dresden 1997 Operational Plan In August 1994, a critical, systematic review was performed to determine the causes of Dresden performance problems and identify the means for correcting them. Based on this review, the Dresden Plan was developed as the overall blueprint for raising the level of station performance. In December 1996, the site completed the Dresden Plan. While the actions taken under the Dresden Plan have improved station performance, we recognize that the desired level of performance improvement was not attained in all areas and that additional actions are necessary. Therefore, as actions under the Dresden Plan were closed, we developed and are implementing additional improvement actions under the 1997 Operational Plan (business plan).
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February 26,1997 USNRC/JSPLTR 97-0041 Page 4 of 5 The business planning process is the foundation for future improvements at Dresden Station. Development of the plan involved the use of continuous improvement teams consisting of both bargaining unit and management personnel. The 1997 Operational Plan sets forth initiatives to improve station performance, targeting the areas of human performance /eiror reduction, materiel condition, and outage execution as specific areas for improvement. The 1997 Operational Plan includes specific performance goals to i ensure accountability exists for specific performance improvements and to permit effective execution of the plan. The broad issues raised by the Independent Safety
- Inspection were already known to Dresden management and were factored into the development of the 1997 Operational Plan.
There is one exception. The Independent Safety Inspection team identified significant concerns with Dresden Station's control of calculations and with the overall performance of site and corporate engineering activities. The results of our assessment of these issues and actions planned to address them are summarized in our Interim Response (Reference (b)). Since these actions are generally beyond the scope of Dresden's Operational Plan, and involve both site and corporate-wide activities, they are being separately tracked, addressed, and acted upon in accordance with our previous correspondence with the NRC, including the NRC's November 21,1996 Confirmatory Action Letter.
4 Conclusion As noted in the Independent Safety inspection report, Dresden Station has improved the overall safety performance of the plant. To continue this improvement, the Station has implemented, or has underway, significant corrective actions for the root causes, deficiencies, unresolved items, and observations identified in the Independent Safety inspection Report. In addition, we have devoted the resources through our business planning process in order to resolve the broader issues remaining to be addressed.
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February 26,1997 ,
j USNRC/JSPLTR 97-0041 -
i . Page 5 of 5 We look forward to the opportunity to brief the Commission, its staff, or the Senior Staff ,
j managers on our response and to answer any questions you may have. We also will i
- continue to provide progress reports on the engineering improvement efforts during the >
] periodic Confirmatory Action Letter status meetings.
i l Sincerely, i 1
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] J, Stephen Perry j- Site Vice President i Dresden Station j JSP/ lad 1 Attachment i
I' cc: A. Bill Beach, Regional Administrator, Region III i P. L. Hiland, Branch Chief, Division of Reactor Projects, Region III J. F. Stang, Project Manager, NRR (Unit 2/3)
Senior Resident Inspector, Dresden OfYice of Nuclear Facility Safety - IDNS File: Numerical
ATTACHMENT CORRECTIVE ACTIONS FOR ISI DEFICIENCIES AND UNRESOLVED ITEMS During the Independent Safety Inspection, the NRC Inspection Team raised several questions involving potential noncompliance with NRC requirements. For those issues where the potential compliance issue could not be readily resolved, Dresden Station implemented prompt corrective action to ensure compliance with NRC requirements and to prevent recurrence of these or similar events. In nearly every case, these corrective actions were implemented prior to, or shortly after completion of the Independent Safety inspection.
When the NRC issued the Independent Safety inspection Report, twenty-three deficiencies and six unresolved items involving potential noncompliance with NRC requirements were identified.
In response, Dresden reviewed the corrective actions taken when the questions were raised during the Independent Safety Inspection to confirm that the action taken or underway was adequate. In a few .:ses, these corrective actions were augmented.
This Attachment summarizes the issues and Dresden's response to each of the deficiencies and unresolved items identified in the Independent Safety Inspection Report. The "lSI Summary" for each deficiency or unresolved item is Dresden's attempt to capture the essence of the issue identified by the Inspection Team using excerpts from the report. The corrective actions are intended to address the full scope of the issue identified by the Team, and not just the ISI Summary. For convenience, we are tracking these deficiencies using the numbering system initiated in the Independent Safety Inspection Report.
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DEFICIENCY 96201-01 ISI Summary:
The team identified that a change to a Unit I annunciator procedure contained an incorrect value for the low count failure alarm setpoint, stating that the annunciator was set to allow 30 minutes before alarming. The annunciator is actually set to allow 10 minutes before alarming. Instrumentation and control personnel encountered problems during alarm setpoint change activities, and did not change the alarm setpoint as planned. This failure was not communicated to operations personnel to ensure that they did not revise the procedure to reflect the intended setpoint change. The j instrument setpoint change and the procedure revision process did not administratively prevent this l type of problem. Failure to maintain an adequate procedure is contrary to 10 CFR Part 50, Appendix B, Criterion V.
1 Corrective Action:
The annuciator procedure referred to above is associated with a low count alarm with a Unit SPING 1
Monitor. Because the SPING Monitor is located in a low background area, it routinely generates momentary low count alarms (less than 52 counts in 10 minutes). To allow the monitor to operate more efficiently in a low background area and thereby eliminate the nuisance alarms, the low count l alarm delay time was to be increased from 10 to 30 minutes. The change required Instrument l Maintenance personnel to reprogram the SPING memory chip and Operations Department l personnel to revise the setpoint description in three procedures. While the setpoint descriptions l were changed and the revised procedures implemented, the actual alarm time delay was not l changed because of problems encountered during reprogramming of the SPING Monitor memory chip. Although this resulted in the alarm setpoint description in the procedure being in error, it did not affect the operator's actions in response to a SPING Monitor alarm or the safety of personnel in the area monitored by the SPING.
l The following corrective actions were taken in response to this deficiency:
(1) To ensure compliance with NRC requirements:
l Immediately following identification of this deficiency, the alarm setpoint descriptions in
- the affected procedures were revised to reflect the correct time delay value (10 minutes).
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(2) To prevent recurrence of this or similar events:
The procedure revision process was reviewed and DAP 09-02, " Procedure and Revision Processing," was changed to require coordination between departments prior to implementing a new procedure when more than one department is afTected.
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DEFICIENCY 96201-02 ISI Summary:
The team determined by a review of radiation work permit (RWP) Folder 96-2020, Revision 2, and by discussions with a radiation protection shift supervisor (RPSS), that on three occasions (10/5/96 to 10/6/96) workers entered the Unit 2 hotwell area before determining the current radiological conditions in the work area. The last documented survey at a similar power level was performed on March 3,1994. Workers were briefed on the radiation levels in the work area using this survey and were briefed on contamination levels in the work area using a radiological survey dated July 24, 1996. These surveys did not record radiological airborne levels in the work area. An airbome survey was not performed by the licensee for this task. The licensee's position was that their historical data indicated that the radiation and contamination levels in the work area have remained constant in the past. The team concluded that the licensee did not assess the potential radiological hazards present in the work area, the concentration of radioactive material, or the radiation levels at the current power level before the workers entered the work area. Failure to survey the work area and assess the potential radiological hazards were contrary to the requirements of the RWP and 10 CFR 20.1501(a).
On October 31,1996, the team observed workers decontaminating a section of the Unit 3 reactor building overhead and asked to review the survey of this work area. Radiation protection management informed the team that they did not perform a survey before the work activity. The failure to survey the work area and assess the potential radiological hazards was contrary to the requirements of 10 CFR 20.1501(a), and represents an additional example of Deficiency 50-237(249)/96201-02.
Corrective Action:
The following corrective actions were taken in response to this deficiency: l (1) To ensure compliance with NRC requirements:
In the first example, work had already been completed under RWP 96-2020 when the l deficiency was identified. To ensure the radiological hazards in the hotwell area were l properly assessed for this work, a new survey was perfomied on October 8,1996, that included dose rates, contamination levels, and airbome concentrations. These survey results were found to be consistent with the results on the historical survey maps used for the initial l
briefings given prior to the hotwell entries on October 5,1996 and October 6,1996. i In the October 31,1996 example, work was still in progress in the overhead of the Unit 3 Reactor Building at the time the deficiency was identified. Work was immediately stopped l
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and a new survey was completed to assess the potential radiological hazards prior to recommencing work.
(2) To prevent recurrence of these or similar events:
Although the October 8,1996 survey results confirmed the accuracy of the historical surveys, the Radiation Protection Department revised its Policy on the use historical surveys in steam sensitive areas to require a new survey if no current survey data is available (i.e., a survey less than one month and seven days old), if the worker has to go above ground level, or if system breach will take place. For all other areas, historical survey data will not be used. Rather, appropriate survey measurements will be performed.
A letter describing this new policy was provided to all Radiation Protection Supervisors and Technicians on October 17,1996. Follow up reviews of entries in three steam sensitive l areas found that all three had properly documented current surveys.
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DEFICIENCY 96201-03 ;
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Radiation Work Pennit 96-2020 did not specify a stay time for entry into the hotwell area. The I failure to specify a maximum stay time on the RWP before beginning the job is contrary to the i j requiremen'.s of Technical Specification 6.12.2.
Il i The team determined after reviewing the Performance Improvement Form (PIF) summary dated !
i October 3,1996, that the licensee identified three opened LHRA doors and one additional LHRA !
i problem subsequent 4 February 1,1996. On March 15,1996, the licensee identified unsecured
- lead blankets covering a dnun of sludge in the Unit I radwaste yard. On August 30,1996, the licensee identified that the Unit I west reactor equipment drain tank room LHRA was unlocked. -
On September 18, 1996, licensee staff identified that the LHRA door to the "B" reactor water cleanup heat exchanger room was open. On September 27,1996, radiation protection personnel i l discovered that the door to the Unit 2/3 m imum recycle demineralizer LHRA door was unlocked.
]- The failure to maintain these door:, or areas ;ocked er controlled is contrary to the requirements of i 10 CFR 20.1601 and Technical Specification 6.12.2 and represents an additional example of
- Deficiency 50-237(249)/96201-03.
t Corrective Action:
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The following corrective actions were taken in response to this deficiency:
i (1) To ensure compliance with NRC requirements:
}. Upon subsequent investigation, it was determined that the Radiation Protection Supervisor }
{ had orally provided the work group with a stay time during his pre-job brief on October 5,
! 1996, based on a late entry into the RWP log (October 6,1996) which noted, among other j things, the allowable stay time for the work.
j Immediately following the identification of each individual high radiation area event, action was taken in response to the specific event in order to restore compliance with the Technical Specifications and 10 CFR 20.601. For example, these actions included:
- Moving the identified drum to radwaste and properly storing it in a locked high radiation area;
- The Unit I west equipment drain tank room was maintained under surveillance
- until a hasp and lock were installed.;
i e The "B" RWCU room was maintained under surveillance until a hasp and lock j were installed; and 1
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- The Maximum Recycle room was maintained under surveillance until a hasp I and lock were installed. l (2) To prevent recurrence of these events or similar events Only Radiation Protection Technicians are allowed to control high radiation area keys.
In addition, a walkdown of all High Radiation Area / Locked High Radiation Area (IIRA/LHRA) doors was performed and a list a materiel deficiencies prepared. To date,26 of 32 of these deficiencies have been corrected and the remaining deficiencies are being tracked by Action Requests. Additionally, the 1997 site Operational Plan (RP06-01.003) addresses this issue.
Finally, a root cause evaluation was performed for the high radiation area events as a group, and the following additional corrective actions are being implemented:
Worker awareness training was given by the Radiation Protection Manager covering the events, the significance of the high radiation area program as it relates to the Technical Specifications, management expectations, and a review of the applicable sections of DAP 12-4, "High Radiation Area Control," focusing on the worker's responsibilities. Additionally, the 1997 site Operational Plan addresses this issue (RP06-01.002).
HRA/LHRA controls were removed from six areas upon reduction of the dose rates to less than 100 mrem /hr. Further reductions are planned for D3R14. Additionally, the 1997 site Operational Plan addresses this issue (RP06-01.005).
The Lead Technical Health Physicist has been named as the owner of the high radiation area program.
The Radiation Protection Greeter policy has been revised to include reminders of high radiation area controls and worker responsibilities.
Swing gates have been placed as barricades to unlocked high radiation areas in order to heighten worker awareness. The 1997 Operational Plan addresses this issue (RP06-01.001). Alarms will be installed on these gates.
High radiation area awareness training will be added to cycle training for Operations, Maintenance and Engineering. This is addressed in the Operational Plan (RP06-01.006).
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DEFICIENCY 96201-04 ISI Summary:
The team identified that the entrance way to the radioactive material area surrounding the Unit I contaminated demineralized water storage tank was not conspicuously posted as radioactive material. Additionally, an area along the side of the tank had a 2 to 3 ft opening that was not barricaded and a caution sign that was missing the words " radioactive material." The failure to -
have a conspicuous radioactive material posting is contrary to the requirements of 10 CFR 20.1902(e).
Corrective Action:
The following corrective actions were taken in response to this event:
(1) To ensure compliance with NRC requirements:
A Radiation Protection Technician was immediately dispatched to investigate when the issue was identified to the Radiation Protection Depanment. The posting was immediately corrected.
The Radiation Protection Technician determined that the wind had blown down a stanchion causing a ponion of the barrier around the demineralized water storage tank to fall to the ground. He picked up the stanchion and restored the barrier. When the temporary area was removed two days later, the postings were placed on the tank manway covers to provide more wind resistant mountings.
(2) To prevent recurrence of this or similar events:
The need to consider wind effects when placing temporary barriers was discussed with the Technicians.
The station's posting signs are being replaced with signs that have the required information printed directly on the sign rather than using inserts. This will eliminate the possibility of the inserts blowing out of the signs during the high winds occasionally experienced at the site.
Weekly inspection of all outside radiological postings was added to Radiation Protection Technician routines.
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DEFICIENCY 96201-05 ISI Summary:
l The team identified that 75 LilRA keys listed on page 2 of Form 12-04B were not properly documented as inventoried on September 21,1996 as required by Procedure DAP 12-04. This
- failure to document an inventory was contrary to the requirements of Technical Specification 6.2.
Overall, the licensee properly maintained the radiation shielding program. However, after review of four temporary shielding packages, the team identified that in all cases documentation was not
- completed in accordance with Procedure DAP 12-12. In several cases, the licensee did not perform
] pre- or post-shielding surveys, and in one case the licensee did not complete an exposure estimate
) to install the shielding. This failure to follow Procedure DAP 12-12 was contrary to the
- requirements of Technical Specification 6.2, and represents an additional example of Deficiency 50-237(249)/96201-05.
4 The team toured work areas and inteniewed licensee radiation workers about their responsibilities 1 for work in the radiological posted area (RPA). The team identified that certain licensee workers inteniewed were not aware of the radiological conditions in their work areas contrary to the
- requirements of Procedure DAP 12-25, which specifically requires plant workers to be familiar j with radiological conditions and postings in work areas. Not knowing the conditions was contrary
, to the requirements of Technical Specification 6.2, and represents an additional example of
] Deficiency 50-237(249)/96201-05.
! The team and a licensee representative toured the Unit 2 torus area, and observed housekeeping material from completed jobs in Bays 1 and 8 contrary to what is required by Procedure DAP 12-
- 25. The failure to return the work area to its "as found" condition was contrary to the requirements i j of Technical Specification 6.2, and represents an additional example of Deficiency 50- l 237(249)/96201-05.
i Corrective Action:
The following corrective actions were taken in response to this deficiency:
(1) To en re compliance with NRC requirements:
Immediately following the identification of the potential inventory deficiency, an inventory of locked high radiation area keys was performed. This inventory was completed 3 satisfactorily and properly documented. In addition, a review was performed of the j inventories conducted between September 21,1996 and the date of discovery and these
- inventories were verified to have been properly documented.
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The shielding coordinator completed a self audit of all 1996 shielding permits. Packages l
with incomplete documentation were updated with the required information. Because the ;
specific workers who were unaware of the radiological conditions were not identified '
during the inspection, personalized training was not pc nible. As an alternative to ensure all workers entering the RPA understood radiological coedhiore in there area, the Radiation Protection Department reinstituted the " Greeter" progr.un. Greeters are RP technicians i stationed at the RPA access point who challenge workers going into the plant on radiological conditions at their work site and on their responsibilities as radiation workers, including their responsibility to retum the work area to the as found condition.
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The Unit 2 torus basement was cleaned and partially decontaminated to reduce contamination levels.
Specific training was conducted for Radiation Protection Technicians on basic radiation
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protection practices, procedures adherence, and management and departmental l
expectations. General training was provided and a letter transmitted to all Radiation l Protection Supervisors and Technicians regarding management's expectation for l
maintaining a continuous presence at locked high radiation area doors until all werkers !
leave the LHRA. )
Specific training was provided on management's expectation for RWP contents. Surveys were performed and the required information was included in the RWPs. j (2) To prevent recurrence of these or similar events:
The inventory of locked high radiation area keys has been added as a sign-off on the RP Supervisors Shift Tumover form.
The Greeter Program was formalized in Policy Memo ADM-10. In addition to providing expectations for greeters, it provides actions to be taken when a worker is encountered who is not prepared to enter the RPA and requires trending of such events.
A RP Rov.er program was also initiated. This program provides RP Technicians who tour the plant looking for radiological deficiencies, observe radiological work practices, provide coaching to the work force, support short duration jobs and document identified deficiencies.
Senior managers and the RP staff performed overview and self evaluations of RP activities.
Identified deficiencies are being trended to identify underlying causes.
A predefined surveillance has been established in EWCS for management to communicate procedure adherence expectations on a monthly basis. Each month a different topic is chosen and discussed with RP personnel.
Policy Memorandum ADM-11 was prepared to address procedure use and adherence expectations in the RP Department and provided to department personnel.
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Several high use RP procedures were are now required to be in hand when performing tasks controlled by those procedures. ' Such procedures cover surveys, posting areas, and
! mstrument source checking and operation. I i
The Radiation Protection Manager (RPM) has established a program to track and trend both
- poor and good radiological work practices. When such an event is identified, a letter is i j provided to the worker's supervisor describing the worker's good or bad performance. Poor j j practices can result in the individual being locked out of the RPA until a meeting is held i with the individual, his supervisor, and the RPM.
Finally, the RP department is tracking Procedure Adherence and Human Performance PIFs to evaluate the effectiveness of these actions.
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DEFICIENCY 96201-06 ISI Summary:
In March 1995, the licensee performed a " clean sweep" survey for uncontrolled radioactive materials outside the RPA and identified 450 items. Between March 3 and June 14,1996, the licensee stafT discovered approximately 50 additional uncontrolled items during a second clean sweep survey. Six additional iteras were identified before the team's arrival. On October 10,1996, the licensee identified a seventh item while the team was reviewing the radioactive material release program.
The licensee stated that these seven cases were recent problems, and not items missed during the prior " clean sweep" surveys. The most significant example was identified on September 23,1996, when a stanchion labeled as radioactive material was found in an uncontrolled area of the Unit 2 turbine building. Contamination levels were 60,000 dpm/100cm2 fixed contamination and 2000 to 5000 dpm/100cm2 loose contamination. The failure to maintain control of radioactive material is contrary to the requirements of 10 CFR 20.1802 (Deficiency 50-237(249)/96201-06). After reviewing the root causes for each problem, the team concluded that the licensee did not implement adequate corrective actions after the prior occurrences.
Correct 6c Action:
The following corrective actions : vere taken in response to this deficiency:
(1) To ensure compliance with NRC requirements:
As with previous items of radioactive material discovered onsite but outside the RPA, a Radiation Protectica Technician immediately took control of the item and retumed it to the RPA.
I (2) To prevent recurrence of these or similar events:
Eliminating radioactive material control events is an ongoing management focus area.
Numerous articles have been written in the station newspaper ar.d " tailgate" training sessions conducted to increase personnel awareness of the need to properly control radioactive material. Approximately 1000 ink pens with the logo " RAM Control - Our Station Goal" have been passed out to station personnel as a reminder of management's l expectations. The RPM delivered a presentation on radioactive material control and worker's responsibilities at the site Vice President's monthly meeting on January 17,1997. j l
l Low level surveys on dumpsters and vehicles leaving the protected area are now performed in order to increase the probability of detection of material.
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l The RP Greeter is being utilized to (1) control material entering the plant which may l inadvertently become contaminated and then be removed, and (2) ensure that material I l entering the plant is authorized by an RP Supervisor.
The RP Rovers were created to provide oversight of Rad Workers. Among other things, the l Rovers are responsible for assisting work groups on getting their material out of the RPA properly and for identifying practices that might lead to the loss of control of radioactive ;
material.
A stanchion control policy has been implemented to prevent contaminated stanchions from !
inadvertentlyleaving the RPA.
A benchmarking trip was conducted at the V. C. Summer plant. The Radiation Protection department will evaluate incorporating outside RPAs and a hot tool facility based on observations made at V. C. Summer.
The Radiation Protection Manager reinforced expectations regarding locking out . material J from the RPA with Radiation Protection Shift Super ? sors. I I
A barrier was installed on the second floor of the Unit 2 side of the Turbine Building to I better separate RPA and non-RPA portions of the Turbine Buildmg. I i ;
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DEFICIENCY 96201-07 ISI Summary:
In general, the team determined that the licensee properly posted and labeled radioactive material in accordance with regulatory requirements. However, during tours of the RPA, the team identified two bags labeled as radioactive material that did not have the required regulatory information recorded on the label. The first involved a scaled radioactive material bag discovered on September 30,1996. The second involved a bag discovered on October 7,1996. Radiological conditions were not recorded or posted; however, the licensee determined that the bags contained contamination levels less than 3000 dpm/100cm2 and dose rates ofless than one millirem per hour. The licensee properly controlled both bags; however, the failure to provide sufficient information, such as, radiation levels, quantity of radioactivity, and date surveyed, is contrary to the requirements of 10 CFR 20.1904.
Corrective Action:
The following corrective actions were taken in response to this deficiency:
(1) To ensure compliance with NRC requirements:
A RP technician was immediately dispatched to place proper information on both bags when the inspector notified the RP department of the deficiency.
(2) To prevent recurrence of this or similar events:
DAP 12-33, " Control of Radioactive Material," and DRP 5010-01, " Radiological Posting and Labeling Requirements," were revised to improve control of radioactive material by eliminating the exemption to not tag radioactive material in bags inside contaminated materials.
The RP Rovers were tasked with the responsibility to look for radioactive unlabeled or improperly labeled radioactive material in the RPA.
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DEFICIENCY 96201-08 ISI Summary:
A review of the PM and surveillance tasks from April 30 to November 1,1996, revealed no overdue tasks. The team subsequently determined that some of the predefined activities that became due during this period were either deleted or deferred. Ilowever, several predefined tasks were not deferrea in accordance with Procedure DAP 11-02, Revision 26," Preventive Maintenance and Predefine Program." Specifically, nine (safety and nonsafety) predefined tasks were overdue and no action requests (ARs) were initiated as required by the procedure.
Corrective Action:
The following corrective actions were taken in response to this deficiency:
(1) To ensure compliano with NRC requirements:
The predefined activities referred to were identified in the October 28,1996, Plan of the Day publication and were being discussed at the Plan of the Day meeting being observed by the inspector. Action Requests for these past due predefines were initiated on October 29, 1996. A PIF (Trend 96-10207) was written on October 29,1996, to address the DAP 11-02 procedural non-compliance associated with the lack of" Action Requests" or " Attachment
'C' Deferral Requests" for the listed predefines. ;
A review of the predefine data base was performed which concluded that no critical dates i for required predefines had been exceeded.
l (2) To prevent recurrence of this or similar events:
The Plant Testing Unit revised DAP 11-02, " Preventative Maintenance and Pre-define l Program," to clarify the time allowed to submit these documents. This DAP requires the ;
submission of an administrative action request if predefine cannot be completed on or before its due date.
The Plant Testing Unit now tracks predefines to identify past due predefines.
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I DEFICIENCY 96201-09 ISI Summary:
The oil sampling frequency of the ECCS system keep fill pump was changed from a 6-month interval to an 18-month interval on the basis of a recommended PM frequency template developed by the corporate organization, without considering plant-specific data. Specifically, the current Unit 2 ECCS keep fill pump oil analysis sample indicated possible bearing wear (later identified as slinger ring wear). No documentation was provided that indicated this oil analysis had been dispositioned before extending the sampling frequency. In addition, a PlF was not written when the results of the oil sample exceeded the acceptance criteria as required by Procedure DAP 02-27, "The Integrated Reporting Process"(Deficiency S0-237(249)/96201-09).
The team identified two events that challenged the operability of the Unit 2,125 VDC batteries. 1 Both of these events involved low battery room temperature conditions during the winter months, !
and revealed a design weakness in the Unit 2,125 VDC battery room heating system. Although the i licensee had written a PIF for the first of these two events, which occurred on January 19,1996,no PIF was written for the second event, which occurred on February 3,1996. The licensee took no i corrective actions as a result of the first event to prevent recurrence. The failure to initiate a PIF is ;
contrary to the requirements of Procedure DAP 02-27, and constitutes an additions. example of I Deficiency 50-237(249)/96201-09.
1 Work Request 96009693601, which was implemented to replace power pack Number 7 on DG 3, did not identify the as-found condition. A coolant supply line connection was tightened on power pack Number 10 and documented under minor maintenance / adjustments without documenting the ;
measuring and test equipment (M&TE) used. The loose coolant line resulted in approximately 13 gallons of water in the DG 3 oil box. A PIF was not initiated to document this condition. The coolant lines to the other power packs were checked using a torque wrench; however, the work l
performed was not documented. The failures to initiate PIFs for these instances is contrary to the l requirements of Procedure DAP 02-27, and constitute additional examples of Deficiency 50-237(249)/96201-09. i Corrective Action; 1
The following corrective actions were taken in response to this deficiency- 1 (1) To ensure compliance with NRC requirements:
To clarify the threshold at which problems are to be reported, DAP 02-27, "The Integrated Reporting Process (IRP)", has been revised to provide more concise direction for site personnel regarding Performance Improvement Form (PIF) initiation criteria. This revision also incorporated Maintenance Preventable Failures (MPFFs) as a criterion for PIF initiation. This procedure revision became effective on October 25,1996.
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Dresden Engineering management has taken several steps to encourage PIF initiation within the Engineering Department. Engineering Senior Management met with engineering organization personnel in order to communicate expectations for PIF initiation and review of the PlF database for 1996 was performed. Applicable Engineering procedures were revised by January 31,1997, to clearly delineate management expectations for PIF generation by Engineering personnel when design discrepancies are identified. During the first eight (8) months of 1996, the engineering organization initiated an average of 49 PlFs per month. During the last four (4) months of 1996, the average increased to 93 PIFs per month, almost double the previous number, indicating that personnel are now more sensitive to PIF initiation requirements.
Nuclear Engineering Procedures were revised to provide specific direction on action (including instructions for documentation and reporting problems on PIFs) to be taken whenever a potential 1 design basis discrepancy is identified. These revisions and associated training were completed by January 31,1997.
(2) To prevent recurrence of this or similar events:
In April 1997, Dresden Station will implement a new Corrective Action Program which was developed by representatives from all six (6) Commonwealth Edison nuclear sites and the corporate office. Thes:: representatives reviewed state-of-the-art corrective action programs in the industry to establish a new corrective action process for the entire Comed nuclear program.
The new process includes several improvements over the current program. It clearly delineates
- and standardizes the threshold for problem identification through Performance Improvement Form (PIF) initiation, and also incorporates a common PIF database that will provide the site with greater versatility in coding, sorting, and analyzing PIF data (PIFs are the main problem-reporting mechanism used at Dresden Station). The corrective action process team has also developed a new procedure that outlines trending analysis requireraents, including a review of industry operating experience for operational events. The new corrective action process
- incorporates human error reduction methodology developed by Failure Prevention International l
(FPI).
The following actions have been initiated to ensure that management expectations are clearly
. understood:
Operations Operations has established a fixed period of time in normal cycle training to discuss and reinforce management expectations. Also a fixed amount of time has been established to cover human performance trends / issues as well as corrective actions that are being implemented so that operators understand and support these activities.
Operations Shift Manage utilize routine crew briefs to reinforce management standards and personal accountability associated with those standards.
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Maintenance Since September 1996, all new maintenance supenisors have been provided training on Station and Maintenance Standards and Expectations. This has specifically included expectations in the Radiation Protection Area.
Maintenance and Station Standards and Expectations are reinforced through:
Weekly staff meetings between the Maintenance Superintendent and Maintenance management and supenision.
Pre-job briefings provided to craft personnel by Maintenance supervision.
Scheduled weekly shop meetings between Maintenance masters, supervision and craft personnel.
Worker knowledge and compliance to Standards and Expectations is further being monitored and reinforced during ongoing Management By Walking Around (MBWA) tours by Maintenance management and supenision. Coaching and/or remedial corrective actions are being taken during the tours to correct identified deficiencies immediately.
In December 1996, Maintenance implemented a required reading policy process as part of the Maintenance Supenisor Policy Manual to ensure understanding of changes to programs and processes as well as an additional method for communication of Standards and Expectations to i management, supenision and craft.
Engineering l
Engineering Management expectations are reinforced at daily " leads meetings" and weekly
" tailgates" with engineering personnel. In addition, these expectations have been included in the first quarter Engineering Support Personnel Training (ESPT). This training began February 20, 1997 and is scheduled to be completed by March 28,1997.
During the first half of 1997, Engineering Performance Improvement training will be developed and implemented. These sessions will be attended by Engineering Department Personnel and will focus on reinforcing improved standards and expectations for the Engineers.
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An Engineering Standards and Expectations IIandbook will be issued by the end of the Second j Quarter,1997. The handbook will clearly enumerate standards and management expectations for !
Engineering Department personnel involving expectations for problem reporting. l 17
DEFICIENCY 96201-10 ISI Summary:
The team identified some WRs that had the incorrect Preventive Maintenance Testing (PMT) assigned or the PMT was not adequately performed. The licensee replaced a Unit 3 ECCS keep fill pump discharge check valve (WR 950060836-01), but the PMT was performed on another check valve in the system. During the performance of WR 960026808, two Unit 2 scram valve position indication limit switches failed the PMT. Although the work package was statused as 54 (failed PMT), the two limit switches were not reworked and the PMTs were not reperfonned contrary to the work instructions (Deficiency 50-237(249)/96201-10).
There were a number of problems in the performance of the corrective action for the repair of a MCC cubicle. The repairs were not performed in accordance with the work instructions, and constitutes an additional example of Deficiency 50-237(249)/96201-10.
Corrective Action:
The following corrective actions were taken in response to this deficiency:
(1) To ensure compliance with NRC requirements and prevent recurrence of this event:
Regarding the ECCS pump discharge check valve, the PMT for system tightness was performed on the correct check valve. A reverse flow seat leakage test was not performed because of system configuration. Therefore, no deficiency existed. '
Regarding the two Unit 2 scram valve position indication limit switches, the two switches I were reworked. Because the plant must be shutdown to perform the PMT, this test will be l' performed during the first unit shutdown.
(2) To prevent recurrence of this or other similar events:
The PMT group was formed in April 1996 to support the Unit 3 refueling outage. DAP 15-10, " Post Maintenance Testing Program," assigns the PMT group the responsibility for reviewing work packages and ensuring PMT completion. This new PMT review responsibility and process should prevent future missed PMTs.
In order to improve human performance, a Policy Deployment Tea'n was assembled and charged with improving human performance by reducing the number of human performance issues associated with procedure adherence. As a result of the Quality Improvement Process and analyses of data by the team, action plans were developed to address specific issues that were identified. In all, there were six action plans consisting of a total of forty-five action steps, that were developed. These action plans include the necessary steps to: Improve Procedure Adequacy, Reduce Procedure Adherence Errors 18
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i Resulting from a Failure to Self-Check, Improve Procedure Adherence by Improving Communications, Improve Procedure Adherence by Improving Supervisor Interaction, improve Procedure Adherence by Improving Schedule Planning, and Improve Procedure Adherence by Improving Training and Qualifications. These action plans have been incorporated into the Dresden Station 1997 Operational Plan. Implementation of the improvement measures will be monitored by the Process Team.
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DEFICIENCY 96201-11 ISI Summary:
The licensee determined that Dresden Station, Unit 1, was not subject to the requirements of the maintenance rule becase it had a possession only license. The guidance currently provided in NUMARC 93-01, dated May 1993, states in part that " Plants that are defueled with a possession-only license will be govemed in accordance with the possession only license." Ilowever, the team determined that the current license for Unit 1 is an amended operating license under Section 104 of the Atomic Energy Act, as described in 10 CFR 50.21(b). Therefore, Dresden Station, Unit I should have been included within the scope of the maintenance rule. The failure to include Unit 1 SSCs within the scope of the maintenance rule is contrary to the requirements of 10 CFR 50.65 (Deficiency 50-010/96201-11).
Corrective Action:
Following the freeze event in January 1994, Dresden developed the Unit 1 Structural Integrity Program and Surveillance Program to monitor and maintain the efTectiveness of the Unit 1 structures, systems, and components associated with the storage, control, and maintenance of the spent fuel in a safe condition.
These programs were not developed or administered under the Unit 2/3 Maintenance Rule Program because the station believed that the Maintenance Rule was not applicable to Unit I when the programs were developed. Prior to August 28, 1996, 10 CFR Section 50.65(a)(1) imposed requirements only on the " holder of a license to operate a nuclear power plant under Sections 50.21(b) or 50.22." Emphasis added. Since Dresden Unit 1 is not licensed to operate but rather only holds a " license to possess" under Section 104 of the Atomic Energy Act, the station believed that Section 50.65 did not apply.
On July 29, 1996, the NRC revised Section 50.65 (efTective August 28, 1996) to apply the Maintenance Rule to Part 50 licensees who's " licenses have been permanently modified to allow possession but not operation of the facility." Since Dresden Unit l's license was permanently modified to allow possession only, i.e., not operation of the facility, the Maintenance Rule became applicable to Dresden Unit 1 effective August 28,1996.'
When this issue arose during the ISI, an evaluation of the Unit I maintenance program spent fuel storage systems was perfomied. This evaluation concluded that the existing Unit 1 Maintenance Program monitored the performance of the SSCs associated with the storage of the spent fuel in a manner sufficient to nrovide reasonable assurance that such SSCs were capable of fulfilling their intended functions, thus the existing program complied with the Maintenance Rule's requirements.
The reviewers also found that the Unit I maintenance program did not include the full administrative controls discussed in NUMARC 93-01 (e.g., no expert panel review was performed).
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i i Therefore, the Unit 1 program was re-constituted using the Unit 2 and 3 maintenance program methodology. While the administration of the program was modined and process enhancements 3 were made to be consistent the Maintenance Rule Program for Units 2 and 3, no changes were j made to the surveillance requirements used to maintain Unit 1 SSCs as a result of this review.
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- This rulemaking did not apply all the Maintenance Rule's requirements to " possession only" licensees. The revision applies to " possession only" licensees only : l
. . . to the extent the licensee shall monitor the performance or condition of all structures, systems, or components associated with the . storage, control, and ,
maintenance of spent fuel in a safe condition, in a manner sufficient to provide reasonable assurance that such structures, systems, and components are capable of fulfilling their intended functions. 61 Fed. Reg. 39301, July 29,1996.
Specifically, the requirement for monitoring against licensee goals was specifically.
deleted.
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DEFICIENCY 96201-13 ISI Summary:
la January 1989, the licensee initiated Procedure DGA-003, Revision 0, " Loss of 250 Vdc Battery Charger Concurrent with a Design Basis Accident." In the 10 CFR 50.59 safety evaluation (DAP 10-2, Revision 1), dated December 23, 1988, the licensee concluded that the Technical 4
Specification (Technical Specification) and the Updated Final Safety Analysis Report (UFSAR) were not affected by the change. Failure to perform an adequate safety evaluation is contrary to the requirements of 10 CFR 50.59 (Deficiency 50-237(249)/96201-13).
The team identified that . . the results of Calculation ATD-2016 clearly demonstrated that even I with the nonconservative assumptions, the 20 psid differential pressure requirement could not be l
maintained for any of the considered CCSW flow / pump combinations. The reduction in margin of ;
the 20 psid pressure requirement appeared to constitute a USQ, as defined in 10 CFR 50.59, that was not submitted to the NRC for review as an application for amendment of the license, contrary to the requirements of 10 CFR 50.59. This constitutes an additional example of Deficiency 50- ;
237(249)/96201-13. !
In response to the team's questions, the licensee stated in a written response that: (1) Valves MO 2(3) 230'-15 and MO 2(3)-2301-49 were normally closed valves that were required to remain closed following the initiation of the HPCI system; and (2) Valve MO 2(3)-2301-48 was normally open and was designed to stay open following the initiation of the HPCI system. However, Valves MO 2(3)-2301-15 and MO 2(3)-2301-49 were actually normally open and automatically reposition closed following the initiation of the HPCI system, and Valve MO 2(3)-2301-48 was normally ;
closed and automatiedly repositions open following the initiation of the HPCI system. The !
original standby alignment of these valves was first changed in 1981 t, compensate for a design I deficiency involving the HPCI Gland Seal Leak-off condenser hotw ell 6til pump. The standby l alignment of the subject valves was changed to provide a flowpath to tia coAensate storage tank. l Although the realignment of these valves was approved in Onsite Review No. 81-2, no 10 CFR 50.59 evaluation was conducted for the change in the standby dignment of these valves. ,
Subsequently, the Unit 2 valves were realigned to their original standby alignment following an event in March 1990 in which feedwater back leakage through Valve MO 2(3)-2301-10 (HPCI test return valve) was identified, as discussed in LER 2-89-029, Revision 4. The valve alignment was subsequently changed again in 1991 following the repair of Valves 2-2301-7, MO 2-2301-8, and MO 2-2301-10. A 10 CFR 50.59 evaluation was performed for changing the Unit 2 valves back to the original standby alignment; however, a 10 CFR 50.59 cvaluation was not performed to change the alignment of these valves back to their current positions. These failures to perform 10 CFR 50.59 evaluations constitute additional examples of Deficiency 50-237J249)/96201-13.
The team identified that UFSAR, Section 6.3.2.3.3.4, and the Design Basis Document, Section 4.1.4.7, indicated that the setpoint for isolation of the steam supply to the HPCI turbine was 100 psig, and that a pending change to the UFSAR, dated August 22,1996, would revise the UFSAR 22 l
value for the setpoint from 100 psig to 80 psig. liowever, the 10 CFR 50.59 evaluation supporting the pending UFSAR change to revise the setpoint from the specified 100 psig to 80 psig was inadequate in that it only considered this change as a clarification and an editorial correction without accounting for the possible impact on IIPCI turbine operability. The failure to perform an adequate 10 CFR 50.59 evaluation represents an additional example of Deficiency 50-237(249)/96201-13.
Although the 10 CFR 50.59 cvaluation correctly assessed the effects of the problems associated with the previous impulse trap and bypass valve operation, it did not address the effect of continued corrosion, which could have been a contributing factor in the piping failures in the reactor and turbine buildings after the implementation of the change. The failure to properly evaluate the llPCI steam trap replacement with an orifice represents a failure to comply with 10 CFR 50.59, and constitutes an additional example of Deficiency 50-237(249)/96201-13.
The licensee continued to rely on the use of temporary system alterations, some of which were unauthorized, to address material and design problems. These examples were not processed as temporary system alterations in accordance with Procedure DAP 05-08, and as a result, safety evaluations were not perfbrmed. The failure to perform safety evaluations for these temporary system alterations constitutes an additional example of Deficiency 50-237(249)/96201-13.
Corrective Action:
The following corrective actions were taken in response to this deficiency:
(1) To ensure compliance with NRC requirements:
Operability Evaluation (Op Eval) 96-31 was prepared to address racking out the auxiliary oil pump breaker and automatic restart of the 11PCI system, and was dispositioned operable - no concems, Comed agrees that the original 50.59 should have addressed issues raised by the ISI and discussed in the Op Eval. Completion of the Op Eval addresses the deficiency.
i Operability Evaluation 96-68 was prepared to address concerns related to CCSW temperatures, i LPCI IIeat Exchanger performance, and the 20 psid difTerential pressure requirement. l Compensatory actions ensured the ability to achieve the required 20 psid differential. The '
operability evaluation was closed by NRC approval of Comed's request for a license amendment.
NRC approva! was received via J. Stang letter to 1. Johnson, dated January 28,1997 "NRC Staff Safety Evaluation Report approving Comed proposed license amendment."
A 50.59 was perfbrmed October 31, 1996 for the changed HPCI System line-up. The 50.59 supported the acceptability of the line-up.
As discussed in the ISI report, an operability evaluation was performed which documents the ability of the HPCI system to perform its safety related function with the isolation setpoint switches at their current value. NTS Item 237-100-96-201-22E due September 21, 1997 was initiated to 23
review and revise calculation NED-I-EIC-00110,0108 and 0111 to address environmental effects.
A 50.59 will be initiated as required with the calculation revision.
The 50.59 for replacement of the HPCI steam trap was revised in December 1996 to address concerns identified during the ISI. The revised safety evaluation concluded that no unreviewed safety question resulted from the change.
During the Independent Safety Inspection , three undocumented temporari alterations were identified by the NRC. Two of them were related to the use of portable faas/ heaters utilizing existing procedures wh;ch were not consistent with the requirements of the terrporary alteration procedure. The third one, which actually occurred in early 1996, was related to closure of a damper without proper documentation and evaluation. A NTS item 237-100-9620ll3A was written to investigate and develop plans to bring the use of portable equipment, as short term substitute for permanent design feature, in compliance with the requirements of Temporary Alteration Procedure and revise affected procedures as appropriate. The third incident that occurred in early 1996 was believed to be due to lack of knowledge about what constitutes a temporary alteration. Formal training was conducted in June of 1996 to increase the level of awareness regarding temporary alterations (NTS commitment 237-200-95-32600-01). Dresden plans to conduct this training on annual basis.
(2) To prevent reoccurrence of similar events.
The ISI inspection report identifies several instances where plant changes were not properly evaluated under the provisions of 10 CFR 50.59. Comed and Dresden initiated several new i programs to address the adequacy of 10 CFR 50.59 Safety Evaluations and screenings.
. Comed Nuclear Division developed a new procedure NSWP A-04, to improve and standardize the safety evaluation and screening process throughout the Division. This procedure also corrects minor deficiencies associated with the previous DAP related to safety evaluation screening and safety evaluation. The procedure will be implemented at Dresden by 3/10/97.
Training to the new procedure has been completed.
- An Engineering Assurance Group (EAG) was established at Dresden to provide oversight and guidance for documents prepared by engineering, including safety evaluations and screenings.
The Engineering Department will trend the quality of 50.59 evaluations reviewed by the EAG to provide a management indicator of overall 50.59 quality.
. The Site Engineering Manager designated senior engineering personnel who are required to provide in-line review of all safety evaluations.
As discussed above, Dresden is aggressively addressing the adequacy of safety evaluations. It is expected that the quality of future safety evaluations and screenings will improve. l l
The off-site review group, which reviews all 50.59's produced by the station, maintains a " grading i system" on the quality of documents reviewed. The attributes graded on 50.59s include consistency, clarity and completeness of responses to the 50.59 questions, consideration of safety significance and justification of conclusions. The trend in 50.59 quality will be monitored by 24
periodically reviewing the data base maintained by the off-site review group. The need for additional training / procedure revisions will be assessed based on EAG and off-site review feedback.
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DEFICIENCY 96201-14 ISI Summary: i The failure to implement corrective actions to prevent the recurrence of significant conditions adverse to quality (battery room temperature problems) is contrary to the requirements of 10 CFR Part 50, Appendix B, Criterion XVI, " Corrective Action" The failure to adequately address the long-standing IST program and implementation problems is contrary to the requirements of 10 CFR Part 50, Appendix B, Criterion XVI, and constitutes an additional example of Deficiency 50-237(249)/96201-14.
The failure to correct long-standing deficiencies affecting the control room ventilation system is I contrary to the requirements of 10 CFR Part 50, Appendix B, Criterion XVI, and represents an '
additional example of Deficiency 50-237(249)/96201-14.
The failure to implement corrective action to account for 250 Vdc battery aging margin constitutes ;
an additional example of Deficiency 50-237(249)/96201-14). 1 i
i The failure to implement corrective actions for the 125 Vdc and 250 Vdc battery charger ;
anchorages is contrary to the requirements of 10 CFR Part 50, Appendix B, Criterion XVI, and ;
constitutes an additional example of Deficiency 50-237(249)/96201-14.
The failure to resolve the SBLC vulnerabilities identified by the VAT in 1992 constitutes an additional example of Deficiency 50-237(249)/96201-14.
Contrary to a commitment made in Licensee Event Report 2-89-029, Sepplement 4, Valves MO 3-2301-8, -9, and -10 were not pressure tested during the previous Unit 3 refueling outage, D3R13.
The licensee stated the commitment was not implemented because it had not been entered into the nuclear tracking system. This constitutes an additional example of Deficiency 50-237(249)/96201-14.
Corrective Action:
The following corrective actions were taken in response to this deficiency:
- In April 1997, Dresden Station will implement a new Corrective Action Program that was developed by representatives from all six (6) Commonwealth Edison nuclear sites and the corporate office. These representatives reviewed state-of-the-art corrective action programs in the industry to establish a new corrective action process for the entire Comed nuclear program. The new process includes several improvements over the current program. It clearly delineates and standardizes the threshold for problem identification through Performance Improvement Form (PIF) initiation, and also incorporates a common PIF database that will provide the site with greater versati!ity in coding, sorting, and analyzing 26
PIF data (PIFs are th'e main problem-reporting mechanism used at Dresden Station). The corrective action process team also developed a new procedure that outlines trending analysis requirements, including a review of industry operating experience for operational events. The new corrective action process incorporates human error reduction methodology developed by Failure Prevcntion International (FPI), including FPI coding and root cause analysis techniques. Included in FPI methodology are problem identification and trend analysis techniques. The new corrective action process will also utilize dedicated root cause analysts and experts, specifically trained and qualified in root cause analysis techniques. Site personnel will receive training on the new corrective action process prior to implementation in April,1997.
Specific training for FPI methodology concerning human error reduction and organizational
& programmatic issues will be conducted as follows:
1 Day Senior Managers and Department Heads Complete 4 Days Root Cause Experts Complete 2 Days First Line Supervisors Schedule under Development 1 Day All Station Personnel Schedule under Development
. DAP 02-38, " Station Self-Assessment," was revised to provide an easier-to-follow format.
Representatives from several of the Sta: ion line organizations participated in the development of the revised procedure, which was in;plemented on January 6,1997. Training on the revised procedure will be conducted for depart ant self-assessment coordinators and department heads prior to their first scheduled self-assessment of 1997.
J e To clarify thresholds at which problems are to be reported, DAP 02-27, "The Integrated Reporting Process," was revised to provide more concise direction for site personnel regarding Performance Improvement Form (PIF) initiation criteria. This revision also incorporated Maintenance Preventable Failures (MPF) as a criterion for PIF initiation. This procedure revision became effective on October 25,1996. Site personnel are being trained to ensure understanding of the revised initiation criteria.
. DAP 02-29, " Corrective Actions Effectiveness Review," was revised to address several issues. This revision eliminated the requirement to initiate a PIF when a corrective action was determined to be ineffective. Instead, the new revision requires that a Nuclear Tracking System (NTS) item (a higher tier tracking item requiring more formal closure) be initiated to track and address the issue. This revision also provides additional guidance on methods for conducting effectiveness reviews and requires that all Level 2 Root Cause Investigation l corrective actions, and all corrective actions associated with Licensee Event Reports (LERs) l be subjected to effectiveness review. Effectiveness reviews conducted for these corrective actions are reviewed and approved by the Corrective Actions Review Board (CARB), thus providing a multi-disciplined review.
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e DAP 02-15," Site Program for Tracking Commitments and Corrective Actions," was revised to provide for inline review and approval of documentation related to closure of commitments and corrective actions. Concurrently, station management implemented actions to provide for increased personnel accountability for actions tracked via the Dresden Administrative Procedure.
- Dresden Engineering management has taken several steps to encourage PIF initiation within the Engineering Department. Engineering senior management met with engineering organization personnel in order to communicate expectations for PIF initiation. In addition, review of the PIF database for 1996 was performed. Engineering procedures were revised to clearly delineate management expectations for PIF generation by Engineering personnel when design discrepancies are identified. During the first eight (8) months of 1996, the engineering organization initiated an average of 49 PIFs per month. During the last four (4) months of 1996, the average increased to 93 PIFs per month, almost double the previous number, indicating that personnel are now more sensitive to PIF initiation requirements.
There were two events in 1996 where the Unit 2 125 Vdc battery room to drop in temperature. Due to a lack of heating being supplied through the East Turbine Building (ETB) ventilation. The first event occurred on January 19,1996 when the operator noticed on his rounds that the battery room temperature had dropped to 64 degrees. The required action at the time this event occurred was to immediately take action to restore the temperature, and notify the shift supervisor. As recorded on the HVO round log sheet dated January 19, 1996, both these immediate actions were implemented. The HVAC system engineer directed the adjustments of the dampers to allow for proper flow of heating through the ETB ventilation. Based on personal interviews, the DC system engineer who had Electrical Maintenance take the pilot cell electrolyte readings because room temperatures were outside of the required limits of 65 to 100 degrees. These readings demonstrated test electrolyte readings remained within operable limits.
In the second event, MCC 25-2 was taken out of service on February 3,1996. As a result, the ETB ventilation dampers failed completely open and introduced ambient outside air directly into the ventilation system feeding the battery rooms. The shift immediately contacted the HVAC and DC system engineers to help resolve the problem. When it was discovered that the vent damperr were open the dampers were wired closed to restore the l battery room temperatures to prevent battery operability. A temporary alteration was written l to follow up the wiring of the dampers. The battery room temperatures dropped to 56 degrees but were restored to above 70 degrees within 1/2 hour after securing the ETB vent dampers. Again, the system engineer had Electrical Maintenance verify that electrolyte temperatures did not drop below operable levels even though the results were not logged at ,
the time. The room temperatures were restored to normal in short order. Based on the )
interviews and the fact that the volume of electrolyte would take significant time to drop to 1 an inoperable level, there was no operability concern. i l
Given that at no time was the operability of the battery in question, there was no requirement to notify the NRC.
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To assure that battery temperatures are documented as operable when room temperatures drop, a step was added to the action statement in the HVO rounds to have electrical maintenance take the pilot cell readings and record them in the rounds.
- Engineering Program performed a Self Assessment of the IST program in 1996. The results of this assessment were captured in LER 2-96-011. Performing a detailed basis review for the program by March 28,1997 was one of the corrective actions in this LER. This review is presently in progress, with all discrepancies being tracked under NTS Item 237-180-9601101-A. We are scheduled to close all IST Basis Review deficiencies under NTS Item 237-1800-9601101-C by August 1,1997, to meet our LER corrective action commitments.
Completion of this review, and resolution of the issues by summer 1997 will resolve the open issues with our progre.m.
The operability of the Control Room HVAC System has been restored. Action plans addressing remaining deficiencies have been developed and are tracked in NTS.
CR HVAC issues arc discussed in LER 2-96-017 and will be discussed at an enforcement conference on February 28,1997.
Regarding battery loading, a calculation to assess battery operability was performed (Calculation DRE 96-0216). This calculation included an increase in the battery aging factor from 1.00 to 1.25 as recommended by standard IEEE 485 "lEEE Recommended Practice for Sizing Large Lead Storage Batteries for Generating Stations and Substations." This calculation determined that the batteries were adequately sized calculation which is used to track and monitor loading on the 250 Vdc battery (PMED-898230-01) was revised and approved December 16,1996
- Installation of safety related seismic anchorages for batteries have been completed as listed below:
Unit 2/3 250 Vdc Battery Charger 2/3-83250 January 17,1997 Unit 2 250 Vdc Battery Charger 2-83250-2 January 26,1997 Unit 3 250 Vdc Battery Charger 3-83250-3 January 8,1997 Unit 3125 Vdc Battery Charger 3-83125-3 January 3,1997 Operability Assessments 96-40 and 96-42 demonstrated that the SBLC system remains operable with no concerns with regard to (1) firing current for squib valves (minimum required current vs. current available) and (2) vulnerability to ground induced failures.
The Vulnerability Assessment Team (VAT) item that stated that "the control power transformer was marginally rated for its duty cycle" is related to the first of these 2 issues.
ER9700521 was initiated to study ways to improve the margin related to the control power transformer and to reduce the circuits vulnerability to ground induce fail.res (NTS item 237-29
i 100-96201148). A separate NTS item is tracking the review of other VAT items for appropriate technical closure (NTS ltem 237-100-9620114A).
e Licensee Event Report 2-89-029, " Elevated HPCI Discharge Piping Temperature Due to Reactor Feedwater System Back Leakage," contains a corrective action to "informationally pressure test 2(3)-2301-9, MOV 2(3)-2301-8 and MOV 2(3)-2301-10 on an every refuel outage frequency." Dresden Station failed to meet this commitment when the test was not performed on Unit 3 during D3R13. The failure to meet this commitment was due to not racking its completion using the station's Nuclear Tracking System (NTS). Current practice is to assign an NTS Item number to all incomplete LER commitments when an LER is issued. The pressure testing of the subject valves was completed for Unit 2 on January 31, 1996 D2R14 and for Unit 3 on November 15,1996 during D3F23.
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4 DEFICIENCY 96201-15 ISI Summary:
- 1 Other problems identified included the safe storage of equipment in the plant. Although Procedure DAP 07-48, " Control of Lay Down, Storage Areas, and Equipment In Use," required that ladders be secured following their use, the team identified four examples in which ladders were not properly secured, including: (1) a ladder on the 544 ft-elevation of the radwaste building; (2) a ladder in the Unit 3 turbine building,517 ft-elevation; (3) a ladder in the reactor feed pump room; and (4) a ladder in the Unit 2 control rod drive west bank accumulators (Deficiency 50-237(249)/96201-15).
Corrective Action:
The following corrective actions were taken in response to this deficiency:
(1) To ensure compliance with NRC requirements:
The four ladders were secured immediately following identification of each deficiency. In l addition, a plant walk through was conducted to search for and secure any other improperly secured equipment. No other unsecured equipment was identified.
(2) To prevent recurrence of this or cimilar events:
Although DAP 07-48, " Control of Lay Down, Storage Areas, and Equipment in Use,"
provided general guidance on control ofladders, a new procedure, DHP 120-07, " Control of Ladders," was issued to establish a formal ladder control program. This program provides better ladder control by imposing specific requirements for ladders.
31
l DEFICIENCY 96201-17 i i
ISI Summary: )
The NRC's 1987 DE of Dresden Station identified Valve MO 2(3)-2301-20 as being within the scope of the IST program but had not been tested. Although the licensee implemented corrective !
actions for other IST valve deficiencies identified by the NRC's 1987 DE, Valve 2(3)-2301-20 was not included within the scope of the IST program for testing. The failure to test Valve 2(3)-2301-20 ,
is contrary to the requirements of the ASME Code, Section XI and Technical Specification 3.0.D l
(Deficiency 50-237(249)/96201-17). l Valve MO 2(3)-2301-15 was included in the IST program beginning in February 1990 even though ,
it appeared that it should have always been tested in accordance with ASME, Section XI and I Technical Specification 3.0D, and constitutes an additional example of Deficiency 50- 1 237(249)/96201-17.
Corrective Action: !
The following corrective actions were taken in response to this deficiency:
(1) To ensure compliance with NRC requirements:
Dresden identified the failure to test valves MO 2(3)-2301-20 during a review in September 1996, and initiated an evaluation to verify that the valves should be included in the IST program. Based on this evaluation, the valves were included in the program as
" disassemble and inspect." The Unit 3 valve was tested in November 1996 during D3F23 )
and found to be in acceptable condition. The Unit 2 valve is being tracked for inspection at the next required outage window in accordance with the IST program.
The HPCI MO-2(3)-2301-15 valves provide a flowpath to the condensate storage tank during periodic system testing. Although this valve is located outside the ISI-class and safety-related boundaries, it is prudent to test it since it receives a safety initiation signal.
These valves were added to Section 12 " Augmented Valves" of the Third Ten-Year Interval Inservice Testing Program, submitted to the NRC on August 31,1995. Since Unit 2 was in a refueling outage (D2R14) when the valves were added to the IST program in August 1995, the HPCI system was not required to be operable. Accordingly, the Unit 2 valve was not tested until February 17,1996, at the end of the outage. The Unit 3 valve was added to the program and quarterly testing of this valve was initiated on August 26,1995. Both Unit's valves have been tested quarterly since their initial testing. No further action is required to restore compliance.
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t (2) To prevent recurrence of this or similar events:
i Based on the IST self assessment performed in 1996 and LER 2-96-011, a detailed IST l program basis review is underway and is scheduled to be completed by March 31,1997 (NTS 237-180-96011B). ;
Discrepancies identified during the review will be tracked by NTS 237-180-9601101 A and l are currently scheduled to be completed by August 1,1997. i
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DEFICIENCY 96201-18 ISI Summary:
Two of the battery tests, which the licensee was using to satisfy Technical Specification requirements, were not performed in accordance with the test procedures. For example, the latest Unit 2,125 Vdc battery service test performed in February 1993 was not tested as specified by the design load profile. The senice test validated the batteries' ability to supply loads under loss-of-power concurrent with loss-of-coolant accident conditions. Procedure DES 8300-28, " Unit 2,125 Volt Station Main Battery Senice Test," required that the battery discharge rate be established in accordance with the load profile, which specified that the battery be discharged at 636.7 amps for one minute and at other specified amperage values for the remainder of the test. However, the test data indicated that the load amperage during the first minute of the test oscillated between 615 and 665 amps. The licensee failed to perform and document an evaluation of this test deficiency before declaring the Unit 2,125 Vdc battery operable. Failure to test the Unit 2,125 Vdc battery at the specified amperage value is contrary to the requirements of Procedure DES 8300-28 (Deficiency 50-237(249)/96201-18).
Corrective Action:
The following corrective actions were taken in response to this deficiency:
(1) To ensure compliance with NRC requirements:
An assessment of battery capacity was performed which confimled battery compliance with IEEE 450 requirements.
(2) To prevent recurrence of this or similar events:
To correct the root cause of the event, plant procedures were revised to specify battery discharge test acceptance criteria reflecting the actual design load profile and to require engineering review of the discharge test results.
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DEFICIENCY 96201-19 ISI Summary:
The licensee concluded that a test performed on a Unit 2 Battery that supplies Unit 3 reactor building loads did not bound all possible loads as part of its senice test. The Unit 2 250 Vdc battery was declared operable after a technical review while Unit 3 remained shut down, with the licensee considering further actions before Unit 3 startup. The failure to demonstrate the performance of an acceptable senice test of the Unit 2, 250 Vdc battery is contrary to the
- requirements of Technical Specification 4.9.A.3 (Deficiency 50-237(249)/96201-19).
Corrective Action:
The following corrective actions were taken in response to this deficiency:
(1) To ensure compliance with NRC requirements:
Unit 2 250 Vdc battery operability was im. mediately restored by taking loads out of senice.
A modification was installed which changed the load profile on the Unit 2 250 Vdc battery to within the parameters tested during the last battery surveillance. This deficiency did not apply to Unit 3 250 Vdc battery due to differences in the load profile used at the time oflast battery test.
(2) To prevent recurrence of this or similar events:
Battery testing procedures have been revised to require use of the most current load profile.
To provide additional assurance, a review is being performed of al! 125Vdc and 250 Vdc battery calculations to determine if any similar situations exist (NTS 237100 96 20100).
35
DEFICIENCY 96201-20 ISI Summary:
1 Surveillance Procedure DTS 5750-06, Revision 3, " Control Room Standby HVAC Air Filtration Unit, and Refrigeration Condensing Unit Performance Requirements," dated August 24,1996 only required 1/8 iwg positive pressure in the control room and did not ensure that pressure was greater than 1/8 iwg for the surrounding areas. In addition, the instrumentation used to verify the control room differential pressure (dp) was not calibrated nor verified to be appropriate for the parameters being measured. Specifically, dp Instruments DPI-2-5740-31/32 and 36 for the control roem and east turbine building had not been calibrated. The licensee also identified that the control room instrumentation was mislabeled with respect to the areas being sensed and, according to the drawings, other sensing lines were misrouted or were broken. The procedures used to test the control room HVAC system and boundaries were not appropriate to circumstances, contrary to 10 CFR Part 50, Appendix B, Criterion V (Deficiency 50-237(249)/96201-20).
Dresden Response:
The corrective actions for the Control Room HVAC issues are discussed in LER 237/96-017. In addition, these issues are the subject ofInspection Report 50-237/96-014, and therefore, they vill be discussed at an Enforcement Conference scheduled for February 28,1997.
The control room HVAC deficiencies stem from modifications that were installed in the 1982-1993 time frame. When the deficiency was identified, aggressive actions were taken to address the operability issues and restore the control room boundary to its original design basis. Actions taken included sealing of the control room boundary and duct work, repair and calibration of instrumentation, and revision of procedures for monitoring control room pressure with respect to adjacent areas.
The Control Room HVAC system was declared operable with no concerns on January 21,1997.
Testing confirmed that the Control Room Emergency Zone was able to maintain positive pressure in the normal mode and maintain a differential pressure of 1/8" w.g. with respect to surrounding areas in the emergency mode. In leakage was also verified to be within UFSAR criteria. The analyzed dose to the operators was in accordance with Dresden's currently licensed methodology.
The auxiliary computer room remains removed from the Control Room Emergency Zone by a Temporary Alteration. A permanent modification will be developed to install permanent ventilation for the auxiliary computer room so that the temporary alteration can be removed.
A PIF was generated after identifying that the East Turbine Building was at negative pressure, contrary to the UFSAR which states that the building is maintained at a positive pressure. The control room pressure instrumentation indicated that the East Turbine Building was positive.
36
An Operability Evaluation (96-04) was completed and the system was declared operable but degraded with compensatory actions in place to monitor potential releases from the east turbine building. The Operability Evaluation is expected to be closed (i.e., the ETB system restored to full operability) by March 31,1997.
The following actions were part of the efforts to restore the ETB HVAC system:
. sealed the building boundary to achieve design positive pressure (complete) e instrument labels installed to reflect correct descriptions (complete)
. trouble shoot and rework instrumentation loops (complete)
. replace fan impeller and vortex damper for ETB Fan C (complete)
. Removal of damaged heating coils and installation of new coils (complete)
. Control valve repair / replacement, piping rework (in work) 37
DEFICIENCY 96201-21 ISI Summary:
The team reviewed the operability assessment for the control room HVAC concerns. On September 26,1996, the licensee identified that the control room was not positive relative to the hallway outside the work execution center, and initiated a PIF. On October 1,1996, the licensee identified that the east turbine building dp relative to the outdoors was not maintained at a positive pressure as identified in the UFSAR. The control room dp instrument indicated that the east turbine building was being maintained at a positive pressure. On October 2,1996, operations initiated an operability determination evaluation. The initial engineering operability judgment was questioned by the Plant operations Review Committee (PORC) on October 7, and a 14-day Dresden Administrative Technical Requirement (DATR) limiting condition for operation was entered. The failure to perform a prompt operability determination within the time specified is contrary to the requirements of Procedure DAP 07-31, Revision 3, " Operability Determinations" (Deficiency 50-237(249)/96201-21).
t Corrective Action:
The following corrective actions were taken in response to this deficiency:
The specific corrective actions for the Control HVAC issues are discussed in LER 237/96-017. In addition, these issues are the subject ofInspection Report 237/96014, and therefore, they will be discussed at an Enforcement Conference scheduled for February 28,1997.
A significant revision to DAP 07-31, " Operability Determinations," had just become effective at the time this deficiency was identified in September 1996.
In October 24,1996, DAP 02-27, "The Integrated Reporting Process (IRP)," and DAP 07- ,
31 were revised to further clarify the various thresholds and decision criteria for PIF initiation, as well as better defining the interfaces between the two procedures.
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DEFICIENCY 96201-22 ISI Summary:
The ISI team identified numerous examples of the failure to translate design information into drawings, specifications, and procedures.
Corrective Action:
Implementation of comprehensive corrective action to improve the quality of engineering activities and verify system operability began promptly during the Independent Safety Inspection as the specific deficiencies were identified. In each case an operability evaluation was performed to assure that the affected systems continued to be capable of performing their intended function.
Moreover, as examples continued to be identified, the Plant Operations Review Committee (PORC) evaluated system operability for the aggregated deficiencies. PORC concluded that the plant systems were operable, and therefore, continued plant operation was acceptable.
As described in our November 8,1996 letter to the NRC, Dresden subsequently performed a series of reviews to reverify that plant conditions were safe and supported continued operation. A review of the ten most safety significant systems was conducted with respect to surveillance acceptance criteria and design functions identified in the Final Safety Analysis Report. In each case, it was determined that the systems were operable and that the equipment would perform its safety function. This conclusion was confirmed by reviewing system performance between 1989 and 1996. In each case, the safety systems operated as required, when called upon.
Corrective actions also focused on ensuring the soundness of current engineering activities. The November 8,1986, letter included a description of the new Engineering Assurance Group's responsibilities and the scope of its activities. This new group was a primary element in our improvement plans. The letter also described the actions to be taken to improve engineering functions, impreve control of vendors and architect engineers, and the verify the adequacy and retrievability of the design basis.
Dresden's planned and implemented corrective actions were summarized in the NRC's November 21, 1996, Confirmatory Action Letter (CAL) to Dresden Station. Subsequently, Dresden has updated the status of corrective actions in monthly " CAL Status Meetings" held alternately at the site or the NRC Region III Office. These status meetings will continue until the CAL's conditions are fulfilled.
39
! DEFICIENCY 96201-23 ISI Summary:
Modifications M12-2(3)-80-03 changed the connection of the instrumentation piping and the ;
setpoint of the dp switch to the current setpoint. However, the UFSAR and the Design Basis
! Document were not revised to reflect these changes. The Design Basis Document and the UFSAR l
were both incorrect, and the licensee issued PIF 96-9559 to correct the noted discrepancies. The failure to update the UFSAR is contrary to the requirements of 10 CFR 50.71(e) (Deficiency 50-237(249)/96201-23).
A September 22, 1994, Comed letter to the NRC, related to a Notice of Violation for NRC l Inspection Report 50-237(249)/94-14, stated that "A change to the Dresden FSAR will be submitted by 12/31/94 to clarify that continued operation of the HPCI system is dependent upon AC electrical components." The team noted that the UFSAR had not been updated as committed in the letter. The licensee identified the UFSAR discrepancy during its August 1996, UFSAR reviews
- and initiated a UFSAR change. The failure to update the UFSAR is contrary to 10 CFR 50.71(e)(4), and constitutes an additional example of Deficiency 50-237(249)/96201-23.
Corrective Action:
l l The following corrective actions were taken in response to this deficiency:
L The required calculations will be verified to support updating the Design Basis Document and changing the UFSAR in the next periodic update (Currently scheduled for June,1997).
l Efforts are underway to prepare UFSAR and Design Basis Document changes to address the new l setpoints associated with MOD M12-2(3)-8-003 for the core spray system. Change packages for the UFSAR and Design Basis Document will be completed by April 30,1997 (NTS Item 2371009620123A). A change package to update the UFSAR to address HPCI operation and AC
! power requirements will be completed on the same schedule (NTS Item 2371009620123B). The 1 existing modification process requires identification of UFSAR and Design Basis Document changes as part of the modification approval process. Update of the UFSAR and Design Basis Documents is tracked by the modification closure process.
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DEFICIENCY 96201-28 ISI Summary:
From 1992 until the end of 1995, a contractor was responsible for implementing the VETIP program at the site. During this period, there was minimal management oversight of the program.
In August 1993, the site quality verification (SQV) department conducted a review of VETIP at Dresden Station to evaluate the status of control in accordance with Procedure DAP 2-10, Revision 4, " Control of Vendor Equipment Technical Information." On August 8,1993, the licensee initiated Corrective Action Record (CAR) 12-93-047, which identified deficiencies with VETIP.
CAR 12-93-047 identified that commitments made in response to NRC GL 83-28, " Required Actions Based on Generic Implications of Salem ATWS Events," and GL 90-03 were not being met. These commitments were stated in the licensee's responses to NRC dated September 26,1990 and July 18,1991, and clarified in the licensee's memorandum dated December 5,1991 (Deficiency 50-237(249)/96201-28).
Corrective Action:
The following corrective actions were taken in response to this deficiency:
This deficiency was self-identified and a long tenn plan to bring Vendor Equipment Technical Information Program (VETIP) into full compliance with NRC requirements was implemented in July 1996. This plan has been followed, with one exception, and is on schedule to place all manuals under the control of Central Files as controlled documents by May 30,1997. The one exception to the plan is that GE SAL receipt and distribution will continue to be the responsibility of the Operating Experience Coordinator (OPEX) in order to conform with the Comed standard procedure for nuclear stations.
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I DEFICIENCY 96201-29 ISI Summary:
As recently as December 1995, an NRC violation was issued for failing to preapprove exemptions from the overtime guidelines and in July 1996, the licensec issued a corrective action record (CAR) for additional violations of overtime requirements. During the root cause evaluation for the CAR, the licensee identified additional violations involving the failure of the electrical maintenance department to track overtime. Consequently, two electrical maintenance workers exceeded the NRC Generic Letter (GL) 82-12, " Nuclear Power Plant Staff Working Hours " limit of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in 7 days during the period of July 17 through 24,1996, and one instrument maintenance supervisor exceeded the GL 82-12 limit of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> on September 18,1996, without preapproval.
Exceeding the limits of GL 82-12 without obtaining preapproval is contrary to the requirements of Technical Specification 6.1.B (Deficiency 50-237(249)/96201-29).
l Corrective Action:
The following corrective actions were taken in response to this deficiency:
(1) To ensure compliance with NRC requirements:
l As provided in DAP 01-09, " Control of Overtime," Overtime Deviation Authorization Fomis were submitted and Post-Authorized for these events.
A subsequent investigation revealed that the Instrument Maintenance Supervisor involved in exceeding the GL 82-12 limits had requested pre-authorization via telephone before the violation occurred. The DAP 01-09 Form was signed the following day by Station Management. Therefore, in his case, a GL 82-12 (Technical Specification 6.1.B) violation did not occur.
(2) To prevent recurrence of this or similar events:
The individuals and supervisor involved in the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in 7 day event were counseled.
Electrical Maintenance Department received " tailgate" training on the event and the need to control overtime within the GL 82-12 requirements.
DAP 01-09 was revised in October,1996 to allow the Department Manager to approve Overtime Deviations that do not violate the GL 82-12 limits.
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UNRESOLVED ITEM 96201-12 i
ISI Sunamary:
The team reviewed program documents and records to evaluate the process that was established to
! set goals and monitor under (a)(1) and (a)(2) of the maintenance rule for the HPCI, circulating water, offgas,125 Vde, and CS systems. In general, the goals or performance criteria were in accordance with good safety practices, and industry-wide operating experience was taken into consideration when setting goals and performance criteria. However, some of the reliability 1 performance criteria and goals were not consistent with reliability values that were assumed in the j PRA. For example, the reliability performance criterion for risk-significant systems was no maintenance preventable functional failures (MPFFs). In essence, this is a reliability of 100 i percent, which may not be consistent with the assumptions in the PRA. Similarly, the reliability goal for the circulating water system was no MPFFs. The use of MPFFs as a measure of reliability
- and its compatibility with the reliability assumptions used in the PRA was the subject of ongoing i discussions between the NRC staff and the Nuclear Energy Institute (NEI) at the time of this inspection. Accordingly, this issue will remain unresolved pending further NRC review (Unresolved Item 50-237(249)/96201-12).
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- Dresden Response
l The Dresden Station Maintenance Rule Program, consistent with industry practice, uses Maintenance Preventable Function Failures per unit of time as an indicator of reliability. However, the ISI inspector questioned whether both failures and successes per time should be measured for a l
true indicator of reliability. The following action was taken as a result:
1 l
An ad hoc committee was assembled to verify that the numerical target values selected for I Maintenance Rule performance criteria are commensurate with safety. This committee determined that the performance criteria selected in the Dresden Maintenance Rule Program are commensurate with safety because adequate conservatism is built into the availability indicators and because all risk significant Maintenance Rule SSC functions modeled in the PRA have a "O MRFFs/ time" reliability performance criteria. Hence, this component of the issue was satisfactorily completed.
The Maintenance Rule Comed Peer Group discussed the reliability indicator issue with other nuclear stations and with the Noclear Energy Institute (NEI). Comed understans that NEI issued a 4 letter to the NRC requesting clarification of the NRC's requirements in this area.
The Comed Maintenance Rule Peer Group attended the October 15,1996, NRC Public Meeting conceming this issue. At the meeting, the NRC stated that a letter clarifying the requirements would be forth coming.
As acknowledged in the ISI Report, no further action is planned until the NRC clarifies its requirements. l l
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{ UNRESOLVED ITEM %201-16 ;
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[ The team detemiined that the licensee's Design Basis Document, Sargent & Lundy Specification K- l l 4080, and Procedure DMS 0040-01 did not contain accurate or consistent design information for j design pressures and relief valve setpoints. The adequacy of ECCS relief valve setpoints and Valve RV 2-2301-23 will remain unresolved pending funher review by NRC (Unresolved Item 50- ,
i 237(249)/96201-16).
! September 18,1996, which identified discrepancies between the LPCI pump suction relief valve !
j setpoint and the design pressure for the system. The Design Basis Documents for the LPCI, CS !
and HPCI systems identified open items since the Design Basis Documents were written regarding d
the setpoint of the relief valves in those systems. The license initiated a PIF for a CS system relief i i valve on July 17,1996, but did not address more than the CS relief valve (Level 4 PIF). The root i l cause and investigation report for this PIF (237-200-96-18200) stated that in June 1996, during a j review of the D2R14 refueling outage activities (February 1996), a concern was raised as to the carrect setpoint for the LPCI pump suction relief valve. The relief valve setting exceeded the l design pressure given in Specification K-4080. A description of relief Valve RV 2-1599-13D, l LPCI pump suction relief valve, attached to the PlF fonn stated that "Since the torus and LPCI .
suction piping will not experience these pressures, the urgency in writing this PIF vice investigating the problem was deemed to be low." The NRC will further evaluate the condition of the LPCI suctica r lief valve, which will be tracked by Unresolved Item 50-237(249)/96201-16.
Dresden Response:
This unresolved item involves a discrepancy between the relief valve setpoints and the controlling )
documentation for the HPCI, LPCI and Core Spray Systems identified by the Dresden ILRT i Coordinator during his review of the D2R14 ILRT results. ]
i Based on the limited information found as part of our investigation, it appears that the relief valve !
setpoints have been unchanged for a significant period of time. For example, the original j procurement specification required the relief valves to be set at their current setpoints.
l Subsequently, the piping calculations performed to evaluate these systems for Generic Letter 79-14, i "Large Bore Piping," and the Mark I analysis used the setpoint data currently in the K-4080 design basis document, i.e., the current setpoints. Similarly, the original issue of DMS 0040-01, " Relief .
Valve Outage Surveillance," received an oversight review on December 14,1989, and the setpoints j have remained unchanged since that time. ,
l To ensure compliance with NRC requirements, operability evaluations were performed which i ve ilied that all sub-systems were within & sign basis limits given the relief valve setpoints. As l pm of the evaluation M relief valve setpoints for relief valves in the IST program and the relief )
Mves for Section III sessels were reviewed and additional valves were included in the evaluation 44
ofitems to be evaluated (IIPCI pump suction, Core Spray discharge, I.PC) lleat exchanger Relief Valves and the LPCI pump suction relief valves).
To prevent recurrence of this or similar events:
Dresden Station will perform a design reconstitution for the 12 most risk significant systems during the next two years. This design reconstitution will include a review of relief valve setpoints.
The Design Basis Documents for each system (Core Spray, IIPCI and LPCI) will be changed to include the revised information (NTS 237-200-96-07 due June 18,1997).
K-4080 is being updated to reflect the higher design pressures of the lines these relief valves protect. (NTS 237-200-96-182-06 due June 18,1997).
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UNRESOLVED ITEM %201-24 ISI Summary:
The licensee implemented a modification to replace :he circuit breaker associated with the Unit 2 HPCI Gland Seal Leak-off condenser hotwell drain pump motor. The breaker was a 35 A Westinghouse Type FA, which had to be replaced because of a broken handle. Since the FA breaker was no longer available, the replacement breaker selected was a 30 A Westinghouse Type HFD breaker.
Calculation DR-E-96-0040, Revision 0, March 12,1996, which was performed to determine the acceptability of the breaker, included a coordination diagram. The team identified several inconsistencies with the coordination diagram, including the determination that the motor thermal damage poir i not been taken into account.
The team reviewed the licensee's response and identified that: (1) the licensee failed to demonstrate that the replacement breaker was adequate for providing proper motor protection because the breaker trip band of 10-19 seconds is an uncertainty band, and tripping at 10 seconds may not occur (typically, an appropriate breaker would provide margin over the maximum expected trip time to ensure that tripping would occur before motor thermal damage); and (2) the licensee's response was incomplete because the licensee did not evaluate tb motor accelerating times at the expected limits of source voltages. The adequacy of the replacement breaker to perform its safety function is unresolved pending further licensee and NRC review (Unresolved Item 50-237(249)/96201-24).
Dresden Response:
The replacement breaker is adequately sized for the intended application. The motor starting / acceleration curve shown on the coordination plot should not be mistaken with a thermal damage curve. Neither the actual maximum design acceleration time, nor the thermal damage curve for the subject motor was available at the time of the calculation. The general philosophy used in selection of protective devices on senices important to safety is to sacrifice protection of the senice in avoidance of spurious operation. Therefore, the protective device was selected using a conservatively assmned 10 second motor acceleration time to ensure that motor acceleration could occur prior to protective device operation. Although this may suuject the motor to potential damage, continuity to the senice which is important to safety is assured. In addition, the calculation shows that the motor feed cable is adequately protected with the selected breaker.
The existing Comed standard takes into account potential changes in system voltage. For example, the 140 percent continuous current rating is intended to allow for a variance in system voltage leading to increases in motor full load current. As the calculation uses a very conservative acceleration time, (a typical acceleration time for a small pump is 3-5 seconds) there is sufficient margin in the calculation to account for voltage variation.
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A prior commitment has been made to revise this calculation. During this revision, the above mentioned factors will be clarified. (NTS Item 2371009620124A, July 15,1997) 1 l
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UNRESOLVED ITEM 96201-25 ISI Summary:
The licensee stated in a letter to the NRC, dated November 6,1996, that the Dresden Station 10 i CFR 50.46 analysis did not include allowances for the instrument uncertainty, but they planned to address the efTects ofinstrument uncertainty by March 1997.
The results of these preliminary calculations indicated that with the existing low margins in pump performance, instrument errors, data nonconservatisms, and pump degradation, system performance could be impacted:
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For Unit 3, the licensee developed an analysis that depicted liPCI pump flow as a function of '
reactor pressure. This analysis predicted that the HPCI pump can deliver a flow of 5000 gpm against a reactor pressure of approxaately 1180 psig. The 10 CFR 50.46 analysis assumed that for l
this flow the reactor pressure will be 1150 psig. Therefore, the available margin is only 30 psid (2.5 l percent). l For the CS system, the predicted combined flow rate was 11,380 gpm for the Unit 3 and 11,400 gpm for Unit 2. The limiting PCT analysis assumed, that at 0 psig reactor pressure, the combined j CS flow will be 11,300 gpm. Therefore, the available margin is only 80 gpm, or 0.7 percent for i Unit 3 and 100 gpm or 0.9 percent for Unit 2.
These performance analyses were developed almost exclusively on the basis of test data from the preoperational tests, IST surveillance tests, and MOV tests, which in some cases, involved multiple instruments with various accuracies. As discussed in Section 4.6.2.2, the licensee identified that l they had failed to meet ASME Code requirements for calibration error and accuracy requirements for indication instrumentation.
The licensee acknowledged the need to accomit for instrument error, but as of the end of the inspection, had not done so either in testing or the latest 10 CFR Part 50.46 analysis. The need to consider instrument inaccuracy will remain unresolved pending further NRC review (Unresolved item 50-237(249)/96201-25).
Dresden Response:
Dresden Station will incorporate an allowance for instrument uncertainty into the LOCA analysis to address the effect ofinstrument uncertainty on ECCS pump performance testing. As stated in the inspection report, Design Engineering is in the process of quantifying the accuracy of the instrumentation used for pump performance test methods and will issue a letter Much 31,1997, describing the planned treatment ofinstrument uncertainties. (This work is being tracked by NTS item 237-123-96-02006). Based on the instrumentation accuracy, an equivalent leakage term will e be established as an input to the D3R14 LOCA analysis.
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I UNRESOLVED ITEM 96201-26 ISI Summary:
i Technical Specification Interpretation No. 22, dated June 2,1995, addressed the Tedmical l Specification requirement pertaining to the Containment Atmospheric Dilution and Purge System, Technical Specification 3.7.A.6. The Technical Specification requires the system to supply nitrogen to containment for atmospheric dilution if needed by post-LOCA conditions. The bases for Technical Specification 3.7.A.6 explains that this capability is required to maintain the oxygen- l hydrogen mixture below the flammable limit. )
l Interpretation No. 22 stated that a General Electric (GE) evaluation concluded that 29 scfm of nitrogen flow is needed to provide the required containment dilution; however, the system can only j supply nitrogen to the containment at 20 scfm. The interpretation also stated that the GE ,
performance requirement of 29 scfm should not apply to Dresden Station because the system l design and installation predated the GE evaluation. Technical Specification 3.7.A.6 has been in !
place since 1974, and the GE evaluation was transmitted to the licensee in 1991.
The Unit 2 nitrogen inerting system was upgraded during the 1996 refueling outage and is capable of supplying nitrogen at a rate that exceeds 29 scfm. The Unit 3 system has not been modified. At the time this problem was identified, Unit 3 was shutdown. The ability of the Unit 3 nitrogen !
incrting system to satisfy the requirements of Technical Specification 3.7.A.6 will remain unresolved pending further licensee and NRC review (Unresolved Item 50-237(249)/96201-26).
I Dresden Response:
The Technical Specification Interpretation discussed in the inspection report was reviewed and determined to be unnecessary. As described in JSPLTR #96-0244, a subsequent Operability Evaluation was performed which verified that sufficient flow (more than twice the required 32 sefm) was available such that neither the Unit 2 nor Unit 3 Nitrogen Make-up Systems were ever in noncompliance with the 1972 32 scfm requirement.
Prior to the ISI, the Station had conducted a review of the existing Technical Specification Interpretations and determined that several were no longer needed, including Technical Specification Interpretation number 22. Therefore, these Technical Specification Interpretations were being deleted. In addition, the remaining nine Technical Specification Interpretations were determined to be technically correct and would remain in force until the Technical Specification Upgrade Program was implemented.
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UNRESOLVED ITEM 96201-27 ISI Summary:
The purpose of DCP 9300122 was to install an MOV actuator that was capable of delivering greater closing thrust than the actuator currently installed on Valves MO 2(3)-2301-15. These valves, as well as Valves MO 2(3)-2301-10 provide isolation between the liPCI system and outside of containment. In 1993 the licensee identified that the actuator for Valve MO 2-2301-15 was not capable of closing against the shutoff head of the liPCI pump. Calculations using the appropriate valve factor indicated that the valve would not open and close against worst-case differential pressure and degraded voltage conditions. The licensee assumed that these results were also applicable to Valve MO 3-2301-15. As compensatory actions, the licensee revised procedures to declare the llPCI system inoperable in the test mode and to trip the IIPCI turbine if either Valves MO 2(3)-2301-10 or Valves MO 2(3)-2301-15 fail to close, when the IIPCI system is operating in the pressure control mode and the IIPCI pump suction supply auto-transfers.
The licensee's initial long-term solution was to install a larger MOV actuator; however, when the licensee walked down the HPCI system, it determined that the larger actuator would not fit in the existing configuration. The licensee determined that other options included rotating or moving the valve, extending the valve yoke, or changing the design basis. In July 1994 DCP 9300122 was canceled and the issue was tracked by Plant Change Request 0326, which was intended to disposition all MOV upgrades under the Generic Letter (GL) 89-10 program. Rather than replace the existing actuator with a larger actuator, the licensee modified the actuator gearing and bypassed the torque switch in the closed direction. These modifications were accomplished under Design Changes P12-2-94-272 and P-12-3-94-288.
Licensee personnel indicated that even after the implementation of these modifications, the actuator for Valve MO 2-2301-15 was marginally sized and the actuator for Valve MO 3-2301-15 was undersized. The licensee also stated that there were no plans to change the actuators because the valves were nonsafety-related and therefore were removed from the scope of the GL 89-10 program. Ilowever, the team determined that the liPCI design basis document still indicated that the HPCI MOVs, including Valves MO 2(3)-2301-15, must close against ilPCI pump discharge pressure.
The licensee's written response to the team's questions regarding DCP 9300122 indicated that this l DCP was not necessaiy because the existing actuator configuration provided enough capability to perform its safety function with acceptable design margin. The same written response also indicated that DCP 9300122 was canceled because it was a duplicate of Design Change E 12-2 272. This information was inconsistent with the information previously discussed, and the licensee subsequently provided a revised response. This issue will remain unresolved pending NRC review of tne acceptability of removing Valves MO 2(3)-2301-15 from the scope of the GL 89-10 program (Unresolveu Item 50-237(249)/96201-27).
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Dresden Response:
As discussed with the ISI Inspectors, the 2(3) 2301-15 valves were removed from the Generic Letter 89-10 program on September 26,1996 on the following basis:
The HPCI system test retum to the Condensate Storage Tank (CST) redundant shutof' MOV does not perform any safety related function. The removal of the valve from the program was reviewed in accordance with MOV White Paper 155. These MOVs have been diagnostically tested to ensure reliable operation.
This issue was subsequently resolved in NRC Inspection Report 50-237/96015(DRS); 50-249/96015(DRS), dated February 14,1997, which stated:
The NRC Independent Safety Inspection (ISI) team identified a concern with the removal of the HPCI return to condensate storage tank isolation valve,2(3)-2301-15. The valve was normally open and would close upon initiation of the HPCI system to isolate the condensate storage tank (CST) from the HPCI test line and the HPCI pump cooling water line. The valve provided a redundant function, since 2(3)-2301-10 was the test line isolation valve and 2(3)-2301-49 was the cooling water isolation valve. Both of these valves were included in the GL 89-10 and insenice test (IST) programs with a closed safety function. The '
inspectors, in conjunction with NRR reactor systems staff, concluded that the valve could be removed from the GL 89-10 program based on the valve not having a safety function.
The 2(3)-2301-15 valve, however, res ;ained in the augmented IST program to ensure the valve would operate, although not under design-basis conditions.
Since the removal of the valves from the program was approved by the NRC, this unresolved item should be closed consistent with the guidance contained in the inspection report.
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