ML17188A074
| ML17188A074 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 03/06/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17188A072 | List: |
| References | |
| 50-237-97-21, 50-249-97-21, NUDOCS 9803130194 | |
| Download: ML17188A074 (23) | |
See also: IR 05000237/1997021
Text
U.S. NUCLEAR REGULATORY COMMISSION
Docket Nos:
License Nos:
Report No:
Report No:
Licensee:
Facility:
Location:
Dates:
Inspectors:
Approved by:
9803130194 980306
ADOCK 05000237
G
REGION Ill
50-237; 50-249
50-2.37/97021 (DRS)
50-249/97021 (DRS)
Commonwealth Edison .Company
Dresden Generating Station, Units 2 and 3
6500 North Dresden Road
Morris, IL 60450-9765
October 21, 1997 through January 27, 1998
George M. Hausman, Reactor Inspector
Gerry F. O'Dwyer, Reactor Inspector
Darrell L. Schrum, Reactor Inspector
Tom Tella, Reactor Inspector
Ronald N. Gardner, Chief
Engineering Specialists Branch 2
Division of Reactor Safety
EXECUTIVE SUMMARY
Dresden Generating Station, Units 2 and 3
NRC Inspection Report 50-237/97021 (DRS); 50-249/97021 (DRS)
An announced core inspection that reviewed the engineering and technical support (E& TS)
organization's effectiveness in the performance of routine and reactive site activities including
identification and resolution of technical issues and problems. As a result of the inspection,
three violations (VIOs) of Nuclear Regulatory Commission (NRC) requirements were identified
and one unresolved item (URI) was issued.
Overall the inspection concluded that the engineering staff was effective in the
identification and resolution of technical issues. Self-assessments exhibited a pro-active
trend in the attempt to disclose performance problems within the engineering
organization. The quality of engineering activities was in most cases technically sound.
(Section All)
The team had concerns that the UFSAR did not accurately characterize the plant's
design-basis or the plant's capability to respond to a potential Dresden Lock and Dam
failure. As a result, the team concluded that further review by the licensee and NRC
was required. An NRC URI was initiated to document.these concerns. (Section E3.4;
URI 50-237/249-97021-01 (DRS))
The team concluded that all commitments and corrective actions identified by
Confirmatory Action Letter (CAL) No. Rlll-96-016, dated November 21, 1996, including
those activities associated with the Dresden Engineering Assurance Group (DEAG)
have satisfied NRC requirements. The CAL was closed. (Section E6).
In November 1994, the licensee identified that a prior inadvertent cha*nge to the Dresden
Station's control ro'om ventilation system design deleted the automatic smoke purge
mode transfer capability. From November 1994 to March 1996, the licensee failed to
perform a written safety evaluation to provide the bases for the determination that the
change did not involve an unreviewed safety question. (Section F2;
VIO 50-237 /249-97021-02(DRS))
.
From November 1994 through November 21, 1997, the Fire Protection Report,
referenced as part of the UFSAR, had not been updated and the required revision
updates submitted to the NRC. (Section F3; VIO 50-237/249-97021-03(DRS))
- As of November 21, 1997, the fire pre-plans had not been updated since
September 1992. (Section F3; VIO 50-237/249-97021-04(DRS))
2
Report Details
Ill. Engineering
E1
Conduct of Engineering
E 1.1
Performance and Effectiveness
a.
Inspection Scope (IP37550: IP40500)
The purpose of the inspection was to evaluate the effectiveness of the E& TS
organization in the performance of routine and reactive site activities including
identification and resolution of technical issues and problems. The inspection focused
on system engineering functions, modifications, technical problem resolution, and
engineering support to other plant organizations. In addition, the licensee's corrective
action proces~ was evaluated.
Ttie criteria used to assess the E& TS performance was quality of technical work
produced, understanding of plant design, and active involvement in preventing and
solving plant problems.
b.
Observations and Findings
Overall, the engineering staff was effective in the identification and resolution of
technical issues. The inspection showed engineers to be knowledgeable and involved
with the work conducted in their respective areas of responsibility. Engineers and
immediate supervisors were cognizant of the current status of assigned systems and
components, as well as, recent problems and deficiencies that had been identified. The
quality of the reviews conducted by the engineering staff was in most cases technically
sound. However, minor discrepancies were observed in many of the engineering
products and activities. These discrepancies indicated that the licensee's engineering
staff should be thorough and exhibit more attention to detail. The DEAG reviews were
in most cases thorough and technically sound. However, the team was concerned that
many DEAG members were no longer employed at the Dresden site and such loss of
experienced personnel might degrade the licensee's ability to maintain an improving
trend in engineering performance.
c.
Conclusions
The inspection team concluded that conduct of engineering was satisfactory.
E1 .2
Problem Identification and Root Cause Determination
a.
Inspection Scope (IP37550: IP40500)
The team reviewed several PIFs generated by the plant staff and verified whether the
PIFs were properly processed for root cause determination and corrective actions.
3
.,
b.
Observations and Findings
The team reviewed selected PIFs for adequate d.ascription of the problem and to verify
whether the PIFs were properly prioritized and followed up as necessary. The team
also reviewed whether the PIFs were reviewed for root cause determination and
corrective actions when required. A nuclear tracking system (NTS) number was
assigned to follow up PIFs. The team reviewed a few NTS items to verify whether they
were adequately followed up by the licensee for completion. The team noticed that the
reasons for NTS due .date extensions were not always adequately justified. An example
was PIF 97-12037 dated January 30, 1997, regarding allowable battery temperatures.
This PIF was tracked by NTS Item 237-201-97-12001. The reason for extending this
NTS item for about five months was "to provide new DC system engineer time to
evaluate other options."
The tepm attended a PIF screening meeting on November 3, 1997. The team noted
that the department managers/supervisors were present as necessary. The PIFs
received were adequately discussed and assigned to the responsible departments for
further follow up.
c.
Conclusion
The team concluded that a low threshold exists for generation of PIFs. The team
observed that the PIFs were promptly processed and assigned to a department for
follow up. The root causes for important PIFs were identified for further corrective
actions. However, adequate justification was not always provided for extending
corrective action due dates.
E2
Engineering Support of Facilities and Equipment
E2.1
4kV Breaker Auxiliary Switch Failures
a.
Inspection Scope (IP 37550: IP40500)
The team reviewed the licensee's corrective actions for Merlin-Gerin 4.1 kV.breaker
auxiliary switch failures.
b.
Observations and Findings
The licensee's cqrrective actions involved the installation of nylon tie-wrap~ around the
breaker's auxiliary switches. The auxiliary switches on the breakers were made of a
phenolic material and were observed to develop cracks at the Dresden and Quad Cities
Stations.
The manufacturer and the local distributor of the breakers, Pacific Breaker Systems, Inc.
and Golden Gate Switchboard Co., were informed of the defects. A 1 O CFR Part 21
notification was issued by Golden Gate Switchboard Co. on April 11, 1997, regarding
the cracking and breakage of the circuit breaker auxiliary switches in the mounting area.
4
The cracking and breakage in the mounting area resulted in unacceptable contact
resistance readings.
The licensee developed a temporary fix that used nylon tie-wraps around the two
auxiliary switches on each breaker and qualified the fix for a period of 18 months. The
qualification test was performed by testing the breaker for 225 cycles and performing a
seismic test at Wyle Laboratories. The Plant Operations Review Committee approved
the modification for only one plant operating cycle.
The team noted that the root cause(s) for the failure of the auxiliary switches had not
been identified by the manufacturer. Potential .corrective actions, such as a change in
the* type of switch material, had not been provided to the licensee.
However, the licensee performed a root cause evaluation during August 1997, which
concluded that the primary root cause(s) for the faiiures were a design weakness in the
auxiliary switch mounting and inappropriate torque values forthe mounting T-bolts. The
evaluation led to the licensee's immediate corrective action of using nylon tie-wraps
around the auxiliary switches.
For a semi-permanent fix, the licensee intended to qualify the nylon tie-wraps for a
period of six years. The breakers were tested with the nylon tie-wraps for 750 cycles at
Commonwealth Edison Company's (ComEd's) C-Team facility and seismically qualified
at the Wyle Lab for six years. The licensee intended to use stainless steel U-bolts (in
place of the tie-wraps) as a permanent fix,
The licensee's root cause evaluation indicated that the original design created tensile
forces where the phenolic material was not sufficiently strong. The team noted that the
tie~-wraps had reduced the tensile forces to some extent; however, the licensee's root
cause effort did not address the weakness of the switch material and the. potential need
to change to an alternate (stronger) material that could withstand the higher tensile
forces.
The team expressed concern that the tie-wrapped auxiliary switches were considered
for extended use, prior to the completion of the manufacturer's root cause evaluation
and without considering an alternate material. The team considered the potential for
cracking the auxiliary switches during operation remained even with the tie-wraps or
U-bolts in place.
c.
Conclusion
The licensee's actions to temporarily extend the life of the auxiliary switches with nylon
tie-wraps were acceptable. However, the licensee's and vendor's failure to address the
weakness of the phenolic material and not considering an alternate (stronger) material
for the auxiliary switches was considered a weakness.
5
E2.2
Plant Walkdowns
a.
Inspection Scope (IP 37550)
The team walked down several areas of the plant to assess the material condition of
equipment and general plant condition.
b.
Observations and Findings
The team walked down the intake strudure and some electrical areas, such as the
diesel generators, switchgear areas, arid battery rooms.
The areas walked down were generally kept clean. The equipment observed, such as
. safety-related batteries, diesel generators and safety-related electrical switchgear were
maintained in good condition.
c.
Conclusions
The team concluded that the plant areas walked down were well maintained and no
deficiencies were observed.
E3
Engineering Procedures and Documentation
E3.1
Design Change Packages. Modifications and Temporary Alterations
a.
Inspection Scope (IP 37550)
The team reviewed the following design change packages (DCPs), modifications and
temporary alteration (Temp Alt):
DCP 9700202
Install 70 Amp Breaker in Cubicle 39-2-C3
DCP 9700207
Change out of Control Transformers in Turbine Oil Tank
Vapor Extractor Breaker
E12-3-95-224
Limit Switch Replacement on Motor OperatedValve
(MOV) 3-205-24
M12-0-97-001A
Auxiliary Electrical Equipment Room Heating, Ventilation,
and Air Conditioning (HVAC) Modification
M 12-2-85-302
- Unit 2 - 125 Volt DC Charger Upgrades
- .
M12-3-96-008
Time Delay Addition on Valve 3-2301-15
P12-3-94-284
Gearset Replacement on MOV 3-1501-28B
Temp Alt 111-09-97
Install Portable Air Compressor Outside 'Crib House
6
b.
Observations and Findings
The team ob.served that the above DCPs and modifications clearly described the
proposed alterations and justifications. Each design change contained an adequate
10 CFR 50. 59 screening or safety evaluation. The design. issues worksheets
considered several additional*issues. Adequate interdepartmental reviews were
performed as necessary.
The team reviewed several calculations made in support of the design changes. The
calculations included acceptable assumptions and were adequately reviewed and
approved. No problems were identified with the calculations.
Several work requests were reviewed that implemented the design changes. The team
found that the design changes did not always include the results of post-modification
testing (PMT). An example was the PMT performed for DCP E12-3-95-224 (level switch
replacement on MOV 3-205-24) that was completed on June 11, 1997. The team had to
obtain a copy of the completed procedure from .central files to verify whether the PMT
was completed.
The team observed that Temp Alt 111-09-97 provided the reasons for the alteration, an
adequate safety evaluation and a date for the expected removal of the alteration (five
months after installation). The team's walk down of the temporary alteration found the
- Temp ALT installation in good condition.
c.
Conclusion
The team concluded that the modifications, DCPs and temporary alteration reviewed
were adequately implemented. However, some DCPs did not include PMT results.
E3.3
Calculations/Evaluatio*n
a.
Inspection Scope (IP 37550}
The team reviewed the following calculations/evaluation and associated DEAG reviews:
Calculation DRE 97-0171, "Determination of Acceptance Criteria for CCSW One
and Two Pump NPSH Testing - Units 2 & 3," Revision O
Calculation DRE 97-0172, "Vortexing at CCSW Intakes - Units 2 & 3," Revision O
Document ID# 5543459, "Evaluation, Re: Low Pressure Coolant Injection (LPCI)
System, Hydraulic Calculation for Containment Cooling and Containment Cooling
Spray Modes," dated October 29, 1997
7
b.
Observations and Findings
Calculation DRE 97-0171:
The team observed that the calculation used the pump suction centerline as the pump
datum plane.
The team determined that this method of calculation was non-conservative and
introduced an error into the calculation. The team's assessment ofthe DEAG review
identified that the DEAG did not detect this error, but did note conservatism in the
calculation. The team determined that the conservatism compensated for the
non-conservative error.
The team observed that not all of the logic thought processes and equation derivations
were documented in the calculation, making the methodology more difficult to
understand (e.g., the gage error effect was not bounded). These weaknesses indicated
a need for more attention to detail. The DEAG review recommended similar
clarifications to make the calculation a better source of information for future users.
Calculation DRE 97-0172:
The Vortexing calculation stated the maximum CCSW intake flow rate was 7,200 gpm.
The calculation's design input reference was the Hydraulic Institute Standards, ANSI/HI
1.3.3.6.1-1.3.3.6.3, American National Standard for Centrifugal Pumps, approved
May 23, 1994.
As flow rates increase the distance between intake centerlines must be increased to
prevent vortexing. The calculation identified the actual distance between the CCSW
intake centerlines as 42 inches. The design input reference stated the minimum
distance between the i_ntake centerlines should be 52 inches for a 7,200 gpm flow rate
and that at 42 inches the flow rate should be limited to 5,400 gpm.
Although the actual distance did not meet the design input reference's recommendation
for a 7200 gpm flow rate, the calculation concluded that the distance was acceptable *
because the CCSW system was required to be maintained at 20 psid higher than the
LPCI system. The 20 psid differential was maintained by throttling the CCSW flow rate
below 7,200 gpm.
The team was concerned that the amount of throttling was not specified and given the
right operating configuration, vortexing might occur due to insufficient distance between
intake centerlines. In response, the licensee obtained and documented in Nuclear
Design Information Transmittal (NDIT) S040-DH-0513 the vendor's confirmation that a
42 inch distance was acceptable for flows as high* as 7,200 gpm. The team determined
that the specific vendor statement took precedence over the general recommendation in
the design input reference. Therefore, the calculation's conclusion that CCSW pump
intake bay dimensions were adequate was correct. The DEAG* reviewer stated the
8
reason he did not comment on the absence of a specified maximum flow rate was that it
was common knowledge within Dresden Engineering that the 20 psid restriction required
throttling the CCSW flow.
Document ID # 5543459:
The 12 System Key Parameter Verification Program (LPCI System Discrepancy #4)
identified that no formal hydraulic calculation existed which demonstrated that the LPCI
system could provide the required 5,000 gpm flow through the containment cooling heat
exchanger to ensure adequate containment cooling.
'
This evaluation documented that the LPCI system could provide the required flow. The
capability was demonstrated primarily by Dresden Operating Surveillance (DOS)
1500-10, "LPCI System Pump Operability Test with Torus Available and lnservice
Testing (IST) Program," Revision 30 and NFS-BSA-D-97-03, "Sensitivity Analysis Post-
LOCA Containment Performance for Dresden Units 2/3," dated March 12, 1997. The
team determined that the evaluation was technically sound.
c.
Conclusions .
The team concluded that the calculations and evaluation were technically .sound.
However, the documentation of logic employed and the common site specific knowledge
used was not always evident and could have been improved with more attention to
detail.
E3.4
Updated Final Safety Analysis Report
a.
Inspection Scope (IP 37550)
The team reviewed sections of the UFSAR and the licensee's corrective action
documentation associated with a potential Dresden Lock and Dam failure.
b.
Observations and Findings
The team expressed a number of concerns with regards to the validity of some UFSAR
statements contained within Section ~.2.5.3.1, "Dam Failure during Normal Operations,"
and Section 9.2.5.3.2, "Dam Failure Coincident with a LOCA."
The team observed that the UFSAR did not accurately characterize the plant's
design-basis or the plant's capability to respond to a potential Dresden Lock and Dam
failure. As a result, the team had concerns with the ability of the plant to respond to a
dam failure as stated in the UFSAR.
The team's review of the licensee's "Summary of Dresden NRC Requirements for
1997," dated September 30, 1997, indicated that the licensee was aware of similar
concerns, although not identical to the team's. The licensee stated that several PIFs
9
related to this issue were in the corrective action process. The PIFs identified were:
PIF 227A-12-1997-012788, "UFSAR Implied One CCSW Pump Operation After
a Dam Failure Coincident With a LOCA," dated February 25, 1997
PIF D1997-05554, "UFSAR CCSW Piping Statement Discrepancy" dated
June 25, 1997
PIF D1997-05955, "UFSAR LPCI Flow Timing Discrepancy," dated June 24,
1997
PIF D1997~06487, "Incorrect Source Document Referenced for Diesel Generator
Cooling Water Pump in a Calculation," dated August 27, 1997
PIF D1997-08290, "NRC Concerns About CCSW System Performance After a
Dam Failure Coincident With a LOCA," dated November 25, 1997
This PIF was issued as a result of the team's concern that no high-point vent
valves were installed to vent trapped air during the reflood of the CCSW intake
bay, which was not considered by DOA-0010-01, "Dresden Lock and Dam
Failure," Revision 6.
In addition, the licensee stated that an evaluati~m had not been completed to determine
. whether the Dresden Nuclear Plant Design Basis required the plant to be capable of a
safe shutdown after a dam failure coincident with a Unit 2 or 3 LOCA and a loss of
offsite power (LOOP) ..
c. * Conclusions
The team had concerns that the UFSAR did not accurately characterize the plant's
design-basis or the plant's capability to respond to a potential Dresden Lock and Dam
failure. As a result, the team concluded that further review by the licensee and NRC
was required. An NRC URI was initiated to document these concerns.
(URI 50-237/249-97021-01 (DRS))
E4
Engineering Staff Knowledge and Performance
a.
Inspection Scope (IP 37550)
The team observed the performance of the engineering staff, interviewed both system
and design engineering personnel, and walked down plant systems with some system
engineers.
b.
Observations and Findings
All engineers interviewed appeared to be experienced and well qualified. However, the
turnover rate for some system engineers appeared to be high. The system engineerfor
r*
10
DC systems was only on the job for about six months. The system engineers for several
other systems were only on the job for about six months to 1 Yi years. However, the
team did not identify any specific problems directly linked to the lack of experience on
the part of the system engineers.
The team noted that the system engineers interviewed maintained good system
notebooks. The system engineers were required to walk down their systems
periodically. The team walked down selected plant systems with the system engineers,
and considered them knowledgeable on their assigned systems.
The team observed a surveillance test on the Unit 2 125 Volt alternate battery. The test
was modified performance test per procedure Dresden Engineering Surveillanee
(DES) 8300-52. As the DC system engineer at Dresden was relatively new to this test,
it was performed under the supervision of a system engineer from Braidwood. The
battery testing was done smoothly and no major problems were observed. The team
noted good communications with operations and maintenance during these tests.
c.
Conclusion
The team concluded that the system engineering department was adequately staffed.
The team determined that the engineers interviewed were qualifie*d and experienced in
the areas assigned. Good inter-departmental communications were noted between
system engineering, operations and maintenance during the special test observed.
E6
Engineering Organization and Administration
a.
Inspection Scope <IP 37550: IP 92703)
The team evaluated the performance and effectiveness of the DEAG to determine if the
CAL commitments and corrective actions were completed and had satisfied NRC
requirements.
b.
Observations and Findings
On November 21, 1996, CAL No. Rlll-96-016, was issued by the NRC as a result of
significant concerns with the station's control of calculations and with the overall
performance of site and corporate engineering activities. The CAL identified various
planned corrective actions to improve the performance of the engineering organization.
One of the planned activities was the formation of an engineering assurance group or
DEAG that was composed of senior ComEd engineering personnel and experienced
outside experts. The function of the group was to provide oversight of key engineering
activities until normal engineering functions had improved to the point where the reviews
were no longer necessary.
In NRC Inspection Report 50-237/249-97008(DRS), the NRC evaluated the CAL
activities and determined that the CAL commitments and corrective actions were
completed, except for those activities associated with the DEAG. The inspection
11
..
identified that initial DEAG implementation was not effective as an oversight
organization. As a result, the CAL remained open until effective DEAG performance
was demonstrated.
'
The team reviewed most of the DEAG review sheets for the period between June 1997
and October 1997, and determined that the DEAG reviews had in most cases,
documented relevant significant problems and appropriately required those documents
to be corrected. As a result, the DEAG reviews have improved the quality of the
engineering products. The DEAG reviews provided good recommendations for
improvements in methodology, technical content, and clarification and documentation
improvements that would make the engineering products a better source of information
for future users.
Since June 1997, the DEAG provided monthly reports to engineering management that
summarized the scope of the DEAG activities and the results of the DEAG reviews. The
DEAG observations were consistent through November 1997, in identifying areas that
needed improvement. The improvement areas were identified as follows:
Understanding of Regulatory or Design-Basis Requirements on Work Performed
- Attention to Detail
lnterdiscipline Reviews
The team observed that the DEAG reviews were generally thorough and technically
sound and produced similar observations with other licensee self-assessment efforts, as
described in Section E7. The DEAG efforts showed that the quality of the engineering
documentation has improved. However, the .team was concerned that many of the
DEAG members, who were engineering contractors, were no. longer employed at the
Dresden site and such loss of experienced personnel might degrade the licensee's
ability to maintain.an improving trend in engineering performance. Full staffing of
qualified personnel in the DEAG was a continuing problem.
c.
Conclusions
The team concluded that all commitments and corrective actions identified by
.
Confirmatory Action Letter (CAL) No. Rlll-96-016, dated November 21, 1996, including
those activities associated with the Dresden Engineering Assurance Group (DEAG)
have satisfied NRC requirements. The CAL was closed.
E7
Quality Assurance in Engineering Activities
a.
Inspection Scope (IP37550: IP40500)
The team reviewed the following self-assessment documents to assess quality and
proposed corrective actions:
12
"'-.
Report Number 237-230-97-00300, "Common Cause Analysis and Investigation
of an Adverse Trend in Human Performance Error-Related Licensee Event
Repo"rt (LER) Rate for the First Two Quarters of 1997 Which Resulted in
Exceeding the Dresden 50.54(f) Performance Criterion Action Level, Caused by
Failure to Make Timely Change and Inadequate Work Practices," Revision O
Report Number 237-251-97-05000, "Plant Engineering Work Management and
Support Responsiveness," dated November 18, 1997
DOC ID# 5549414, "Assessment of Engineerfng Department Safety Evaluation,"
Revision 0
b.
Observation and Findings
The team's review of the documents identified above indicated that the licensee had
taken .a pro-active position in an attempt to disclose the performance problems within
the organization. Many of the weaknesses identified described similar problems
previously identified by the NRC, but the make-up and the openness of the licensee's
conclusions indicated a positive trend. For example, the LER common cause analysis
investigation identified that the most prevalent problems were associated with personnel
acceptance of insuffiCient time to perform consistent quality technical reviews due to
shortcuts taken and inaccurate assumptions made during validation and verification
activities. The licensee stated t_hat the same type of errors were occurring station wide
and in a variety of processes. In addition, as discussed in Section E6, the DEAG *
consistently identified that the problems associated with engineering rework were
predominately due to inattention to detail as a result of not taking the time to perform ah
adequate detailed review.
The team observed that the self-assessment documents identified above were focused,
provided detailed and relevant observations, and provided a quality product. The
self-assessment corrective action recommendations were appropriate for the identified.
weaknesses. For example, the insufficient time pressure problem was addressed by the
LER common cause analysis investigation by the implementation of an Engineering
Rapid Response Team (ERRT) to remove short duration emergent work activities from
the system engineer's responsibility. In addition, an engineering reporting system (ERS)
was developed and implemented to provide a workload scheduling and tracking tool to
assist engineering personnel in managing workload.
The team observed that the proposed self-assessment corrective actions have not been
fully effective for all proposed recommendations. For example, the ERRT was effective
in reducing some of the readive workload; however, the ERS was too complex and not
user friendly to effectively prioritize and manage the engineers workload. The DEAG, as
discussed in Section E6; provided quality reviews that contributed to the overall
effectiveness of the licensee's self-assessments activities .
13
c.
Conclusions
The team concluded that the licensee's self-assessment activities were pro-active and
for the most part effective.
IV. Plant Support Areas
F2
Status of Fire Protection Facilities and Equipment
a.
Inspection Scope (IP40500: IP92904)
The team reviewed the licensee's corrective actions concerning problems associated
with the cpntrol room's HVAC system automatic smoke purge mode.
b.
Observations and Findings
During testing of the control room's HVAC system exhaust ducts in November 1994, the
licensee discovered that a prior inadvertent change to the control room's HVAC system
deleted the automatic smoke purge mode transfer capability as described in UFSAR,
Section 6.4.4.3. A URI 50-237/249-96002-07 was generated to track the concern and is
discussed further in Section F8.2.
The UFSAR stated that the control room's HVAC system was designed to isolate and
maintain design conditions within the control room during fires. In the event of smoke in
. the control room, the smoke purge mode would allow 100% outside air intake with no
recirculation of exhaust air into the control room HVAC zone (envelope). The UFSAR
further stated that smoke detectors automatically switched the control room's HVAC
system (Train A) to the smoke purge mode.
The licensee concluded that the problem occurred as a result of control room
modifications M12-2/3-82-1, M12-0-87-005, and M12-0-86-006. The smoke detectors
were inadvertently isolated as a result of modifications to the control room's envelope,
which deleted the automatic smoke purge mode transfer capability. As a result, control
room operators were required to take manual action to initiate the HVAC smoke purge
mode. A safety evaluation to ascertain whether the problem was an unreviewed safety
question was not initially performed by the licensee. Following NRC concerns, the
licensee performed a safety evaluation prior to startup from the 1996 Unit 2 refuel
outage. The licensee concluded that an unreviewed safety question did not exist.
A recent modification, M12-0-96-001, "Control Room HVAC Fire Protection System
Modification" corrected the deleted automatic smoke purge mode transfer capability by
installing smoke detector's in the remaining ventilation system. However, the team
identified that the description of the system's automatic initiation capability had been
removed from the UFSAR. Removal of the UFSAR's reference to the control room's
automatic transfer to the smoke purge mode was made during the performance of the
safety evaluation made in March 1996, just prior to the Unit 2 startup. The UFSAR
14
.,
..
change was made to accommodate the inadvenent change to the control room HVAC
system by only referencing the manual mode. The licensee stated that as a result of
two engineers not communicating, one engineer had taken the description for the
automatic initiati9n of the smoke purge mode out of the UFSAR.
I
- The safety evaluation performed in June 1996, for Modification M 12-0-96-001,
neglected to identify that a change to the UFSAR was required. As a result, during this
inspection, the licensee issued PIF# 01997-08239 to'correct the affected UFSAR
sections concerning the control room HVAC system's automatic initiation.
c.
Conclusions
The failure to perform a safety evaluation from November 1994 until March 1996, until
identified by the NRC, was a violation of 10 CFR 50.59.
(VIO 50-237 /249-97021-02(DRS))
F3
Fire Protection Procedures and Documentation
a.
Inspection Scope (IP40500: IP92904)
The team reviewed the licensee's corrective aGtions concerning problems associated
with the Fire Protection Report (FPR).
b.
Observations and Findings
The NRC previously identified that polyvinyl chloride (PVC) drain piping was installed
during a 1986 control rod drive modification and that the licensee had not performed a
safety evaluation nor added the increased combustible fire loading to the FPR's Fire
Hazards Analysis (FHA). In addition, the NRC also identified that the construction of a
turbine deck concrete building, which was another combustible fire load, had not been
added to the FHA. The licensee committed to perform a safety evaluation,
investigate/identify other unevaluated plant PVC usage, and specifically evaluate PVC
usage during the modification process and to include the identified combustible fire
' loads in the next update to the FHA. A URI 50-237/249-96002-09(DRS) was generated
to track the concern and is discussed further in Se.ction F8.3.
Branch Technical Position Auxiliary Power Conversion System Branch
(BTP APCSB) 9.5.1, "Guidelines for Fire Protection for Nuclear Power.Plants," dated
May 1976, was an FPR requirement, which required the minimization of PVC usage in
the plant. The team determined that the safety evaluation completed as part of the
licensee's corrective action was acceptable. During the licensee's investigation,
additional in-plant PVC usage was identified. In addition, the licensee had changed the
modification process to ensure that PVC usage was minimized in the plant.
The team observed, however, that the combustible fire load items were never added to
the FHA, which included the PVC usage and turbine deck concrete building previously
identified. The reason that the combustible fire load items had not been incorporated
15
into the FHA was that the FPR had not been updated since 1994. The FHA is part of
the FPR and the FPR was considered part of the :.JFSAR.
Generic Letter (GL) 86-10, "Implementation of cire Protection Requirements," dated
April 24, 1986, stated that fire protection plans and programs shall be incorporated as
part of the UFSAR and therefore, would be updated and submitted to the NRC in
accordance with the requirements of 10 CFR 50.71(e). GL 86-10 also stated, "All
changes to the approved program shall be reported annually to the Director of the Office
of Nuclear Reactor Regulation, along with the UFSAR revisions required by
1 O CFR 50.71 (e)." The failure to submit revised portions of the FPR to the NRC was a
violation of 10 CFR 50.71(e). (VIO 50-237/249-97021-03(DRS))
The team also observed a weakness within the licensee's corrective action-process
concerning these earlier identified FPR problems. Following NRC Inspection
Report 96002 (February 14, 1996, through March 29, 1996), Quality and Safety
Assessment (Q&SA) wrote Corrective Action Record (CAR) 12-96-151 "Fixed
Combustible Loading." The CAR identified that, contrary to the requirements of
GL 86-1 O .and Engineering Procedure ENC-QE-85, "Control and Revision of the Fire
Protection Program Documentation," updates to the FHA Report, which was part of the
FPR, had not been submitted to the NRC. A PIF and NTS item were generated on
December 12, 1996, 10 months after the identification of the earlier FPR problems.
NTS history indicated that completion of the FPR update was extended from June 30,
1997, to September 1, 1997, and then to December 18, 1998. In addition, on
November 7, 1997, Q&SA identified that there was no process to receive, evaluate,
track, and update FPR information.
On November 19, 1997, the licensee opened NTS Item #237-225-97R12-97242 to track
the development of a procedure to control updating of the FPR and provide interim
tracking of FHA changes. A due date of September 4, 1998, was assigned to the NTS
item .. Currently, the FPR does not represent plant conditions. The identification and
corrective actions for FPR problems were not timely.
The team further identified that fire risks associated with the additional combustible fire
loading had not been incorporated into the fire pre-plans. Technical Specification (TS) 6.2.A stated that written procedures shall be established and implemented covering
these activities. Dresden Fire Protection Procedure (DFPP) 4100-01, Revision 1, "Fire
Protection Program," required that Fire Pre-Plans be updated annually. The Dresden
"Fire Pre-Plans," Revision 2, had not been updated since September 1992. The
licensee's failure to comply with these requirements was a violation of TS 6.2.A.
(VIO 50-237 /249-97021-04(DRS))
c. Conclusions
Failure to update and submit the revised portions of the FPR to the NRC was a violation
of 10 CFR 50.71(e). (VIO 50-237/249-97021-03(DRS)) Failure to update the fire
pre-plans was a violation of TS 6.2.A (VIO 50-237/249/97021-04(DRS))
16
F8
Miscellaneous Fire Protection Issues (IP92904)
FB.1
(Closed) VIO 50-237/249-96002-05B(DRS): This violation was issued for not
performing a full 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> discharge test on 47 Appendix R emergency lighting units as
required by DES 4153-04, "Emergency Lighting Discharge Test," Revision 0. The
licensee changed the procedure/surveillance to ensure that the batteries were discharge
tested for the full 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The team reviewed two years of surveillance data and
determined that the licensee's corrective actions were effective. This item was closed.
FB.2
(Closed) URI 50-237/249-96002-07(DRP): This unresolved item was issued for
inadvertently deleting the control room HVAC system automatic smoke purge mode
transfer capability as described in UFSAR, Section 6.4.4.3. The change to the
automatic smoke purge mode had been made as a result of a control room modification.
A recent modification corrected the control room's HVAC system automatic smoke
purge mode problem. However, a violation was issued for not performing a safety
evaluation as discussed in Section F2. This item was closed.
FB.3
(Closed) URI 50-237/249-96002-09(DRS): This unresolved item was issued for using
PVC during a modification without performing a safety evaluation. The licensee
completed the safety evaluation and concluded there was no unreviewed safety
question. During the team's review, the FPR was observed as not having been updated
for PVC usage and the addition of a turbine deck concrete building. As a result, a
violation was issued for not having updated the FPR since 1994 as discussed in
Section F3. This item was closed.
V. Management Meetings
X1
Exit Meeting Summary
The team presented the final inspection results to members of licensee management at the
conclusion of the inspection on January 27, 1998. The team initially met with the licensee's
representatives to summarize the scope and findings of the on-site inspection activities on
November 26, 1997. During both of these meetings, the team questioned licensee personnel
as to the potential for proprietary information being included or retained in the inspection report
material as discussed at the exits. No proprietary information was identified as included or
retained.
17
PARTIAL LIST OF PERSONS CONTACTED
Licensee
G. Abrell, NRC Coordinator; Regulatory Assurance
D. Ambler, Regulatory Assurance Supervisor (Acting), Regulatory Assurance
. H. Anagnostopoulos, Corrective Action Process (CAP) Supervisor, Quality & Safety Assessment
R. Book, CAP Staff, Quality & *Safety Assessment
A. Casillo, Mechanical Lead (M1 ), Design Engineering
W. Clover, Design Engineer, Design Engineering
J. Dawn, DEAG Supervisor, Plant/Engineering Programs
F. Fink, Business Manager, Dresden
M. Friedmann, HP Technical Lead, Health Physics
R. Freeman, Site Engineering Manager, Dresden
W.Halcott, Auxiliary System Lead, Systems Engineering
M.Heffiey, Site Vice President, Dresden
K. Housh, ISEG Engineer, Quality & Safety Assessment
L. Jordan, Training Manager (Acting), Training
A. Khanna, Design Lead, Design Engineering
J. Kish, CCSW System Engineer, Systems Engineering.
W. Lipscomb, Assessor, Site Vice PresidentStaff
R. Mahendranathan, Mechanical Engineer, Design Engineering
T. McGowan, DC System Engineer, Electrical System & Components
E. Netzel, Director, Supplier Evaluation Services/Nuclear Oversight
K. Peterman, Supervisor, Configuration & Administration Management; DEAG Member
P. Planing, Superintendent, Systems Engineering
P. Racicot; AC System Engineer, Electrical System & Components
C. Richards, Audit Supervisor, Quality & Safety Assessment
E. Salinas, System Engineer, Systems Engineering .
B. Shete, Mechanical Engineer, Design Engineering
F. Spangenberg, Regulatory Assurance Manager, Dresden
D. Spencer, Electrical System & Components Lead, Systems Engineering
S. Tutich, Electrical Lead, Design Engineering
L. Weir, Superintendent, Design Engineering
D. Winchester, Manager, Quality & Safety Assessment
ComEd Contractors
H. Campbell, Member, DEAG (Titan)
C. Kinstler, Engineer (Sargent & Lundy)
H. McCullough, Site Lead (Acting), Design Basis Initiative (Sargent & Lundy)
18
[j
LIST OF INSPECTION PROCEDURES USED
IP 37550:
Engineering
IP 40500:
Effectiveness of Licensee Controls in ldentifyin!J, Resolving, and Preventing Problems
IP 92703:
- Followup of Confirmatory Action Letters
IP 92904:
Followup - Plant Support
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
50-237/249-97021-01 (DRS)
50-237 /249-97021-02(DRS)
50-237 /249-97021-03(DRS)
50-237 /249~97021-04(DRS)
Closed
50-237 /249-96002-05B(DRS)
50-237 /249/96002-07 (DRP)
50-237 /249-96002-09(DRS)
UFSAR Dam Failure Discrepancies
Failure to Perform 50.59 Evaluation
Failure to Update FPR and Submit to NRC
Failure to Update Fire Pre-plans
Failure to Adequately Test Emergency Lighting
Untimely Resolution of Operability Evaluations
Polyvinyl .Chloride (PCV) Usage Not Well Controlled
19
'
ATTN
ccsw
CFR
Com Ed
OAP
DEAG
DES
DFPP
DTI
E&TS
GL
ISEG
JSPLTR
LPM
NEP
NOC-BOD
NRC
NTS
Q&SA
rs*
LIST OF ACRONYMS
Attention
Boiling Water Reactor
Confirmatory Action Letter
Corrective Action Record
Containment Cooling Service Water
Code of Federal Regulations
Commonwealth Edison
Dresden Administrative Procedure
Dresden Engineering Assurance Group
Dresden Engineering Surveillance
Dresden Fire Protection Procedure
. Division of Reactor Projects
Division of Reactor Safety
Desk Top Instruction
Engineering and Technical Support
Generic Letter
Heating, Ventilation, and Air Conditioning
Independent Safety Engineering Group
ComEd (J.S. Perry) Letter
Loss of Coolant Accident
Low Pressure Coolant Injection
Licensing Project Manager
Mean Sea Level
Nuclear Engineering Procedure
Nuclear Operating Committee-Board of Directors
Nuclear Regulatory Commission
Office of Nuclear Reactor Regulation
Nuclear Tracking System
Public Document Room
Problem Identification Form
Polyvinyl Chloride
Quality and Safety Assessment
Reactor Building Closed Cooling Water
Regulatory Guide
Senior Resident Inspector
Technical Specification
Updated Filial Safety Analysis Report
Unresoived Item
Unreviewed Safety Question
Violation
20
PARTIAL LIST OF DOCUMENTS REVIEWED
DOCUMENT
REVISION OR
NUMBER
DOCUMENT DESCRIPTION
DATE ISSUED
CAL No. Rlfl-96-016
Confirmatory Action Letter
November 21, 1996
CAR 12-96-151
Fixed Combustible Loading
December 23, 1996
CAR 12-97-105
Fire Protection Report
November 7, 1997
OAP 02-27
The Integrated Reporting Process (IRP)
Revision 7
OAP 21-03
Processing Plant Design Changes
Revision 13
DEAG Review Sht 8.10
Removal of Description of Acid & Caustic
August 21, 1997
Equipment from UFSAR
DEAG Review Sht 8. 11
Troubleshooting of a Stator Leak
August22, 1997
DEAG Review Sht 8.16
Clarification of Information on an Overhead Crane
August 22, 1997
DEAG Review Sht 8.17
Clarification of Spent Fuel Pool Liner Thickness
August22, 1997
DEAG Revi~w Sht 8.28
Security Position Title Change in the UFSAR
August28, 1997
DES 4153-04
Emergency Lighting Discharge Test
Revision 0
DFPP 4100-01
Revision 1
Dresden Engineering Assurance Group Activities
June 26, 1997
for May, 1997 (1st DEAG Monthly Report)
DOC ID# 0005458065
Dresden Engineering Assurance Group Activities
July 11, 1997
for June, 1997
DOC ID# 0005491140
Dresderi Engineering Assurance Group Activities
August 18, 1997
for July, 1997
DOC ID# 0005503264
Dresden Engineering Assurance Group Activities
September 8, 1997
for August, 1997
DOC ID# 0005558157
Dresden Engineering Assurance Group Activities
November 17, 1997
for October, 1997
21
,.
- .
DRE 97-0171
Calculation for Determination of Acceptance
Revision O
Criteria for CCSW One and Two Pump NPSH
Testing - Units 2 & 3
DRE 97-0172
Calculation to Determine Submergence for
Revision 0
Excessive CCSW Intake Vortexing Prevention.
DTl-DE-15 .
Roles and Responsibilities of the Dresden
Revisions 0, 1, 2
Engineering Assurance Group
ENC-QE-85
Control and Revision of the Fire Protection
Program Documentation
Eval Doc ID #5543459
CAL Action Item Update Report Following First
December 30, 1996
Monthly Status Meeting Held December 19, 1_996
Implementation of Fire Protection Requirements
April 24, 1986
JSPL TR: 97-0005
ComEd Interim Response to NRC Independent
January 13, 1997
Safety Inspection Report
JSPL TR: 97-0041
ComEd Response to NRC Independent Safety
February 26, 1997
Inspection Report
JSPL TR: 97-0043
Verification Screening of Key Parameters for
Revision 0
Twelve Risk Significant Systems
M 12-0-96-001
Control Room HVAC Fire Protection System
Modification
NEP-04-01 DR
Dresden Plant Modification Site Appendix
Revision 2
NEP 10-03
Disposition of Design Basis Discrepancies
Revision 0
N~P 12-01
Preparation, Review, and Approval of Design
Revision 2
Input Requirements
NEP 12-02
Preparation, Review, and Approval of
Revision 4
Calculations
NSWP-A-15
ComEd Nuclear Division Integrated Reporting
Revision 0 & 1
I
Program
OP EVAL 97-81
Minimum Water Level in CCSW Intake Bay
July 8, 1997
PIF # D1997-05554 *.
UFSAR CCSW Piping Statement Discrepancy
June 25, 1997
22
.t
PIF # D1997-05556
PIF # D1997-05955
PIF # D1997-06487
PIF # D1997-08239
PIF # D1997-08290
PIF # 227A-12-1997-012788
Report Base NTS Number:
237-251-97-05000
UFSAR Safety Grade Cold Shutdown Capability
June 25, 1997
Discrepancy
UFSAR LPCI Flow Timing Discrepancy
June 24, 1997
Incorrect Source Document Referenced for Diesel
August 27, 1997
Generator Cooling Water Pump in a Calculation
UFSAR Deletion/Addition
November 21, 1997
NRC Concerns About CCSW System
November 25, 1997
Performance After a Dam Failure Coincident With
a LOCA
UFSAR Implied One CCSW Pump Operation
February 25, 1997
After a Dam Failure Coincident With a LOCA
PlanUPrograms Engineering Sel Assessment
November 18, 1997
3-7 Nov 97
Fire Protection Report (FPR)
- Amendment 1 O
December 1994
23