ML17191A598
| ML17191A598 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 04/14/1998 |
| From: | Jeffrey Jacobson NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17191A594 | List: |
| References | |
| 50-237-98-07, 50-237-98-7, 50-249-97-07, 50-249-97-7, NUDOCS 9804170103 | |
| Download: ML17191A598 (12) | |
See also: IR 05000237/1998007
Text
U.S. NUCLEAR REGULATORY COMMISSION
Docket Nos:.
License Nos:
Report No:
Licensee:
Facility:
Location:
Dates:
Inspector:
Approved by:
9804170103 980414
ADOCK 05000237
G
REGION Ill
50-237; 50-249
50-237/98007(DRS}, 50-249/98007(DRS)
Commonwealth Edison Company
Dresden Generating Station, Units 2 and 3
6500 North Dresden Road
Morris, IL 60450-9765
February 12 through March 5, 1998
Gerard F. O'Dwyer, Reactor Inspector
John Jacobson, Chief
Lead Engineering Branch
Division of Reactor Safety
EXECUTIVE SUMMARY
Dresden Generating Station, Units 2 and 3
NRC Inspection Report 50-237/98007(DRS); 50-249/98007(DRS)
An announced regional initiative inspection that reviewed portions of the M& TE calibration
control program and issues related to Dresden Station's response to a hypothetical failure of
the Dresden Dam.
Overall the inspection concluded that the M& TE calibration control program was
effective. M& TE storage controls were adequate and plant temperature instrumentation
was properly calibrated. One violation of NRC requirements was identified.
(Section M 1.1, VIO 50-2371249-98007-01 (DRS))
Various scenarios were identified concerning dam failure both with and without a LOCA,
as described in the UFSAR, which the station may not be able to accommodate.
(Section E3)
2
M1.1
a.
Report Details
II. Maintenance
Measuring and Test Equipment (M&TEl Control Program
Inspection Scope (IP 61725)
The inspector reviewed portions of the M& TE control program, M& TE calibration
re90rds, Technical SpeCifications and M&TE vendor information.
b.
Observations and Findings
The inspector reviewed selected Problem Identification Fc;>rm (PIF) packages related to
the M&TE program.initiated during 1997 and 1998 and found the PIFs appropriately
initiated and dispositioned.
Temperature-initiated safety-related automatic actions were actuated by temperature
switches that were calibrated appropriately to the necessary accuracy to meet the
TS requirements. The control room operators indicated that safety-related manual
. actions (e.g., Emergency Operating Procedure actions) were taken based on the
outputs of annunciators and computer alarms which were also calibrated with
appropriate accuracy.
The M& TE storage controls were found to be adequate. The storage requirements for
the M&TE were very broad, e.g., the vendor-recommended storage range for a Gordon
model 5060 meter was from -40°F to 140°F with no restriction on humidity. The
temperature in the M&TE storage areas was controlled between 60 and 90°F by normal
ventilation.
Calibration lab personnel were required to be trained on the procedures that were used
to calibrate the M& TE. The training records indicated that the calibration lab personnel
had the appropriate training.
Technical Specification 6.8.A.1 required the implementation of certain Regulatory
Guides and American National Standards Institute (ANSI) standards including ANSI
N45.2.9-1974, "Requirements for Collection, Storage, and Maintenance of Quality
Assurance Records for Nuclear Power Plants." Appendix A of ANSI N45.2.9-1974
required that M& TE calibration records be retained as quality records for five years.
However, the licensee identified that prior to December 22, 1997, the Master Records
Retention Schedule of General Procedure GP 136, dated September 15, 1995,
"Retention of Company Records," failed to designate M& TE calibration records as
quality assurance (QA) records and the procedure allowed the calibration records to be
disposed after three years. This was a severity level IV violation of TS 6.8.A.1
(VIO 50-237/249-98007-01 (DRS)); however, safety consequences were minimal
because the licensee had not needed to retrieve any calibration records older than three
years. Also, all M&TE was calibrated at least once a year, creating a more recent
3
calibration record so the potential to need an older record was reduced each year. On
January 19, 1998, the Dresden Central File Supervisor approved a change to the
Master Record Retention Schedule to require the M& TE calibration records to be
retained as QA records for at least five years. The M& TE supervisor informed the
inspector that: 1) the Master Schedule was for all Com Ed sites and changes sometimes
required an extensive time to be implemented, therefore the Dresden Nuclear Power
Station Record Retention Schedule had been created; 2) the change still had to be
approved and implemented on the Master Schedule by ComEd corporate personnel;
and 3) while the Master schedule was being corrected, the Dresden schedule had
already been changed to meet the requirements. When responding to the violation, the
licensee should specify when the change to the Master schedule will be implemented.
c.
Conclusions
The M& TE calibra,ion control program was effective in maintaining plant temperature
instrumentation properly calibrated. PIFs related to M& TE were properly dispositioned.
Failure to designate and maintain M&TE calibration records as QA records for five years
was a severity level IV violation of TS 6.8.A.1. (VI~ 50-237/249-98007-01 (DRS))
Ill. Engineering
E3.1
Updated Final Safety Analysis Report (UFSAR) Section 9.2.5.3.2 Review
a.
Inspection Scope OP 37550)
The inspector reviewed section 9.2.5.3.2, "Dam Failure Coincident with a LOCA," of the
UFSAR, Revision 2, docketed letter dated October 16, 1968, from the Atomic Energy
Commission (AEC) staff to Dresden staff amendments 9 and 1 O to the applications for
the operating licenses for Unit 2 & 3, and docketed letter dated March 13, 1998, from
the site vice president to the NRC.
b.
Observations and Findings
The im:;pector noted that Section 9.2.5.3.2 stated that the Dresden Station could be
safely shutdown if there was a catastrophic failure of the Dresden dam coincident with a
design basis loss of coolant accident (LOCA ) in either Unit 2 or Unit 3, without using
. any seismic Class II systems and with a loss of offsite electrical power (LOOP). The.
inspector determined that it could not be demonstrated that Section 9.2.5.3.2 could be
met without using Class II systems for isolation condenser make up and the cognizant
senior design engineer agreed. The inspector also determined that even if the Class II
systems were assumed to be operational, the Containment Cooling Service Water
(CCSW) system would not maintain the required 30 pounds per square inch differential
(psid) greater than the low pressure coolant injection (LPCI) system and Part 100 limits
may be exceeded. The Site Engineering Manager and the licensee's NRC Coordinator
informed the inspector that the Com Ed position was that this capability was not required
by the license or the design basis. However, the inspector noted that by docketed letter,
4
c.
dated October 16, 1968, the Atomic Energy Commission (AEC) staff had requested the
Dresden staff to provide an evaluation to complete the application for the operating
licenses for Units 2 & 3. The AEC required the evaluation to describe the effect of an
earthquake which disabled the dam, all Class II systems, offsite power and caused a
design.basis LOCA in one of the two units. By docketed letter, dated February 28,
1969, Dresden answered, in amendments 9 and 10, to the applications for the operating
licenses for Unit 2 & 3 that the station can safely shutdown after all those coincident.
events. Dresden also placed the. evaluation in the FSAR. By docketed letter, dated
March 31, 1998, the Dresden site vice president stated that a dam failure coincident with
a LOCA was beyond the design basis of the Dresden Station and that clarifications to
the UFSAR would be made through the 10 CFR 50.59 provisions.
Conclusions
It was. not demonstrated that the station could shutdown per the UFSAR statements in
Section 9.2.5.3.2 post dam failure coincident with a LOCA. This will be reviewed as part
of a previously opened unresolved item. (URI 50-237/249-97021-01A (DRS))
E3.2
Review of UFSAR section 9.2.5.3.1. "Dam Failure During N.ormal Plant Operation"
a.
Inspection Scope
The inspector reviewed UFSAR section 9.2.5.3.1, "Dam Failure During Normal Plant
Operation."
b.
Observations and Findings
The inspector noted that section 9.2.5.3.1 stated that the Dresden station could be
safely shutdown after a dam failure with no LOCA, without using Seismic Class II
systems and with a LOOP. The section discussed only methods of shutdown which
relied on Class II systems that required special lineups to be powered from the
Emergency Diesel Generators. The inspector noted that similar statements were in the
AEC-requested evaluation discussed in section E3.1. The inspector determined that the
Dresden station could not be safely shutdown by only Class I systems under the stated
conditions because all Isolation Condenser Makeup methods used Class II systems and
the containment cooling service water (CCSW) system would not function without
Class II systems. The. licensee stated that this requirement was not part of the design
basis of the Dresden Station and that UFSAR clarifications will be made as discussed in
paragraph E3.1.
Even if the Class II systems were assumed to operate, the Dresden station might still
not be able to shutdown safely after a catastrophic dam failure that lowered the intake
canal water level to the postulated 495' MSL. The inspector was eoncemed that the
Unit 2 & 3 reactor vessels (RV) might not be adequately cooled. Section 9.2.5.3.1
stated that the isolation condensers (IC) would be used to cool the .RVs. All Isolation
Condenser makeup methods used Class II systems; however, the licensee assumed
one would still function. The AEC-requested evaluation stated that there would be nine
5
million gallons available for IC makeup in the Ultimate Heat Sink (UHS) because the
discharge canal level would fall to 498' MSL after a dam failure. The inspector identified
that the discharge canal level would fall to 495' MSL which would result in six million
gallons available in the UHS. Such a significant underestimation of UHS capacity might
invalidate the ability of the Unit 2 & 3 ICs to cool the RV. The licensee stated that this
UFSAR statement will be corrected by a 50.59 UFSAR change.
The UFSAR discussed using the Service Water Pumps (SWPs) if the intake level
dropped to 495 feet MSL after a dam failure. The inspector was concerned because the
Hydraulic Institute Standards, ANSI/HI~ 1994 Edition, Figure 1.66 contained the general
recommendation of a minimum of five feet submergence to prevent excessive vortexing
of a pump with 15,000 gpm rated capacity such as each SW pump. If intake level
dropped to 495 feet, the SWPs would only have 1 foot 2.5 inch submergence. This
might allow enough air-entraining vortexing for the SWPs to lose their prime and to not
function. The licensee was generating calculations to demonstr~te that the Dresden
SWPs will operate adequately with intake at the postulated 495 feet level. If the SWPs
do not function at full capacity, the SWPs might not be able to be used to cool the
reactor building closed cooling water (RBCW) and achieve cold shutdown via the
shutdown cooling heat exchangers as stated in the UFSAR. Also, the inspector
identified that since the SWPs indirectly cooled the reactor recirculation pump (RRP)
seals, the seals may eventually fail if the SWPs are at less than full capacity. *
Additionally, loss of CCSW, as discussed in S~ction E3.3, may impact long term cooling.
of the plant.
c.
Conclusions
It was not demonstrated that Dresden station could be safely shutdown during normal
operation following a dam failure as described in Section 9.2.5.3.1 of the UFSAR, using
seismic Class I systems only. The licensee does not believe that this requirement is
part of the design basis and intends to clarify the UFSAR. However, assuming no
damage to Class II systems, the UHS inventory is less than that assumed in the
evaluation and SWP performance at the 495' MSL has not been evaluated. These
concerns will be reviewed as part of a previously opened unresolved item.
(URI 50-237/249-97021-01 B(DRS))
.
E3.3
Procedural Problems
a.
Inspection scope <IP 37550. 92903)
The inspector reviewed the CCSW suction piping drawings and Procedure
DOA 0010-01, Rev. 7, "Dresden Lock and Dam Failure."
b.
Observations and Findings
The UFSAR stated that a CCSW pump would be put in-service after a LOCA coincident
with a dam failure that lowered the intake water level to 495 feet MSL. The bottom of
6
the CCSW suction line was at 498 feet MSL, so the suction piping would drain and
- approximately 200 feet of horizontal piping for each pump would fill with air. The
UFSAR stated that the CCSW intake bay would be sealed and the water level raised so
that the CCSW intakes would be covered and the CCSW pumps 6ould provide water for
containment cooling. However, the inspector identified that there were no high-point
vent valves to vent the air that would be trapped inside the piping when water level was
restored. With the water level restored, the trapped air volume would prevent CCSW
pump operatiqn. Previous to December 1997, Procedure DOA 0010-01, Revision 7,
"Dresden Lock and Dam Failure," did not have provisions for venting the trapped air.
The licensee revised Procedure DOA 0010-01 to cut holes in the top of the CCSW
suction piping in the CCSW bay to vent the air. Since the holes would be at 500' MSL,
the inspector was concerned that the CCSW bay water level might not be able to be.
restored high enough to prevent excessive air-entraining vortexing through the holes
which may prevent the CCSW pumps from pumping adequate water.
The UFSAR stated that the CCSW intake bay would be sealed by installing stoplogs
across the two openings to the bay. The licensee did not quantify the maximum leakage
expected. The inspector was concerned that the stoplogs would leak excessively,
further reducing the CCSW capacity. Evaluatior) of CCSW pump performance with vent
holes cut in the suction piping and the stoplogs in place is considered an unresolved
item. (URI 50-237/249-98007-02(DRS))
c.
Conclusions
Prior to December 1997, Procedure DOA 0010-01 would not have supported CCSW
pump operation after a dam failure as stated in the UFSAR. Furthermore, adequate
pump performance with suction piping vents and intake bay stoplogs has not been
demonstrated.
E3.4
Other UFSAR Discrepancies
a.
Inspection scope OP 37550. 92903)
The inspector reviewed UFSAR section 9.2.5.3.1, "Dam Failure During Normal
Operations," PIF D1998-00455, "Suction of Unit 2 & 3 Diesel-Driven Fire Pump
Uncovered Following Dam Failure," and interviewed the Diesel Fire Pump (DFP)
System Engineer.
b.
Observations and Findings
The UFSAR stated that the DFP's suction was at 492', the DFP took suction from the
CCSW bay and the DFP would operate after a dam failure lowered CCSW bay level to
- 495' MSL. However; the inspector identified that-the suction could not be at 492'
because the CCSW bay floor elevation was at 493' 8" MSL. PIF D1998-00455 was
initiated which stated that the vendor-recommended minimum intake water level was
498' 10." Engineering personnel stated that the CCSW bay level.would not be restored
7
c.
until two hours after a dam failure. The inspector was concerned because even if bay
level was restored, the DFP would be competing with the CCSW pumps for water if the
DFP was needed. The docketed letter dated March 31, 1998, from the site vice
president stated that the license did not require the DFP to function after a dam failure.
The licensee stated the DFP suction level specified in the UFSAR will be corrected by
the UFSAR 50.59 change process. The NRC will review the safety evaluation for this
change. (IFI 50-237/249-98007-03(DRS))
Conclusions
The DFP intake water level post dam failure may not be adequate to support DFP
operation. The acceptability of a two hour delay and reduced performance for the DFP
after a dam failure will be reviewed when the 50.59 evaluation is made available.
E3.5
Canal Pikes
a.
Inspection Scope (IP 37550)
The inspector performed a walkdown of some of the canals and reviewed portions of the
safety evaluation of Hydrology SEP topics 11.3.C, SafetY Related Water Supply (Ultimate
Heat Sink) contained in docketed letter dated June 21, 1982.
b.
Observations and Findings
A dike is located at the south side of the hot canal to the Dresden lake near the
intersection of Collins and Dresden roads. All the canal dikes were seismic Class II and
assumed to fail after an earthquake strong enough to destroy the dam. It appeared that
the failure of this dike might lower intake level to below 495' MSL. This failure did not
appear to have been analyzed in the safety evaluation of Hydrology SEP topics 11.3.C,
Safety Related Water Supply (Ultimate Heat Sink). Review of the potential for dike
failure to lower the intake level below the assumed 495' MSL is considered an
unresolved item. (URI 50-237/249-98007-04(DRS))
c.
Conclusions
The inspector identified a potential for dike failure to lower the intake bay below the
495' MSL. Evaluation of this potential could not be located during the inspection .
8
t
V. Management Meetings
X1
Exit Meeting Summary
The inspector presented the final inspection results to members of licensee management at the
conclusion of the inspection on March 5, 1998. During the meeting, the inspector 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.
9
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PARTIAL LIST OF PERSONS CONTACTED
Licensee
G. Abrell, NRC Coordinator, Regulatory Assurance
R. Book, CAP Staff, Quality & Safety Assessment .
A. Casillo, Mechanical Lead (M1), Design Engineering
P. Channel, Control Rod Drive System Engineer, Systems Engineering
M. Crowley, Circulating Water System Engineer, Systems Engineering
G. Feige, M& TE Supervisor
J. Fox, Senior Design Engineer, Design Engineering
R. Freeman, Site Engineering Manager, Dresden
W. Halcott, Auxiliary System Lead, Systems Engineering
J. Kish, CCSW System Engineer, Systems Engineering
K. Peterman, Supervisor, Configuration & Administration Management; DEAG Member
P. Planing, Superintendent, Systems Engineering
W. Poppe, Reactor Recirculation System Engineer; Systems Engineering
B. Shete, Mechanical Engineer, Design Engineering
F. Spangenb.erg, Regulatory Assurance Manager, Dresden
L. Weir, Superintendent, Design Engineering
LIST OF INSPECTION PROCEDURES USED
IP 37550:
Engineering
IP 61725 *
Surveillance Testing and Calibration Control Program
Follow up - Engineering
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
50-2371249-98007-01 (DRS)
50-237 /249-98007-02(DRS)
50-237 /249-98007-03(DRS)
50-237 /249-98007-04(DRS)
Discussed
50-237/249-97021-01 (DRS)
Improper M&TE record retention requirements
CCSW Pump Operability, post dam failure
IFI
50.59 to evaluate DFP use during dam failure
Evaluate effect of dike failure on intake bay level
UFSAR Dam Failure Discrepancies
10
ATTN
ANSI
ccsw
CFR
Com Ed
DES
E&TS
GL
JSPLTR
LPM
NEP
NRC
NTS
psid
Q&SA
TS.
LIST OF ACRONYMS
Attention
American National Standards Institute
Boiling Water Reactor
Containment Cooling Service Water
Code of Federal Regulations
Commonwealth Edison
Dresden Administrative Procedure
Dresden Engineering Surveillance
Diesel Fire Pump
Division of Reactor Projects
Division of Reactor Safety
Engineering and Technical Support
Generic Letter
ComEd (J.S. Perry) Letter
Loss of Coolant Accident
Loss of Offsite Electrical Power
Low Pressure Coolant Injection
Licensing Project Manager .
Mean Sea Level
Nuclear Engineering Procedure
Nuclear Regulatory Commission
Office* of Nuclear Reactor Regulation
Nuclear Tracking System
Public Document Room
Problem Identification Form
pounds per square inch differential
Quality and Safety Assessment
Reactor Building Closed Cooling Water
Regulatory Guide
Systematic Evaluation Program
Safety Evaluation Report
Senior Resident Inspector
Technical Specification
Updated Final Safety Analysis Report
Unresolved Item
Unreviewed Safety Question
Violation
11
DOCUMENT
NUMBER
PIF # D1997-08290
PIF # 227A-12-1997-012788
PARTIAL LIST OF DOCUMENTS REVIEWED
REVISION OR
DOCUMENT DESCRIPTION
DATE ISSUED
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
12