ML20217D067
| ML20217D067 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 03/23/1998 |
| From: | Jeffrey Jacobson NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20217D010 | List: |
| References | |
| 50-373-97-23, 50-374-97-23, NUDOCS 9803270293 | |
| Download: ML20217D067 (11) | |
See also: IR 05000373/1997023
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U.S. NUCLEAR REGULATORY COMMISSION
REGIONlli
Docket Nos:
50-373;50-374
License Nos:
Report Nos:
50-373/97023(DRS); 50-374/97023(DRS)
Licensee:
Commonwealth Edison Company
Facility:
LaSalle County Station, Units 1 and 2
Location:
2601 N. 21st Road
Marseilles,IL 61341
Dates:
December 22,1997, through March 6,1998
Inspector:
Eric Duncan, Reactor Engineer
Approved by:
John M. Jacobson, Chief
Lead Engineers Branch
Division of Recctor Safety
9803270293 980323
ADOCK 05000373
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- EXECUTIVE SUMMARY
LaSalle County Station, Units 1 and 2
NRC inspection Report 50-373/97023(DRS); 50-374/97023',DRS)
Engineering
The inspector reviewed the licensee's response to information Notice 87-10 related to
the potent;al for waterhammer in the residual heat removal (RHR) system if a Loss-Of-
Coolant-Accident (LOCA) concurrent with a Loss-Of-Offsite-Power (LOOP) were to
occur while the system was aligned for suppression pool cooling. (Section E8.1)
Allowable tolerances in the construction of the drywell may cause a delay in leakage
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entering the floor drain sump for detection. However, technical specification leakage
requirements would be still be met assuming the worst case drywell floor holdup volume.
(Section E8.2)
The shell side of the Unit 1 and Unit 2 RHR pump seal coolers did not meet design
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pressure requirements and were not procured as required by 10 CFR 50, Appendix B,
Criterion IV," Procurement Document Control." _ (Section E8.3)
A design modification to add screens to the Unit 2 floor and equipment drain sumps to
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prevent foreign material intrusion into the sump piping and containment isolation valves
was not controlled as required by 10 CFR 50, Appendix B, Criterion ill, " Design Control."
(Section E8.4)-
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Report Details
Exercise of Enforcement Discretion
A violation described in Section E8.3 of this report is based upon licensee activities which were
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identified after, but occurred prior to the licensee announcing, in December 1996, an extended
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shutdown of the LaSalle County Station. This violation satisfies the appropriate criteria in
Section Vll.B.2," Violations identified During Extended Shutdowns or Work Stoppages," of the
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" General Statement of Policy and Procedures for NRC Enforcement Actions,"(Enforcement
Policy), NUREG-1600, and a Notice of Violation is not being issued for this violation because
the criteria specified in Section Vll.B.2 were met, which allows enforcement discretion to be
applied. Specifically, the violation was licensee-identified as a result of a comprehensive
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program for problem identification and correction that was developed in response to the
shutdown, the violation would not be categorized at a severity level higher than Severity Level
ll, and the violation was not willful. In addition, actions specified in Confirmatory Action Le#er
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Rlll-96-0088 effectively prevent the licensee from starting up LaSalle County Station witl at
implicit NRC approval.
Ill. Engineering
E8
Miscellaneous Engineering issues
E8.1
(Ocen) Insoection Follow uo item 50-373/97013-01: 50-374/97013-01: Review of
As discussed in inspection report 50-373/97013; 50-374/97013, the inspector reviewed
the licensee's response to Information Notice 87-10 related to the potential for
waterhammer in the RHR system if a Loss-Of-Coolant-Accident (LOCA) concurrent with
a Loss-Of-Offsite-Power (LOOP) were to occur while the system was aligned for
suppression pool cooling.
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As part of that effort, the inspector reviewed the RHR waterhammer analysis prepared
by Sargent & Lundy (S&L) which concluded that although a waterhammer would occur,
the RHR system would maintain its pressure boundary integrity, structural stability, and
functional capability during the waterhammer event. The inspector questioned the
methodology which the licensee employed in the calculation including the basis for the
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assumptions made and the basis for the analysis acceptance criteria. Inspection follow
up item 50-373/97013-01; 50-374/97013-01 was opened pending further NRC review.
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During this inspection, the inspector obtained the Office of Nuclear Reactor Regulation's
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(NRR) respon.se to Task Interface Agreement (TIA) 96-0389, " Quad Cities, Unit 1 and 2,
Regarding NEDC-32523 Applicability to RHR Water Hammer Potential," dated October
12,1997. In that response, NRR stated the following:
In accordance with General Design Criteria 35," Emergency Core Cooling,"
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licensees are required to address unavailability of either onsite or offsite power
(whichever is more limiting) concurrent with a LOCA and the consequences of
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the event. If the loss of offsite power is more limiting, the licensee is required to
consider the LOOP concurrent with a LOCA.
Since the probability of a waterhammer event increases as the amount of time
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the system is operated in the suppression pool cooling (SPC) mode increases,
and the likelihood of damage to the system increases with the frequency of
waterhammer events, operating in the SPC mode more often that assumed in
the Updated Final Safety Analysis Report (UFSAR) may be an unreviewed
safety question.
If licensee's determine that the frequency of use of the SPC mode of RHR is
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greater than that assumed in the UFSAR, then LOCA occurrence during SPC
mode should be postulated and the corresponding draindown and waterhammer
should be addressed.
Therefore, based on the discussion in TIA 96-0389, a waterhammer analysis was not
required if operation in the SPC mode of RHR was less than that assumed in the
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UFSAR. The inspector discussed this information with licensee personnel.
Subsequently, the inspector determined that although no specific amount of time spent
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in shutdown cooling was addressed or prescribed in the UFSAR, historically the time
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spent in this configuration was low which indicated that a valid waterhammer analysis
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may not be required. At the end of the inspection, the licensee was in the process of
establishing a maximum SPC operation limit, above which an acceptable waterhammer
analysis would be required.
This item will remain open pending NRC review of this established operation limit.
E8.2 (Closed) Licensee Event Reoort (LER) 50-373/97021-00: Undrainable Low Areas in the
Drywell Floor Resulting in a Degradation of the Leak Detection System (LDS) Due to
increased Delays in Detection of Unidentified Leakage.
As discussed in inspection report 50-373/97013; 50-374/97013, and LER
50-373/97021-00, the licensee determined that the ability of the drywell floor to
accumulate water was inconsistent with the UFSAR description. Specifically, the
licensee's response to UFSAR question 212.17 stated that there were no undrainable
low points in the primary or secondary containment which would result in a delay in the
detection of leakage. Contrary to this description, there were undrainable areas which
would result in the delay of the detection of leakage.
During the licensee's investigation of this problem, an additional problem related to the
reliability of instrumentation associated with portions of the LDS was identified.
Regulatory Guide 1.45 required that the sensitivity and response time for the LDS
should be adequate to identify a leakage rate of 1 gallon per minute (gpm) in less than 1
hour. To meet this requirement, a capacitance probe was used to measure
instantaneous sump level which is electronically converted to a flow rate. However,
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operating experience had demonstrated that the capacitance probe frequently drifted
and was unreliable. As a result, the recurrent failure of the electronic level indication
resulted in the LDS not meeting design basis requirements.
As part of the licensee's immediate corrective actions, the LDS was declared inoperable.
In addition, the licensee planned the following long-term actions:
Resolution of the discrepancy between the as-built configuration of the plant and
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the description contained in response to UFSAR question 212.17.
Improving the reliability of the sump level monitoring instrumentation.
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Confirming that there were no other hold up volumes in the containment which
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could result in unacceptable delays in the detection of unidentified leakage.
During this inspection period, the licensee reviewed documents associated with the as-
built configuration of the plant and identified that the floor of the Unit 1 and Unit 2
drywells were poured to conform to American Concrete Standard (ACl) 301-72,
" Specifications for Structural Concrete for Buildings," as required by S&L Standard
Specification for Concrete Work (Form 1715-Q) and were certified by quality control
inspectors on the " pour checkout cards." Table 4.3.1 of ACI 301-72 allowed up to a 3/4-
inch variation from the level or from the grades specified in the contract documents.
Therefore, although the UFSAR stated that there were no undrainable low points in the
primary or secondary containment which would result in a delay in the detection of
leakage, in fact, there was a potential that holdup volumes in the drywell floor may exist,
which would delay the detection of RCS leaksge.
The licensee reviewed this information and subsequently determined that assuming a
worst case with the floor drain 3/4 inch above all the rest of the floor and that 15 percent
of the floor was taken up with equipment mounting, then the calculated holdup volume
was about 1800 gallons.
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Technical Specification (TS) 3/4.4.3.2, " Reactor Coolant System Operational Leakage,"
required that RCS leakage shall be limited to a 2 gpm increase in unidentified leakage
over any 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period. The inspector verified that given a 2 gpm leakage rate, and
assuming a worst case holdup volume, that the leakage would be conducted to the floor
drain sump well within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and therefore the ability for detection to meet the
requirements of TS 3/4.4.3.2 would still be present.
The inspector concluded that allowable tolerances in the construction of the drywell may
cause a delay in leakage entering the floor drain sump for detection, although the
licensee indicated in the UFSAR that there were no undrainable low points in the
primary or secondary containment which would result in a delay in the detection of
leakage. However, the inspector also concluded that technical specification leakage
requirements would be still be met assuming the worst case drywell floor holdup volume.
However, it also appeared that the as-built construction of the drywell may be outside
the plant's licensing basis since undrainable low points may exist in the drywell although
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the UFSAR stated that there were no undrainable low points in the primary containment.
Resolution of the discrepancy between the as-built configuration of the plant and the
description in the UFSAR as well as improvements to sump level monitoring
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instrumentation is an unresolved item (URI 50-315/97023-01(DRS);
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50-316/97023-01(DRS)) pending NRC review of the licensee's corrective actions.
E8.3 (Closed) LER 50-373/98018-00: RHR Pump Seal Coolers Do Not Meet Design
Pressure Requirements Because Requirements Were Not included in Original Purchase
Specification to Pump Manufacturer.
As discussed in the subject LER, the licensee identified that the shell side of the Unit 1
and Unit 2 RHR pump sea! coolers did not meet design pressure requirements.
Specifically, the licensee determined that the shell side of the coolers had a design
' pressure of 75 pounds per square inch gauge (psig) although the design pressure of the
RHR Service Water (RHRSW) system that supplied the cooling water had a design
pressure of 150 psig.
The licensee performed a root cause investigation and determined that the coolers were
purchased during initial plant construction without regard to pressure requirements and
that, as a result, the coolers were procured with a shell side design pressure of 75 psig
(which was about normal system operating pressure) vice the 150 psig design pressure
requirements.
To determine the significance of this event, the licensee obtained the hydrostatic testing
results for the cooler casings and determined that although the coolers were rated at 75
psig, they were able to withstand significantly higher pressures. Specifically, in addition
to successfully hydrostatically testing each cooler's casing to twice the design pressure,
the manufacturer had also performed a hydrostatic test to destruction of an identical
cooler casing. The destruction test pressure where the casing was first noted to be
leaking was found to be 450 psig. The licensee concluded that the seal coolers would
not fall catastrophically when exposed to a pressure of 150 psig and would remain intact
and operational.
As part of the licensee's corrective actions, the affected seal coolers were replaced with
coolers rated at a design pressure of 150 psig. In addition, the licensee verified that
similar procurement problems did not exist for other coolers.
The inspector reviewed this event and verified that modifications were installed to
replace the cast iron RHR seal coolers with cast steel seal coolers which met system
design pressure requirements. In addition, the inspector reviewed the hydrostatic test
results for the coolers removed as well as the replacement coolers and had no
additional concems.
10 CFR 50, Appendix B, Criterion IV," Procurement Document Control," required that
measures shall be established to assure that the applicable regulatory requirements,
design bases, and other requirements which are necessary to assure adequate quality
are suitably included or referenced in documents for procurement. The failure to include
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in procurement documents for the Unit 1 and Unit 2 RHR seal coolers specifications
regarding shell side design pressure was an example where the requirements oi10
CFR 50, Appendix B, Criterion IV, were not met and was a violation. However,
because this violation was based upon activities prior to the events leading to the
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current extended plant shutdown and satisfy the criteria in Section Vll.B.2," Violations
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Identified During Extended Shutdowns or Work Stoppages," of the " General Statement
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of Policy and Procedures for NRC Enforcement Actions"(Enforcement Policy), NUREG-
1600, a Notice of Violation is not being issued (50-373/97023-02; 50-374/97023-02).
Specifically, the violation was licensee-identified as a result of a comprehensive program
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for problem identification and correction that was developed in response to the
shutdown, the violation would not be categorized at a severity level higher than Severity
Level ll, and the violation was not willful. In addition, actions specified in Confirmatory
Action Letter Rlll-96-0088 effectively prevent the licensee from starting up LaSalle
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County Station without implicit NRC approval.
E8.4 (Closed) LER 50-374/95006-00: 50-374/95006-01: Primary Containment Maximum
Allowable Leakage Exceeded Due to Local Leak Rate Test (LLRT) Failure.
As discussed in the subject LER, the licensee identified on March 20,1995, that the Unit
2 maximum allowable primary containment leakage rate was exceeded during th3
performance of a Local Leak Rate Test (LLRT). Specifically, the 2RE024 and 2RE025
Drywell Equipment Drain (RE) Sump containment isolation valves had been leak rate
tested and the leak rate was determined to be excessive (test volume could not be
pressurized). The cause of the leakage was determined to be seat leakage through
both valves. Upon inspection, the seat was found damaged due to foreign material.
As part of the licensee's corrective actions, both valves were repaired and successfully
leak rate tested. In addition, a plant modification was performed which installed
permanent screens in the floor drain and equipment drain sumps to prevent foreign
materialintrusion into the RE piping and isolation valves.
The licensee concluded that since the drywell RE sump would normally be filled with
water which would tend to seal any air leakage, the safety significance of the event was
minimal. In addition, the licensee concluded that in the event that air leakage eventually
occurred through the containment isolation valves, the downstream piping was normally
filled with water and provided additional isolation with normally closed automatic valves
that are designed to open with pump flow.
The inspector reviewed this event including design change packages (DCPs) 9500086
(Unit 1) and 9500087 (Unit 2) which controlled the installation of the foreign material
exclusion (FME) screens in the floor drain and equipment drain sumps. The inspector
reviewed the modification package for Unit 1. No deficiencies were identified. However,
the inspector noted the following weaknesses regarding the licensee's implementation of
the modification on Unit 2:
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Unit 2 Drawings Were Not Upoated as Appropriate
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The inspector identified that although Piping and Instrumentation Drawings
(P&lDs) associated with the floor and equipment drain system had been updated
to reflect the addition of the screens on Unit 1, similar drawings for Unit 2 had not
been updated to reflect the change.
The Unit 2 Design Change Package Was inappropriately Canceled
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Although the Unit 1 DCP (9500086) was statused as complete, the Unit 2 DCP
(9500087) was statused as canceled. Upon further review, the inspector
identified that although the work was documented in the Electronic Work Control
System (EWCS) as accomplished and had been accomplished according to the
cognizant system engineer, the DCP was canceled on March 20,1997, because
the DCP could not be located following completion of the work. As a result,
documentation associated with the work was not available for review.
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The inspector discussed this information with licensee personnel. As a result, Problem
Identification Form (PlF) L1997-07512 was initiated to document the issue. At the end
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of the inspection, the licensee planned to re-construct a new DCP to identify any
required post-modification inspections, revise drawings as appropriate, and perform a
walkdown of the system.
The inspector concluded that the design of the modification to add screens to the Unit 1
and Unit 2 floor and equipment drain sumps was good. In addition, no deficiencies were
identified in the implementation of the modification on Unit 1. However, the inspector
also concluded that the implementation of the modification on Unit 2 was poor since the
DCP was canceled when the DCP paperwork could not be located and design drawings
were not updated to reflect the installation of the modification.
The inspectors determined that the design modification to add screens to the Unit 2 floor
and equipment drain sumps was not controlled as required by 10 CFR 50, Appendix B,
Criterion 111, " Design Control," and was a violation (50-374/97023-03(DRS)).
This LER is closed.
E8.5 [Clqged) LER 50-373/96021-00: Inadequate Review of Modification of Main Control
Room Atmospheric Control System Radiation Monitoring Logic Results in an
Unreviewed Safety Question.
This event was discussed in inspection report 50-373/97003; 50-374/97003. No new
issues were revealed by the LER.
This LER is closed.
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VI. Management Meeting
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Exit Meeting Summary
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The inspector presented the results of these inspections to licensee management at an
exit meeting on March 6,1998. The licensee acknowledged the findings presented.
The inspector asked the licensee if any materials examined during the inspection should
be considered proprietary. No proprietary information was identified.
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PARTIAL LIST OF PERSONS CONTACTED
Comed
F. Dacimo
Site Vice President
G. Poletto
Site Engineering Manager
E. Connell
Design Engineering Supervisor
R. Palmieri
System Engineering Supervisor
P. Bames
Regulatory Assurance Manager
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J. Damron
System Engineering
G. Kats
System Engineering
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INSPECTION PROCEDURES USED
Engineering
Onsite Engineering
In-Office Review of Written Reports of Nonroutine Events at Power Reactor
Facilities
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Onsite Follow-Up of Written Reports of Nonroutine Events at Power Reactor
Facilities
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ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
50-373/97023-01; 50-374/97023-01 URI
Floor and Equipment Drain System Sump Level
Monitoring Problems
50-373/97023-02; 50-374/97023-02 NCV RHR Pump Seal Coolers Do Not Meet Design
Pressure Requirements
50-373/97023-03; 50-374/97023-03 VIO
Inadequate Drywell Sump Screen Modification
Closed
50-373/97021-00
LER
Undrainable Low Areas in the Drywell Floor
Resulting in a Degradation of the LDS
50-373/96018-00
LER
RHR Pump Seal Coolers Do Not Meet Design
Pressure Requirements
50-374/95006-00; 50-374/95006-01 LER
Primary Containment Maximum Allowable Leakage
Exceeded Due to LLRT Failure.
50-373/96021-00
LER
Inadequate Review of Modification of MCR
Atmospheric Control System Radiation Monitoring
Logic
Discussed
50-373/97013-01; 50-374/97013-01 IFl
Review of Information Notice 87-10
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LIST OF ACRONYMS USED
American Concrete Institute
American Society of Mechanical Engineers
CFR
Code of Federal Regulations
Design Change Package
Division of Reactor Safety
EWCS
Electronic Work Control System
gpm
gallons per minute
IFl
Inspection Follow up item
LDS
Leak Detection System
LER
Licensee Event Report
Local Leak Rate Test
Loss Of Coolant Accident
Non-Cited Violation
Office of Nuclear Reactor Regulation
Public Document Room -
P&lD
Piping and Instrumentation Drawing
Problem Identification Form
psig
pounds per square inch gauge
RE
Drywell Equipment Drain System
Residual Heat Removal Service Water
S&L
Sargent & Lundy
Suppression Pool Cooling
Task Interface Agreement
TS
Technical Specification
Updated Final Safsty Analysis Report
Unresolved item
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