IR 05000373/1998026
| ML20198B232 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 12/18/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20198B225 | List: |
| References | |
| 50-373-98-26, 50-374-98-26, NUDOCS 9812180112 | |
| Download: ML20198B232 (15) | |
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U.S. NUCLEAR REGULATORY COMMISSION
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REGIO.Nlil Docket Nos:
50-373;50-374
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License Nos:
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Report Nos:
50-373/98026(DRS); 50-374/98026(DRS)
l Licensee:
Commonwealth Edison Company Facility:
LaSalle Nuclear Generating Station, Units 1 and 2
- Location:
2605 N. 21st Road Marseilles, Illinois 51341-9756 Dates:
October 26 through November 6,1998 Inspector:
E. Duncan, Reactor Engineer Approved by:
John M. Jacobson, Chief Lead Engineers Branch Division of Reactor Safety i
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9812180112 981118 PDR ADOCK 05000373 G
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l EXECUTIVE SUMMARY l
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LaSalle Nuclear Generating Station, Units 1 and 2 l
NRC Inspection Reports 50-373/98026; 50-374/98026 The purpose of this inspection was to review the effectiveness of completed corrective actions to address a number of previously identified violations and inspection followup items. In addition, an assessment of the actions planned or completed to address a number of events discussed in licensee event reports was also performed.
Enoineerina Design change packages to install relief valves for containment penetration piping
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susceptible to overpressurization did not identify appropriate controls and interlocks for intervening isolation valves. (Section E8.8)
Corrective actions to address the potential failure of ductwork in the primary containment
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vent and purge system during a loss-of-coolant-accident were appropriate. Weaknesses regarding the use of engineering judgement for final disposition in the stress evaluation of a piping support, and the exemption of a safety switch from the environmental qualification program were identified. (Section E8.10)
Corrective actions to address unanalyzed steam lines used for alternate decay heat
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cooling were appropriate. However, structural support loading calculations inappropriately incorporated the use of engineering judgement for final disposition. A Non-Cited Violation was identified. (Section E8,15)
A fire detection circuit module in the Unit 1 and Unit 2 standby gas treatment system was
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not included in the environmental qualification program as required. A Non-Cited Violation was identified. (Section E8.18)
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Report Details Ill. Engineering E8 Miscellaneous Engineering issues E8.1 (Closed) Violation 50-373/97008-01: 50-374/97008-01: Overtime Approval Requirements Not Met.
The inspector verified the corrective actions described in inspection report 50-373/97008; 50-374/97008 to be reasonable and complete. No similar problems were identified.
E8.2 (Closed) Violation 50-373/97023-03: 50-374/97023-03: Inadequate Drywell Sump Screen Modification.
The inspector verified the corrective actions described in the licensee's response letter, dated May 8,1998, to be reasonable and complete. No similar problems were identified.
E8.3 (Closed) Violation 50-373/98005-03: 50-374/98005-03: Fire Loading Calculation Errors.
The inspector verified the corrective actions described in the licensee's response letter, dated July 3,1998, to be reasonable and complete. No similar problems were identified.
E8.4 (Closed) Violation 50-373/98005-04: 50-374/98005-04: Diesel Generator Airbox Drain Unauthorized Temporary Alteration.
The inspector verified the corrective actions described in the licensee's response letter, dated July 3,1998, to be reasonable and complete. No similar problems were identified.
E8.5 (Closed) Violation 50-373/98005-08: 50-374/98005-0_8: Failure to Provide Annual 10 CFR 50.59 Report.
The inspector verified the corrective actions described in the licensee's response letter,
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dated July 3,1998, to be reasonable and complete. No similar problems were identified.
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E8.6 (Closed) Violation 50-373/98005-15: 50-374/98005-15: Relay Replacement Error.
The inspector verified the corrective actions described in the licensee's response letter, dated July 3,1998, to be reasonable and complete. No similar problems were identified.
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E8.7 (Closed) Violation 50-373/98015-02: 50-374/98015-02: Fire Protection System Valve Testing.
l The inspector verified the corrective actions described in the licensee's response letter, dated August 21,1998, to be reasonable and complete. No similar problems were identified.
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E8.8 (Closed) Inspection Followuo item (IFI) 50-373/97018-01: 50-374/97018-01: Use of Isolation Valves in Relief Discharge Path.
Design change packages to install relief valves for containment penetration piping susceptible to overpressurization contained isolation valves in the relief valve flow path.
The inspector questioned whether locking open the isolation valves with administrative controls over the locks met the requirements of American Society of Mechanical Engineers (ASME) Section lil, Article NC-7000," Protection Against Overpressure,"
Paragraph NC-7153, " Provisions When Stop Valves Are Used," regarding " controls and interlocks."
During this inspection, the inspector reviewed this issue with the Office of Nuclear Reactor Regulation and determined that the " controls and interlocks" planned by the licensee were not sufficient to meet the Code. However, the inspector concluded that since the licensee's planned actions would have provided some measure of protection against inadvertent isolation, the event was of minor significance.
10 CFR 50, Appendix B, Criterion lil, " Design Control," requires that measures shall be established to assure that the regulatory requirements and the design basis for structures, system and components are correctly translated into specifications and drawings. The failure to ensure that the requirements of Article NC-7000 of ASME Section ill were translated into modifications to address piping overpressurization concerns is an example where this requirement was not met and was a violation.
However, this failure constitutes a violation of minor significance and is not subject to formal enforcement action.
E8.9 (Closed) IFl 50-373/98005-05: 50-374/98005-05: Battery Intercell and Interack Resistances.
The inspector questioned wheth 3r the battery terminal connection resistance limits in Technical Specification 3.8.2.3, ' Direct Current Distribution - Operating," were appropriate.
During this inspection, the inspector discussed this issue with battery specialists from the Office of Nuclear Reactor Regulation and determined that these limits were similar to those approved for other utilities and were acceptable.
E8.10 (Closed) LER 50-373/97005-04: Potential Loss of Both Trains of Standby Gas Treatment (SBGT) System.
The NRC reviewed and closed LER 50-373/97005-00/01/02/03 in inspection report 50-373/98005; 50-374/98005. Supplement 4 to this LER was provided to document the progress of planned corrective actions, and to document additional planned corrective actions. In particular, the supplement identified that primary containment vent and i
purge system ductwork would failin the event of a loss-of-coolant-accident during l
inerting or de-inerting which could potentially impact safety-related equipment. To address this issue, the licensee created a new harsh environmental qualification (EQ)
zone and evaluated equipment located in the zone in addition, a modification to remove two piping supports to address thermal growth concerns was also completed.
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l During this inspection, the inspector reviewed safety evaluation L98-177 which addressed the modification and a number of calculations which addressed the equipment potentially affected in the new EQ zone. The following deficiencies were
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identified:
Support M09-VQ02-1038X Evaluation Deficiency
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To address piping stresses due to the failure of the non-seismic portion of the primary containment vent and purge system, piping supports were evaluated to i
ensure that anticipated stresses were within American Society of Mechanical l
Engineers (ASME) Code limits. Although the vast majority of the supports analyzed had a significant margin between the allowable and anticipated load, support M09-VQ-02-1038X was an exception. The new calculated service level l
D load was 12,261 pounds. Also, the allowable service level D load was 12,229 pounds. Therefore, the anticipated load exceeded the allowable load.
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However, the licensee concluded by engineering judgement that since 12,229 pounds - 12,261 pounds, the support assembly was acceptable and the allowable load was not exceeded.
The inspector discussed this issue with licensee management and determined that the use of engineering judgement to conclude that the support assembly was acceptable instead of the completion of additional calculations did not meet management expectations. As a result, Problem Identification Form L1998-07159 was generated to identify this issue for entry into the corrective actions program. Subsequently, the licensee determined that adequate margin existed and the calculation would be revised accordingly.
Switch Evaluation Weakness
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Components located in the new harsh EQ zone were evaluated to determine whether they were required to be included in the EQ program. The inspector reviewed calculation L-001239 and identified that documentation justifying the exemption of Square D Safety Switch Model DU-221-RB was incomplete.
According to the manufacturer, the safety switch hcd a maximum safe operating threshold of 145*F. However, the maximum temperature the switch was expected to be subjected to was 160*F. The calculation then documented that the materials of construction were " typically the same" as those used for terminal blocks and fuse blocks which were capable of withstanding temperatures in excess of 200*F. The inspector questioned the validity of this evaluation in light of the manufacturer's certification and the lack of documentation to establish the similarity of materials. Subsequently, the licensee produced documentation from the manufacturer which stated that the switch was composed of a phenolic material which was neither temperature-sensitive nor age-sensitive. The
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inspector concluded that although the exemption of the switch from the program was acceptable, the calculation lacked adequate documentation to justify the conclusion.
The inspector concluded that overall, corrective actions to address the failure of ductwork in the primary containment vent and purge system during a loss-of-coolant-
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accident were appropriate. Weaknesses regarding the use of engineering judgement in the stress evaluation of a piping support, and the exemption of a safety switch from the EQ program were identified.
E8.11 (Closed) LER 50-373/97012-01: Low-Low Setpoint Function of Main Steam Safety Relief Valves Not Tested.
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The NRC reviewed and closed this LER as documented in inspection report 50-373/98015; 50-374/98015. Supplement 1 to this LER was provided to document the completion of Unit 1 testing. No new issues were revealed by this LER supplement.
E8.12 (Ocen) LER 50-373/97015-00: Diversion of Low Pressure Coolant injection Flow to l
Suppression Pool.
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As discussed in the subject LER and inspection report 50-373/98005; 50-374/98005, the l
licensee identified that if the residual heat removal test return valve was in the process of being manually opened, and a simultaneous emergency core cooling system (ECCS)
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actuation occurred, the manual overide logic would be activated and the valve would not automatically re-position to inject into the vessel.
Section 6.3.2.8, " Manual Actions," of the Updated Final Safety Analysis Report stated that with regard to the ECCS sys'em, "the initiation of the ECCS is completely automatic. No operator action is assumed for at least 10 minutes after initiation."
However, with the valve in the manually throttled position, and barring manual action to l
re-position the valve closed, low pressure coolant injection flow would be diverted to the suppression pool, bypassing the reactor vessel. The diversion of low pressure coolant injection flow to the suppression pool (due to an ECCS actuation while opening the test return valve) in combination with a single active failure could result in a condition with
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fewer available systems than assumed in the accident analysis.
Based on the low probability of the condition occurring, the licensee considered the event to be outside their design basis. The NRC questioned whether the use of probability to justify exclusion of this event from the design basis was appropriate. This question was forwarded to the Office of Nuclear Reactor Regulation (NRR) for technical review.
During this inspection, the inspector received NRR's response to the technical issue, in that response, NRR concluded that there was no probability threshold that a licensee can apply to exclude events from its existing design or licensing basis for two reasons.
First, regulations require the consideration of various types of possible accidents which may impose a risk to the public, including those that would not be expected during a plant's operationallife. Second, the Commission's guidance on the use of probabilistic risk assessment admonishes against making decisions based solely on risk.
This LER will remain open pending licensee formulation of corrective actions.
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8.13 (Closed) LER 50-373/97020-01: Recirculation Flow Converter Calibration and Functional Testing Deficiencies.
The NRC reviewed and closed this LER as documented in inspection report 50-373/98015; 50-374/98015. Supplement 1 to this LER was provided to update the progress of planned corrective actions. No new issues were revealed by this LER supplement.
E8.14 (Closed) LER 50-373/97030-01: Containment Integrated Leak Rate Test Error.
As discussed in LER 50-373/97030-00, the Unit 1 and Unit 2 containment integrated leak rate tests were performed with containment liner weld channel vent plugs installed.
The NRC reviewed and closed this LER as documented in inspection report 50-373/98005;50-374/98005. Gupplement 1 to this LER was issued to identify that 18 channels for each unit could not be demonstrated to maintain their integrity with the plugs installed during a design basis accident due to their location with respect to the suppression pool. As a result, the Unit 1 channel vent plugs on each of the 18 channels were permanently removed and a local leak rate test was successfully performed to demonstrate the integrity of the underlying containment liner welds. A similar action was planned prior to Unit 2 startup.
During this inspection, the inspector determined that by letter dated May 4,1998, the Office of Nuclear Reactor Regulation documented their review and approval of these actions.
E8.15 (Closed) LER 50-373/97040-00: Unanalyzed Condition for Alternate Shutdown Cooling Steam Lines.
On October 24,1997, no piping analysis could be located which evaluated the use of reactor core isolation cooling system piping and main steam lines as an alternate shutdown cooling method as directed in LaSalle Operating Procedure LOP-RH-17,
" Alternate Shutdown Cooling." In particular, the lines had not been analyzed as filled with water and addressing seismic concerns.
As part of the licensee's immediate corrective actions, the required analyses were performed for Unit 1 and demonstrated that the piping was adequately supported and within the ASME Code allowables. The licensee planned to perform a similar evaluation prior to Unit 2 startup.
During this inspection, the inspector reviewed calculation L-001496, " Evaluation of Piping Subsystem 1MS02 for Procedure LOP-RH-17 During Safety Relief Valve Removal," Revision 4, dated November 20,1997 and calculation L-001497, " Evaluation of Piping Subsystem 1R101 for Procedure LOP-RH-17," Revision 4, dated November 20,1997. Both calculations concluded that the affected piping subsystems would be
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able to withstand a seismic event with the steam lines filled with water for alternate
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10 CFR 50, Appendix B, Criterion Ill, " Design Control," requires that measures shall be established to assure that the regulatory requirements and the design basis for
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structures, systems, and components are correctly translated into specifications and drawings. The failure to have adequate calculations to support alternate decay heat removal methods is an example where the requirements of 10 CFR 50, Appendix B, l
Criterion lll were not met and was a violation. However, this non-repetitive, licensee-identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy (NCV 50-373/98026-01; 50-374/98026-01).
During the review of calculation L-001497, the inspector noted that the anticipated load associated with spring hangar Rl41-1095V was calculated to be 2262 pounds which exceeded the load at which the hangar was expected to bottom out which was 2210 pounds. No followup calculations were performed to further evaluate this issue, j
although the calculation stated that by engineering judgement, if hangar stiffness was taken into account, the calculated load would decrease.
The inspector discussed this issue with licensee management and determined that the use of engineering judgement to conclude that the hangar loading was acceptable instead of the completion of additional calculations did not meet management expectations. As a result, Problem identification Form L1998-07158 was generated to identify this issue for entry into the corrective actions program. Subsequently, the licensee determined that although additional justification was not documented, adequate i
margin existed and the calculation would be revised accordingly.
The inspector concluded that overall, the corrective actions to address unanalyzed steam lines used for alternate decay heat cooling were acceptable. However, structural support loading calculations incorporated engineering judgement in a manner which did not meet management expectations.
E8.16 (Closed) LER 50-373/97044-01: Potentially Unanalyzed Condition for Automatic Depressurization System Accumulator Capacity.
This event was discussed in inspection report 50-373/98012; 50-374/98012. No new issues were revealed by this LER.
E8.17 (Closed) LER 50-373/97046-02: Potential Pressurization of Turbine Building Ventilation Exhaust Tunnel.
The NRC reviewed and closed LER 50-373/97046-00/01 in inspection report 50-373/98015;50-374/98015. Supplement 2 to this LER was provided to update the progress of planned corrective actions. No new issues were revealed by this LER supplement.
E8.18 (Closed) LER 50-373/98006-00: Local Fire Detection Module in Standby Gas Treatment System (SBGT) Located in Harsh Environment.
On June 22,1998, engineering personnel discovered that a local fire detection circuit
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module in the Unit 1 and Unit 2 SBGT system was not included in the EQ program as
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required. The SBGT local fire detection module was not safety-related and was not l
required to operate during or following an accident. However, it was located in a harsh l
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environment and a short to ground could have impaired the functional capability of safety-related SBGT flow control circuits.
The licensee conducted a root cause investigation and determined that during the development of the LaSalle EQ program, the potentialimpact of this nonsafety-related component on safety-related equipment was not adequately addressed. In addition, the licensee determined that although the fire detection module was not environmentally qualified, the component was subjected to environmental qualification testing and that the test results enveloped all of the parameters required for environmental qualification except radiation dose. To correct this problem, the installation of a separate power feed for the SBGT system local fire detection circuit was planned. In addition, an
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assessment of the SBGT and hydrogen recombiner systems verified that the error was l
isolated.
During this inspection, the inspector reviewed "LaSalle County Station Unit 1 and 2 Assessment of Safety-Related Components Located in Harsh EQ Zones for LER 98006," dated July 21,1998. Although no additional EQ-related issues were identified, a number of drawing errors, EQ binder omissions, and other documentation deficiencies were discussed in the report which indicated that the review was thorough and comprehensive. The inspector also verified that design change package 9800161 installed a separate power feed for the local SBGT fire detection circuit. No deficiencies were identified.
10 CFR 50.49(a), " Environmental Qualification of Electrical Equipment important to Safety for "Jelear Power Plants," requires that licensee's establish a program for qualifying electrical equipment important to safety, including safety-related and nonsafety-related electric equipment relied upon to remain functional during and following design basis events. The failure to include the Unit 1 and Unit 2 SBGT local fire detection modules in the EQ program was an example where the requirements of 10 CFR 50.49(a) were not met and was a violation. However, this non-repetitive, licensee-identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy (NCV 50-373/98026-02; 50-374/98026-02).
E8.19 (Closed) LER 50-373/98008-00: Fire Protection Valves Not Tested Per Technical Specifications.
This event was discussed in inspection report 50-373/98015; 50-374/98015. No new issues were revealed by this LER.
V. Management Meetings X1 Exit Meeting Summary The inspector presented the inspection results to members of licensee management at the conclusion of the inspection on November 6,1998. The licensee acknowledged the findings l
presented. The inspector asked the licensee whether any materials examined during the l
inspection should be considered proprietary. No proprietary information was identified.
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PARTIAL LIST OF PERSONS CONTACTED Commonwealth Edison J. Amburgey, Design Engineer P. Barnes, Regulatory Assurance Manager L
G. Campbell, Site Engineering Manager E. Connell, Design Engineering Supervisor i
L R. Garber, System Engineer i
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J. Pollock, Support Engineering Supervisor l
lNSPECTION PROCEDURES USED i
. Followup - Engineering l
I? 92700 Onsite Followup of Written Reports of Nonroutine Events at Reactor Facilities
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ITEMS OPENED, CLOSED AND DISCUSSED i
Opened
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50-373/98026-01;50-374/98026-01. NCV Alternate Shutdown Cooling Steam Lines 50-373/98026-02;50-374/98026-02 NCV Fire Detection Module in Harsh Environment
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Closed E
50-373/98026-01:50-374/98026-01 NCV Alternate Shutdown Cooling Steam Lines i
50-373/98026-02;50-374/98026-02 NCV Fire Detection Module in Harsh Environment L
50-373/97008-01:50-374/97008-01 VIO - Overtime Approval Requirements Not Met.
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50-373/97023-03;50-374/97023-03 VIO Inadequate Drywell Sump Screen Modification 50-373/98005-03:50-374/98005-03 VIO Fire Loading Calculation Errors.
50-373/98005-04;50-374/99005-04 VIO Diesel Generator Air Box Drain Unauthorized t
Temporary Alteration 50-373/98005-08;50-374/98005-08 VIO Failure to Provide Annual 10 CFR 50.59 Report 50-373/98005-15;50-374/98005-15 VIO Relay Replacement Error 50-373/98015-02:50-374/98015-02 VIO Fire Protection System Valve Testing i
e 50-373/97018-01;50-374/97018-01 IFl Use of Isolation Valves in Relief Discharge Path l
50-373/98005-05;50-374/98005-05 IFl Battery Intercell and Interack Resistances 50-373/97005-04 LER Potential Loss of Both Trains of SBGT System 50-373/97012-01 LER Low-Low Setpoint Function of Main Steam Safety Relief Valves Not Tested 50-373/97020-01 LER Flow Converter Calibration / Testing Deficiencier,
.50-373/97030-01 LER Containment Integrated Leak Rate Test Error
~ 50-373/97040-00 LER Alternate Shutdown Cooling Steam Lines 50-373/97044-01 LER Potentially Unanalyzed Automatic Depressurization
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System Accumulator Capacity
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_.50-373/97046-02 LER Overpressurization of Turbine Bui! ding Exhaust l
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50-373/98006-00 LER Fire Detection Module in Harsh Environment
50-373/98008-00 LER Fire Protection Valves Not Tested Per Technical
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Discussed 50-373/97015-00 LER Diversion of Low Pressure Coolant injection Flow
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to Suppression Pool
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American Society of Mechanical Engineers CFR Code of Federal Regulations ECCS Emergency Core Cooling System EQ Environmental Qualification IFl inspection Followup Item LER Licensee Event Report
.NCV Non-Cited Violation-NRR Office of Nuclear Reactor Regulation SBGT Standby Gas Treatment VIO Violation
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l LIST OF DOCUMENTS REVIEWED
The following is a list of documents reviewed during the inspection, including documents
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prepared by others for the licensee, inclusion on this list does not imply that the NRC inspector reviewed the documents in their entirety, but rather that selected sections or portions of the
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documents were evaluated as part of the overallinspection effort, inclusion of a document in p
this list does not imply NRC acceptance of the document, unless specifically stated in the inspection report.
Procedures
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LAP-1200-24, "Roadmap for Safety Evaluation," Revision 1, dated April 17,1998.
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LES-NB-101 A, " Unit 1 Division 1 Automatic Depressurization System Relay and Low-Low Set Function Logic Test," Revision 5, dated July 18,1997.
LES-NB-101B, " Unit 1 Division 2 Automatic Depressurization System Relay and Low-Low Set Function Logic Test," Revision 4, dated July 18,1997.
LIS-RR-101 A, " Unit 1 Recirculation Flow Converter A Calibration," Revision 0, dated May 28,
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LIS-RR-101B, " Unit 1 Recirculation Flow Converter B Calibration," Revision 0, dated June 3,
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LIS-RR-101C, " Unit 1 Recirculation Flow Converter C Calibration," Revision 0, dated June 8, 1998.
'LIS-RR-101D, " Unit 1 Recirculation Flow Converter D Calibration," Revision 0, dated June 8, 1998.
LTS-600-3,." Primary Containment inspection,". Revision 7, dated April 21,1998.
Comoleted Surveillances LOP-FP-A7, Attachments A-H," Fire Protection Filter Unit Deluge Flow Path Valve Cycling
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- Test," completed July 31,1998.
Desian Chanae Packaaes
.9800161 Isolate 1VG002 Damper Control Logic From SBGT Fire Protection Logic Calculations L-001496 Evaluation of Piping Subsystem 1MSO2 for Procedure LOP-RH-17 During Safety
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Relief Valve Removal.
L-001497 Evaluation of Piping Subsystem 1R101 for Procedure LOP-RH-17.
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. Evaluation of GE HMA111 Relays, l
L-001861 Environmental Qualification of Equipment Added to LaSalle EQ Program per the Component Operability Assessment.
L-062513 Stress Report Primary Containment Vent and Purge System Piping.
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10 CFR 50.59'Safetv Evaluations
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L98-144 Diesel Generator Air Box Drains L98-177 -
Primary Containment Vent and Purge System Overpressurization During a Loss-l Of-Coolant-Accident Drawina Chanae Reauests -
980304 Diesel Generator Air Box Drains Corrective Action Records'
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01 96-077 Overtime
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Licensee Event Reoorts 50-373/97005-04 Potential Loss of Both Trains of Standby Gas Treatment System
- 50-373/97012-01 Low-Low Setpoint Function of Main Steam Relief Valves Not Tested 50-373/97015-00 Diversion of Low Pressure Coolant injection Flow to Suppression Pool 50-373/97020-01 Recirculation Flow Converter Calibration and Testing Deficiencies
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50-373/97030-01 Containment Integrated Leak Rate Test Error
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50-373/97040-00-Unanalyzed Condition for Alternate Shutdown Cooling Steam Lines
- 50-373/97044-01 Unanalyzed Automatic Depressurization System Accumulator Capacity 50-373/97046-02 Overpressurization of Turbine Building Exhaust 50-373/98006-00 Fire Detection Module in Harsh Environment
'50 373/98008-00 Fire Protection Valves Not Tested Per Technical Specifications
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Miscellaneous August Overtime Deviation Report-September Overtime Deviation Report Offsite Review Report 1-98-168, "98-383 Notice of Violation Response," dated July 9,1998
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Memorandum dated June 9,1998 from H. Pontius to E. Connel lil, " Review of Design Change Packages to Enture Cancellations Were Appropriate"
' _ LaSalle County Station Unit 1 and 2 Assessment of Safety-Related Components Located in
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Harsh EQ Zones for LER 98006," dated July 21,1998 l
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Distribution:
F WU C D X.U;,..~ i;L....
uo u6 SAR (E-Mail).
RPC (E-Mail)
Project Mgr., NRR w/enci J. Caldwell, Rlli w/enci '
C. Pederson, Rlli w/enci B. Clayton, Rlli w/enci SRI LaSalle w/ encl DRP w/enct TSS w/ encl
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DRS (2) w/enci Rlli PRR w/ encl..,.
,m RUBLIC IE-01.wlenci -
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" Docket Files /enci GREENS IEO (E-Mail)
DOCDESK (E-Mail)
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