IR 05000382/1999015

Insp Rept 50-382/99-15 on 990719-23 with in Ofc Insp Until 0819.No Violations Noted.Major Areas Inspected:Assessment of 10CFR50.59 Safety Evaluation Program & Review of Previous Insp Findings Involving Tornado Missile Protection

ML20212C596
Person / Time
Site:	Waterford
Issue date:	09/14/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20212C592	List: IR 05000382/1999015 ML20212C588
References
50-382-99-15, NUDOCS 9909220042
Download: ML20212C596 (14)

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ENCLOSURE U.S. NUCLEAR REGULATORY COMMISSION  ;

REGION IV I

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Docket No.: 50-382 License No.: NPF-38 Report No.: 50-382/99-15 Licensee: Entergy Operations, In Facility: Waterford Steam Electric Station, Unit 3 Location: Hwy.18 Killona, Louisiana Dates: July 19-23, with continuing in-office inspection until August 19,1999 j inspector: M. Runyan, Senior Reactor inspector Engineering and Maintenance Branch Approved By: Dr. Dale A. Powers, Chief, Engineering and Maintenance Branch Division of Reactor Safety i l

ATTACHMENT: Supplemental Information

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9909220042 990914 PDR ADOCK 05000382 O PM

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-2-1 EXECUTIVE SUMMARY Waterford Steam Electric Station, Unit 3 NRC Inspection Report No. 50-382/99-15 The purpose of the inspection was to assess the 10 CFR 50.59 safety evaluation program and to review a previous inspection finding involving tornado missile protectio Enaineerina

• The licensee's policy on increase in consequences of an analyzed accident was not consistent with 10 CFR 50.59. The licensee defined an increase in consequences to indicate that the calculated dose exceeds an applicable regulatory limit established in 10 CFR Part 100 or the General Design Criteria. However, the NRC considers an increase in consequences to constitute any increase in dose beyond that previously i

calculated and reported in the Updated Final Safety Analysis Report. Over the past 3

years, no examples existed where the licensee's calculated accident doses had been i increased (Section E2.1).  ;

• Safety evaluations were well written and explained in depth the reason for responses made to the evaluation questions. The evaluations were stand-alone documents, meaning that a reader could understand the changes and the bases for the unreviewed safety question determinations without a need to consult individuals or other document The evaluations, for the most part, reflected a high safety awareness and conservative engineering judgement (Section E2.1).

• The licensee failed to identify that a change made to the Updated Final Safety Analysis Report constituted an unreviewed safety question and, therefore, implemented the change without the approval of the Commission, in conflict with the requirements of 10 CFR 50.59. The change involved a correction to the Updated Final Safety Analysis Report to state that several nonsafety-related loads were automatically resequenced to the Class 1E bus following a loss-of-offsite power event. The previous revision stated that nonsafety-related loads were only reconnected manually under administrative I controls. This change involved an increase in the possibility of a malfunction of equipment important to safety (the Class 1E electrical bus) and, therefore, should have been identified as an unreviewed safety question. This issue was identified as an ,

unresolved item pending further NRC review to determine whether the change, if j submitted as required, would have been approved by the Commission (EA 99-220) 1 (Section E2.1). l l

• Two procedure changes were observed to have been reviewed for applicability to !

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10 CFR 50.59 with only the revised version of the procedure and change bars available l l to the licensee's reviewer. The reviewer's lack of information regarding the previous i version of the procedure was considered to be a vulnerability in the licensee's program for handling procedure changes under 10 CFR 50.59. No examples were identified where this vulnerability resulted in a faulted review (Section E2.1).

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- * The licensee's use of a previous safety evaluation to pre-screen a field change made to a newly installed modification was considered inappropriate and illustrative of a process i vulnerability. The nontrivial nature of the field change should have resulted in a documented basis for concluding that the change was fully addressed by the previous safety evaluation (EA 99-221) (Section E2.1).

• The licensee's probabilistic risk analysis of damage from tornado missiles was flawed. It failed to establish a definitive acceptance criterion for risk levels that would necessitate corrective actions and used a statistically incorrect method (failed to account for unreported tornadoes) to estimate the probability of a tornado striking the sit Conservative assumptions addressing other aspects of the calculation were observed to be sufficient to compensate for the observed discrepancies. As a result, an immediate operability concern did not exist (Section E8.1).

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\ l 4 1 Report Details Summary of Plant Status l

The plant was in Mode 1 during the inspection.

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lil. Enaineerina E2 Engineering Support of Facilities and Equipment E CFR 50.59 Proaram Inspection Scoce (37001)

The inspector reviewed procedures and training associated with the 10 CFR 50.59 safety evaluation program and reviewed safety evaluation reports and screenings performed during the past 3 years. The inspector also reviewed the last safety evaluation annual report submitted by the licensee to the NR Observations and Findinos Proaram and Trainina The inspector reviewed Procedure W2.302,"10 CFR 50.59 Review Program,"

Revision 4, dated June 30,1998. The inspector found that this procedure accurately implemented the requirements of 10 CFR 50.59, but contained a vulnerability in a definition that could result in a violation (discussed below).

The program provided a three-stage approach to 10 CFR 50.59. The first stage was termed a prescreening,in which a safety evaluation screening could be determined to be unnecessary if certain conditions were met. These conditions included, for example, that the change was editorial in nature, was controlled by another regulation, was completely addressed in a previous safety evaluation, or had been previously approved I by the NRC. The inspector considered these exclusion criteria to be appropriat The second stage was termed a screening and involved answering the safety evaluation applicability questions of 10 CFR 50.59(a)(1). The third stage was the actual performance of the safety evaluatio The inspector observed that the program did not include the concept of a " trivial" change, where modifications of a minor nature are not subject to a screening for 10 CFR 50.59 applicability because of their unsubstantial nature. In other words, physical plant modifications, no matter how minor, were subject to the same three-step process as major modifications.

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-5-The inspector reviewed Lesson Plan W-3-LP-ESPO-5059, " Performing 10 CFR 50.59 Reviews," Revision 4, and "Entergy Operations, Inc.,10 CFR 50.59 Review Program Guidelines," Revision 1. These documents provided comprehensive explanations concerning the implementation of the program. The inspector noted the licensee's definition of an increase in consequences c. an accident, which identified that an increase in calculated dose resulting from a plant change would not be positively considered an increase in consequences under 10 CFR 50.59 unless the revised dose exceeded a regulatory limit as defined in 10 CFR Part 100 or the General Design Criteria. The word " positively"is used in the previous sentence because the licensee's guidance encouraged the identification of an unreviewed safety question when in doubt, though no quantitative threshold was established in this case. The licensee's policy on increase in consequences was not consistent with the NRC position, which establishes the existing dose calculations in the Updated Final Safety Analysis Report (UFSAR) as the basis for determining if an increase in consequences had occurred. Use of the licensee's policy could result in a violation of 10 CFR 50.59. Over the past 3 years, two changes were made to the dose calculations. In each of these cases, fortuitously, the calculated doses were adjusted downward.

The program did not require refresher training on a periodic basis, but did delineate special training sessions when needed to discuss program changes. The licensee's representative stated that the 10 CFR 50.59 program was to be revised by August 1, 1999, to conform to a new Entergy Operations, Inc., corporate procedure. The licensee's representative stated that the new procedure would not substantially change the current training program.

Imolementation The inspector reviewed 13 safety evaluations performed during the past 3 years. The evaluations were well written and explained in depth the reason for responses made to the evaluation questions. The inspector noted that the evaluations were stand-alone documents, meaning that a reader could understand a change and the basis for the ,

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unreviewed safety question determination without a need to consult individuals or other documents. The evaluations, for the most part, reflected a high safety awareness and conservative engineering judgement.

Safety Evaluation 99-036, dated June 23,1999, reviewed Engineering Request ER-W3-98-0936-00-00," Evaluation of Discrepancy in UFSAR Chapter 8 Section 8.3.1.1.2.13f and Table 8.3-1." The UFSAR, Section 8.3.1.2.15(e)7, stated that nonsafety-related electrical loads were not automatically sequenced onto the Class 1E bus during a load sequence following a diesel generator start, and could only be reconnected manually under administrative control. This statement was incorrec Some nonsafety-related loads, including emergency lighting, the uninterruptible power i supply (UPS) for the plant computer, the diesel generator air compressors, and several l

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l-6-i l nonsafety power distribution panels were configured to automatically reconnect to the l Class 1E bus following a loss-of-offsite power event. Although these loads were listed in UFSAR, Table 8.3-1, as being automatically sequenced to the Class 1E bus, they

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were not identified as being nonsafety related. The automatic sequencing of nonsafety-related loads to the Class.1E bus was identified by the licensee to be not consistent with i

the intent of Regulatory Guide 1.75, " Physical Independence of E! .ctrical Systems."

l Of particular concern, the UPS for the plant computer was isolated from the Class 1E bus by only one Class 1E protective device (a circuit breaker). The other nonsafety loads had double protection. In the safety evaluation, the licensee concluded that an unreviewed safety question did not exist, principally because the UPS, though not seismically or environmentally qualified, was of inherit rugged design and that the non-Class 1E circuit breaker within the UPS provided additional circuit isolation. The evaluation stated that the information available from the plant computer for mitigating the consequences of an event outweighed the risk associated with automatically resequencing this power source. The licensee decided to make no changes to the existing design and processed a change to correct the UFSAR under 10 CFR 50.5 The inspector questioned the licensee's determination that an unreviewed safety question did not exist, specifically, in light of the 10 CFR 50.59 evaluation question asking whether the change (in this case, a de facto change based on discovery of a error in the UFSAR) increased the possibility of a malfunction of equipment important to safety. The change from the previously reviewed condition, which included the understanding that no nonsafety loads were automatically reconnected to the Class 1E bus, to the existing condition, with these loads being automatically reconnected, constituted an increase in the possibility of a malfunction of the Class 1E bus. This was because potentially faulted conditions affecting the nonsafety loads (which are more vulnerable because of being not seismically or environmentally qualified, as well as, having less rigorous controls pertaining to other attributes) could provide additional challenges to the safety-related isolation devices protecting the Class 1E bus. When these loads are manually reconnected to the bus, as was the understood practice based on the previous UFSAR text, administrative controls can be used to ensure circuit integrity. The inspector considered the safety evaluation statement that information from the plant computer outweighed the risk of it being automatically resequenced to be j

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invalid because this information could still be available after manually reconnecting the loa The inspector was also concerned that the UFSAR change, had it been submitted to the ;

Commission, may not have been approved. This consideration was based on the uniqueness of this design feature, in that few other, if any, cases were known to the inspectors where the NRC had licensed Class 1E electrical buses with automatic resequencing of nonsafety-related loads. The issue of compliance to 10 CFR 50.59 and the technicalissue of whether the change would have been approved, if submitted, was identified as an unresolved item (EA 99-220) (50-382/9915-01), pending further review by the NR The inspector reviewed two safety evaluation screenings. Both of these documents were consistent with 10 CFR 50.59 requirement m .

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l l-7-l The inspector reviewed four safety evaluation prescreenings. Two of these involved a procedure change (Procedures 01-002-000, " Annunciator, Control Room Instrumentation and Workarounds Status Control," Revision 16, and E-lGP, " internal Gas Purging," Revision 2). The inspector observed that, in each of these cases, the 10 CFR 50.59 reviewer was provided a copy of the revised procedure with revision bars in the margins. These bars indicated where the procedure was changed. However, the previous revisior to the procedure was not provided. As a result, the reviewer could have had difficulty determining the nature of the change and, consequently, how the procedure change should be handled under 10 CFR 50.59. The inspector discussed l this vulnerability with the licensee's representative who stated that it would be evaluated j for a potential program enhancemen i The inspector reviewed a prescreening of Engineering Report ER-W3-99-3546-00-04,

" Modification to SUPS 3A-S,3B-S,3B-S,3MA-S,3MB S,3MC-S & 3AB," March 14, 1999. This was a field change that was made following a post-modification test that revealed a problem with the design of the now static uninterruptible power supply (SUPS) units. The original 30 amp feeder circuit breaker was not adequately sized as revealed by field measurements showing that the bypass transformer was drawing 34.3 amps while carryirg no load. After consulting the vendor, the licensee decided to l replace the 30 amp breaker with a 70 amp breaker. The original 30 amp breaker was a '

Gould thermal magnetic circuit breaker. Since a Gould 70 amp thermal magnetic breaker was not available, the replacement was made with a 70 amp Westinghouse '

thermal magnetic circuit breaker. Other changes were also made: (1) the existing Gould contactor, Size 1, was replaced with a Gould contactor, Size 3, (2) due to the size of the existing cubicle, manipulation and rework of several cubicles were performed, and the new 70 amp circuit breaker and the Gould, Size 3, contactor were relocated to a different cubicle, and (3) the installation test was revised to perform the load rate test at a different amperage rating. The inspector noted that the licensee did not perform a new screening or safety evaluation for these changes, instead taking credit for the existing safety evaluation associated with the original design of the new SUPS unit This action was presumably based on the licensee reviewer's opinion that the changes did not affect the previous safety evaluation, though no documentation existed to explain the basis for this decisio The inspector verified that the licensee's program allowed, as a prescreening criterion, the use of a previous screening or safety evaluation to preclude the performance of a new screening or evaluation. Procedure W2.302, Section 5.2.1.3, states that, "if an existing screening or 50.59 evaluation is used (in lieu of a new screening or evaluation),

it must be reviewed to ensure that: (1) it covers the proposed change entirely and (2) subsequent changes to LBDs [ licensing basis documents) have not impacted the previous screening of the 50.59 safety evaluation. The 50.59 evaluation number must

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be referencM or a copy of the screening or 50.59 evaluation attached to the original document.' 7t e prescreening in this case referred to Safety Evaluation 99-088, which addressed the original modification that replaced the old with the new SUPS units. The inspector reviewed Safety Evaluation 99-088 and did not find any references to the feeder breaker or cubicle configuration The inspector was concerned that the disposition of this change did not meet the intent of 10 CFR 50.59. The analysis associated with the original safety evaluation may have considered aspects of the design that were not explicitly documented. When the design was changed in this case, it would appear, at the least, that a new screening should have been performed, in which the 10 CFR 50.59 applicability questions would be reconsidered, answered, and documented. A prescreening of a change to an existing modification would appear to be appropriate only for items of a truly minor nature, such j as rerouting instrument cabling because of interference, a change to a bolt specification, !

or a revision to a tag or label. When a change involves an actual change in the '

operation of the design, as occurred in this example, the licensee should formally re-enter the process to determine whether 10 CFR 50.59 applies. This may be limited to the licensee's screening process, where a determination may be made, with a documented basis, that the change does not have the potential to alter the related j information, operation, or function of a safety-related component as described in the safety analysis report. In this case, this question was, by default, answered in the prescreening, without a documented basis. Because of the nontrivial nature of the change, the inspector concluded that the use of a prescreening in this case was inappropriate and contrary to the intent of 10 CFR 50.59. However, in the absence of evidence suggesting that the above change constituted an unreviewed safety question, a specific violation of 10 CFR 50.59 did not exist. Rather, the inspector considered this example to expose a vulnerability in the licensee's 10 CFR 50.59 program (EA 99-221).

Annual Report The inspector reviewed the annual report of 10 CFR 50.59 evaluations, which was documented in Letter W3F198-0192, dated November 24,1998. This letter appeared to acceptably implement the licensee's requirement under 10 CFR 50.59 to provide an annual summary of changes made under the rul c. Conclusions Safety evaluations were well written and explained in depth the reason for responses made to the evaluation questions. The evaluations were stand-alone documents, meaning that a reader could understand the changes and the bases for the unreviewed safety question determinations without a need to consult individuals or other document The evaluations, for the most part, reflected a high safety awareness and conservative engineering judgemen l

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. The licensee's policy on increase in consequences of an analyzed accident was not consistent with 10 CFR 50.59. The licensee defined an increase in consequences to indicate that the calculated dose exceeds an applicable regulatory limit established in i 10 CFR Part 100 or the General Design Criteria. However, the NRC considers an l increase in consequences to constitute any increase in dose beyond that previously calculated and reported in the UFSAR. Over the past 3 years, no examples existed where the licensee's calculated accident doses had been increase The licensee failed to identify that a change made to the UFSAR constituted an t, , reviewed safety question, and, therefore, implemented the change without the approval of the Commission, in conflict with the requirements of 10 CFR 50.59. The i I

change involved a correction to the UFSAR to state that several nonsafety-related loads were automatically resequenced to the Class 1E bus following a loss-of-offsite power event. The previous revision stated that nonsafety-related loads were only reconnected manually under administrative controls. This change involved an increase in the possibility of a malfunction of equipment important to safety (the Class 1E electrical bus)

and, therefore, should have been identified as an unreviewed safety question. This issue was identified as an unresolved item pending further NRC review to determine whether the change,if submitted as required, would have been approved by the Commission (EA 99-220).

Two procedure changes were observed to have been reviewed for applicability to 10 CFR 50.59 with only the revised version of the procedure and change bars available to the licensee's reviewer, The reviewer's lack of information regarding the previous version of the procedure was considered to be a vulnerability in the licensee's program for handling procedure changes under 10 CFR 50.59. No examples were identified where this vulnerability resulted in a faulted revie The licensee's use of a previous safety evaluation to prescreen a field change made to a newly installed modification was considered inappropriate and illustrative of a process vulnerability. The nontrivial nature of the field change should have resulted in a documented basis for concluding that the change was fully addressed by the previous .

safety evaluation. No unreviewed safety question was identified, so no violation I

occurred (EA 99-221).

E8 Miscellaneous Engineering issues (92903)

E8.1 (Closed) Inspection Followup Item 50-382/98-201-09: emergency feedwater tornado missile protectio The NRC identified that the turbine-driven emergency feedwater pump steam supply l piping and exhaust stack were not adequately designed to withstand the effects of l tornado missiles. The licensee's UFSAR, Section 3.1.2, cites a commitment to General

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Design Criterion ll," Design Bases for Protection Against Natural Phenomena." This i

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regulation requires that the structure, system, or component important to safety is

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l designed to withstand the effects of natural phenomena. This includes tornado wind and/or missile loads. The exposed sections of the emergency feedwater system are not housed in a protective structure and are not designed to be capable of withstanding certain missile impacts. Therefore, this design configuration does not meet the UFSAR I commitment to General Design Criterion 1 )

Following discussions with the NRC, the licensee agreed that an unreviewed safety question existed and intended to submit to the NRC for approval a 10 CFR 50.90 request to change the design and licensing basis for the emergency feedwater exposed system components. Specifically, the licensee will request that the licensing basis be amended to allow use of a probabilistic analysis consistent with Regulatory Guide 1.117

" Tornado Design Classification," showing that the probability of a missile striking the emergency feedwater exhaust stack and supply piping is sufficiently low to be considered negligible. This request was scheduled to be submitted to the NRC by October 30,1999. The licensee submitied Licensee Event Report 50-382/99-007 on July 23,1999, to formally report this issue to the NR The licensee's probabilistic analysis was described in a letter to the NRC dated June 4, 1997. This analysis concluded that the probability of a tornado missile striking the turbine-driven emergency feedwater pump exhaust stack was 3.96E-12 occurrences per year and the probability of a strike on the emergency feedwater steam supply piping was 3.00E-11. The cumulative probability for all safety-related targets, which included other safety-related systems, was 1.64E-7. The licensee was using this analysis to establish interim operability for the turbine-driven emergency feedwater system, as well as, other unprotected safety-related targets pending NRC approval of the revision to the design and licensing basis, it was the licensee's intent to avoid additional construction ;

of tornado missile barrier The inspector reviewed the probabilistic analysis described above and noted several discrepancies. The analysis did not establish a definitive acceptance criterion for l determining whether potential targets should be protected. The analysis referred to the reference value of 1.0E-7 contained in Regulatory Guide 1.117, but considered it as only a guideline for comparison purposes. The total strike probability of all targets,1.64E-7, was in excess of this figure but, according to the licensee's representative, was not noted as failing the criterion because it was on the same order of magnitude. The inspector noted that the NRC has approved a limit of 1E-6 for similar reviews of other plants, with this limit applying to the cumulative damage probability of all exposed safety-related equipmen To determine the probability of a tornado striking the site, the analysis used a meteorological study of a 4 square degree (latitude and longitude) area centered on the site for the period of 1950 to 1986. This study was based on reported tornadoes in this area. The inspector noted that the lack of habitation in much of this area during this time could have resulted in some tornadoes not being reported. Therefore, the calculated frequency of tornado strikes at the plant,4.6E-5 occurrences per year, or approximately once every 21,700 years, might be less than actua r-

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-11-The inspector noted that the calculation also included an analysis based on the Electric Power Research Institute (EPRI) Report NP-2005, " Tornado Missile Simulation Methodology - Computer Code Manual," August 1981, which, based on the tornado region (Region l) that includes the Waterford site, resulted in a probability of 47.544E-5, or about 10 times the value of the more localized study. This value was referenced in the calculation, but was adjusted by a conversion factor of 0.1 to normalize it to the local analysis. The inspector considered the use of the conversion factor to be not justified because the EPRI analysis included a correction factor to account for unreported tornadoes while, as noted above, the local analysis was uncorrected. If the EPRI probability had been used without the conversion factor, the inspector determined that the total missile strike probability of all safety-related targets would have been approximately 1.7E-6. This figure is significantly greater than the value contained in Regulatory Guide 1.117, which is 1E-7, and is also in excess of previous NRC p1sitions of acceptable risk (1E-6). If this figure had been used in the licensee's letter of June 4, 1997, it would have weakened the licensee's basis for not erecting additional missile protectio However, the inspector was aware of other conservatisms in the calculation that could be removed to lower the probability, including the licensee's assumption that some of I the protected targets were unprotected and, in the sizing of piping targets, the use of the internal surface area of piping in lieu of the exposed projected area. The inspector believed that once these conservatisms were removed from the calculation, the final probability would be less than 1E-6, even if the EPRl-generated tornado frequency was I used. On this basis, the licensee's interim operability deterrnination was considered acceptable. The licensee's representative stated that the calculation would be revised to account for the findings of this inspection. The inspector understood that the NRC Office of Nuclear Reactor Regulation, in response to the 10 CFR 50.90 submittal, will evaluate the long-term acceptability of the current configuration. Any enforcement aspects to this matter will be addressed following the 10 CFR 50.90 review by the Office of Nuclear Reactor Regulation and in conjunction with the closure of Licensee Event Report 50-382/99-00 l The inspector observed that the tornado missile probabilistic analysis was flawed, in l that, it failed to establish a definitive acceptance criterion and used a statistically incorrect method of estimating the risk of a tornado striking the sit V. Manaaement Meetinas X1 Exit Meeting Summary An exit meeting was conducted on July 23,1999. Following additional in-office review, a supplemental telephonic exit was conducted on September 14,1999. The licensee's management acknowledged the findings of the inspection. The licensee's management stated that no proprietary information had been reviewed during the inspectio l-l

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ATTACHMENT l SUPPLEMENTAL INFORMATION PARTIAL LIST OF PERSONS CONTACTED Licensee H. Brodt. Senior Lead Engineer C. DeDeaux, Licensing Supervisor R. Douet, Maintenance Manager E. Fields, Senior Lead Engineer T. Fleischer, Senior Lead Engineer J. Halman, Design Engineer E. Lemke,10 CFR 50.59 Coordinator T, Leonard, General Manager, Plant Operations S. Matharu, Engineering Supervisor S. Munshi, Senior Staff Engineer E. Perkins, Acting Director, Nuclear Safety and Regulatory Affairs A. Wrape, Director, Engineering NRC T. Farnholtz, Senior Resident inspector INSPECTION PROCEDURES USED I

37001 10 CFR 50.59 Safety Evaluation Program 92903 Followup-Engineering ITEMS OPENED AND CLOSED Opened 50-382/9915-01 URI Failure to Identify an Unreviewed Safety Question (Section E2.1)

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50-382/98201-09 IFl EFW Tornado Missile Protection (Section E8.1) l l

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i DOCUMENTS REVIEWED Procedures W2.302,"10 CFR 50.59 Review Program," Revision 4 Miscellaneous i

Lesson Plan W-3-LP-ESPO-5059, " Performing 10 CFR 50.59 Reviews," Revision 4 Report of Facility Changes, Tests, and Experiments, W3F1-98-0192, November 24,1998 Entergy Operations, Inc.,10 CFR 50.59 Review Program Guidelines, Revision Safety Evaluations99-034, DCP-3521, Revision 1, " Route DCP Sumps Discharge to Circulating Water System,"

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May 19,1999 99-037, PC-8020," Broad Range Gas Monitor-Control Room Vent Supply," June 10,1999 99-036, ER-W3-98-0936-00-00," Evaluation of Discrepancy in FSAR Chapter 8 Section 8.3.1.1.2.13f and Table 8.3-1," June 23,1999 99-040, ER-W3-1178-00-00," Update FSAR to include LPSI Throttle Valve Failure Effects,"

June 24,1999 j 99-017, ER-W3-98-1220-00-00,"Limitorque Technical Update 98-01 Modifications,"

February 23,1999 98-110, ER-W3-98-1263-00," Changes to DC-3533, Provide Containment Isolation for PASS from SIS Sump," December 21,1998 99-005, ER-W3-98-0934-00-00," Revise DBD-20 and DBD-6 to Clarify MFIV and MSIV Actuator Heaters," January 28,1999 96-117, Temporary Alteration Request TAR 96-012. " Gag Closed CC-807A and CC-823A,"

November 5,1996 98-005, LDCR 98-0041, " Revision to LOCA Radiological Dose Consequences Documented in FSAR 15.6.3.3.5," January 22,1998 97-175, Work Authorization 01161072," Resetting of Relief Valves ACC 121 A(B) on Shell Side of CCW Heat Exchangers," July 31,1997 i

98-096, Calculation EC-S96-011,"LOCA Offsite and Control Room Radiological Dose l . Consequences," Revision 1, November 5,1998 l l I

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-3-97-148, LDCR 97-0234, " Revision to ECCS Performance Minimum Containment Pressure Analysis," June 13,1997 l 97-159, Calculation EC-E90-006, " Emergency Diesel Generator Loading and Fuel Oil l

Consumption," Revision 2, July 4,1997 10 CFR 50.59 Review Screenina Forms ER-W3 97-0325-00-00, " Evaluate the Feasibility for Replacing the Existing Seals on the Fuel

Pool Pumps with a Different Seal Type," July 28,1997 l

MM-006-053," Check Valve Inspection (Swing)," Revision 2, December 16,1997 10 CFR 50.59 Figview Pre-Screenina Forms ER-W3-99-0431-00-00, " Excessive Pen Movement on Control Room Recorders," April 27, 1999 O1-002-000, " Annunciator, Control Room Instrumentation and Workarounds Status Control,"

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E-lGP, " Internal Gas Purging," Revision 2, May 12,1998 l ER-W3-99 3546-00-04," Modification to SUPS 3A-S,38-S,38-S,3MA-S,3MB-S,3MC-S &

l 3AB," March 14,1999

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