IR 05000413/1997010
| ML20198H891 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 09/11/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20198H890 | List: |
| References | |
| 50-413-97-10, 50-414-97-10, NUDOCS 9709220216 | |
| Download: ML20198H891 (20) | |
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U. S. NUCLEAR REGULATORY COMMISSION iiEGION 11 Docket No.:
50-413, 414 License No.:
NPF-35 and 52
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Report No.:
50-413/97-10, 50414/97-10
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Licensee:
Duke Power Company I
Facility:
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Catawba Nuclear Station, Units 1 & 2 l
Location:
York, South Carolina
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Dates:
June 23 - 27 and July 21 - 31,1997 Team Leader:
P. Kellogg, Senior Project Manager Engineering Branch Division of Reactor Safety inspectors.
L. Moore, Reactor Inspector D. Starkey, Resident inspector, Sequoyah K. Mortensen, l&C Branch, NRR Approved By:
Harold O. Christensen, Chief Engineering Branch Division of Reactor Safety l
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9/09220216 970911 PDR ADOCK 05000413
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EXECUTIVE SUMMARY Catawba Nuclear Station
- NRC Inspection Report 50-413/97-10, 50-414/97-10
- This inspection included a review of the licensee's calculations, analysis, and other engineering documents that were used to support the Residual Heat Removal (NO) system performance during normal and accident or abnormal conditions. The report covered a three-week period of inspection.
Overall the inspection found the system to be operational. The calculations to support system performance were adequate. The quality of later calculations had improved.
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However, thel.' was still room for improvement as attention to detail appears to be lacking in some instances.-
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Maintenance
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in general, plant material condition and housekeeping observed during walkdowns wes a good indication that equipment was being adequately maintained. Preservation of equipment
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i by painting was considered to be good (Section M2.1).
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Enoineerina The inspectors concluded that methodologies used for the current setpoint calculations were wnsistent with current industry standards and practices, were technically sound, and were sufficiently comprehensive for their intended purposes (Section E2.1).
The inspectors concluded that temporary modifications had not impacted the operability of the ND system and that the licensee's program for the installation and removal of temporary modifications was adequate (Section E2.2).
Errors were identified in the ND/ECCS calculations which demonstrated a weakness in the independent review function of calculation design control. The identified ECCS flow analysis calculations errors occurred in 1987 through 1997 which indicated that this was a continuing.
engineering performance deficiency. - Generous margins and performatico test benchmarking provided assurance that the errors did not invalidated the ECCS flow model (Section E3.1).
- An engineering evaluation of the IST program for check valves was still on-going at the conclusion this inspection. This item will remain unresolved pending completion of the licensee's evaluation (Section E.7).
The inspectors concluded that while the quality of calculations was improving, additional management emphasis was required to communicate management expectations to the engineering staff level (Section E.7).
The problem identification process was applied in an effective manner. The number of PIPS indicate a weakness in design control _ (Section E.7).
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Report Details introduction The primary objective of this Safety System Engineering inspection (SSEI) was to assess the adequacy of calculations, analysis, and other engineering documents that were used to support system oerformance during normal and accident or abnormal j
i conditions. The secondary objective of the SSEI was to determine the quality of safety evaluations performed by the licensee in support of engineering modifications performed on the selected system. The inspe.ction was performed by a team of inspectors that included a Team Leader, one Region 11 Inspector, one Resident inspector, and one NRR Instrument and Contro's (l&C) Engineer.
1, Operations
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Operations Procedures and Documentation l
03.1 fleview of Emeraency Procedure Setooints Document l
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Inspection Scope (37550)
The inspector reviewed those portions of Engineering Calculation CNC-1552.08-00-0195, Emergency Procedure Setpoints, Revision 7, related to FWST levels and the j
ND elapsed time operating restrictions and flow setpoints. The purpose of the review was to ensure that the setpoints were consistently used throughout Emergency
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Operating Procedures, Abncnnal Procedures and Operating Procedures, b.
Observations and Findinas The inspector verified that the references (UFSAP., Technical Specifications, Engineering Calculations), from which the Emergency Procedure Setpoints calculations were develcped, were correctly referenced and used in the Emergency Procedure Setpoints engineering calculation document. The inspector then reviewed emergency procedures, abnormal procedures, and operating procedures (See List of Documents Reviewed Section), which referenced the ND and FWST systems, to ensure that the setpoints were used consistently throughout the procedures.
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Conclusions-The inspector conc luded that the setpoints, as developed in the Emergency Procedure Setpoints engineering calculation, related to FWST levels and ND elapsed time and flow setpoints, were used consistently throughout EOP, AP, and OP procedures.
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11. Maintenance
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M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Material Condition Walkdowns
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Insoection Scope (37550)
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The inspectors walked down accessible portions of both the Unit 1 A-train NC and the FWST using flow drawings and elevation drawings (isometric) of
the systems. (See List of Documents Reviewed Section). The purpose of the inspection was to verify the as-built system was consistent with the reference
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drawirigs. Specific attention was directed to verification of pipe hanger / pipe support installation as depicted on the elevation drawings. - Additionally, the inspector,
observed the systems for general housekeeping, equipment condition, excessive j
leakage, deficiency tags and adequacy of equipment identification.
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Observations and Findinos
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The inspectors began the walkdown in the Unit 1 1 A ND pump room and traced the accessible portions of the 1 A ND suction, discharge, and mini-flow piping and the FW piping from the FWST to the 1A ND pump suction. The walkdown included the 1A ND heat exchanger room. During the walkdown the inspectors noted several examples of pipe supports which were depicted on the reference drawings but which had been removed as part of a snubber reduction program. In each example, an interim drawing was available which showed the correct as-built number of snubbers. The interim drawings will be incorporated into the as-built drawings upon completion of the snubber reduction project. The inspector noted that equipment material condition and housekeeping were good and that, in most cases, pipe supports were clearly labeled.
The inspectors were accompanied on the walkdowns by the responsible System Engineers. The engineers demonstrated a good level of knowledge and familiarity with their assigned system.
In general, housekeeping in the general areas around equipment and material was good. Piping and components were painted, and very few indications of corrosion, oil leaks, or water leaks were evident.
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Conclusions In general, plant material condition and housekeeping observed duriag walkdowns was a good indication that equipment was being adequately maintained.- Preservation of equipment by painting was considered to be good. The inspectors concluded that the as-built Unit 1 A-train ND system and the FWST system were consistent with the flow and elevation drawing _ _ - _ _ _ _ - _ _ - _.
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lil. ENGINEERING E2 Engineering Support of Facilities and Equipment
'2.1 Review of Selected Instrument Setpoints
a, Scope of Review (37550)
The l&C inspection focused on the adequacy of the design and configuration of selected safety-related and important-to-safety LPI systems and components. The inspectors also examined engineering calculations to determine whether (1) selected instrument setpoints were properly derived such that automatic actions would occur to prevent safety limits from being exceeded, (2) calculations supporting these setpoints considered all appropriate uncertainties, and (3) setpoint calculation methods,were technically consistent with accepted standards. The adequacy of the above methodology was assessed against the requirernents of the TS, UFSAR,10 CFR Part 50, SERs, and design basis documents. Seven setpoint-related engineering calculation _s were reviewed.
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Findinas l
The inspectors performed a walkdown inspection of the LPI I&C in the control room.
l The inspectors reviewed the LPI l&C functions, such as measurement and indication i
of the refueling water storage tank level, level int'ication for the containment sump, flow rates, and valve position indications and controls. The inspectors found that the display of measured parameters and the provision of controls for pumps and valves in the control room for the LPI functions were appropriate and adequate, t
The inspectors reviewed the licensee's setpoint methodology. The licensee indicated that the majority of the setpoints were established during the initial plant design and construction phases by Duke Power, various equipment vendors, and the respective NSSS vendor. At that time, the current industry standards and regulatory guidelines regarding setpoint determinations did not exist. Duke Power has since issued Engineering Directives Manual, Part 102 (EDM-102), instrument Setpoint/ Uncertainty Calculations. The purpose of EDM-102 was stated in Section 102.1 as follows:
The procedure provides a consistent, programmatic methodology for the development of setpoint and uncertainty calculations, and is intended to support modification activities and evaluation of existing setpoints on an "as-needed' basis. This procedum is not intended to necessitate backfit of existing calculations (performed by Duke, NSSS vendor, or other vendor), which were performed prior to its issuance.
The inspectors determined that the engineering calculations had been updated over the years. The inspectors reviewed successive versions of calculations for some of the LPI associated setpoints. Each successive calculation was found to conform more nearly to current industry standards. However, the licensee was not committed to use the current industry standards, such as the current versions of Regulatory Guide (RG)
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-1.105 and American National Standards Institute / Instrument Society of America
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_ (ANSl/ISA) S67.04.J The inspectors found that methodologies used for the current versions of the engineering calculations that the inspectors reviewed were consistent with current industry standards and practices, were technically sound, and were sufficiently_ comprehensive for their intended purposes.
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Conclusions -
The l&C design cf the LPI systems and components was adequate. Although tho licensee was not committed to use the current industry standards, such as the current -
versions of RG 1.105 and ANSI /ISA S67.04, the inspectors concluded that
methodologies used for the current setpoint calculations were consistent with current industry standards and practices, were technically sound, and were sufficiently
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comprehensive for their intended purposes.
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E2.2 Temocrary Modifications a.
Inspection Scopef37550)
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The inspectors reviewed the temporary modifications to the ND system which had been initiated for the ND system since June 25,1992, b.
Observations and Findinas
- There had been one temporary modification instal!ed and removed from the ND
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system ciuring this time. This modification (97008350 and 97008351, Unit 1 and 2)
temporarily installed thermocoupies to monitor the upper and lower motor bearing temperatures and motor stator temperature. The modification was installed and removed in accordance with the licenste's program for temporary modifications.
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Conclusions The inspectors concluded that temporary modifications had not impacted the
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Engineering Procedures and Documentation E3.1 ND System Desian Documentation a.
inspection Scope (37550)
The inspectors reviewed ND system calculations, design documentation, and.
- performance testing. Calculations were dated from 1987 to 1997. Applicable regulatory requirements included 10 CFR 50 Appendix B, ANSI N45.2.11 - 1974, Quality Assurance Requirements for the Design of Nuclear Power Plants,-UFSAR, and implementing procedures.
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b.
Observations and Findinos
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The calculations were generally adequate in that design inputs, assumptions, methodology, and conclusions were appropriately supported and documented.
However, an exception was noted in several calculatic,ns related to the ECCS flow analysis. Minor errors were identified in the original ECCS flow model calculation and incorrect, unsupported design input errors were identified in a recently developed calculation.
Calculation CNC-1552.08-00-0016. Catawba Units 1 & 2. ECCS RETRAN Model.
Revision O. dated September 28.1987
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This calculation provided the initial model for ECCS flow used for acciden, analysis.
The inspectors verified the ND Train 1 A field configuration was consistent with the as-built isometric / elevation drawings. The as-built drawings were verified to be consistent with the ECCS RETRAN flow model configuration for this train. The
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l inspectors noted minor errors in the model calculation. These included the omission
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of a 90 degree elbow in the 1A train and an incorrect value determined for the loss coefficient for the ND HX In the equation used to determine the loss coefficient for the ND HX, a velocity value was incorrectly inserted for the dp parameter. A later calculation also noted that the ND HX loss coefficient value was not corrected for the appropriate junction defined pipe diameter. Other quality deficiencies in this calculation included frequent line-outs and corrections without dates, initials, or a
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comment sn tion to justify the changes. There was no source identified for an equation frecuently used to adjust loss coefficient values to junction defined pipe diameters. The inspectors noted that adequate margins were incorporated in the calculation to absorb the impact of the errors. The performance of the independent reviewer / checker was weak in that the errors were not identified during the independent review process.
CNC 1552.08-00-0118. Evaluation of Tech. Spec. 4.5.2.h ND Flow Reauirements.
Revision O. dated January 31.1991 This calculation corrected the ND HX loss coefficient error of 'he previous calculation and included the loss due to a ND line discharge orifice omitted from the original RETRAN model. There were no encrs noted in.nis calculation.
CNC 1552.08.00-0181. Safety Iniection Flows for Safety Analveis. Rev. 9. dated September 9.1996 This calculation originated in 1993 and contained an unverified arsumption that NC pump seal flow from the NV was less than 80 gpm during a LBLOCA. Although this statement was included in the Test Acceptance Criteria, it was not tested or analyzed until December,1996, as discussed in the following calculation.
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CNC 1249,00-00-0064. Operability Determination for PIP 0-C96-2467. ECCS Flow I
Balance and NC Pumo Seal Injection. Rev.1. dated December 31.1996 During a Technical Specification review, Engineering identified that the 80 gpm assumption had not been verified and initiated PIP 0-C96-2467 to address this issue.
The associated calculation analyzed operability. This calculation contained errors that were not identified by the indepenaent reviewer / checker. The methodology analyzed i
potential LBLOCA NC pmp seal flow based un NV pump parameters predicted for LBLOCA and normal operation. The NV pump dp run out parameter (LBLOCA condition) was not supported by the referenced performance test and was not consistent with the vendor's pump performance curve. The normal condition NV pump dp parameter was correctly defined in the variable list but a different value was inserted in the equation. The resulting 65.5 gpm seal flow was incorrect. The inspector used more appropriate parameters and the conclusion was 76 gpm which s
was acceptable but much closer to the 80 gpm limit. The 80 gpm assumption of the ECCS flow analysis was not adequately verified by this calculation. The errors demonstrated weak performance during the independent review process in this recent calculation.
The above calculations were used to develop the ECCS flow model for accident analysis. The model was upgraded by benchmarking from system testing. The result of the benchmarking and generous conservatism in the calculation was that the impact of the calculation errors was limited.
CNC-1381.05-0004. Desian Verification for Penetration Protection. Rev. 5 dated June 30.1993
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The inspectors reviewed this calculation to ensure those weaknesses that were identified in the EDSFl had been adequately addressed. The weaknesses reviewed were low short circuit levels used as inputs, inconsistent impedance values, resistance values were not corrected to minimum ambient temperature conditions, and the thermal damage characteristics for the feeder cables should have been based on insulation failure rather than conductor fusing. The inspectors concluded that the weaknesses had been addressed in this revision of the calculation and that the calculation verifies sufficient protection of the Type A, B, C, D, F, and G electrical penetrations to preserve their mechanical and electricalintegrity in the event of an overload or short-circuit fault.
CNC-1381.05-00-0144. Maximum Short-circuit Currents For the 13 8 KV and 6.9KV p_on-safety switchaears and - the 4.16KV Safety Related Switchaears Rev 0 The EDSFI findings identified a lack of formal short circuit studies for the medium voltage essential bus switchgear. The inspector reviewed calculation CNC-1381.05-00-0144 Maximum Short-circuit Currents For the 13.8 KV and 6.9KV non-safety switchgears and the 4.16KV Safety Related Switchgears, Rev 0, dated 6/1/93 and concluded that the short circuit studies had been adequately addressed.
Closing and Latching current level remained within their respective ratings. The most realistic worst cace short circuit for a fault of the 13.8KV switchgear remains within the
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- breaker rating, the 6.9KV switchgears remain within their breaker ratings and the 4.16KV switchgears remained inside their breaker ratings.
CNC-1381.05-00-0159. Simulation of Emeroency Diesel Generator Acceptina LOCA Loadr. Usino CYME. Rev. 3 The inspector reviewed the results of calculation CNC-1381.05-00-0159 Simulation of Emergency Diesel Generator Accepting LOCA Loads Using CYME Rev 3. This simulations indicated that the EDG would function properly when aenpting LOOP /LOCA loads. Throughout the simulation, the generator voltage, frequency and speed were within the limits of Reg. Guide 1.9 Rev 0; large motors (4KV) started and ran without problems because their terminal voltages did not drop below 90%; small motors (575K) started and ran without any problems because their terminal voltages either did not drop below 80% or recovered to we!! above 80% following the iryitial dip such that the over all effect was better than if the voltages were to remain conctant at
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80% during the entire starting period.
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Conclusions
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Errors were identified in the ND/ECCS calculations which demonstrated a weakness in the independent review function of calculation design control. The identified ECCS flow analysis eticulations errors occurred in 1987 through 1997 which indicated this is a continuing engineering performance deficiency. Generous margins and performance test bench marking provided assurance that the errors did not invalidate
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the ECCS flow model. The unverified assumption of LBLOCA NC pump seat flow limits demonstrated deficient Engineering performance in follow up of design
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assumptions.
t:.7 Quality Assurance in Enaineerina Activities a.
Inspection Scope (37550)
The inspectors reviewed the PIPS noted in the list of documents reviewed. This review was conducted to perform an initial assessment of the safety significance of the identified problems, b.
Observations and Findinas The licensee had generated 41 PIPS on the ND system in the two years period on which this inspection focused. Twenty-four of these PIPS were generated during the inspection. While this demonstrates an excellent problem identification process, the inspectors considered it a weakness in that engineering had not identified these problems during earlier reviews.
The inspectors reviewed informal licensee analyses of the PIPS that raised concerns about operability of safety or safety-related systems and concluded that the discrepancies and inaccuracies did not affect operability. However, these items represent examples of weak design contro,
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Four PIPS, 0-C97-0126, 0-C97-0379, 0-C97-2050, and 0-C97-2340 were written against ASME Section XI check valve testing.-- At the conclusion of this inspection the license was stillin the process of determining the extent.of condition. The licensee issued LER 50-413/97-05 on 7/23/97. Part of the corrective action for the LER, an engineering evaluotion of the IST program for check valves was still on-going at the conclusion this inspection. This item will remain unresolved pending completion of the-licensee's evaluation (URI 50-413,414/97-10-01).
Six PIPS,0 C97-0236, 0-C97-2023, 0-C97-2338, 0 C97-2417, 0-C97-2418, and O.
C97-2425 were written to document inadequate design control problems associated with ND and FWST system calculations. -These problems included minor design input-discrepancies which may affect the overall conservatism of the calculations. At the
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conclusion of the inspection, the licensee verified and the inspectors observed that -
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adequate margins were incorporated in the calculation to absorb the impact of the errors. These PIPS were additional examples of the poor performance of the independent reviewers / checkers in that the errors were not identified during the i.
independent review process.- The licensee identified in PIP 0-C97-2283 that EDM 101
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should be revised to provide more consistent calculation quality. The inspectors conclud6d that while the quality of calculations was improving, additional _ management emphasis was required to communicate management expectations to the engineering staff. The closecut of PIP 0-C07-2283 will be followed up during a subsequent inspection (IFl 50-413,414/97-10-2).
PIP 0-C97-2417, dated July 25,1997, documents inconsistencies with Calculation CNC-1552.08-00-0264, FWST Level Setpoints, dated June 25,1997. The inspectors reviewed Calculation CNC-1552.08-00-0264 and concluded that the licensee's
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assessment that the assumed maximum 32-gpm flow from the FWST-(Assumption 1 under Section 10.0) appeared to be low. The PIP further discusses that the TS time for swap to the containment sump was 60 seconds. The inspectors identified that the calculation that was based upon the flow from the FWST during the 37 seconds that both valves are fully open appear to be non-conservative. On the basis of the above concems, the inspectors identified that a potential exists that the swap-over sequence may not be completed before reaching the minimum height in the FWST that is-needed to prevent vortexing. This issue will be reviewed during a future inspection (IFl 50-413,414/97-10>03). The inspectors also identified that the licensee may not
- have given sufficient attention to the different results obtained when the FORTRAN code for the routine,- DEPLET, is compiled on_th; IBM RISC 6000 lnstead of the IBM mainframe computer (Comment i under Section 11.0). Although the difference here was less than 0.3 percent (not significant for this case), the inspectors questioned whether the licensee ensured that the difference was not sigrificant when applied to other calculations that also use DEPLET. This item will be reviewed during a future-inspection (IFl 50-413,'414/97-10-04).
- Six PIPS,0 C97-738,- 0-C97-1354, 0-C97-1974, 0-C97-2110, 0-C97 2111, and 0-C97-
' 2332 were written to document miscellaneous discrepancies between the UFSAR and other documents. The licensee had previously committed, by letter to the NRC dated June 16,1997, to perform an UFSAR accuracy review in addition to a review of -
selected licensing basis correspondence for information that should be reflected in the i
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3 UFSAR for completeness. This accuracy review had not started at the conclusion of this inspection. However, the inspectors reviewed the UFSAR Review Project Methodology dated,4/28/97 and concluded that this process should identify items similar to thoje identified in the above noted PlPs.
Five PIPS,0-C97-2035,0 C97 2381,0-C97-2420,0-C97 2424, and 0-C97 2339 were written to document inconsistencies in engineering documents. These were incorrect references and incorrect items in engineering data bases.
10 CFR 50, Appendix B, Criterion Ill, Design Control requires that measures shall be established to assure that the design basis are correctly translated in to specifications, drawings, procertures, and instructions. The above examples of design control inadequacies and discrepancies were examples of a violation of Criterion Ill. This l
licensee identified violation is being treated as a Non Cited Violation, consistent with Section Vll.lE Y the NRC Enforcement Policy (50-413,414/97-10 05).
The rema e, 3 PIPS were written to identify potential issues that were later determing not to be problems.
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Conclusions The inspectors concluded that while the problem identification process was being applied in an effeuive manner, many of the above mentioned problems should have been identified during engineering's previous reviews of these documents. The inspectors also concluded that while the quality of calculations was improving, additional management emphasis was required to communicate management expectations to the engineering staff.
V. MANAGEMENT MEETINGS X1 Exit Meeting Summary The Team Leader discussed the progress of the inspection with licensee representatives on a daily basis and presented the results to members of licensee management and staff at the conclusion of the inspection on July 30 and 311997.
The licensee acknowledged the findings presented.
PARTIAL LIST OF PERSONS CONTACTED 1lCENSFA C. Bigham, System Engineer M. Birch, Safety Assurance Manager J. Forbes, Er'gineering Manager R. Jones, Station Manager S. Hoss, System Engineer
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J. Kammer, Engineering Point of Contact M. Kitlan, Regulatory Compliance Manager K. Nicholson, Compliance Specialist G. P terson, Catawba Site Vice President NRC:
P. Balmain, Resident Inspector H. Christensen, Branch Chief. Engineering Branch LIST OF INSPECTION PROCEDURES USED IP 37550, Engineering
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IP 9380X, Safety System Engineering inspection
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LLST OF ITEMS OPENED 50-413,'414/97-10-01 URI Completion of corrective actions for check valves
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in the IST program - Section E.7 50-413, 414/97-10-02 IFl Review revision to EDM-101 for calculation quality
- Section E.7 50 413, 414/97 10-03 IFl Resolution of FWST level setpoint inconsistencies
- Section E.7 50-413, 414/97-10-04 IFl Difference results as a result of the use of DEPLET on different computers - Section E.7 50-413, 414/97-10-05 NCV Failure to maintain adequate design controls -
Section E.7 LIST OF ACRONYMS USED AP Abrormal Procedure
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CFR Code of Federal Regu'ations
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DP Diffurential Pressure
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ECCS Emergency Core Cooli1g Systems
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EDG Emergency Diesel Generator
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EDM Engineering Directives Manual
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EDSFI Electrical Distribution System Functional inspection
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EOP Emergency Operating Procedure
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FSAR Final Safety Analysis Report
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FWST Refueling Water Storage Tank
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GPM Gallons Per Minute
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HX Heat Exchanger
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I&C Instrumentation & Control
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Inspector Followup Item
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IP Inspection Procedure
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IST Inservice Test
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LBLOCA Large Break Loss Of Coolant Accident
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LER Licensee Event Report
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LPI Low Pressure injection
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LOCA Loss of Coolant Accident
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ND Residual Heat removal System
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NRC Nuclear Regulatory Commission
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NRR Office of Nuclear Reactor Regulation
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NSSS Nuclear Steam System Supplier
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NUMARC Nuclear Management and Resources Council, Inc.
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-NV Chemical and Volume Control bystem
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OP Operating Procedure
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PDR Public Document Room
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PIP Problem Identification Process
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RG Regulatory Guide
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Safety Evaluation Report
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SSC Structure, System, or Component
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SSEl Safety System Engineering inspection
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TS Technical Specification
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UFSAR Updated Final Safety Analysis Repori
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URI Unresolved item
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LIST OF DOCUMENTS REVIEWE_l,.
Emergency Operating Procedures (eon)
EP/1/A/5000/ECA 1.2 LOCA OUTSIDE CONTAINMENT 11/20/05 EP/1/A/5000/ES 3.1 POST SGTR COOLED USING BACKFILL 8/6/96 EP/1/A/5000/ES-3.2 POST-SGTR COOLDOWN USING BLOWDOWN 8/6/96 EP/1/A/5000/ES-3.3 POST-SGTR USING STEAM DUMP 8/6/96 EP/1/A/5000/ECA-0.2 LOSS OF ALL AC POWER RECOVERY WITH S/l REQUIRED 5/8/97 EP/1/A/5000/ECA-1.1 LOSS OF EMERGENCY COOLANT RECIRCULATION 3/24/97 EP/1/A/5000/ECA.i.1 SGTR WITH LOSS OF REACTOR COOLANT - SUBCOOLED RECOVERY 4/14/97 EP/1/A/5000/ECA-3.2 SGTR WITH LOSS OF reactor COOLANT - SATURATED RECOVERY 5/8/97 EP/1/A/5000/FR S.1 RESPONSE TO NUCLEAR POWER GENERATION /ATWS 7/25/96 EP/1/A/5000/FR-C,1 RESPONSE TO INADEQUATE CORE COOLING 1/15/97 EP/1/AI5000/FR-C.2 RESPONSE TO DEGRADED CORE COOLING 1/15/97 EP/1/A/5000/FR-C.3 RESPONSE TO SATURATED CORE COOLING 10/11/95 EP/1/A/5000/FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK 2/20/97 EP/1/A/5000/FR P.1 RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITION 8/16/96 EP/1/A/5000/ECA-2.1 UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS 8/28/96 i
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EP/1/A/5000/E 0 REACTOR TRIP OR SAFETY INJECTION 4/3/97 EP/1/A/5000!ES 1.1 SAFETY INJECTION TERMINATION 1/31/97 EP/1/A/5000/ES 1.2 POST LOCA COOLDOWN AND DEPRESSURIZATION 7/29/96 EP/1/A/5000/ES 1.3 TRANSFER TO COLD LEG RECIRCULATION 3/24/97 EP/1/A/5000/E 3 STEAM GENERATOR TUBE RUPTURE 4/29/97 Operating Procedures (OP)
l OP/1/A/6150/06 Draining The Reactor Coolant System, Change 34 l
OP/1/A/6150/01 Filling and Venting the Reactor Coolant System, Change 58
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OP/1/A/6200/13 Filling, Draining and Purification of Refueling cavity, Change 28 OP/1/A/6200/04M Residual Heat removal System Drain, Fill and Vent, Change 6
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OP/1/A/6200/04 Residual heat Removal System, Change 68
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UFSAR Changes Changes to UFS4R Sections 5.4 Component and Subsystem Design,6.2, Containment Systems,6.3 Eme gency Core Cooling,7.4 Systems Required for Safe Shutdown,7.6 All Other Systems Required for Safety,15.6 Decrease in Reactor Coolant Inventory Performance Testing (PT)
PT/1/A/4400/01 ECCS Flow Balance Change 25 dated 2/11/97 PT/2/A/4400/01 ECCS Flow Balance Change 16 dated 4/10/96 PT/1/A/4200/10A Residual Heat removal Pump 1 A Performance test Change 46 PT/1/A/4150/0/C NC systera Controlled Leakage Verification, dated 6/15/97 TP/1/A/1200/03C Safety injection Pumps and Flow Adjustment Functional Test (pre-op test) dated 9/28/83 Abnormal Procedures (AP)
AP/1/A/5500/19 Loss of Residual Heat Removal System, Change O AP/1/A/5500/05 Reactor Trip or inadvertent S/l Below P-11, Change O AP/1/A/5500/07 Loss of Normal Power, Change 0 AP/1/A/5500/10 Reactor Coolant Leak, Rev 24 AP/1/A/5500/17 Loss of Control Room, Change 0 Other Documents CNS-1561.ND-00-0001 Design Basis Specification for the Residual Heat Removal (ND)
System Rev 5
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c-Final Safety Analysis report Section 5.4.7, Residual Heat removal System, Dated 30 Nov 1995 Final Safety Analysis report Sections 6.1 and 6.3 Dated 30 Nov 1995 Engineering Directives Manual EDM 102: Instrument Setpoint/ Uncertainty Calculations Rev i EDM 101 Engineering Calculation / Analysis Rev 6,6/30/97 CNS Test Acceptance Criteria, Unit 1 & 2 ND System, ND (Residual Heat Removal) Pumps A and B Safety Analysis Performance Rev 2 CNS Test Acceptance Criteria, Unit 1 & 2 ND System Flow Rate Verification Rev1
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Lesson Plans (LP)
Residual Heat Rimovat System Familiarization (Simulator Exercise GuHe), Rev. 5, dated 4/17/90 Residual Heat Removal System (ND), lesson Plan, Rev 20, dated 1/6/97 Shutdown Operations (SO) Lessen Plan, Rev 5, dated 3/31/97
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Leak in Residual Heat Removal System ( Simulator Exercise Gulde), Rev 2, dated 11/14/95 Loops Dralned/Midloop Operational Characteristics and Response (simulator Exercise Guide), Rev 1, dated 11/15/95 Emergency Core Cooling System (Simulator Exercise Guide, Rev 3 dated 11/27/95 Safaty injection System (NI) Lesson Plan, Rev 22, dated 4/30/97 Engineering Calcuiations (EC)
CNC-1210.04 00 0001 Containment Building LevelInsWment Control Loop Accuracy Rev 6 CNC-1552.08-00-0016 ECCs RETRAN Model, Catawba Units 1 & 2, Rev 0 dated 9/28/87 CNC-1552.04-00-0079 Refueling Water Storage Tank Uncertainty Calculations Rev 0 CNC-1552.08-00-0118 Evaluation of Tech Spec 4.5.2.h ND Flow Requirement Rev 0, dated 1/31/91 CNC-1551.08-00-0181 Safety injection Flows for Safety Analysis, Rev 9, dated 9/9/96 CNC 1552.08-00-0264 FWST Level Setpoints Rev 0 DPC 1552.08-00-0136 ECCS Injected and Auxiliary Containment Spray Flow Assumptions for MNS and CNS Mra and Energy Release / Containment Response Report Rev 2 CNC-1552.00-0240 Flood Volumes in Containment For Large Break LOCA Rev 0
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CNC-1223.1100-0021 Design Parameters For NDHX Bypass Flow Control Rev 1 CNC 1223.12 00 0050 Design Parameters for Containment Sump LevelInstrumentation NILT5260,5270 Rev 0 CNC 1223.1100-0001 Design Parameter Verification for the RHR (ND) System Rev 0 i
CNC 1223.02-00-0005 Minimum Height of Water above Outlet Nozzle Required to i
Prevent Vortexing in Outflow of Tanks Rev 0 CNC 1223.1100-0003 Residual Heat Remova!(ND) Relief Valves 1 &2 ND 3,31. 35.
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38, and G4 Rev 0 l
CNC 1552.08-00-0265 FWST Missile Wall Design Basis Evaluation Rev 0 CNC 1552.08-00-0195 Emergency Procedure Setpoints Rev 7 CNC 1223.21-00-0004 Refueling water Storage Tank Level Setpoints Rev 1 CNC 1381.05-00 0004 600 Penetration Fault Currents, Fusing times and Fusing Currents for Various Conductor Size Rev 5 s
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instrument Procedures (IP)
IP/1/A/3222/0008~
Analog Channel Operational Test Channel 117300 Rev 72 IP/1/Af3222/010 Refueling Water Storage Tank Level Channel 2 Change 11
- 3 IP/1/A/3146/003 (NI) Containment Sump Level Calibration Rev 7 IP/1/A/3144/02A Calibration procedure for Residual Heat Removal (ND) Heat Exchanger 1 A Instrumentation Change 0 IP/1/B/3222/05 Calibration Procedure for Residual Heat Removal Heat Exchangers A and B Bypass Flow Control, Change 11 IP/1/B/3101/01 Calibration Procedure for Refueling Water System (FW) Change
IP/0/A/3816/10 Barton Model 580 and 581 Switch Calibration Change 9 IP/1/A/3144/001B Calibration Procedure for ND Miniflow Control Pressure Switches Rev 12 Drawings (DR)
CN-15501.0 Symbols for Flow Diagrams Rev 7 CN 1550-1.1 Symbols for Flow Diagrams Rev 5 CN 1550-2.0 Catawba Symbols and Abbreviations for Ficw Diagrams Rev 7 CN-1550-2.1 Index of Flow Diagrams (Unit 1) Rev 8 CN-1550 2.2 Index of Flow Diagrams (Unit 1) Rev 1 CN 1561-1.0 Flow Diagram of Residual Heat Removal System (ND) Rev 14 CN 1561-1.1 Flow Diagram of Residual Heat Removal System (ND) Rev 10 CN-1571-1.0 Flow Diagram of refueling Water System (FW) Rev 19 CN 1492-ND001 Auxiliary Building Residual Heat Removal (ND) Rev 15 CN 1492-ND003 Auxiliary Building Residual Heat Removal (ND) Rev 9 CN-1492 ND004 Auxiliary Building Residual Heat Removal (ND) Hav 14 CN 1492-ND020 Auxiliary Building Residual Heat Removal (ND) Rev 5 CN 1492 ND021 Auxiliary Building Residual Heat Removal (ND) Rev 8 CN 1492-ND022 Auxiliary Building Residual Heat Removal (ND) Rev 12 CN 1492-ND023 Auxiliary Building Residual Heat Removal (ND) Rev 12
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~CN 1492 ND024 Auxiliary Building Residual Heat Removal (ND) Hev 15 CN 1492 ND025 Auxiliary Building Residual Heat Removal (ND) Rev 7 CN 1492 ND020 Auxiliary Building Re:idual Heat Removal (ND) Rev 18 CN 1492 ND027 Auxiliary Building Residual Heat Removal (ND) Rev 13 CN 1492 ND042 Auxiliary Building Residual Heat Removal (ND) Rev 9
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CN 1492 ND049 Auxiliary Building Residual Heat Removal (ND) Rev 10 CN 1491 NIO15 Reactor Building Safety injection System (NI) Rev 24 CN 1491 N1026 Reactor Building Safety injection System (NI) Rev 17
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CN 1491 N1049 Reactor Building Safety injection System (NI) Rev 18 CN 1491 N1050 Reactor Building Safety injection System (NI) Rev 5 CN 1491 N1051 Reactor Building Safety injection System (NI) Rev 9 CN 1491 N1052 Reactor Building Safety injection System (NI) Rev 10 l
CN 1491 N1053 Reactor Building Safety injection System (NI) Rev 11 CN 1491 N1054 Reactor Building Safety injection System (NI) Rev 4
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CN 1491 NIO55 Reactor Building Safety injection System (NI) Rev 8 CN 1492 N1039 Auxiliary Building Safety injection System (NI) Rev 8 l
CN 1571 1.0 Flow Diagram of Refueling Water System (FW) Rev 19
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CN 1492 FWOO1 ~
Auxiliary Building Refueling Water System Rev 8 CN 1492 FWO17 Auxiliary Building Refueling Water System Rev 6 CNM 1148.00-0050 Refueling Water Storage Tank Rev D12 CNM 1148.00 0051 Details - Refueling Water Storage Tank Rev DS
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CN 1785 0.1.12-01 Conne-tion & Outline Diagram Process Control System (EIA) Misc.
t Terminal Boxes Rev 16 CNEE 0115-01.20 Elementary Diagram 4160 Switchgear 1ETB Unit #9 RHR PMP. MTR.
1B Rev 9
CNEE 0115 01.09 Elementary Diagram 4160 Switchgear 1ETB Unit #9 RHR PMP. MTR.
1A Rev 8 CN 1702 05.02 One Line Diagram Essential & Blackout Auxiliary Power Systems 4.160 KV/600V Systems, EPC, EPE, ETC Rev 8 CNEE 0141-01.01 Elemental Diagram RHR System (ND) NC Loop 2 Supply to ND Train 1 A Isolation Valve 1ND001B Rev 11 CNEE-C141-01.01 Elemental Diagram RHR System (ND) Status Indication Valves 1ND001B & 1ND036B Rev 0 CNEE-014101.02 Elemental Diagram RHR System (ND) NC Loop 2 Supply to ND train isolation Valve 1ND002A Rev 12 CNEE-0141-01.02-01 Elemental Diagram RHR System (ND) NC Loop 2 Supply to ND Train 1 A Isolation Valve 1ND002A Rev 8 CNEE-014101.03 Elemental Diagram RHR System (ND) ND HX 1 A Outlet to Letdown HX
- 1 Isolation Valve 1ND024A Rev 4 CNEE 014101.04 Elemental Diagram RHR System (ND) ND Pump 1 A Miniflow Valve 1ND025A Rev 9 CNEE-0141-01.05 Elemental Diagram RHR System (ND) ND HX Outlet to Centrifugal Charging Pump 1A &18 Isolation Valve 1ND028A Rev 10 CNEE 0141-01.06 Elemental Diagram RHR System (ND) ND Train 1 A Hot Leg injection Return isolation Valve 1ND032A Rev 7 CNEE 0141-01.07 Elemental Diagram RHR System (ND)NC Loop 3 Supply to ND Train 18 Isolation Valve 1ND 036B Rev 14
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CNEE-0141-01.07.01 Elemental Diagram RHR System (ND) Status Indication for Valves 1ND002A & 1ND037A Rev 0 CNEE 014101.08 Elemental Diagram RHR System (ND)NC Loop 3 Supply to ND Train 1B Isolation Valve 1ND037A Rev 8 CNEE-014101.08 01 Elemental Diagram RHR System (ND)NC Loop 3 Supply to ND Train 18 Isolation Valve 1ND037A Rev 8 CNEE-014101.09 Elemental Diagram RHR System (ND) ND HX 18 Outlet to Letdown HX
- 1 Isolation Valve 1ND058B Rev 4 CNEE 014101.10 Elemental Diagram RHR System (ND) ND Pump 18 Miniflow Valve 1ND0598 Rev 8 CNEE 014101.06 Elemental Diagram RHR System (ND) ND Train 1B Hot Leg injection l
Retum Isolation Valve 1ND065B Rev 7 CNEE-014101.12 Elemental Diagram RHR System (ND) Misc. Control Circuits 1NDFS 5130,5140,5220, & 5230 Rev 7
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CNEE 014101.13 Elemental Diagram RHR System (ND) Misc. Alarm Points Rev 15
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I CNEE-014101.14 Elemental Diagram RHR System (ND) Hx 1A Bypass Flow Cont VI, 1ND027 Rev 5 CNEE-0141-01.14.-01 Elemental Diagram RHR System (ND) Alaim Polnts Rev 70 CNEE 014101.1'$
Elemental Diagram RHR System (ND) HX 1B outlet Flow Cont. VL.
1ND060 & 1B Bypass Flow Cont. VL.1ND001 Rev 5 CNEE-014101.16 Elemental Diagram RHR System (ND) ND Train A Aux. Pzr Spray isol.
Viv.1ND090 Rev 1 CNEE 0141-01.17 Elemental Diagram RHR System (ND) ND Train B Aux Pzr Spray Isol.
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Viv.1ND091 Rev 1
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CNEE 014101.18 Elemental Diagram RHR System (ND) Isolation Valves 1ND 001B &
1ND037A Temp. Emer Fire Protection Controls Rev 2 CNFE 014101.19 Elemental Diagram RHR System (ND) ND HX A(B) Bypass Selector Station CONTROL 1ND027 &1ND061 Rev 7 CNEE 014101.20 Elemental Diagram RHR System (ND) Pump Motor 1 A & 1B Low Flow Alarm Rev 4 Technical Specifications Unit 1 Technical Specifications Sections 3.1.2.5.B. 3.1.2.6., 3.5.4, 3.4.1.3, 3.4.1.4.1, 3.4.1.4.2, 3.5.2, 3/4.3.3.6 Table 4.3.7(14), 3.2.2 Table 3.3 3, 3.3.3.6 Table 3.3-10, 2.2.1 Table 2.21, 3.3.1,3/4.1 Tables 3.3-1, 3.3.3, 3.3.4, 3/4.2 Table 4.3.2, 4.4.1.1 Table 3.3-4, 3.5.4.a, Problem Investigation Process Problem investigation Forms PIP #
Brief Problem Description
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1 C96-1700 1ND-35 Relief valve failed IVW surveillance test,7/8/96 1-C96-1796 Valve 1ND 1B would not stay closed when operated form the ASP,7/16/97 1-C96-1887 Valve 1ND 36B did not meet its reference stroke time,7/24/96
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1 C96 2359 ND pump tripped within a few minutes of establishing NCS loop flow 8/31/96 1 C96 2419 Valve 1ND 1B would not stroke from the control room,9/6/96 2-C96 3250 ND pump 2B failed its quaderly performance test 21/11/96 I
0 C96 3304 Westinghouse TB 96 03 RO need to be evaluated for implementation, 12/18/96 1 C97-0028 ND system for unit i is classified as Ai per the maintenance rule due to exceeding MPFF criteria 1/6/97 2-C07 0157 Ni pump discharge pressure is indicating the same as CLA pressure O C97 0126 Possible inadequate ASME Section XI Surveillance Test for FW and ND Check Valves concerning backlog testing,1/15/97 0-C97 0190 CNC 1223.2100-0005 is inadequate to support the design basis of the FWST missile wall,1/21/97
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0-C97-0236 Design control on FWST levelinstrumentation elevation appears to be inadequate in that a water leg correction was not applied,1/26/97
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0-C97 0379 Performance test does not ensure that all ASME Section XI components are tested prior to startup (ND and FW check valves, backflow testing),2/11/97 0 C97 0738 Certain periodic leak rate tests are not being performed as per the UFSAR (Leakage external to the containment),3/18/97 Change UFSAR to delete the
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lealage rate.
0-C971236 TS 3.4.1.3 appears to be inconsistence with the safety analyses for boron dilution accident,4/16/97 0 C97-1354 License amendment process does not have adequate controls to ensure the UFSAR is maintained up to date,4/23/97 0 C971974 Discrepancies between the unit's Data Curve Books and UFSAR data,6/16/97 2 C97 2003 Flow could not be established to the ND heat Exchanger 2B 8/4/96 0 C97 2021 Cedain " time critical" items identified in the UFSAR and Design Basis are not contained in OPS lesson plans,6/19/97 0 C97 2023 Calculations for design parameters of valves 1 & 2ND 28A do not clearly address the effects of containment pressure of the dp the valve may be expected to see,6/19/97 0-C97-2035 EDM101, Engineering calculations and analysis contains incorrect references, 6/20/97 0-C97-2050 Three Nl check valves have not been reverse flow tested as called for by the ASME Section XI test program, 6/23/97-0-C97-2110 UFSAR and design documents differ about ND suction relief capabilities, 6/28/97 0-C97 2111 UFSAR is inconsistent on design assumptions used to calculate RHR NPSH, 6/28/97 0-C97-2276 NSD 408, Testing, has not been fully implemented, 7/15/97 0-C97 2283 EDM 101 should be improved to provide for more consistent calculation quality,7/15/97 0 C97 2332 An Engineering review found miscellaneous discrepancies in the UFSAR 7/18/97 0 C97-2333 TS 4.5.2.b.1 appears to be non-conservative with respect to the UFSAR, 7/18/97 0-C97 2338 Errors were identified in the FWST levelinstrument calibration procedure concerning boron concentration, 7/20/97 0-C97 2339 EPs are inconsistent for switching to hot leg recirculation, 7/20/97
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0 C97 2340 IWP testing was not performed under assumed conditions,7/20/97 0 C97-2341 Flow to the NC pump seal could exceed 80 gpm in a large break LOCA, 7/20/97 0-C97 2343 Several problems were identified with the ND Design Basis Document,7/20/97 0-C97 2381 Modification Engineering Librcry DBDs are not maintained with Interim As Built" information,7/23/97 0-C97 2417 Calculation CNC 1552.03 00-0264, FWST Level Setpoints, was identified to have minor design input d!screpancies which may affect the overall conservatism of the calculation. 7/25/97 0-C 97 2418 Calculation DPC 1552.08-00-0019, Safety injection Flows for Safety Analysis, dated 9/5/96 does not discuss the level of conservatism associated with the i
assumptions for pump suction pressures. 7/25/97 0-C97 2420 Several restricting flow elements could not be located in Zeus (an engineering
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data base program containing equipment records) 7/26/97
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0 C97-2424 inconsistencies as errors were identified in the design documentation l
associated with containment sump level instrumentation, 7/27/97 0 C97 2425 A calculation could not be located that identifies the minimum volume of water l
transferred to the containment sump is adequate to prevent vortexing in the -
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coritainment sump following a swap to the sump under Small Break LOCA Conditions.- 7/27/97 0-C97-2493 CNC 1381.05 00 0144 was not reviewed within three years as required by the specified Type I retview frequency, Dated 7/30/97 0-C97 2494 RHR lesson plan revision 20 dated 1/6/97 contained minor discrepancies
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