IR 05000395/1998003
| ML20248J341 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 05/29/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20248J325 | List: |
| References | |
| 50-395-98-03, 50-395-98-3, NUDOCS 9806090081 | |
| Download: ML20248J341 (29) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket No..
50-395 License No..
NPF-12 Report No.
50-395/98-03 s
Licensee:
South Carolina Electric & Gas Company Facility:
V. C. Summer Nuclear Station Location:
Jenkinsville. South Carolina Dates:
March 30 - April 3 and April 13 - 17. 1998 Team Leader:
J. Lenahan. Senior Reactor Inspector Engineering Branch Division of Reactor Safety Inspectors:
L. Moore. Reactor Inspector J. Coley. Reactor Inspector J. T. Beard. Engineering Consultant Approved By:
Kerry D. Landis. Chief Engineering Branch Division of Reactor Safety Enclosure 2 9806090001 980529 PDR ADOCK 05000395 O
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l EXECUTIVE SUMMARY l
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l Summer Nuclear Station NRC Inspection Report 50-395/98-03 l
This inspection included a review of the licensee's calculations, analyses.
Other engineering documents, and maintenance practices that were used to support the emergency feedwater (EFW) system performance during normal and
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accident or abnormal conditions. The report co.ered a two-week period of inspection.
Overall the inspection found the system operation to be consistent with the design and licensing basis.
Doerations Automatic EFW start signals were in agreement with the design and
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licensing basis.
Manual operator actions necessary to mitigate abnormal or accident conditions were specified in written procedures consistent with the design and licensing basis.
Maintenance Material condition of EFW equipment and components, and housekeeping in
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the areas around equipment and components was very good.
Maintenance of the EFW system components has been effectively implemented.
The EFW system has performed reliably.
This was identified as a strength.
Surveillance test procedures reviewed for the EFW system effectively
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implemented the design and licensing basis.
Enaineerino The quality of EFW-design calculations was poor.
Although current
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licensee procedures for calculation development were adequate, the design calculations prepared prior to 1995 were not reviewed against
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these standards.
Similar problems with design calculation quality could apply to other safety-related systems.
Calculation quality was identified as a weakness.
An accurate EFW system hydraulic model provided assurance of design and
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licensing criteria that was not provided in the design calculations.
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The 1997 EFW flow balance test used to bench mark the system hydraulic
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model was an example of good engineering support due to the integration of Engineering. Maintenance, and Operations activities.
I The first example of a violation was identified for failure to follow
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procedures for control of M&TE in that a required evaluation was not performed for an identified out-of-tolerance M&TE instrument.
The Design Basis Document (DBD) was a comprehensive consolidation of
design and licensing basis criteria for the EFW system.
Numerous
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deficiencies in design documentation references were self-identified and entered into a corrective action program for tracking and resolution.
l Instrument setpoint calculations used an approved methodology and
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considered appropriate sources of instrumentation inaccuracies.
The second example of a violation was identified for failure to follow
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the 10 CFR 50.59 screening procedure to determine if a design change required a 10 CFR 50.59 Safety Evaluation.
The design change did not involve an unreviewed safety question.
A non-cited violation was identified for failure to analyze a condition
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which had potential to effect environmental qualification of equipment.
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l Renort Details Introduction The primary objective of this Safety System Engineering Inspection (SSEI) was to assess the adequacy of calculations, analysis, other engineering documents, and maintenance practices that were used to l
support EFW system performance during normal and accident or abnormal l
conditions.
The secondary objective of the SSEI was to determine the quality of safety evaluations perforrr.ed by the licensee in support of engineering modifications performed on the selected system.
The i
inspection was performed by a team of inspectors that included a Team Leader, two Region II Inspectors, and one engineering consultant.
Prior to this inspection, the licensee retained an independent engineering consultant to perform a detailed self-assessment of the design, operation, and maintenance of the EFW system.
The consultant. Sargent &
Lundy, performed the self-assessment from February 2 through March 13.
1998. with a five man dedicated project team. Approximately 40 man-weeks were spent onsite in performance of the assessment.
The self-assessment results are discussed in Section E7.1. below.
1. Doerations
Operations Procedures and Documentation 03.1 Review of Emeraency Doeratina Procedures a.
Insoection Scooe (93809)
The procedures related to the accident or abnormal operation of the EFW system were reviewed to verify that they were consistent with the design capabilities of the system and that necessary manual actions are included.
b.
Observations and Findin2E The safety-related operation of the EFW is to provide feedwater to the steam generators for abnormal plant shutdowns so that the steam generators will be available as a heat sink for the reactor coolant system.
In addition, the EFW plays a vital role in the mitigation of postulated accidents such as a main steam line break (MSLB) and a main
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feedwater line break (MFLB). The EFW system is started automatically under the following conditions which were documented on SCE&G Drawing IMS-41-011. sheet 14:
Condition MD-EFW Trains TD-EFW Train 1.
Trip of All Main Feedwater pumps Yes No l
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S/G level low on 2-of-3 channels Yes No l
for any 1 S/G l
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S/G level low on 2-of'3 channels Yes Yes for any 2 S/Gs
Station Blackout (Undervoltage on Yes Yes both Buses)
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Safety Injection signal Yes No 6.
AMSAC (ATWS protection) signal Yes Yes These start conditions agree with the conditions in the UFSAR.
Review of SCE&G drawing C-203-010 shows that the undervoltage signal that starts the TD-EFW requires loss of voltage on both safety buses.
In addition, the only intentional time delay related to the bus loss of voltage is 0.25 seconds.
Therefore, the TD-EFW train will be actuated before the emergency diesel generators have time to start, so in effect.
the TD-EFW train will be actuated on any complete loss of offsite power (LOOP) regardless of the availability of the diesel generators.
The degraded voltage sensor relays on the safety buses have a 3.0-second time delay associated with them. Other than this slight delay, the TD-EFW will similarly be automatically started upon degraded voltage on both safety buses.
This is in compliance with design and licensing requirements'.
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The EFW design includes provisions to detect a faulted steam generator, based on high EFW flow.
The EFW flow control valves will be closed automatically within 48 seconds after EFW operation starts.
The 48 seconds is based on an intentional time delay of 30 seconds to avoid spurious closures plus 18 seconds allccated to the valve stroke time.
In the event of postulated single failures of the EFW automatic isolation system manual operator actions would be required. The licensing basis of the plant assumes that the faulted generator might not be isolated for as long as 30 minutes to account for these operator
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Initially, the plant operators would close the flow control valves from the Main Control Board in the control room within the first 10 minutes.
If the valves could not be closed from the control room, operator actions would be to close these valves locally within the-following 20 minutes.
The inspector verified that the plant emergency operating procedures (EOPs) stipulate these local actions.
'In addition, the User's Guide associated with the E0Ps provides direction that if the automatic EFW isolation has not occurred and a loss of DC power has occurred, the operator is required to:
If the "A" Train dc power is lost, manually close the 2030 valve
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from the CREP to isolate the steam to the TD-EFW or L
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If the ~B" Train de power is lost. manually trip the "A" MD-EFW pump from the main control board in the control room.
These manual operator actions are consistent with the operational aspects of the EFW described in the licensing basis. However, as discussed in Section E7.1. below. the licensee's self-assessment identified an accident scenario wherein these actions could not be performed in some situations.
Resolution of this issue is discussed in Section E7.2.
The team confirmed that the use of the turbine-driven EFW train assumed in the Station-Blackout Evaluation that was approved by the NRC pursuant to 10 CFR 50.63 was consistent with the current design and its capabilities, as supplemented by manual operator actions. that are addressed in emergency operr. ting procedures.
_.c. Conclusions
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The team verified.that tne automatic EFW ' start signals were in
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agreement with the design and licensing basis.
In addition. the team verified that there.were written procedures directing those manual operator. actions necessary to mitigate abnormal or accident I
conditions consistent'with the design' and licensing basi.
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II. Maintenance
M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Material Condition of Emeroency Feedwater System I
a.
Insoection Scoce (IP-93809)
The team reviewed maintenance records and conducted walkdown inspections to determine the condition of the emergency feedwater system and the material condition of equipment and components within this system.
b.
Observations and Findinas The team reviewed maintenance records and discussed system maintenance practices with the EFW system engineer to determine design, maintenance, and testing practices related to the EFW system.
EFW components included in this review were the steam turbine driven emergency feedwater pump (TDEFP). the motor driven EFW pumps, instrumentation and flow control valves. and portions of the EFW piping systems.
The maintenance, and testing practices were compared to information and events described in NRC Information Notices 88-09, 93-51, 94-66 Supplement 1. and 86-14 Supplement 1 & 2 to determine if the equipment discrepancies reported were applicable to the plant. This review disclosed that design and configuration of components within the EFW system has a history of reliable performance. Maintenance practices have been adequate.
The work history for the EFW system was reviewed for the period from April 1992 until the present and founo to be satisfactory with minimal backlog.
Pump and motor vibration data from January 1996 until the present was examined by the team and found satisfactory.
Engineers were interviewed and data reviewed to determine whether the emergency feedwater system and the main steam supply line to the turbine driven pump had experienced any water hammer events.
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erosion / corrosion problems or service induced discrepancies.
No problems were identified in these 6reas.
The team also held discussions with the service water system engineer and reviewed documentation to verify the reliability of this system. The service water system interfaces with the emergency feedwater system to provide additional cooling water if the normal source of water from the condensate storage tank is lost. The inspectors walked down the accessible portions of the emergency feedwater piping. including the main steam supply piping
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to the turbine driven pump to determine the condition cf these
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components. Approximately 40 seismic supports, including snubbers and spring cans, were examined by the team to determine if they were installed in accordance with drawing requirements and were properly set and loaded. A few minor discrepancies were identified by the team during the walkdown which tne licensee l
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documented on a condition evaluation report. The team noted that material condition of equipment and components examined was very l
good as well as housekeeping in the general areas around equipment and components.
System engineers demonstrated a high level of knowledge and familiarity with their assigned systems.
Engineers assigned specialties such as vibration testing, water hammer analysis. and erosion / corrosion were knowledgeable and aware of industry experience in their area of expertise.
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Conclusion Material condition of EFW equipment and components examined. and housekeeping in the areas around the equipment and components was very good.
Maintenance of emergency feedwater system components has been
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effectively implemented. The EFW system has performed reliably.
j M3 Maintenance Procedures and Documentation M3.1 Surveillance Test Procedure (STPs)
a.
Insoection Scooe (IP-93809)
The inspectors reviewed surveillance test procedures for the emergency feedwater system to determine if acceptance criteria were consistent
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with the design and licensing bases.
b.
Observations and Findinos
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l The team reviewed the surveillance test procedures listed in the
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appendix to this report to determine if the acceptance criteria in the procedures were in accordance with the design and licensing bases.
One minor weakness was identified during this review.
Surveillance Test
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Procedure (STP) 220.002. Revision 2. did not specify the acceptance criteria for Technical Specification requirement 3.3-5 Item lla, for the Turbine Driven Emergency Feedwater Pump response time test.
STP 220.002 provided data to STP 550.001. which was used for evaluation of the data for technical specification acceptance.
The licensee issued a change to
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STP 220.002 which refer.ences STP 550 001 to clarify requirements for evaluation of the response time test to the Technical Specification acceptance criteria.
c.. Conclusions Surveillance test procedures reviewed for the EFW system effectively implemented the design and licensing bases.
III. ENGINEERING E1.
Conduct of Engine ring
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El.1 Design Change Control and 50.59 Processes a.
Insoection Scoce The team reviewed the licensee's procedures which control the design change process, b.
Observations and Findinos The team reviewed the current revisions of the licensee's design control procedures.
The team concluded that the procedures adequately addressed the'following: design input, design verification, control of design output documents, post modification testing. control of field changes.
' training requirements, and 10 CFR 50.59 reviews. The team concluded that the procedures provided adequate controls to ensure effective
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Procedure SAP-107 provides the instructions for completing the 10 CFR
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50.59 safety review.
The procedure requires that only trained and certified personnel ' perform or independently review 50.59 screens and
safety evaluations.
Originators of 50.59 screenings are required to document the technical basis in the safety evaluation, and that the
. basis -(justification) be sufficiently detailed such that a qualified independent' reviewer can review and understand the justification and i
verify its' adequacy without recourse to the originator.
The independent
- reviewer must_ be a person who had not been directly involved with the change being considered.
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Ti independent reviewer is required to confirm and substantiate the tecnnical completeness, consistency, and correctness of the safety evaluation.
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Conclusions The team concluded toat the licensee's design control procedures complied with the requirements of 10 CFR 50.59 and 10 CFR 50. Appendix B. Criterion III.
E3.
Engineering Procedures and Documentation
- E3.1 Mechanical EFW Design Calculations a.
Insoection Scoce (93809)
The team assessed the quality of calculations which support the design i-and licensing basis for the EFW system.
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Observations and Findinas The majority of the system design calculations were developed by the architect /(.ngineer (AE). Gilbert Associates, during the initial design and' licensing of the station.
In 1995, the licensee assumed the design authority of the plant and began the transfer and assimilation of design documentation from the AE. The assimilation process entered the calculations into the licensee's document control program but did not specifically include an assessment of the quality of the calculations.
-During the development of the design b0 sis documents (DBDs) the licensee
' identified the design and licensing basis requirements for the system
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and referenced the_ design documentation or testing which supported these requirements.
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During this inspection and the licensee's pre-inspection self-
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assessment, numerous deficiencies were identified in the EFW system design calculations. The signifjectrce of the deficiencies was appropriately evaluated and entered in'ro a corrective action tracking program.
Examples of deficienc1M i'icluded calculations which were not updated to address changing plant tonditions since original design such
as the 1994 Steam Generator (SG) r@lacemerts, or the 1996 power uprate.
In many cases, the original calcubi.1% was replaced by a new
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However, the old.carM6 Son was not removed from the active status.
As a result, severai calculations provide inconsistent
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information regarding a specific design or licensing requirement. This inconsistent information could impact the accuracy of a later operability evaluation or design change, particularly with respect to the available margins assumed in analysis. Other calculations included methodology which was unclear or design inputs which were not properly referenced.
For example, several calculations were used to validate a design requi'rement that the turbine driven EFW pump (TDEFWP) would provide adequate feed flow during natural circulation cooldown, i.e. with the SG pressure at 100 psia and provide 200 gallons p~er minute (gpm) feed flow.
Calculation DC0-5220-050. Determination of TDEFW Performance at Minimum Conditions for Natural Circulation, originally dated September 4. 1987, concluded that 100 psia SG pressure would result in 200 gpm feed flow from the TDEFWP.
Calculation DC0-5220-29. EFW Turbine Pump. originally dated January 6, 1978, and revised March 11. 1998, stated that 100 psia-SG pressure would provide 415 gpm total EFW flow.
Both calculations were on active status.
Another example was calculation DCO-5220-072. Minimum and Maximum EFW Flow Under FSAR Postulated Accident Conditions, originally dated July 23. 1976, which evaluated maximum available flow to verify that the accident assumption limit of 1000 gpm to a faulted SG was met.
The methodology was unclear and design inputs were not consistently referenced such that an independent reviewer was forced to make assumptions regarding the originator's methodology and inputs.
Additionally, the calculation manually constructed system curves on a pump curve and derived critical values from the interpretation of this curve. The final worst case value of 996 gpm provided little margin from the 1000 gpm limit and the error assumptions were not addressed.
Additionally the calculation had not been upgraded from the SG
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replacement or power upgrade. The maximum flow limit was validated by more recent system flow modeling.
However, calculation DCO-5220-072 was not retired.
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The team reviewed a recent calculation and noted that the qu'ality of this document was improved from the older calculations.
Calculation DC0-5220 077. Condensate Storage Tank EFW Nozzle Hydraulic Conditions, dated April 8.1998, evaluated the potential EFW pump CST suction line air entrainment due to vortexing. The methodology and design inputs were clearly stated. The independent review which was attached to the calculation was thorough and challenging.
This indicated that the
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current procedure guidance for calculation' development was adequate, however, the calculations assimilated from the AE had not been examined to this standard.
As a result of the poor quality of.EFW design calculations it was difficult.to validate.various design and licensing requirements.
The licensee was aware of the deficiencies in system design documentation.
The DBD process and the corrective action process provided a mechanism-for resolving these deficiencies. The team noted that the majority.of l
the calculations were assimilated in 1995. yet the numerous deficiencies were recently identified and being addressed.
The quality of EFW design t
calculations indicated that other safety-related systems' design
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. documentation for which the licensee assumed design authority could be of similar quality These systems have not received the attention focused on the EFW system..The inspectors noted that these systems did not include the Nuclear Steam Supply Systems (NSSS) for which Westinghouse _ remains the design authority. The team identified the quality of EFW design calculations as a weakness.
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Conclusions The quality of EFW design calculations was poor. Although current licensee procedures for calculation development were adequate, the design calculations prepared prior to 1995 were not reviewed against these standards. Similar problems with calculation' quality could apply to other safety-related systems.
Calculation quality was identified as a weakness.
E3.2 System Hydraulic Flow Models a.
Scone (93809)
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The team reviewed the accuracy and application of system hydraulic flow
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modeling used for validating the design and licensing basis criteria for the EFW system.
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Observations and Findinas There were several flow models which were used to evaluate the system response to design base accident scenarios. The PIPF hydraulic model was developed by the AE in 1988. A more recent model was developed by the licensee and documented in Engineers Technical Work Record LC 13109.
- dated November 4. 1997, and calculation DCO-5220-076. ESW PIPE-FLO.
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dated March 2, 1998.
This model was bench marked, or normalized to test data, against actual system performance from a flow balance test performed on November, 4. 1997, per STP 220.011, EFW Flow Balance Verification.
FSAR accident scenarios were input to the model to verify the. system performance assumed in the accident analysis..The performance criteria included minimum and maximum flow, high flow isolation, and flow split to the SGs.
The team reviewed similar accident scenarios run on both the AE and more
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recent licensee models and noted that the results were consistent in demonstrating system performance. Accident scenarios which required EFW input included Main Steam Line Break. Main Feed Line Rupture, Loss of Main Feedwater, Loss of Coolant Accident, and Loss of Offsite Power.
The team reviewed a portion of the system model configuration and verified that it was consistent with the as-built configuration. The s'ystem hydraulic models provided assurance that the EFW system response was consistent with the design and licensing bases.
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Conclusion The EFW system had been accurately modeled and the model was used to verify that system performance was consistent with the design and licensing bases. The system modeling provided assurance of design and licensing criteria that was not provided in the design calculations.
E3.3 Test Documentation a.
Insoection Scooe (93809)
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The team reviewed test documentation to assesses the adequacy of station testing to validate design and licensing basis criteria.
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Observations and Findinas
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A specific criteria not validated by design calculations or hydraulic models was the one minute initiation time for EFW flow assumed in the accident analysis. This criteria was periodically verified in the Engineered Safeguards Features (ESF) response time test. The team reviewed STP 550.001A. "A" Train Reactor Trip /ESF Total Response Time Test, dated November 19. 1997, which verified that the EFW system was initiated within one minute of the reactor trip /ESF actuation signal.
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Additionally, the team reviewed SAP-134. the EFW Flow Balance Verification, dated November 4.1997. This was the test used to bench mark the system hydraulic model.
Documentation of the cest was generally good and demonstrated good integration of Engineering.
Maintenance, and Operations organizations.
A deficiency was noted in the flow balance test activity related to the control of Measuring and Test Equipment (M&TE). A calibrated test instrument used during the test was found to be out of tolerance during the post-test calibration. This was test transmitter TU-8. The applicable station procedure. SAP-141. Control and Calibration of M&TE.
revision 11. required an out-of-tolerance (00T) evaluation of the instrument impact on the test. This 00T evaluation was not performed.
Following identification of this issue by the team, the licensee initiated CER 98-0376. dated April 16. 1998, to evaluate the 00T instrument and determine why the required review was not performed.
The evaluation indicated the test results were not impacted. This failure to meet the procedure requirements for M&TE control was identified as an example of Violation 50-395/98-03-01. Failure to Follow Procedures.
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Conclusions Testing to validate design and licensing basis criteria was generally good.
The 1997 f7W flow balance test was an example of good integration of Engineering Maintenance. and Operations activities. A violation was identified for failure to follow procedure for control of M&TE in that a required evaluation was not performed for an identified out-of-tolerance M&TE instrument.
E3.4 EFW System Design Basis Document (DBD)
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Insoection Scooe (93809)
The teams review of the system design and licensing basis included the system DBD and the station blackout analysis.
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Observations and Findinas I
The current revision of the DBD was revision 10. dated March 18, 1998.
This revision incorporated deficiencies identified by the licensee's i
self-assessment of the EFW system. The DBD was comprehensive in its I
listing of design and licensing criteria and functional requirements for
the system and equipment. Although the poor quality of design
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L calculations impacted the reliability of the design references, the DBD indicated a thorough understanding of the system and equipment design function. As previously discussed in paragraph E3.1, above, deficiencies in design documents have been evaluated for significance and entered into a corrective action program for resolution. The licensee's threshold for deficiency identification was low which contributed to the large number of identified deficiencies and demonstrated a commitment to establish a reliable and comprehensive DBD.
The team reviewed GAI Report 2782. Station Blackout Evaluation. Revision 6. dated May 27, 1992. This report discusses the licensee's methodology
for complying with the requirements of 10 CFR 50.63. Loss of All Alternating Current Power (Station Blackout).
Review of the report disclosed some discrepancies between the 5B0 analysis and the current plant configuration.
None of the discrepancies invalidated the
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licensee's analysis, or the NRC Safety Evaluation Report. The licensee stated that the SB0 analysis would be revised and updated to reflect current plant configuration.
In addition, the SB0 analysis will be added to the checklist of potentially affected documents. to be considered during subsequent plant modifications.
Failure to maintain the SB0 report current was identified to the licensee as a weakness.
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Conclusions The DBD was a comprehensive consolidation of design and licensing basis criteria for the EFW system.
Numerous deficiencies in design documentation references were self-identified and entered into a corrective action program for tracking and resolution. A weakness was identified for failure to update the SB0 analysis to reflect current
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E5.5 Review of Instrumentation Setpoint Calculations sa.
Insoection Scone
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The team reviewed setpoint calculations for instrumentation associated with the EFW system to ensure that the design basis for the plant was being maintained; b
0 observations and Findinos
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.The team reviewed setpoint calculations and evaluated the application of
- setpoint methodology to determine'if appropriate types of inaccuracies
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(or. uncertainties) were considered in the calculations. Three setpoint calculations were reviewed. These were the level instrumentation for-the condensate storage tank'(Calculation No. DC0-9620-013), the pressure instrumentation at the pump suction that is used to automatically switch the EFW water. source from the condensate storage tank to the plant service water system (Calculation No. DCO-9630-018). and flow instrumentation that is used to sense and isolate a faulted steam
' generator (Calculation No. DC0-9630-024). The team confirmed that the uncertainty analyses included in the three selected setpoint calculations were consistent with NRC approved setpoint methodology, and l
that the calculations considered appropriate sources of instrumentation inaccuracies.
l During the review of the high flow setpoint calculation for the TD-EFW train. a discrepancy was identified which indicated that the TD-EFW pump would have insufficient net positive suction head (NPSH) at virtually the same flow rate.as the high flow setpoint. The concern was that damage to the TD-EFW pump could occur prior to the pump tripping due to a high flow rate.
In response to NRC questions. the licensee reevaluated the calculation using an updated computer model. The results of the new analysis demonstrated that there was adequate NPSH at flow rates well in excess of the high flow setpoint value stated in the calculation.
This discrepancy was similar to those discussed in Paragraph E3.1.b. above. regarding examples in which several active calculations or analyses contained conflicting information.
In conjunction with the suction pressure instrumentation calculation.
calculation number ~DCO-5220-070 was reviewed. This calculation
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The calculation assumed that a tornado strikes the
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condensate storage _ tank and severs the piping and instrumentation tubing at ground level while all three EFW pumps are running. The calculation showed that the water standing in the piping above the suction of the pumps would provide sufficient head of water to preclude a pump NPSH problem for.several seconds longer than is necessary for the automatic switchover to' service water. The time required to initiate the switchover includes an intentional 5-second time delay, even if all offsite power was lost at the instant the switchover was to be initiated and included the time delays associated with starting and sequential loading the emergency diesel generators.
The design output from the setpoint calculations listed determination of uncertainties in engineering units such as inches water column. gallons u
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per minute (gpm), or pounds per square inch gauge (psig). General maintenance procedure GMP-100.016. Instrumentation and Controls Scaling Document, was used to convert the engineering units into instrumentation-values. The results of the setpoint calculations become inputs to the periodic surveillance calibration' procedures. The team reviewed surveillance. test procedure'(STP-396.007) for the pump _ suction pressure instrumentation. The team verified that the results of the corresponding setpoint calculation and scaling document were specified
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Conclusions The team concluded that-the setpoint calculations used an approved methodology and considered appropriate sources of instrumentation inaccuracies. The team concluded that the licensee is maintaining the.
design basis.
E3.6 Modifications and 50.59 Reviews a.
Insoection Scooe The team reviewed modifications to the EFW system to determine if the design basis was being maintained and NRC requirements were followed in review and approval of the modifications.
b. ~ Observations and Findinas The team reviewed the modifications listed below in order to: 1)
determine the adequacy of the 10 CFR 50.59 safety evaluations: 2) verify
.that modifications were reviewed and approved in accordance with Technical _ Specification.and applicable administrative controls: 3)
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verify that applicable design bases were included: and, 4) verify that UFSAR commitments were being maintained. Modifications reviewed were as
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-follows:
MRF-21222 A temporary (non-permanent) modification to the EFW system which installed special pressure transducers and associated
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instrumentation to investigate pressure spikes that had caused the
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l-failure of the normal pressure instrumentation.
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-MRF-21098 Temporary installation of an additional flow path for testing and providing a seismically-qualified mounting for local flow gauge for -
the TD-EFW train. A subsequent MCN made this temporary modification into a permanent modification.
MRF-21514 Changes to the setpoints for the high flow instrumentation that are used to detect and isolate the EFW in the event of a faulted steam generator.
i c.
Conclusions l
The team concluded that the modifications were performed in accordance with the change control procedures.
E.7. Quality Assurance in Engineering Activities E7.1 Licensee Self Assessments a.
Insoection Scooe(37550)
The team reviewed the licensee's pre-inspection self-assessment of the EFW system, b.
Observations and Findinas-The licensee retained an independent consultant to performed a detailed,
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in-depth self-assessment of the.EFW system prior to this inspection.
The consultant. Sargent & Lundy, spent approximately 40 man-weeks onsite-performing the self-assessment from February 2 through March 13. 1998.
The results of the self-assessment were documented in a report titled Emergency.Feedwater System Engineering Assessment. dated April 13. 1998.
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The team reviewed the licensee's report which documented the Assessment
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findings.
The report discusses the system description, design basis, the design, operations.. maintenance, and test requirements for the EFW components, a description of ongoing improvement programs, e. g., the FSAR review project. and summary and conclusions of the self-assessment.
A total of;119 issues were identified as a result of the self-assessment. These issues-included minor drawing discrepancies.
inadequate-documentation.in 10 CFR 50.59 evaluations.
inconsistencies / discrepancies in calculations, inconsistencies between the DBD and other documents, minor hardware discrepancies, lack of
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documentation in some instrunent setpoint calculations, and minor procedural inconsistencies / discrepancies.
None of the issues identified resulted in a unreviewed safety question. The assessment findings were documented by the licensee in condition evaluation reports (CERs). The calculation issues were similar to those discussed in Section E3.1, above. These involved lack of documentation in the calculations, failure to update calculations to reflect current plant design conditions, the fact that several active calculations contained conflicting information, and lack of a calculation to validate various design assumptions.
However, none of the calculation issues resulted in an unanalyzed condition. A significant finding from the self-assessment identified a possible unanalyzed plant operating condition which was identified to NRC as Licensee Event Report (LER) 1998-004.
This issue was documented in CER 98-0259 and is further discussed in Section E7-.2.
below.
c.
Conclusions The team concluded that the licensee's pre-inspection self-assessment which was performed on the EFW system was very thorough and effective in identifying discrepancies in configuration. design, and operation of the EFW system.
E7.2 Corrective Actions to Resolve EFW Nonconformances a.
Insoection Scooe The team reviewed nonconformances initiated to document and disposition discrepancies identified involving the EFW system,
_b.
Observations and Findinas The team reviewed the licensee's corrective actions to disposition the nonconformances discussed below.
Prior to 1996, nonconformances implemented as required by 10 CFR 50. Appendix B. Criterion XVI, were i
identified under the licensee's program as Nonconformance Notices (NCNs).
The licensee's nonconformance program was revised in 1996 to a more encompassing Condition Evaluation Report (CER) system.
The team reviewed the CERs and NCNs discussed below:
CER 96-0052. TDEFP Oscillation Problem The Operability and Deportability determinations were appropriate. When minor modifications
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were made to the design as recommended by the vendor engineering
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representative, the 10 CFR 50.59 safety evaluation process was performed
in accordance with NRC requirements.
CER 96-0225. TDEfW Exceeded Maximum Soeed The overspeed situation was I
l identified as a repetitive problem. The corrective action was to readjust the governor. The effectiveness of this correction will be monitored during routine surveillance testing.
CER 97-0356. Check Valve Operability Following a plant trip on 04/22/97. the operators noted that the differences between the EFW flows to the three steam generators were significantly greater than previous
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plant experiences. A subsequent test of the flow capabilities of the EFW check valves indicated that one valve did not pass an adequate flow rate.
It was also determined that mechanical stops in the flow control valves were not adjusted properly. The technical evaluation was comprehensive and considered alternate test methods supported by NUREG-1482. The evaluation also developed and conducted special investigative tests, recognized the contributions of instrument uncertainties, and utilized the licensee's computer model of the EFW system to analyze the effectiveness of the corrective actions.
The team confirmed that the mechanical stops on the flow control valves are now part of a periodic surveillance action.
CER 97-1276. Install / Removal of Heise Gauae The operability evaluation and corrective actions were adequate.
CER 98-0154. CST Volume This problem concerned failure to revise the minimum condensate storage tank volume available for the EFW system to account for the increased volume required for the uprated power level modi fication. The licensee determined operability was maintained because the condensate storage tank had actually been maintained in
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excess of the minimum level required for accident conditions.
However, the Technical Specification values and CST level alarms were not correct. This issue was identified as a non-cited violation (NCV) in
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NRC Inspection Report number 50-395/98-01.
CER 98-0259. Unanalyzed Condition for Eauioment Qualification The licensee's self-assessment identified a possible issue regarding the inability to isolate EFW flow to a faulted steam generator.
This issue involved a main steam line pipe break outside containment coincident with loss of offsite power and loss of the "A" train DC power.
The postulated accident scenario would result in the inability to close the valves supplying steam to the TDEFW pump from the main control board.
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and due to the high energy line break outside containment, local access
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to the valves or to the control room evacuation panel (CREP) would be inhibited. The'resulting accident would result in exposure of some environmentally qualified electrical equipment to temperatures and steam in excess of the 10 minute period assumed in the licensee's accident analysis.
The licensee took~immediate corrective actions to resolve this problem by stationing an operator in the CREP. The operator was_in communication with the' control room via telephone and two-way radio.
The licensee was developing.a modification to' permit closure of the.
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steam supply valves from the control room.
Further review of the sequence of events required to result in the accident scenario described above disclosed that the probability of their occurrence was very low.
The team identified the fact that an unanalyzed accident scenario was identified and had not been previously considered by the licensee in their accident analysis as a violation of NRC requirements.
However, this event had very low safety consequences, and the licensee initiated very comprehensive and immediate corrective actions.
This non-repetitive, licensee-identified and corrected violation is being treated as a Non-cited Violation (NCV) consistent with Section VII.B.1 of the NRC Enforcement Policy.
This was identified to the licensee as NCV 50-395/98-03 02. Unanalyzed Environmental Condition for Secondary Pipe Break outside Containment.
CER 98-0384. TDEFW Pumo TS Surveillance Inadecuacy As part of the EFW DBD improvement effort, the licensee discovered that they did not have a formal design calculation which validated the surveillance parameters in TS 4.7.1.2.a.2 for measurement of the TDEFW pump flow and discharge pressure.
The licensee initiated calculation DC0-5220-079 to calculate
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the design requirements.and compare them to the TS surveillance requirements. The calculation showed that the TS values were incorrect.
The team reviewed the licensee's operability assessment for this issue.
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The results of the operability assessment showed that, although the values in the TS were nonconservative, the speed settings used by the licensee for the TDEFW pump resulted in adequate flow from the TDEFW
- pump which would have exceeded minimum flow requirements.
NCN #5464. Incorrect Flow Transmitters This NCN involved replacement of the instrument transmitters used for EFW flow to the steam generators.
The original plant equipment used transmitters with a range of 0 to 750 inches. water column (in wc). However, the plant had experienced several
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instances of the original transmitters being unreliable and drifting out of tolerance.
It was found in May 1996 that the instrument manufacturer (Rosemount) did not recommend use of this particular transmitter for applications where the measured range'was less than 0-125 in we.
In this application, the plant had been calibrating these transmitters for a range of 0-99 in we and the manufacturer indicated that this range was the likely'cause of the excessive drift that had been experienced. The disposition of the Nonconformance Notice was to replace the instruments with instrument transmitters that had a range of 0-150 in we, which is compatible with the 0-99 in wc application.
As-part of the disposition of this NCN a 50.59 screening was performed to determine if this change required a 50.59 safety evaluation. The plant procedure that governs 50.59 process is SAP-107, 10 CFR 50.59 Unreviewed Safety Question Review Process Revision 0.
The 50.59
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process requires the safety reviewer to answer four screening questions.
If the answers to any of the screening questions are "Yes", the safety reviewer is required to perform a review to determine if the change involves an unreviewed safety question. Appendix II to procedure SAP-107 contains the instructions for answering the screening questions.
Paragraph B.3.g.1 specifies that a "YES" answer is required to the screening question "Does the activity represent a change to the facility as described in the FSAR?" when the range of a replacement instrument is not equivalent to the existing instrument being replaced.
Review of the 50.59 screen showed the reviewer answered the question "No".
The reviewer's basis for that answer stated that in part that: "The FSAR/FPER does not state the model number of the transmitters used for this application." The failure to perform the 50.59 screening in accordance with the. requirements of procedure SAP-107 was identified to the licensee as a second example of violation item 50-395/98-03-01.
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Failure to Fcllow Procedures.
The team concluded that this issue did not result in an unreviewed safety question.
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.c.
Conclusions
,The team concluded that the corrective actions to resolve the above nonconformances were appropriate. An additional violation example was identified for fcilure to follow the 10 CFR 50.59 screening procedure.
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V. MANAr# MENT MEETINGS X1 Exw
' ting 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 manageant and staff at the conclusion of the inspection on April 17.~ 1998.
The licensee acknowledged the findings presented.
PARTIAL LIST OF PERSONS CONTACTED l
. LICENSEE:
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S. Byrne. General Manager. Nuclear Plant Operations S. Furstenberg. Manager. Maintenance Services D. Lavique, General Manager. Nuclear Support Services G. Moffatt. Manager. Design Engineering l
A. Rice Manager. Nuclear Licensing and Operating Experience G. Taylor. Vice President. Nuclear Operations
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R. Waselus. Manager. Systems and Component Engineering B. Williams, General Manager. Engineering Services EC:
B. Bonser. Senior Resident Inspector J. Jaudon. Director. Division of Reactor Safety M. King. Resident Inspector LIST OF INSPECTION PROCEDURES USED IP 93809.
Safety System Engineering Inspection
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LIST OF ITEMS OPENED
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50-395/98-03-01 VIO Failure to Follow Procedures - Sections E3.3 and E7.2 50-395/98-03-02 NCV Unanalyzed Condition for Equipment Qualification During a Secondary Line Break Outside Containment - Section _ _ _ _ _ - _ _ _ _ _ - _ _ _ - - _
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LIST OF DOCUMENTS REVIEWED Emeraency Doeratina Procedures E0P-3.0 Rev. 9. Faulted Steam Generator Isolation E0P-6.0 Rev.13. Loss of All ESF AC Power E0P-15.4 Rev. 5. Response to Steam Generator Low Level OAG-103.4 Rev. 2C. E0P/ADP User's Guide l
l Maintenance Procedures
I GMP-100.016 Rev. 2. Instrumentation and Controls Scaling Document Development Program MMP-300.039. dated Septemoer 11. 1997. Motor and Turbine Driven EFW Pump Flow Control Valve and Actuator Maintenance Station Administration Procedures
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SAP-107 Rev. O and 1. 10 CFR 50.59 Unreviewed Safety Question Review Process SAP-133 Rev. 9. Design Control / Implementation and Interface SAP-139 Rev. 18, Procedure Development. Review. Approval. and Control SAP-300 Rev. 8C. Conduct of Maintenance Activities SAP-1141 Rev. 6A. Non-Conformance Control Program SAP-1205 Rev. O. Setpoint Program Enaineerina Services Procedures ES-109 Rev. 2. Conduct of the Principal Engineering Review Group ES-110 Rev. 08. Review and Verification of Controlled Engineering Documents ES-416 Rev.138. Design Modification Change Process and Control ESr419 Rev. 7. Limited Scope Design Changes Request for Engineering Evaluations and Equal To/Better Than Evaluations ES-452 Rev. 00. Design Control: Modification Screening ES-453 Rev. 1. Design Control: Commercial Change
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ES-454 Rev. OB Design Control: Plant Enhancement ES-455 Rev. 1B. Design Control: Plant Modification ES-509 Rev. 3B Disposition of Site Nonconformances Surveillance Test Procedures
'STP-120.003, dated November 7,1995. Emergency Feedwater Valve Verification STP-120.004. Rev. 12. Emergency Feedwater Valve Operability Test STP-120.005. Rev. 6. Emergency Feedwater Actuation Test
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t STP-120.006. Rev. 6. Emergency Feedwater-Valves Backup Air Supply Test
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l STP-130.003. dated November 3, 1997. Valve Operability Testing j
l STP-130.005E. dated January 30, 1995. EF Valve operability Testing
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STP-220 001A Rev. 5. Motor Driven Emergency Feedwater Pump and Valve Test STP-220.002 Rev. 2. Turbine Driven Emergency Feedwater Pump and Valve Test STP-220.007. dated January 2. 1996. Backup Air Supply Check Valve Test for Emergency Feedwater Valves STP 220.007A. dated November 12. 1997 IST of EFS and MS Backup Air Supply Valve i
STP-342.001 Rev. 5. Condensate Storage Tank Level Instrument (ILT03621)
l Calibration STP-342.002 Rev. 4. Condensate Storage Tank Level Instrument (ILT03631)
Calibration STP-390.001 Rev. 5. Emergency Feedwater Flow to Steam Generator A Flow Instrument (IFT03561) Calibration STP-390.002 Rev. 5. Emergency Feedwater Flow to Steam Generator B Flow Instrument (IFT03571) Calibration STP 396.001, dated October 1. 1997. EFW to SG A Flow Instrument (1FT03561)
Calibration STP-396.007 Rev. 7. Emergency Feedwater Pump Suction Pressure Instrumentation (IPT03632) Calibration STP-396.011 Rev. 4. Emergency Feed Pump Suction Pressure I Instrument (IPT03632) Operational Test STP 0550.001A dated November 19.1997. "A" Train Reactor Trip /ESF Total Response Time STP-803.002 Rev. 7. Mechanical Snubber Visual Examination STP-803.003 Rev. 5. Mechanical Snubber Operational Test Calculations DC0-5220-029. originally dated January 6.1978 revised March 11, 1998. EFW
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Turbine Driven Pump operation DCO-5220-036. dateo April 16, 1987. EF Flow Control Valve Emergency Closure
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DCO-5220-044, dated April 28. 1995. Modeling of VCS EFW Motor Driven Pump Operation DCO-5220-050. originally dated September.4.1987. Determination of EFW Pump Performance of Minimum Conditions for Natural Circulation DC0-5220-52, originally dated April 23, 1987. Determination of EFW Flow Split to 2/3 Unfaulted SGs DCO 5225-055, dated August 24, 1987. Emergency Feedwater System Flow Rates for Various' Configurations DC0-5220-058. originally dated December 14. 1987. Determination of Flow and NPSH Requirements for One MDEFW Pump
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DC0'-5220-063, originally dated November 21, 1988. Evaluation of EFW Pumps for Concerns identified in NRC Bulletin 88-04 DC0-5220-064, originally dated December 14, 1987, Determination of Pressure Loss to the.TDEFW Pumps:
DC0-5220-069. dated' July 16, 1976. EFW System Control Valve Design Conditions DC0-5220-070, dated March 26, 1998. Evaluation of Allowable Time to Accomplish Switchover from Condensate. Storage Tank to-Service Water DC0-5220-072, originally dated July 23, 1976. EFW FSAR Data DC0-5220-077, dated April 8,1998, Condensate Storage Tank EFW Nozzle Hydraulic conditions
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DC0-5220-108, dated December 15 1993. EFW Hydraulic Analysis
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0C0-9610-23. dated March 19, 1998. EF Motor Driven High Flow Loop Accuracy calculation
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DC0-9610-024, dated March 19, 1998. EF Turbine Driven High Flow Loop Accuracy L
Calculation
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DCO-9620-013. dated April 3.1998. Condensate Storage Tank Level Instrument Loop Error Analysis DC0-9630-018. dated March 19, 1998. EF Pump Suction Swapover Pressure Setpoint Calculation Enaineerina Drawinas SCE&G drawing D-302-085 Rev. 38 (FSAR Figure 10.4-16) System Flow Diagram.
Emergency Feedwater (Nuclear)
SCE&G drawing IMS-41-011. Rev. 7 Sheet 14, (Westinghouse drawing 108D837)
Functional Diagrams. Auxiliary Feedwater, Pumps Startup SCE&G drawing C-'203-010 Rev. 4, 7.2 KV Bus 1DA-1DB Undervoltay;: i; 9 ying Logic Diagram SCE&G. drawing B-208-032 Rev 7. Sheet 2. Electrical-Elementary Diagram Motor i
Driven Emergency Feedwater Pump B (XPP21B)
Fisher drawing 54A7513 Rev. 6G, 3 Body 87 Actuator., 657-ET, Diaphragm Actuated
' Control Valve-SCk&G drawing E-302-089 Rev. 23. Emergency Feedwater Intermediate Building
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. Plans Below El. 436' 0"
'SCE&G. drawing B-208-032 Rev. 8, Sheet 35B. EF Isolation From Motor Driven PP's
.to Steam Generator "B" IFV-3541
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SCE&Gl drawing.B-208-032..Rev. 13.. Sheet 37A. EF Isolation From Turbine Driven PP's to Steam Generator "A" IFV-3536
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SCE&G drawing VCS-IFT03531-EF Rev. 3. Instrument Loop Diagram, Emergency Feedwater to Steam Generator 'A' Flow. Train B
- SCE&G drawing VCS-IFT03571-EF Rev. 2. Instrument Loop Diagram. Emergency
~Feedwater to Steam Generator 'B' Flow
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Pittsburgh-Des Moines Steel Co. Drawing F4 Rev. A3 Condensate Storage Tank.
General Arrangement q
l SCE&G drawing B-814-304 Rev. 6. I&C Instrument Diagram. Yard El.
435' 0"
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Condensate Storage Tank. Level LS-3172. LS-3173. LS-3174. LT-3621. LT-3631 l
SCE&G Drawing 1MS-39-142 Rev. 7. Yard Piping. Circuits #7. 8. & 9 l
SCE&G Drawing VCS-ILT03621-EF. Rev. 2. Instrument Loop Diagram, Condensate
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Storage Tank Level. Train A SCE&G Drawing D-302-011 Rev. 33 (FSAR Figure 10.3-1) Steam Flow Diagram (Main Steam Nuclear)
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l SCE&G Drawing D-302-085 Rev. 38. System Flow Diagram. Emergency Feedwater (Nuclear)
SCE&G Drawing C-314-011 Re'. 3. Sheet 9. Main Steam Supply to Emergency leedwater Pump SCE&G Drawing C-314-085 Rev. 6. Sheet 1. Piping Analysis Diagram.
Emergency Feedwater - Pump XPP-8-EF Discharge to 18" Feedwater Header SCE&G Drawing C-314-85. Rev. 7. Sheet 5. Piping Analysis Diagram. Turbine Driven Emergency Feedwater Pump Suction Technical Specifications Technical Specifications Sections 4.0.5. 4.7.1.2.a.1. 4.3.2.1 & Table 4.3-2 item 6a and 6b, 4.7.1.2a.2. Table 3.3-5 item lla. 4.6.4, 4.3.3.5 Table 4.3-6.
4.7.1.2c.1. 4.7.1.2c.1. 4.7.1.2c.4. and 4.7.1.2c.2
' Modification Packaces MRF-21098. Temporary installation of an additional flow path for tF TD-EFW train for surveillance testing purposes and seismic installation for a local flow gauge.
MRF-21222. Temporary installation of pressure instrumentation to investigate
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pressure instrumentation. failures.
MRF-21514. Changes to the setpoints for the EFW high flow bistables.
Condition Evaluation Reoorts CER-96-0052. Turbine Driven Emergency Feedwater Pump Oscillation CER-96-0056. Taper Pin Found in Governor and was Determined to be Loose CER-96-0060. TDEFP Will Not Run at 2100-2300 RPM CER 96-0225. During Surveillance Test TDEFP Exceeded %s hasimum Allowed Speed CER-96-0484 Switch for "A" MDEFP has Experience Repeated Problems CER 97-0154 Install / Removal of Heise Gauge CER-97-0175. During STP for TDEFP, Test was Started From Wrong Train CER 97-0356. Check Valve Operability
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CER-97-0366. Measured Stroke for IFV 03531-EF is Greater Than Design
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CER-97-0369. Measu,ed Stroke for IFV 03541-EF is Greater Than Design
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CER 97-0371. During STP Valve XVB01035-EF Would Not Fully Open ER-97-0380. High Bearing Temperature on Inboard Bearing of XPP0021A CER-97-0418. "A" Train Relay Test for EF-Pump XPP 0008 Exceeded Quarterly Surveillance Requirement l
CER-97-0445. Failure to Perform CST N2 Purge Valve Test l
CER-97-0460. 1 of 4 Mounting Bolts was Found Loose on Motor Bracket l
CER-97-0540. 0-Ring Not Seated on Transmitter Cover CER-97-0788,- Steam Generator Sodium Level Increase CER-97-0898. When Removing Pedig m late on 4 inch Pipe. Grinder Cut Into Pipe CER-97-0942. STP-130.003 Showed Acceptance Criteria Not Met for Steam Generator A&B TDEFP Supply Stop/ Check Valve CER-98-0154 The TS Condensate Storage Tank Required Volume was Identified as non-conservative CER-98-0205. Frequency of Internal Pump Inspection Did Not Meet NRC Bulletin 88-04 Commitment CER-98-0259. Unanalyzed Condition for Equipment Qualification During a Secondary Line Break Outside Containment CER-98-0384 TDEFW Pump Technical Specification Surveillance Inadequacy Miscellaneous
Non-conformance Notice (NCN) #5464. Replacement of EFW flow post-accident monitoring instruments Updated Final Safety Analysis Report, through Amendment 98-01. February 1998.
System NRC Safety Evaluation Report on V. C. Summer Nuclear Station. NUREG-0717 and
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Supplements 1 through 5 SCE&G Design Basis Document " Emergency Feedwater System." Revision 10.
03/18/98 SCE&G Design Basis Document "Setpoint Bases." Revision 1. 03/02/98 SCE&G " Station Blackout Evaluation." GAI Report 2782. Revision 6. 05/27/92 NRC Safety Evaluation Report on Station Blackout Analysis, transmittal letter dated January 30. 1992, and associated Technical Evaluation Report by SAIC y
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NRC Information Notice No. 86-14 with Supplements 1 & 2. PWR Auxiliary Feedwater Pump Turbine Control Problems NRC information Notice No. 88-09, Reduced Reliability of Steam-Driven Auxiliary Feedwater Pumps Caused by Instability of Woodard PG-PL Type Governors l
NRC Information Notice No. 93-51. Repetitive Overspeed of Turbine Driven Auxiliary Feedwater Pumps WRC Information Notice No. 94-66 and Supplement 1. 0verspeed of Turbine Driven Pumps Cause by Governor Valve Stem Binding
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