IR 05000250/1998009
| ML17354B150 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 09/22/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML17354B149 | List: |
| References | |
| 50-250-98-09, 50-250-98-9, 50-251-98-09, 50-251-98-9, NUDOCS 9810300015 | |
| Download: ML17354B150 (60) | |
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION II
Docket Nos.:
License Nos.:
50-250 and 50-251 DPR-31 and DPR-41 Report No:
.
50-250/98-09 and 50-251/98-09 Licensee:
Facility:
Florida Power and Light Company Turkey Point Units 3 and 4 Location:
9760 S.
M. 344 Street Florida City. FL. 33035 Dates:
Team Leader:
July 27 - 31.
and August 10 -
14 '998 N. Merriweather, Senior Reactor Inspector Engineering Branch Division of Reactor Safety Inspectors:
K. Jabbour.
Turkey Point Project Manager.
NRR C. Jones.
Mechanical Engineering Consultant L. Moore. Reactor Inspector J.
Reyes, Resident Inspector (Sections MZ.Z/M8.2)
Approved by:
Kerry D. Landis, Chief Engineering Branch Division of Reactor Safety Enclosure 98i0300015 980922 PDR ADOCK 05000250
EXECUTIVE SUMMARY Turkey Point Units 3 and 4 NRC Inspection Report 50-250/98-09, 50-251/98-09 This inspection included a review of the licensee's calculations.
analysis, and other engineering documents that were used to support the Auxiliary Feedwater (AFW) system performance during normal and accident or abnormal conditions.
In addition to its independent inspection effort. the team reviewed the licensee's self assessment results.
The team's primary goal was to determine whether there was reasonable assurance that the AFW system would have been able to adequately perform its design safety functions, recent upgrades by the licensee notwithstanding.
While some testing deficiencies were identified, the team concluded that the system would have performed adequately.
In addition, the report includes a review of the Emergency Core Cooling System swap over and the
CFR 50.59 Program.
The report covered a
three week period of inspection in which two weeks were on site and one week was in the Region II office in Atlanta. Georgia.
Q~erati ons
~
Emergency and normal operating procedures were consistent with the design and licensing bases for the AFW system.
(Sections 03/E3. 1.3)
aintenance The plant material condition and housekeeping observed during the walkdown indicated that the auxiliary feedwater (AFW) equipment was being adequately maintained.
AFW piping and equipment was painted with few indications of corrosion. oil leaks or water leaks.
(Section MZ. I)
~.
Acceptance criteria in AFW system maintenance and surveillance procedures were generally consistent with the design and 1icensing
. basis.
One exception was the In-service Test (IST) procedure acceptance criteria for AFW secondary boundary outside containment isolation valves and Condensate Storage Tank (CST) discharge check'alves which did not meet Section XI requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code.
This deficiency was identified as a Non-cited Violation.
(Section H2.2)
~
Maintenance history indicated that the licensee had adequately resolved AFW equipment problems to maintain system reliability.
The team found.
'hat maintenance activity on the system was consistent with the design and licensing bases.
(Section M2.3)
En ineerin
~
The team determined that the licensee's safety evaluation program conforms to the requirements of 10 CFR 50;59 and is effective.
The program implementing procedures were generally of sufficient detail to ensure that proposed activities were processed in accordance with the regulation.
Overall quality of safety evaluations reviewed was good.
One exception was an inadequate safety review for a tempora'ry procedure
'hat placed the Unit 3 S/G Blowdown Heat Exchanger in Service on August 2.
1997.
This deficiency'as identified as an NCV. (Section E1.1)
The configuration management methods used in the licensee's engineering program are significantly enhanced by its design basis documentation.
As a result. the licensing basis for the AFW system is well understood and controlled.
(Section E1.2)
While the licensee's AFW self assessment did not uncover any significant functional deficiencies'ome calculation updates were needed.
(Section E2.1)
The team found that the CST level instrumentation was consistent with the design and licensing basis for the AFW system.
(Section E2.2. 1)
The calculation that sized the TOL relay heaters for AFW valve actuators was consistent with the engineering design guidance and the inputs and assumptions used in the calculation were deemed to be adequate.
(Section E2.2.2)
The calculations performed to evaluate reduced voltage starting of the AFW NOVs were found to be complete and technically adequate.
(Section E2.2.3)
Engineering support for AFW system modifications was good. With one exception.
documentation of AFW system modifications was adequate.
The exception was a modification installed in 1984 and was not representative of current design activities.
(Sections E3.2 and E3.2. 1)
Engineering support for assessing condition reports was good.
Technical documentation was comprehensive and was systematically developed and.
clearly presented.
Condition Report 98-276 for the February 16.
1998 Unit 3 trip (on loss of turbine and steam leak) was a good example of effective engineering. support.
(Sections E3.3)
The licensee's self assessment of the AFW system was very detailed and thorough and is considered a strength.
The assessment identified weaknesses in the IST Program and operating procedures for safety-related pumps at minimum flow conditions.
The assessment also resulted in several discrepancies in plant documentation being identified and corrected.
(Section E7)
The licensee was responsive to and in conformance with information and requirements from external industry and regulatory sources.
The inspection team determined that the licensee's evaluations and implementation of appropriate precautions and verifications in plant procedures were adequate.
(Section E8. 1)
Design documentation, emergency operating and calibration procedures.
and alarm set points demonstrated that adequate NPSH was available for the ECCS pumps during the injection and recirculation phases of safety injection.
(Section E8.2)
Introduction Re ort Details The primary objective of this Safety System Engineering Inspection (SSEI) was to determine if the Auxiliary Feedwater (AFW) System was designed, operated.
and maintained consistent with the design and licensing bases.
A secondary objective was to determine the quality of 10 CFR 50.59 safety evaluations performed by the licensee in support of engineering modifications performed on the selected system.
This inspection also reviewed the Emergency Core Cooling System (ECCS) design and Refueling Water Storage Tank (RWST) instrumentation to assess the adequacy of the swap over initiation point with respect to assuring adequate ECCS pump conditions are maintained.
In addition to its independent inspection effort. the team reviewed the licensee's self assessment results.
The inspection was performed by a team of inspectors which included a Team Leader, one Region II inspector.
the Turkey Point NRR Project Manager.
and one mechanical design contractor.
Also the Resident Inspector assisted the team in the followup and closeout of'reviously identified open items.
During the course of the inspection.
both Units 3 and 4 were operating at full power.
'
I. 0 erations 03.
Operations Procedures and Documentation a.
Ins ection Sco e
93809 The inspectors reviewed the normal and emergency operating procedures to determine if'uxiliary Feedwater System (AFW) operation was consistent with the design and licensing bases.
b.
Observations and Findin s Overall, emergency and operating procedures were consistent with the AFM system design and licensing bases.
During the pre-inspection system self assessment the licensee identified a deficiency in the emergency operating procedures (EOPs).
The AFW'pump mini-flow line capacity of 10 gallons per minute (gpm) was insufficient for extended pump operation at low flow conditions and this was not noted in the EOPs.
The design basis requires the pumps to be operated up to 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br /> during a design base accident.
The issue was identified in condition report (CR) 98-1069.
dated July 14.
1998.
A safety evaluation was performed that determined the AFM low flow restrictions which were then documented in a night order and included in revisions to the EQPs.
The EOPs were revised to direct the operators to secure AFW pump operation.
when not needed.
during accident conditions.
The inspectors noted that there was no installed instrumentation for monitoring pump and turbine bearing or oil temperatures during pump operation.
The system operating procedures and surveillance procedures did not require or provide for temperature monitoring.
The vendor manuals for both the pump and Terry turbine specified limitations on
these parameters.
The pump vendor manual.
VTH No. A2223. Ingersol Rand, Installation, Operation.
and Maintenance of CNTAH Pump.
dated March 24.
1995, stated the pump should be stopped if the bearing oil temperature exceeds 160 degrees Fahrenheit (F).
The Terry 'Turbine vendor manual, VTH No.
V000126, Terry Turbine Corp.
AFW Pump Turbine. dated June 17.
1998. stated that the maximum temperature rise across the bearing should not exceed 50 degrees F,
and oil temperature leaving the bearing should not exceed 180 degrees F.
The AFW pump Inservice Test (IST) procedure.
0-OSP-075. 11. dated March l. 1994.
operates the pump for approximately one to two hours.
The monitored parameters do not include oil or bearing temperatures.
The licensee had monitored bearing temperatures with portable instrumentation until approximately 1990.. This was discontinued when the requirement for bearing temperature monitoring was deleted from the ASME Code IST recommended parameters in favor of increased vibration monitoring to detect degraded bearings.
The test data from before 1990 indicated that operating pump and turbine bearing temperat'ures were stable below the vendor specified limits.
Only one bearing fai lure has occurred and this was due to improper installation which was identified during the post maintenance test.
Previous test data and equipment history indicate that pump and turbine reliability has not been impacted due to not monitoring bearing or lube 'oil temperatures.
Conclusion Emergency and operating procedures were consistent with the AFW system design and licensing bases.
The licensee identified and corrected an EOP deficiency related to limitations on AFW pump low flow operation.
II. Maintenance Maintenance and Material Condition of Facilities'nd Equipment Material Condition Walkdowns Ins ection Sco e
93809 The inspectors performed general field walkdowns to assess the material condition of the AFW system.
Observations Findin s.
and Conclusions The AFW system equipment was located on the open turbine deck areas which were exposed to the environment.
The general material condition of the AFW equipment was good.
Housekeeping in the equipment areas was good.
Valves in particular were well maintained.
Piping and equipment was painted with few indications of corrosion, oil leaks or water leak Maintenance and Surveillance Procedures Ins ection Sco e
93809 The inspectors reviewed maintenance and surveillance procedures to determine if acceptance criteria were consistent with the design and licensing bases.
This rev'iew included the licensee's incorporation of'FW equipment into the Inservice Test (IST) Program.
Observations Findin s and Conclusions The inspectors reviewed those procedures listed in the references to this report and reviewed a sample of completed surveillance tests and maintenance work orders.
The post maintenance test acceptance criteria specified in the maintenance procedures was consistent with the design parameters specified in the design and licensing bases.
Acceptance criteria in the surveillance tests were consistent with the design and licensing bases.
The licensee's pre-inspection system self assessment identified deficiencies in the incorporation of AFW system components into the IST program.
These included the'ondensate Storage Tank (CST) discharge check valve testing which did not verify full flow accident conditions as required by the American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel (B&PV) Code,Section XI and ASME/ANSI-1987.
Operation and Maintenance of Nuclear Power Plants.
The valves were tested at a flow less than the analyzed accident i'low value of 467 gpm.
The tested flow of 390 gpm was based on the Technical Specification individual train operability requirement rather than the worst case analyzed condition.
The item was identified on CR 98-0566 and CR 98-0761.
dated June 19, 1998 and May 7.
1998 respectively.
and LER 50-'50,251/98-003 was submitted.
The licensee promptly tested the check valves to the greater flow capacity and revised the acceptance criteria in the IST procedures.
An additional IST deficiency identified by the licensee was that the AFW secondary boundary outside containment isolation valves were not tested as specified by ASME B&PY Section XI and ASME/ANSI-1987.
The train one boundary isolation valves were check valves CV-140.
-240.
and -340 which the code requires to be reverse flow tested.
The. train two isolation valves were flow control valves FCV-2816.
-2817.
- 2818.
-2831.
- 2832.
- and-2830 for which the code required closure time stroke testing.
The previous IST testing of these valves only performed full flow testing and open stroke time testing.
This discrepancy was documented in CR 98-1070, dated July 14, 1998.
The operability evaluation documented in CR 98'-1070 determined that adequate containment boundary isolation barrier capability was provided as demonstrated by system surveillance testing and open stroke testing of both trains'low control valves.
The corrective actions included revision to'he IST program to include closure stroke time testing of both trains'CVs and
M2.3 a proposed revision to the Final Safety Analysis Report (FSAR)
designating the FCVs as the outside containment secondary boundary isolation valves for the AFW system.
The IST program deficiencies discussed above are a violation of regulatory requirements incorporated in Technical specification 4.0.5 which requi res the licensee to implement an Inservice test program consistent with ASME BSPV Section XI.
Corrective actions for this self-identified deficiency were adequate.
In accordance with Section VII.B.1 of the Enforcement Policy. this non-willful..non-repetitive violation is identified as Non-cited Violation (NCV) 50-250.251/98-09-01, Inadequate IST Criteria for AFW Valves.
Conclusion Acceptance criteria in AFW system maintenance and surveillance procedures were consistent with the design and licensing basis.
One exception was the IST procedure acceptance criteria for AFW secondary boundary containment isolation valves and the CST discharge check valves which were not consistent with ASME BSPV Section XI.
This deficiency was identified as an NCV.
Maintenance Histor Ins ection Sco e
93809 The inspectors reviewed the maintenance history of AFW equipment for the previous five year period to determine if performance of corrective maintenance was appropriate to maintain equipment reliability consistent with the design and licensing bases.
Observations and Findin s
. The maintenance history indicated that system equipment problems were appropriately resolved.
Condition reports and corrective maintenance work orders'indicated that equipment problems were adequately evaluated by engineering and resolved.
For example.
an industry issue related to inadvertent over speed tripping of Terry turbines due to inappropriate material of the governor valve stem was identified at Turkey Point in February.
1996 and documented on CR 95-135.
The corrective action appropriately addressed all AFW pump turbines and a program was implemented to periodically cycle the valves and replace the susceptible stem material with the approved substitute.
A more recent related industry issue involving an interact~on with the new valve stem material and the stem carbon washers.
which could result in binding and subsequent speed control problems.
was analyzed for applicability 'at Turkey point.
The analysis determined this was not a concern at -this station.
Another example of resolution of equipment problems was the modification to remove the electronic trip function of the turbine which was causing unnecessary trips of the turbine.
The modification improved the
M8
reliability of the AFW pumps.
The electronic trip components were vulnerable to the high temperature environment.
This was a redundant trip functions the mechanical trip mechanism provided the remaining over speed trip protection.
Conclusion Maintenance history indicated that the equipment problems had been adequately resolved to maintain system reliability.
Maintenance activity has been consistent with the design and licensing bases.
Miscellaneous Maintenance Issues (92903)"
N8.1 Closed IFI 50-250 251/98-04-02: Auxiliar feedwater Pum Tri and Throttl e T8T Val ve Fai lures This item addressed several failures of TST valves to operate properly during Inservice surveillance tests which occurred between Oecember, 1997 and March.
1998.
On two occasions the valves failed to open.
following a manual trip test.
On one occasion a T8T valve failed to shut when mechanically tripped.
Root cause investigations did not identify a specific cause for the opening failures.
The cause appeared to be related to the electrical control circuit.
Relays were inspected and replaced.
In October.
1998. the control room 'panel switch is to be replaced on the B AFW pump.
No additional open fai 1ures have occurred.
The cause of the close failure appeared to be a time dependent phenomena which resulted in increased friction between mechanical parts, of the trip mechanism and degraded trip spring strength.
The mechanical parts
.
were cleaned and the spring replaced and the test frequency was increased from monthly to bi-monthly.
The inspectors reviewed the test results from the increased frequency exercising of the valves and noted that no additional closure fai lures have occurred.
The performance of the T&T valves since the increased testing was initiated ~ndicated that the TKT valve operating problem had been resolved.
N8.2 Closed LER 50-250 251/98-003:
AFW Check Valves Inade uate Surveillance This item addressed the inadequate IST of the AFW check valves on the discharge of the Condensate Storage Tank.
The valves were not full flow tested as required by ASNE Boiler and Pressure Vessel Code Section NI.
The valves were tested at a flow of 390 gpm consistent with the Technical Specification for single train flow from one CST.
The worst case analyzed condition was approximately 467 gpm for one train from both CSTs.
This was identified. during a licensee self assessment and corrective actions included performance of the full flow test at 467 gpm.
This issue was addressed in paragraph H2.2 of this repor III.
En ineerin Conduct of Engineering Safet Evaluations Ins ection Sco e
93809/37001 The inspectors reviewed the adequacy of the licensee's controls for making changes to the plant in accordance with 10 CFR 50.59.
The team reviewed the major system modifications completed since licensing (see Section E2, below) and assessed the quality of the associated safety evaluations.
This review included design packages implemented in response to industry experience.
including changes intended to enhance system redundancy and upgrade components such as pumps and control valves.
In addition to significant hardware upgrades, Turkey Point engineers reevaluated the AFW system for a thermal uprate project.
Thus, there has been a significant amount of engineering effort expended on this system since licensing'esulting in many engineering safety evaluations in modification packages.
Observations and Findin s The licensee was able to retrieve comprehensive record documentation for all of the modifications requested.
which spanned about 18 years.
The team reviewed this documentation and identified no safety ev'aluation concerns for the AFW system.
The team found that the safety evaluation documentation has improved over time in that the more recent modifications included more detai 1 in the safety evaluation area.
Detailed explanations of safety considerations were much more likely to be found in the more recent modification packages.
A more generic evaluation of 10 CFR 50.59 documentation was performed in conjunction with this SSEI.
The generic evaluation included a review of procedures that support the design change process and a random selection of 12 safety evaluations (SEs) associated with non-AFW specific plant modifications and procedure changes.
.The inspectors also interviewed cognizant licensee personnel and attended two Plant Nuclear Safety Committee (PNSC) meetings on July 28 and 30 '998.
Only one deficiency was noted.
involving the safety review for a Temporary Procedure (TP)97-041 that placed the Unit 3 S/G Blowdown Heat Exchanger in service on August
~
1997.
During the performance of this procedure, operations personnel recognized that the procedure utilized a
flowpath that bypassed a high radiation isolation trip documented in the FSAR without performing a
CFR 50.59 safety evaluation for the change.
The procedure flowpath would have allowed the S/G Blowdown Recovery Tank to discharge to a storm drain via the Condensate Recovery Tank if a S/G Tube Leak/Rupture event had occurred while this procedure was in effect.
Implementation of this TP lasted from 1900 to 2320 on August 2.
1997.
At that time Operations recognized the problem and the TP was
El. 2 terminated.
The problem was documented on Condition Report 97-1184.
dated 8/2/97 for root cause determination and corrective actions.
T The root cause of the event was determined to be an inadequate
CFR 50.59 safety review of the TP.
The preparerlreviewer incorrectly concluded in the screening of the TP, that 10 CFR 50.59 was not applicable and subsequent program reviews failed to identify their error.
The licensee's response to a screening question
"Does this request affect the facility or procedures as described in the FSAR" should have been, answered
"Yes" in lieu of'he licensee's answer as
"No".
The corrective actions implemented by the licensee included providing training on this event and its causes to all those plant personnel who perform 10 CFR 50.59 screenings.
and reducing the number of qualified screeners and reviewers.
To evaluate the safety consequences of the event.
the licensee performed an analysis assuming maximum concentration in the Blowdown Tank and concluded that the offsite dose would only have been a fraction of the
CFR ZO limit (i.e.. approximately 5.38E-Z} if a steam generator tube leak or rupture event had occurred while this TP was in effect.
The corrective actions were assessed as adequate by the team, and the licensee's safety evaluation program and associated training were improved as a result.
The licensee's failure to perform a safety evaluation for TP 97-041 is a
violation of 10 CFR 50.59.
However. this non-willful. non-repetitive.
licensee identified and corrected violation is being treated as a
Non-'ited Violation (NCV). consistent with Section VII.B.1 of the NRC Enforcement Policy.
This item is identified as NCV 50-250.251/98-09-02.
Failure to perform a 50.59 Safety Evaluation for Temporary Procedure
.
Conclusion The team determined that the licensee's Safety Evaluation program conforms to the requirements of 10 CFR 50.59 and is effective.
The program implementing procedures were generally of sufficient detail to ensure that proposed activities were processed in accordance with the regulation.
Overall quality of safety evaluations reviewed was good.
'ne exception was an inadequate safety review for a temporary procedure that placed the Unit 3 S/G Blowdown Heat Exchanger in Service on
'ugust 2.
1997.
This deficiency was identified as an NCV.
En ineerin Su ort of Confi uration Mana ement Ins ection Sco e
93809 The team reviewed the licensee's design basis documentation for the AFM system.
This review included design information at the system levels the component level.
and related accident analyses design basis document b.
Observations and Findin s During its safety system functional inspection. (SSFI) performed in 1985.
the NRC found that the design basis information available at the Turkey Point Plant was not adequate to support configuration management in the engineering area.
Since that time, the licensee has reconstituted its design basis for 18 safety related systems.
The licensee provided the team with a detailed and comprehensive set of design basis documents (DBD) for the AFM system and the system components that provide the plant engineering staff with-an effective means of managing the system configuration changes.
Also, the licensee provided the team with the Turkey Point Accident Analysis Basis Document (AABD).
These design basis and related accident analysis documents are updated annually, so the inspection team was also provided information regarding the planned updates.
The licensee showed that the plant's design calculation management program includes the use of DBD historical information to guide the design process in identifying the relevant design basis calculations.
Similarly, the DBD provides engineers with extensive information on previous system modifications. including detailed discussions of the engineering rationale associated with each modification.
These discussions provide a clear technical context within which the plant engineers are able to address new modifications and emergent issues.
In view of the technical detail and descriptive historical qualities of the DBD, the inspection team determined that the licensee's overall configuration management program was very good.
In particular, the licensee's methods are effective in resolving some of the complex
'anagement and documentation problems that are often found in this area.
Nevertheless.
both the licensee and the inspection team found that the normal flow of information across organizational interfaces could be improved.
c E2 E2.1 Conclusion The configuration management methods used in the licensee's engineering program are significantly enhanced by its design basis documentation.
As a result. the licensing basis for the AFM system is well understood and controlled.
Engineering Support of Facilities and Equipment Mechanical Desi n Review Ins ection Sco e
93809 The team selected 34 of the more than 500 design basis calculations currently classified by the licensee as active or otherwise applicable f'r use in the design engineering process.
The calculation sample included all of the important system functional areas, including the
system hydraulic flow model, net positive suction head (NPSH)
~
and condensate storage tank (CST) level uncertainties.
Observations and Findings Although the team initially had concerns regarding some of the key calculations, the licensee provided additional calculations and revisions to calculations that resulted from the licensee's self assessment and from a separate independent assessment contractor.
The licensee recorded the seven discrepancies found in calculations, the DBD. and drawings in Condition Report 98-1056.
The new information rovided to the team included revised assessments of NPSH
~ the system ydraulic model.
and CST level uncertainty.
In its evaluation of these self assessment related discrepancies and calculation changes, the team determined that no significant issues were involved.
Although the functional performance margins for the system were found to be adequate, these discrepancies support the team's overall programmatic concern with how the licensee manages its otherwise superior design basis documentation.
The team notes that the condition report did not pursue a programmatic root cause or the generic implications for the deficiencies identified.
Nevertheless'he licensee's effectiveness in this area was very good.
Conclusion While the licensee's AFW self assessment did not uncover any significant functional deficiencies, some calculation updates were needed.
E2.2 Electr ical/I8C Desi n Review E2.2.1 Review of CST Level Instruments Ins ection Sco e
93809 The team assessed the adequacy of the CST level instrumentation.
Observations and Findin s The team found that the CST level instrumentation met the design and licensing basis for the AFW system.
The team also noted that applicable instrument calibration procedures included appropriate design inputs from the CST level instrumentation scaling documents.
The team verified by field walkdown inspection that the as-built instrument installations were consistent with design.
Conclusion The CST level instrumentation was consistent with the design and licensing basis for the AFW syste E2.2.2. MOV Thermal Overload TOL Heater Sizin Calculations a.
Ins ection Sco e
93809 The team assessed the adequacy of the sizing for the thermal overload (TOL) relay heaters for the AFW system motor operated valves by review of design calculations.
Observations Findin s and Conclusions The calculation that sized the TOL relay heaters for AFW valve actuators was consistent with the engineering design guidance and the inputs and assumptions used in the calculation were deemed to be adequate.
E2.2.3 MOV Reduced Volta e Startin Anal sis Ins ection Sco e
93809 Those electrical calculations shown in the references to this report were reviewed to assess the adequacy of the AFW MOVs for reduced voltage starting.
Observations Findin s and Conclusions The calculations performed to evaluate reduced voltage starting of the AFW MOVs were found to be complete and technically adequate.
The inputs and assumptions used in the calculations were reasonable.
The calculations showed that at reduced starting voltages the actuators would still be capable of producing the required stem thrust and torque required to operate the valve under worst case design basis conditions.
Engineering Procedures and Documentation E3. 1 Calculations
'
E3.1.1 Review of AFW Calculation Control Pro ram And AFW Calculation Ade uacv a.
Ins ection Sco e
93809 The,inspection team selected about 30 AFW system design calculations from the licensee's listing of more than 500 such calculations.
The team later determined that the licensee's self assessment had included all of the calculations reviewed 'by the team.
After the team had requested additional information on the calculations, the licensee provided information regarding calculation changes that had resulted from the licensee's own self assessment.
Newly developed calculations and revisions to older calculations were provided to the team and were reviewe b.
Observations and Findin s In spite of the new information, the inspection. team identified several out of date assumptions and results in the calculations reviewed and questioned the licensee's calculation management program.
The team was particularly concerned that calculations were not being superseded when required by the applicable plant procedure (fNG QI 1.5)
on calculations.
The procedure specifies and provides guidance for the superseding and voiding calculations, but neither of these categories was used much for.
the AFW system at the Turkey Point Plant.
Although most of the system had been redesigned and rebuilt-over the years.
the calculation database only included about four voided calculations and five superseded calculations out of more than 500 calculations.
Rather than superseding calculatidns and having to deal with the documentation cross-reference updating complexities that would result, plant engineers have relied on the DBD historical descriptions and on the plant documentation database to identify relevant design information.
They have also used an informal and undocumented
"co'nfi rmed" category in which new calculations validate older calculations'esults under new conditions or assumptions.
That is; rather than superseding or revising previous calculations that include out-of-date assumptions and other conditions.
a new calculation may be developed and issued that conclude the old calculation's results were confirmed.
The revised assumptions or conditions addressed in the new calculation did not result in a need to modify the previous calculation's results.
The licensee explained that this approach avoids mistakes by engineers
'ecause they are required to review both the old and new calculations.
basing any new design work on a more detailed understanding of al,l of the historical system design changes.
Thus. for example.
there are at
. least six calculations related to condensate storage tank volume that an engineer would have to consider..
Not all of these calculations are described in the DBD. but they are retrievable from the licensee's documentation database.
The licensee also explained that the DBD provides a detailed historical record and explanations of the key calculations for the AFW system.
Using the DBD, an engineer can readily discern the current system configuration as well the previous configurations and their design rationale and limitations.
The team verified this position in its review of the DBD and the supporting documentation databases.
The DBD is comprehensive in including historical information and descriptions of the key system calculations and modifications.
However.
the licensee in its recent self assessment of the AFM system found and-corrected several calculation issues.
In its review of these licensee identified issues.
the inspection team determined that none of the issues or updates were operationally significant due to the design safety margins in the AFW syste.
Horeover, based on experience, the team determined that other administrative methods would not necessarily have avoided the deficiencies identified in the self assessment process.
Indeed, it is likely that the DBD facilitated the self assessment review and the resulting refinements of the system calculations.
No significant functional deficiencies were found in the calculations.
although the inspection team and the licensee found that there was inadequate design analysis regarding one modification.
This modification involved providing cooling water from the pump second stage to the lube oil cooler and it is discussed in Section E3.2.1 below.
Conclusion
.Overall, AFM system design calculation documentation for AFM was good.
Plant engineering procedures do not fully describe the documentation management process.
E3.1.2 Self assessment of Calculations Ins ection Sco e
93809 The team reviewed the licensee's self assessment of engineering calculations'ocusing on what was changed as a result of the self assessment.
Observations and Findin s P
Some key system calculations were either written or revised.
For example.
as a result of the self assessment.
the licensee issued a new net positive suction head calculation (PTN-BFSN-98-0010) for the AFM pumps at a potentially limiting case under current performance requirements.. This calculation is in addition to the 1985 Bechtel.
calculation M08-462-01 done following extensive changes to the system
'piping.
The primary va'lue of this latest calculation is that it uses the latest flow requi rements and reflects an extreme case that exceeds design and operational requirements for CST level.
Thus, it would supersede all less demanding NPSH cases.
The licensee indicated that the older calculation was not being formally superseded by the new one.
Under the licensee's calculation program. older calculations may be retained in a historical status rather than being superseded.
Plant engineers perceived this as being advantageous'lthough the practical distinction was not evident to the inspection team.
Also, it is apparent that calculations performed by other design organizations a'e not considered subject to 'later updating and revision by the plant engineering staff.
This results in additional complexity and the potential for confusion and error l3
In its review of the calculation changes resulting from the self assessment.
the team determined that no actual. operational or safety issues were identified that would indicate the AFW system was unable to perform its requi red safety functions in 'a reliabl'e manner.
Conclusion To the extent that revisions were required and based on how this was done.
the licensee's self assessment indicates'that there would be some value in extrapolating this experience to other systems and clarifying plant procedural guidance regarding calculations.
E3.1.3 Im lementation of Desi n Data in Plant Procedures Ins ection Sco e
93809 The team verified that the plant emergency and abnormal operating procedures were adequate to ensure the AFW system design capabilities would be used effectively in 'response to applicable plant conditions.
Also, the licensee's LERs associated with AFW system were reviewed.
which helped the inspection team verify the performance of the system automatic functions under emergency conditions.
Observations and Findin s Normal operating procedures (e.g..
3-OP-075) were in place that were consistent with the design capabilities of the AFW system.
and the procedures implement design requirements in a conservative manner.
For.
example.
procedure prerequisites require the CST to be available with at least 210.000 gallons available.
as compared with the 191, 100 minimum required in design calculations.
Operational requirements and constraints such as not operating a
pump against a shutoff head for.
extended periods due to low pump recirculation flow were also conservatively incorporated.
although this was a recent change as a
result of the self assessment finding and.,is discussed in detail in Section 03 of this report.
The AFW system was designed to be actuated automatically in response to several different emergency conditions and requires no operator actions until well into an event. usually 30 minutes.
This automatic actuation capability has been demonstrated and
'ocumented in tests and in several LERs in which the system started automatically on emergency conditions such as loss of main feedwater.
In addition to monthly reliability checks that ensure the system would be able to provide sufficient secondary side cooling water. the operators have CST level alarms that ensure the system has an adequate volume of water available.
The inspection team reviewed the applicable alarm response procedure (3-ARP-097.CR)
and found it to be adequate.
Conclusion
The AFM system would start automatically and provide adequate secondary cooling water flow to steam generators under emergency conditions, consistent with its design parameters and the technical specifications.
Nodifications Ins ection Sco e
93809 The =inspection team requested documentation on 20 system modifications and reviewed 12 that were considered to be most significant.
Observations and Findin s The. licensee was able to retrieve comprehensive documentat'ion for all of the modifications requested.
The team's review of these modifications resulted in one concern regarding the engineering analysis of the adequacy of cooling water f'r the lube oil cooler.
A similar issue was identified by the licensee in its self assessment.
The sample of AFW modification packages extended back about 20 years.
Significant improvements were noted in engineering documentation over this period.
The inspection team noted that older modifications of similar importance had engineering evaluations that were comparatively sketchy and under-developed.
For example.
PCN 80-55 upgraded AFM system flow control and initiation redundancy to meet NUREG 0578 requirements.
but the inspection team was not able to find any engineering evaluation in the approximately 1.000 record pages for this modifica'tion.
Similarly, PCH 80-78 modified the
'FM steam supply valves to include QC as well as AC motors.
Out of, the 1.000 pages of record documentation.
the only engineering evaluation was that DC power supplies are more reliable than AC power supplies.
Another example of this is PCH 79-124.
AFW Pump Control Hodification.
In contrast.
PCH 95-141 raised the CST low level alarm and included several pages of engineering discussion.
Another recent example of good engineering support is PCH 97-033 for the removal of the AFW pump turbine electrical over speed trip. in which every engineering and safety aspect of modification was described in great detail.
Conclusion Overall. the engineering and safety evaluations provided for modification packages has improved at this plant since licensing.
Engineering support for modifications was good.
Design related information was provided in detailed descriptions'ystematic checklists.
and fully developed technical evaluations and safety related justifications under
CFR 50.5 '
E3.2.1 Review of Plant Chan e Modification PCM 78-104 a.
Ins ection Sco e
93809 The inspectors reviewed AFW modifications to assess the quality of design documentation and determine if'ystem modifications were consistent with the design and licensing bases.
Findin s and Observations The inspectors noted one example of poor modification documentation, for AFW system design changes.
Plant change modification (PCM)78-104.
AFW Pump Lube Oil Cooler, installed March 31.
1984, changed the pump and turbine lube oil cooler cooling water supply from a non-safety related service water to the second stage discharge from the pump which was safety related.
There were no design inputs included in the modification documentation provided to the inspectors.
Such design inputs would include equipment heat loads, bearing temperature limits.
temperature ranges of lube oil and cooling water which could be used for sizing of, piping and tlow requirements.
The modification addressed the need for an in-line orifice to protect the cooler from over pressure but did not give the parameters for the orifice.
There was a minimum eight.
gpm flow referenced in connection with the orifice.
The IST data sheets.
which included bearing temperature's until 1990.
demonstrated that bearing temperatures were stable below the vendor recommended limits.
This indicated that adequate cooling water tlow was provided from this modification and that there was not a safety concern with respect to the pump design.
The inspectors'oncern was the poor quality of the design documentation=related to this modi fication.
Additional review of installed AFW modi fications demonstrated that the more recent modification documentation was of considerably improved quality.
c.
Conclusion With one exception.
documentation of system modifications was adequate.
The exception was a modification installed in 1984 and was not representative of current design activities.
E3.3 Review of Condition Re orts Ins ection Sco e
93809 The inspection team assessed the licensee's condition reports (CR)
associated with the AFW system.
Observations and Findin s The team assessed the technical adequacy of 35 condition reports related to this system, including the 12 CRs that resulted from the licensee's self assessment.
The team's review of CRs resulted in finding one recent potential technical problem that had been identified but not
resolved.
The licensee resolved this concern during the assessment and determined that the potential technical problem was not valid.
Conclusion E7 Engineering support for assessing condition reports was good.
Technical documentation was comprehensive and was systematically developed and clearly presented.
Condition Report 98-276 for the February 16.
1998 Unit 3 trip (on loss of turbine and steam leak) was a good example of engineering support.
Quality Assurance in Engineering Activities
a.
Ins ection Sco e
93809 The team reviewed the findings from the licensee's pre-inspection self assessment report on the AFW system.
b.
Observations and Findin s
The team noted that the self assessment was a comprehensive and detai
review of the design, testing.
operation.
and maintenance of the AFW system.
The report was clear and concise.
A total-of 55 findings were documented in the report.
Of those.
36 required revisions to plant documents
-
17 were classified as enhancements.
and 19 were identified as discrepancy corrective actions.
Weaknesses were noted in the IST program and in operating procedures for safety-related pumps during minimum flow conditions.
Overall. the assessment resulted in twelve (12) Condition reports and one Licensee Event Report (LER) 98-03) being.
identified.
The LER is discussed in Section H8.2 of this report.
The remaining issues had minor safety significance or were program enhancements.
The corrective actions for both the assessment findings and condition reports were reviewed by the team and found to be acceptable.
One Non-cited Violation was identified for inadequate IST criteria for AFW valves and it is discussed in Section H2.2 of this report.
The overall conclusion from the licensee's self assessment was that the AFW system was designed.
operated.
tested, and maintained in conformance with the licensing and design bases requirements.
Although.
some discrepancies in the FSAR, OBD. and calculations were identified. these discrepancies did not invalidate the overall conclusion of the assessment.
c.
Conclusion The licensee's pre-inspection self assessment of the AFW system was thorough in its review of engineering documentation for the system and is considered a strength.
The assessment resulted in weaknesses being identified in the IST Program and operation of safety related pumps at minimum flow conditions.
It also resulted in several discrepancies in plant documentation being identified and correcte ES E8.1 Miscellaneous Engineering Issues Res onse to Generic Communications and Industry.O eratin Ex erience Ins ection Sco e
93809 E8.2 b.
The team evaluated the licensee's response to various NRC generic communications and industry operating experience information related to the AFW system.
Observations and Findin s The licensee's self assessment.
provided sufficient information and references to applicable documentation on the requirements or issues raised by external communications.
The team reviewed samples of this documentation and determined the AFW system complies with those that are applicable to the Turkey Point Plant.
Moreover. the system engineer was found to have signiAcant technical experience related to the AFW system based on his previous industry experience.
The team reviewed several specific issues such as the potential for pump steam binding (NRC IE Bulletin 85-01),to determine whether the licensee's action or response was appropriate.
Nevertheless, the licensee determined in its self assessment that improvements were needed in the area of addressing external communications.
Conclusion The licensee was responsive to and in conformance with informati'on and requirements from external industry and regulatory sources.
The inspection team determined that the licensee's evaluations and implementation of" appropriate precautions and verifications in plant procedures were adequate.
Refuelin Water Stora e Tank RWST Level Instrumentation and Emer enc Core Coolin S stem ECCS Swa Over Ins ection Sco e
93809 The inspectors reviewed the ECCS design and RWST instrumentation to assess the adequacy of the swap over initiation point with respect to assuring adequate ECCS pump conditions are maintained.,
Observations and Findin s Swap over of ECCS from the RWST injection mode to the containment sump recirculation mode is manually initiated in accordance with the emergency operating procedures.
EOP 3/4-EOP-E-1.
Loss of Reactor or Secondary Coolant.
dated June 23.
1997.
and 3/4-EOP-ES-1.3.
Transfer to Cold Leg Recirculation.
dated November 18.
1997.
The first phase of the swap over was initiated at the RWST low level alarm set point of 155.000 gallons and the final phase completed at the low-low alarm set point of 60,000 gallon The critical level for ECCS pump NPSH considerations was the low level alarm set point of 155,000 gallons which was based on the most limiting NPSH requi rement at maximum safeguards pump configuration, which was for the containment spray pumps.
The RWST low level alarm set point development included two deficiencies which the licensee identified in the pre-inspection self assessment.
The first was in the conversion of the minimum requi red water height in the NPSH analysis in feet to the RWST level measurement parameter in gallons.
The conversion of the requi red 22 feet of water in the tank to 155,000 gallons was approximate and was not clearly documented.
The 155,000 gallons was measured from the elevation of the level transmitters at 13 to 16 inches above the tank bottom rather than from the bottom as indicated in the NPSH analysis.
This was a conservative deficiency'hat added approximately one foot to the NPSH Margin.
The instrument loop calibration did not include instrument uncertainty due to density variation associated with the tolerance allowed in Boron concentration and temperature of the RWST.
This was a negati ve impact on the NPSH available of approximately 0.75 feet.
The net impact of the deficiencies was a slight positive margin increase.
The licensee's corrective actions included revision of the uncertainty calculation, PTN-BFJI-94-006.
RWST Level Uncertainty Determination.
dated July 15, 1998 and evaluation of the impact of the deficiencies on the NPSH analysis.
procedures and calibration.
Additionally. the licensee determined there was additional margin due to containment conditions at the time of the swap over.
The Containment pressure profile for LOCA indicated a Containment pressure of 30 psig at the time when the RWST level is at 155.000 gallon, which reduced the NPSH required by 4:5 feet.
c.
Conclusion Design documentation.
emergency operating and calibration procedures.
and alarm set points demonstrated that adequate NPSH was available for the ECCS pumps during the injection and recirculation phases of safety injection.
Exit Meeting Summary V.
MANAGEMENT MEETINGS The Team Leader discussed the progress of the inspection with the licensee representatives on a daily basis and presented the results to members of licensee management and staff at the conclusion of the inspection on August 14, 1998.
The licensee did not have any dissenting comments at the meeting.
The licensee acknowledged the findings presente Licensee
PARTIAL LIST OF PERSONS CONTACTED M. Chambers'echanical System Engineer R. Earl, Quality Assurance Supervisor W. Haley. Operations Supervisor 0.
Hanek, Licensing Engineer R. Hovey. Site Vice-President D. Jernigan, Plant (general Manager H. Jurmain, Maintenance Manager
- Acting M. Lacal, Training Manager J.
LaDuca, Mechanical Design Engineer W. Parker.
SSEI Project Engineer R.
Rose, Work Control Manager W. Skelly. Plant Engineering Manager E.
Thompson. Site Engineering Manager D. Tomaszewski, Systems Engineering Manager J. Trejo. Health Physics/Chemistry Supervisor Other licensee employees contacted included engineers.
technicians.
operators.
and electricians.
NRC:
F.
Hebdon, Director, Project Directorate II-3, NRR K. Landis. Chief. Engineering Branch, Division of Reactor Safety INSPECTION PROCEDURES USED
,.
IP 37001:
CFR 50.59 Safety Evaluation Program IP 92903:
Follow-up - Engineering IP 93809:
Safety System Engineering Inspection ITEMS OPENED, CLOSED, AND DISCUSSED
~oened
~T e
Item Number NCV 50-250,251/98-09-01 NCV 50-250.251/98-09-02 Descri tion and Reference Inadequate IST Criteria for AFW Valves (Section MZ.2)'.
Failure to perform a required 10 CFR,50.59 Safety Evaluation for a Temporary Procedure
[TP 97-041] (Section El. 1).
,~
Closed
~T e
Item Number
Description and Reference IFI 50-250.251/98-04-02 AFW Pump Trip and Throttle (T8T) Valve Failures (Section N8.1).
LER 50-250.251/98-003 AFW Check Valves Inadequate Surveillance (Section M8.2).
LIST OF ACRONYMS USED AC AFW AP ANSI ARP ASME BB(PV CFR CR CST DBD
.
DC DEEP ECCS EOP EP FCV FSAR GPN I8C IE IFI IP IST LER LOCA NOV NBS NCV
'PSH NRR PCM PSIG RPS RWST SSEI SSFI S/G TOL TP TSAR T&T VTN Alternating Current Auxi 1 iary Feedwater Abnormal Procedure American National Standards Institute Alarm Response Procedure American Society Of Mechanical Engineers Boiler and Pressure Vessel Code of Federal Regulations Condition Report Condensate Storage Tank Design Basis Document Direct Current Design Equivalent Engineering Package Emergency Core Cooling System Emergency Operating Procedure Engineering Package Flow Control Valve Final Safety Analysis Report Gallons Per Minute Instrumentation and Control Office of Inspection and Enforcement.
NRC Inspector Followup Item Inspection Procedure Inservice Test Program Licensee Event Report Loss of Coolant Accident Motor Operated Valve Nitrogen Backup System Non-Cited Violation Net Positive Suction Head Office of Nuclear Reactor Regulation Plant Change Modification Pounds Per Square Inch Gauge Reactor Protection System Refueling Water Storage Tank Safety System Engineering Inspection Safety System Functional Inspection Steam Generator Thermal Overload Temporary Procedure Tracking System Activity Report Trip and Throttle Vendor Technical Manual
Calculations:
REFERENCES 162-001/048
[Bechtelj,=AFW Suction (piping supportsj.
Revision 2.
'62-008/009 LBechtelj, Auxiliary Feedwater Turbine Steam Supply fpipe supportsJ, Revision 9.
CIV-TPS-808.
AFW Pump Enclosure [foundationsj, Revision 0.
CN-TA-94-180, Loss of Normal Feedwater/Station Blackout for the Power Uprating, Revision 0.
N08-093-01 'team Pressures at Inlet of the (High Pressure)
Auxiliary Feedwater pump Turbine, Revision 0.
H08-093-03. Auxiliary Feedwater Flow at Minimum Steam Conditions Revision 1.
N08-420-01.
AFW Discharge Flow Limitation Requirements.
Revision 0.
N08-420-28. Auxiliary Feedwater Nitrogen Backup System Design Basis.
Revision l.
NOB-462-01. Verification of Auxiliary Feedwater Pump Net Positive Suction Head, Revision 1.
H08-469-01, Functionality Stress Analysis.
Aux. Feedwater Steam Supply Piping.
Revision l.
N08-498-01.
Pressure Drop to AFW Pump Turbine. Revision 2.
M08-498-03.
Minimum Steam Pressure at which Two AFW Pumps Can Be Run from One Steam Generator.
Revision 0.
H08-498-07.
AFW Flow Control Valve Trim Change Analysis of Pre-op Procedure Results.
Revision 0.
H08-498-08. Analysis of the Operability of the Auxiliary Feedwater Steam Supply NOV's, Revision 0.
M08-498-11, Nitrogen Consumption of Auxiliary Feedwater Control Valves.
Revision 0.
H08-498-12, Determine the Flow Coefficient (Cv) and Valve Position for AFW Flow Control Valve for Limiting Conditions of Flow. Following a Hain Steam Line Break. with Refurbished Impellers.
Revision 0.
N08-498-13, Verify Minimum Required AFW Flow to Steam Generators.
Revision 0.
N08-592-01, Calculation of impact of steam jet from feedwater line break locations 4 and 5.
on AFW system flow transmitters located below feedwater platform. Revision 0.
M12-383-03. Vortex Design Evaluation of Condensate Storage Tank. Revision PTN-0-S-H-90-001.
Calculation of Setpoint for Initiating Makeup to the Condensate Storage Tanks Revision 0.
PTN-0-S-M-90-0013.
Allowance for Machining AFW -A" Terry Turbine Casing.
Revision 0.
PTN-BBJM-68-001, Condensate Storage Tank Volume Requirement for Various Conditions'evision 0.
PTN-BBJM-68-019, Auxiliary Feedwater Required vs.
Time, Revision 0.
PTN-BBJM-89-102.
Reactor Cooldown with Fire in AFW Pump Area or Steam Supply Valve Area and loss of Power (Appendix R Open Action Item 218), Revision 1.
PTN-BCJE-90-014.
Evaluation of voltage drop created by the addition of the AFW DC HOV shunt field surge suppression resistors, Revision l.
PTN-BFJE-90-006, Motor Operated Valve Voltage Drop Calculations
- Generic Letter 89-,10. Revision 14.
PTN-BFJE-92-028, Valve Actuator Motor Thermal Overload Heater Sizing Generic Letter 89-10, Revision 11.
PTN-BFJE-92-032,125 Vdc Valve Actuator Motor Voltage Drop Calculation GL 89-10. Revision 0.
PTN-BFJF-90-12.
Determination of CST Inventory Requirement as a Function of Time After Trip, Revision 0.
(1980; source of original 185,000 gallons)
PTN-BFJI-94-010.
Condensate Storage Tank Level Indication Uncertainty Determination.
Revision 2.
PTN-BFJM-90-079.
NRC Generic Letter 89-10 HOV Actuator Evaluation.
Revision 18.
PTN-BFJH-92-012..
Aux Feedwater System Pipe-Flo/Net-Flo Computer Model.
Revision 0.
PTN-BFJM-95-008.
CST Volume/Setpoints Revision 3.
PTN-BFJN-95-001, Calculation Related to AFW Flow During a Hain Steam Line Break Event, Revision 0.
PTN-BFSM-97-032, Minimum Containment Sump Level for ECCS Switch over.
dated July 23.
1998.
PTN-BFSM-98-007, AFW Hydraulic Analysis for FCV Stem Travel Stop Tolerance.
Revision 0. (from self assessment)
PTN-BFSM-98-009. Validation of AFW Flows Assumed for Containment Response.
Revision 0.
(from self assessment)
PTN-CR95-1281, Condensate storage tank level calculations for Units 3 and 4.
Revision 1.
SE-SS-FPL-2090, Minimum Condensate Storage Tank Volume [summaryj, Revision 1.
Condition Reports:
CR 95-0654.
Governor valve binding and OST. 8/Z4/95.
CR 95-0739, Water in lube oil, 9/4/95.
k CR 96-0135, AFW pump A over speed trip, 2/9/98.
CR 96-0694.
AFW steam supply vs. ISI, 7/9/98.
(Self"Assessment reconsidered)
CR 97-0902, AFW pump ATWS flowrates, 5/27/98.
CR 97-1184, Safety Screening for TP inadequate, 8/2/97 CR 97-1374.
Checkvalve repairs, 9/10/97.
.CR 97-Z031, AFW pump A vibration, 12/4/97.
CR 98-0059.
Steam trap repairs'2/22/97.
CR 98-0109.
Solenoid valve wiring issue overlooked.
~
~
CR 98-0276, Unit 3 trip and steam leak
~ 2/16/98.
CR 98-0302.
Steam traps vs. drawings.
2/18/98.
CR 98-0966.
AFW Pump Discharge Testing.
(Self Assessment)
CR 98-0974.
Flow control valves, 6/22/98.
(Self Assessment)
CR 98-1034.
CST Outlet Check Valve Testing.
(Self Assessment)
CR 98-1050.
Check valves and Code boundaries.
7/10/98.
(Self Assessment)
CR 98-1052 '50F vice 325F. 7/10/98.
(Self Assessment)
'R 98-1053 'FW DBD Discrepancies.
7/10/98.
(Self Assessment)
CR 98-1054.
FSAR inconsistencies.
7/10/98. (Self Assessment)
CR 98-1056, Minor discrepancies noted in calculations.
the DBD. and drawings.
7/10/98.
(Self Assessment)
CR 98-1066.
AFW 1&C calculation issues.
7/14/98.
(Self Assessment)
CR 98-1069.
AFW pump minimum flow limitations. (Self Assessment)
CR 98-1070.
AFW system checkvalve testing.
3/14/98.
(Self Assessment)
e
Plant Change Hodifications (PCN):
11/3/81.
PCH 79-124. Auxiliary Feedwater Pump Control Modification. implemented PCM 80-117, Upgrade Auxiliary Feedwater Suction, Discharge.
and Steam Supply Piping. completed 10/8/85.
4. 7/14/87.
PCN 86-009 'uxiliary Feedwater Steam Supply Valve Replacement, Unit 4, change PCM 88-286/287.
AFW FCV Stem Travel. Revision l.
PCM 89-531.
Component Material Replacement for MOV-~-1403.
1404.
1405.
Revision 1.
PCN 90-040, Part Number Change.
Revision 0.
PCM 90-156, 'A'FW Pump Turbirie Thrust Bearing Failure.
3/ZZ/90: including O-PMM-075.8. Auxiliary Feedwater Pump Oi 1 Change.
Attachment l.
3/27/90.
. PCM 91-019. Modify Auxiliary Feedwater Turbine Trip and Throttle Valve Above and Below Seat Drains for HOV-6459A. B.
and C, completed 9/17/93.
PCN 9?-033.
Elimination of Electrical Trip to Aux. Feedwater Turbines.
approved 10/27/97.
PCH 95-114.
CST Low Level Alarm Modification, Unit 3, Revision 1.
PCH 95-141/142.
Relocate CST Level Switch. Revision 1. 8/30/96.
PCM 80-105. Auxiliary Feedwater Pump'Turbine Modification. 5/25/82.
PCH 80-055, Upgrade of Auxiliary Feedwater System Flow Control and Initiation.
10/4/82.
Drawings:
5610-C-18-393.
Turkey Point Units 354, Primary and Refueling Water Storage Tanks.
Revision
5610-E-Z6.
Sht 51A. Elementary Diagram CST Level Alarm. Revision Z
5610-J-1400-1.
Instrument Loop Diagram CST Level. Revision
5610-J-1400-2.
Instrument Loop Diagram CST Level. Revision
5610-J-1400-3.
Instrument Loop Diagram CST Level. Revision
5610-J-1400-4, Instrument Loop Diagram CST Level. Revision
5610-H-3075. Auxiliary Feedwater System.
Revision
5613-H-3018.
Condensate Storage System.
Revision
~
'
5613-H-3075, Auxiliary Feedwater System, Revision
5614-H-3018.
Condensate Storage System.
Revision 18, 5614-M-30?5, Auxiliary Feedwater System, Revision 8 5610-H-34-11, Turkey Point Units 3 and 4 AFW Pump Curves Nos.
03686.
036887.
and 036888.
Revision
5613-H-313, Turkey Point Unit 3 Setpoint List, Revision
5613-P-599-S, Turkey Point Units 3&4 High Head Safety Injection System.
SYS No.
62 Outside Containment Stress Problem 031. Revision
Safety Evaluations:
PCH 97-033, Elimination of Electrical Trip to AFW Turbines PCH 97-002, Replacement of Unit 3 ICW System Valve 3-50-360 PCM 95-097, Replacement of Main Steam Safety Relief Valve Discharge Piping PCM 96-008, Potential EDG Lockout Following Normal Stop SECP-95-046, SE for Units 3&4 Twenty-Fifth Year Containment Tendon Survei llances SEHS-96-037, SE for the Temporary Installation of a Freeze Seal and Blind Flange per TSA-04-96-046-11 to, Support HOV-4-350 Valve Repairs SEMS-96-048, Safety Evaluation for Measuring ICW Header Flow per TP-96-070 SEFJ-97-030.
CFR 50.59 SE for Performing the Rod Drop Test from All Rod Out Condition SEHS-97-033 'E for Unit 3 Reactor Core Foreign Object SENS-97-084, SE for Containment Sump Screen Design Requirements PTN-ENG-SEES-98-0024.
Safety Evaluation for EDG 3A Loading Consideration Due to Manual Loading of Electrically Droven Fire Pump PTN-ENG-SENS-96-079.
SE for NEI Initiative for Licensing Basis Conformance Plant Procedures:
3-PHI-018. 1.
CST Loop L-3-6384A and LI-3-6584 Calibration.
Dated 1/7/97 3-PHI-018.2.
CST Loop L-3-6384B Calibration.
Dated 1/8/97 4-PMI-018.1.
CST Loop L-4-6384A and LI-4-6584 Calibration, Dated 1/7/97 4-PHI-018.2.
CST Loop L-4-6384B Calibration.
Dated 1/8/97 3/4-OP-075. Auxiliary Feed Water System, dated February 10.
1998
L0
0,
pl
4-EOP-FR-S. 1, Response to Nuclear Power Generation/ATWS, dated February 6, 1998 3-EOP-E-O, Reactor Trip or Safety Injection. dated April 28, 1998 3-EOP-ECA-O.O, Loss of All AC Power.
September 27, 1997 3-EOP-ES-0.1.
Reactor Trip Response, April 29.
1997 3-0SP-075.6, AFW Train 1.
Backup Nitrogen Test, dated February 28.
1998 3-0SP-075.7, AFW Train 2 Backup Nitrogen Test, dated July 16.
1996 3-0SP-075.2.
AFW Train 2 Operability Verification, dated February 27.
1998 3-0SP-075.1.
AFW Train 1 Operability Evaluation, dated February 27, 1998 0-OSP-075. 11, AFW Inservice Test, dated August 18.
1995 O-PHM-075.2.
AFW Pump General Inspection, dated June 23.
1997 O-PME-075.2, A AFW Pump Turbine Speed Monitor Calibration.
dated December 9.
1997 O-PHH-075.4.
AFW Control Valve Inspection, dated November 10.
1997 O-PMI-075.4, AFW control Valve Actuator Overhaul.
dated August 5.
1997 O-PHH-075.5, AFW Pump Turbine and Turbine Oil Pump Inspection and Overhaul.
dated September 9,
1997 O-PMM-075.6, AFW Pump Rotating Element Removal and Replacement.
May 4.
1998 O-PHM-075.7.
AFW Pump Rotating Element Overhaul. July 23.
1997 0-ADM-502. In-Service Testing (IST) Program, dated June 15.
1998 3/4-EOP-E-l, Loss of Reactor or Secondary Coolant.
dated June 23.
1997 3/4-EOP-ES-1.3
~ Transfer to Cold Leg Recirculation.
dated November 18.
1997 3-PM}-062.35.
RWST Level Instrumentation Channel L-3-6583A Calibration,'ated April 21.
1998 3-ARP-097.CR, Control Room Annunciator Response, 8/5/98 3-OP-'075. Auxiliary Feedkater System.
2/10/98 4-OSP-075.5.
Auxiliary Feedwater System Flowpath Verification. 3/22/96
I
Other References:
3-OSP-075. 1, Auxiliary Feedwater Train 1 Operability. Verification. completed 10/5/96 4-0SP-075.2, Auxiliary Feedwater Train 2 Operability Verification, completed 12/16/96 O-OSP-075.11, Auxiliary Feedwater Inservice Test, completed 10/5/96 (Pump A)
and 12/16/96 (Pump C)
5610-075-DB-001.
Design Basis Document, Volume 1. Auxiliary Feedwater System.
Revision 8. 9/28/98 5610-075-08-019, Design Basis Document, Accident Analysis Basis Document.
Revision 8, 10/15/97 Engineering Guide M-3.6. Bolting Material Selection.
1/18/77 Engineering Guide MN-3.26, Valve Applications Engineering Guide STD-M-030, Piping Pressure Drop Calculations.
Revision
Engineering (ENG) Quality Instruction (QI) 1.0.
Design Control ENG QI 1.1. Engineering Package ENG QI 1.2. Minor Engineering Package ENG QI 1.3.
Drawing Change Requests ENG QI 1.5. Calculations ENG QI 1.7.
Design Input/Verification ENG QI 2.0
~ Engineering Evaluations ENG QI 2.1.
10 CFR 50.59 Screening/Evaluation O-ADM-100, Preparation, Revision.
Review. Approval and Use of Procedures Guidance for Performing
CFR 50.59 Safety Evaluations FOP-.82-50 and FOP-81-160, Operating Experience Feedback (OEF) Report.
Auxiliary Feedwater Piping Overpressurization:
NSAC/INPO SER 92-81 FOP-90-007.
SER 3-90, OEF Report.
Inadequate AFW Testing and Preventative Maintenance FOP-87-70, OEF Report evaluation on single failures in AFW System FOP-89-142.
OEF Report evaluation on Westinghouse Technical Bulletin NSD-TB-89-05
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FOP-94-095Sl.
OEF Report Evaluation on over speed of turbine driven pumps caused by binding in stems of govenor valves FPL letter L-86-77. reply to IE Bulletin 85-01, Steam'Binding of Auxiliary Feedwater Pumps, 2/28/86 IE Bulletin 85-01, Steam Binding of Auxiliary Feedwater Pumps LER 92-007. Automatic Auxiliary Feedwater Start on Hain Feedwater Pump Trip (U4). 9/29/92 LER 96-004; Auxiliary Feed Water Start on S/G Level (U3). 5/24/96 LER 97-004. Auxiliary Feedwater Automatic Start Upon the Trip of all Hain Feedwater Pumps (U3), 4/11/97 LER 9?-009, Console Switch Position Disabled Auxiliary Feedwater Auto Start Logic on 3B Steam Generator Feedwater Pump Trip (U3). 11/14/97 LER 98-001, Manual Reactor Trip Due to Loss of Turbine Control Oil Pressure.
with Steam Leak in Auxiliary Feedwater Steam Supply Piping, 2/16/98 NRC Information Notice 87-34, Single Failure in AFW System NRC Information Notice 93-51, Repetitive Tripping of-AFW Pump Turbine NRC Inspection Manual.
Inspection Procedure 93809, Safety System Engineering Inspection (SSEI).
issued 6/23/97 NRC Inspection Report 50-250/85-32:
50-251/85-32.
Safety System Functional Inspection Report.
10/1/85 Turkey Point Units 3 and
Vendor Technical Manual for Auxiliary Feedwater Pump and Terry Turbines Westinghouse letter JPN-PTN-94-5136, Thermal Power Uprate Project Non-LOCA Input Assumptions. letter 2/6/95 Westinghouse letter MIP-FPL-1476, Thermal Power Uprate Project.
Revised CST Evaluation and Atmospheric Dump Valve Summary, 2/2/95 Westinghouse letter MIP-FPL-1556, Thermal Power Uprate Project, CST Minimum Required Usable Volume [199.100 gallons]. 4/18/95
'I Westinghouse Technical Bulletin NSD-TB-'89-05.
Minimum Auxiliary Feedwater Flow Rates.
'10/31/89
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