ML20234E550
| ML20234E550 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 09/16/1987 |
| From: | Carey J DUQUESNE LIGHT CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| 2NRC-7-201, NUDOCS 8709220430 | |
| Download: ML20234E550 (38) | |
Text
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'3%p3C (412) 393-6000 on. oxeoro sntre 2NRC-7-201 301 crant street j
Pittsburgh, PA 15279 September 16, 1987.
l U. S. Nuclear Regulatory Commission Attn:
Document Control Desk Washington, DC 20555
Reference:
Beaver Valley Power Station, Unit No. 2 i
Docket No. 50-412, License No. NPF-73 Operational Self Assessment Gentlemen:
In a meeting held on May 18 and in our letter dated June 29,
- 1987, Duquesne Light Company informed you of our plans for performing Operational Assessment of Beaver Valley Unit No.
2.
The an assessment would take special inputs from groups involved in the Power Ascension Test Program.
These inputs would then be periodically reviewed by a committee of senior management which would monitor the attributes of operational readiness for safe and reliable station operations.
This letter forwards the Operational Assessment Report for Beaver Valley Unit No.
2.
This report covers the interval between Initial Criticality, 8/4/87, to completion of the 50% power plateau testing on 9/6/87.
As indicated in our June 29, 1987 letter, the assessment focused on the following areas:
I.
Organizational interfaces II.
Plant configuration control III.
Procedural adequacy and compliance IV.
Teamwork and communications V.
Operations Quality Assurance effectiveness VI.
Responsiveness and timeliness of Engineering and Construction Support' Services VII.
Control room operations and effectiveness of training VIII. Adequacy of design based on test
- program, technical specifications and FSAR 8709220430h00012
,' 00 J PDR ADOCK ppg P
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r~V Beaver Valley-Power Station, Unit No. 2 Docket No. 50-412, License'No. NPF-73' Operational Self Assessment Page l The agenda for the weekly meetings normally-included the following items:
Previous Open Items and. Minutes Chemistry Report e
Test Deficiency Reports Special Evaluations.(ISEG/OSC) e Quality Assurance Reports Incident and Unit Off Normal Reports Plant Status Summary Status of INPO SOER's, SERs and O&MR's Review of Control Room Annunciator Status STA Surveillance Reports on Operating Activities While the assessment identified a
number' of areas requiring H
improvements, our overall evaluation concludes that the:
Unit startup and power ascension was evenly
- paced, e
deliberate and well disciplined.
Problems were identified, assessed and corrected in a. timely i
manner.
Support groups provided timely support.
l l
Experience gained through operation. of Unit 1
and e
transferred to Unit 2
has been a noted asset.
Disciplined Control Room operations were evident throughout this period.
Modeling of the BV-2 technical ~ specifications-after the BV-1 specifications has substantially minimized the overall impact on training, procedure development.and implementation.
l The magnitude and types of problems experienced during this l
period of startup were generally less than those experienced by similar
- plants, as documented in NUREG-1275 dated July 1987.
The enclosures submitted herein include the following:
ATTACHMENT "A" Evaluation Summary APPENDIX "A" Summary of Unit Off Normal Reports end Incident Reports
boaver Valley Power Station, Unit No. 2 Docket No. 50-412, License No. NPF-73 Operational Self Assessment Page 3 i
Summary of Technical Specification Action APPENDIX "B" Statements and NUREG-1275 Comparison APPENDIX "C" ISEG Evaluations Summary of STA Surveillance APPENDIX "D" Summary of Testing APPENDIX "E"
.PPENDIX "F"
Summary of Engineering Support APPENDIX "G" Summary of Unit 2 Annunciator Trends APPENDIX "H" Summary of OSC Assessment Chemistry Report APPENDIX "I"
Summary of Milestones for Unit 2 APPENDIX "J"
This self-assessment report satisfies our written and oral I
commitments in this regard.
Beaver Valley 2 is now in a maintenance 1
outage with a
scheduled. return to service date of September 21, 1987.
Since we have completed testing at the 50% power plateau, our intentions are to proceed with testing at the 75% power plateau following this outage and complete the remainder of the testing necessary to achieve commercial operation by mid-October.
Very tr y yours, 1
l
. J.
Carey Vice President Nuclear Group Attachment cc:
Mr.
F.
I. Young, Sr. Resident Inspector.(Unit 1)
Mr. J. Beall, Sr. Resident Inspector (Unit 2)
Mr. W. T. Russell, NRC Region I Administrator Mr.
P. Tam, Project Manager
ATTACHMENT A l
OPERATIONAL SELF ASSESSMENT
}
j l
I.
Organizational Interfaces i
Purpose l
To evaluate departmental interfaces and assess their ability to support efficient station operations.
Included were the evaluation of engineering response times, adequacy of evaluations for design or test deficiencies, the preplanning of normal operation evolutions and interface with test program initial l
conditions.
Also included were maintenance and construction schedule adherence and response to corrective maintenance items.
l Summary of Findings j
1 Prior to fuel
- load, the Technical Work Planning Group was
-l instituted and situated in an area within close proximity to the j
l control room.
The group was staffed with representatives-from l
Operations, Maintenance, Engineering, Planning Outage Management, QC, Construction, Testing and QA Surveillance.
A j
Shift Test Director with experience in operations headed the organizations to interface with the Nuclear Shift Supervisor and 2
testing supervisors.
The Shift Test Director provided for the j
integration of all necessary activities to support control room I
operations during the test program.
At the disposal of the Technical Work Planning Group were startup field operators, operations, and construction personnel.
This allowed coordination of field work with all responsible groups.
Two events were evaluated with respect to organization interface problems:
On August 7 a Reactor Trip occurred as a result of four dropped control rods.
Based on previous BV-1 experience, and lack of detailed trovtieshooting, circuit boards were changed out and inadequate communications resulted in delaying effective post-maintenance testing.
A second reactor trip resulted from the same problem.
In-depth troubleshooting occurred after the second event which resulted in finding a bad circuit thyristor.
Thyristors have not been a
problen at Beaver Valley so a complete set of additional rod control traces were taken at this time.
The root cause of the problem was the lack of i
adequate post-maintenance testing.
The fact that rod movement was a
necessity to troubleshoot was not communicated strengly enough to overcome this sensitivity.
Post maintenance testing was to be done during the subsequent startup but the lack of adequate communications and a breakdown in interfacing, precluded such monitoring.
A second event associated with an isolated pressure transmitter is discussed in more detail in the next section under Configuration Control.
L______________-._
ATTACEMENT A Page 2 Conclusions The dropped control rod event required a manual reactor trip.
The corrective action required the assignment of specific individuals to review post maintenance testing.
The Plant Manager required a
- review, by plant personnel, of a letter stressing the importance of Post Maintenance Testing.
Although deficiencies were noted under organizational interfaces that resulted in plant challenges, all deficiencies were identified and aggressively addressed by management.
II. Plant Configuration Control Purpose To evaluate plant configuration control in support of plant operations and pre-operational testing.
Areas evaluated were control of valve
- lineups, verifying annunciator status against plant
- status, operating manual changes, test procedure changes, l
temporary operating procedures and Type 1 drawing (those required in the control room) updates from engineering.
Summary of Findings In
- general, there were few problems found relating to configuration management.
Areas evaluated include the following:
1 Problems arose in the area of configuration control of plant instrument valves related to an auxiliary feedwater pump instrument.
As a
result of this event, safety related valves and power supplies were re-verified in their required position.
Of the systems surveilled, a small number of problems were identified.
The problems were judged to be as the result of inadequate interface between proof test, pre-operational test, and operations groups through the transition prior to
- and, in some cases, after fuel load.
Annunciator status verification was evaluated and is noted with conclusions and corrective actions in Appendix G of the appendices.
1 Operating Manual and Test Procedure changes were e
routinely evaluated by the shift Technical Advisors (STA's) and the Onsite safety Committee (OSC) for changes to plant
- status, configuration control and design adequacy.
Actions taken and conclusions are contained appendices D and H of this report.
l
ATTACHMENT A Page 3 i
f Type 1
drawings (those required in the control room) were continuously reviewed for-adequacy through use by all groups in the process of engineering, operation and j
maintenance activities.
There were no problems found.
Personnel were assigned for drawing updates on an expedited basis.
If drawings were not issued as final as-built, they were red-lined to accurately depict the as-built status at that time.
Design changes were not j
turned over to the Plant Manager until updates were complete.
Conclusions j
l The necessary corrective actions have been taken regarding plant configuration control in the areas of valve lineups and annunciators.
Conclusions related to the other areas evaluated are contained in the noted appendices.
III. Procedural Adequacy and Compliance Purpose To evaluate the adequacy of and compliance to various procedures used in the administration, testing, operations and maintenance of BV-2 during the evaluation period.
Summary of Findings Procedural inadequacies and a
few administrative procedure violations were the the major problem noted through STA surveillance, OSC or QA surveillance reviews.
These reviews are noted in Appendices B, D and Section V with conclusions.
Conclusions Procedure adequacy has been the area with the greatest number of corrective actions identified.
One of the objectives of the test program was to identify problems with procedures.
As operations personnel gain additional at-power experience, changes will be required to enhance procedures.
No procedure inadequacies were found that compromised plant safety.
QA surveillance deficiencies were normally documented through Test l
Deficiency
- reports, expediting the resolution of deficiencies.
Overall no adverse trends have resulted from these evaluations as noted in the appendices.
l l
ATTACHMENT A l
Page 4 l
IV. Teamwork and Communications l
l Purpose To evaluate control room operations,
- teamwork, and the communications necessary to maintain control over the plant status with the large number of activities occurring in the station.
l Summary of Findings Adequate control was maintained, with the exception of three activities, throughout the evaluation period.
i e
Problems arose in the area of communications in meeting j
initial test conditions as reported in Section I
Organizational Interfaces.
\\
Other problems occurred relative to NRC deportability j
and the timeliness and adherence to reporting requirements.
During the recovery from a reactor trip initiated for i
- testing, the electrical buses were restored and the l
motor-driven auxiliary feed (AFW) pumps were shutdown while the main feed system was being restored.
When the main feed pumps were being restarted, the operator failed to hold the switch in the " start" position long enough for the recirculation valve to open.
This caused the l
motor-driven auxiliary feed pumps to start on a loss of l
the main feed pumps.
One pump had failed to start and l
concurrently, the operator secured the AFW pumps to restart the main feed pump.
After the main feed pump was
- started, the AFW pumps were put in " Auto", but there was an overcurrent trip locked in on the AFW pump.
This condition was identified later during subsequent testing on the AFW pump.
Conclusions
- Overall, adequate teamwork and communications were evident i
I throughout the test program.
During tLe period of the detailed operations assessment there were no Technical Specification l
violations or loss of safety system function events.
A comparison with other Westinghouse planta during the startup l
phase is included in Appendix B.
All significant problems associated with this operational phase will be presented to the BV-2 operators during licensed l
retraining as part of our experience feedback program.
This effort has been very effective in reducing recurrent personnel related problems and also serves as a
useful forum for the operators to identify items that require additional management attention.
ATTACHMENT A Page 5 V.
Operations Quality Assurance Effectiveness This section itemizes the Quality Assurance activities during the period of the assessment.
Summary of Findings The following Quality Assurance activities were conducted:
Two (2) surveillance of Beaver Valley Tests'(BVTs)
Twenty-eight (28) surveillance of Initial Startup Testing covering fifteen (15) tests.
One (1) surveillance of Maintenance Surveillance Procedures (MSPs).
One (1) surveillance of Preventive Maintenance Procedures (PMPs)
Three (3) surveillance of Operations Surveillance Tests (OSTs)
Five (5) surveillance of Pre-Operational Testing covering three (3) tests.
l l
Twenty (20) surveillance of System Operability l
=
Verification Testing covering twelve (12) tests.
Four (4) surveillance of Special Test Procedures covering two (2) tests.
During this period no Surveillance Deficiency Reports were issued I
as Test Deficiency Reports were utilized to resolve any j
procedural problems.
No trends adverse to the quality aspects of I
the Testing Program were identified and adherence to l
administrative requirements was indicated to be satisfactory.
l i
Three audits of operational readiness resulted in one finding and ten observations.
The general conclusion drawn from these audits was that there were no adverse trends in the principle areas i
reviewed which would affect the Unit 2 Power Ascension Program.
l l
Five audits of plant support organizations and activities were conducted which resulted in 2 findings and 19 observations.
No significant concerns or program weaknesses were indicated as a j
result of the audits performed during this period.
One Nonconformance and Corrective Action Report (NCAR) was issued.
This NCAR resulted from a procedure violation which did i
not affect Power Ascension activities, i
i I
ATTACHMENT A Page 6 Conclusion A
review of the QA Surveillance, Audits and NCAR's did not indicate any adverse trends which would prevent safe and reliable operation of Unit 2.
The organizational interfaces between QA and other plant organizations appeared to be well established and very effective in satisfying QA requirements.
No significant concerns or program weaknesses were identified which would adversely affect the operations Quality Assurance Program.
The i
general consensus of the Quality Assurance Unit is that the various organizations involved with Beaver Valley Unit 2 are l
working together as a team to ensure that the unit is operating i
in a safe and reliable manner.
VI. Responsiveness and Timeliness of Engineering and Construction Support Services Purpose To evaluate the effectiveness of the engineering and construction support during plant startup.
Also evaluated were the adherence to administrative controls, OSC review, software updates, and the ability to respond to additional tasks in a timely manner.
Summary of Findings l
Responses to Engineering Memoranda have generally been timely I
relative to resolution of testing deficiencies.
Engineering personnel provided coverage around the clock as part of the Technical Work Planning Group that was previewed in Section I.
Being part of this team fostered teamwork and communications that I
carried over through this assessment phase.
Appendix F contains l
a summary of engineering support data.
Construction support forces have been managed by DLC supervisors and have responded to station needs in an expeditious manner.
Support services personnel
- have, in some
- cases, worked in a maintenance capacity for a
number of years and this fact minimized problems during the transition phase.
Conclusion Engineering and construction support services have been satisfactory throughout the assessment period.
VII. Control Room Operations and Effectiveness of Training Purpose To evaluate control room operators on their ability to work together as a
team and their ability to maintain adequate l
l l
ATTACHMENT A Page 7
'l controls during the power ascension phase of the test program.
Training effectiveness has also been evaluated.
Feedback will be provided to the Training Department as part of the operations i
experience program during licensed retraining.
Summary of Findings Phase II, or Pre-operational testing personnel, were tasked to manage the testing program to conclusion.
Along with the testing
- charter, the operations group is required to maintain control I
over operation of the plant within established administrative and license requirements.
Control was maintained in an exemplary manner with problems being held to a minimum.
Problems that.did arise in the areas of organizational interfaces, plant configuration control and reporting requirements have been cited in Sections I,
II and IV.
Management and craft experience at I
BV-1, and the use of the BV-1 plant specific simulator, were i
judged to be key factors in the success of the BV-2 Power Ascension Test Program.
Many historical PWR problems, such as trips due to feedwater control
- problems, have been avoided I
through this experience and training.
Some testing was executed on the simulator to avoid unexpected problems and check the fidelity of the model.
Conclusions control room activities have been managed well and the supervisor / operator control over the plant and outside influences have been conservative and deliberate.
Problems that did occur were adequately tracked and analyzed with a minimum number of recurrent problems.
Deficiencies noted are identified in sections I, II and IV and conclusions noted therein.
The effectiveness of the training and operating programs will continue to be assessed and adjusted based on experience gained through power and shutdown operations.
VIII. Adequacy of Design Purpose To evaluate the adequacy of the Beaver Valley Unit 2 design based on the test
- program, Technical Specifications and Final Safety Analysis Report (FSAR).
The type and magnitude of problems noted during the test phase are an indicator of the adequacy of plant design.
l l
l l
l
9
' ATTACHMENT A i
Page.8 Summary of Findings l
There were a number of requests for Engineering to evaluate test l
results which did not meet the stated test acceptance criteria.
j For
- example, heat transfer rates for the excess letdown, seal
- water, and non-regenerative heat exchangers tested below the l
acceptance criteria.
Engineering evaluation revealed that the I
test procedure acceptance criteria were very conservative and that the reported test results were acceptable to meet design objectives.
I e
The Beaver Valley Unit 2 circulating water chlorination I
system was non-operational and a hypochlorite addition i
tank was implemented as a substitute design feature.
An
.{
Appendix B
Environmental Evaluation was performed before l
hypochlorite was added to the circulating water system.
1 Beaver Valley Unit 2
experienced a loss.of all three l
Reactor Coolant Pumps on underfrequency during a fast transfer from onsite power to offsite sources.
The reason was due to the installed IPAC underfrequency relays not being adequately designed to reset after a i
fast bus transfer to offsite power.
Original specifications indicated that the design was adequate, but the end result showed that the transfer function had j
inadequate reset time on a loaded bus.
The IPAC relays on all three reactor coolant pump buses were replaced with Hathaway relays which have been used successfully at l
Unit 1.
Subsequent Unit 2 testing on the fast transfer l
l showed acceptable results.
l Duquesne Light has found no instances where the l
Technical Specifications required an emergency Technical Specification change.
The majority of desired changes l
involved administrative or format corrections which are being resolved through the normal revision process.
l t
Changes are being processed for the next revision to the Beaver Valley Unit 2
FSAR.
However, the majority of these are minor changes remaining from the construction phase.
There are no significant changes pending from the initial criticality to fifty percent power ascension
{
- phase, j
1 Conclusion i
Although there are a number of differences between Beaver Valley Units 1
and 2,
much of the design is very similar.
Duquesne Light's experience with Beaver Valley Unit 1
has minimized problems encountered with BV-2 design.
NRC endorsement for modeling the Beaver Valley Unit 2 Technical Specifications after j
the Beaver Valley 1
Technical Specifications has had a very j
positive impact on minimizing historic problems associated with 3
implementation, j
i l
1
APPENDIE A Summary of Unit Off Normal Report (UONR),
Incident Reporting (IR), Licensee Event Reporting (LER). System f
l i
PURPOSE To evaluate in-house off normal events that occurred from the time of 1
initial criticality (8/4/87) to 50%
power operation.
This will l
assess the operation of Unit 2
in the following categories:
organizational interfaces, plant configuration control, procedural adequacy and compliance, teamwork and communications, responsiveness i
and timeliness of engineering and construction support services, l
control room operations and effectiveness of training, and adequacy i
of design based on test program, Technical Specifications and FSAR.
I This will determine if there are any common or recurring problems which could impact the safe and reliable operation of Unit 2.
l ACTIONS TAKEN i
Data for each off-normal event regarding root cause determination, corrective action initiation and designated evaluation criteria were placed in a
matrix (attached) to determine frequency of cccurrence and to perform trend analysis.
The reports were grouped on the matrix as either a
Licensee Event Report (most significant), or a Unit Off Normal Report. A review of the Licensee Event Reports shows the most frequent cause to be human error (operator, maintenance, personnel),
followed by equipment malfunctions and procedural inadequacies.
Although human errors have been designated as the root cause for 50%
of the LERs generated, no specific common cause could be established between events.
Only two of the five human error events involved similar circumstances which were related to reactivity control with a
near positive moderator temperature coefficient.
This was corrected by additional administrative guidelines.
The equipment and procedural problems were also determined to be random and not the result of a common cause.
A l
review of the Unit Off Normal reports shows that the majority of these events involve equipment problems.
However, in only one case involving turbine thrust bearing trip actuation did a common trend occur.
Actions to resolve this problem are being pursued.
All other noted problems do not appear to be limited to a specific system or evolution.
Although a number of events appear to involve common root cause and corrective action initiations, from a review of the matrix, no major common mode or detrimental trend conditions exist regarding the safe and reliable operation of Unit 2.
CONCLUSIONS A
review of all of the information shown in the matrix indicates that the repair of a
component is the most frequent corrective action taken.
Retraining of selected personnel on specific items is also performed frequently.
However, these personnel errors / retraining do not indicate or identify any trends that are significant, considering this phase of plant operation.
Within the areas focused on for
' APPENDIX A Page 2 i
i evaluation
- purposes, the
" Adequacy of Design Based on Test Program, j
Technical Specifications, and FSAR" category had high totals.
Even-though numerous occurrences fall into.this.' category, no trends could be identified, as the incidents in this category ranged from inadequate setpoints, to equipment malfunctions to-inadequate i
post-maintenance testing methods.
'i 4
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APPENDIX A j
l i
IR No.
LER No.
Title 87-20 87-13 Inadvertent ESF Actuation Due to Loss of ERF 4KV Bus 'G' l
87-21 87-12 Manual Reactor Trip Due to Dropped Rods j
87-22 87-12 Manual Reactor Trip Due to Dropped Rods UONR 87-23 Gaseous Waste Surge Tank Rupture Disc Problem 87-24 87-14 Reactor Trip on Low-Low Steam Generator Level 87-25 87-15 Reactor Trip on Low-Low Steam Generator Level 87-26 87-16 Auto Start of 23B Motor Driven Auxiliary Feedwater Pump 87-27 87-17 Inadvertent Feedwater Isolation Due to Initial Startup Test Atmospheric Steam Dump Valve Setpoint Drift UONR 87-28 87-29 87-18 Reactor Trip Due to Shorting of Rod Control Power Supply j
CO2 Dump Test Incident UONR 87-30 UONR 87-31 Turbine Trip Due to Misoperation of EHC Control System Steam Driven Auxiliary Feedwater Pump Trip on Overspeed UONR 87-32 Steam Driven Auxiliary Feedwater Pump Speed Low UONR 87-33 Steam Driven Auxiliary Feedwater. Pump Trip on Overspeed UONR 87-34 Source Range Detector N32 Startup Meter Spiking UONR 87-35 Reactor Critical Below Rod Insertion Limit UONR 87-36 87-37 87-19 Reactor Trip on Turbine Overspeed Signal-UONR 87-38 Turbine Trip Due to Faulty Thrust Bearing Trip Detector Operation UONR 87-39 Turbine Trip Due to Faulty Thrust Bearing Trip Detector Operation UONR 87-40 Turbine Trip Due to Thrust Bearing Trip Detector Operation Offscale Dosimeter After Containment Entry UONR 87-41 Out-of-Service Effluent Velocity Probes UONR 87-42 UONR 87-43 Improper Level Alarm Setpoint on SIS Accumulator 87-44 87-020 Reactor Trip During Main Steam Isolation Valve Closure Test
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APPENDIX B j
SUMMARY
OF TECHNICAL SPECIFICATION ACTION STATEMENTS PURPOSE This section itemizes components covered by the BV-2 Technical Specifications which were declared inoperable during the period 8/17/87 through 9/6/87.
This interval was chosen because the full power license was issued on 8/14/87, invoking all Mode 1
requirements, and the plant subsequently entered Mode 1 for the first l
time on August 15, 1987.
l l
EQUIPMENT TECHNICAL OUT OF SERVICE DESCRIPTION SPECIFICATION INTERVAL REMARKS 1.
Control Room Chlorine 3.3.3.7 8/17 - Present Equipment time response j
Detectors does not meet vendor specification values.
2.
Control Room Radiation 3.3.3.7 4 days Revised time response Monitors criteria in Technical i
Specifications.
3.
Liquid Waste Effluent 3.3.3.9 15 days Monitor 4.
Gaseous Waste Radia-3.11.2.5 1 day tion Monitor 5.
Rod Insertion Limit 3.1.3.6/
8/17 - Present Four-hour rod position Monitor / Digital Rod 3.1.3.2 checks in effect Indication Deviation 6.
Liquid Waste Flow 3.3.3.9 8/16 - Present Four-hour flow estimates Recorder during discharge i
7.
Pressurizer PORV 3.4.11 8/17 - Present Limit' switch ' pNblem I
(455D) only I
8.
Pressurizer PORV 3.4.11 3 days (456) i !
9.
Gaseous Waste Radia-3.3.3.10 36 houre tion Monitor
~
'Page 2 APPENDIX B j
-EQUIPMENT TECHNICAL OUT OF SERVICE DESCRIPTION SPECIFICATION
' INTERVAL REMARKS
- 10. Fuel Building Venti-3.3.3.1 8/18 - Present Restricts fuel movement lation Radiation
-Monitors
- 11. SLCRS Damper 3.7.8.1 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
- 12. Condensate Polishing 3.3.3.10 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> Radiation Monitor
- 13. Containment Air 3.6.15 7 days Temperature i
- 14. Main Steam Radiation 3.3.3.1 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Monitor
- 15. Feedwater Flow Trans-3.3.1.1 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> mitter i
- 16. (3) Radiation Monitor 3.11.2.1 8/26 - Present Flow estimated every 4 Flow Instruments hours
- 17. Lecon Building Venti-3.3.3.10 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> lation Monitor
- 18. SLCRS Radiation 3.3.3.1 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> Monitor i
- 19. Waste Gas Storage 3.3.3.10 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Vault Monitor
- 20. Loop "A" Delta T 3.3.1.1 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
- 21. RCP Underfrequency 3.3.1.1 18 hour2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> Design change required to Relays modify relays
- 22. Turbine AFW Pump 3.7.1.2 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
- Page-3' APPENDIX B EQUIPMENT.
' TECHNICAL OUT OF SERVICE DESCRIPTION SPECIFICATION INTERVAL REMARKS
- 23. 21B LHSI Pump 3.5.2 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />
. Pump would not hand rotate
- 25. Control Room Rad--
3.3.3.7
.8/29 - Present Control Room Ventilation-iation Monitors Isolated
- 26. Elevated Release 3.3.3.10 4 days
' Radiation Monitor
- 27. 21A Reactor Plant 3.7.4.1 8/31 High Vibration - 1 other River Water Pump pumps available f
i
- 28. A Loop Bypass flow 3.3.1.1 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br />
' Transmitter calibrated
- 29. Containment Isolation 3.6.3.1 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Valve 2CVS-SOV151B
- 30. Delta Flux Alarm 4.2.1.1 9/2/87 -
Delta flux being logged-Present every 30 minutes
- 31. Elevated Release 3.3.3.10 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> Radiation Monitor
GENERAL COMMENT
S 1.
Eight items placed the plant into ACTION STATEMENTS requiring shutdown in a
specified period (Ref. items 11, 15, 20, 21, 22, 23, 28, 29).
None of these items had recurrent failures during-the evaluation period.
The total out of service time for these components was 170 hours0.00197 days <br />0.0472 hours <br />2.810847e-4 weeks <br />6.4685e-5 months <br /> out of a'504-hour interval.
Of the eight. items, only two (items 22 and
- 23).can be considered a-partial loss of the safety function since all others represented instruments or equipment that was effectively placed into its accident
- mode, i.e.,
bistables tripped, valves' closed, etc. upon failure detection.
2.
A separate effort is being conducted to correct the radiation monitor related
- problems, which represented 16 out of the'31 items in 'the data base.
The Digital Radiation Monitoring System consists of approximately 84 process, effluent and area radiation
.l monitors.
This failure rate should be substantially reduced-with
]
resolution 'of the engineering problems and increased personnel a
familiarity with the DRMS.
]
_y 1
I
'Page 4 APPENDIK B General Comments, (Cont'd)
I 3.
Items 24 and 27 have installed redundancy to meet Technical j
Specification requirements.
4.
Items 5
and 30 are being met through increased surveillance by j
the operators.
ACTIONS REQUIRED 1.
Engineering and/or maintenance efforts to resolve items 1, 6, 10 16 should be specifically identified and scheduled for resolution.
i 2.
Operations, engineering and computer personnel should resolve items 5
and 30 during the scheduled _ outage to minimize the increased surveillance and enhance startup and load swing monitoring capability.
3.
Item 7 should be resolved by maintenance during the outage.
4.
Items 24 and 27 should be actively worked through the outage to provide operational flexibility.
1 CONCLUSION The Technical Specification development effort involved a continuous interface with the NRC and plant groups by the Licensing Section.
Therefore, the.relatively few significant problems that have been encountered are believed to be a
positive indication that organizational interfaces, communit.ations, procedural adequacy and test programs have functioned at an acceptable level in this regard.
Consistent with the recommendations of NUREG-1275 dated July 1987, we believe that significant improvements in the overall technical specification compliance area can be obtained through issuance of the specifications.far in advance of their required need.
This would enable the utility to perform a more comprehensive implementation program and maximize the familiarity of the facility organizations to the Technical Specifications.
The impact of this was minimized at BV-2,
- however, by NRC endorsement of the use of the BV-1 technical specifications for the baseline BV-2 document.
All necessary corrective actions are scheduled with completion dates to support BV-2 operations.
In review of the startup experience documented in NUREG-1275 for plants with a Westinghouse NSSS and turbine-generator, BV-2 appears to be above average in the performance categories associated with Technical Specification problems.
'Page 5 APPENDIX B NUREG-1275 COMPARISON 1
This section compares Beaver Valley 2
performance to four other l
facilities with Westinghouse NSSS and Turbine Generators as l
identified in NUREG-1275.
i REACTOR TRIPS l
l OL to Criticality FPL to Criticality to FPL Commercial TOTAL l
Plant A 0
1 10 11 Plant B 0
3 9
12 Plant C 0
2 17 18 Plant D 0
0 22 22 BV2 2
2 5**
9**
l l
l ESF ACTUATIONS I
)
Criticality to FPL Commercial TOTAL Plant A 1
1 3
5 Plant B 8
9 8
25 Plant C 5
1 14 20 Plant D 0
44 30 74 BV2 6
1 7**
14**
2 Safety Injections SLCRS Actuation Feedwater Isolation, AFW Pump Start AFW Pump Start Feedwater Isolation (5) AFW Actuations on Rx trips
- 1 Diesel Start AFW Actuation Commercial ~7? ration l
not attained
'Page 6 APPENDIE B NUREG-1275 Comparison, (Cont'd)
TS VIOLATIONS OL to Criticality FPL to Criticality to FPL Commercial TOTAL Plant A 8
1 22 31 Plant B 9
4 8
21 Plant C 1
0 13 14 Plant D 21 2
23 46 BV2 1
0 0**
1**
Rx trip breakers closed with IR surveillance not performed.
LOSS OF SAFETY SYSTEM FUNCTION OL to Criticality FPL to Criticality to FPL Commercial TOTAL Plant A 1
0 4
Plant B 3
0 2
5 l
Plant C 0
0 2
2 Plant D 3
0 1
4 BV2 2
0 0**
2**
(MSIVs, CIB to CREBAPS)
Partial Loss of Safety System Functions were:
l l
FWE-P23A Start Failure Containment Airlock l
Pressurizer Safety Valves Commercial operation not attained
j l
APPENDIX C Summary of Independent Safety Evaluation Group (ISEG) Assessments PURPOSE
)
l In accordance with Beaver Valley Unit No. 2 Technical Specification l
No.
6.2.3, the Independent Safety Evaluation Group (ISEG) shall l
function to examine unit operations characteristics to find areas for l
improving plant safety.
The Operations Assessment Committee made 1
requests to the ISEG to evaluate several issues which they felt i
needed attention.
l 1
ACTIONS TAKEN Procedure Readiness l
The Independent Safety Evaluation Group evaluated Beaver Valley Unit No.
2 procedural readiness to meet the Technical Specification surveillance requirements to be able to enter Mode 4 operations.
The evaluation was a
100%
review of the first eight sections in the Technical Specifications (3/4.1 through 3/4.8).
All surveillance procedures were reviewed to verify that:
l l
- 1. A BVPS-2 procedure exists to accomplish surveillance on Technical Specifications listed for Mode 4.
- 2. The test surveillance frequency encompasses the Technical i
specification surveillance.
l
- 3. The test acceptance criteria is the same as given in the Technical Specifications.
- 4. The test lists steps in its procedure to attempt to evaluate the appropriate acceptance criteria.
- 5. Any reference to the Technical Specifications in the test procedure conforms to the appropriate Technical Specifications, t
ISEG diti not attempt to judge the correctness or completeness of the test pr)cedure
- methods, but evaluated the procedure conformance to Technical Specification criteria.
I t\\fter the Mode 4 review was completed, a similar review was also done ffor Modes 3,
2 and 1.
Eight reports were issued which detailed I
l potential items where additional changes or corrections may have been needed.
Although numerous potential items were initially identified by ISEG, most were not significant corrections.
'Page 2 APPENDIX C Master Punchlist Evaluation e
ISEG reviewed the BVPS Unit No. 2 Master Punchlist (MPL) containing open items required for Mode 2.
The review attempted to determine the accuracy of this list after a substantial reduction (2300 to 300 items) over a short time period which occurred during the first week of August.
ISEG made a
quick review of the 2000 items removed from the Mode 2 MPL and selected 30 in safety related systems for further review.
Upon completion of a
detailed review by the 1SEG and Operations Assessment Committee, the removal of all 30 items were found to be acceptable.
- Thus, from this representative
- sampling, the MPL reduction was judged to be acceptable since no instances of incorrect dispositions were found.
Startup Checklist Assessment e
ISEG reviewed the BVPS-2 startup checklist against the OST required frequency to ensure that surveillance are properly checked as being performed upon plant mode changes.
Checklists B (Mode 5 to 4), C (Mode 4
to
- 3) and D (Mode 3 to 2 & 1) were reviewed.
Several plant mode discrepancies between the OSTs and startup checklists were found.
A review showed that all OSTs were actually performed during the initial plant startup.
However, the checklists are being revised l
to correct the identified deficiencies and will be completed by the time the outage is completed.
l NRC Information Notice 87-25, Wrong Unit, Train and Component l
ISEG is performing an evaluation focusing on actual performance of operations, mechanics, electricians, MCR's and testing and plant l
performance personnel.
A preliminary report was provided to the Operations Assessment Committee on August 19.
The report reviewed and provided recommendations in the following areas:
1.
Plant equipment tagging, markings and identification 2.
Procedures 3.
Training 4.
LERs and Incident Reports 5.
SOERs 6.
Personnel Performance 7.
Industry Evaluations 8.
NRC Information This evaluation is continuing and final recommendations are expected during October 1987.
The Operations Assessment Committee recommended that a
survey be issued to all electrical and mechanical maintenance
- workers, I&C technicians and operators requesting their input on how to reduce / eliminate possible human errors which could occur on the wrong unit, wrong train or wrong component.
This group of people was targeted since they are the personnel who actually conduct the work and could potentially see these problems.
The ISEG final report will factor in this input.
Page 3 APPENDIX C CONCLUSIONS The people contacted during all of the ISEG evaluations were open and quite helpful in providing requested information.
ISEG results were readily accepted and action taken to remedy identified problems.
Teamwork and communications appear to be a
BVPS-2 strength.
Procedural adequacy and compliance also appeared to be strongly followed and deficiencies, when identified, were actively resolved.
l Interfacing between organizations was evident, for example, in the Master Punchlist evaluation.
Collective efforts allowed the Master l
Punchlist to be an effective tool in ensuring that required work was accomplished during the progression of the startup phase.
l l
l 1
l l
APPENDIX D
SUMMARY
OF SHIFT TECHNICAL ADVISORY SURVEILLANCE PURPOSE Evaluate in-house evolutions / operation from 8/4/87 to 9/9/87 to assess the operation of Unit 2
in the areas of organizational interfaces, plant configuration
- control, procedural adequacy and compliance, teamwork and communications, Operations Quality Assurance surveillance effectiveness, responsiveness and timeliness of engineering and construction support
- services, control room operations and effectiveness of
- training, and adequacy of design based on test program, Technical Specifications, and the FSAR.
These surveillance were performed to determine if there were any common or recurring problems which would impact the safe and reliable operation i
l of Unit 2.
i ACTIONS TAKEN A
review of the problems noted for the surveillance performed (see following table) shows procedural inadequacies to be the category most frequently identified.
Problems were noted in other areas.
- However, these areas showed no significant trend which would impact I
the safe and reliable operation of Unit 2.
The procedural inadequacies found involved minor setpoint changes and differing revisions between controlled copies in a relatively small number of operating surveillance tests and abnormal operating procedures.
All probleras identified were immediately forwarded to the NSS and the appropriate supervisory personnel in the Operations Group for corrective action.
All actions taken were performed in a timely manner and it should be noted that the procedures were usable at all times.
I CONCLUSIONS As a
result of procedural inadequacies identified above, procedural revisions accounted for the majority of corrective actions taken.
A number of these problems were corrected immediately upon j
identification, while the remainder were forwarded to Operations supervisory personnel for discussions with the procedure writers and the clerical staff, for resolution.
These areas will be periodically monitored for corrective action implementation and effectiveness.
i 1
1 i
Appendix D BVSJ V
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1 i
APPENDIX E
)
d 1
SUMMARY
OF TESTING
, PURPOSE Testing performed during the period described in this report consisted of Initial Criticality Testing, Low Power Testing and Power Level Escalation Testing up to and including the 50%' Power Plateau.
The sequence of testing and the plant conditions under which tests were performed was controlled by a master test procedure entitled, "Startup Testing Program".
This procedure also ensured that the results of tests were reviewed prior to Mode changes 'and/or escalations in power level.
Furthermore, the test results of all Engineered Safeguards or reactivity control system tests were reviewed and approved by the Joint Test Group and Onsite Safety
)
Committee prior to Mode changes and/or power level escalations.
This requirement helped ensure operational readiness for safe and reliable i
station operations as the startup was performed in a controlled deliberate manner.
ACTIONS TAKEN Initial criticality was achieved on August 4, 1987 by a combination of shutdown and control bank withdrawal and reactor coolant system baron concentration dilutions.
Throughout this period, samples of the primary coolant were obtained and analyzed for boron o
concentration.
Low Power Testing was conducted to obtain reactor physics measurements which verified that basic static and kinetic characteristics of the core were as expected.
The measurements included:
verification of calculated values of control rod assembly group reactivity worths, of isothermal temperature coefficient under various core conditions, of differential boron concentration reactivity
- worth, and of critical boron concentration as a function of control rod assembly group configuration.
In
- addition, measurements of relative power distributions were made.
Concurrent tests were conducted on instrumentation including the source and intermediate range nuclear channels.
After the operating characteristics of the reactor and unit were verified by the low power
- testing, a
program of power level escalation began on August 16, 1987 at 1430 hours0.0166 days <br />0.397 hours <br />0.00236 weeks <br />5.44115e-4 months <br />.
Both reactor and unit operational characteristics were closely examined at each stage.
Measurements were made to determine the relative power distribution in the core as functions of power level and control assembly group position.
Secondary system heat balance calculations ensured that indications of power level were consistent and provided the basis for calibration of the power range nuclear channels.
The ability of the reactor control system to respond effectively to signals from primary and secondary instrumentation under a variety of conditions encountered in normal operations was verified.
At prescribed power levels the dynamic response characteristics of the reactor coolant and steam systems were evaluated.
Response of the.
systems was measured for design step-load
- changes, rapid load
'Page 2 APPENDIX E reduction, and plant trips.
Adequacy of radiation shielding was verified by gamma and neutrot radiation surveys at selected points l
inside containment and throughout the station site at the 5 and 50%
l power plateaus.
Periodic sampling was performed to verify the chemical and radio-chemical analysis of the reactor coolant.
Beginning with fuel
- load, the Onsite Safety Committee reviewed all test changes and intent changes were reviewed prior to I
implementation.
This process enhanced procedural compliance and adequacy.
During this period 114 test changes were made of which 58 were intent changes.
Test deficiencies and their resolutions, generated during this period I
were monitored by the Operational Assessment Committee, with particular emphasis placed on deficiencies against safety related systems.
The Startup Program required that deficiencies be reviewed as part of the test resulte review process.
Once again, this ensured l
- that, as a
- minimum, deficiencies involving Egineered Safeguards or l
reactivity control systems tests were reviewed prior to Mode changes l
and/or power level escalation.
None of the deficiencies generated l
were reportable to the NRC as defined in 10 CFR 50.72 and 73.
There were 139 deficiencies open at the time of initial criticality, all of l
which were evaluated and determined not to be detrimental to safe plant operation.
Sixty-five (65) new deficiencies were generated during the subject test period and 101 deficiencies were closed during this period, resulting in a total of 103 open deficiencies as of September 10, 1987.
Closure dates for these remaining deficiencies and any new deficiencies that may be generated are tied to the completion dates for the remaining test procedures.
Test deficiencies generated during this report period may be categorized as follows:
Safety Non-safety Systems (%)
Systems
(%)
Equipment Failure 6
24 Procedural Deficiencies 0
7 Procedure Acceptance Criteria Not Met 6
40 Personnel Error 1
2 Other (equipment not available to test, 1
13 etc.)
l It should be noted that guidelines for determining test procedure acceptance criteria were developed very early in the Startup Program.
When vendor criteria were not
- provided, conservative generic criteria are used.
Many times when deficiencies were generated for this
- reason, the criteria were met by performing maintenance,
- repair, recalibration, etc.
If the criteria still were not
- met, a deliberate engineering evaluation was performed to ensure that all technical requirements and margins of safety were maintained by accepting the results "as is".
Finally, it should be noted that I
the equipment failure category of deficiencies generally covers items that required adjustment, recalibration and/or repair.
As stated earlier, none of these items were reportable per 10 CFR 50.72 and 73.
1
w.
1
'Page 3 APPENDIN E 1
l 1
i l
l l
CONCLUSION l
l l
Organizational interfaces between Testing, Operations, Maintenance, l
Engineering, Quality Assurance and Licensing were very effective.
l The Startup Program has been able to respond to plant problems and still accomplish required plant testing in a timely manner.
l l
l l
l l
APPENDIX F l
SUMMARY
OF ENGINEERING AND CONSTRUCTION SUPPORT PURPOSE To assist the startup of Beaver Valley Unit 2 through responsive and timely support from engineering and construction services.
ACTIONS TAKEN Engineering Memoranda (EM) - a total of 457 ems have been issued e
with 50 outstanding and awaiting engineering completion.
Design Changes
- 704 Engineering and Design Coordination Reports and 13 Design Change Packages have been issued, all of which have received appropriate OSC review.
I l
Type 1
Drawings (Drawings required in the Control Room and e
All Type 1 Drawings are in the Emergency
Response
Facility)
Control Room and ERF.
1 VOND (Valve Operating Number Diagram)
Jointly developed by Engineering and Operations, these documents which cover all systems in the plant are 100% complete and issued to the Control Room.
The Nuclear Plant Reliability Data System list is NPRDS complete and being maintained by DLC Engineering.
BV-2 Records They are in the process of being turned over to DLC.
BV-2 Master Equipment List (MEL)
Over 41,000 items are currently listed and engineering data is being input to those items.
CONCLUSIONS Engineering and Construction have actively supported actions needed to address problems and changes encountered during the power ascension.
Response to Engineering Memorandums (ems) from operations has been timely and helped foster the communications and teamwork needed to keep the startup steadily progressing.
Engineering has continuously interfaced with all the groups involved with the startup program at Beaver Valley Unit 2.
Duquesne Light has also maintained a
cognizant Stone Webster organization onsite to address design concerns as they arose which has helped maintain a smooth startup and transition from Stone & Webster engineering during construction.
l
APPENDIK G I
SUMMARY
OF UNIT 2 ANNUNCIATOR TRENDS PURPOSE Because of an Operational Assessment Committee request, an evaluation of the number of annunciators alarmed in the Unit 2 control room from 8/4/87 to 9/9/87 was performed to determine if any problems are identified which could impact the operation of Unit 2.
This evaluation identified the total number of alarmed annunciators, the number of alarmed annunciators which were determined to be significant, the total number of out-of-service annunciators and the number of out-of-service annunciators which were determined to be significant.
These totals were trended on a weekly basis, with the results shown in the table below.
l ACTIONS TAKEN i
As shown in the table, the total number of alarmed and out-of -service annunciators has been classified as being significant or not significant.
The annunciators which involved safety-related systems / equipment and equipment required operable for the e.pplicable plant mode were classified as significant.
The annunciators involving equipment out-of-service, level alarms on tanks which were under the supervision of the operating crew, improper setpoints, and alarms which were expected for the applicable plant mode trere classified as not significant.
This same method of classification was used for the out-of-service annunciators.
A review of the table shows a
32%
decrease of alarmed annunciators during the time period l
trended.
A decrease of significant alarmed annunciators is also I
shown.
The out-of-service annunciator totals remain relatively the i
- same, due to the direction of the maintenance efforts toward the I
plant startup.
An additional review of the remaining significant l
alarmed annunciators shows that 50%
of'these are due to equipment being out of service, 25% are due to parameters being intentionally kept in the conservative direction and 25% are valid alarms.
Of all the significant and out-of-service annunciators that were or remain in
- alarm, no documented operational abnormalities were identified which could be attributed to these annunciators.
A task force has been formed to address engineering and operating procedure changes to l
reduce the number of lit annunciators.
CONCLUSIONS l
After initiation of this trend, a number of alarms were cleared by I
performing minor plant evolutions (filling tanks, draining tanks, l
opening / closing
- breakers, etc.),
while others were cleared by l
initiating computer software changes.
The alarms which could not be immediately cleared were written up under maintenance work requests l
to ensure completion.
]
l i
_---__-_--__N
APPENDIX G UNIT 2 ANNUNCIATOR TREND (cont'd.)
l 8/4/87 8/11/87 8/18/87 8/25/87 9/1/87 9/9/87 No. of Alarmed Annunciators 93 102 97 73 85 63 No. of Significant Alarmed Annunciators 16 15 14 11 11 13 No. of 00S Annunciators 14 14 14 14 15 15 No. of Significant OOS Annunciators 7
7 7
7 7
7 i
__.____._______.J
ys J
APPENDIX H i
SUMMARY
OF ONSITE SAFETY COMMITTEE (OSC)
ASSESSMENT OF UNIT 2 OPERATIONAL READINESS PURPOSE 1
l To evaluate and report to the Unit 2 operations assessment committee the operational readiness of Unit 2
regarding organizational interfaces, plant configuration
- controls, procedural adequacy, and adequacy of design based on the test
- program, Technical l
Specifications, and FSAR through OSC reviews of Unit 2 station procedures and design changes from 8/4/87 to 9/9/87.
l l
ACTIONS TAKEN The OSC held ten scheduled meetings between initial criticality (8/4/87) and 50% power operation and conducted 39 special meetings or polls.
A total of approximately 760 items related to new or modified procedure and design changes were reviewed and a breakdown into general categories is attached.
On an average, the OSC attendance consisted of an advisor from seven to eight of the nine areas described in the plant's technical specifications for the ten scheduled meetings.
The major trend or problem identified in this time frame involved a number of nonconformances by various groups to Technical Specification 6.8.3,
" Administrative Control for Temporary Procedure Changes".
Specifically, various groups failed to comply with the requirement that the proposed change is reviewed by the OSC and approved by the Plant Manager within 14 days of implementation, and in adapting to the administrative definition for non-intent changes.
There were no unreviewed safety questions identified and no identified disagreements between the OSC and the Plant Manager requiring notification of the Senior Manager of Nuclear Operations.
Another possible concern noted by the OSC involved the large number of special meetings or polls conducted (39).
Due to around-the-clock l
- testing, a
flexible testing schedule and plant equipment availability, the large number does not indicate that organization l
interface problems or procedural inadequacies exist.
Consistently, all initial test programs require many more test procedures and OSC involvement than during normal full power operation.
The majority of procedural revisions by the groups were due to administrative changes like the implementation of the full power technical specifications, l
adapting from a single unit to a dual unit administrative format, and test results.
l The OSC did not identify any adverse trends regarding procedural inadequacies but did see a
favorable trend regarding responsive organizational interfaces when procedural changes were required.
In
- addition, good overall teamwork by OSC members and alternates was exhibited by consistently exceeding the minimum quorum requirements at all scheduled meetings.
Page 2 APPENDIE H Based on these conclusions, Unit 2 has met or exceeded all cf the required attributes to demonstrate operational readiness for safe'and reliable station operation.
CONCLUSION The OSC has reviewed the corrective action by all groups that had difficulty with Technical Specification 6.8.3.
In all cases, the action initiated to correct the problem has and should continue to prevent similar incidents in the future.
The OSC has not identified a single recurrence since the corrective actions were implemented.
1 i
l l
?
i I
l h
--_m.-._____-_
l l
l APPENDIX H ATTACHMENT OSC REVIEW ITEM BREAKDOWN Total Items Total % Items DCPs,ACNs, WARS,UDCPs,E&DCRs,(Design Changes) 56 8%
Field Revisions of all Groups 213 30%
TRRs (Test Result Reports) 63 9%
Tests (SOVs,Pos,ISTs,BVTs) 96 13%
j Revisions (Procedures other than tests) 276 37%
Admin. (SUMS, SAPS) 14 2%
irs, LERs 3
1%
i TOTAL 721 l
Approximate No. Poll Items 39 TOTAL 760 100%
i 1
l i
l l
f j
l
l APPENDIX I l
SUMMARY
OF CHEMISTRY PURPOSE Chemistry supported the BVPS
- 2 initial startup activities with chemical
- analyses, chemical treatment of operating and shutdown
- systems, and recommendations to operations personnel.
The following is a
brief synopsis of chemistry conditions / prob; ems encountered during the initial startup through operation at 50% power.
ACTIONS TAKEN No problems were encountered in chemically treating and maintaining l
the Reactor Coolant System (RCS) and its auxiliary systems within chemical specifications.
Reactor coolant hydrogen and lithium were maintained within their respective control bands and chemical additions resulted in the expected chemical increase within the coolant system.
An unexpected delay did occur in initially placing the Chemical Volume and Control System (CVCS) filtera and domineralizers in service.
The delay was due to Operation's concern on affecting RCS boron concentration during initial core physics testing.
The CVCS demineralizers were placed in serviced prior to initial criticality.
Isotopic monitoring of the RCS showed no unusual results, with the exception of a
crudburst which occurred shortly after initial criticality consisting solely of Iron isotopes.
Later crudburst monitoring showed only expected isotopes of Cobalt and Manganese.
No Iron was detected after the initial crudburst.
The accumulation of other fission and activation products continued, as expected, as time at power accumulated.
Several problems were encountered maintaining secondary chemistry specifications.
The majority of the problems resulted from inadequate chemical treatment, initial startup of systems, condenser
- leaks, and inexperience with the chemical behavior of the condensate polisher.
Most of the problems involved inadequate chemical treatment.
Of the seven available treatment pumps, only one was operational for several periods during the startup.
The cause of the failures was diverse, ranging from malfunctioning tank level switches and pump discharge relief valves to improperly primed pumps.
Direct contact between chemistry and maintenance supervision was required to troubleshoot and correct the
- problems, and make all pumps available for operation.
Even with all pumps available for operation, difficulties were experienced in maintaining feedwater and steam generator pH.
Several mass balances were performed, and it was decided that too much hydrazine was being
- added, too little ammonium hydroxide was being
- added, and that virtually all the treatment was being removed by the condensate polishers.
Appropriate changes were made to the types of chemical treatment, and an ammoniated resin was ordered to reduce treatment losses across the polishers.
L
l
-Page 2 APPENDIX I i
l Two condenser leaks occurred during power operation.
The leaks, once diagnosed, were quickly located and
- repaired, and the condensate polishing system was used for rapid cleanup.
Diagnosis of the leaks was difficult,
- however, due to the return of untreated steam generator blowdown to the hotwell.
The blowdown demineralized
- system, scheduled for installation during the September outage, is l
expected to enhance the present capabilities of condenser inleakage detection.
Until the blowdown demineralizers are operational, the best tool for determining that condenser inleakage exists is l
condensate polisher run life.
i Most auxiliary systems had been chemically treated and placed in l
operation prior to hot functional testing and few problems were experienced during initial startup.
Problems did exist with the service and circulating water chlorination systems.
The chlorination system appears to be inadequate, as it presently exists, to complete a biological " kill" within the system during the 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per day chlorination permitted by the EPA.
System engineers are evaluating the problems, in the l
- interim, tanks,
- pumps, and piping were installed to pump liquid sodium hypochlorite into the circulating water system to augment the i
l present chlorination system.
l CONCLUSION Few problems have arisen with chemistry control of Beaver Valley #2.
i Startup of the steam generator blowdown demineralizers, coupled with the use of ammoniated resins in the condensate polisher, should l
correct the secondary pH control problem.
Use of liquid sodium
)
hypochlorite will mitigate the temporary inadequacy of the circulating water chlorination system.
l 1
i i
_m._~______.-___________.m
APPENDIK J i
SUMMARY
OF MILESTONEm MILESTONES DATE Low Power License (NPF-64) Issued May 27, 1987 Initial Fuel Load (Entry to Mode 6) Commenced May 29, 1987 l
l Initial Fuel Load Completed June 1, 1987 1
Mode 5 Entered (Tensioned Vessel Head Studs)
June 6, 1987 Containment Closed - Vacuum Established July 9, 1987 Mode 4 Entered July 15, 1987 l
l
)
Turbine Placed on Turning Gear July 16, 1987 l
Mode 3 Entered July 17, 1987 Mode 2 Entered August 3, 1987 Initial Criticality Achieved August 4, 1987 Low Power Physics Testing August 4, 1987 l
to August 6, 1987 Full Power License (NPF-73) Issued August 14, 1987 1
Mode 1 Entered August 15, 1987 Turbine Rolled, Generator Synchronized August 17, 1987 to Grid 10% Reactor Power Achieved August 17, 1987 30% Reactor Power Achieved August 22, 1987 30% Reactor Power Testing (Steady State)
August 27, 1987 to September 2, 1987 40% Reactor Power Achieved September 4, 1987 l
50% Reactor Power Achieved September 5, 1987 Maintenance Outage Entered September 11, 1987 Mode 5 Entered for Outage September 13, 1987 Maintenance Outage Completed (projected)
September 21, 1987
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