IR 05000410/1988014
| ML17055E045 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 07/18/1988 |
| From: | Marilyn Evans, Lange D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17055E044 | List: |
| References | |
| 50-410-88-14, NUDOCS 8808020291 | |
| Download: ML17055E045 (26) | |
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-410/88-14 Docket No.
50-410 License No.
NPF-69 Licensee:
Nia ara Mohawk Power Cor oration 301 Plainfield Road S racuse New York 13212 Facility Name:
Nine Mile Point Nuclear Station Unit 2 Inspection At:
Scriba New York Inspection Conducted:
June 6-10 1988 Inspector:
1~Uu-4. A.
M. Evans, Operations Engineer date Approved by D.
Lange, Ch f,
BWR S
tion Operations Branch, DRS date Ins ection Summar:
Ins ection on June 6-10 1988 Re ort No. 50-410/88-14 A~.
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d(
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of licensee action on previous inspection findings, and of the power ascension test program following the completion of testing activities including overall program, plateau review, test results evaluation and gA interface.
Results:
No violations were identified.
The licensee has satisfactorily completed the Nine Mile Point Nuclear Station, Unit 2, Power Ascension Test Program.
Note:
For acronyms not defined,,refer to NUREG-0544,
"Handbook of Acronyms and Initialisms."
8808020291 88072Q PDR ADOCN, 05000410
PNU
DETAILS 1.0 Persons Contacted Nia ara Mohawk Power Cor oration
- S. Agarwal, Lead Licensing Engineer
"J. Beratta, Manager, Security
- J. Conway, Power Ascension Manager
- P. Eddy, Site Representative, NYSPSC
- P. MacEwan, Site Representative, NYSEG
- R. Nield, Technical Assistant to Station Superintendent
- A. Pinter, Licensing Engineer
"R. Smith, Operations Superintendent
~P.
Wi lde, gA Surveillance Supervisor NRC Personnel, WE Cook, Senior Resident Inspector
'W. Schmidt, Resident Inspector The inspector also contacted other members of the licensee's Operations, Technical, Test and gA staffs.
2.0 Licensee Action on Previous Ins ection Findin s
(Closed) Violation (410/86-16-02)
Licensee performed a test to deter-mine the effect of isolation of a single feedwater line on feedwater temperature stratification without a written procedure and without the performance of a written safety evaluation to determine that an unreviewed safety operation did not exist.
The licensee took immediate corrective action on June 4,
1987 through performance of a safety evalu-ation to address the issue of operating with one feedwater line.
During the current inspection, the inspector reviewed the following documents:
NMPC IOC NMP24719 from R.
B. Abbot to SSS's and ASSS's dated June 5,
1987.
NMPC IOC NMP24770 from R.
B. Abbot to SSS's, ASSS's and Site Engineers dated September 10, 1987.
NMPC IOC NMP28572 from K.
P. Buckly to J. J.
Bebko dated October 5,
1987.
NMP-2 Operations Lesson Plan for Licensed Operator Requalification Training Cycle VIII Schedule - 1987, dated October 6, 1987.
Training Modification Recommendation TMR ¹02-87.227, dated October 1,
198 The inspector discussed the corrective actions with a licensee operation's department representative.
As part of the corrective actions, documenta-tion regarding this issue was included in the operation's department Lessons Learned Book.
However, the representative stated that during a
recent reformatting of the Lessons Learned Book, the documentation was deleted.
The representative stated, that based upon the significance of the event, a
summary was being written and would be incorporated in the current Lessons Learned Book.
Based upon review of the above documents, verification that all licensed operators had received -the requalification training and discussions with the licensee representative, this violation is closed.
3.0 Power Ascension Test Pro ram PATP 3.1 References
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Regulatory Guide 1.68, Revision 2, August 1987, "Initial Test Program for Water Cooled Nuclear Power Plant."
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"Administrative Controls and Quality Assurance for Operations Phase of Nuclear Power Plants."
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Nine Mile Point Unit 2 (NMP-2) Technical Specifications, July 2, 1987.
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Nine Mil'e Point Unit 2 Final Safety Analysis Report (FSAR)
Chapter 14, "Initial Test Program."
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Nine Mile Point Unit 2 Safety Evaluation Report.
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Nine Mile Point 2 AP-1.4, Startup Test Phase, Revision 5.
3.2 Overall Power Ascension Test Pro ram The inspector held discussions with the Power, Ascension Manager (PAM) and other members of the PATP staff to assess the licensee's closeout of the PATP.
The last power ascension test (N2-SUT-78-6, BOP Pipe Expansion)
was completed on March 8, 1988.
The final plateau review for 100% rod line testing was formally accepted by the Station Operations Review Committee on March 11, 1988 (meeting No. 88-40)
and by the General Superintendent on March 25, 1988.
The inspector reviewed N2-PP-4, Test Plateau 4 Procedure (Test Conditions 5 and 6 ), Revision 0, and determined that all planned testing had been completed.
In addition, the inspector verified that all test exceptions documented in N2-PP-4 had been satis-factorily resolve ~Findin s
No unacceptable conditions were identified.
The inspector concluded that the licensee had satisfactorily completed the Nine Mile Point Nuclear Station, Unit 2, Power Ascension Test Program.
This inspection represents the final NRC Region I review of the Power Ascension Test Program.
3.3 Power Ascension Test Results Evaluation
~Sco e
The power ascension test results listed in Attachment A and discussed below were evaluated for the attributes identified in Inspection Report No. 50-410/86-64, Section 2. 1.
Discussion N2-SUT-1-6 Chemical and Radiochemical Reactor water chemistry was within Technical Specification limits.
Three Level 1 test exceptions were identified for certain water quality standards for fuel warranty not being met.
Test results indicated dissolved oxygen and metallics slightly outside Fuel Warranty Acceptance criteria.
Results accepted as is based upon G.E. analysis.
Chemistry department to pursue long-term improvements.
N2-SUT-2-6 Radiation Measurements This procedure was performed to monitor radiation levels in the plant during TC-6 and during TIP System operations.
Two Level 1 test exceptions written for areas monitored in which radiation levels were in excess of acceptance criteria.
Areas are being administratively controlled in accordance with 10 CFR 20 requirements until rezoning or shielding modifications can be implemented.
Work is being tracked by problem reports 07728, 07729, 07730, 07731 and 07734.
N2-SUT-5-6 CRO S stem This test was performed in conjunction with the planned scrams during the Generator Load Rejection Test (N2-SUT-27-6)
on March 5, 1988 and the MSIV.Full Isolation Test (N2-SUT-25-6)
on February 15, 1988.
The scram times of the 4 selected rods (coordinates 42-31, 26-39, 54-43 and 30-59) were measured and verified to be well within the accept-ance criterion and Technical Specification limit of 7.0 seconds for each tes N2-SUT-11-6 LPRM Calibration The LPRM calibration was performed at 98.5% reactor power using the process computer.
Acceptance criteria were met with each LPRM reading within 10% of its calculated value.
N2-SUT-12-6 APRM Calibration The APRMs were calibrated by means of a heat balance performed by the process computer.
The test was performed on February 27, 1988 at 98. 1% reactor power.
All acceptance criteria were satisfactorily met.
N2-SUT-13-6 Process Com uter Acceptance Criteria Verification verified that process computer was correctly calculating core thermal limits by verifying that MCPR, MLHGR, MAPLHGR and LPRM GAF agreed-within 2% with those calculated by the off line program BUCLE.
Two test exceptions not related to acceptance criteria were identified and subsequently resolved through retests.
N2-SUT-16-6 Water Level Measurements With the reactor at rated conditions in the TC-6 window, the B Loop Recirculation Pump was tripped to verify reactor vessel temperature non-stratification.
All acceptance criteria were met.
In addition, narrow, wide and upset range level indications were compared at steady state reactor conditions for various flow rates.
No exceptions were identified.
N2-SUT-18-6 TIP Uncertaint The acceptance criterion regarding reproducibility of the TIP System readings was met.
Calculated total TIP uncertainty was 1.73% (criterion < 6.0%).
N2-SUT-19-6 Core Performance Verification of the core thermal hydraulic limits was performed at 98. 1% of rated thermal power and 99.2% of rated core flow.
All acceptance criteria were satisfied with the results as follows:
Parameter LHGR (KW/ft)
Measured Value 11.67 1.479 10.53 Limit s 13.4 1.25
< 12.01
No unacceptable conditions were identified.
N2-SUT-22-6 Pressure Re ulator This test was performed at 98.0% of rated thermal power to demonstrate adequate response and stability with the main turbine control valves alone controlling the transient and with the main turbine bypass valves alone controlling the transient.
Two level
test exceptions were identified and successfully resolved.
N2-SUT-23-6 Feedwater S stem All Level 1 test acceptance criteria for the Feedwater System were met.
A Level 2 and a Level 3 test exception were identified, reviewed by GE and accepted as is.
N2-SUT-24-6 Turbine Valve Surveillance Response of the reactor at several test points was observed and the maximum power level for performance of surveillance tests for turbine control and stop valves along the 100% power flow control line was established.
Test successfully performed with no test exceptions.
N2-SUT-25-6 Full Reactor Isolation This test was performed on February 16, 1988 at 95.3% reactor power.
All acceptance criteria were satisfied.
All required safety systems functioned properly without manual assistance.
MSIV stroke times were well within the 3-5 second criteria.
N2-SUT-27-6 Generator Load Rejection This test was performed on March 5, 1988 with the reactor at 99.6%
power.
The generator load rejection was initiated by simulating a
345 KV line high differential voltage/current with relay 87A-I/L23.
The acceptance criteria for turbine control valve and stop valve closure and bypass valve opening were satisfied.
A Level 1 test exception was written for recirculation pump failure to meet flow coastdown criteria.
A GE analysis was performed and the observed coastdown rate was found acceptable.
Several other acceptance criteria test exceptions were identified because of the loss of feedwater caused by the fast transfer of house loads during the load rejection not occurring as expected.
These test exceptions were analyzed by GE and accepted as is, based upon review of water
level and feedwater flow response during previous Power Ascension Tests.
Modification 88MX014 was completed to assure that the fast transfer of house loads occurred as required.
No unacceptable conditions were identified.
N2-SUT-29-6 Recirculation Flow Control Level 1 test criteria were satisfactorily verified.
Several Level
test exceptions were identified involving delay times and overshoot for position and flow demand steps.
All exceptions were analyzed and accepted as is.
N2-SUT-30-6 Reactor Recirculation S stem This test successfully demonstrated that the feedwater control system satisfactorily controls water level during the recovery from the trip of one recirculation pump without causing a turbine trip/scram.
Two test exceptions were identified and satisfactorily resolved.
N2-SUT-33-6 Or well Pi in Vibration This test verified reactor recirculation system piping vibration to be within design criteria during steady state conditions and operating transient load testing.
A Level 1 and a Level 2 test exception were identified and accepted as is based upon engineer-ing evaluations that the exceptions were caused by malfunctions or damaged instrumentation.
N2-SUT-35-6 Recirculation S stem Flow During this test, installed recirculation system flow instrumentation was calibrated at conditions between 95-100% rated thermal power.
Two minor Level 2 test exceptions were identified and resolved.
N2-SUT-71-6 Residual Heat Removal S stem The shutdown cooling mode of RHR Loop "A" was successfully tested during plant shutdown following the MSIY closure scram (SUT-25-6)
on February 16, 1988.
Loop "B" was tested following the Turbine Generator Load Rejection Test (SUT-27-6)
on March 5, 1988.
For both tests, the RHR system was proven capable of operating in the shutdown cooling mode at a heat exchanger capacity equivalent or greater than design.
No exceptions were written.
N2-SUT-74-6 Off as S stem This test was performed at 95.76% of rated thermal power.
The Level 1 acceptance criterion was easily satisfied with the measured release of radioactive gaseous and particulate effluents a small percentage of the Technical Specification limits.
Test exceptions
were noted involving instrumentation (moisture element, hydrogen analyzer, flow rate measurement)
and an unidentified source of system inleakage.
Test exceptions accepted as is based upon written justification by engineering and engineering's plans for a followup modification to the hydrogen monitor and an inleakage reduction program.
N2-SUT-75-6 Dr well Coolin S
stem This test was performed at rated steady state conditions of reactor temperature and pressure within the TC-6 envelope and following the transient induced during the load rejection test.
During the steady state testing, the Level 1 acceptance criterion for average drywell temperature was easily satisfied with a measured value of 108'F (Design
< 150'F).
Three Level 2 test exceptions were identified including the maximum drywell area temperature at a point in the upper area of the drywell exceeding the 150 F limit (actual 168.2'F)
and two points in the drywell reaching 236 F and 226 F within five minutes after the scram during the load reject test.
These excep-tions were evaluated by engineering and accepted as is for the short term.
Engineering has identified the equipment which could be affected by the high temperatures and during the October 1988 outage will add additional thermocouples near the equipment and evaluate the need for an accelerated equipment replacement schedule.
N2-SUT-76-3 ESF Area Coolin Completion of this test was deferred from TC-3 to allow conduct of retests which could only be done during monthly and quarterly surveillance tests.
Ten test exceptions wer e identified during conduct of this test including two Level 1 and six Level
exceptions.
Problems encountered included SBGT Building Filter Rooms and Division II Diesel Generator Control Room temperatures not maintained above design limits and several unit coolers not meeting their performance criteria.
Reasons for the exceptions included incorrect service water valve lineup and improper instru-mentation setup and data collection method.
All exceptions were analyzed by engineering and resolved with satisfactory retests performed as necessary.
N2-SUT-76-6 ESF Area Coolin The Levels 1 and 2 acceptance criteria for maintaining the ESF area temperatures below design limits were satisfied for both RHR A and B
Pump Rooms.
N2-SUT-77-6 BOP and Small Bore Pi in Vibration Two Level 1 test exceptions were identified for transient vibration.
These involved the main steam system during SRV lift (MSIV Full Isolation Test)
and during the turbine generator load reject tes Both exceptions were analyzed by engineering and satisfactorily resolved.
N2-SUT-78-2 BOP S stem Ex ansion The completion of this test was deferred from TC-2 to TC-3 because the Main Steam Reheaters had not been placed in service.
Comple-tion was then deferred from TC-3 because the plant was not shutdown to allow the return to ambient measurements.
The system monitored was the main steam piping to the reheaters.
The test was performed twice, once at rated conditions and again at ambient shutdown conditions.
One Level 1 test exception was identified during the ambient walkdown when a snubber was observed in direct contact with a support beam.
A work request was written to move the snubber clamp and provide clearance.
The retest was satisfactory.
Three Level 2 exceptions were also identified and adequately resolved by engineering.
N2-SUT-78-6 BOP S stem Ex ansion Systems monitored included condensate, feedwater, RHR and RCIC.
Twenty-nine test exceptions were identified.
All exceptions were evaluated by engineering with 28 accepted as is and one rework to provide additional clearance.
No unacceptable conditions were identified.
N2-SUT-79-6 Reactor Internals Vibration All Level 1 acceptance criteria were met.
One Level 2 test exception was identified because two strain gage sensors on the jet pump riser braces exceeded the acceptance criteria of 10,000 psi.
GE analyzed, these results and concluded that no adverse effects on the plant would occur for a total of 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> of operation at minimum flow condition.
Additional analyses by,GE is currently ongoing for greater than 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> of operation at minimum flow conditions.
Operations department is maintaining a log of hours of operating in conditions causing the Level 2 stress conditions.
N2-SUT-80 Emer enc Recirculation Ventilation The ability of the Reactor Building Emergency Recirculation Ventilation system to maintain required reactor building area temperatures within design limits under, postulated accident con-ditions was verified.
One Level 1 test exception was identified for several temperatures on the 1?5', 196'nd 240'levations being below the design minimum of 70 F.
Results were evaluated by engineering and accepted as i N2-SUT-81-6 Penetration Coolin All acceptance criteria for drywell penetration temperatures were successfully verified.
N2-PP-3 Test Plateau 3 Procedure The inspector reviewed this procedure to insure that all planned testing for Test Condition 3 was completed and to verify that open exceptions were identified and carried over to the next Plateau (Test Condition 5 and 6).
The inspector noted that portions of three tests had been deferred to Plateau 4.
(N2-SUT-76-3, N2-SUT-77-3 and N2-SUT-78-2).
N2-PP-4 Test Plateau 4 Procedure The inspector verified that Test Plateau 3 open items had been resolved and that all testing for Test Conditions 5 and 6 was complete.
The inspector also reviewed all test exceptions from Test Condi-tions 5 and 6.
All ninety-five (95) exceptions identified were satisfactorily resolved.
The more significant exceptions for Test Condition 6 are identified in the above discussions of specific tests.
~Findin s
No unacceptable conditions were identified.
Chan es to Initial Startu Test Pro ram During review of N2-SUT-23-6, Feedwater System, the inspector noted that the portion of the test requiring performance of manual flow steps on Test Condition 6 had been deleted.
The inspector discussed the deletion with the PAM.
In addition, the inspector reviewed Safety Evaluation (SE)
No.88-019, (SORC approved March 9, 1988)
and the GE analysis justifying deletion of the test.
The inspector found the justification for deletion to be acceptable.
However, upon review of the SE and further discussion with the PAM, the 'inspector noted that a required change to FSAR Table 14.2-222 had never been implemented.
The PAM stated that licensing had made a decision not to change the FSAR table because it was a minor change (one word).
License Condition C(6) requires that any changes to the Initial Test Program described in Section
of the FSAR be made in accordance with 10 CFR 50.59 and reported within one month of such change.
The PAM stated that he would contact licensing to incorporate the change to the FSA This issue is considered unresolved (50-410/88-14-01)
pending implementation of the FSAR change and verification that all changes to the Initial Test Program as described in Section 14 of the FSAR were made and reported as required.
4.0 uality Assurance Interface with the PATP The inspector reviewed QA Surveillance Reports 88-10048, 88-10193, 88-10172, 88-10201 and 88-10202.
The inspector verified that the sur-veillances were performed in accordance with applicable QA procedures and the commitments made in the surveillance plan for the power ascension test program.
In addition, the inspector verified that all comments generated in the QASR's were adequately resolved.
No unacceptable conditions were noted.
5.0 Unresolved Items Unresolved items require additional information to determine whether they are acceptable, an item of noncompliance or a deviation.
An unresolved item identified during this inspection is discussed in Paragraph 3.4.
6.0 Exit Interview A management meeting was held at the conclusion of the inspection on June 10, 1988, to discuss the inspection scope, findings and observa-tions as detailed in this report.
(See Paragraph 1 for attendees'
)
No written information was provided to the licensee at any time during this inspection.
The licensee did not indicate that any proprietary information was contained within the scope of this inspectio ATTACHMENT A POWER ASCENSION TEST RESULTS EVALUATED N2-SUT-1-6 N2-SUT-2-6 N2-SUT-5-6 N2-SUT-11-6 N2-SUT-12-6 N2-SUT-13-6 N2-SUT-16-6 N2-SUT-18-6 N2-SUT-19-6 N2-SUT-22-6 N2-SUT-23-6 N2-SUT-24-6 N2-SUT-25-6 N2-SUT-27-6 N2-SUT-29-6 Chemical and Radiochemical, 95-100%
Power Test, Revision 0, Results Accepted March 11, 1988 Radiation Measurements TC-6, Revision 1, Results accepted March 10, 1988 CRO System, Revision 1, Results Accepted March 10, 1988 LPRM Calibration TC-6, Revision 1, Results Accepted March 8, 1988 APRM Calibration TC-6, Results Accepted March 8, 1988 Process Computer TC-6, Revision 0, Resul'ts Accepted March ll, 1988 Water Level Measurements and Selected Process Temperatures TC-6, Revision 2, Results Accepted March 8, 1988 TIP Uncertainty, Revision 1, Results Accepted March 8, 1988 Core Performance TC-6, Revision 1, Results Accepted March 8, 1988.
Pressure Regulator TC-6, Revision 1, Results Accepted March 10, 1988 Feedwater System, Revision 0, Results Accepted March ll, 1988 Turbine Valve Surveillance TC-6, Revision 0, Results Accepted March 8, 1988 Full Reactor Isolation TC-6, Revision 0, Results Accepted March 8, 1988 Generator Load Rejection, Revision 0, Results Accepted March 1),
1988 Recirculation Flow Control TC-6, Revision 1, Results Accepted March 11, 1988
Attachment A
N2-SUT-30-6 N2-SUT-33-6 N2-SUT-35-6 N2-SUT-71-6 N2-SUT-74-6 N2-SUT-75-6 N2-SUT-76-3 N2-SUT-76-6 N2-SUT-77-6 N2-SUT-78-2 N2-SUT-78-6 N2-SUT-79-6 N2-SUT-80 N2-SUT-81-6 N2-PP-3 N2-PP-4 Reactor Recirculation System,.Revision 1, Results Accepted March 10, 1988 Drywell Piping Vibration TC-6, Revision 2, Results Accepted March 10, 1988 Recirculation System Flow Cali,bration, Revision 1,
Results Accepted March 10, 1988 Residual Heat Removal System, Revision 1, Results Accepted March 8, 1988,.
Offgas System TC-6, Revision 2, Results Accepted March 10, 1988 Drywell Cooling System, Revision 2, Results Accepted March 11, 1988 I
ESF Area Cooling System TC-3, Revision 2, Results Accepted March 11, 1988 ESF Area Cooling TC-6, Revision 1, Results Accepted March 11, 1988 BOP and Small Bore Piping Vibration, Revision 1,
Results Accepted March 10, 1988 BOP System Expansion, Revision 0, Results Accepted March 8, 1988 BOP System Expansion, Revision 2, Results Accepted March 10, 1988 Reactor Internals Vibration Measurements, 100% Load Line, Revision 2, Results Accepted March 10, 1988 Emergency Recirculation Ventilation, Revision 1,
Results Accepted March 11, 1988 Penetration Cooling TC-6, Revision 1, Results Accepted March 8, 1988 Test Plateau 3 Procedure, Revision 0, Results Accepted January 21, 1988 Test Plateau 4 Procedure, Revision 0, Results Accepted March ll, 1988