IR 05000410/1987028
| ML17055D164 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 08/13/1987 |
| From: | Eselgroth P, Wink L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17055D163 | List: |
| References | |
| 50-410-87-28, NUDOCS 8708280283 | |
| Download: ML17055D164 (22) | |
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-410/87-28 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:
Jul 20 - Jul
1987 Inspector: ~
~
~
~
L. J. Wink, Reactor Engineer, DRS 9 <a BP ate da e
Approved by:
8'.
W.
Ese gr h, Chief Test Progra s Section, DRS Ins ection Summar
Ins ection on Jul
Jul
1987 Re ort No. 50-410/87-28 Areas Ins ected:
Routine, unannounced inspection by one region-based inspector of the overall power ascension test program including test procedure review and test results evaluation, licensee action on previous inspection findings, engineering support of the PATP, gA interfaces, and independent measurements and verifications.
Results:
No violations were identified.
Note:
For acronyms not defined, refer to NUREG-0544,
"Handbook of Acronyms and Initialisms."
8708280283 870821 PDR ADOCK 05000410 G
DETAILS 1.0 Persons Contacted Nia ara Mohawk Power Cor oration
- R. Abbott, Station Superintendent G. Carlisle, Lead STD'ngineer J.
Conway, Power Ascension Manager
"P.
Eddy, Site Representative, New York State, PSC R. Gayne, Assistant Superintendent of Operations D. Helms, Lead Shift Test Supervisor G. Moyer, Station Shift Supervisor
"R. Neild, Technical Assistant to Station Superintendent D. Oakes, Startup Test and Operations Engineer H. Pao, Shift Test Supervisor
"A. Pinter, Site Licensing Engineer P. Wilde, Supervisor, Operations Surveillance (QA)
NRC Personnel
"W. Cook, Senior Resident Inspector C. Marschall, Resident Inspector
- W. Schmidt Resident Ins ector p
"Denotes those present at the exit meeting on July 24, 1987.
The inspector also contacted other members of the licensee's Operations, Technical, Test and QA staffs.
2.0 Licensee Action on Previous Ins ection Findin s
(Closed)
Unresolved Item (410/87-06-01)
Adequacy of planned testing for single reactor recirculation loop operation (SLO).
The licensee plans to test in four power-to-flow conditions to bound possible operation with a single reactor recirculation loop (see Inspection Report 50-410/87-16).
During the current inspection, the licensee responded to the inspector questions concerning the testing that would be required at these power-to-flow conditions for thermal expansion, the transversing incore probe (TIP) and the process computer.
The. licensee stated that additional thermal expansion testing will not be required based on General Electric analysis of the recirculation piping with an isolated (gate valves closed)
loop.
The analysis demonstrated that the increase in thermal stress does not result in exceeding any ASME Code or pipe break postulation criterion.
Based on this analysis and the successful completion of recirculation pipe thermal expansion testing during Test Condition Heatup, the inspector agreed that additional thermal expansion testing was not require The licensee stated that additional TIP testing to evaluate the uncer-tainties associated with SLO would not be necessary.
TIP uncertainty is composed of a geometric and a random noise component.
The total TIP uncertainty for two recirculation loop operations will be measured during the performance of power ascension test N2-SUT-18-6.
The geometric com-ponent is unaffected by the mode of recirculation operation and the effects on the random noise component, which is a function of neutronic, electronic and boiling noise, of SLO have been benchmarked using data obtained at Brown's Ferry Unit 1 and.have been included in the uncertain-ties used in the determination of the fuel cladding safety limit. The inspector agreed that additional TIP testing was not required.
The licensee stated that the process computer correctly calculates core flow (WT) from direct flow signals from the jet pump diffusers.
During SLO, the idle loop's jet pump diffuser flows are subtracted from the active loop's flows to account for reverse flow in the idle loop diffusers.
A potential problem exists since, during SLO, the process computer will compare this direct measure of core flow (WT) to an inter-nal array (MTSUB) which is currently based on two recirculation loop operation.
If WT does not agree with WTSUB within 5%, then MTSUB will be used to calculate thermal limits.
To prevent the process computer from incorrectly using WTSUB during SLO testing, the licensee proposes to manually enter the correct value of core flow (MT) into the process computer.
This will force the process computer to accept this value and will result in an accurate calculation of thermal limits.
The proposal will allow testing of SLO to be performed.
During this testing, the licensee will be able to obtain the data for SLO required to establish a
MTSUB array for SLO.
This array would be used in the future if operation in SLO were necessary.
The inspector agreed that this proposal was acceptable.
The inspector had no further questions on the scope of SLO testing.
This item is closed.
3.0 Power Ascension Test Pro ram PATP 3, 1 References Regulatory Guide 1.68, Revision 2, August 1978, "Initial Test Program for Mater Cooled Nuclear Power Plants."
"Administrative Controls and guality Assurance for Operations Phase of Nuclear Power Plants."
Nine Mile Point Unit 2 (NMP-2) Technical Specifications, July 2, 1987.
Nine Nile Point Unit 2 Final Safety Analysis Report (FSAR)
Chapter 14,."Initial Test Program."
Nine Mile Point Unit 2 Safety Evaluation Report.
Nine Mile Point Unit 2 AP-1.4, Startup Test Phase, Revision
3.2 Overall Power Ascension Test Pro ram The inspector held discussions with the Power Ascension Manager (PAM), the Lead Startup, Oesign and Analysis (STOM) Engineer and other members of the PATP staff to assess the status of testing, the test results evaluation process and the preparation and approval of test procedures.
In addition, the inspector attended the daily Power Ascension Management meetings and Site Operations Review Committee (SORC) meet.ings involving the PATP.
At the beginning of the inspection period, the unit was in the process of restarting and heating up to rated conditions.
The unit had scrammed on July 11, 1987 when the EHC hydraulic line to the ¹4 Main Turbine Control Valve failed and depressurized the EHC system.
This caused the bypass valve to fail close and resulted in increasing reactor pressure and a High Reactor Pressure scram.
The restart was delayed until July 19, 1987 due to service water intake temperatures exceeding technical specification limits and standby gas treatment system operability concerns related" to reactor building to service water temperature differential.
Ouring the inspection period, operations were limited to less than 5% of rated power while troubleshooting continued on the offgas system ( see discussion in Section 4.0).
The test results review of Test Condition Heatup was completed by the SORC on July 17, 1987 and accepted by the General Superintendent on July 21, 1987.
Approval to commence TC-1 testing was given in a SORC meeting on July 20, 1987 subject to the completion of all technical specification mode 1 sur-veillances and a satisfactory review of outstanding work items.
At the conclusion of the inspection, troubleshooting was continuing on the offgas system and unit operation remaining constrained to less than 5% of rated power.
3.3 Power Ascension Test Procedure Review
~Scn e
The procedures of Attachment A were reviewed for the attributes identified in Inspection Report Ho. 50-410/86-38, Section 4.3.
Discussion The procedures reviewed were new revisions of previously reviewed pro-cedures which reflect an ongoing licensee follow-up review of issued procedures.
~Findin s
The procedures reviewed were found to be acceptable.
No deficiencies were identifie Power Ascension Tests Results Evaluation
~Sco e
The power ascension test results listed in Attachment B and discussed below were evaluated for the attributes identified in Inspection Report No. 50-410/86-64, Section 2.1.
Discussion N2-SUT-14-HU RCIC S
stem A portion of these test results was reviewed during a previous inspection (Inspection Report 50-410/87-23);
During this inspection, the results of the CST injection testing at a reactor pressure of 150 psig were reviewed.
The test was performed at a reactor pressure of 156 psig.
The inspector verified from GETARS traces that the RCIC system was capable of reaching and maintaining rated flow (600 GPM) within 30 seconds.
The actual time was 24. 1 seconds.
The inspector also verified that the maximum turbine speed (2747 RPM) was less than the acceptance criterion limit of 4777 RPM (overspeed trip avoidance margin).
All other acceptance criteria were satisfied.,
I N2-SUT-16-HU Selected Process Tem eratures and Water Level Measurements Test Condition Heatu Portions of these test results were reviewed during a previous inspection (Inspection Report. 50-410/87-23).
During this inspection, the results of temperature stratification testing with minimum recirculation flow (mini-mum flow control valve position and recirculation pumps running on the LFMG sets)
and during single.recirculation loop operation were evaluated.
In addition, the inspector reviewed a reperformance of the water level calibration check.
This testing was repeated at the higher reactor building temperatures required to satisfy standby gas treatment system operability concerns.
The temperature stratification testing satisfied all acceptance criteria both for single recirculation loop operation and two recirculation loop operation at minimum flow.
The maximum temperature differentials measured (bottom head to reactor steam dome) were 15'F for single recirculation loop operation and 67'F for two recirculation loops at minimum flow (acceptance criterion < 145'F).
The reperformance of the water level calibration check at higher reactor building temperatures yielded similar results to the initial run but with small margins to the acceptance cri-terion limit s4
N2-SUT-77-HU BOP and Small Bore Pi in Vibration Steady state and transient piping vibration were measured for selected portions of the RCIC and CRD Systems.
Five test exceptions (TEs) were identified, one Level 1 and four Level 2.
The Level 1 test exception was identified during the CRD scram testing, at 0 psig reactor pressure, of rod 58-31.
The measured displacement of the CRD hydraulic line was
mils (acceptance. criterion 11 mils).
An investigation revealed that the acceptance criterion was based on the assumption of no movement at a
support point.
The actual support point consisted of a sliding clamp with a design gap of 1/16 inch.
To verify that this was the cause of the high vibration reading, a test was conducted to evaluate the piping response to a test engineer's manual shaking of the line.
The measured response was 42 mils.
Based on this information, the results of the test were accepted
"as-is".
The four Level 2 test exceptions (one for RCIC and three for CRD 58-31) were dispositioned as acceptable
"as-is."
N2-PP-HU Heatu Plateau Procedure The inspector reviewed this procedure to insure that all planned testing for Test Condition Heatup had been completed and that all open Test Exceptions could be safely carried forward to later test conditions.
The licensee has deferred three planned tests to later test conditions and has deferred the completion of the analysis of a fourth test, The testing deferred includes N2-SUT-70-HU, Reactor Mater Cleanup; N2-SUT-71-HU, RHR Suppression Pool Cooling Mode; and N2-SUT-74-HU, Offgas System.
RMCU testing could not be performed due to operational restric-tions imposed because of the feedwater temperature stratification problems.
The test will be run in Test Condition 2 following the corrective actions for the feedwater temperature stratification problem.
Currently this modification is scheduled for the outage following Test Condition 1.
The inspector reviewed the safety evaluation covering this deferral (SER 87-091 dated July 9, 1987).
The testing planned for the Suppression Pool Cooling mode of RHR could not be performed due to insufficient differential temperature between the service water and the suppression pool.
The temperature in the sup-pression pool was administratively limited to 90~F to allow the continuation of heatup testing while the RRCS was out of service for troubleshooting.
This testing is now planned for Test Condition l.
The testing planned for the Offgas System could not be performed due to operability problems with this system (see discussion in Section 4.0).
Testing will be performed following the correction of system problems and prior to exceeding 5% of rated power.
The analysis of the final portion of the BOP Thermal Expansion Test, N2-SUT-78-HU, was deferred to allow the comprehensive analysis of the test data.
All Level 1 acceptance criteria have been verified and all Level
test exceptions have been resolve The inspector also reviewed all open test exceptions from Test Condition Heatup.
Only six test exceptions remained open, four Level 2 exceptions and two other exceptions not directly related to acceptance criteria.
The inspector concluded that the identified tests and open test exceptions could be safely carried forward to subsequent test conditions.
~Findin e
All test results reviewed and "the resolutions of test exceptions were found to be acceptable.
No unacceptable conditions were identified.
4.0 En ineerin Su ort Durin the PATP
~Sco e
The inspector followed the troubleshooting efforts involving the Offgas System to assess the quality and effectiveness of the engineering support provided to resolve problems identified during the PATP.
The Offgas System was selected since the inability to place this system in service has significantly delayed the PATP.
Discussion Efforts to place the Offgas System in service have been in process for at least two weeks prior to the start of the current inspection.
The lead responsibility had been assigned to Operations, with Station Shift Super-visors routinely working 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of overtime each day to direct this effort.
During the current inspection period, engineering support was provided in an attempt to expedite the identification and resolution of problems being experienced with the system.
On July 21, 1987, the inspector attended an engineering coordination meeting called to assess progress to date and assign responsibilities and establish priorities for the resolution of identified problems.
The engineer-in-charge used a preexisting list of possible problems with the system in an attempt to determine the current status.
Disagreements were noted among various participants concerning the significance and cur-rent status of problems noted, These disagreements arose because of the anecdotal nature of the observation with an apparent lack of documentation of the troubleshooting actions taken and results achieved.
The meeting resulted in a list of twenty (20) items requiring resolution.
Priorities and responsibilities were assigned for each item.
Engineering responsi-bilitiess were limited to verification of design.
Five hardware related items, including steam supply pressure to the recombiner preheaters, recombiner condenser level control, freeze-out dryer freon leaks and replacement of an inoperable moisture sensing element were identified as required to be complete before another attempt was made to place the system in service;
On July 22, 1987, Operations personnel attempted to place the system in service.
The engineer-in-charge was present in the control room for this attempt.
System performance was unsatisfactory with little evidence that the steam jet air ejectors (SJAEs)
were effective in removing noncon-densible gases from the main condenser.
Operations personnel performed a system walkdown and identified a piping configuration problem at the discharge of the second stage SJAEs.
The configuration resulted in a loop seal which inhibited the effective functioning of the jets.
Opera-tions personnel proposed the addition of drain valves, to be used for system startup, to remove the water which had collected in the pipe.
This would be a temporary solution pending correction of the poor piping configuration.
Engineering was able to expeditiously generate a modifi-cation package to install these valves.
At the conclusion of this inspection on July 24, 1987, the new startup drains had been installed but an attempt to place the system in service revealed additional problems including air in-leakage from an unidenti-fied source, possible tube leakage in the SJAE intercondenser and a valve operability problem in the blowdown line of the recombiner preheater.
The inspector will continue to follow the Offgas System problems and engineer-ing support of their resolut'ion during a future routine inspection.
~Findin s
No violations were identified.
A Interface With the PATP The inspector reviewed the gA Surveillance Reports listed below:
gASR-87-10425,
"Heatup Plateau Test," dated 7/20/87 gASR-8?-10432,
"RCIC System Test Performance,"
dated 7/20/87 gASR-87-10435,
"MSIV Test Performance,"
dated 7/8/87 gASR-87-10514,
"HSIV Test Results Review," dated 7/7/87 gASR-87-10541,
"RCIC System Test Results Review," dated 7/15/87 The inspector verified that the surveillances were performed in accordance with applicable (}A procedures and the commitments made in the Surveillance Plan for the Power Ascension Test Program.
No deficiencies were identified during this revie.0 Inde endent Measurements and Yerifications During the evaluation of the results of power ascension test, N2-SUT-14-HU, RCIC System, as discussed in paragraph 3.4, the inspector independently calculated the time for the system to reach and maintain rated flow and the margin to the overspeed trip setpoint, using GETARS traces, and verified that the associated acceptance criteria were satisfied.
The inspector's measurements and verifications agreed with the licensee's.
No unacceptable conditions were noted.
7.0 Exit Interview At the conclusion of the inspection on July 24, 1987, an exit meeting was held with licensee personnel (identified in Section 1.0) to discuss the inspection scope, findings and observations as detailed in this report.
At no time during the inspection was written material provided to the licensee by the inspector.
Based on the NRC Region I review of this report and discussions held with licensee representatives during the inspection, it was determined that this report does not contain information subject to
CFR 2.790 restriction ATTACHMENT A POWER ASCENSION TEST PROCEDURES REVIEWED N2-SUT-11-1 N2-SU T-13-1 N2-SUT-'19-1 N2-SUT-23-1 N2-SUT-29-1 N2-SUT-33-1 LPRM Flux Response, Revision 2, Approved= February 4,
1987 Process Computer Test Condition 1, Revision 1, Approved February 10, 1987 Core Performance
- Test Condition 1, Revision 1, Approved April 16, 1987 Feedwater System, Revision 1, Approved February 18, 1987 Recirculation Flow Control Test Condition 1, Revision 2, Approved February 10, 1987 Drywell Piping Vibration - Test Condition 1, Revision 1,
Approved December 23, 1986
ATTACHMENT B POWER ASCENSION TEST RESULTS EVALUATED N2-SUT-14-HU NZ-SUT-16-HU N2-SUT-77-HU N2-PP-HU RCIC System, Revision 4, results accepted July 15, 1987 Selected Process Temperatures and Water Level Measurements Test Condition Heatup, Revision 3, results accepted July 15, 1987 BOP and Small Bore Piping Vibration, Revision 2, results accepted July 15, 1987 Heatup Plateau Procedure, Revision 0, results accepted July 21, 1987