IR 05000410/1987035
| ML17055D370 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 11/05/1987 |
| From: | Lange D, Wink L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17055D368 | List: |
| References | |
| 50-410-87-35, NUDOCS 8711180169 | |
| Download: ML17055D370 (24) | |
Text
'0 U'.S.
NUCLEAR REGULATORY,'COMMISSION
.- -REGION I
'Report No.
d 30-410/87-35 Docket No..
50-41'0 License No.
NPF-69 Licensee:
Nia ara Mohawk Power"Gor oration 301 Plainfield Road S racuse New York Fac'ility Name., Nine Mile Point Nuclear Station Unit 2 Inspection At:
- Scriba New York.
'nspection'onducte Se tember '21-23 and October 3-9 '"'1987 Inspectors:
L. J. Mink, Operations ngineer te Approved by:
O.
.
ange, C ief BWR Section, ORS date Ins ection Summar
Ins ection on Se tember 21-23 and October 5-9 1987 Re ort No. 50-410/87-35 A~I:
R l, dl l
l-R d
inspector of the overall power ascension test program including procedure review, test witnessing and test results evaluation, 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."
t 87iii80169 87iii0 PDR ADQCK 050004i0
Details.
1.0 -'ersons Contacted Nia ara Mohawk'Power Cor oration R. Abbott, Station Superintendent
...,,G..Carlisle, Lead STOMA Engineer J.
Conway, Power Ascension Manager A. DeGracia, Station Shift Supervisor
-* P.
Eddy,- Site 'Representative, New 'York State PSC-R.:,Gayne, Assistant Superintendent of Operations
- '. Helms, Lead Shift Test Supervisor
- P.. MacEwan,.Site Representative, NYSEG
- R.'eild, 'Assistant to the Station Superintendent
" A.-'inter, Licensing Engineer B. Rudd, Shift Test. Supervisor
- R. Smith, Superintendent of Operations
'Imt. Wambsgan, Assistant Superintendent of Operations NRC Personnel
- W. Cook, Senior Resident Inspector C. Marschall, Resident Inspector W. Schmidt, Resident Inspector
" Denotes those present at the exit meeting on October 9,
1987.
The inspector also contacted other members of the Licensee's Operations, Technical, Test and gA staffs.
2.0 Power Ascension Test Program (PATP)
2.1 References Regulatory Guide 1.68, Revision 2, August 1978, "Initial Test Program for Mater Cooled Nuclear Power, Plants."
ANSI N18.7-1976, "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 Mile Point Unit 2 Final Safety Analysis Report (FSAR)
Chapter 14, "Initial Test Program."
Nine Mile Point Unit 2 Safety Evaluation Report.
Nine Mile Point 2 AP-1.4, Startup Test Phase, Revision =- -. -3-2.2 Overall Power Ascension Test Pro ram The"inspector.:held di.scussions with the power ascension manager (PAM), the lead startup test, design -and analysis (STD&A) 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.
On'eptember 3,
1987; the unit was"shutdown to correct condenser tube leaks and other problems identified during initial testing in Test Condition 2.
At the beginning of the inspection period the
'utage was continuing and preparations were under way to conduct a
.
..cold.shutdown loss of offsite. power test
[N2-SUT-31-2(SD).
Loss of
"Turbine-Generator and Dffsite'Power-Cold Shutdown.'ase].
The licensee had elected to. perform this additional testing to insure that overall plant response was acceptable prior to conducting the
'..test at power.
In addition, the testnwould serve as. a <raining exercise and final proof test of the procedures to be used.
The test was performed on September 23, 1987 (See paragraph 2.4 for discussion).
.f'ol.lowing aompletion of the cold shutdown loss of offsite power test and correction of problems-identified=during the test, the unit'was re-started on September 29, 1987.
On October 1,
1987, with the reactor at 500 PSIG and approximately 1% power, a reactor scram occurred on high IRM flux due to a cold water injection transient.
The reactor was restarted on October 2, 1987, and testing resumed in Test Condition 2.
At the conclusion of the inspection period, the unit was at 43% of rated thermal power and testing was continuing in Test Condition 2.
2.3 'ower Ascension Test Procedure Review
~Sco e
The procedures of Attachment A were reviewed for the attributes identified in Inspection Report No. 50-410/86-38, Section 4.3.
Discussion The procedures reviewed were generally acceptable.
The inspector had several questions and comments concerning these procedures which were satisfactorily addressed during discussions with the PAM.
Find~ mis No unacceptable conditions were identifie "2:4 Power Ascension Test Witnessin
~Sco e
The inspector witnessed the performance of the power ascension test discussed below.
The performance of this test was witnessed to verify the attributes previously defined in Inspection Report No.
50-410/86-64, Section 2.3.
Discussion N2-SUT-31-2 SD Loss of Turbine-Generator and Offsite Power-Cold Shutdown Case This test was performed on September 23, 3.987, with the reactor in cold shutdown (Operational Condition 4) to verify that all systems that would be required for the actual loss of offsite power demonstration were functioning properly and that the proposed test procedure was adequate to accomplish the objectives of the test.
Prior to the start of the test, the inspector witnessed the verification of prerequisites and the briefing of operations personnel and test engineers by the station shift supervisor and
. shiest <est supervisor.
"Tke briefing was comprehensive and the responsibilities of the participants were clearly assigned.
Conditions were discussed which would cause the termination of testing and contingent actions required to restore offsite power were reviewed.
At the command of the station shift supervisor, the test was initiated at 2246 by simultaneously opening two breakers in the Scriba switchyard to remove offsite power from the station.
All three EDGs started and successfully loaded on to their respective busses.
Operations and test personnel closely monitored systems performance and control room indications to assure that all required equipment was operating properly.
During the test the inspector observed overall good coordination and control of test activities.
Participants performed their assigned duties in a competent and professional manner.
At 2336 the test was concluded and operations personnel began to restore normal power to the station.
The test revealed a few minor equipment problems.
The control power for the Drywell Unit Coolers (non-safety related)
was found to be unavailable from a non-interruptable source.
A modification was initiated to provide control power from the same stub busses that provide motive power for the drywell fan The 'B'rain of the Standby Gas Treatment System (SBGTS) failed to
'c-'-r'="-'."."automatically start, as required, during the test.
An investigation
" revealed that the problem was associated with a "fail-safe" feature
"-'f tI1e suction and discharge valves (dampers)
of the SBGTS fans.
-'.: The hydraulically actuated valves were designed to fail close on a
loss of AC Power.
This was accomplished by depressurizing the hydraulic fluid on a sensed loss of AC power.
When AC power was restored (from the emergency busses)
the hydraulic system would have to re-pressurize before the valves could operate.
An interlock was also provided to inhibit ~he start of,the SBGTS fan unless its discharge valve was open.
Since the SBGTS receives a start signal 30 seconds following the re-energization of the emergency busses, the valve actuating hydraulic system m'ust re-pressurize within 30 seconds.
Test showed that this could not be reliably accomplished
'..
and, hence, could 'result in one or both SBGTS trains failing to automatically start.
To correct this problem a modification was made to the "fail-safe". logic of the suction and discharge valves of
.the SBGTS.
A timer was.installed to prevent depressurization of the hydraulic fluid for.l5-seconds.following a sensed loss of AC power.
This time is sufficient to allow the re-energization of the emergency busses (required within 13 seconds)
but will still cause-the valves to close on a sustained loss of AC power.
The licensee is evaluating this problem for Part 21 reportability.
Findimis No unacceptable conditions were identified.
2.5 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-02-2 Radiation Measurements This test was still in progress during the inspection.
The inspector reviewed the results of Section 6. 1, Area Radiation Measurements.
This portion of the test was performed on August 25 and 26, 1987, with the unit at 24% of rated thermal power.
All radiation measurements were well within acceptance criteria limits and consistent with current plant radiological zone N2-SUT-10-1 IRM Performance The results of this test were previously evaluated in Inspection Report No. 50-410/87-33.
During the current. inspection, the inspector verified
.,-.the. completion of the licensee's review of the results and the resolu-
. tion of the Level 2 test exception for failure to achieve the desired full decade overlap with the APRNs.
The licensee accepted the test exception 'as-is'or the four IRNs which did not achieve a full decade overlap with the APRMs.
The acceptance was justified based on all IRMs meeting the technical specification requirement for
~~~ decade overlap.
It was also noted that satisfactory SRM/IRN overlap existed and that readjustments to meet the full decade overlap with the APRMs would require an undesirable decrease in the sensitivity of the affected IRNs.
The inspector concurred with this evaluation.
N2-SUT-12-2 APRN Calibration The APRMs were calibrated by means of a heat balance performed by the process computer.
The initial performance of the test on September 1,
1987 (reactor power 28. 1%) resulted in a level 2 test exception when technical specifications required that the APRMs -be set more than 2% higher than 'actual core thermal power due to CMFLPD exceeding FRTP (T=.834).
The test was repeated on September 2,
1987 (reactor power 41.4%)
and all acceptance criteria were satisfied.
N2-SUT-13-2 Process Com uter This test was still in progress during the inspection.
The inspection reviewed the results of Section 6. 1, Dynamic System Test Case, and Section 6.2, Acceptance Criteria Verification.
The Dynamic System Test Case verified proper operation of NSSS programs by comparing their outputs to manual calculations, off-line computer calculations and other process computer programs.
Five minor test exceptions were identified during testing and satisfactorily resolved.
The Acceptance Criteria Verification verified that the process computer was correctly calculating the core thermal limits.
This was accomplished by verifying that MCPR, MLHGR, MAPLHGR and LPRM GAFs agreed (within 2%)
with those calculated by the off-line program BUCLE.
All acceptance criteria were satisfied.
N2-SUT-14-2 RCIC S stem The results of this test were previously evaluated in Inspection Report No. 50-410/87-33.
During the current inspection, the inspector verified the completion of the licensee's review of the results and the resolution of a level 1 and a level 2 test exceptio The level l,test exception involved the failure of RCIC to achieve rated Slow. within 30 seconds when tested at a reactor pressure of 153 psig in <he:CST-to-CST mode.
The licensee accepted this
- " exception"as-is'ased on satisfactory system performance during
'ctual RPV-injection testing at 150 psig.
The fai lure to achieve the desired time.to rated flow in the CST-to-CST mode was attributed to excessive line losses in this mode.
The level 2 test exception for minor steam leaks on the RCIC Governor Valve and Trip Throttle Valve'was resolved through normal corrective maintenance.
N2-SUT-14 RCIC S stem This additional testing'f the RCIC System was performed to evaluate the effect on system performance. of a modification made to the startup steam bypass line.
The modification involved increasing the size of a restricting orifice in the line to improve its efficiency during the initial roll of the turbine from cold conditions.
The test involved
- both cold and hot system quick starts at rated reactor pressure in the CST-to-CST mode.
A comparison was made with tests performed before the modification.
These results were still being reviewed by the licensee but initial evaluation showed little change in overall system performance.
A level 2 test exception was identified during both the cold and hot quick starts when peak turbine speed exceeded the overspeed trip avoid-ance limit of 4777 RPM (4801 RPM-cold and 4794 RPM-hot).
The exception appears to be the result of a calibration problem with the speed sensing loop for GETARS since, with the turbine at rest
,
a speed of 209 RPM is indicated.
The inspector will verify the completion of the licensee's results-review and the formal resolution of the test exception during a
future routine inspection.
N2-SUT-16-2 Water Level Measurements With the reactor at rated conditions in the Test Condition 2 window, the environmental conditions in the drywell and reactor building were measured and -calibration calculations performed for the wide, narrow and upset range water level instrumentation to verify that the scale end point error was <
1%.
The acceptance criterion was met for the wide and narrow range instrumentation but a level 2 test exception was identified for the upset range instrumentation with an. endpoint error of 1. 12% (equivalent to an error of about 2 inches of level).
The upset range instrumentation does not perform any automatic safety function and is used for indication only.
The inspector will review the resolution of this test exception during a future routine inspectio.. N2-SUT-19-2 Core Performance I
t I
,~ -'-Verification'of the Core thermal-hydraulic limits was performed at 28. 1% of rated thermal power (933 MW)and 38.6% of rated core flow
. '(41-.87 mlb/hr).
All acceptance criteria-vere satisfied and the results are summarized below:
Parameter Measured Value Limit
.-.<HGR (kW/ft)
CPF APLHGR (kW/ft)
4. 52 3.436 4.00
< 13.4, 1.55
< 11.88
~Findin s
No unacceptable conditions were identified..
3.0 Control Room Observation
~Sco e The inspector observed overall control room activities during the day shift on. October 7, 1987.
The inspector evaluated operator attentiveness and responsiveness to changesmn plant parameters and conditions during a power ascension from 26% of rated thermal power to 44% of rated thermal power including the initial operation with the reactor recirculation pumps in high speed.
Adherence to plant operating procedures, technical specifications and administrative requirements were verified.
The inspector assessed over-all coordination and control of activities and appropriate supervisory support during the shift.
Discussion At 0700 on October 7, 1987, the reactor was at 26% of rated thermal power (873 MW) with the generator on-line carrying a load of 178 MWe.
The reactor engineer was in the process of completing a whole core LPRM calibration (OD-1).
The orders for the shift called for a power ascension to approxi-mately 45% of rated thermal power following completion of the OD-1 and verification of core thermal-hydraulic limits (P-1).
The run of P-1 following the completion of the LPRM calibration revealed unexpected problems with the process computer.
The P-1 edit showed a number of base crit codes which are never expected following an LPRM calibration.
The edit also indicated a very restrictive T-factor (ratio of.
FRTP to CMFLPD).
Following consultation with the reactor engineering.supervisor, the PAM and plant management, the deci sion was made to commence power ascension utilizing the restrictive T-factor while troubleshooting was performed on the process computer.
The APRM GAFs were set at approximately
.7 At 0915 a power"'increase was begun*.to approximately 40% power by means a rod pulls to establish the,-conditions:required to shift the reactor recirculation
.....pumps. to fast speed::!'-.At:.1345 reactor power was 37% and the required feed-
." 'water low"*flow"(30% of rated) imterlock was cleared and..preparations were
-" " made to shift to fast speed pumps.
The 'A'ump was selected to be the first pump shifted to.fast speed and its flow control valve (FCV) was slowly closed to minimum to satisfy the slow-to-fast speed shift interlock.
During the closure of the FCV, core flow decreased from 43% of rated to 30% and reactor power decreased to 34%
and the low feedwater flow interlock was actuated (conservative setpoint).
-
Under the direction of the shift reactor engineer, control rods were with-
..drawn. until the low feedwater flow interlock cleared, and at 1420 the
'A'ump was successfully shifted to fas't speed.
The FCV for the 'B'ump was
- then slowly closed in preparation for its shift.to fast.
During its closure
. core flow and reactor power again decreased and the,low feedwater flow
":'-'nterlock was'actuated, automatically causing the
'A';,pump to transfer to slow"speed.
At 1445, under the direction of the shift reactor engineer, controls rods were again pulled to increase power and provide additional margin to the low.feedwater flow interlock.
Rod pulls were. continued until'
Rod Block Monitor Rod "Block occurred (A conseq'uence"of the conservative
'APRM 'GAFs).
At 1455 the 'B'ump was successfully shifted to fast speed and 1B FCV then opened approximately 10% to provide added margin, prior to attempting the transfer of the 'A'ump.
At 1505 the 'A'ump was successfully shifted to fast speed and reactor power stabilized at 44% (1467 MW) with the main generator carrying a load of 392 MWe.
Preparations were then begun to re-perform a whole core LPRM calibration and verify conformance with core thermal-hydraulic limits.
The process computer troubleshooting had been successful in identifying and correcting the previous software problem.
~indincis The inspector observed good overall operator performance during the complex evolutions conducted on this shift.
Supervisory support was provided by the reactor engineering supervisor, the PAM and senior operations personnel.
All operating, technical specification and administrative requirements were satisfied.
No unacceptable conditions were identifie I
-10-
":.4.D.Interface'with <he"PATP
':The inspector 'noted.during <he review 'of, power'ascension"procedures that
'they had been reviewed'y QA prior to issuance.
"In addition, during the evaluation of power ascension test results, the inspector noted that the packages were reviewed by QA prior to the presentation of the results to SORC for acceptance.
No unacceptable conditions were noted.
5.0 Inde endent Measurements and Verifications
'"During the witnessing of"the cold shutdown loss of offsite power test as discussed.in paragraph 2.4, the inspector independently verified the expected
';.'ystem responses. 'n addition; during the evaluation 'of the results of power
~scension test N2-SUT-34, RCIC System,,
as discussed in paragraph 2.5, the inspector measured the time to rated flow and peak turbine speed, using
.
GETARS t;races, for both the cold and hot quick starts.
The inspector's measurements and verifications agreed with those made by the licensee.
No unacceptable conditions were noted.
6.0 Exit Interview At the conclusion of the inspection on October 9, 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 materials 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 'Ascensi'on Test'-'Procedures Reviewed N2-SUT-02-6 N2-SUT;05-6 Radiation Measurements--
TCG, Revision 1, approved November 3, 1986 Control Rod Drive System, Revision 1, approved February 18, 1987
.. N2-SUT-3,1-.6
.
.LPRM Calibration - Test Condition G,-Revision 1,
. approved October '29,'j.986 N2-SIJT-12-6
'2-SUT-16-6
."APRM Calibration -- Test Condition. 6, Revision 1,
- approved February 12,. l987 Mater Level Measurements
- Test Condition 6, Revision 1, approved December 17, 1986 N2-SUT..18=6 TIP Uncertainty, Revision 0, approved August 16, 1986 N2-SUT-19-6 N2-SUT-71-6 N2-SUT-74-6 N2-SUT-75-6 N2"SUT-76-6 N2-SUT-77-6 Core Performance - Test Condition 6, Revision 1, approved April 16, 1987 Residual Heat Removal System, Revision 0, approved December 23, 1986 Offgas System TC-6, Revision 1, approved February 12, 1987 Drywell Cooling System, Revision 1, approved August 24, 1987 ESF Area Cooling Test Condition 6, Revision 0, approved August 5, 1986 BOP and Small Bore Piping Vibration, Revision 0, approved October 3, 1986 N2-SUT-78-6 BOP System Expansion, Revision 0, approved August 15, 1986
"ATTACHMENT B
- "Power Ascension 'Test Results=.Evaluated N2-SUT-02-2
-
Radiation Measurements TC2, Revision 1, in progress
'2-SUT-10-1 -~.:IRM Performance; Revision; 2,. results accepted
'September 17, 1987 N2-SUT-12-2 APRM Calibration=- Test Condition 2, Revision 1, resvlis.accepted September 17, 1987 N2-SUT-13-2'rocess'Computer, Revision 0, in progress
'"N2-SUT-14-.2
.
'RCIC System, 'Revision.3, res'ults accepted'September.
17,,...1987 N2-SUT-14 RCIC System, Revision 0, completed October 6, 1987, in review N2-SUT-16-2 Selected Process Temperatures and Mater Level Measurements
" Test Condition'.2;. Revision 0, results accepted September 17, 1987 N2-SUT-19-2 Core Performance - Test Condition 2, Revision 1, results accepted September 17, 1987