IR 05000410/1987017
| ML17054C192 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 06/19/1987 |
| From: | Eselgroth P, Marilyn Evans, Florek D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17054C191 | List: |
| References | |
| 50-410-87-17, NUDOCS 8706300197 | |
| Download: ML17054C192 (16) | |
Text
'
U.S.
NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-410/87-17 Docket No.
50-410 License No.
NPF-54 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:
Ma 19-29 1987
, Inspectors:
M. Evans, Reactor Engineer I
date D. F'lorek, Acting Chief, Test Programs Section, OB, DRS date Approved by:
P. Eselgroth Acting Chief, Operations Branch, DRS d te Ins ection Summar:
Ins ection on Ma 19-29 1987 Re ort No. 50-410/87-17 A~I:
I, d
inspectors of overal
power ascension witnessing, power ascension test results pendent calculations and verifications and Results:
No violations were identified inspection by two region based test program, initial criticality evaluation, QA/QC interfaces, inde-plant tours.
NOTE:
For acronyms not defined, refer to NUREG-0544 "Handbook of Acronyms and Initialisms."
8706300i97 8706i9 PDR ADOCK 050004i0
DETAILS 1.0 Persons Contacted Nia ara Mohawk Power Cor oration R. Abbot, Station Superintendent C.
Beckham, Manager, Quality Assurance Operations M. Boyle, Compliance and Verification J.
Bunyon, Engineering G. Carlisle, Lead STD&A Engineer M. Colomb, Station Shift Supervisor J.
Conway, Power Ascension Manager P.
Eddy, Site Representative, New York State, PSC M. Jones, Operations Superintendent P.
MacEwan, Site Representative, New York State'lectric and Gas T.
Newman, Supervisor, Operations Surveillance T. Pao, Shift Test Supervisor T. Perkins, General Superintendent J. Perry, Vice President, Quality Assurance A. Pinter, Site Licensing Engineer L. Wolf, Site Licensing Engineer Other NRC Personnel W. Cook, Senior Resident Inspector C. Marshall, Resident Inspector W. Schmidt, Resident Inspector Denotes those present at exit meeting on May 29, 1987.
The inspectors also contacted other members of the licensee's technical, QA and Operations staff.
2.0 Power Ascension Test Pro ram PATP 2. 1 References Regulatory Guide 1.68, Revision 2, August 1978 "Initial Test Program for Water Cooled Nuclear Power Plants.".
ANSI N18.7-1976 "Administrative Controls and Quality Assurance for Operations Phase of Nuclear Power Plants."
Nine Mile Point Unit 2 (NMP-2) Technical Specifications, Revision 0, October 31, 1986.
NMP-2 Final Safety Analysis Report (FSAR) Chapter 14 "Initial Test Program."
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NMP-2 Safety Evaluation Report
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NMP-2 AP-1.4, Startup Test Phase, Revision
'.2 Overall Power Ascension Test Proqram The inspector held discussions with the Power Ascension Manager (PAM) and the Lead Startup Test, Design and Analysis (STD&A) Engineer and other members of the PATP staff to assess the overall readiness to begin low power testing.
Shift Test Personnel had begun continuous shift coverage on May 20th to support initial criticality.
Working copies of all low power test procedures had been issued.
The inspector witnessed several licensee planning meetings.
Plan. s:atus, problems, resolutions and future plans were discussed.
2.3 Initial Criticalit Witnessin 2.3.1
~Sco e
The inspector witnessed portions of the initial criticality testing sequence to assess that:
~
Licensee was complying with the technicai specification requirements
=or initial criticality.
SRMi/IRMi instruments were properly cai '.b. a-eG anQ nac appropriate outpu:s and
. esponse.
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RPS shorting links were removed.
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Adequate staffing existed.
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Current revisions of procedures were in.se and being followed.
2.3.2
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Procedure prerequ'sites were satisfiec.
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Required data was obiainea.
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Acceptance criteria were satisfied.
Discussion During the current and one previous inspection (50-410/87-15)
the inspector monitored the licensee's surveillance program to verify compliance with technical specification requirements for entry into Operational Condition 2.
In addition, on May 21, 1987 the inspector witnessed conduct of portions of surveillance procedure N2-ISP-NMS-W*007, APRM/LPRM Channel Functional tes i
'
On May 22, 1987, the inspector attended the Station Operat.ions Review Committee (SORC) meeting to witness management review of activities and to assess overall readiness for initial criticality.
On May 23, 1987, the inspector attended the shift turnover meetings for the day and swing shifts and witnessed two shift test briefings by power ascension test engineers covering the planned initial criticality testing sequence.
All preparatory activities were adequately completed.
Prior to the commencement of test activities the inspector reviewed the General Electric Company, Startup Data Book, 23A1840 and the Control Rod Pull Sheets and performed an independent calculation of the expected critical position and the +1:.'eactivity anomaly limits.
The inspector verified that official working copies of the following procedures were available in the control area:
N2-0P-101A, Plant Start-up N2-SUT-6-HU, SRM Performance N2-SUT-4-HU, Full Core Shutdown Margin Demons:ra-ion N2-0SP-NMS-SU001, SRM/IRM Overlap Check N2-RPSTP-2, Cold Critical Comparison By independent observations and review of records-he inspector verified procedural prerequisites such a's
. emoval of the RPS shorting links, SRM> operability including
>ainimum>
counts anc signal-to-noise ra:io, and SRM>
scram and ro='lock se-ooin.s.
All prerequisites were satisfied.
During the course of the initial criticality sequence
- he inspector also monitored control room 'staffing levels and the conduct of the operations shift personnel.
Control room mannino was more than adequate to support testing and satisfy technical specification requirements.
The station shift, supervisors were effective in maintaining the proper control room atmosphere including positive control of the numbers and conduct of non-participating observers.
At 2210 on May 22, 1987 the Mode Switch was placed in "Startup" and the unit entered Operational Condition 2..
The surveillances for the Rod Worth Minimizer (RWM)
and Rod Sequence Control System (RSCS)
were performed and the systems declared operational.
At 0920, on May 23, 1987, the reactor startup commenced.
At 1040 while withdrawing the 8th control rod (38-07)
a Rod Drive Control System (RDCS)
"Inop" alarm was received when the rod settled at notch position 48 (full out).
The problem was identified as a failed transponder card in the RDCS and a.Work Request (WR)
was written to replace the card.
Following
replacement of the transponder card, rod coupling was verified for control rod 38-07 and the RWM was reinitialized.
Rod pulls were resumed and continued uninterrupted until a
second RDCS transponder card failed while withdrawing control rod 22-15 at 1508.
Following replacement of the card, the inspector verified conduct of the control rod coupling check and RWM reinitializa-tion prior to continuation of rod pulls.
At 2007 the reactor was declared critical on control rod 50-35 at notch position 08 with a
moderator'emperature of 110'F.
Following initial criticality the inspector verified SRM/IRM overlap when all
IRMs came on scale prior to the SRMs exceeding their rod block set points.
During the initial criticality sequence the inspector monitored the maintenance of 1/m plots per N2-SUT-6-HU,
"SRM Performance" and randomly performed the necessary calculations.
In addition, the inspector independently performed the in-sequence critical shutdown margin (SDM)
calculations following initial criticality. The inspector's value for the SDM did not match the value calculated and verified by the Power Ascension Test Engineers.
The inspector informed the Lead STD&A Engineer of the discrepancy and it was determined that the value calculated by the test engineer was incorrect in the conservative direction.
The inspector discussed this with the Lead STD&A Engineer and stressed the need for more carefully performing all calculations.
The Lead STD&A Engineer acknowledged the inspector's concerns.
This was subsequently discussed with the Power Ascension Manager.
The inspector was informed that the individuals involved were counseled and the test engineers were again reminded of their responsibilities in the performance and verification of test calculations.
Witnessing of the initial criticality testing sequence was also performed by the Resident Inspectors and is discussed in their inspection report covering this period.
2.3.3
~Findin s
No violations were identified.
2.4 Power Ascension Test Results Evaluation 2.4.1
~Sco e
The power ascension test results discussed below were evaluated for the attributes identified in Inspection Report 50-410/86-6.4.2 Discussion The test results reviewed were currently in the review cycle
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The inspector assessed the data against the acceptance criteria and performed selected independent calculations.
A summary of the test results reviewed follows.
N2-SUT-4-HU Full Core Shutdown Margin.
A shutdown margin of 2.3% was determined which satisfied the te'st criteria of greater than
.47%.
N2-SUT-6-HU Source Range Monitor Performance.
SRM/IRM overlap was demonstrated.
SRM/IRM values at overlap were; SRM IRM
.6x10'
.3x10'
.7x10'.5
.6x10'.3 6.0 5.3 10.0 6.8 N2-SUT-10-HU IRM Performance.
An acceptable range 6/7 overlap was demonstrated N2-SUT-12-HU APRM Calibration.
APRMs were adjusted to read equal to or greater than actual core thermal power.
The adjusted values were; APRM Desired As-left
.76
.79
.79
.77
.86
.95
.8
.8
.8
.8 1.1 1.1
~Findin s
No violations were identified.
2.5 Other Test Witnessin Activities The inspector witnessed Operations performance of surveillance testing to determine ADS operability.
Technical specifications require this testing to be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving'00 psi This test may be performed at pressures greater than 100 psig, but within
hours of achieving 100 psig.
This test was performed at 133psig, 3.2;.'eactor power with one turbine bypass valve open and one 90%
open.
The licensee conducted a
thorough test briefing, discussed the procedure, discussed potential problems, and determined the necessary steps that must be taken should a
valve stick open.
This procedure involved coordinating at least seven licensee personnel.
The licensee conducted a
dry run to assure the steps could be performed smoothly and assure communications were established.
This activity was effectively controlled by operating shift supervisory personnel.
During the performance of the test, plant responses were being monitored by the operations personnel.
After each valve was tested, operations personnel assured the plant was stable and direct evidence existed that the valve sucessfully reclosed.
The test identified that valve PSY-134 did not open when energized from the "C" solenoid.
This did not affect the ADS function.
A work request was issued to correct this condition.
Plant response to this test other than PSY-134 was acceptable.
Follow-up surveillance on drywell vacuum breakers were also planned to be conducted within two hours of relief valve operation per the technical specifications.
The inspector noted at the exit that the licensee plans, briefing, dry run and implementation of this test demonstrated an ability to conduct complex coordinated tests ir. an acceptable manner.
2.6 Containment Entry The inspector reviewed the licensee plans
,and preparations for a
containmenw en-ry to be performed in order to perform thermal expansion inspections as part of the startup program at approximately 350~F.
In addition, this entry was further complicated by the fact that the personnel airlock inner door was declared inoperable from a
technical specifica.ion point of view due to the discovery that the
',nner door equalizing valve seal had an expired technical specifi-cation surveillance.
As such, initial entry had to be accomplished through tne emergency airlock.
Based on discussions with Operations, Radiological Protection and IKC Supervisors and review of procedure S-RP-9, "Initial Entry to the Drywell", Revision 2 dated June 14, 1986 the inspector determined that initial entr'y would be to perform a radiological survey and determine air quality using two man teams with self contained breathing apparatus.
When air quality and radiological conditions are determined acceptable, additional personnel would be allowed to enter to perform the testing on the personnel air lock equalizing valve.
Until the personnel airlock was declared operable inaccor-dance with technical specifications, the number of personnel in the containment would be limited to six, the number that could be
accommodated with one opening of the emergency airlock.
In addition for personnel safety concerns an additional person would be stationed at the personnel airlock in the event the emergency airlock could not be opened and an emergency exit was required through the personnel airl ock.
The licen'see preparations and plans were reasonable for the drywell entry.
3. ~/I f*
During witnessing of initial criticality activities, the inspector noted continuous QC coverage of these activities in the control room.
4.0 Inde endent Calculations and Verifications
'0 Prior to the occurrence 'of the initial criticality sequence the inspector made independent calculations of the expected critical position (ECP)
and the
+1%
Reactivity Anomaly Limits.
During rod pulls the inspector monitored the maintenance of the 1/m plots and randomly performed the necessary calculations.
When the licensee declared the reactor critical the inspector independently verified the determination by examination of nuclear instrumentation indications.
During the shutdown margin test the inspector independently calculated the shutdown margins In all cases the inspector's calculations matched those of the licensee except as noted in paragraph 2.3.2.
5.0 Plant Tours The inspector made several tours of the drywell, reactor building and control building to observe work in progress, general housekeeping and cleanliness controls, and overall status of the facility to support initial criticality and low power testing.
No unacceptable conditions were identified.
6.0 Exit Interview
,A management meeting was held at the conclusion of the inspection on May 29, 1987, to discuss the inspection scope, findings and observations as detailed in this report (see Paragraph I 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 I