IR 05000220/1979023
| ML17053B276 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 10/11/1979 |
| From: | Bettenhausen L, Caphton D, Graham P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17053B269 | List: |
| References | |
| 50-220-79-23, NUDOCS 7912200332 | |
| Download: ML17053B276 (20) | |
Text
U.S.
NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region I
0
. ~50-220 99-23 Docket No.
50-220 License No. DPR-63 Priority Category C
Licensee:
Nia ara Mohawk Power Cor oration 300 Erie Boulevard West
Syracuse, New York 13202 Facility Name:
Nine Mile Point Nuclear Station, Unit
Inspection at:
Scriba, New York Inspection conducted:
eP<<mbe 18-20, 1979 Inspectors:
P.
D. Graham, Reactor Inspec r
L. H. Bettenhausen, Reactor nspector
a e sign d
at si ed Approved by:
C D. L. Capht n, Nuclear Support Section, No. 1, RO&NS Branch date signed fd rr 7 ate signed Ins ection Summar
Ins ection on Se tember 18-20, 1979 (Re ort No. 50-220/79-23 Areas Ins ected:
Routine, unannounced inspection of post refueling activities inc uding plant operations; startup testing; core power distribution; core thermal power evaluation; shutdown margin determination; and LPRM and APRM calibration.
The inspection involved 41 inspector hours onsite by two NRC regional based inspectors.
Results:
Of the seven areas inspected, no items of noncompliance were identified in ssx areas; one item of noncompliance was identified in one area (Deficiency - failure to properly document and review plant testing procedures, paragraph 5a).
Region I Form 12 (Rev. April 77)
33~
V91SSOP ~
DETAILS 1.
Persons Contacted
"M. Drews, Reactor Analyst
"T. Lempges, General Superintendent Nuclear Generation
"M. Silliman, Results Superintendent C. Stuart, Supervisor, Operations B. Taylor, Supervisor, I&C
".denotes those present at the exit interview.
The inspector also talked with and interviewed other licensee employees during the inspection.
2.
Licensee Action on Previous Ins ection Findin s
(Closed) Unresolved item (77-25-01):
Deletion of LPRM inputs.
The licensee received the computer program change and incorporated the change into procedure Nl-RAP"2, Process Computer Software.
Successful testing of the program change
.has been accomplished by the licensee.
This item is closed.
3.
Post Refuelin Plant 0 erations
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~
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The inspector reviewed the facility procedures, check-off lists and valve lineups to determine that:
systems disturbed during the refueling outage were returned to service in accordance with approved procedures; and, control rod withdrawal sequence and authorization were available and in effect prior to the startup after refueling.
The following procedures/check-off lists/valve lineups were reviewed:
Nl-OP-43, Startup and Shutdown Procedure, including check lists, Revision 12, performed, June 12 through June 29, 1979 Nl-ST-g8, Main Steam Line Isolation Valve Position Instrument Channel Test, Revision 2, performed June 22, 1979 Nl-ST-g3, High Pressure Coolant Injection Pump Operability Test, Revision 2, performed June 22, 1979 Nl-ST-V3, Rod North Minimizer Operability Test, Revision 2, performed June 20, 1979
Nl-ST-V4, Operability of Feedwater and Main Steam Line Power Operated Isolation Valves Test, Revision 1, performed June 24, 1979 Nl-OP-2, Valve Line-Up, Core Spray System, performed May 25, 1979 Nl-OP-3, Valve Line-Up, Clean-Up System, performed May ']9, 1979 Nl-OP-5 Valve Line-Up, Control Rod Drive Hydraulic System, performed June 2, 1979 Nl-OP-12 Valve Line-Up, Liquid Poison System, performed May 22, 1979 Nl-OP-45, Valve Line-Up, Diesel Generators, performed June 13, 1979 Nl-OP-1, Valve Check-Off List For Inservice Vessel Hydro, performed June 6, 1979 Nl-ST-C6, Source Range Monitor Operability Test, Revision 2, performed June 1,
1979 Nl-ST-V5, Suppression Pool Temper'ature Monitoring During Relief Valve Testing, Revision 1, performed June 22, 1979 Nl-ST-C5, Secondary Containment Leak Rate/Reactor Building Ventilation Duct Radiation Monitor Test, Revision 3, performed June 17, 1979 Nl-ST-R2, Loss of Coolant and Emergency Diesel Simulated Automatic Initiation Test, Revision 2, performed June 16, 1979.
~Findin s
Review of the Valve Check-Off List For Inservice Vessel Hydro indicated that the typical method of operators initialling and dating the check-off for each check of a valve position had not been followed.
Instead of initialling and dating the check-off valve position, the operator wrote in the space provided the required valve position, either OPEN or CLOSED.
When the inspector brought this to the attention of a licensee represen-tative, the licensee representative agreed that this was not the proper way to complete a valve check-off list.
Licensee representatives committed to a procedural change to Nl-OP-1 by November 21, 1979, that would insure that the operators initial each check of a valve position.
This item is unresolved pending incorporation of the change and inspector review (220/
79-23-01).
Surveillance test, Nl-ST-V5, requires that the operators log the suppression pool temperature and the time at the completion of relief valve operation.
The inspector noted that the operators performing the test on June 22, 1979, had logged the suppression pool temperature after all relief valves had been tested.
Since heat is added to the suppression pool dur ing each relief valve test, the inspector stated that'the logging requirement should be completed after each valve is tested.
Licensee representatives agreed with the inspector's comments and committed to a procedural change by November 21, 1979 that would require logging the suppression pool temperature after each valve is tested.
This item is unresolved pending the incorporation of the above change and inspector review (220/79-23-02).
The inspector had no further questions in this area at this time.
Post Refuelin Startu Re ort Review The inspector discussed with the licensee the submission of the Cycle 6 Startup Report.
The licensee commented that the report had been submitted to the Nuclear Regulatory Commission (NRC) pursuant to the requirements of the NRC as detailed in the letter of January 19, 1979 to D. Disc (Niagara Mohawk Power Corporation)
from T. Ippolito (NRC).
Although the final Startup Report was not available for inspector review, the inspector did review the data input to the report to verify the following:
The report would include the required information; and, The test results were consistent with design predictions and performance specification.
No items of noncompliance were identified.
Post Refuelin Startu Testin The inspector reviewed tests, checks and documents described below to verify that startup testing was conducted in accordance with technically adequate procedures and as required by Technical Specification (TS)
and the Cycle 6 licensing submittal.
Criticality predictions and predicted physics parameters are contained in the General Electric letter, Nine Mile Point-1 Cycle 6 Nuclear Information, dated May 30, 1979, and NED0-24155, Supplement Reload Licensing Submittal For Nine Mile Point Nuclear Power Station Unit 1 Reload No. 7, dated November 1978, respectivel a.
Control Rod Drive.Scram Insertion Times The licensee verified that the control rod drive scram insertion times for all control rods were within the requirements of T.S.
3. 1. l.c by the performance of procedure Nl-RAP-3, Full Core Control Rod Scram Timing Sequence, Revision 1, on June 22, 1979.
The inspec-tor reviewed the completed procedure, and with the exception of the following item had no further questions.
In the results section of the procedure, a signature block signi-fying that the T.S.
3. 1. 1. c(2) requirements, individual rod scram times, was "N/A'd" by the technician.
This signature block should have been signed in this case to indicate that the individual rod speeds had been reviewed and were in compliance with T.S.
The inspector discussed this with a licensee representative who com-mented that the review had been accomplished and that the block should have been signed.
The inspector reviewed the individual rod speeds and verified compliance with T.S.
requirements.
This item is the first example of the licensee's failure to properly document and review completed test procedures.
This example, taken collectively with two other examples discussed later in the report, is considered to constitute a deficiency level item of noncompliance (220/79-23-03).
b.
Calibration of Local Power Ran e Monitors LPRM)
The inspector reviewed the following approved plant procedures.
Nl-RPSP-7, LPRM Calibration, Revision 0, completed June 25, July 9, 26, and September 7,
1979.
Nl-RCPCP-9, LPRM Sensitivity Monitoring, Revision 1, approved November 1, 1976.
Nl-RCPCP-16, LPRM Calibration by TBAR Method, Revision 2, approved August 13, 1979.
Except as noted below, the inspector had no further questions in this area at this time.
Procedure Nl-RPSP-7, step VII.l.f places the Average Power Range Monitors (APRM) in bypass.
The inspector noted that a ".return to normal" step, which would insure that the APRM's are returned to
"OPERATE", was not included in the procedure.
The inspector brought
.this discrepancy to the attention of a licensee representative and a
change was made to the procedure before the inspector completed the inspection.
This item will be followed up during a subsequent inspec-tion to insure that the Onsite Review Committee (SORC) reviewed the revised procedure (220/79-23-04).
Procedure, RCPCP-16, had not been updated to reference the Cycle 6 data necessary for the THAR method of calibrating LPRM's.
A change to the procedure was made before the inspector completed the inspection.
c.
Calibration of Avera e Power Ran e Monitors APRM The licensee calibrates the APRM's daily pursuant to the requirements of T.S. 4.6.2.
The inspector reviewed the following approved procedures:
Nl-RPSP-1, Daily Surveillance, Revision 4, performed June
through July 13, 1979, and September 8 through September 18, 1979; RCPCP-ll, APRM Calibration, Revision 2.
Except as noted below, the inspector had no further questions in this area at this time.
Procedure RCPCP"11 had not been updated to reflect the total Peaking, Factor for 8 x 8R fuel used in calculating the core thermal power.
A change to the procedure was made before the inspector completed the inspection.
d.
Shutdown Mar in SDM The licensee corn'plied with shutdown margin demonstration requirements of T.S.
4. l. la(l) by determining that the reactor SDM with. the strongest control rod fully withdrawn was at least equal to 0.50K delta k/k vice the required T.S.
SDM value of 0.25K delta k/k.
The inspector reviewed, the following procedure relating to the SDM demonstration.
RPSP-4, Reactivity Margin - Core Loading, Revision 1, completed June 21, 1979.
Except 'as noted below, the inspector had no further questions in this area at this time.
RPSP-4 requires that a second, independent verification of the rod selected for movement be made.
This verification is documented on the procedure by the initials of the verifiers.
The inspector noted that both verifications were initialled by the same technician.
Discussions with a licensee representative indicated that the independent verifications had been made, but were improperly documented.
This item is the second example of the licensee's failure to properly document and review completed test procedures, and is considered part of the deficiency level item of noncompliance discussed in paragraph 5.a (220/79-23-03).
RPSP-4 relies on the rod configuration provided by General Electric for demonstrating SDM.
This rod configuration assumes a reactor moderator temperature of 68'.
At the time the procedure was performed moderator temperature was 139'.
The inspector noted that the procedure did not reference or'incorporate any temperature correction.
The inspector discussed the necessity for a temperature correction with a licensee representative, and determined that the licensee demonstrates an excess SDM to account for this correction.
In this particular case an excess SDM of 0.25K delta k/k was demonstrated which more than compensates for the difference in temperature..
Because of the importance of this test, the inspector-.stated that the procedure should make reference to the temperature correction.
A licensee representative agreed with the inspector, and a change to the procedure was made before the inspection was completed.
This item will be followed up during a subsequent inspection to insure SORC review of the procedure (220/79-23-05).
Core Power Distribution The procedures and methods used by the licensee to verify that the plant is operated within the power distribution. limits of T.S.
3. 17 were reviewed and discussed with cognizant licensee personnel.
The licensee uses the on-line process computer.to monitor core parameters and make core calculations to insure compliance with T.S.
The Traversing Incore Probe (TIP) System is used to measure the neutron flux distribution in the monitored core locations.
The OD-l program normalizes the flux measurements from the TIP machines.
The P-1 program performs a core power distribution calculation based on the normalized flux measurements.
The OD-6 program provides calculations and edits of Linear Heat Generation Rates (LHGR) and Minimum Critical Power Ratio (MCPR).
The inspector reviewed the following plant procedures and data to verify that the plant is operated in compliance with the limits defined in the T.S.
Nl-RCPCP-12, TIP Intercalibration, Revision 2, approved August 13, 1979 Nl-OP-39, TIP, Revision 5, approved February 16, 1978 OD-1, Mhole Core LPRM Calibration and Base Distribution, completed June 23 and July 25, 1979 P-l, Periodic Core Performance Log, completed June 23 and July 25, 197. OD-6, Option 3, The 12 Bundles Closest to CPR Limits, completed June 23 at 1920 hours0.0222 days <br />0.533 hours <br />0.00317 weeks <br />7.3056e-4 months <br /> and July 25, 1979 at 0703 hours0.00814 days <br />0.195 hours <br />0.00116 weeks <br />2.674915e-4 months <br />.
OD-6, Option 4, The 12 Highest Ratios of a Bundle Maximum Average Planar Linear Heat Generation Rate to its Limiting Linear Heat Generation Rate, completed June 23 at 1920 hours0.0222 days <br />0.533 hours <br />0.00317 weeks <br />7.3056e-4 months <br /> and July 25, 1979 at 0703 hours0.00814 days <br />0.195 hours <br />0.00116 weeks <br />2.674915e-4 months <br />.
TIP Traces performed on June 23, 27, July 25, and September 9,
1979.
Nl-RPSP-1, Daily Surveillance, Revision 4, performed June
through July 23, 1979.
Except as noted below, the inspector had no further questions in this area at this time.
The acceptance criter ia in Nl-RPSP-1 for Linear Heat Generation Rate (LHGR) 'was incorrect.
The LHGR stated in the procedure did not take into account the power spiking penalty factor of 2.2X that is discussed in T.S.
A change to the procedure was made before the inspection was 'completed, and the procedure is awaiting SORC review.
This item will be reviewed. in a subsequent inspection (220/79"23-06).
Core Thermal Power The procedures and methods used by the licensee to determine core thermal power were reviewed and discussed with cognizant personnel.
Procedure Nl-RPSP-l, Daily Surveillance, Revision 4,'s performed daily to insure compliance with T.S.
The plant on-line computer scans the appropriate data points and performs the core thermal power determination.
A check of the computer calculated core thermal power is performed at least once a month by a hand calorimetric calculation in accordance with procedure Nl-RCPCP-8, Core Thermal Power, Revision 3.
Hand calorimetrics were reviewed for the period July 17, 1979 to September 12, 1979.
Except as noted below, the inspector had no further questions.
Procedure Nl-RCPCP-8 was revised on August 13, 1979.
The revised procedure incorporated a new revised data sheet 3.
The inspector noted that calculations performed on August 16 and September 12, 1979, still utilized the old revision of data sheet 3.
This discrepancy was brought to the attention of the cognizant personnel and the blank, outdated data sheets were removed from circulatio g.
Core Loadin Verificati on As part of the Startup Test Program, Nine Mile Point Unit 1 was required to perform a core loading verification.
The verification was performed in accordance with procedure Nl-FHP-22, Core Post-Alteration Inspection and Verification, Revision 0.
No items of noncompliance were identified.
h.
Reactivit Anomal The procedure and method used by the licensee to determine the reactivity anomaly was reviewed and discussed with cognizant personnel.
Procedure Nl-RPSP-3, Reactivities Anomalies, Revision 1, is performed every effective full power month to insure compliance with T.S. 4. 1. l.f.
Procedure Nl-RPSP-3 was also performed during startup testing to measure the criticality position.
The following table illustrates the predicted versus actual critical rod position.
Control Rod Notches Mithdrawn
+ 1X delta k position Predicted position-1X delta k position Actual critical position 624 480 1358 Except as noted below the inspector had no further questions in this area at this time.
Procedure Nl-RPSP-3 performed during the Startup Test Program on July 3, 1979 did not document the control rod sequence being used as is required.
The inspector also noted that step 4. a. 1 had been completed by the technician yet plant conditions were such that the step should not have been corn'pleted.
These items are consider ed to be the third example of the licensee's failure to properly document and review completed test procedures, and is considered part of the deficiency level item of noncompliance discussed in paragraphs 5.a and 5.d (220/79-23-03).
TIP Uncertaint Determination As part of the Startup Test Program, Nine Mile Point Unit 1 was required to perform a TIP Uncertainty Calculation.
The calculation was performed in accordance with procedure Nl-RCPCP-18, TIP Uncertainty Calculation, Revision 0.
Except as noted below, the inspector had no further questions in this area at this time.
e.
The inspector noted that procedure Nl-RCPCP-18 did not specify an acceptance criteria for the TIP uncertainty, and it did not.
'ave a review and sign-off sheet.
A licensee representative commented that the procedure was a calculational procedure only,-
, and that the administrative procedures did not require calculational procedures to have acceptance criteria and sign-off sheets.
Since the TIP Uncertainty Calculation was performed as part of the Startup Test.Program, the inspector stated that the procedure should contain an acceptance criteria and a review and sign-'off sheet to insure the calculations acceptability.
The licensee representative agreed and a change to the procedure was made
'efore the in'spection was complete'd.
Unresolved Items
,Items about which more information is required to determine acceptability are considered unresolved.
Paragraph 3 of this report contains unresolved items.
7.
Exit Interview The inspector met with licensee representatives (denoted in Paragraph 1) at the conclusion of the inspection on September 20, 1979.
The inspector summarized the scope and findings of the inspection as they are detailed in this report.
During this meeting, the unresolved items and apparent item of noncompliance were identifie 'l