ML20031G036
| ML20031G036 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 09/25/1981 |
| From: | Caphton D, Chung J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20031G028 | List: |
| References | |
| 50-245-81-13, NUDOCS 8110200680 | |
| Download: ML20031G036 (15) | |
See also: IR 05000245/1981013
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U.S. NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
Region I
Report No.
50-245/81-13
Docket No.
50-245
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License No.
Priority
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Category
C
Licensee:
Northeast Nuclear Energy Company
P.O. Box 270
Hartford, Connecticut 06101
Facility Name:
Millstone Nuclear Power Station Unit 1
Inspection at:
Waterford, Connecticut
Inspection con etcd: August 10-14, 1981
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Inspectors:
m /O R
c.1 /
in W. Chung
date signed
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date signed
date signed
Approved by: MMNM4+<- A
9 -M S/
,
D. L. Caphton, Chief, Test Program
date signed
Section, Engineering Inspection Branch
Inspection Summary:
Inspection on August 10-14,1981 (Report No. 50-245/81-13)
Areas Inspected: Routine, unannounced inspection of cycle 8 post refueling startup
testing, which includes pre-critical scram times and hydro tests; shutdown margin
and critical rod configuration; and power escalation tests.
The inspection involved 31 inspector-hours on site by one region-based inspector.
Resul ts:
One item of noncompliance:
Failure to provide acceptance critoria and
detailed instructions; to follow written procedures; and to review and document
test results properly - section 4).
Region I Form 12
(Rev. April 77)
0110200600 811001
PDR ADOCK 05000245
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_ DETAILS
1.
Persons Contacted
Principal Licensee Employees
- R. Herbert, Unit 1 Superintendent
- E. J. Lindsay, Reactor Engineer
- E. J. Mroczka, Station Superintendent
- R. J. Palmier, Engineering Supervisor
F. W. Teeple, Unit 1, I&C Supervisor
- K. B. Thomas, Senior Engineer
J. Stetz, Unit 1, Operations Assistant
- D.
R. Lipinski, Resident Inspector
The inspector also interviewed other licensee employees during the inspec-
tion, including Reactor Operators, performance and administrative personnel.
- Licensee representative at the Exit Interview
-* denotes those present at the Exit Interview
2.
Cycle 8 Startup Testing - Precritical Test
a.
The inspector reviewed functional and calibration tests'and results to
verify the following:
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Procedures were provided with detailed instructions;
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Technical content of procedures was sufficient to result in
satisfactory component calibration and test;
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Instruments and calibration equipment used were traceable to the
National Bureau of Standards;
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Acceptance and operability criteria were observed in compliance
with Technical Specifications.
)
b.
The following tests were reviewed:
(1) Control Rod Drive; Scram Insertion Time
Scram insertion time had been measured on April 19, 1981 for 5%,
20%, 50%, and 90% insertions from fully withdrawn positions.The
test procedure employed was SP 1051, Control Rod Scram Time Test,
Revision 3, November 12, 1979. The inspector verified that the
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average scram insertion times were all within the limits specified
in Technical Specification 3.3.C except control rod drive 22-51.
Control rode drive 22-51 did not meet the 7.0 second insertion
limit for 90% insertion as required by Technical Specification 3.3.C.3.
The failed drive 22-51 was retested on April 20, 1981, again
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exceeding 7.0 second insertion specification. This drive was
subsequently replaced and the scram insertion times were
successfully demonstrated on June 27, 1981.
The documents reviewed include:
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Scram insertion time tests performed April 19, 1981 for
all control rods; April 20, 1981 and June 17, 1981 for
control. rod 22-51.
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Job Order No. 181-202, 22-51 Drive Replacement, May 4, 1981.
Findings
Procedure SP 1051 requires that scram insertion times be recorded on
RE Form 1051-3 and the average values of 145 rod insertion times be com-
pared with the acceptance limits.
Test records for April 19, 1981' showed
that the average values were not recorded-in the RE 1051-3 data sheet.
Technical Specification 3.3.C.3 specifies that the maximum scram
insertion time for 90% insertion of any operable control rod be less
that 7.0 seconds. The scram time tests performed April 19,
1981 at 145 MWe identified that the 90% insertion time for the
control rod 22-51 was 7.46 seconds, exceeding the required 7.0
seconds. However, neither subsequent action per Technical
Specifications nor any corrective actions such as a maintenance
request / replacement job order were specified in the test'
package.
Instead, the failed rod was retested the following
day, April 20. 1981 at 80 MWe; again failing the 7.0 second
Technical Specification. Again, no subsequent actions were
documented in the test record. However, it was determined
that corrective action was taken May 4, 1981.
During this inspection it took more than a day for a cognizant licensee
engineer to identify the Job Order Number from the separate file based
on his personal memory. These failures to follow written procedure
and to document the test results properly, constitute an example of
noncompliance as summarized in Sec. tion 4.
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(2) Hydrostatic Pressure Test
The inspector verified by review of test procedures, changes, and
test data that the pressure tests of the Reactor Vessel and
Class I piping system and components had been completed in
accordance with procedures, in that:
--
The hydro test was performed April 8-9, 1981 using procedures
SP o0-1-25, Hydrostatic Pressure Test of the Reactor Vessel
and Class I Piping System and Components, Revision 1,
June 18, 1980 and Operations Form 80-1-25-1, RPV Hydrostatic
Test Inspection Checklist, Revision 1, March 14,1981;
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The pressure test of Standby Liquid Control System was
performed March 17, 1981 using a special test procedure,
SP 80-1-38, Hydrostatic Pressure Test of the Standby Liquid
Control System Valves 1-SL-9 and 1-SL-7, Revision 0, June 18,
1980;
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A portion of piping penetrations for,the Reactor Core Spray
System was replaced with new material per Plant Design
Change Request (PDCR) No. 1-58-76 and PDCR Design Review and
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Safety Evaluation dated October 12, 1980.
The separate
hydrostatic test for the new core spray penetrations was
performed March 9, 1981 in accordance with the procedure SP
80-1-16, Hydrostatic Test of Reactor Core Spray Loops A and
B, Revision 0, June 18, 1980.
No unacceptable conditions were identified.
(3) Reactor Core Verification
The inspector determined that the core verification test was
performed April 1,1981 using procedure RE 1077, Reactor Core
Verification, Revision 3, March 20,1981, in which the following
were identified:
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Fuel assembly serial numbers were in accordance with the
core maps;
Fuel assemblies were properly oriented and seated;
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Channel fasteners were intact;
No foreign matter was found on fuel assemblies.
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The inspector had no further questions.
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3.
Cycle 8 Startup Testing - Post-critical Tests
a.
The inspector reviewed selected test programs to verify the follow-
ing:
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The test programs were implemented in accordance with Cycle 8
Refueling Sequencing Procedures;
Step-wise instructions for test procedures were adequately
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provided, including Precautions, Limitations and Acceptance
Criteria, in conformance with the requirements of the Technical
Specifications;.
Provisions to recover from anomalous conditions were provided;
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Methods and calculations were clearly specified and the tests
were performed accordingly;
Review, Approval, and Documentation of the results were in
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accordance witn the requirements of the Technical Specifi-
cations and the licensee's administrative controls.
b.
The following programs were reviewed:
(1) Shutdown Margin Demonstration
The inspector verified by an independent calculation and
review of the Shutdown Margin (SDM) procedure and test data per-
formed April 17, 1981, that the SDM with the strongest control
rod fully withdrawn was 2.13%Ak/k. Technical Specification 4.3.A.~1 requires SDM value greater than 0.33%Ak/k.
The SDM was determined by the "in-sequence" method in the
Xenon-free state and with tae moderator temperature of 172'F.
The reactor achieved criticality at 2147 hours0.0248 days <br />0.596 hours <br />0.00355 weeks <br />8.169335e-4 months <br />, April 17,
1981. The critical rod configuration was attained with
the rod 22-15 in group 2 at 16 notch position > " with the
reactor period of 290 seconds.
The test results were:
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Reactivity Contribution
Reactivity, %Ak/k
Control Rod Worth Less
(Strongest Rod: 30-23)
- 2.86
Rod Worth at Criticality
+ 5.10
Temperature Coefficient Correction
- 0.08881
Period Correction
- 0.023
Maximum decrease from BOC(R)
G0
2.13
Technical Specification 3.3.A.1
0.33
The documents reviewed were:
SP 690B, Reactivity Margin - Core Lo'ading Shutdown Margin
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Test, Revision 4, October 31, 1979.
Operations Form 690B-1, Data Sheet, Revision 3, November 14,
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.1979. Test performed April 17, 1981.
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Cycle 8, Millstone Unit 1 Cycle Management Report, Revision 0,
GE Document No. 22A6889, November 14, 1980.
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Findings
Operations Form 6908-1 requires extensive utilization of the
figures and tables in the cycle management report in order to
calculate the various reactivity contributions to the SDM. The
inspector identified the following inadequacy and incorrect
instruction in the Operations Form 6908-1:
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The formula in step 7.4.5 of the Operations Form 6908-1
specified that the item (4) be subtracted from 1.000 for
the SDMK
calculation.
A correct sign should be "+"
eff
rather than " ".
Consequently, the shutdown margin obtained
was incorrect;
Item (1) in step 7.4.1 requires data from Table 5, Section 7
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of the cycle management report. This was not specified in
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the procedure;
Step 7.4.1, item (2), and step 7.4.3 require information
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from figure 10,_B-sequence, and figure 11, respectively,
in the management report. -The procedure did not specify
the references nor were the figures attached;
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Table 5, Section 7 of the cycle management report is required
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to calculate the quantities identified in the procedural
steps 7.4.2 and 7.4.4.
The procedure did not provide the
above information.
The above failures constitute an example of noncompliance summarized
in section 4.
(2) Critical Rod Configuration and Reactivity Anomaly
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The inspector reviewed procedure SP 1050, Critical Rod Configuration
Comparison, Revision 1, July 26,1978 and test data of SP Form
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1050-1 performed June 29, 1981, and verified that the actual control
rod configuration was within 1%Ak/k of the predicted value as
specified in Technical Specification 3.3.E.
The table below
summarizes the reactivity anomaly test results. The inspector.
calculated the results iadependently from the data sheet 1050-1
and figures in the GE Cycle Management Report:
Steady State
Critical
(157.43 MWD /ST Burnup)
Notches Withdrawn
Actual
1204
6390
Predicted
1008
6344
Reactivity, %Ak/k
Actual
5.35
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Predicted
4.45
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Difference
0.90
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Acceptance *
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Notches Inserted
Actual
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570
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Predicted
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616
Acceptance **
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( <l .0%Ak/k)
- figure 10, cycle 8, GE Cycle Management Report
- figure 6, Cycle 8, GE Cycle Management Report
Findings
The procedure SP 1050 requires recording of a reactor period
in seconds on the SP Form 1050-1. The test data for June 29,
1981, did not have the period recorded.
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Furthermore, neither procedure SP 1050 nor SP Form 1050-1
specified acceptance criteria or re ferences to determine the
acceptability of the test data. The inspector further determined
that the control rod notches withdrawn as specified in the data
form were not directly comparable with the figures in the cycle
management report to determine the test acceptability. The summary
presented in the previous table was constructed by the inspector,
employing the following calculational steps:
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The actual and predicted notches withdrawn at the
critical state were 1204 and 1008 notches, respectively,
on the data sheet SP Form 1050-1. These notches had to
be converted into the numbers of rods withdrawn by
dividing by 48. Entering these numbers (25.1 and 21
rods each) into the abscissa of the figure 10, cycle
management report, the ordinate readings would give the
reactivity worths for the predicted and the actual
configurations. The difference of the two would be
compared with Technical Specification limit. None of
the above steps and reference figures were specified in the
procedure.
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During the steady state operation with a significant
number of rods withdrawn, the upper and the lower curves-
in figure 6 of the report would provide the upper
and the lower bounds of the Technical Specification limits
in terms of the number of rods inserted, and the middle
curve would be the predicted number of rods inserted. The
number of notches withdrawn on SP Form Form 1050-1 had to
be converted into the notches inserted by subtracting the
withdrawn notches from 145x48 = 6960, and then by dividing
the notches inserted by 48.
None of the above steps and reference figures were either
identified or attached in t's procedures.
The above findings constitute an example of noncompliance as
suinmarized in section 4.
(3) Traversing Incore Probes (TIP) - Asymmetry Check
The inspector reviewed the TIP traces and test conducted on
July 28, 1981 to ascertain plant TIP uncertainty. The inspector
determined that the test was performed in accordance with the
procedure RE 1058, TIP uncertainty, Revision 0, June 29, 1980 and
the TIP uncertainty of 5.699% was within the acceptable deviation
of 7% recommended by the General Electric Special Procedure
78-1-37, Revision 0, March 19, 1978.
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The Standard Deviations for the TIP noise and the location
uncertainty were 1.806% and 5.978% respectively.
The TIP
uncertainty due to the location and the noise perturbations was
obtained by taking a geometric average of the two separate
contributions.
The inspector had no further questions.
(4) TIP - Hot Alignment
The inspector reviewed procedure IC-405L. Axial Alignment of
Traversing Incore Probes, Revision 2, June 18, 1980 and I&C
Form 405L data sheet performed July 28, 1981. The inspector
verified that Turn-Around-Margin of all four TIP's were
greater than 2 inches. This satisfied the requirements.
The inspector had no further questions.
(5) Jet Pump Baseline Data Collection
The inspector verified that the baseline data, which included
recirculation flow, motor generator set speed, and scoop tube
position, were obtained in accordance with Procedurc SP 1052,
Revision 2, July 10,1980.
The jet pump baseline data reviewed were:
Power, %
Test Date
62.00
7-3-81
68.89
7-3-81
79.50
7-3-81
90.20
.7-4-81
99.20
7-8-81
The inspector had no further questions.
(6) Fuel Parameter Verification
The inspector reviewed procedure RE 1057, Process Computer
Fuel Parameter Input Verification, Revision 0, March 28, 1981,
and determined that the process computer inputs of April 16, 1981,
for Maximum Planar Linear Heat Generation Rate (MPLHGR) and
Minimum Critical Power Ratio (MCPR) were in accordance with
Technical Specification 3.11.1, figures a, b, c, d, e, f, and g.
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Findings
The objective of this test was to verify the process computer
inputs against MAPLHGR and MCPR limits specified in Technical Specifications 3.11.1 figures for seven different fuel types.
The test data packages did not specify the acceptance criteria
or references, nor was there any objective evidence that the
process computer input for April 16, 1981 was verified against
the figures in Technical Specifications.
These findings are an example of noncompliance summarized in
section 4.
(7) Exposure Surveillance
The inspector reviewed procedure SP 1041, Planar Average Exposure
Surveillance, Revision 2. September 10, 1980, and tests
conducted on June 26 and July 31, 1981, including the "on-demand"
computer printouts.
The inspector verified by an independent
review of the test results that both results were within the
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- requirements specified in Technical Specifications.
The two test results were:
6/26/81
7/31/81
Fuel Assembly XX-ZZ
with >23,5000 MWD /T
03-16
03-16
Maximum Nodal Exposure
ZZ
08
08
Exposure, MWD /T
29,438
29,809
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Acceptance, MWD /T
30,000
30,00'
MAPLHGR, KW/ft
3.343
3.608
LIMLHGR, KW/ft
10.756
10.719
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Findings
The precedure SP 1041 requires that fuel bundles exposed to
more than 23,500 MWD /T in average, be identified from the on-
demand computer program 0D-10, option 24. The procedure further
requires that the Maximum Average Nodal Exposure, its MAPLHGR
and LIMLHGR be calculated for the identified bundles from tha
0D-6, Option 2, to ensure that the maximum nodal exposure is less
than 30,000 MWD /T and that the thermal hydraulic parameters
are within the limits.
It also requires that the computer
calculations are summarized onto the RE Form 1041-1.
The inspector identified the following deficiencies for the
test performed June 26, 1981:
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Test package file did not have the on-demand program output
of 00-10, option 24, as required by the procedure SP 1041.
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Incorrect MAPLHGR and LIMLHGR were recorded. The correct
values for 03-16-08 bundle should be 3.343 KW/ft and
10.756 KW/ft respectively, instead of 3.484 and 10.891
KW/ft as entered on the data sheet.
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Shift Supervisor and Department Head who reviewed the data
failed to recognize the errors.
The above failures to document and to properly review the test
results constitute an example of noncompliance summarized in
section 4.
(8) Calibration of Intermediate Range Monitors (IRM)
The inspector verified that IRM readings were calibrated against
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Average Power Range Monitor (APRM) readings on April 19,.1981,.
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in accordance with procedure SP 1055, IRM Calibration
Data, Revision 1, September 24, 1979.
The test results showed
9% differer.ce between IRM and APRM readings at 15.5% core thermal
power. Subsequent IRM corrective calibrations were performed
April 27, 1981, using SP 402C, IRM Calibration Test, Revision 4,
June 18, 1980.
Findings
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The inspector noted that the acceptance criteria were not
specified in the procedure SP 1055, and the corrective calibration
was not cross-referenced in the test package, even though the
excessiva deviation between IRM and APRM readings were recognized
and a subsequent corrective. calibration was performed April 27, 1981.
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The above findings constitute an example of noncompliance
summarized in section 4.
9) local Power Range Monitors (LPRM) Calibration
1
The inspector verified that LPRM calibration and gain adjustment
were performed June 24, 1981, in accordance with the procedure
RE 1003, Revision 3, July 1,1977.
The inspector had no further questions.
(10) Core Thermal Power
The inspector verified that APRM's were calibrated as per
procedure SP 1040, APRM Calibration Using Heat Balance, Revision 1,
May 4, 1979. The calibration / gain adjustment factors reviewed
were the test data performed July 2, 10, 17, 31, and June 21,
1981.
The core thermal power was measured using procedure RE 1002,
Core Heat Balance - Power Range, Revision 7, June 18, 1980.
Core thermal power calculations by different methods were in
good agreement as shown in the following:
Method
Core Thermal Power, MW
t
Hand Calculations
2005.54
P-4*
2004.10
00-3*
2000.53
instantaneous heat balance calcula.tton.
The inspector had no further questions.
4.
Noncompliance
The inspector identified one item of noncompliance in the activities and
procedures related to
quality and
safety:
Failure to provide
adequate instructions and acceptance criteria for procedures; to properly
review and document the test results; and to follow written procedures.
Appendix B, Part 50, 10 CFR, specifies requirements for the activities
affecting quality'in that:
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Criterion V, " Instructions, Procedures, and Drawings," requires such
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activities be prescribed by procedures and be accomplished in
accordance with the instructions or procedures, and further requires
that the instructions or procedures shall include appropriate
quantitative or qualitative acceptance criteria for determining
satisfactory accomplishment of the activities.
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Criterion XI, " Test Control," specifies that all testing be performed
in accordance with the written procedures, which include the require-
ments and the acceptance limits contained in applicable design
documents. The criterion further requires that test results be
documented and evaluated to assure that test requirements have been
satisfied.
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Criterion XVI, " Corrective Action," requires to identify
malfunctions, deficiencies, and defective equipment, and to correct
the adverse conditions promptly. The criterion further specifies that
the corrective action shall be documented.
Criterion XVII, " Quality Assurance Records," requires that test
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records include, among others, the results, the acceptability, and
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action taken in connection with deficiencies noted. Also, the records
shall be identifiable and retrievable.
Contrary to the above, the following examples of findings were identified
as detailed in 2.b(1), 3.b(1), 3.b(2), 4.b(4), 4.b(5), and 4.b(6):
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Details in item 2.b(1): Average scram insertion times for test
performed April 19, 1981 were not calculated and entered as per
procedure SP 1051 and RE Form 1051-3; ' there was no objective evidence
of documentation in the package that corrective action was taken as
per Technical Specification 3.3.C when the test performed April 19,
1981 failed.
The Replacement Job Order number was not documented in
the test package.
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Details in item 3.b(ll:
Procedure SP 690B and Operations Form 6908-1
did not provide sufficient information or references to conduct the
test calculations of SDM; formula in step 7.4.5 of Operations Form
6908-1 was incorrect; consequently, the SDM calculated on April 17,
1981 was incorrect.
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Details in item 3.b(2):
Reactor period was not entered in the data
sheet, SP Form 1050-1, during the test performed June 29, 1981;
Procedure SP 1050 and the associated data form did not provide
sufficient data reduction details, references, and acceptance
criteria.
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Details in item 3.b(6):
Reactor Engineering Proceuure RE 1057 did not
have acceptance criteria or references; documents for the verification
test conducted on April 16, 1981 did not have any objective evidence
that the input parameters were verified against the figures in Technical Specifications 3.11.1.
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Details in item 4.b(7):
The on-demand program outputs 00-10, option
24, for June 26, 1981 had not been documented as required by the
procedure SP 1041;
incorrect fuel parameters were recorded in RE Form
1041-1 for the test performed June 26, 1981; reviewers failed to
recognize the errors.
Details in i; tem 3.b(8): Acceptance criteria were not provided in the
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procedure SP 1055; corrective calibration performed April 27, 1981 was
not identitled in the test packages, in which the deficiency was
identified.
The licensee representative stated that proper remedial steps would be
taken to correct the above findings, and the necessary procedure revisions
would be completed prior to September 1, 1982.
The above failures to meet the requirements specified in the Code of Federal
Regulations, Appendix B to 10 CFR 50, Criteria V, XI, XVI, and XVII, collect-
ively constitute an item of noncompliance (50-245/81-13-01).
5.
Post Refueling Startup Test Review
a.
The inspector reviewed the Cycle 8, Startup Book, April 15, 1980, to
verify that:
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Test sequence was consistent with the requirements;
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Testing was conducted in accordance with the sequencing Startup
Book;
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Sequencing Startup Book was properly reviewed and documented;
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The startup report was issued.
b.
The inspector noted, during the review of the Cycle 8 Startup Book,
+ hat the final review of the test results was not completed. A licensee
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representative stated that as soon as the departmental reviews were
completed, the Startup Book would be signed off and be reviewed by
Plant Operations Review Committee (PORC). The unreviewed tests were:
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Hot TIP alignment
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Jet pump baseline data
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TIP asymmetry check
The Cycle 8 startup test summary report was not yet av:ilable, and
the inspector was informed that the summary report would be submitted
to NRC shortly as required by Technical Specifications.
No unacceptable conditions were identified.
6.
Control Room Observations and Facility Tours
The inspector observed Control Room Operation for shift turnover and log
sheets, and facility operation in accordance with the administrative
procedures and Technical Specification requirements.
Inspection tours of
the Reactor Vessel penetrations and pipings to and from the Isolation
Condenser were conducted to visually observe the evidence of movements of
the pipings and the support structures caused by the recent water hammer.
The inspector noted that the several bolts holding the piping supports to
the concrete floor for the isolation condenser had been lifted up to 1/8",
possibly due to the impa-t forces. The inspector did not find any
permanent structural da....ges from apparent piping movements. The inspector
discussed the observations and the concerns with the resident inspectors.
The inspector had no further questions.
7.
Entrance and Exit Interviews
Licensee management was informed of the purpose and scope of the inspection
at the entrance interview, and the findings of the inspection were
periodically discussed and were summarized at the conclusion of the
inspection on August 14, 1981. Attendees at the exit interview are denoted
in paragraph 1.
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