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{{Adams | |||
| number = ML20197C249 | |||
| issue date = 05/02/1986 | |||
| title = Insp Rept 50-423/86-09 on 860218-0314.Violation Noted: Failure to Implement Defined QC Surveillance Program | |||
| author name = Eselgroth P, Prell J, Wen P | |||
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) | |||
| addressee name = | |||
| addressee affiliation = | |||
| docket = 05000423 | |||
| license number = | |||
| contact person = | |||
| document report number = 50-423-86-09, 50-423-86-9, NUDOCS 8605130275 | |||
| package number = ML20197C228 | |||
| document type = INSPECTION REPORT, NRC-GENERATED, INSPECTION REPORT, UTILITY, TEXT-INSPECTION & AUDIT & I&E CIRCULARS | |||
| page count = 13 | |||
}} | |||
See also: [[see also::IR 05000423/1986009]] | |||
=Text= | |||
{{#Wiki_filter:. | |||
~. | |||
. | |||
U.S. NUCLEAR REGULATORY COMMISSION | |||
REGION I | |||
Report No. 50-423/86-09 | |||
Docket No. 50-423 | |||
License No. NPF-49 | |||
Licensee: Northeast Nuclear Energy Company | |||
P. O. Box 270 | |||
Hartford, Connecticut 06141-0270 - | |||
Facility Name: Millstone Nuclear Power Station, Unit 3 | |||
Inspection At: Waterford, Connecticut | |||
Inspection Conducted: February 18-March 14,1986 | |||
Inspectors: .' 1h [ | |||
J rell, Reactor Engineer / daYe | |||
v& c. w- | |||
~ | |||
w/ uhs | |||
' da te' | |||
P. Wen,RactorEng/neer | |||
Approved by: - | |||
5/2./86 | |||
F. Ese'Tgrot , Chief, Test Programs 'ddte | |||
Section, , DRS | |||
Inspection Summary: | |||
Routine Unannounced Inspection Conducted On February 18-March 14, 1986 | |||
(Report No. 50-423/86-09) | |||
Areas Inspected: Startup program review, power ascension test procedure | |||
review, test results review, test witnessing, and review of licensee actions | |||
on previous findings. | |||
Results: One violation was identified (failure to implement the defined QC | |||
surveillance program). | |||
. | |||
gnos'd @#dl | |||
O | |||
- | |||
. | |||
. . | |||
. | |||
* | |||
DETAILS | |||
1. Persons Contacted | |||
*G. Clossius, QA Supervisor, NNECO | |||
+J. Crockett, MP-3 Unit Superintendent, NNECO | |||
P. Finck, Startup Engineer | |||
i | |||
E. Fries, Startup Engineer, NNECO | |||
-*J. Jensen, QA Specialist, NNECO | |||
r E. Laware, QA Engineer, NNECO | |||
, C. Libby, Operations QA Supervisor, NUSCO | |||
*+D. McDaniel, Reactor Engineer | |||
D. Miller, Startup Manager, NNECO | |||
D. Moore, Assistant Operations Supervisor, NNECO | |||
M. Pearson, Assistant Operations Supervisor, NNECO | |||
, +W.-Richter, Assistant Startup Supervisor, NNECO | |||
U.S. Nuclear Regulatory Commission | |||
+F. Casella, Resident Inspector | |||
T. Rebelowski, Senior Resident Inspector | |||
+T. Shedlosky, Senior Resident Inspector | |||
. | |||
I + Denotes those present at nini exit meeting on February 28, 1986 | |||
* Denotes those present at the exit meeting on March 14, 1986. The | |||
inspectors also interviewed other personnel during this inspection period. | |||
2.0 Licensee Action of Previous Inspection Findings | |||
(0 pen) Unresolved Item (50-423/85-34-03) pertaining to qualification of | |||
the containment High Range Radiation Monitor, Kaman Model KDI-1000 and | |||
< Mineral Insulation Cable Assembly. | |||
; -- | |||
Qualification of Model KDI-1000 High Range Containment Area | |||
Radiation Detector and Mineral Insulation Cable Assembly. Report | |||
No. 46-0036-001, Revision A. | |||
4 -- | |||
Qualification Report for Model KDI-1000 High Range Containment Area | |||
Radiation Detector and Mineral Insulation Cable System. Report No. | |||
460036-002, Revision 2. | |||
l | |||
-- | |||
Installed Specification for Mineral Triaxial Cable Penetration | |||
Assembly. Report No. 460036-002, Revision A. | |||
. | |||
' | |||
-- | |||
Installation Specification for Mineral Insulation (MI) Cables for | |||
High Range In-Containment Area Monitors. | |||
In reviewing the above documents the inspector concluded that additional | |||
test data needs to be made available for review in order to conclude that | |||
the Containment High Range Radiation Monitor, Kaman Model KDI-1000 and MI | |||
Cable Assembly, will function as required in a LOCA environment. | |||
. | |||
. | |||
3 | |||
. | |||
The basis for this determination is the numerous deviations, anomalies | |||
and failure reports (DAFR) identified in the qualification report No. | |||
46C036-002, Revision 2 which reduced acceptance criteria for problem | |||
areas. | |||
For example; | |||
e | |||
-- | |||
DAFR-9, -10, -13 reduced torque cycle requirements in an attempt to | |||
correct relaxation problems of torqued connectors. | |||
-- | |||
DAFR-14 deletes the voltage withstand test of signal cables as a pre- | |||
seismic functional test. | |||
-- | |||
DAFR-16 requires the chemical spray to be applied only during non- | |||
superheated conditions. | |||
-- | |||
The thermal transients behavior effect inside the cable during rapid | |||
temperature transients add a false signal to the radiation signal. | |||
-- | |||
There are numerous references to test failures attributed to | |||
moisture intrusion / contamination due to a relaxation of torqued | |||
connectors whicn compromises the seal. | |||
-- | |||
The critical assembly / handling requirements for connector / cables re- | |||
quires a clean room type facility when servicing or calibrating the | |||
monitoring system. This is not found inside containment. | |||
-- | |||
The MI cables are subject to rupture from mishandling, surface | |||
scratches and breaking of the sheath due to excessive bending. | |||
The licensee has experienced considerable difficulties in installing a | |||
functional Kaman High Range Containment Monitor due to problems in finding | |||
an acceptable length of MI cable assembly. As a result, the licensee has | |||
received an NRC exception to use an acceptable shortened cable assembly | |||
with one Kaman monitor installed at a lower containment elevation. Two | |||
other monitors (General Atomic) have been located inside containment at | |||
the proper elevation. | |||
This item remains open pending receipt of conclusive environmental test | |||
data that ensures equipment availability in compliance with the require- | |||
ments of NUREG-0737, Table II F.1-3. | |||
3.0 Startuo Test Program | |||
3.1 Procedure Review | |||
Scope | |||
The following approved power ascension test procedure was reviewed | |||
for technical and administrative adequacy and to verify that test | |||
planning satisfies regulatory guidance and licensee commitments: | |||
3-INT-8000 Appendix 8032, Revision 0, Generator Trip From 100% Power. | |||
_ - _ _ - | |||
, | |||
... | |||
. | |||
. | |||
4 | |||
. | |||
Discussion | |||
The above procedure was examined for: management review and approval; | |||
procedure format; clarity of stated test objectives; prerequisites; | |||
environmental conditions; acceptance criteria; source of acceptance | |||
criteria; references; initial conditions; attainment of test objec- | |||
tives;. test performance documentation and verification; degree of | |||
detail for test instructions; restoration of system to normal after | |||
testing; identification of test personnel; evaluation of test data; | |||
independent verification of critical steps or parameters and quality | |||
control and assurance involvement. | |||
Findings | |||
The review indicated that the procedure was consistent with regula- | |||
l tory requirements, guidance, and with the Licensee's commitments. | |||
No discrepancies or unacceptable conditions were identified. The | |||
inspector had no further questions on this procedure. | |||
3.2 Test Witnessing | |||
Scope | |||
The following power ascension tests were witnessed te verify the 11- | |||
censee's conformance to regulatory and procedural requirements, to | |||
observe the performance of the operating staff, and to ascertain the | |||
adequacy of test program' records, including preliminary evaluation | |||
of test results: | |||
3-INT-8000 Appendix 8023, Revision 0, Reactor Trip and Shutdown From | |||
Outside the Control Room. | |||
3-INT-8000 Appendix 8006, Revision 0, Secondary Plant Performance | |||
Test. | |||
Discussion | |||
The above tests were witne:/,ed to verify that; the operating crew | |||
was using the most current test procedure, minimum crew requirements | |||
were met, all test prere:uisites and initial conditions were met, | |||
required test equipment was properly calibrated, the implementing | |||
procedures were technically adequate to perform the test, crew | |||
actions during the test were correct and timely, a quick analysis , | |||
and evaluation was made to assure proper plant response to the test, i | |||
all test data were collected for analysis by the appropriate per- l | |||
sonnel, overall test acceptance requirements were met and adherence , | |||
to TS requirements was met for any LCO's affected by the test. i | |||
! | |||
I | |||
l | |||
l | |||
, | |||
. | |||
. . | |||
5 | |||
. | |||
FINDINGS | |||
,' | |||
-- | |||
Appendix 8023, " Reactor Trip and Shutdown from Outside the Con- | |||
trol Room" - The inspector verified that the test satisfied R.G. | |||
1.68 Appendix A, paragraph 5.d.d and FSAR requirements. Credit | |||
was taken for a test performed during the pre-core hot func- | |||
tional test program which demonstrated that the plant could be | |||
placed from a hot standby to a hot shutdown condition. There- | |||
fore, this portion of the test was not conducted. | |||
1 | |||
[ -The inspector observed the pretest briefings of both the Test crew, | |||
located at the Auxiliary Shutdown Panel, and the Normal crew located | |||
in the Control Room. The briefing of the Test crew included a step | |||
L by step review of the procedure, identification of the various con- | |||
l trols and predicted plant indications and alerting the crew to actions | |||
to be taken in the event unsafe conditions developed during the test. | |||
The Normal crew was briefed on the purpose of test and the actions | |||
they could and could not perform during the test. | |||
It was verified that communications were established and maintained | |||
during the test between the Auxiliary Shutdown Panel (ASP) and the | |||
Control Room, Control Transfer Panels (orange and purple trains), | |||
the Auxiliary Building and the Engineer Safeguards Features Building. l | |||
i | |||
i | |||
The reactor was manually tripped from 15% power with the CRDM breaker | |||
located in the Auxiliary Building. Emergency Operating Procedure E0P | |||
3503 " Shutdown Outside Control Room," Revision 2 was used to shutdown | |||
the reactor and transfer control to the ASP. It was verified that | |||
the test operators were able to maintain Tave at 557*F for over 1/2 | |||
hour from the ASP and then transfer control of the plant back to the | |||
Control Room. No problems were identified with the test, test results l | |||
or performance of the operating crew. A problem was identified with i | |||
QC involvement with this test. See paragraph 4.0 "QA/QC Interface" l | |||
for details. - | |||
i | |||
-- | |||
Appendix 8006, " Secondary Plant Performance Test". This test is | |||
performed at various power levels - 30%, 40%, 50%, 75%, 90% and | |||
100% - in order to obtain base line data and identify any sec- | |||
ondary plant problems. The inspector witnessed this test at the | |||
30% power level. It was verified that the latest procedures | |||
were being used, the test crew was knowledgeable of the program | |||
and their responsibilities and the test data was being reviewed | |||
and evaluated against expected results. | |||
No problems were identified. | |||
l | |||
l | |||
- _ _ _ . | |||
. | |||
, | |||
. | |||
6 | |||
. | |||
3.3 Startup Test Results Evaluation | |||
Scope | |||
The test data results from the tests listed in Appendix A were review- | |||
ed to verify that adequate testing had been accomplished. The results | |||
were also reviewed to verify if regulatory guidance and licensee com- | |||
mitments were satistied and to ascertain whether uniform criteria were | |||
being applied in the evaluation of completed tests in order to assure | |||
their technical and administrative adequacy. | |||
Discussion | |||
The inspector reviewed the test results and verified the licensee's | |||
evaluation of test results by review of: test changes; test excep- | |||
tions; test deficiencies, "as-run" copy of the test procedure; accep- | |||
tance criteria; performance verification; recording of the conduct of | |||
tests; QA inspection records; restoration of system to normal after | |||
the test; independent verification of critical steps or parameters; | |||
identification of personnel cenducting and evaluating test data; and | |||
verification that the test results had been reviewed and approved by | |||
licensee management. | |||
FINDINGS | |||
Turbine Overspeed Test ( Appendix 8016) | |||
The capability of the mechanical overspeed trip device for the main | |||
turbine generator was tested on February 15, 1986. Three (3) actual | |||
trips occurred at 1962, 1963, 1963 rpm, respectively. This test was | |||
accepted as satisfactory since the measured values were within test | |||
procedure established criterion of s1998 rpm. The Backup Overspeed , | |||
Trip (BOST) Normal Mode is set approximately 0.5% above the mechanical 1 | |||
overspeed trip setting, to provide additional protection against tur- | |||
bine overspeed during plant normal operation. The BOST electronic | |||
circuit was also tested at 105% rated speed prior to performing me- | |||
chanical overspeed trip test. The test result was also satisfactory. | |||
Reactor Trip / Shutdown Outside Control Room (Appendix 8023) | |||
The objectives of this test was to demonstrate the following: | |||
a. Verification that the unit can be safely shutdown from outside | |||
the control room using the Auxiliary Shutdown Panel, | |||
l | |||
- -- . .- . - - | |||
-____ _ _ _ _ _ _ _ _ - _ _ _ _ - - - _ . .__ | |||
. | |||
. | |||
. | |||
. | |||
y | |||
b. Verification that the unit can be maintained in a hot standby | |||
condition from outside the control room. | |||
c. Verification that the unit can be cooled down to the hot shutdown | |||
condition from outside the control room. | |||
Through test results review and test witnessing, the inspector noted | |||
the licensee has demonstrated the remete shutdown capability and met | |||
the above test objectives with the exception of item c. | |||
The cold shutdown capab'lity (item c) was demonstrated earlier during | |||
preoperational test, 3-INT-3000, Appendix 3014, " Remote Shutdown with | |||
Cooldown" on November 1, 1985. | |||
The remote shutdown capability test was accepted as satisfactory by | |||
the JTG. | |||
Automatic Reactor Control (Appendix 8017) | |||
The performance of the automatic control system in raintaining reactor | |||
coolant average temperature within programmed value was verified per | |||
test procedure 3-INT-8000, Appendix S017 on February 17, 1986. The | |||
inspector independently reviewed test data and noted that automatic | |||
reactor control system responded well during the test. Tavg returned | |||
to programmed value withia specified oscillation amplitude, and no | |||
unstable response was observed. The corresponding system responses | |||
such as pressurizer pressu. e, pressurizer level and steam generator | |||
level responded well with no unusual behavior noticed. | |||
During performance of this test, Step 5.5 required the feedwater pump l | |||
speed control system to be set per 3-INT-3000, Appendix 3010. A test l | |||
engineer misread Appendix 3010 to be Appendix 8010. Since Appendix | |||
8010, Neutron Shield Tank Testing, has nothing to do with this test, | |||
Step 5.5 was deleted during the test. The inspector informed the 11- | |||
censee of this discrepancy. The test engineer re- eviewed the test | |||
result and determined that this test change did not have adverse im- | |||
pact on the test outcome, since Appendix 3010 had already been com- | |||
pleted. The inspector concurred with the licensee's assessment and | |||
had no further questions. | |||
Power Coefficient (Appendix 8020) | |||
The doppler-only power coefficient measurements are to be obtained at | |||
30, 50, 75, 90 and 100 percent power. The measurement at 30% power | |||
level was performed on February 17, 1986. | |||
I | |||
_ _ _ _ _ = _ _ - _ _- - - | |||
- _____ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . | |||
- . | |||
- | |||
. | |||
8 | |||
Since it is difficult to directly measure the reactivity change due | |||
to fuel temperature change during a power change, an indirect measure- | |||
; ment technique was utilized. This technique involved measuring the | |||
l primary side responses such as the change in core average temperature | |||
(ATAVG) and the corresponding change in RCS loop parameters with re- | |||
spect to a small change in turbine load. The doppler-only power co- | |||
efficient was then calculated from this set of data and isothermal | |||
i temperature coefficient information as previously derived from Zero | |||
. | |||
Power Physics Test. | |||
I | |||
All test results, including five sets of power swing data, met the | |||
test acceptance criteria. | |||
The inspector noted that an assumption of dTm=dff (Where T,is | |||
moderator temperature and T f | |||
is average fuel temperature) was used | |||
in the licensee's measuring methodology for the doppler-only power l | |||
coefficient derivation. The validity of this assumption was discussed ' | |||
with licensee reactor engineer and consultation provided by a NRR | |||
Core Physics specialist. This assumption is valid when there is no | |||
power change. From the test results and information provided by the l | |||
fuel nndor (Westinghouse), it appeared that this assumption is also ! | |||
valid under test cordition with minimum power change. | |||
l | |||
l | |||
The inspector had no further questions. i | |||
' | |||
hatural Circulation Test (Appendix 7006) | |||
' | |||
The test witness and preliminary test result review was documented in | |||
the Inspection Report 86-07. Although the licensee completed this , | |||
test and demonstrated that plant core heat can be removed satisfac- l' | |||
totily by using natural circulation, the detailed test results had | |||
not been thoroughly analyzed. From the test result review, the in- | |||
spector noted that PORV lifted 3 times during the initial stage of | |||
the natural circulation test. Plant behavior and lessons learned | |||
from this test will be evaluated by the licensee and incorporated in | |||
the licensee operator training program and possible procedure enhance- , | |||
ment. ! | |||
I | |||
The licensee management agreed that this detailed evaluation and ! | |||
incorporation into operator's training will be completed by May 1, ! | |||
1986. This is an unresolved item (50-423/86-09-01). | |||
RCS Flow Coastdown (Appendix 5017) | |||
The purpose of this test was to verify that the RCS responded as de- | |||
signed to a partial and complete loss of forced reactor coolant flow. | |||
With four reactor coolant pumps (PCPs) operating, partial loss of , | |||
forced reactor coolant flow was established by tripping one RCP. Com- l | |||
plete loss of reactor coolant flow was established by simultaneously | |||
tripping all four RCPs. ( | |||
_. - _ _-____ _ _ - - _ _ _ _ _ _ _ _ _ _ - _ - _ . _ _ _ _ _ _ _ - _ _ _ .- | |||
..., | |||
_ | |||
- | |||
. | |||
9 | |||
The inspector verified that low flow alarm times, control rod drop | |||
times,-core flows and low flow alarm values met FSAR Chapter 15 Table | |||
15.3-1 requirements. This was determined by reviewing the strip | |||
chart recordings and the sequence of events computer printouts for | |||
reactor coolant loop flow, RCP alarm, reactor trip times, and low | |||
flow' alarms. It was also verified that the licensee had correctly | |||
translated the information from the strip chart recorders and computer | |||
print-out to the data sheets and, by performing independent calcula- | |||
tions, that the final results were correct. | |||
Post Core Hot Functional Tests UNSATS | |||
A review of the disposition of all UNSATs foentified during the Post | |||
Core Hot Functional Test Program was made. These UNSAT's, identified | |||
in Appendix B, were reviewed for conformance with administrative re- | |||
quirements and proper disposition of problems. No problems were | |||
identified. | |||
4.0 QA/QC INTERFACE | |||
While observing the performance of 3-INT-8000 Appendix 8023, Reactor Trip / | |||
Shutdown Outside the Control Room, the inspector questioned a QC inspector | |||
also witnessing the test as to his involvement with the test. From these | |||
discussions the following information was determined: | |||
-- | |||
Although the QC inspector stated that he was there to verify operator | |||
compliance to E0P 3503, he had not yet reviewed E0P3503 or Appendix | |||
8023 nor did he have a copy of the procedures with him. | |||
-- | |||
The QC inspector had not been briefed by his management on what to | |||
look for or expect during the test. This was later verified by his | |||
immediate supervisor. | |||
-- | |||
The QC inspector had not reviewed the FSAR or RG 1.68 requirements | |||
pertaining to this test although he was aware he should do so. | |||
-- | |||
The QC inspector was not aware of what was taking place as evidenced | |||
by the fact he did not know from what power level the reactor had | |||
been tripped. | |||
The NUSCO QA and NNECO QC departments share responsibility for performing | |||
surveillances of startup tests. This responsibility is loosely coordin- | |||
ated with NUSCO QA having primary responsibility. During this particular | |||
test, representatives from both organizations were present. The auditor | |||
from the QA section appeared knowledgeable of the test procedure and what | |||
was taking place. | |||
t | |||
I | |||
_ __ _______ _ _ _ _ _ __ ___ _____________________ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ | |||
. | |||
*' | |||
10 | |||
The above occurrence of an inspector being assigned an inspection task | |||
for which that inspector had not been trained or briefed constitutes a | |||
violation (50-423/86-09-02). | |||
At the exit meeting, the licensee QA representative agreed that: | |||
-- | |||
More specific guidance will be provided to QC inspectors who will | |||
cover the remaining startup tests, as to which areas / criteria to | |||
look into. | |||
-- | |||
NUSCO QA will continue to provide surveillances during the startup | |||
test program. | |||
The following NUSCO Operations QA surveillance reports were reviewed to | |||
determine the adequacy of 00A's involvement with the Startup Test Program: | |||
NUSCO 00A SURVEILLANCE NO. TITLE | |||
TC 3950 Digital Rod Position Indication | |||
TC 3960 Boron Endpoint Measurements | |||
TC 3960A Boron Endpoint Measurements | |||
TC 3961 Initial Criticality | |||
TC 3961A Initial Criticality | |||
TC 3968 Natural Circulation | |||
TC 3972 Preparations for Power Ascension | |||
Testing | |||
TC 3976 Power Ascension Test | |||
TC 3986 Integrated Plant Testing | |||
TC 3986A Integrated Plant Testing | |||
TC 39868 Integrated Plant Testing | |||
The surveillance appeared to be thorough with good follow-up of identified | |||
concerns. No problems were identified. | |||
5.0 Independent Calculations | |||
3-INT-5000 Appendix 5010, "RTO Bypass Loop Verification" obtains data | |||
which is used to calculate the hot leg and cold leg flow and transport | |||
times through the RTO Bypass Lines. Using the formulas: | |||
F | |||
H,= Ft Fc' = Ft-Fgg | |||
(1+ Fc) , | |||
F l | |||
H - | |||
Fc' = Calculated cold leg flow | |||
FH ,= Calculated hot leg flow | |||
Fg= Measured Total flow | |||
l | |||
l, | |||
r 4 | |||
1 | |||
o | |||
. . | |||
11 | |||
l | |||
i | |||
Fc = Measured cold leg flow | |||
FH = Measured hot leg flow | |||
The inspector verified the flows and transport time for the hot and cold | |||
leg RTO bypass lines. | |||
Power coefficient measurement requires lengthy data reduction. The | |||
inspector independently verified that the predicted reactor physics | |||
parameters were correctly taken from the nuclear design reference. The | |||
inspector also performed an independent calculation and confirmed that ; | |||
the first power swing case data were being correctly reduced. | |||
6.0 Plant Tours | |||
The inspector made several tours of the facility durtag the course of the | |||
inspection. This included tours of the control building and control room. | |||
A review of the work in progress, security, cleanliness and housekeeping | |||
was made. | |||
7.0 Exit Meeting | |||
An exit meeting was held on March 14, 1986 to discuss the inspection | |||
scope and findings, as detailed in this report (see paragraph 1.0 for | |||
attendees). | |||
At no time was written material given to the licensee. The inspector | |||
determined that no proprietary information was utilized during this | |||
inspection. | |||
i | |||
, | |||
d | |||
.. | |||
.. | |||
APPEN0IX A | |||
TEST DATA REVIEWED | |||
TEST NUMBER TITLE | |||
3-INT-5000 Appendix 5001 Shutdown Margin | |||
3-INT-5000 Appendix 5010 RTD Bypass Loop Verification | |||
3-INT-5000 Appendix 5015 Oigital Rod Position Indication | |||
Operation Test | |||
3-INT-5000 Appendix 5016 Loose Parts Monitoring | |||
3-INT-5000 Appendix 5017 RCS Flow Coastdown | |||
3-INT-5000 Appendix 5031 Chemical and Volume Control System | |||
3-INT-6000 Initial Criticality | |||
3-INT-7000 Appendix 7006 Natural Circulation | |||
3-INT-8000 Appendix 8023 Reactor Trip / Shutdown Outside | |||
Control Room | |||
3-INT-8000 Appendix 8016 Turbine Overspeed Test | |||
3-INT-8000 Appendix 8017 Automatic Reactor Control | |||
3-INT-8000 Appendix 8020 Power Coefficient | |||
.. | |||
.. | |||
APPENDIX B | |||
POST CORE HOT FUNCTIONAL TEST UNSATS REVIEWED | |||
TEST PROCEDURE UNSAT #'s | |||
3-INT-5000, Anpendix 5001 7497 | |||
3-INT-5000, Appendix 5002 7471 | |||
3-INT-5000, Appendix 5004 7341, 7342 | |||
3-INT-5000, Appendix 5006 7495, 7492, 7493 | |||
3-INT-5000, Appendix 5007 7485, 7486, 7489, 7496 | |||
3-INT-5000, Appendix 5009 7466 | |||
3-INT-5000, Appendix 5015 7487 | |||
3-INT-5000, Appendix 5016 7475, 7479 | |||
3-INT-5000, Appendix 5017 7504, 7510 | |||
3-INT-5000, Appendix 5031 7472, 7473, 7474, 7476 | |||
7477, 7478, 7484, 7488 | |||
7490, 7491, 7499 | |||
3-INT-5000, Appendix 5033 7378, 7417, 7420 | |||
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}} |
Latest revision as of 07:34, 2 January 2021
ML20197C249 | |
Person / Time | |
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Site: | Millstone |
Issue date: | 05/02/1986 |
From: | Eselgroth P, Prell J, Wen P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
To: | |
Shared Package | |
ML20197C228 | List: |
References | |
50-423-86-09, 50-423-86-9, NUDOCS 8605130275 | |
Download: ML20197C249 (13) | |
See also: IR 05000423/1986009
Text
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No. 50-423/86-09
Docket No. 50-423
License No. NPF-49
Licensee: Northeast Nuclear Energy Company
P. O. Box 270
Hartford, Connecticut 06141-0270 -
Facility Name: Millstone Nuclear Power Station, Unit 3
Inspection At: Waterford, Connecticut
Inspection Conducted: February 18-March 14,1986
Inspectors: .' 1h [
J rell, Reactor Engineer / daYe
v& c. w-
~
w/ uhs
' da te'
P. Wen,RactorEng/neer
Approved by: -
5/2./86
F. Ese'Tgrot , Chief, Test Programs 'ddte
Section, , DRS
Inspection Summary:
Routine Unannounced Inspection Conducted On February 18-March 14, 1986
(Report No. 50-423/86-09)
Areas Inspected: Startup program review, power ascension test procedure
review, test results review, test witnessing, and review of licensee actions
on previous findings.
Results: One violation was identified (failure to implement the defined QC
surveillance program).
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DETAILS
1. Persons Contacted
+J. Crockett, MP-3 Unit Superintendent, NNECO
P. Finck, Startup Engineer
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E. Fries, Startup Engineer, NNECO
-*J. Jensen, QA Specialist, NNECO
r E. Laware, QA Engineer, NNECO
, C. Libby, Operations QA Supervisor, NUSCO
- +D. McDaniel, Reactor Engineer
D. Miller, Startup Manager, NNECO
D. Moore, Assistant Operations Supervisor, NNECO
M. Pearson, Assistant Operations Supervisor, NNECO
, +W.-Richter, Assistant Startup Supervisor, NNECO
U.S. Nuclear Regulatory Commission
+F. Casella, Resident Inspector
T. Rebelowski, Senior Resident Inspector
+T. Shedlosky, Senior Resident Inspector
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I + Denotes those present at nini exit meeting on February 28, 1986
- Denotes those present at the exit meeting on March 14, 1986. The
inspectors also interviewed other personnel during this inspection period.
2.0 Licensee Action of Previous Inspection Findings
(0 pen) Unresolved Item (50-423/85-34-03) pertaining to qualification of
the containment High Range Radiation Monitor, Kaman Model KDI-1000 and
< Mineral Insulation Cable Assembly.
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Qualification of Model KDI-1000 High Range Containment Area
Radiation Detector and Mineral Insulation Cable Assembly. Report
No. 46-0036-001, Revision A.
4 --
Qualification Report for Model KDI-1000 High Range Containment Area
Radiation Detector and Mineral Insulation Cable System. Report No.
460036-002, Revision 2.
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Installed Specification for Mineral Triaxial Cable Penetration
Assembly. Report No. 460036-002, Revision A.
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Installation Specification for Mineral Insulation (MI) Cables for
High Range In-Containment Area Monitors.
In reviewing the above documents the inspector concluded that additional
test data needs to be made available for review in order to conclude that
the Containment High Range Radiation Monitor, Kaman Model KDI-1000 and MI
Cable Assembly, will function as required in a LOCA environment.
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The basis for this determination is the numerous deviations, anomalies
and failure reports (DAFR) identified in the qualification report No.
46C036-002, Revision 2 which reduced acceptance criteria for problem
areas.
For example;
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DAFR-9, -10, -13 reduced torque cycle requirements in an attempt to
correct relaxation problems of torqued connectors.
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DAFR-14 deletes the voltage withstand test of signal cables as a pre-
seismic functional test.
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DAFR-16 requires the chemical spray to be applied only during non-
superheated conditions.
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The thermal transients behavior effect inside the cable during rapid
temperature transients add a false signal to the radiation signal.
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There are numerous references to test failures attributed to
moisture intrusion / contamination due to a relaxation of torqued
connectors whicn compromises the seal.
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The critical assembly / handling requirements for connector / cables re-
quires a clean room type facility when servicing or calibrating the
monitoring system. This is not found inside containment.
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The MI cables are subject to rupture from mishandling, surface
scratches and breaking of the sheath due to excessive bending.
The licensee has experienced considerable difficulties in installing a
functional Kaman High Range Containment Monitor due to problems in finding
an acceptable length of MI cable assembly. As a result, the licensee has
received an NRC exception to use an acceptable shortened cable assembly
with one Kaman monitor installed at a lower containment elevation. Two
other monitors (General Atomic) have been located inside containment at
the proper elevation.
This item remains open pending receipt of conclusive environmental test
data that ensures equipment availability in compliance with the require-
ments of NUREG-0737, Table II F.1-3.
3.0 Startuo Test Program
3.1 Procedure Review
Scope
The following approved power ascension test procedure was reviewed
for technical and administrative adequacy and to verify that test
planning satisfies regulatory guidance and licensee commitments:
3-INT-8000 Appendix 8032, Revision 0, Generator Trip From 100% Power.
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Discussion
The above procedure was examined for: management review and approval;
procedure format; clarity of stated test objectives; prerequisites;
environmental conditions; acceptance criteria; source of acceptance
criteria; references; initial conditions; attainment of test objec-
tives;. test performance documentation and verification; degree of
detail for test instructions; restoration of system to normal after
testing; identification of test personnel; evaluation of test data;
independent verification of critical steps or parameters and quality
control and assurance involvement.
Findings
The review indicated that the procedure was consistent with regula-
l tory requirements, guidance, and with the Licensee's commitments.
No discrepancies or unacceptable conditions were identified. The
inspector had no further questions on this procedure.
3.2 Test Witnessing
Scope
The following power ascension tests were witnessed te verify the 11-
censee's conformance to regulatory and procedural requirements, to
observe the performance of the operating staff, and to ascertain the
adequacy of test program' records, including preliminary evaluation
of test results:
3-INT-8000 Appendix 8023, Revision 0, Reactor Trip and Shutdown From
Outside the Control Room.
3-INT-8000 Appendix 8006, Revision 0, Secondary Plant Performance
Test.
Discussion
The above tests were witne:/,ed to verify that; the operating crew
was using the most current test procedure, minimum crew requirements
were met, all test prere:uisites and initial conditions were met,
required test equipment was properly calibrated, the implementing
procedures were technically adequate to perform the test, crew
actions during the test were correct and timely, a quick analysis ,
and evaluation was made to assure proper plant response to the test, i
all test data were collected for analysis by the appropriate per- l
sonnel, overall test acceptance requirements were met and adherence ,
to TS requirements was met for any LCO's affected by the test. i
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FINDINGS
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Appendix 8023, " Reactor Trip and Shutdown from Outside the Con-
trol Room" - The inspector verified that the test satisfied R.G.
1.68 Appendix A, paragraph 5.d.d and FSAR requirements. Credit
was taken for a test performed during the pre-core hot func-
tional test program which demonstrated that the plant could be
placed from a hot standby to a hot shutdown condition. There-
fore, this portion of the test was not conducted.
1
[ -The inspector observed the pretest briefings of both the Test crew,
located at the Auxiliary Shutdown Panel, and the Normal crew located
in the Control Room. The briefing of the Test crew included a step
L by step review of the procedure, identification of the various con-
l trols and predicted plant indications and alerting the crew to actions
to be taken in the event unsafe conditions developed during the test.
The Normal crew was briefed on the purpose of test and the actions
they could and could not perform during the test.
It was verified that communications were established and maintained
during the test between the Auxiliary Shutdown Panel (ASP) and the
Control Room, Control Transfer Panels (orange and purple trains),
the Auxiliary Building and the Engineer Safeguards Features Building. l
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The reactor was manually tripped from 15% power with the CRDM breaker
located in the Auxiliary Building. Emergency Operating Procedure E0P
3503 " Shutdown Outside Control Room," Revision 2 was used to shutdown
the reactor and transfer control to the ASP. It was verified that
the test operators were able to maintain Tave at 557*F for over 1/2
hour from the ASP and then transfer control of the plant back to the
Control Room. No problems were identified with the test, test results l
or performance of the operating crew. A problem was identified with i
QC involvement with this test. See paragraph 4.0 "QA/QC Interface" l
for details. -
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Appendix 8006, " Secondary Plant Performance Test". This test is
performed at various power levels - 30%, 40%, 50%, 75%, 90% and
100% - in order to obtain base line data and identify any sec-
ondary plant problems. The inspector witnessed this test at the
30% power level. It was verified that the latest procedures
were being used, the test crew was knowledgeable of the program
and their responsibilities and the test data was being reviewed
and evaluated against expected results.
No problems were identified.
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3.3 Startup Test Results Evaluation
Scope
The test data results from the tests listed in Appendix A were review-
ed to verify that adequate testing had been accomplished. The results
were also reviewed to verify if regulatory guidance and licensee com-
mitments were satistied and to ascertain whether uniform criteria were
being applied in the evaluation of completed tests in order to assure
their technical and administrative adequacy.
Discussion
The inspector reviewed the test results and verified the licensee's
evaluation of test results by review of: test changes; test excep-
tions; test deficiencies, "as-run" copy of the test procedure; accep-
tance criteria; performance verification; recording of the conduct of
tests; QA inspection records; restoration of system to normal after
the test; independent verification of critical steps or parameters;
identification of personnel cenducting and evaluating test data; and
verification that the test results had been reviewed and approved by
licensee management.
FINDINGS
Turbine Overspeed Test ( Appendix 8016)
The capability of the mechanical overspeed trip device for the main
turbine generator was tested on February 15, 1986. Three (3) actual
trips occurred at 1962, 1963, 1963 rpm, respectively. This test was
accepted as satisfactory since the measured values were within test
procedure established criterion of s1998 rpm. The Backup Overspeed ,
Trip (BOST) Normal Mode is set approximately 0.5% above the mechanical 1
overspeed trip setting, to provide additional protection against tur-
bine overspeed during plant normal operation. The BOST electronic
circuit was also tested at 105% rated speed prior to performing me-
chanical overspeed trip test. The test result was also satisfactory.
Reactor Trip / Shutdown Outside Control Room (Appendix 8023)
The objectives of this test was to demonstrate the following:
a. Verification that the unit can be safely shutdown from outside
the control room using the Auxiliary Shutdown Panel,
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b. Verification that the unit can be maintained in a hot standby
condition from outside the control room.
c. Verification that the unit can be cooled down to the hot shutdown
condition from outside the control room.
Through test results review and test witnessing, the inspector noted
the licensee has demonstrated the remete shutdown capability and met
the above test objectives with the exception of item c.
The cold shutdown capab'lity (item c) was demonstrated earlier during
preoperational test, 3-INT-3000, Appendix 3014, " Remote Shutdown with
Cooldown" on November 1, 1985.
The remote shutdown capability test was accepted as satisfactory by
the JTG.
Automatic Reactor Control (Appendix 8017)
The performance of the automatic control system in raintaining reactor
coolant average temperature within programmed value was verified per
test procedure 3-INT-8000, Appendix S017 on February 17, 1986. The
inspector independently reviewed test data and noted that automatic
reactor control system responded well during the test. Tavg returned
to programmed value withia specified oscillation amplitude, and no
unstable response was observed. The corresponding system responses
such as pressurizer pressu. e, pressurizer level and steam generator
level responded well with no unusual behavior noticed.
During performance of this test, Step 5.5 required the feedwater pump l
speed control system to be set per 3-INT-3000, Appendix 3010. A test l
engineer misread Appendix 3010 to be Appendix 8010. Since Appendix
8010, Neutron Shield Tank Testing, has nothing to do with this test,
Step 5.5 was deleted during the test. The inspector informed the 11-
censee of this discrepancy. The test engineer re- eviewed the test
result and determined that this test change did not have adverse im-
pact on the test outcome, since Appendix 3010 had already been com-
pleted. The inspector concurred with the licensee's assessment and
had no further questions.
Power Coefficient (Appendix 8020)
The doppler-only power coefficient measurements are to be obtained at
30, 50, 75, 90 and 100 percent power. The measurement at 30% power
level was performed on February 17, 1986.
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Since it is difficult to directly measure the reactivity change due
to fuel temperature change during a power change, an indirect measure-
- ment technique was utilized. This technique involved measuring the
l primary side responses such as the change in core average temperature
(ATAVG) and the corresponding change in RCS loop parameters with re-
spect to a small change in turbine load. The doppler-only power co-
efficient was then calculated from this set of data and isothermal
i temperature coefficient information as previously derived from Zero
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Power Physics Test.
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All test results, including five sets of power swing data, met the
test acceptance criteria.
The inspector noted that an assumption of dTm=dff (Where T,is
moderator temperature and T f
is average fuel temperature) was used
in the licensee's measuring methodology for the doppler-only power l
coefficient derivation. The validity of this assumption was discussed '
with licensee reactor engineer and consultation provided by a NRR
Core Physics specialist. This assumption is valid when there is no
power change. From the test results and information provided by the l
fuel nndor (Westinghouse), it appeared that this assumption is also !
valid under test cordition with minimum power change.
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The inspector had no further questions. i
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hatural Circulation Test (Appendix 7006)
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The test witness and preliminary test result review was documented in
the Inspection Report 86-07. Although the licensee completed this ,
test and demonstrated that plant core heat can be removed satisfac- l'
totily by using natural circulation, the detailed test results had
not been thoroughly analyzed. From the test result review, the in-
spector noted that PORV lifted 3 times during the initial stage of
the natural circulation test. Plant behavior and lessons learned
from this test will be evaluated by the licensee and incorporated in
the licensee operator training program and possible procedure enhance- ,
ment. !
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The licensee management agreed that this detailed evaluation and !
incorporation into operator's training will be completed by May 1, !
1986. This is an unresolved item (50-423/86-09-01).
RCS Flow Coastdown (Appendix 5017)
The purpose of this test was to verify that the RCS responded as de-
signed to a partial and complete loss of forced reactor coolant flow.
With four reactor coolant pumps (PCPs) operating, partial loss of ,
forced reactor coolant flow was established by tripping one RCP. Com- l
plete loss of reactor coolant flow was established by simultaneously
tripping all four RCPs. (
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The inspector verified that low flow alarm times, control rod drop
times,-core flows and low flow alarm values met FSAR Chapter 15 Table
15.3-1 requirements. This was determined by reviewing the strip
chart recordings and the sequence of events computer printouts for
reactor coolant loop flow, RCP alarm, reactor trip times, and low
flow' alarms. It was also verified that the licensee had correctly
translated the information from the strip chart recorders and computer
print-out to the data sheets and, by performing independent calcula-
tions, that the final results were correct.
Post Core Hot Functional Tests UNSATS
A review of the disposition of all UNSATs foentified during the Post
Core Hot Functional Test Program was made. These UNSAT's, identified
in Appendix B, were reviewed for conformance with administrative re-
quirements and proper disposition of problems. No problems were
identified.
4.0 QA/QC INTERFACE
While observing the performance of 3-INT-8000 Appendix 8023, Reactor Trip /
Shutdown Outside the Control Room, the inspector questioned a QC inspector
also witnessing the test as to his involvement with the test. From these
discussions the following information was determined:
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Although the QC inspector stated that he was there to verify operator
compliance to E0P 3503, he had not yet reviewed E0P3503 or Appendix
8023 nor did he have a copy of the procedures with him.
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The QC inspector had not been briefed by his management on what to
look for or expect during the test. This was later verified by his
immediate supervisor.
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The QC inspector had not reviewed the FSAR or RG 1.68 requirements
pertaining to this test although he was aware he should do so.
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The QC inspector was not aware of what was taking place as evidenced
by the fact he did not know from what power level the reactor had
been tripped.
The NUSCO QA and NNECO QC departments share responsibility for performing
surveillances of startup tests. This responsibility is loosely coordin-
ated with NUSCO QA having primary responsibility. During this particular
test, representatives from both organizations were present. The auditor
from the QA section appeared knowledgeable of the test procedure and what
was taking place.
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The above occurrence of an inspector being assigned an inspection task
for which that inspector had not been trained or briefed constitutes a
violation (50-423/86-09-02).
At the exit meeting, the licensee QA representative agreed that:
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More specific guidance will be provided to QC inspectors who will
cover the remaining startup tests, as to which areas / criteria to
look into.
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NUSCO QA will continue to provide surveillances during the startup
test program.
The following NUSCO Operations QA surveillance reports were reviewed to
determine the adequacy of 00A's involvement with the Startup Test Program:
NUSCO 00A SURVEILLANCE NO. TITLE
TC 3950 Digital Rod Position Indication
TC 3960 Boron Endpoint Measurements
TC 3960A Boron Endpoint Measurements
TC 3961 Initial Criticality
TC 3961A Initial Criticality
TC 3968 Natural Circulation
TC 3972 Preparations for Power Ascension
Testing
TC 3976 Power Ascension Test
TC 3986 Integrated Plant Testing
TC 3986A Integrated Plant Testing
TC 39868 Integrated Plant Testing
The surveillance appeared to be thorough with good follow-up of identified
concerns. No problems were identified.
5.0 Independent Calculations
3-INT-5000 Appendix 5010, "RTO Bypass Loop Verification" obtains data
which is used to calculate the hot leg and cold leg flow and transport
times through the RTO Bypass Lines. Using the formulas:
F
H,= Ft Fc' = Ft-Fgg
(1+ Fc) ,
F l
H -
Fc' = Calculated cold leg flow
FH ,= Calculated hot leg flow
Fg= Measured Total flow
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Fc = Measured cold leg flow
FH = Measured hot leg flow
The inspector verified the flows and transport time for the hot and cold
leg RTO bypass lines.
Power coefficient measurement requires lengthy data reduction. The
inspector independently verified that the predicted reactor physics
parameters were correctly taken from the nuclear design reference. The
inspector also performed an independent calculation and confirmed that ;
the first power swing case data were being correctly reduced.
6.0 Plant Tours
The inspector made several tours of the facility durtag the course of the
inspection. This included tours of the control building and control room.
A review of the work in progress, security, cleanliness and housekeeping
was made.
7.0 Exit Meeting
An exit meeting was held on March 14, 1986 to discuss the inspection
scope and findings, as detailed in this report (see paragraph 1.0 for
attendees).
At no time was written material given to the licensee. The inspector
determined that no proprietary information was utilized during this
inspection.
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APPEN0IX A
TEST DATA REVIEWED
TEST NUMBER TITLE
3-INT-5000 Appendix 5001 Shutdown Margin
3-INT-5000 Appendix 5010 RTD Bypass Loop Verification
3-INT-5000 Appendix 5015 Oigital Rod Position Indication
Operation Test
3-INT-5000 Appendix 5016 Loose Parts Monitoring
3-INT-5000 Appendix 5017 RCS Flow Coastdown
3-INT-5000 Appendix 5031 Chemical and Volume Control System
3-INT-6000 Initial Criticality
3-INT-7000 Appendix 7006 Natural Circulation
3-INT-8000 Appendix 8023 Reactor Trip / Shutdown Outside
Control Room
3-INT-8000 Appendix 8016 Turbine Overspeed Test
3-INT-8000 Appendix 8017 Automatic Reactor Control
3-INT-8000 Appendix 8020 Power Coefficient
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APPENDIX B
POST CORE HOT FUNCTIONAL TEST UNSATS REVIEWED
TEST PROCEDURE UNSAT #'s
3-INT-5000, Anpendix 5001 7497
3-INT-5000, Appendix 5002 7471
3-INT-5000, Appendix 5004 7341, 7342
3-INT-5000, Appendix 5006 7495, 7492, 7493
3-INT-5000, Appendix 5007 7485, 7486, 7489, 7496
3-INT-5000, Appendix 5009 7466
3-INT-5000, Appendix 5015 7487
3-INT-5000, Appendix 5016 7475, 7479
3-INT-5000, Appendix 5017 7504, 7510
3-INT-5000, Appendix 5031 7472, 7473, 7474, 7476
7477, 7478, 7484, 7488
7490, 7491, 7499
3-INT-5000, Appendix 5033 7378, 7417, 7420
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