ML20212Q178
| ML20212Q178 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 08/13/1986 |
| From: | Eselgroth P, Florek D, Wink L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20212Q155 | List: |
| References | |
| 50-354-86-35, NUDOCS 8609040039 | |
| Download: ML20212Q178 (14) | |
See also: IR 05000354/1986035
Text
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-354/86-35
Docket No. 50-354
License No. NPF-50
Licensee:
Public Service Electric & Gas Company
Post Office Box 236
Hancocks Bridge, New Jersey 08038
Facility Name: Hope Creek Generating Station, Unit 1
Inspection At:
Hancocks Bridge, New Jersey
Inspection Conducted: July /-24, 1986
Inspectors:
7M. fa,tw+
.1,.,
__7!/S!f6
0. Florek, Lead Reacftor Engineer
d&te
'
S
Y 19 Th
%
L. Wink, R c or En
neer
' date
Approved by:
f//3 [fd
,
P..Eselgro W, Chief,
/ dite
Test Prog Nms Section, 08, DRS
Inspection Summary:
Inspection on July 7-24, 1986 (Inspection Report Number
50-354/86-35)
Areas Inspected:
Routine unannounced inspection of the overall power
ascension test program including procedure reviews, test witnessing and test
results evaluation, surveillance test activities, independent measurements and
evaluations, QA/QC interfaces and tours of the facility.
Results:
Two violations were identified. One involved a procedural
deficiency (paragraph 2.4) and the other involved failure to follow a startup
<
test procedure and failure to adequately review the completed test (2.5).
Note: For Acronyms not defined, refer to NUREG-0544 " Handbook of Acronyms
and Initialisms."
8609o40039 860s26
{DR
ADOCK 05000354
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DETAILS
1.0 Persons Contacted
Public Service Electric and Gas Company (PSE&G)
,
- o R. Beckwith, Station Licensing Engineer
G. Chew, Power Ascension Results Coordinator
- o G. Connor, Operations Manager
E. Dalton, Nuclear Systems Engineer
G. Daves, Senior Engineer, Operations
P. Dempsey, Shift Test Coordinator
M. Dick, Startup Test Engineer
- o M. Farshon, Power Ascension Manager
B. Forward, Power Ascension Administrative Coordinator
- o A. Giardino, Manager, Station QA
D. Hosmer, Lead Shift Test Coordinator
R. Hovey, Senior Nuclear Shift Supervisor
P. Krishna, Assistant to the General Manager
o S. La8runa, Assistant General Manager
- o M. Metcalf, Principal QA Engineer
L. Newman, Senior Nuclear Shift Supervisor
E. Rush, I&C Supervisor
.
- o R. Salvesen, General Manager, Hope Creek Operations
W. Schell, Power Ascension Technical Director
E. Skeehan, GE Operations Manager
W. Thomas, Shift Test Coordinator
L. Zull, Lead Startup Test, Design and Analysis Engineer
U.S. Nuclear Regulatory Commission (NRC)
o D. Allsopp, Resident Inspector
R. Borchardt, Senior Resident Inspector
o S. Collins, Chief, Reactor Projects Branch No. 2
!
o L. Norrholm, Chief, Reactor Projects Section, No. 2B
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- o R. Summers, Project Engineer
o Denotes those present at the interim exit meeting on July 17, 1986.
- Denotes those present at the exit meeting on July 24, 1986.
The inspector also contacted other members of the licensee's staff
including Senior Nuclear Shift Supervisors, Reactor Operators, Test
Engineers and other members of the Technical Staff.
2.0 Power Ascension Test Program (PATP)
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2.1 References
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Regulatory Guide 1.68, Revision 2, August 1978, " Initial Test
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Programs for Water-Cooled Nuclear Power Plants"
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ANSI N18.7-1976, " Administrative Controls and Quality Assurance
for the Operational Phrase of Nuclear Power Plants"
Hope Creek Generating Station (HCGS) Technical Specifications,
Revision 0, April 11, 1986
HCGS Final Safety Analysis Report (FSAR), Chapter 14 " Initial
Test Program"
HCGS Safety Evaluation Report (SER), Chapter 14, " Initial Test
Program"
Station Administrative Procedure, SA-AP.ZZ-036, Revision 3,
" Phase III Startup Test Program"
Specification NEB 0 23A4137, Revision 0, " Hope Creek Startup
Test Specification"
HCGS Power Ascension Test Matrix, Revision 7.
2.2 Overall Power Ascension Test Program
Scope
The inspector observed various activities associated with the conduct
of low power testing - these included daily management meetings,
operations shift turnovers, planning and scheduling meetings, test
results reviews by the Test Review Board (TRB) and the Station Opera-
tions Review Committee (SORC) and meeting to address specific prob-
lems identified during the performance of power ascension tests. The
inspector also observed the general conduct of operations and the
coordination of activities among the various departments.
Findings
No unacceptable conditions were identified.
2.3 Power Ascension Test Procedure Review
Scope
The Power Ascension Test Procedures listed in Attachment A were
reviewed for their conformance to the requirements and guidelines of
the references in paragraph 2.1 and for the attributes previously
defined in Inspection Report 50-354/86-03.
Findings
No unacceptable conditions were identified.
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2.4 Power Ascension Test Witnessino
Scope
The inspector witnessed portions of the power ascension tests
discussed below. The performance of these tests were witnessed to
verify that:
The current, approved revision of the test procedures was
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available and in use by all participants.
The minimum crew requirements, as defined in the approved
--
procedure and technical specifications, were being met.
Required prerequisites and initial conditions were
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established.
Test equipment was calibrated and operated in accordance with
--
procedure.
The procedure was technically adequate and appropriate to the
--
circumstances.
Crew performance was correct and timely.
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Coordination of test related activities was adequate.
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A preliminary analysis of all data collected was expeditiously
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performed following completion of the test.
The test was performed in accordance with all administrative
--
requirements.
Discussion
TE-SU.BF-052, Scram Testing'of Selected Rods. This test was
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performed on July 16, 1986 at a reactor pressure of 800 psig.
The inspector observed the scram tests performed on control rods 34-07 and 34-23.
The overall test crew performance was
satisfactory and coordination with ongoing surveillance testing
was well controlled by the shift operations personnel. The
inspector observed pre-scram data taking, initiation of testing
for each rod and data reduction following the testing of each
rod. All test results were well within acceptance criteria
Ifmits.
TE-SU.BD-141, Reactor Core Isolation Cooling System Condensate
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Storage Tank Injection.
This test was performed on July 9,
1986 at a reactor pressure of 150 psig.
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The inspector witnessed the performance of a hot quick start to
the Condensate Storage Tank.
Prior to the performance of this
portion of the test, the inspector observed an on-station test
briefing of the test crew by the test engineer.
This test
brief was in addition to the required pre-test brief.
This
additional briefing was very thorough and timely and
contributed to the subsequent smooth performance of the test.
During the performance of the quick start, the inspector
independently verified that the RCIC pump achieved a flow rate
greatar than 600 GPM within the required 30 seconds against an
appropriate discharge head.
The inspector also observed ap-
propriate overall system performance and response.
TE-SU.BD-142, RCIC Reactor Vessel Injection. This test was
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performed on July 23, 1986 at rated reactor pressure.
The inspector witnessed test preparations, test briefings and
portions of the actual performance of the test.
Test prepara-
tions were very thorough with operations personnel taking the
lead to assure that the test was effectively planned and imple-
mented.
The test briefings were acceptable, the planned testing
was described and precautions discussed.
Test personnel,
augmented by senior operations personnel, discussed several
potential problems that could occur during the test.
Several
potential problems discussed were based on inspector questions
during the test preparation phase regarding the possibility of a
RCIC pump. suction swap from the Condensate Storage Tank to the
Torus.
The inspector witnessed the licensee's initial attempt'to
operate RCIC at rated flow into the reactor vessel.
Because
reactor power was not quite high enough, reactor vessel level
continued to increase, even with no flow from the feedwater
pumps.
The licensee subsequently, raised power to 4.1% of rated
and was then able to perform the remainder of the test.
Having established stable conditions, with full RCIC flow into
the reactor vessel, the licensee prepared to perform step 5.3.1
of the procedure which required RCIC turbine speed to be
reduced to 2150 (+1C0, -0) RPM and the subsequent performance
of speed steps.
The inspector inquired as to the appropriate-
ness of this step, since this low speed would result in no flow
to the reactor vessel. After review by the licensee, it was
determined that the step was not appropriate and the procedure
was changed to require a specific pump flow rather than turbine
speed.
This procedure sequence could have resulted in equipment
damage and low reactor vessel water level if feedwater pump flow
was not manually increased.
10 CFR 50, Appendix B, Criterion V
states, in part, that " activities affecting quality shall be
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prescribed by documented procedures appropriate to the circum-
stances." Technical Specification 6.8.la through its reference
to Regulatory Guide 1.33 Revision 2, 1978 and ANSI N18.7. 1976
Section 5.3 contain similiar requirements. Contrary to these
requirements, step 5.3.1 of procedure TE-SU.BD-142 was not appro-
priate to the circumstances and is considered a violation
(354/85-35-01).
The licensee took prompt action to correct this
specific procedural problem at tFe time it was identified by the
inspector.
The inspector also witnessed the hot quick start portion of
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this test. RCIC was observed to achieve rated flow into the
reactor vessel within twenty seconds.
TE-SU.AE-236, Feedwater Startup Controller Test. This test
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was performed on July 7, 1986 at a reactor pressure of 145
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psig.
The inspector witnessed the test briefing and performance of
level step changes in both manual and automatic modes of con-
trol.
The inspector noted that while the controller was not
divergent (Level 1 acceptance criterion) in either the manual
or automatic modes it did exhibit limit cycle behavior in auto-
matic.
This limit cycle behavior made it essentially impossible
to achieve the stable level conditions required prior to the
performance of the level step changes in automatic and impos-
sible to distinguish the response of the system to these changes
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from the limit cycling. The inspector also noted that the Level
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2 acceptance criterion for these step changes was not applicable
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to the current test condition.
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Following the completion of the test and examination of the
recorded test data, the test engineers decided to abort the test
and attempt to reperform it at a later time. A retest was
performed on July 9,1986.
TE-SU.GT-721, Drywell and Steam Tunnel Cooling System - Normal
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operation performance.
This test yas performed on July 16, 1986
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at a reactor pressure of 800 psig.
The inspector observed the post test evaluation of recorded
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data and comparison to acceptance criteria. The licensee deter-
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mined that three Level 2 acceptance criteria were not satisfied
and issued results deficiency form No. 048 to document and track
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the resolution of these problems. The problems all involved
local temperatures in excess of the 150* F acceptance limit and
circumferential variations in temperature greater than the 10* F
acceptance limit. The Level 1 acceptance criterton for the
volumetric average drywell temperature of 135* F was satisfied.
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The inspector observed a licensee meeting, which was held in the
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control room, to discuss the test results, impact on future
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testing and possible resolution to the problem.
Present for
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this discussion were the Operations Manager, Power Ascension
Manager, Senior Nuclear Shift Supervisor, Shift Test Coordinator
and Test Engineer. The discussion was comprehensive and inte-
grated all available data, including the results of a special
drywell entry made to assess the situation. Short term plans
were formulated to ameliorate the problem, future tests were
discussed and plans were made for a formal engineering meeting
to assess the data, success of the short term plans and formu-
late any required long term solutions.
Findings
One violation was identified in the above discussion for pro-
cedure TE-SU.BD-142, which was inappropriate to the
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circumstances.
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2.5 Power Ascension Tests Results Evaluation
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The power ascension test results discussed be kw were evaluated for
the attributes identified in Inspection Report 50-354/86-24.
Otscussion
TE-SU.22-018, Process Radiation Monitoring Baseline Data. Test
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completed June 18, 1986, results review completed July 22, 1986,
pending management acceptance.
This test contains no acceptance criteria.
It gathers baseline
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data (prior to initial criticality) which will be used to es-
tablish setpoints for the Main Steam Line Radiation Monitors.
TE-SU.22-041, Full Core Shutdown Margin Demonstration.
Test
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completed June 29, 1986, results review completed July 10, 1986
and accepted by management on July 17, 1986.
All acceptance criteria were met.
Initial criticality was
achieved on RWM step 84, RSCS Group 3, Rod 46-27 at Notch 8
(2272 total notches) at a moderator temperature of 123' F.
The
inspector independently calculated the shutdown margin and veri-
fled the licensee's value of 2.4% delta K/K. The rod notch
limits to satisfy the 1% delta K/K reactivity anomaly criterion
at 123' F were 1428 to 2364 notches. Criticality was achieved
within these limits.
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During a detailed review of the test package the inspector
discovered a calculational error which had not been identified
during the results review.
Prerequisite step 2.15 requires the
test engineer to make a " Predicted Critical Calculation" in
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accordance with Attachment 2 of the procedure and to use the
results in Caution Statement 5.2.2.
Attachment 2 provides the
equations to be used to perform a calculation of the estimated
critical rod position and calculations of minimum and maximum
allowable rod withdrawal positions based on the 11% reactivity
anomaly limits.
These limits on rod withdrawal are then entered
in caution statement 5.2.2 to prevent test personnel from un-
knowingly exceeding these limits. Attachment 2 requires that
the completed calculations be independently reviewed.
The inspector noted that the calculation of the expected
critical position was correctly performed but that the
calculation of the minimum and mayimum Tod withdrawal limits
were not performed in accordance with the procedure.
The
procedure indicated that the calculations had been reviewed by
a senior test engineer.
The result of the error was that the
minimum allowable rod withdrawal limit was non-conservative.
The potential consequences of this error are of minor safety
sigt'ificance since the reactivity anomaly limits were being
independently monitored by the licensee's reactor engineer
personnel during the performance of this test and these er-
roneous calculations were not used to verify acceptance criteria
due to moderator temperature changes that occurred during the
test. However, the error was made and the licensee's detsiled
review of the test results failed to identify the deficiency.
10 CFR 50, Appendix B, Criterion XI states, in part,
that " testing shall be performed in accordance with written
test procedures," and further states that " test results shall
be evaluated to assure that the test requirements have been
satisfied." Contrary to these requirements the calculation of
minimum and maximum rod withdrawal limits were not performed in
accordance with the written procedure and the review of the test
results failed to identify this deficiency. This is considered
,
a violation (354/86-35-02).
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TE-SU.SE-062, Source Range Monitor Response to Rod Withdrawal.
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Test completed June 28, 1986, results review completed July 10,
1986 and accepted by management on July 14, 1986.
This test contains no acceptance criteria.
It records SRM
response to each step of the Rod Pull Sheet for reference during
future startups.
It also produces 1/M plots for the purpose of
anticipating criticality.
TE-SU.SE-064, Source Range Monitor Non-Saturation
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Demonstration.
Test completed June 29, 1986, results review
completed July 10, 1986 and accepted by management on July 14,
1986.
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This test contains no acceptance criteria.
It denonstrates
that the SRMs would not saturate below 150% of their normal
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scram setpoint (2 x 10 cps). All SRMs were demonstrated to be
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capable of recording a count rate of at least 5 x 10 cps.
TE-SU.SE-101, Source Range Monitor / Intermediate Range Monitor
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Overlap Verification. Test completed June 29, 1986, results
review completed July 10, 1986 and accepted by management on
July 14, 1986.
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During the initial performance of this test IRM "A"
failed to
respond to increasing neutron flux. Results Deficiency Forms
Nos. 38 and 39 were written to document the failure of IRM "A"
to satisfy the Level 1 and Level 2 acceptance criteria and to
allow continued operation and testing in accordance with
technical specifications.
IRM "A" was repaired and
successfully retest on July 1,1986 to close RDFs Nos. 38 and
39. All acceptance criteria were satisfied.
TE-SU.SE-102, IRM Range 6-7 Continuity Verification. Test
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completed June 30, 1986, results review complete July 22, 1986,
pending management acceptance.
This test contains no acceptance criteria.
It verifies IRM
continuity when switching from the low frequency amplifier
(Ranges 1-6) to the high frequency ampitfier (Ranges 7-10).
All IRM channel required a range correlation adjustment to
achieve the desired continuity.
IRM "A" was inoperable during
the initial performance of this test and was tested separately
on July 1, 1986.
TE-SU.BD-141, Reactor Core Isolation Cooling Condensate Storage
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Tank Injection. Test performed July 21, 1986.
This test was aborted due to a RCIC turbine trip. The turbine
tripped on overspeed due to controller hardwrce problems and
lack of a flowpath to the CST. This test also caused an over-
pressure condition to exist in the RCIC discharge piping (pres-
sure extrapolated from available test data to be 1900 psig for
less than five seconds).
To prevent a reoccurrence, a procedure
change was made to assure that the CST test return valve and the
RCIC steam admission valve are opened simultaneously rather than
in series. The inspector reviewed the safety evaluation per-
form 1d on the affected piping which indicated that the piping
was not damaged during the event.
TE-SU.AB-251, Main Steam Isolation Valve Functional Test.
Test
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completed July 21, 1986.
Licensee review of the test results is in progress.
The in-
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spector performed independent calculations of several MSIV
stroke times based on recorded data.
For those calculations
performed, the acceptance criteria of 3 to 5 seconds were
satisfied.
TE-SU.GT-721, Drywell and Steam Tunnel Cooling System - Normal
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Operation Performance. Test completed July 17, 1986.
Licensee review of the test results is in progress.
The
calculated drywell volumetric average temperature of 107" F
satisfied the Level 1 acceptance criterion.
Level 2 acceptance
criteria failures occurred for maximum local air temperature and
maximum temperature differentials and are being processed via
the administrative controls defined. The inspector questioned
the licensee on the minimum temperature limits for the under-
vessel area. The inspector noted that this area was being
cooled to 70* F.
The inspector inquired as to whether this
temperature was consistent with General Electric Design Require-
ments for the reactor vessel support skirt. The licensee re-
sponse was not formulated at the conclusion of this inspection.
Pending the response of the licensee, this item is considered
unresolved (354/86-35-03).
In addition to the test results evaluated above, the inspector
reviewed the status of 17 results deficiency forms (RDFs)
generated during low power testing.
Six RDFs were resolved but
not formally closed. The remaining RDFs still require
resolutions.
Findings
One violation was identified for failure to perform testing in
accordance with written procedures and for failure to
adequately review. One unresolved item was identified for
minimum temperature limits in the drywell undervessel area.
3.0 Surveillance Test Activities
Scope
The inspector witnessed portions of the operations surveillance tests
discussed below. The tests witnessed are closely related to testing
performed in the power ascension test program and were evaluated for the
attributes defined in paragraph 2.4.
Discussion
OP-ST.SN-001, ADS and Safety Relief Valve Manual Operability Test -
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18 months.
Test completed July 8, 1986.
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The inspector witnessed the test preparations, test briefing and
test performance of Section 5.1 of the test procedure to satisfy, in
part, the technical specification requirements for ADS valves
operability. All aspects of the testing were performed in a
thoroughly competent and professional manner.
Four of the five ADS
were satisfactorily tested and the inspector independently verified
various diverse indications of actual steam flow through the valves.
One ADS valve failed to respond to the manual actuation. Operations
personnel responded appropriately and in accordance with technical
specification requirements.
Immediate plans were made to isolate and
identify the cause of the failure.
Subsequent investigation identi-
fled the problem as a crimped Containment Instrument Gas Line to the
valve pilot. The line vas repaired and the valve was successfully
retested.
OP-ST.BJ-002, HPCI System Functional Test (Low Pressure) - 18
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months.
Test completed July 10, 1986.
The inspector witnessed the performance of Section 5.2 of the test
procedure to satisfy, in part, the technical specification require-
ments for HPCI response time. During initial attempts to perform the
quick start required for the test difficulties were encounter in
stroking the fell flow test return line to establish a flow path back
to the CST. A procedure change was made to allow positioning of the
valve prior to manually initiating HPCI,
Following this change a
quick start was successfully performed but the time to rated flow
exceeded the technical specification requirement of 27 seconds.
It
was noted that the discharge head was significantly higher than re-
quired (700 psig versus 300 psig) due to the position of the full
flow test return valve.
The valve was throttled open slightly and
the quick start was satisfactorily completed. All aspects of
operations personnel performance were judged to be satisfactory.
During the performance of this test the inspector noted that the
observed correlations between HPCI turbine speed, pump flow and
discharge pressure were not consistent with expected relationships.
The inspector obtained preliminary data from the performance of
Power Ascension Test TE-SU.BJ-151, HPCI Condensate Storage Tank
Injection, which was performed subsequent to the surveillance test
and confirmed these observations. The inspector questioned the
licensee on this anomalous correlation and the licensee agreed that
the data recorded was not consistent with the HPCI pump curves.
Subsequent investigation revealed a instrumentation problem with the
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turbine speed indication.
This problem should not affect the
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validity of the surveillance test but may result in invalidating all
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or part of the power ascension test. The inspector will review the
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licensee's determination of the effect of this problem during a
subsequent routine inspection.
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Findings
No unacceptable conditions were identified.
4.0 Independent Measurements and Verifications
The inspector performed multiple independent measurements and
verifications during the witnessing of power ascension (paragraph 2.4)
and surveillance (paragraph 3.0) testing and during the evaluation of
Power Ascension Test Results (paragraph 2.5).
Except as discussed in
paragraph 2.5, the inspector's measurements and verifications agreed with
those of the licensee.
5.0 QA/QC Interface with the Power Ascension Test Program
The inspector reviewed QA Surveillance Report No.86-187 covering the
performance of Power Ascension Test TE-SU.ZZ-041, Full Core Shutdown
Margin Demonstration. The inspector also observed QA involvement in the
review of power ascension test results.
No unacceptable conditions were
noted.
6.0 Tours of the Facility
In the course of witnessing power ascension and operations surveillance
test activities the inspector made tours of various areas of the facility
to observe work in progress, housekeeping and cleanliness controls. No
unacceptable conditions were identified.
7.0 Unresolved Items
Unresolved items are matters about which more information is required in
order to determine whether they are acceptable, an item of noncompliance
or a deviation. Unresolved items disclosed during the inspection are
discussed in paragraph 2.5.
8.0 Exit Interview
At the conclusion of the site inspection on July 24, 1986, an exit
meeting was conducted with the licensee's senior site representative
(denoted in Section 1.0).
In addition, an interim exit meeting was held
with the licensee on July 17, 1986.
At no time during the inspection was written material provided to the
licensee by the inspector. Based on the NRC Region I review of this
report and discussions held with licensee representatives during the
inspection, it was determined that this report does not contain
information subject to 10 CFR 2.790 restrictions.
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ATTACHMENT A
POWER ASCENSION TEST PROCEDURES REVIEWED
TE-SU.ZZ-013, Chemical and Radiochemical 15 to 25% Power Tests, Revision 1,
Approved May 16, 1986.
TE-SU.ZZ-017, Verification of Compliance for Discharge, Revision 0, Approved
June 10, 1986.
TE-SU.ZZ-018,
Process Radiation Monitoring Baseline Data, Revision 0,
Approved June ?7, 1986.
TE-SU.SE-103,
Intermediate Range Mor.itor/ Average Power Range Monitor Overlap
Verification, Revision 1, Approved April 28, 1986.
TE-SU.SE-112,
Local Power Range Monitor Calibration, Revision 1, Approved
April 30, 1986.
TE-SU.SE-122, APRM Calibration at Power, Revision 1, Approved
February 11, 1986.
TE-SU.ZZ-170,
Residual Heat Removal System Piping Expansion During Shutdown
Cooling Operation, Revision 3, Approved June 5, 1986.
TE-SU.ZZ-176, Balance of Plant Systems Piping Expansion During Power
Operations, Revision 1, Approved April 30, 1986.
TE-SU.BB-221, Pressure Regulator Test - Control Valves Controlling,
Revision 1, Approved March 26, 1986.
TE-SU.BB-223, Pressure Regulator Test - Bypass Valves Controlling, Revision 1,
Approved March 15, 1986.
TE-SU.AE-231,
Feedwater System Level Setpoint Steps, Revision 0, Approved
October 15, 1985.
TE-SU.AE-235, Manual Feedwater Flow Step Change Test, Revision 0, Approved
October 15, 1985.
TE-SU.AB-251, Main Steam Isolation Valve Functional Test, Revision 1,
Approved June 10, 1986.
TE-SU.ZZ-262, Relief Valve Response During Major Trips, Revision 1, Approved
June 25, 1986.
TE-SU.BB-303, Recirculation System Performance, Revision 0, Approved
October 28, 1985.
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TE-SU.AB-331, NSSS Main Stean Piping Steady State Vibration, Revision 2,
Approved May 20, 1986.
TE-SU.BB-332,
Recirculation System Piping Steady State Vibration, Revision 2,
Approved May 10, 1986.
TE-SU.AB-341,
NSSS Main Steam riping Dynamic Response, Revision 1, Approved
May 23, 1986.
TE-SU.BB-343,
BOP Main Steam Piping Dynamic Response, Revision 3, Approved
April 22, 1986.
TE-SU.AB-344, Main Steam Relief Valve Discharge Piping Dynamic Response,
Revision 2, Approved April 15, 1986.
TE-SU.GT-723, Drywell Cooling System Post Trip Performance Test, Ravision 1,
Approved June 19, 1986.
-TE-SU.GT-724,
Containment Penetration Cooling Normal Operation Performance
Test, Revision 1, Approved June 16,-1986.
TE- SU . HA-741, Gaseous Radwaste System Performance Test, Revision 0, Approved
October 9, 1985,
1E-SU.ZZ-761,
In-Plant Safety Relief Valve Shakedown Test, Revision 2,
Approved June 24, 1986.
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