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{{Adams | |||
| number = ML20198H771 | |||
| issue date = 01/10/1986 | |||
| title = Insp Rept 50-458/85-77 on 851031-1130.Violation Noted: Failure to Provide Clear Design Instructions on Approved Change to safety-related Design Mod | |||
| author name = Bennett W, Chamberlain D, Chanberlain D, Farrell R, Jaudan J, Jaudon J, Jones W, Mullikin R | |||
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) | |||
| addressee name = | |||
| addressee affiliation = | |||
| docket = 05000458 | |||
| license number = | |||
| contact person = | |||
| document report number = 50-458-85-77, NUDOCS 8601310093 | |||
| package number = ML20198H734 | |||
| document type = INSPECTION REPORT, NRC-GENERATED, INSPECTION REPORT, UTILITY, TEXT-INSPECTION & AUDIT & I&E CIRCULARS | |||
| page count = 13 | |||
}} | |||
See also: [[see also::IR 05000458/1985077]] | |||
=Text= | |||
{{#Wiki_filter:,- | |||
4 | |||
APPENDIX B | |||
U. S. NUCLEAR REGULATORY C0!DtISSION | |||
REGION IV | |||
NRC Inspection Report: 50-458/85-77 License: NPF-47 | |||
Docket: 50-458 | |||
Licensee: Gulf States Utilities Company (GSU) | |||
P. O. Box 2951 | |||
Beaumont, Texas 77704 | |||
Facility Name: River Bend Station (RBS) | |||
Inspection At: River Bend Station, St. Francisville, Louisiana | |||
Inspection Conducted: October 31 through November 30, 1985.. | |||
Inspectors: l , | |||
( ' | |||
A I2'/2- W | |||
D. D.VChamberlain, Senior Resident Inspector Date | |||
(pars. 1, 2, 3, 4, 5, 6, 7, 10, and 11) | |||
%&& w | |||
W. B. Jones,-Residdnt Inspector | |||
Q- 12'W | |||
Date | |||
(pars. 3,.4, 7, 8, 9, 10, and 11) | |||
, | |||
DA l N | |||
. ( Mu(liki , Reactor Inspector" ' Date | |||
(par. 3) | |||
, | |||
/ | |||
_ | |||
/d N | |||
R. Farfell, enior Resident InTp6ctor Date | |||
( ars. Q_and 10) | |||
8601310093 860127 | |||
PDR ADOCK 05000458 | |||
G PDR | |||
- . _ - .- - . . _ . ._ __ _. | |||
- | |||
I | |||
i | |||
2 | |||
k Ibh4 | |||
W. R. Bennett, Project Engineer Date | |||
(pars. 2 and 10) | |||
Approved: 62 //4 dM | |||
, ~J/. P Jauff on, Chief,' Project Section A, Date | |||
Me ob Projects Branch | |||
Inspection Summary | |||
Inspection Conducted October 31 through November 30, 1985 (Report 50-458/85-77) | |||
Areas Inspected: Routine, unannounced inspection of licensee action on previous | |||
ir.spection findings, status of facility operating license conditions, site tours, | |||
control and use of field change notices, licensee quality concern program | |||
review, initial criticality witness, pipe support and restraint systems testing, | |||
human engineering discrepancies (HEDs) review, startup test procedures review | |||
and startup test witness. The inspection involved 483 inspector-hours onsite by | |||
five NRC inspectors. | |||
Results: Within the areas inspected, one violation was issued in the area of | |||
control and use of field change notices (failure to provide clear and concise | |||
,' | |||
design instructions on an approved change to a safety-related design | |||
modification, paragraph 5), | |||
1 | |||
a | |||
I | |||
) | |||
L | |||
. , - - . _ . --. ~,,m- _ , _ . , . _ . . , , ,, r----- ._-v-,-- .- e- w -- . - , , -#m--- --,__.r- ,w - - - - - - - - - - - - | |||
3 | |||
DETAILS | |||
1. Persons Contacted | |||
Principal Licensee Employees | |||
*W. J. Cahill, Jr., Senior Vice President, River Bend Nuclear Group | |||
*E. M. Cargill, Superintendent, Radiological Programs | |||
*T. C. Crouse, Manager, Quality Assurance (QA) | |||
*J. C. Deddens, Vice President River Bend Nuclear Group | |||
*D. R. Derbonne, Supervisor, Startup and Test | |||
*Jan Evans, Stenographer | |||
S. Finnegan, Control Operating Foreman | |||
*P. E. Freehill, Superintendent, Startup and Test | |||
*A. D. Fredieu, Assistant Operations Supervisor | |||
- | |||
D. R. Gipson, Assistant Plant Manager, Operations | |||
*P. D. Graham, Assistant Plant Manager, Services | |||
*E. R. Grant, Supervisor, Nuclear Licensing | |||
D. Hartz, Shift Supervisor, Operations | |||
*R. W. Heinick, Director, Projects | |||
B. D. Hey, Licensing Engineer | |||
K. C. Hodges, Supervisor, Quality Systems | |||
R. Jacksor , Shift Supervisor, Operations | |||
D. Jernigan, Engineer, Startup and Test | |||
*G. R. Kimmall, Supervisor, Operations QA | |||
R. King, Engineer, Licensing | |||
A. D. Kowalczuk, Assistant Plant Manager, Maintenance | |||
D. J. Krueger, Supervisor, Engineering Administration | |||
T. Lacy, Shift Supervisor, Operations | |||
*H. M. McClellan, Senior Compliance Analyst | |||
*I. M. Malik, Supervisor Quality Engineering | |||
A. Middlebrooks, Shift Supervisor, Operations | |||
E. R. Oswood, Engineer, QA | |||
G. A. Patrissi, Engineer, QA | |||
*T. L. Plunkett, Plant Manager | |||
S. R. Radebaugh, Assistant Superintendent, Startup and Test | |||
*W. J. Reed, Director Nuclear Licensing | |||
*D. Reynerson, Director, Nuclear Plant Engineering | |||
*J. E. Spivey, Engineer, QA | |||
*R. B. Stafford, Director, Quality Services | |||
*P. F. Tomlinson, Director, Operation QA | |||
C. Warren, Shift Supervisor, Operations | |||
Stone and Webster | |||
F. W. Finger, III, Project Manager, PTO | |||
*B. R. Hall, Assistant Superintendent, Field Quality Control | |||
R. L. Spence, Superintendent, Field Quality Control | |||
4 a., .\ ~ 4 & -+-Au -2--- u - w , 4 . | |||
. | |||
4 | |||
The NRC inspectors also interviewed additional licensee, Stone and Webster | |||
(S&W), and other contractor personnel during the inspection period. | |||
* Denotes those persons that attended the exit interview conducted on | |||
December 12, 1985. The NRC resident inspector (RI), W. B. Jones, also | |||
attended the exit interview. | |||
2. Licensee Action on Previous Inspection Findings | |||
. | |||
a. (Closed) Open Item (8425-02): Wetting of neutron monitoring system | |||
cables below the reactor pressure vessel. | |||
The NRC inspector reviewed the licensee's records of megger testing | |||
and continuity testing of neutron monitoring system cables located | |||
below the reactor pressure vessel. All of the testing was witnessed | |||
by licensee quality assurance (QA) engineers. | |||
- | |||
This item is closed. | |||
'b. (Closed) Open Item (8558-04): Procedures identified for performance | |||
of surveillance testing of Emergency Core Cooling Systems (ECCS) were | |||
not approved for use. | |||
The NRC inspector reviewed the licensee's surveillance test | |||
procedures (STPs) for ECCS systems and determined that they were | |||
adequate to meet Technical Specification (TS) requirements. | |||
This item is' closed. | |||
i (Closed) Open Item (8558-05): Procedures that were similarly | |||
^ | |||
.c. | |||
performed contained inconsistencies and errors existed in procedures | |||
- | |||
, | |||
that had not been walked down. | |||
The NRC inspector reviewed the STPs for ECCS systems and determined | |||
that they were consistent and adequate to meet requirements. | |||
This item is closed. | |||
d. (Closed) Open Item (8558-06): Procedures contained "LATERS" and did | |||
not meet requirements of TS 4.6.1.2.a. | |||
The NRC inspector reviewed STP 000-0702 and determined that it meets | |||
the requirements of TS 4.6.1.2.a. | |||
This item is closed. | |||
. | |||
i | |||
_, ._ _ _ __ __ ___ - | |||
..- .- ,_ , _ _ _ _ _ _ _ _, _ . | |||
. . | |||
I | |||
< | |||
, | |||
. | |||
1 | |||
5 | |||
e. (Closed) Open Item'(8558-08): Procedures identified to meet the | |||
requirements of TS 4.6.4.2 were not approved for use. | |||
The NRC inspector reviewed the STPs for TS 4.6.4.2 and determined | |||
that they are adequate to meet TS requirements. | |||
" This item is closed. | |||
f. (Closed) Open Item (8558-26): Changes to TS have not been | |||
implemented into STPs. | |||
The NRC inspector reviewed the licensee's reference tracking system | |||
~for implementing changes to TS into procedures and determined that | |||
the program is adequate to ensure that procedures meet TS | |||
requirements. | |||
This item is closed. , | |||
g. =(Closed) Open Item (8558-28): Procedure STP 055-0101 (dated | |||
July 22, 1985) is not responsive to TS 4.9.12.4 in that it'does not | |||
verify access. interlocks and palm switches in certain locations are | |||
operable as required. | |||
The NRC inspector reviewed STP 055-0101 (dated September 23, 1985)- | |||
and determined that it is adequate'to meet.the. requirements of | |||
TS 4.9.12.4. | |||
. | |||
This item is closed. | |||
h. (Closed) Open Item (8527-04) (License Condition 3.a. Part 4): This | |||
item involved the completion of procedure A0P-0031, " Shutdown From | |||
Outside the Main Control Room," and the training of required | |||
personnel in the procedure. | |||
The NRC inspector reviewed a copy of approved procedure AOP-0031 and | |||
performed a walkdown of this procedure. The procedure appeared | |||
adequate to provide the necessary instructions to personnel to | |||
remotely shutdown the plant due to a fire in the main control room. | |||
All required actions were verified as being able to be performed in-a | |||
timely manner. In addition, the required training of operators has | |||
been performed. | |||
lhis open item and part 4 of license condition 3.a. are closed, | |||
i. (Closed) Open Item (8527-06) (License Condition 3.a. Part 3): This | |||
item involved the completion of all modifications, including diesel | |||
generator electrical circuits, to isolate the control room circuitry | |||
from the safe shutdown panel. | |||
T: | |||
> . | |||
. | |||
< | |||
, | |||
d | |||
6 | |||
, | |||
. | |||
~ | |||
The NRC inspector reviewed the results'of the following | |||
preoperational tests: (1) Special Situation Test Procedure 1-SST-50, | |||
' | |||
" Remote Shutdown," Revision 0 and (2) Special Situation Test | |||
' | |||
Procedure 1-SST-59, " Division I Remote Shutdown Tests," Revision 0. | |||
These tests were found to be complete and acceptable. | |||
This open item and part 3 of license condition 3.a. are closed. | |||
j. (Closed) Open Item (8527-08) (License Condition 3.a. Part 4): This | |||
. Item concerned the possible need for radio communications between | |||
operators during the implementation of A0P-0031. | |||
A review of the procedure and a walkdown by the NRC inspector. | |||
revealed no areas where remote communications were required. | |||
This open item and all parts of license condition 3.a. are closed. | |||
3. Status of Facility Operating License Conditions | |||
Facility Operating License NPF-40 for River Bend Station was issued on | |||
August 29, 1985, and pending Commission approval, operation is restricted | |||
to power levels not to exceed 5% of rated power. Attachment 1 to this | |||
license contains items to be completed to the satisfaction of NRC Region IV | |||
prior to achieving certain operational conditions. The following status | |||
is provided for the Attachment I license cond; Lions: | |||
a. (Closed) License Condition 3.a.: | |||
Complete the fire protection and prevention items prior to exceeding | |||
5% rated power. | |||
Parts 1 and 2 of this license condition were closed in NRC Inspection | |||
Report 50-458/85-69. Parts 3 and 4 of this license condition were | |||
closed in paragraph 2 of this report along with open items 8527-04, | |||
8527-06,and 8527-08. | |||
All four parts of license condition 3.a. are closed. | |||
b. (Closed) License Condition 3.b.: | |||
Complete testing of liquid, gaseous, and solid radwaste systems and | |||
place these systems in service prior to exceeding 5% rated power. | |||
The liquid radwaste system was addressed in paragraph 3 of NRC | |||
Inspection Report 50-458/85-70. The gaseous radwaste system was | |||
l | |||
addressed in paragraph 3 of NRC Inspection Report 50-458/85-69. | |||
. | |||
._ _ . _ . _ _ ._ . . | |||
' | |||
4 | |||
7 | |||
- | |||
The solid radwaste system was addressed in paragraph 2 of NRC | |||
Inspection Report 50-458/85-53. All of these systems have been tested | |||
and are operational. | |||
This license condition is closed. | |||
c. (Closed) License Condition 3.d.: | |||
Complete installation and testing of post-accident sampling system | |||
(PASS) and place system in service prior to exceeding 5% rated power. | |||
On November'10, 1985, the RI witnessed a demonstration of the | |||
licensee's ability to provide representative liquid samples from | |||
within the primary containment using the manual grab post-accident | |||
sampling system. The demonstration consisted of taking two liquid | |||
samples from the B jet pump header using the liquid sample station | |||
located in the auxiliary building. The licensee was able to draw two | |||
samples and analyze the first sample for pH, dissolved oxygen, | |||
hydrogen, boron, conductivity and radionuclides within the required 3 | |||
hours from the time the decision was made to take the samples. The | |||
RI verified that approved procedures.were used for the PASS demonstra- | |||
tion and sample analysis.' In addition, the RI verified that work on | |||
PASS has been completed and the system turned over to operations. | |||
This license condition is closed. | |||
d. ,(Closed) License Condition 3.e.: | |||
Load reduction modifications to reduce maximum emergency service load | |||
to 2884 kw for the Division I diesel generator and to 2780 kw for the | |||
Division II diesel generator prior to exceeding 5% of rated power. | |||
The licensee initiated Modification Request (MR) 850291 to reduce the | |||
loads on the Division I and II emergency diesel generators. As a | |||
result of this request, S&W engineering company initiated Engineering | |||
and Design Coordination Requests (E&DCR) S-10017 and S-10018. E&DCR | |||
S-10017 adjusted the fan blade settings on Division I and II diesel | |||
generator exhaust fans 1HVP*FN2A and B, respectively. E&DCR S-10018 | |||
reduced the heating capacity of the standby cooling tower remote air | |||
intake room duct heaters 1HVY*CH6A and B from 30 kw to 12 kw egh. | |||
These modifications were performed under Maintenance Work Requests | |||
(MWR) 003497, 008678, and 14680. MWRs 003497 and 008678 adjusted the | |||
fan blade pitches from 37 to 20 for Division I and II diesel | |||
generators respectively, while MWR 14680 disconnected 9 of the 15 | |||
heater elements from HVY*CH6A and 68. These modifications resulted | |||
in final diesel generator loading of 2882.84 kw for Division I and | |||
2766.01 kw for Division II. The licensee has also submitted a Final | |||
Safety Analysis Report (FSAR) change notice to the NRC reflecting the | |||
above modifications. | |||
This license condition is clor,ed. | |||
I | |||
L | |||
. | |||
" | |||
~ | |||
* | |||
. | |||
8 | |||
l | |||
' | |||
4. Site Tours | |||
, | |||
The SRI and RI toured areas of the site during the inspection period to | |||
gain knowledge of the plant and to observe general job practices. | |||
1 | |||
l No violations or deviations were identified in this area of inspection. | |||
5. Control and Use of Field Change Notices (FCNS) | |||
During a routine followup from a previous NRC inspection, the SRI reviewed | |||
the completed Field Change Notice (FCN) 19 to MR 85-0397. FCN-19 had | |||
. | |||
become a 96-page document and the SRI found that the design instructions | |||
l | |||
' | |||
were often misleading and difficult to follow. For example, page 36 of | |||
FCN-19 added terminal board 43B contacts (contacts 7/8 and 3/4) on an | |||
electrical schematic for remote shutdown operation of the diesel generator | |||
; output breaker, and page 57 of FCN-19 removed the same contacts. Also, | |||
l several pages of FCN-19 were found to be illegible due to poor reproduction | |||
quality. The failure to provide clear and concise design instructions on | |||
. an approved change to a safety-related design modification was identified | |||
by the SRI as an apparent violation (8577-01). The SRI immediately | |||
informed the licensee of this finding and the licensee took immediate | |||
action to review other safety-related design modifications for similar | |||
problems. Although similar documentation problems were found, hardware | |||
inspections revealed no problems with the installed hardware. | |||
6. Licensee Quality Concern Program Review | |||
The licensee's quality concern prouram was described in NRC Inspection | |||
Report 50-458/84-25. The SRI continued to review licensee actions on | |||
' | |||
individual cases and completed the review of the current status of all | |||
open quality concerns, | |||
t | |||
; No violations or deviations were identified in this area of inspection. | |||
: | |||
7. Initial Criticality Witness | |||
i | |||
l This area of inspection was conducted to review licensee preparations for | |||
initial critical and to witness initial criticality in order to ascertain l | |||
l confomance to license and procedural requirements and to observe | |||
l operating staff performance. The licensee completed preparations for | |||
' | |||
initial criticality on Octooer 30, 1985, and began pulling the first | |||
control rod at about 9:30 p.m. The reactor was declared critical at | |||
12:38 a.m. on October 31, 1985, on step 27 of the rod withdrawal sequence | |||
with 2032 notches withdrawn. Nuclear instrumentation was monitored closely | |||
during approach to criticality and appeared to respond as expected. | |||
Also, during this inspection period, the RI completed the review of the | |||
results of 13 STPs performed prior to achieving initial criticality. The | |||
RI verified that the STP had been properly signed by the personnel | |||
performing the test, that the data was collected and recorded as required, | |||
. | |||
O | |||
that all deficiencies had been addressed and corrective action completed, | |||
and that the surveillance tests had been performed within the required | |||
frequency. | |||
The. SRI concluded that the licensee preparations for initial criticality | |||
were cautious and the operating staff performance was controlled and | |||
cautious. License and procedural requirements appeared to have been | |||
implemented for a safe and cautious initial criticality at River Bend. | |||
No violations or deviations were identified in this area of inspection. | |||
8. Pipe Support and Restraint S_vstems Testing | |||
During the initial nuclear heat up cycle, the R1 visually examined system | |||
pipe supports to ascertain whether system thermal expansion vrould result | |||
in any apparent failures to the dynamic pipe supports, fixed pipe supports, | |||
or their associated component support structures. These visual examirationr. | |||
were performed at ambient plant temperature and again at a plant | |||
temperature of 350 F. Pipe supports and structures were examined to | |||
verify where applicable, that; (a) deterioration, corrosion, physical | |||
damage or deformation were not noticeable; (b) bolts, nuts, washers, were | |||
tight and secure; (c) the equipment was not in a " lock up" or " frozen" | |||
position; and (d) pipes, supports or other associated equipment or | |||
compcnents were not in contact or cause rubbing due to thermal expansion. | |||
The following pipe supports and associated support structures were examined: | |||
Type System Identification No. | |||
Spring Residual Heat Removal RHS-PSSH-30775 | |||
Spring Recirculation Pump Housing H-306A | |||
Spring det Pump Header H354A | |||
Spring Low Pressure Core Spray CSL-PSSH-3001 | |||
Spring Low Pressure Coolant Injection RHS-PSSH-3089 | |||
Snubber Feed Water FWS-PSSP-3021 | |||
Snubber Residual Heat Removal RHS-PSSP-3078 | |||
Snubber det Pump Header S-354A | |||
Snubber det Pump Header S-3598 | |||
Snubber det Pump Header S-304B | |||
Snubber Residual Heat Removal Rl15-PSSP-3011 | |||
Snubber Recirculation Pump Motor S-371B | |||
Snubber Recirculation Pump Housing S-306A | |||
Snubber Feedwater FWS-PSSP-3015 | |||
Hanger Feedwater FWPRR-804 | |||
11 anger Feedwater FWS-PRR-823 | |||
Hanger Feedwater FWS-PRR-808 | |||
Hanger Residual Heat Removal RHS-PRR-831 | |||
Bracket Main Steam MSS-814 | |||
Bracket Main Steam MSS-822 | |||
Bracket Main Steam MSS-823 | |||
. _ _ . _ . . _ - _ . _ _ . . . _ - _ .._-.__.__ ___ _ _ _ | |||
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- | |||
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4 | |||
> | |||
10 | |||
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i .. | |||
! The RI did not identify any conditions which would appear to adversely | |||
affect the operation of the examined system pipe supports. The system | |||
expansion data for Startup Test Procedure (ST) 1-ST-17, " System | |||
Expansion," will be reviewed to ensure that expansion problems do not - | |||
exist at operating temperature and pressure for the above pipe supports i | |||
i and associated structures. The results of this review will be documented | |||
in a subsequent inspection report. | |||
No violations or deviations were identified in this inspection area. | |||
9. Human Engineering Discapancies (HEDs) Review | |||
During this inspection period the NRC RI reviewed Human Engineering ? | |||
Discrepancy (HED) records to verify that identified improvements in the | |||
control room / operator interface, for selected HEDs, had been completed. | |||
- ,. The HEDs were documented in the River Bend Station Detailed Control Room | |||
l | |||
Design Review Summary Report, issued in October 1984. | |||
The following HEDs were verified complete: | |||
. | |||
847 Label "SRV Low-Low Set / Reset" for A and B train. | |||
4 | |||
833 Label recorders 1CliS-PR2A and 2B to indicate containment | |||
pressure. | |||
i | |||
8 Mimic of RHR system. l | |||
! 25 Establish lines of demarcation for RHR Loop B and C. | |||
! | |||
. 830 Label switch for manual initiation of containment as | |||
l " Isolation." | |||
: | |||
248 Label as ADS safety relief valve. | |||
, | |||
747 Add escutcheons for manual steam isolation manual initiation | |||
! push bottoms. | |||
] 431 Modify total feedwater flow recorder to read in % flow. | |||
l 328 Annunciator windows consistent in type and style. | |||
l 422 Identify operational limits or warnings on meters. | |||
I | |||
q 575 Accurately identify valves (up to root valves) by tagging. | |||
307 Mcdify "HPCS NOT READY FOR AUTO START" annunciator circuit to | |||
actuate for any of the seven conditions which will render HPCS | |||
incapable of auto start. | |||
: | |||
: | |||
o | |||
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! | |||
. - - - - . - _ _ _ . _ _ . - - - - _ _ . - | |||
' | |||
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11 | |||
No violations or deviations were identified in this area of inspection. | |||
10. Startup Test Procedures Review | |||
The purpose of this area of the inspection was to review selected ST | |||
procedures for compliance with regulatory requirements, FSAR commitments | |||
and Technical Specifications. The NRC inspectors reviewed ST Procedure | |||
1-ST-19. " Core Performance," Revision 0, ST Procedure 1-ST-26, " Safety | |||
Relief Valves," Revision 0, and a draft major change request to 1-ST-26, | |||
and ST Procedure 1-ST-28, " Shutdown From Outside the Control Room," | |||
Revision 0. | |||
No violations or deviations were identified in this area of inspection. | |||
11. Startup Test Witness | |||
During this inspection period, the NRC inspectors witnessed startup testing | |||
and operational activities conducted under the low power testing program. | |||
Testing and operational activities witnessed included a reactor startup on | |||
November 17,~1985, a reactor core isolation cooling (RCIC) turbine run on | |||
November 17, 1985, and initial turbine roll on November 26, 1985. | |||
River Bend received a full power operating license (NPF-47) on November | |||
20, 1985, and reactor power was raised above 5% on November 25, 1985. | |||
No violations or deviations were identified in this area of inspection. | |||
12. Apparent Power Level Exceeding Licensed Level of 5% | |||
On November 18, 1983, the apparent power level exceeded the licensed | |||
level of 5%. The SRI, assisted by a Region IV supervisor, conducted an | |||
inspection to determine the facts of this occurrence, | |||
a. Background | |||
On November 15, 1985, the NRC commissioners voted to allow the NRC | |||
staff to issue a full power license to the River Bend facility. The | |||
full power license was subsequently issued on November 20, 1985. | |||
b. _0_ccurrence | |||
During the night shift of November 17-18, 1985, the on-duty shift | |||
supervisor directed power be raised to approximately 6.5% to | |||
facilitate turbine steam chest warm up. Although other on-watch | |||
personnel questioned this order, it was carried out. The shift | |||
supervisor on watch stated to NRC personnel that he understood th6t | |||
the Commission vote authorized operation above 5% power. The | |||
indicated power of approximately 6.5% power was maintained for about | |||
. . _ - - - . -. -- . .. -. . _ . - | |||
_ | |||
c. | |||
12 | |||
2 hours. The event was tcrminated when the oncoming shift supervisor | |||
* | |||
questioned why power had been raised above 5%. | |||
c. Determination of Actual Power | |||
The NRC inspectors reviewed the licensee's two methods of | |||
quantifying actual power during the event. The first method by a | |||
secondary calorimetric calculation. This calculation indicated that | |||
actual power was about 4.5%; however, the accuracy of this | |||
calculation was not considered good because feed flow was very low. | |||
The measurement of feed flow was not considered to be accurate at | |||
the very low end of the indicating range of the measuring device. | |||
The second calculation was based on the valve position of the | |||
turbine by-pass valves. There are two turbine by-pass valves at | |||
River Bend. Each of these valves is rated at 5% of total power. | |||
One of the valves was shut and the other valve was 50% open. Data | |||
from General Electric taken on identical valves at other plants | |||
indicated that these bypass have linear response or flow | |||
characteristics (e.g., a valve that is 50% open would pass one-half | |||
of its total capacity). The licensee' calculated that the one bypass | |||
valve which was 50% open accounted for 2.5% of reactor power. ! | |||
Allowing for other heat losses such as steam traps and the reactor | |||
water clean up system,' the licensee calculated that actual power was | |||
less than 3.5% | |||
d. Inspection Findings | |||
' | |||
The NRC inspectors noted that nuclear instruments could not be set | |||
accurately until power was at a higher level so that an accurate | |||
heat balance'could be made. Nuclear instrument gains were set high, | |||
which is the conservative approach. Indicated power was thus | |||
greater than actual power, and actual power had probably not | |||
exceeded the licensed limit. The NRC inspectors concluded that | |||
there had been a serious breakdown in training and communications. | |||
As the result of interviews with operators and supervisors, it | |||
appeared that this communication breakdown was limited to the one | |||
shift supervisor who had directed the power increase. The NRC | |||
inspectors noted that the licensee had removed this shift supervisor | |||
from licensed duties for retraining and evaluation. It was also | |||
found that the licensee had taken measures to improve communications. | |||
These included the presence of an operations management individual | |||
at all shift turnover briefings, written instructions to operators on | |||
power limit and the resetting of rod blocks to 5%. | |||
It was concluded that continued close monitoring of shift performance | |||
communications was warranted to assess the effectiveness of the | |||
corrective actions taken. | |||
-- | |||
= - | |||
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13. Exit Interview. , | |||
, | |||
j An exit interview was conducted on December.12,.l'985, with licensee | |||
' | |||
representatives (identified.in paragraph 1). During this interview, the | |||
. SRI-reviewed the scope and findings of.the' inspection. | |||
, | |||
f | |||
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}} |
Latest revision as of 08:32, 15 December 2020
ML20198H771 | |
Person / Time | |
---|---|
Site: | River Bend |
Issue date: | 01/10/1986 |
From: | Bennett W, Chamberlain D, Chanberlain D, Farrell R, Jaudan J, Jaudon J, William Jones, Mullikin R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
To: | |
Shared Package | |
ML20198H734 | List: |
References | |
50-458-85-77, NUDOCS 8601310093 | |
Download: ML20198H771 (13) | |
See also: IR 05000458/1985077
Text
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APPENDIX B
U. S. NUCLEAR REGULATORY C0!DtISSION
REGION IV
NRC Inspection Report: 50-458/85-77 License: NPF-47
Docket: 50-458
Licensee: Gulf States Utilities Company (GSU)
P. O. Box 2951
Beaumont, Texas 77704
Facility Name: River Bend Station (RBS)
Inspection At: River Bend Station, St. Francisville, Louisiana
Inspection Conducted: October 31 through November 30, 1985..
Inspectors: l ,
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A I2'/2- W
D. D.VChamberlain, Senior Resident Inspector Date
(pars. 1, 2, 3, 4, 5, 6, 7, 10, and 11)
%&& w
W. B. Jones,-Residdnt Inspector
Q- 12'W
Date
(pars. 3,.4, 7, 8, 9, 10, and 11)
,
DA l N
. ( Mu(liki , Reactor Inspector" ' Date
(par. 3)
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R. Farfell, enior Resident InTp6ctor Date
( ars. Q_and 10)
8601310093 860127
PDR ADOCK 05000458
G PDR
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W. R. Bennett, Project Engineer Date
(pars. 2 and 10)
Approved: 62 //4 dM
, ~J/. P Jauff on, Chief,' Project Section A, Date
Me ob Projects Branch
Inspection Summary
Inspection Conducted October 31 through November 30, 1985 (Report 50-458/85-77)
Areas Inspected: Routine, unannounced inspection of licensee action on previous
ir.spection findings, status of facility operating license conditions, site tours,
control and use of field change notices, licensee quality concern program
review, initial criticality witness, pipe support and restraint systems testing,
human engineering discrepancies (HEDs) review, startup test procedures review
and startup test witness. The inspection involved 483 inspector-hours onsite by
five NRC inspectors.
Results: Within the areas inspected, one violation was issued in the area of
control and use of field change notices (failure to provide clear and concise
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design instructions on an approved change to a safety-related design
modification, paragraph 5),
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DETAILS
1. Persons Contacted
Principal Licensee Employees
- W. J. Cahill, Jr., Senior Vice President, River Bend Nuclear Group
- E. M. Cargill, Superintendent, Radiological Programs
- T. C. Crouse, Manager, Quality Assurance (QA)
- J. C. Deddens, Vice President River Bend Nuclear Group
- D. R. Derbonne, Supervisor, Startup and Test
- Jan Evans, Stenographer
S. Finnegan, Control Operating Foreman
- P. E. Freehill, Superintendent, Startup and Test
- A. D. Fredieu, Assistant Operations Supervisor
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D. R. Gipson, Assistant Plant Manager, Operations
- P. D. Graham, Assistant Plant Manager, Services
- E. R. Grant, Supervisor, Nuclear Licensing
D. Hartz, Shift Supervisor, Operations
- R. W. Heinick, Director, Projects
B. D. Hey, Licensing Engineer
K. C. Hodges, Supervisor, Quality Systems
R. Jacksor , Shift Supervisor, Operations
D. Jernigan, Engineer, Startup and Test
- G. R. Kimmall, Supervisor, Operations QA
R. King, Engineer, Licensing
A. D. Kowalczuk, Assistant Plant Manager, Maintenance
D. J. Krueger, Supervisor, Engineering Administration
T. Lacy, Shift Supervisor, Operations
- H. M. McClellan, Senior Compliance Analyst
- I. M. Malik, Supervisor Quality Engineering
A. Middlebrooks, Shift Supervisor, Operations
E. R. Oswood, Engineer, QA
G. A. Patrissi, Engineer, QA
- T. L. Plunkett, Plant Manager
S. R. Radebaugh, Assistant Superintendent, Startup and Test
- W. J. Reed, Director Nuclear Licensing
- D. Reynerson, Director, Nuclear Plant Engineering
- J. E. Spivey, Engineer, QA
- R. B. Stafford, Director, Quality Services
- P. F. Tomlinson, Director, Operation QA
C. Warren, Shift Supervisor, Operations
Stone and Webster
F. W. Finger, III, Project Manager, PTO
- B. R. Hall, Assistant Superintendent, Field Quality Control
R. L. Spence, Superintendent, Field Quality Control
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The NRC inspectors also interviewed additional licensee, Stone and Webster
(S&W), and other contractor personnel during the inspection period.
- Denotes those persons that attended the exit interview conducted on
December 12, 1985. The NRC resident inspector (RI), W. B. Jones, also
attended the exit interview.
2. Licensee Action on Previous Inspection Findings
.
a. (Closed) Open Item (8425-02): Wetting of neutron monitoring system
cables below the reactor pressure vessel.
The NRC inspector reviewed the licensee's records of megger testing
and continuity testing of neutron monitoring system cables located
below the reactor pressure vessel. All of the testing was witnessed
by licensee quality assurance (QA) engineers.
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This item is closed.
'b. (Closed) Open Item (8558-04): Procedures identified for performance
of surveillance testing of Emergency Core Cooling Systems (ECCS) were
not approved for use.
The NRC inspector reviewed the licensee's surveillance test
procedures (STPs) for ECCS systems and determined that they were
adequate to meet Technical Specification (TS) requirements.
This item is' closed.
i (Closed) Open Item (8558-05): Procedures that were similarly
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performed contained inconsistencies and errors existed in procedures
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that had not been walked down.
The NRC inspector reviewed the STPs for ECCS systems and determined
that they were consistent and adequate to meet requirements.
This item is closed.
d. (Closed) Open Item (8558-06): Procedures contained "LATERS" and did
not meet requirements of TS 4.6.1.2.a.
The NRC inspector reviewed STP 000-0702 and determined that it meets
the requirements of TS 4.6.1.2.a.
This item is closed.
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e. (Closed) Open Item'(8558-08): Procedures identified to meet the
requirements of TS 4.6.4.2 were not approved for use.
The NRC inspector reviewed the STPs for TS 4.6.4.2 and determined
that they are adequate to meet TS requirements.
" This item is closed.
f. (Closed) Open Item (8558-26): Changes to TS have not been
implemented into STPs.
The NRC inspector reviewed the licensee's reference tracking system
~for implementing changes to TS into procedures and determined that
the program is adequate to ensure that procedures meet TS
requirements.
This item is closed. ,
g. =(Closed) Open Item (8558-28): Procedure STP 055-0101 (dated
July 22, 1985) is not responsive to TS 4.9.12.4 in that it'does not
verify access. interlocks and palm switches in certain locations are
operable as required.
The NRC inspector reviewed STP 055-0101 (dated September 23, 1985)-
and determined that it is adequate'to meet.the. requirements of
.
This item is closed.
h. (Closed) Open Item (8527-04) (License Condition 3.a. Part 4): This
item involved the completion of procedure A0P-0031, " Shutdown From
Outside the Main Control Room," and the training of required
personnel in the procedure.
The NRC inspector reviewed a copy of approved procedure AOP-0031 and
performed a walkdown of this procedure. The procedure appeared
adequate to provide the necessary instructions to personnel to
remotely shutdown the plant due to a fire in the main control room.
All required actions were verified as being able to be performed in-a
timely manner. In addition, the required training of operators has
been performed.
lhis open item and part 4 of license condition 3.a. are closed,
i. (Closed) Open Item (8527-06) (License Condition 3.a. Part 3): This
item involved the completion of all modifications, including diesel
generator electrical circuits, to isolate the control room circuitry
from the safe shutdown panel.
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The NRC inspector reviewed the results'of the following
preoperational tests: (1) Special Situation Test Procedure 1-SST-50,
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" Remote Shutdown," Revision 0 and (2) Special Situation Test
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Procedure 1-SST-59, " Division I Remote Shutdown Tests," Revision 0.
These tests were found to be complete and acceptable.
This open item and part 3 of license condition 3.a. are closed.
j. (Closed) Open Item (8527-08) (License Condition 3.a. Part 4): This
. Item concerned the possible need for radio communications between
operators during the implementation of A0P-0031.
A review of the procedure and a walkdown by the NRC inspector.
revealed no areas where remote communications were required.
This open item and all parts of license condition 3.a. are closed.
3. Status of Facility Operating License Conditions
Facility Operating License NPF-40 for River Bend Station was issued on
August 29, 1985, and pending Commission approval, operation is restricted
to power levels not to exceed 5% of rated power. Attachment 1 to this
license contains items to be completed to the satisfaction of NRC Region IV
prior to achieving certain operational conditions. The following status
is provided for the Attachment I license cond; Lions:
a. (Closed) License Condition 3.a.:
Complete the fire protection and prevention items prior to exceeding
5% rated power.
Parts 1 and 2 of this license condition were closed in NRC Inspection
Report 50-458/85-69. Parts 3 and 4 of this license condition were
closed in paragraph 2 of this report along with open items 8527-04,
8527-06,and 8527-08.
All four parts of license condition 3.a. are closed.
b. (Closed) License Condition 3.b.:
Complete testing of liquid, gaseous, and solid radwaste systems and
place these systems in service prior to exceeding 5% rated power.
The liquid radwaste system was addressed in paragraph 3 of NRC
Inspection Report 50-458/85-70. The gaseous radwaste system was
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addressed in paragraph 3 of NRC Inspection Report 50-458/85-69.
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The solid radwaste system was addressed in paragraph 2 of NRC
Inspection Report 50-458/85-53. All of these systems have been tested
and are operational.
This license condition is closed.
c. (Closed) License Condition 3.d.:
Complete installation and testing of post-accident sampling system
(PASS) and place system in service prior to exceeding 5% rated power.
On November'10, 1985, the RI witnessed a demonstration of the
licensee's ability to provide representative liquid samples from
within the primary containment using the manual grab post-accident
sampling system. The demonstration consisted of taking two liquid
samples from the B jet pump header using the liquid sample station
located in the auxiliary building. The licensee was able to draw two
samples and analyze the first sample for pH, dissolved oxygen,
hydrogen, boron, conductivity and radionuclides within the required 3
hours from the time the decision was made to take the samples. The
RI verified that approved procedures.were used for the PASS demonstra-
tion and sample analysis.' In addition, the RI verified that work on
PASS has been completed and the system turned over to operations.
This license condition is closed.
d. ,(Closed) License Condition 3.e.:
Load reduction modifications to reduce maximum emergency service load
to 2884 kw for the Division I diesel generator and to 2780 kw for the
Division II diesel generator prior to exceeding 5% of rated power.
The licensee initiated Modification Request (MR) 850291 to reduce the
loads on the Division I and II emergency diesel generators. As a
result of this request, S&W engineering company initiated Engineering
and Design Coordination Requests (E&DCR) S-10017 and S-10018. E&DCR
S-10017 adjusted the fan blade settings on Division I and II diesel
generator exhaust fans 1HVP*FN2A and B, respectively. E&DCR S-10018
reduced the heating capacity of the standby cooling tower remote air
intake room duct heaters 1HVY*CH6A and B from 30 kw to 12 kw egh.
These modifications were performed under Maintenance Work Requests
(MWR) 003497, 008678, and 14680. MWRs 003497 and 008678 adjusted the
fan blade pitches from 37 to 20 for Division I and II diesel
generators respectively, while MWR 14680 disconnected 9 of the 15
heater elements from HVY*CH6A and 68. These modifications resulted
in final diesel generator loading of 2882.84 kw for Division I and
2766.01 kw for Division II. The licensee has also submitted a Final
Safety Analysis Report (FSAR) change notice to the NRC reflecting the
above modifications.
This license condition is clor,ed.
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4. Site Tours
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The SRI and RI toured areas of the site during the inspection period to
gain knowledge of the plant and to observe general job practices.
1
l No violations or deviations were identified in this area of inspection.
5. Control and Use of Field Change Notices (FCNS)
During a routine followup from a previous NRC inspection, the SRI reviewed
the completed Field Change Notice (FCN) 19 to MR 85-0397. FCN-19 had
.
become a 96-page document and the SRI found that the design instructions
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were often misleading and difficult to follow. For example, page 36 of
FCN-19 added terminal board 43B contacts (contacts 7/8 and 3/4) on an
electrical schematic for remote shutdown operation of the diesel generator
- output breaker, and page 57 of FCN-19 removed the same contacts. Also,
l several pages of FCN-19 were found to be illegible due to poor reproduction
quality. The failure to provide clear and concise design instructions on
. an approved change to a safety-related design modification was identified
by the SRI as an apparent violation (8577-01). The SRI immediately
informed the licensee of this finding and the licensee took immediate
action to review other safety-related design modifications for similar
problems. Although similar documentation problems were found, hardware
inspections revealed no problems with the installed hardware.
6. Licensee Quality Concern Program Review
The licensee's quality concern prouram was described in NRC Inspection
Report 50-458/84-25. The SRI continued to review licensee actions on
'
individual cases and completed the review of the current status of all
open quality concerns,
t
- No violations or deviations were identified in this area of inspection.
7. Initial Criticality Witness
i
l This area of inspection was conducted to review licensee preparations for
initial critical and to witness initial criticality in order to ascertain l
l confomance to license and procedural requirements and to observe
l operating staff performance. The licensee completed preparations for
'
initial criticality on Octooer 30, 1985, and began pulling the first
control rod at about 9:30 p.m. The reactor was declared critical at
12:38 a.m. on October 31, 1985, on step 27 of the rod withdrawal sequence
with 2032 notches withdrawn. Nuclear instrumentation was monitored closely
during approach to criticality and appeared to respond as expected.
Also, during this inspection period, the RI completed the review of the
results of 13 STPs performed prior to achieving initial criticality. The
RI verified that the STP had been properly signed by the personnel
performing the test, that the data was collected and recorded as required,
.
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that all deficiencies had been addressed and corrective action completed,
and that the surveillance tests had been performed within the required
frequency.
The. SRI concluded that the licensee preparations for initial criticality
were cautious and the operating staff performance was controlled and
cautious. License and procedural requirements appeared to have been
implemented for a safe and cautious initial criticality at River Bend.
No violations or deviations were identified in this area of inspection.
8. Pipe Support and Restraint S_vstems Testing
During the initial nuclear heat up cycle, the R1 visually examined system
pipe supports to ascertain whether system thermal expansion vrould result
in any apparent failures to the dynamic pipe supports, fixed pipe supports,
or their associated component support structures. These visual examirationr.
were performed at ambient plant temperature and again at a plant
temperature of 350 F. Pipe supports and structures were examined to
verify where applicable, that; (a) deterioration, corrosion, physical
damage or deformation were not noticeable; (b) bolts, nuts, washers, were
tight and secure; (c) the equipment was not in a " lock up" or " frozen"
position; and (d) pipes, supports or other associated equipment or
compcnents were not in contact or cause rubbing due to thermal expansion.
The following pipe supports and associated support structures were examined:
Type System Identification No.
Spring Residual Heat Removal RHS-PSSH-30775
Spring Recirculation Pump Housing H-306A
Spring det Pump Header H354A
Spring Low Pressure Core Spray CSL-PSSH-3001
Spring Low Pressure Coolant Injection RHS-PSSH-3089
Snubber Feed Water FWS-PSSP-3021
Snubber Residual Heat Removal RHS-PSSP-3078
Snubber det Pump Header S-354A
Snubber det Pump Header S-3598
Snubber det Pump Header S-304B
Snubber Residual Heat Removal Rl15-PSSP-3011
Snubber Recirculation Pump Motor S-371B
Snubber Recirculation Pump Housing S-306A
Snubber Feedwater FWS-PSSP-3015
Hanger Feedwater FWPRR-804
11 anger Feedwater FWS-PRR-823
Hanger Feedwater FWS-PRR-808
Hanger Residual Heat Removal RHS-PRR-831
Bracket Main Steam MSS-814
Bracket Main Steam MSS-822
Bracket Main Steam MSS-823
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! The RI did not identify any conditions which would appear to adversely
affect the operation of the examined system pipe supports. The system
expansion data for Startup Test Procedure (ST) 1-ST-17, " System
Expansion," will be reviewed to ensure that expansion problems do not -
exist at operating temperature and pressure for the above pipe supports i
i and associated structures. The results of this review will be documented
in a subsequent inspection report.
No violations or deviations were identified in this inspection area.
9. Human Engineering Discapancies (HEDs) Review
During this inspection period the NRC RI reviewed Human Engineering ?
Discrepancy (HED) records to verify that identified improvements in the
control room / operator interface, for selected HEDs, had been completed.
- ,. The HEDs were documented in the River Bend Station Detailed Control Room
l
Design Review Summary Report, issued in October 1984.
The following HEDs were verified complete:
.
847 Label "SRV Low-Low Set / Reset" for A and B train.
4
833 Label recorders 1CliS-PR2A and 2B to indicate containment
pressure.
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8 Mimic of RHR system. l
! 25 Establish lines of demarcation for RHR Loop B and C.
!
. 830 Label switch for manual initiation of containment as
l " Isolation."
248 Label as ADS safety relief valve.
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747 Add escutcheons for manual steam isolation manual initiation
! push bottoms.
] 431 Modify total feedwater flow recorder to read in % flow.
l 328 Annunciator windows consistent in type and style.
l 422 Identify operational limits or warnings on meters.
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q 575 Accurately identify valves (up to root valves) by tagging.
307 Mcdify "HPCS NOT READY FOR AUTO START" annunciator circuit to
actuate for any of the seven conditions which will render HPCS
incapable of auto start.
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No violations or deviations were identified in this area of inspection.
10. Startup Test Procedures Review
The purpose of this area of the inspection was to review selected ST
procedures for compliance with regulatory requirements, FSAR commitments
and Technical Specifications. The NRC inspectors reviewed ST Procedure
1-ST-19. " Core Performance," Revision 0, ST Procedure 1-ST-26, " Safety
Relief Valves," Revision 0, and a draft major change request to 1-ST-26,
and ST Procedure 1-ST-28, " Shutdown From Outside the Control Room,"
Revision 0.
No violations or deviations were identified in this area of inspection.
11. Startup Test Witness
During this inspection period, the NRC inspectors witnessed startup testing
and operational activities conducted under the low power testing program.
Testing and operational activities witnessed included a reactor startup on
November 17,~1985, a reactor core isolation cooling (RCIC) turbine run on
November 17, 1985, and initial turbine roll on November 26, 1985.
River Bend received a full power operating license (NPF-47) on November
20, 1985, and reactor power was raised above 5% on November 25, 1985.
No violations or deviations were identified in this area of inspection.
12. Apparent Power Level Exceeding Licensed Level of 5%
On November 18, 1983, the apparent power level exceeded the licensed
level of 5%. The SRI, assisted by a Region IV supervisor, conducted an
inspection to determine the facts of this occurrence,
a. Background
On November 15, 1985, the NRC commissioners voted to allow the NRC
staff to issue a full power license to the River Bend facility. The
full power license was subsequently issued on November 20, 1985.
b. _0_ccurrence
During the night shift of November 17-18, 1985, the on-duty shift
supervisor directed power be raised to approximately 6.5% to
facilitate turbine steam chest warm up. Although other on-watch
personnel questioned this order, it was carried out. The shift
supervisor on watch stated to NRC personnel that he understood th6t
the Commission vote authorized operation above 5% power. The
indicated power of approximately 6.5% power was maintained for about
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2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The event was tcrminated when the oncoming shift supervisor
questioned why power had been raised above 5%.
c. Determination of Actual Power
The NRC inspectors reviewed the licensee's two methods of
quantifying actual power during the event. The first method by a
secondary calorimetric calculation. This calculation indicated that
actual power was about 4.5%; however, the accuracy of this
calculation was not considered good because feed flow was very low.
The measurement of feed flow was not considered to be accurate at
the very low end of the indicating range of the measuring device.
The second calculation was based on the valve position of the
turbine by-pass valves. There are two turbine by-pass valves at
River Bend. Each of these valves is rated at 5% of total power.
One of the valves was shut and the other valve was 50% open. Data
from General Electric taken on identical valves at other plants
indicated that these bypass have linear response or flow
characteristics (e.g., a valve that is 50% open would pass one-half
of its total capacity). The licensee' calculated that the one bypass
valve which was 50% open accounted for 2.5% of reactor power. !
Allowing for other heat losses such as steam traps and the reactor
water clean up system,' the licensee calculated that actual power was
less than 3.5%
d. Inspection Findings
'
The NRC inspectors noted that nuclear instruments could not be set
accurately until power was at a higher level so that an accurate
heat balance'could be made. Nuclear instrument gains were set high,
which is the conservative approach. Indicated power was thus
greater than actual power, and actual power had probably not
exceeded the licensed limit. The NRC inspectors concluded that
there had been a serious breakdown in training and communications.
As the result of interviews with operators and supervisors, it
appeared that this communication breakdown was limited to the one
shift supervisor who had directed the power increase. The NRC
inspectors noted that the licensee had removed this shift supervisor
from licensed duties for retraining and evaluation. It was also
found that the licensee had taken measures to improve communications.
These included the presence of an operations management individual
at all shift turnover briefings, written instructions to operators on
power limit and the resetting of rod blocks to 5%.
It was concluded that continued close monitoring of shift performance
communications was warranted to assess the effectiveness of the
corrective actions taken.
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13. Exit Interview. ,
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j An exit interview was conducted on December.12,.l'985, with licensee
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representatives (identified.in paragraph 1). During this interview, the
. SRI-reviewed the scope and findings of.the' inspection.
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