ML20136H390
| ML20136H390 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 08/05/1985 |
| From: | Johnson E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | William Cahill GULF STATES UTILITIES CO. |
| References | |
| NUDOCS 8508200290 | |
| Download: ML20136H390 (2) | |
See also: IR 05000458/1985031
Text
4
o
AUG 0 51985
/
/
In Reply Refer To:
Docket: 50-458/85-31
50-458/85-43
Gulf States Utilities
ATTN: William J. Cahill, Jr.
Senior Vice President
River Bend Nuclear Group
P. O. Box 2951
Beaumont, Texas 77704
Gentlemen:
This refers to the above-referenced inspection reports which were sent to
you on August 2, 1985. During assembly of these reports the appendices were
inadvertently switched. Enclosed you will find the inspection reports along
with their correct appendices.
Sincerely,
/
Orla!nct E!-+ !hj
E. H. J " :i
E. H. Johnson Chief /
Reactor Projec Br hch
cc:
Gulf States Utilities
ATTN: J. E. Booker, Manager-
Engineering, Nuclear
Fuels & Licensing
P. O. Box 2951
Beaumont, Texas 77704
Louisiana State University,
Government Documents Department
Louisiana Radiation Control Program Director
RIV:RPB Y
EHJohnson/lk
7/f/85
8508200290 050005
gDR ADOCK 05000458 ,
i I
t
i
4
Gulf States Utilities -2-
bec to DMB (IE01)
bec distrib. by RIV:
RPB R. P. Denise, DRSP
Resident Inspector R. D. Martin, RA
SectionChief(RPB/A) *D Weiss, LFNB (AR-2015)
MIS System
RSTS Operator
EP&RPB
RIV File
.'
- .
,
,.'.
,,
o
.
AUG 0 21985
In Reply Refer To:
Docket: 50-458/85-31
Gulf States Utilities
ATTN: William J. Cahill, Jr.
Senior Vice President
River Bend Nuclear Group
P. O. Box 2951
Beaumont Texas 77704
Gentlemen:
This refers to the inspection conducted by Ms. K. A. Whittlesey and Mr. J. 1.
Tapia of this office during the period April 3-8, 1985, of activities
authorized by MRC Construction Permit CPPR-145 for River Bend Station, and to
the discussion of our findings with you, and other members of your staff at the
conclusion of the inspection.
Areas examined during the inspection included witnessing the containment
structural integrity test and integrated leak rate test. Within these areas,
the inspection consisted of selective examination of procedures and representa-
tive records, interviews with personnel, and observations by the inspector.
These findings are documented in the enclosed inspection report.
During this inspection, it was found that certain of your activities were in
violation of NRC requirements. Consequently, you are required to respond to
this violation, in writing, in accordance with the provisions of Section 2.201
of the NRC's " Rules of Practice," Part 2. Title 10, Code of Federal
Regulations. Your response should be based on the specifics contained in the
Notice of Violation enclosed with this letter.
The response directed by this letter and the accompanying Notice is not subject
to the clearance procedures of the Office of Management and Budget as required
by the Paperwork Reduction Act of 1980, PL 96-511.
RIV:RPS RPB RPS2/8 RPO B1
ittlesey JITapia DHunnicutt JPJaGdon ghnson
/M/85 7/l%/85 1/d/85 )/(/85 *
/\/85
, .~ ~L
g r + N
e
'
i
,.,
.
Gulf States Utilities -2-
Should you have any questions concerning this inspection, we will be pleased to
discuss them with you.
Sincerely,
.. . -. '..~y
3. . J ..an
E. H. Johnsoi, Chief
Reactor Projset Branch 1
Enclosures:
1. Appendix A - Notice of Violation
2. Appendix 8 - NRC Inspection Report
50-458/85-31
. cc w/ enclosure:
Gulf States Utilities
ATTN: J. E. Booker, Manager-
Engineering, Nuclear
Fuels & Licensing
P. O. Box 2951
Beaumont, Texas 77704
Louisiana State University,
Government Documents Department
Louisiana Radiation Control Program Director
bec to DM8 (IE01)
bec distrib. by RIV:
'*RP81 R. P. Denise DRSP
- Resident Inspector R. D. Martin, RA
- Section Chief (RP81/A) *D. Weiss, LFM8 (AR-2015)
- MIS System
- RSTS Operator
- EP&RPB
- RIV File
- w/766
.
,
,
.
APPENDIX,A_
NOTICE OF VIOLATION
Gulf States Utilities Docket: 50-458
River Bend Station Permit: CPPR-145
During an NRC inspection conducted April 3-8, 1985, a violation of NRC require-
ments was identified. The violation involved an inadequate procedure. In
accordance with the10
Enforcement Actions," " General
CFR PartStatement
2. AppendixofC Policy (and Procedure for NRC1985), t
below:
Inadequate Test Procedure
10 CFR Part 50, Appendix B, Criterion V states that procedures, drawings,
and instructions shall be used to control activities affecting quality
related activities.
Contrary to the above Preonerational Procedure 1-PT-57-1 failed to
identify six differential pressure instruments and one motor operated
valve required for the valve lincup to conduct the Intergrated Leak Test.
This is a Severity Level V Violation. (Supplement II.E)(458/85-31-01)
Pursuant to the provisions of 10 CFR 2.201, Gulf States Utilities is hereby
required to submit to this office, within 30 days of the date of this Notice, a
written statement or explanation in reply, including: (1) the reason for the
violation if admitted, (2) the corrective steps which have been taken and the
results achieved, (3) the corrective steps which will be taken to avoid further
violations, and (4) the date when full compliance will be achieved. Where good
cause is shown, consideration will be given to extending the response time.
Dated at Arlington, Texas
this 2d day of Aug., 1985
-
.
.
'
,
,
.
.
APPENDIX 8
U. S. NUCLEAR REGULATORY COMMISSION
REGION IV
NRC Inspection Report: 50-458/85-31 License /CP: CPPR-145
Docket: 50-458
Licensee: Gulf States Utilities
P. O. Box 2951
Beaumont, Texas 77704
Facility Nau: River Bend Station
Inspection At: River Bend Station, St Francisville, Louisiana
Inspection Conducted: April 3-8, 1985
Inspectors: . / b 7 17 86
K. A W ttinsey,'Re ~ or Inspe16 tor, Project Date
Sect , Reactor P ect Branch 2
) /) ] l~1 ! S5
J. Tapid, NactQ Inspector, Project Branch 8 Date
Rea Project BrMch 2
/
Approved; /]]) N $l SY
~
J P audg6 Chief, Project Section A. Reactor Date
r et granch 1
Inspection Summary
Inspection Conducted April 3-8, 1985 (Report 50-458/85-31)
, , .. , -
r 1
,_
- .- . . _ _ - . . _- - - .-._.-._ __ _- _ _ _ _ _ - __ .- _ . _ .
.
'
.
,
. .
2
Areas Inspected: Routine, announced inspection of containment structural
'i integrity test and integrated leak rate test. The inspection involved 51
inspector-hours onsite by two NRC inspectors.
Results: Within the two areas inspected, one violation was identified (failure
to provide adequate procedure, paragraph 3).
!
r
e
1
1
a
I
.
l
4
.
e
-!
l
4
. . - - _-_-.--.-..-._m, -- - . - .- -.__ _ . . _ , _ . _ , , .- , - . - ,-,.,.----.m - --- , ,, , ., - - - - -- -r
_ _ _ .. _ - _ _ . __-
.
'
'
.
. .
3
L
DETAILS
1. Persons Contacted
Gulf States Utilities (GSU)
'
- C. M. Coones, Civil Engineer
- T. L. Crouse, Manager Quality Assurance
- P. J. Dautel, Licensing Staff Assistant
- J. C. Deddens, Vice President
- P. E. Freehill, Superintendent Startup and Test
- E. R. Grant, Supervisor Licensing
J. E. Lozes, Senior QC Inspector
- G. R. Kimmell, Supervisor Operations Quality Assurance
- G. V. King, Plant Services Supervisor
"E. R. Oswood, Quality Assurance Engineer
C. D. Payton, Field Quality Control Level II
"T. F. Plunkett, Plant Manager
J. E. Redmond, Senior QC Inspector
"T. E. Suhrke, Manager Project Planning and Coordination
Stone and Webster
- R. H. Bernier, Senior Advisory Engineer
- J. L. Busa, Assistant to the Chief Engineer
"F. W. Finger III, Project Manager, Preliminary Test Organization
- R. I. Parry, Supervisor, Mechanical Test Engineering
- Indicates presence at exit interview conducted April 8, 1985.
2. Structural Integrity Test (SIT)
The purpose of the SIT is to demonstrate the ability of the containment
vessel to withstand internal loads imposed by pressurizing to 1.15 times
the design pressure of 15 psig. Preoperational Test Procedure No. 1
PASIT.001, Revision 1, " Pressure Test of the Steel Containment," was
reviewed and determined to meet NRC requirements and licensee commitments
listed in the' Final Safety Analysis Report. The test, already in progress
at the time of the NRC inspector's arrival on site, was being conducted in
accordance with the reviewed and approved test procedure.
The inspectors reviewed the procedure for ultrasonic monitoring of
electrical penetrations (attachment 10.2 to 1 PASIT.001) to be performed
,
under step 7.19. The ultrasonic leak detection method was implemented in
lieu of local pneumatic tests of the circumferential butt welds in
electrical penetration nozzles. Calibration of acoustic monitoring
equipment on a mockup in the annulus area was observed prior to
.
l
l
_ _ . - - -- . - _ . -
. l
.
n.
.
. 9
4
commencement of the ultrasonic inspection by field quality control
personnel. The NRC inspectors observed and independently monitored
inspection of several penetrations and noted that no leaks were detected.
The method for discerning the equipment response to leakage from response
to background noise was described to the inspectors as well as
demonstrated.
1
At the completion of the SIT, the containment was depressurized.
The preoperational containment integrated leak rate test conducted in
accordance with Preoperational/ Acceptance Test Procedure 1-PT-57-1
" Integrated Leak Rate Test," was addressed during this portion of the
inspection. The inspection included procedure and records review, test
- witnessing, and independent calculations by the NRC inspectors. The
inspection was performed in order to ascertain whether testing was
conducted in accordance with approved procedures and satisfied the
specified acceptance criteria of 10 CFR 50, Appendix J and the Final
Safety Analysis Report.
After a period of time at atmospheric pressure to allow for degassing of
structures and components inside containment subsequent to the SIT,
pressurization of the containment vessel for the ILRT commenced.
Stabilization commenced after internal pressure reached 8.6 psig (23.3
'
psia), the compressors were shut down and isolated. The atmosphere is
considered stabilized when the rate of change of containment temperature
averaged over the last 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> minus the rate of change in containment
About
temperatureaveragedoverthelasthourislessthan0.5'F/@that
.
11:00 a.m. on April 5, 1985, the NRC inspectors were advise
containment atmosphere stabilization had been achieved and the official
24-hour test had begun. Initial calculated leakage was excessive and
attempts to identify the leakage source resulted in the following sequence
of events.
ILRT configuration includes a pneumatic block of main steam isolation
valves (MSIVs). An increasing pressure trend was noticed on the pressure
gauge on the outboard MSIVs indicating leakage across the inboard MSIVs.
After the piping between inboard and outboard MSIVs equalized with
containment pressure, the downstream piping was sealed at a pressure of
8.5 psig to minimize the effect of this leakage. It should be noted that
the MSIVs are supplied with a positive leakage control system, which would
be pressurized above peak postulated accident pressure in the event of a
design basis accident. Additionally, Valve 1 RHS-V15, an instrument root
valve on Residual Heat Removal B pump discharge piping was found out of
position and leaking a steady stream of water. Although the valve lineup
sheet and control room tagout log both indicated the valve to be in the
I
l
l
._
. - - - _ _ _ _ - - _ - _ . . - - .
-- _ - _ _ - - -
_ __
,
.
..
. .
5
'
closed position, it was found open by the licensee. The importance of
compliance with tagging procedures was discussed at the exit meeting,
although this was considered an isolated case. The valve was closed, but
calculated leakage remained excessive, indicating a remaining unidentified
leakage source. Further investigation by the licensee identified a direct
leakage path from the containment vessel to the annulus via three
instrument lines associated with the containment to annulus differential
pressure monitoring portion of the annulus pressure control system.
Instrument root, isolation, and equalization valves were open for 1
HVR*PDT GOA, 1 HVR*PDT 60C, and 1 HVR*PDT 60E. This allowed a direct
leakage path from containment to the annulus because six differential
pressure instruments were omitted from the ILRT valve lineup. The subject
instrument and associated root valves were added to the ILRT lineup via
Minor Change Request (MCR) Number 4; a correct lineup was achieved, and
the ILRT was restarted.
After the restart of the test, motor operated valve 1 DFR*MOV 146, a
designated containment isolation valve, w s determined to be in the closed
condition. 10 CFR 50, Appendix J requir,s that closure of containment
isolation valves for the ILRT be accompl shed by normal operation. In
this case, normal operation would indicate response of the valve to a
containment isolation signal. However, the valve, having been recently
installed under authorization of Engineering and Design Coordination
Report Number P13043B, was not yet connected to electrical supply so it
could not be closed by normal means. The valve had bten hand closed prior
to initiation of the test, and it was not included in the ILRT valve
lineup, although the expressed intent had been to call 1 DFR*MOV 146 open
in the lineup and rely on check valves 1 DFR*V131 and 1 DFR*V132. The
omission of differential pressure instruments 1 HVR* POT 60A, 1HVR*PDT 608,
1 HVR*PDT 60C, 1 HVR*PDT 600, 1 HVR*PDT 60E, and 1 HVR*PDT 60F and the
omission of 1 DFR*MOV 146 from the valve lineup for 1-PT-57-1 constitute a
violation for failure to have adequate procedures (458/8531-01).
1 DFR*MOV 146 was opened, and the test continued. There was no noticeable
perturbation in the test data which could be attributed to the valve
manipulation, and open drain valves outboard showed no sign of water
leakage. It should be emphasized that as a portion cf the reactor
building floor drain system for pump back suppression pool water
inventory, the referenced valve would be exposed to a water seal rather
than directly exposed to containment atmosphere. Pending final acceptance
and test demonstrating a fluid leakage rate within technical specification
limits for 1 DFR*MOV 146, this item remains open. (485/8531-02).
Continuation of the test indicated convergence of the calculated leak rate
and the upper confidence limit below the allowable leakage. At completion
of the 24-hour test, the superimposed leak verification portion of the
test was performed with results between the calculated and imposed
leakages within the 25% La limit.
.
..- *.
6
Subsequent to the performance of the test, the NRC inspectors obtained the
raw data and computed the leakage rate in accordance with the Mass Point
Data Analysis technique. The computations performed by the NRC inspector
were compared with the licensee's results for the purpose of verifying the
calculational procedure and confirming the results. This analytical
technique confirmed the acceptability of the results obtained by the
licensee.
,
4. Exit Interview
The NRC inspector met with the licensee representatives denoted in
paragraph I at the conclusion of the inspection. The NRC inspector
summarized the scope and findings of the inspection.
.
,
. .' ;
AUG 0 21985
In Reply Refer To:
Docket: 50-458/85-43
Gulf States Utilities
ATTN: William J. Cahill, Jr.
Senior Vice President
River Bend Nuclear Group
P. O. Box 2951
Beaumont Texas 77704
Gentlemen:
This refers to the inspection conducted by Messrs. D. D. Chamberlain, J. P.
Jaudon, and J. R. Boardman of this office during the period May 1 through June
15, 1985, of activities authorized by NRC Construction Permit for River Bend
Station, and to the discussion of our findings with Mr. J. C. Deddens and other
members of your staff at the conclusion of the inspection.
Areas examined during the inspection included review of licensee action on
previous inspection findings, site tours, reactor protection system pre-
operational test witness, reactor pressure vessel leakage test witness, control
rod drive system full core scram test witness, reactor coolant system
hydrostatic test results evaluation, construction quality assurance (QA)
program review, Inspection and Enforcement (IE) Bulletin follow up, and
allegation follow up. Within these areas, the inspection consisted of
selective examination of procedures and representative records, interviews with
personnel, and observations by the inspector. These findings are documented in
the enclosed report.
During this inspection, it was found that certain of your activities were in
violation of NRC requirements. Consequently, you are required to respond to
this violation, in writing, in accordance with the provisions of Section
2.201 of the NRC's " Rules of Practice," Part 2. Title 10. Code of Federal
Regulations. Your response should be based on the specifics contained in the
Notice of Violation enclosed with this letter.
The response directed by this letter and the accompanying Notices are not
subject to the clearance procedures of the Office of Management and Sudget as
required by the Paperwork Reduction Act of 1980 PL 96-511.
'
AIV:$ - R R A B
C riain/lw rdman J ydon pnson
d. / 5 // /85 ///85 g///85
,e p ,
dD
e 'J ?
. - _ _ _ _ _-- . _ . ___ .__ ._.
.
.
-
s
-2- .
Should you have any questions concerning this inspection, we will be pleased to
discuss them with you.
Sincerely,
L::;. .. . :.f
J. P. J u; n
E. H. Johnson, Chief
Reactor Projects Branch
Enclosures:
1. Appendix A - Notice of Violation
2. Appendix B - NRC Inspection Report -
50-458/85-43
cc w/ enclosures:
Gulf States Utilities
ATTN: J. E. Booker, Manager-
Engineering, Nuclear
Fuels & Licensing
P. O. Box 2951
Beaumont, Texas 77704
Louisiana State University
Government Documents Department
Louisiana Radi.ition Control Program Director
bec to DMB (IE01)
bec distrib. by RIV:
9PRPB1 R. P. Denise, DRSF
- Resident Inspector R. D. Martin, RA
- Section Chief (RPB1/A) *D. Weiss, LFMS (AR-2015)
- MIS System
- RSTS Operator
- EP&RPB
-
- RIV File l
- w/766
-_- _ - . - . . . _ _ _-. ..
. - . _ _- - . - _ _ - . .
- - ___ __ - - ._ . _ _
.
'
,. .
APPENDIX A
NOTICE OF VIOLATION
Gulf States Utilities Docket: 50-458/85-43
River Bend Station Permit: CPPR-145
During an NRC inspection conducted during the period of May 1 through June 15,
1985, a violation of NRC requirements was identified. The violation involved
the failure to obtain NRC approval of a QA program change. In accordance with
the " General Statement of Policy and Procedure for NRC Enforcement Actions,"
10 CFR Part 2, Appendix C (1985), the violation is listed below:
Failure to Obtain NRC Approval of a QA Program Change
10 CFR 50.55(f)(3) requires that after March 11, 1983, each construction
permit holder described in paragraph (f)(1) of this section may make a
'
change to a previously accepted QA program described, included or
referenced in the Safety Analysis Report (SM<), provided the change does
not reduce the commitments in the program description previously accepted
, by the NRC. Changes to the QA program tiescription that do reduce the
commitments must be submitted to NRC and receive NRC approval before
implementation. -
Contrary to the above, the licensee revised and implemented SAR Section
17.1.2.4.A (previously Section 17.1.2.6) without NRC approval which
reduced periodicity of GSU review of all controlled documents from 1 year
to 2 years.
This is a Severity Level V violation (Supplement I.E.) (458-8543-01).
.
Pursuant to the provisions of 10 CFR 2.201, Gulf States Utilities is hereby ~
'
required to submit to this office within 30 days of the date of the letter
transmitting this Notice, a written statement or explanation in reply,
a including: (1) the reason for the violation if admitted; (2) the corrective
I steps which have been taken and the results achieved; (3) corrective steps
which will be taken to avoid further violations; and (4) the date when full
compliance will be achieved. Where good cause is shown, consideration will be
given to extending the response time.
Dated at Arlington, Texas,
this 2d day of Aug. , 1985.
1
__ . _ _ _ _ -. _- _ -- . _ - - .-- -_ .- - - . . . .
, !
. ..e
.
. .
APPENDIX B
U. S. NUCLEAR REGULATORY COMISSION
REGION IV
NRC Inspection Report: 50-458/85-43 CP: CPPR-145
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 l
Inspection Conducted: May 1 through June 15, 1985
Inspect : ul 44^
D. 7. C6am@ rTain, Senior Resident Inspector Date
(pars. 1,'2, 3, 4, 5, 6, 7, 11)
[
til N
Date
h
T.B rtiman, Reactor Inspector, Operations
Setti n, Reactor Project Branch (par. 8)
,
q
(Ll/ NV) E/
J. . aud Chief, Project Section A, Reactor Date
roj{ect anch
Approved: c_ - /) 4 / 4A [ ~
J./ P. pa n, Chief, Protect' Section A, Date~
[Reac'to Projects Branch
- Inspection Summary
Inspection Conducted May 1 through June 15, 1985 (Report 50-458/85-43)
Areas Inspected: koutine, unannou.1ced inspection of licensee action on
previous inspection findings, site tours, reactor protection system
preoperational test witness, reactor pressure vessel leakage test witness,
control rod drive system full core scram test witness, reactor coolant system
v. ,
n __
g upQ wK)d'J.u
_ - _ - _ _ . _ - , _ , - - _ -
.
.
.
. .
'
2-
hydrostatic test results evaluation, construction quality assurance (QA)
program review, Inspection and Enforcement (IE) Bulletin followup, and
allegation followup. The inspection involved 147 inspector-hours onsite by 1
three NRC inspectors.
Results: Within the areas inspected, one violation was issued in the area of
construction QA program review (failure to obtain NRC approval of a QA
program change, paragraph 8).
l
.
i
_ _ _ _ _ _ . _ _ _
. . .
,
,
.; e
.
6
-3-
DETAILS
I
1. Persons Contacted
Principal Licensee Employees
t
K. Arnstedt, Quality Assurance (QA) Engineer
- W. H. Cahill, Jr. , Senior Vice President, River Bend Nuclear Group
- T. F. Crouse, QA Manager
J. Davis, QA Engineer
- P. J. Dautel, Licensing Staff Assistant
- J. C. Deddens, Vice President, River Bend Nuclear Group
D. R. Derbonne, Supervisor, Startup and Test
S. Finnegan, Control Operating Foreman
- P. E. Freehill, Superintendent, Startup and Test
- D. R. Gipson, Assistant Plant Manager, Operations
P. D. Graham, Assistant Plant Manager, Services
R. W. Helmick, Director, Projects
K. C. Hodges, Supervisor, Quality Systems
D. Jernigan, Engineer, Startup and Test
- G. R. Kimmell, Supervisor, Operations QA
- G. V. King, Supervisor, Plant Services
J. L. Pawlik, Engineer, Startup and Test
- T. L. Plunkett, Plant Manager
- S. R. Radebaugh, Assistant Superintendent, Startup & Test
- S. F. Sawa, Control Superintendent, Startup & Test
- J. E. Spivey, QA Engineer
R. B. Stafford, Director, Quality Services
K. E. Suhrke, Manager Project Planning & Coordination
L. Sutton, QA Engineer
- P. F. Tomlinson, Director, Operations QA
- A. Valenzuela, Startup and Test .
- J. Venable, Mechanical Maintenance Supervisor
D. White, Engineer, Startup and Test
Stone and Webster
D. P. Barry, Superintendent of Engineering
W. I. Clifford, Senior Construction Manager
F. W. Finger, III, Project Manager, Preliminary Test Organization (PTO)
M. Fischete, Engineer, Startup and Test
- P. H. Griffin, Site Advisory Manager
B. R. Hall, Assistant Superintendent, Field Quality Control (FQC)
Q. E. Harper, Hydro Test Engineer
D. Hill, Maintenance Engineer
R. L. Spence, Superintendent, FQC
The NRC senior resident inspector (SRI) also interviewed additional
, . - -
__
- _ - - . . - - - - - _ - - _ . - - . -._ . _ _ . _ _ _ _ _.
. ..-
,.
5
l -4-
'
licensee, Stone and Webster (S&W), and other contractor personnel during
j the inspection period.
I * Denotes those persons that attended the exit interview conducted on
5
June 21, 1985.
2. Licensee Action on Previous Inspection Findings
! a. (0 pen) Open Item (458/8408-01): Review to determine if and how the
. diesel generator loading restrictions of calculation 12210-E-122 are
j implemented in plant operating procedures.
,
- The SRI obtained a copy of calculation 12210-E-122, Revision 4,
- " Standby Diesel generator Loading Calc. " dated March 1, 1984. This
!
revision of the calculation is based on a 3500 KW loading limit on
i the diesel and does not reflect the latest load restriction of 3130
! KW. However, a GSU letter R8G - 20,086 dated February 6, 1985,
- contains revisions to the FSAR "to establish a qualified load for
- each of the diesel generators." These revisions include, for diesel
manually tripped after 2.0 hr of LOCA, depending upon the available
diesel generator sets" and for diesel 1EGS*EG1B loading, a requirement
that "RHR C is stripped manually by the operator after 2.0 hr of
. operation after LOCA, depending upon the combination of the available
- diesel generator sets."
4
l' Abnormal Operating Procedure A0P - 0004, Revision 1, " Loss of Offsite !
Power," dated April 10, 1985, appears to address the latest loading
+ restrictions (3130 KW) for the diesels, but the stated action for
- clarification. For example, it is not clear under what conditions
'
that RHR A pump is tripped instead of the LPCS pump. This item will
i remain open pending issue of an approved calculation reflecting
- qualified diesel loading and pending the required clarifications in
procedure Section 5.8.
l
i b. (0 pen) Unresolved Item (458-8408-04): Review of licensee program
j for tracking of commitments to the NRC.
I
GSU has developed and implemented a commitment tracking program at
i River Bend. Project Procedure No. 8.2, " Identifying and Tracking
Project Commitments," was issued on August 9, 1984, to " provide
Guidelines for River Bend Nuclear Group (PBNG) organizations for
1 identifying, documenting, tracking, and closing commitments made to
regulatory agencies." Nuclear licensing is responsible for the
i tracking system with QA responsible for verification of completion
} of commitments on a sampling basis. GSU uses the " TRAC" computer
! program and they have identified approximately 2,152 commitments to
! date. The present status of the 1,191 open commitments is 717 high
j priority commitments required for fuel load and 156 required after
- fuel load, and there are 280 low priority commitments required for
4
fuel load and 38 required after fuel load.
!
!
!
.
-- .-._. - -. - -. - _ - _ _ _ _ - _
_ -
_ -. . . -- _ .- .- . - -_--_ _.
.
.
'k
,
. .
4
'
-5-
4 This item will remain open for the SRI to evaluate the GSU method of
identifying commitments that require closure prior to fuel load and
for identifying those that can be completed after fuel load.
c. (Closed) Open Item (458/8434-01): Implementation of preoperational
test commitments by the control rod drive (CRD) hydraulic
preoperational test procedure (1-PT-052).
The specific items of concern were implementation of the following
Final safety Analysis Report (FSAR), Chapter 14, test commitments:
"1.e. To verify the failure mode of the CRD
system on loss of power."
"3.j. The CRD pumps are tripped and the time for
accumulator inoperable alarms to occur is
- recorded as baseline data."
, "4.f. All scram valves open on a loss of
instrument air to the CRD system."
'
. The above items were addressed in the following manner:
- 1.e. The failure mode of the CR0 system was
tested by verifying the scram function on a
,
loss of power to the scram pilot solenoids.
3.j. A minor change request (MCR 09) was issued
,
'
to require recording of the times for all
hydraulic control units which exhibited low
accumulator pressure alarms within 10
minutes after tripping of the CR0 pump.
i 4. f. An acceptance criteria step 10.9 was added
i by a major change request (MRC 06) to
reference the backup scram valve test which
{ demonstrates that the scram valves open on a
loss of instrument air.
This item is closed.
- d. (Closed) Violation (458/8507-01)
- Procedures were not implemented
. to maintain Class B cleanliness requirements in the spent fuel storage
area where the new fuel was to be stored in accordance with the
w
Special Nuclear Material License issued on January 15, 1985.
l GSU took immediate action to issue an unsatisfactory inspection
report and fuel receipt was delayed 1 day to allow removal and
} inspection of the spent fuel racks and clean up of the spent fuel
j pool floor. Following the cleaning, the spent fuel racks were )
,
i
___ __ _ _._ .__ _ _._ _ _. _ __ _ _ _ _ _ _ _ . _ _ _ _ _ _ _
_ ____, _ -- _ J
- . _ _ _ _ _ _
.
.
. *
.
. .
-6-
reassembled in the pool and fuel receipt progressed as scheduled.
Also, to prevent recurrence of this type of item, housekeeping and
cleanliness procedures shall be implemented as a prerequisite to the
governing procedures.
This item is closed.
e. (Closed) Deviation (458/8507-02): Preoperational test procedures
are not being provided for NRC review 60 days prior to the scheduled
test performance in all cases.
Only four preoperational test procedures remained to be submitted for
NRC review at the time of this deviation. The four remaining
procedures were expedited and all have now been submitted.
This item is closed.
f. (Closed) Open Item (458/8522-07): GSU has installed motor operated
valve (MOV) circuit breaker trips that can be reset either manually
or automatically. GSU has not established control to verify that all
such resets are in the manual mode. ,
Motor control center starters for MOVs at River Bend have both
thermal overload trip and magnetic trip devices. The magnetic trip
devices trip the manual circuit breaker to remove the overload
condition. The magnetic trip device then resets automatically, but
the circuit breaker must be manually reclosed to provide power to the
starter. The thermal overload trip device opens the circuit to the
motor starter to remove the overload condition. The thermal overload
device has a hand / automatic option on reset. GSU has chosen to place
all of the thermal overload devices in the hand reset position.
Temporary Change Notice No.85-131 has been issued to revise Procedure
No. CMP-1026, " Corrective Maintenance of MCC Starters," to include a
step for verifying that hand automatic reset selectors are in the
hand position for thermal overload trip devices. GSU operations was
notified of this condition per memorandum APM-M-85-94 dated June 13,
1985. It was also noted during the review of this item that certain
loss of coolant accident initiated MOVs would have not thermal
_'
overload trip devices installed.
This item is closed.
> g. (Closed)) Open Item (458/8522-12): A system had not been provided
for assuring that each piece of measuring and test equipment (M&TE)
is calibrated and adjusted on or before the date required.
Procedure ADM-0029, Revision 4, " Control of Measuring and Test
Equipment (M&TE)," has been revised via temporary Change Notice
85-733 to clarify the system used and the responsibilities for the
. recall of M&TE for calibration. Also, in addition to the recall
-. - . - - . . _ - . _ - . - ._ -_ _ ___ _ _ _ . _ - . _ _ _ _ - . . _ - - -
. -. . ._. - _ _ .-.
.
,
...
. .
2
-7-
requirements, the M&TE issue facil:ty must verify that the M&TE
calibration due date is current pri1r to issue of the M&TE and users
of M&TE are required to verify that the calibration of M&TE is
current prior to use. All of there requirements are intended to *
,
preclude the use of any M&TE for waich the calibration has expired.
This item is closed.
3. Site Tours
The SRI toured areas of the site during the inspection period to gain
knowledge of the plant and to observe general job practices. The site
tours conducted included special tours on separate occasions with
Commissioner Bernthal and with a group from NRC Nuclear Reactor Regulation
headed by Harold Denton. Both of these tours included the conduct of mock
scenarios on the River Bend plant simulator.
No violations or deviations were identified in this area of inspection.
4. Reactor Protection System Preoperational Test Witness
The SRI witnessed portions of the reactor protection system response time
measurements testing conducted during this inspection period. The
specific testing witnessed included reactor pressure sensor response
timing, reactor vessel level sensor response timing and drywell pressure
sensor response timing. Testing personnel experienced several problems
with the set up of the response time test equipment which caused testing
delays. They also experienced problems with interpreting the response
time curves such that the proper ramp was generated for acceptance
criteria purposes. The vendor for the test equipment was brought to the
site and the test equipment problems were corrected. Also, a uniform
method for interpreting the response time curves was formulated. The SRI
conducted a preliminary review of several response time curves and it
appeared that the response times were within acceptance criteria limits.
The major testing remaining for the reactor protection system
preoperational test at the end of this inspection period was the
intermediate range monitor (IRM) and average power range monitor (APRM)
response time measurement testing.
No violations or deviations were identified in this area of inspection.
5. Reactor Pressure Vessel Leakage Test Witness
This special reactor pressure vessel (RPV) leakage test (1.MPRV.002) was
performed in order to disposition a Nonconformance and Disposition (N&D)
Report No. 11275. This N&D resulted when the review of the N-5 data
reports on the reactor pressure vessel indicated that no hydrostatic test
was performed subsequent to the installation of reactor internals or
+
rework on nozzle safe ends. This included installation of items such as
control rod drive (CRD) housings, incore housings, recirculation and
!
.
I
l
. .- '
..
-8-
feedwater safe end rework, jet pump penetration seals, etc. The reactor
I
pressure vessel system hydro procedure (1-G-ME-15) and associated
documentation did not identify the welds for these items as being within
the scope of the RPV hydro test inspections. Therefore, during the RPV
hydro conducted in May 1984, documentary evidence of the inspection of
o these welds was not obtained.
The SRI witnessed the RPV leakage testing and weld inspections performed
on May 16, 1985. The leakage testing was performed at a design pressure
of 1250 psig. Initially, trouble was experienced with obtaining the test
pressure due to excessive leakage through the gaged safety relief valves.
Test personnel obtained approval from the relief valve vendor which
allowed the relief valve gags to be torqued to 30 foot pounds psig at a
RPV pressure of 1000 psig. This was accomplished and they were then able
to obtain the required test pressure. The test pressure was held for a
minimum of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> prior to the performance of the official inspections. l
'
The inspections were performed and no problems were identified.
No violations or deviations were identified in this area of inspection.
6. Control Rod Drive System Full Core Scram Test Witness
During this inspection period, two back up. scram valve full core reactor
scram tests were performed to complete the control rod drive system
preoperational testing. Also, a special full core reactor scram test was
performed to evaluate the scram discharge volume level instrument response.
The SRI witnessed the performance of the special scram test on May 29, 1985.
This special scram test was performed in conjunction with reactor
protection system response time testing and the scram was initiated by a
reactor vessel low water signal.
No violations or deviations were identified in this area of inspection.
7. Reactor Coolant System Hydrostatic Test Results Evaluation
The SRI conducted a review of the completed test results for the RPV ,
system hydrostatic test (Procedure 1-G-ME-15) and for the subsequent
reactor pressure vessel leakage test (Procedure 1.MPRPV.002). The
specific areas reviewed and findings noted included the following: ,
'
l
,
a. Changes to the test procedures were documented and implemented in
accordance with the licensee's administrative controls,
I
b. The system boundary either included all piping and equipment
protected by the safety relief valves or documentation was provided
to show that separate hydros were performed on equipment or piping
that could be isolated from the RPV.
i c. The water quality met all requirements.
1
(
- . _ . - -
. - - _ _ _ _ - - _ _. .. _____ _
.
. ..'
. .
-9-
d. The licensee held the maximum test pressure (1.25 times the design
pressure) for at least 10 minutes during the RPV system hydro test.
e. The hydrostatic test pressure did not exceed the minximum pressure
allowed.
f. The reactor coolant temperature was maintained above the nil
ductility transition temperature throughout the hydro and leak
testing,
g. All identified test exceptions have been resolved, but a concern was
identified with Test Exception TE-13 for Procedure 1-G-ME-15. This
test exception addressed certain flexible hoses that were not
' installed at the time of the original hydro. There were 40 hoses
identified on the test exception and they were apparently identified
on system punch lists for installation and hydro at a later date.
The SRI selected 4 flexible hoses (Nos. 114, 121, 122, and 140) for a
review of the rework documentation to determine if the required
hydros were performed. Of the four selected, documentation was
obtained to verify that the flex hoses did receive a subsequent hydro
test. However, it was noted that the flange connections on these
flex hoses had been blanked during the hydro and only two of the four
rework control forms required a subsequent operational leak test
(OLT). This was discussed with test personnel and it was determined
that the performance of OLTs, on a flange connection that is completed
after a hydro, has been normal practice at River Bend. Further
review revealed that three of the four hoses received an OLT during
the RPV leakage test per a startup trouble ticket (STT). The fourth
hose was not specifically mentioned on the STT, but it was identified
on working drawings as being inspected. The SRI believes that no
problem was created by failure to note an OLT requirement on the
rework document and test personnel stated that the normal practice
will continue for performance of OLTs.
h. The test results have been reviewed and approved by those personnel
charged with the responsibility.
The SRI also reviewed selected vendor supplied pump and valve hydro
records and no problems were noted.
ha violations or deviations were identified in this area of inspection.
8. Construction QA Program Review
A review of the licensee construction QA program revealed that on June 10,
1983, GSU fonvarded for approval a revision of their construction QA
program as required by 10 CFR 50.55(f)(2). Discussions with licensee
personnel revealed that revised Section 17.1.2.4.A reducing the periodicity
of GSU review of all controlled documents from 1 year to 2 years was made
after March 11, 1983. Licensee personnel further stated that this change
had been implemented, although it had never been approved by NRC.
_ - _ _
.
. .o
.
.
l -10-
1
Implementationofthischangeisinviolationof10CFRPart50.55(f)(3),
, which requires prior NRC approval of licensee QA program changes made
! after March 11, 1983, when such changes reduce commitments. (8543-01)
l 9. Inspection and Enforcement (IE) Bulletin Follow Up
The purpose of this inspection was to followup on licensee action taken in
response to Inspection and Enforcement Bulletins (IEBs),
a. IEB 84-03
This bulletin concerned the consequences of a failure of the
refueling cavity seal. The NRC inspector reviewed the following
licenseecorrespondencetotheNRC(RegionIV).
Letter Serial Date Items Addressed '
RBG-19487 11-29-83 . Cross Seal Failure
. Maximum Leak Rate
Because of Seal
Failure 3
.Make Up Water
Capacity
. Potential Effect on
Stored Fuel and
Fuel in Transfer
.0ther Consequences
RBG- 20042 02-01-84 . Emergency Operating
Procedures
RBG- 20635 04-05-85 . Time to Cladding
Damage Without
Operator Action
RBG-21023 05-15-85 . Time to Cladding
Damage Without
Operator Action
These four letters address all of the points required by IEB 84-03.
The design of the seal used at River Bend is a stainless steel
bellows assembly welded to its support structure. The maximum
credible leakage rate is within make up capacity. Total failure of
the seal without operator action could result in a problem for fuel
in transit between the reactor vessel and the containment fuel
storage pool. This is addressed in licensee Procedure A0P-0032,
which requires the fuel to be placed in either the vessel or storage
racks. All other fuel would remain covered with water, and no vital
equipment would be flooded by a complete draining caused by a bellows
failure. The bellows is also protected from direct impact by a
radiation shield and a guard ring.
. ..-
. <
-11-
IEB 84-03 is considered closed.
b. IEB 77-06
This bulletin concerned General Electric Series 100 containment
electrical penetrations. River Bend does not use this type of
electrical penetration. Therefore, there is no action required at ,
River Bend for IEB 77-06.
IEB 77-06 is considered closed.
c. IEB 79-15
This IEB addressed deep draft pump deficiencies and the long tee
operability of these pumps.
The NRC inspector found that the licensee had addressed opvability ;
of pumps in the FSAR, and this was recognized in NUREG-0989, the
safety evaluation report for River Bend. Additionally, the final
draft technical specifications contained surveillance requirements
for monthly demonstration of pump operability in accordance with the
ASME code,Section XI and an 18-month system operability test. Since
the question of deep draft pump operability is being addressed by the
nomal review process and implementation of the requirements to test
is under the routine inspection program, no additional tracking of
this IEB is warranted.
IEB 79-15 is closed for record purposes.
d. IEB 79-23 -
This IEB concerned the potential failure of emergency diesel
generators. The failure could result if there was a large
circulating current between the exciter transformer and the
generator. Such a circuit could be set up by connection through a
common ground. It was found that the design of all three emergency
diesel generators at RBS was such that exciter transformers had a
floating primary neutral. Subsequent testing of the emergency diesel
generators did not disclose any problems of the type discussed in -
this IEB. -
IEB 79-23 is considered closed.
e. IEB 79-27
This IEB concerned the loss of non-Class 1-E instrumentation and
control power. This IEB was not specifically directed to River Bend,
but it was included in the FSAR review, becoming Question 421.003 and
as Confirmatory Item 31 of NUREG-0989, the Safety Evaluation Report.
Since the action required by this IEB is being tracked as a confirma-
tory item, IEB 79-27 is closed for record purposes.
-. _. .
._ _
.,
1
.* t
, . .
. .
-12-
,
f. IEB 80-08
This IEB addressed radiography of flued head design penetrations of
the containment. The licensee was found to use flued head design in
both the containment and the drywell. The licensee committed to use i
radiography on all flued head design penetrations of the containment
and other nondestructive tests on other penetrations.
This IEB is closed.
g. IEB 80-16
This IEB concerned Rosemont pressure transmitters, Models 1151 and
1152. When these transmitters were fitted with either "A" or "D"
output code cards, it was possible for the transmitters to have an
ambiguous output and the input signal was either an over pressure or
a reverse pressure signal. The licensee found that there were four
Rosemont 1152 transmitters with an "A" output code. These four
transmitters were modified to have an "N" output board.
IEB 80-16 is considered closed.
10. Allegation Follow Up
The NRC inspector did a followup inspection of an allegation.
. Background. An anonymous letter was sent to both Gulf States
Utilities (GSU) and the NRC. This letter forwarded an internal piece
of Stone and Webster (S&W) correspondence. This S&W correspondence
was a letter signed by engineers in the design group for small bore
pipe. The letter complained that the group was being required to
account for on-the-job time in a log and alleged that such a time
accounting procedure was inimical to quality assurance.
. Licensee Action. GSU conducted a quality assurance review of the
small bore piping group March 4-22, 1985.
. NRC Review. The NRC inspector reviewed the report of the licensees'
review. It was noted that the licensee had concluded that the
allegation was not substantiated in that the use of a time log to
account for time spent on various charge items does not have a direct
relationship to quality assurance. It was further noted that there
were four additional concerns noted by the quality assurance review.
The NRC inspector noted that the licensee had followed up on these
four concerns and closed them.
. Conclusion. The NRC inspector concluded that the allegation was
substantiated in that the time log was kept but that it was invalid
as a safety concern.
1 This item is closed.
_ ___ _ __ _ ____ __ _ _ _ _ _ _.. _ . . _ _ - - . _ _ _ _ _ _
i-
.
i . . . ,, e
- . .
l -13-
l
'
11. Exit Interview ,,
An exit interview was conducted on June 21, 1985, with licensee
representatives (identified in paragraph 1). During this interview, the
SRI reviewed the scope and findings of the inspection.
l
l
l
,
.
\)
.
.
i
!
!
- _ _ - - - _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _