IR 05000458/1990002
| ML20062A897 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 10/08/1990 |
| From: | Collins S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | Deddens J GULF STATES UTILITIES CO. |
| References | |
| NUDOCS 9010230209 | |
| Download: ML20062A897 (2) | |
Text
[~
.
n-
.
.
[
-
00T - 81990
-
p
.
.
.In Reply Refer To:
,
Docket: 50-458/90-02
!
Gulf States Utilities j..
ATTN: James C. Deddens SeniorVicePresident(RBNG)
,
P.O. Box 220
St. Francisv111e, Louisiana 70775 Gentlemen:
Thank you for your letter of September 18, 1990,. which provided a revised response to our letter and Notice of Violation dated April 6, 1990. We have reviewed your revised schedule for full compliance and find that it remains responsive to the concerns raised in our Notice of-Violation. We will review the implementation of your corrective actions during a future inspection to N
i determine that full compliance has been achieved and will be maintained.,
Sincerely, OdelnalGigned By:
.
Thomas P.Gwynn Samuel J. Collins, Director Division of Reactor Projects l
cc:
Gulf States Utilities ATTH:
J. E. Booker, Manager-Nuclear Industry Relations P.O. Box 2951 Beaumont Texas 77704 Bishop. Cook, Purcell & Reynolds
' '
ATTN: Mark. Wetterhahn, Esq.
.
1401 L Street, N.W..
L L
Washington, D.C.
20005 i-
,
Gulf States Utilities
'
ATTN: Les England, Director Nuclear Licensing-P.O. Box 220 St. Francisville Louisiana 70775 RIV:RI:PSS C6 C:PS$h D:DR f
- RP LJCa lanP'N p/D/ g/90 CJohnson/cjg
/f TStetka JCoJ11ns to /4 /90 g /0/4 /90 p/4 /90 s
'
'
J
/
9010230fo9 9o2 cog
/
t
,
PDR ADOCK 05000458 l
PDC l
g.
-
y-v;
,c
'
-
t
+
.
.
.
,
..
.!
.-
,
..
+
[
Gulf States Utilities.
1 -2-
,
.
p.
,
'Mr. J. David McNeill, III
,>
.
William G. Davis Esq.
'
Department of Justice.
'l
- Attorney General's-Office P.O.~ Box 94095 i
Baton Rouge, Louisiana-70804-9095-
H. Anne Plettinger
'
,
L 3456 Villa Rose Drive i
'
t
,
L Baton Rouge, Louisiana -70806
'
President of West Feliciana E
Police Jury
-'
P.O. Box 1921 St. Francisville,. Louisiana 70775-
'
,
Cajun Electric Power Coop. Inc.
ATTH: Philip G. Harris 10719 Airline Highway P.O. Box 15540 Baton Rouge, Louisiana 70895 Department of Environmental Quality ATTN: William H. Spell, Administrator Radiation Protection Division P.O.. Box 14690 Baton Rouge, Louisiana. 70898 U.S. Nuclear Regulatory Commission ATTN: Resident Inspector
'P.O. Box 1051 St. Francisville, Louisiana 70775'
U.S~ Nuclear Regulatory Commission
- ATTN: Regional Administrator, Region IV 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011
[
'bectoDMB(IE01)
bec distrib by RIV:-
R. D. Martin Resident Inspector
- DRP; Section Chief (DRP/C)
Lisa Shea, RM/ALF Mls System l
DRSS-FRPS RSTS Operator-ProjectEngineer(DRP/C)
RIV File
-
L DRS Senior Resident Inspector, Cooper l
Senior Resident. Inspector, Fort Calhoun T. Stetta C. Johnson
'
..
i
,
,1 m
.
- ..
,...
-
,
'
- f.-
i..
s e
auzr svares urzzzrres coaerawr Alvth 0END 5t Att0N 0-0$f Of 8lCf DDR 220 St fitANCisvtLLt L0utslANA ?O776 ARE A CODI 604 C35 6364 346 4661
$ $ $ 0Y/[$ %
September 18, 1990
~
RBG 33596 File Nos. G9.5, G15.4.1 SEP 211990
)
'
-
U. S. Nuclear Regulatory Commission
~~
Document Control Desk Washington, D.C.
20555 Gentlemen:
,
River Bend Station - Unit 1 Refer to: Region IV Docket No. 50-458/90-02 Pursuant -to 10CFR2.201, this letter revises Gulf States Utilities Company's (G$U) response dated May 7,
1990 to the Notice of Violation for NRC Inspection Report No. 50-458/90-02.
The inspection was conducted by Messrs.
Johnson, Singh and Murphy during the period of January 22 26, 1990 of
-
activities authorized by NRC Operating License NPF-47 for River Bend Station-Unit 1(RBS).
This letter is being submitted at this time pursuant to a conversation with Mr. L. Constable today.
Revisions to the original response-are denoted in the attachment by change bars in the margin.
'
Should you have any questions, please contact Mr.
L.
A.
England at (504)381-4145.
Sincerely, f
I t
W.
. Odell Manager-Oversight River Bend Nuclear Grcup Attachment cc:
'U. S. Nuclear Regulatory Comiss16n
'
Region IV j
611 Ryan Plaza Drive Suite 1000
>
Arlington TX 76011 i
Senior Resident Inspector Post Office Box 1051 St. Francisville LA 70775-9&ieWo//o ?W-
-0 o
,'
..
.
'
,'
.
'i UNITED STATES OF AMERICA
,
NUCLEAR REGULA'!ORY COMMISSION STATE OF IDUISIANA
)
PARISH OF WEST FELICIANA
)
-
Docket No. 50-458
'
In the Matter of
)
GULF STATES (ITILITIES COMPANY
)
(River Bend Station - Unit 1)
.
'
W. H. Odell, being duly sworn, states that he is a Manager
- Oversight for Gulf States Utilities Company; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the documents attached hereto; and that all such documents are true and correct to the best of his knowledge, information and. belief.
0.
,
.
W.
H. Odell
'
.
Subscribed and sworn to before me, a Notary Public in and for the State and Parish above named, this /8A day of
/M1d-mlHA 1990.
My Commission expires with Life.
,
0AllLLOYL 0. ANAAOY Claudia F.
Hurst Notary Public.in and for West Feliciana Parish, Louisiana
.
--
e
..
' ',,
,o.
-
,
ATTACWENT
'
-
REPLY TO NOTICE OF VIOLATION 50-458/9002-02 (SEVERITY LEVEL III)
,,_
'
REFERENCES Response to Violation - Letter from J. C. Deddens to U. S. NRC, dated May 7,
,
1990.
Notice of Violation - Letter from S. J. Collins to J. C. Deddens, dated April 6, 1990.
Enforcement Conference Summary - Letter from S. J. Collins to J. C.
Deddens, i
dated March 26, 1990.
Notice of Enforcement Conference - Dated March 6,1990.
Inspection Report
Letter from
S.
J.
Collins to J. C. Deddens, dated
'
-
February 26, 1990.
Licensee Event Report No.89-036 - Letter from J. E.
Booker to NRC, dated
I
November 16, 1989, Rev. I dated January 31, 1990.
VIOLATION
Operating License NPF-47
Section C.10., states that GSU shall comply with
the requirements of the fire protection program as specified in " Attachment
4."
'
Attachment
to
Operating License NPF-47,
.
Requirements," states that GSU shall implement and maintain
in effect all
'
provisions of the approved fire protection program as described in the Final
Safety Analysis Report for the facility through Amendment 22 and as approved
in the SER dated May 1984 and Supplement 3 dated August 1985 subject to
Provisions 2 and 3 below (which are not applicable here).
Tables 2 and 5 of GSU design specification 240.201,
" Fire Analysis and
Evaluation Criteria and Evaluation Method Including Results and Conclusions
for 10 CFR 50, Appendix R Fire Hazards Analysis," part of the approved fire
protection program described above,
list motor-operated valves for which
electrical powgr is assumed to be removed during plant operations.
Contrary to the above, from November 1985 to October 1989, GSU did not
implement and maintain in effect all
provisions of the approved fire
protection program in that when River Bend Station was operating during this
period, electrical
power had not been removed from 19 motor-operated valves
!
listed in Tables 2 and 5 of design specification 240.201 as having power
j
removed during plant operations.
Page 1 of 6
- - - - - -.
-.
-
- - - - - - - -.
E,
-
,
'
...
,
,
'
REASON FOR THE VIOLATION
- -
Of the nineteen valves assumed in the FHA to have electrical power removed,
four were listed in the Final
Safety Analysis Report (FSAR),
Section
9A.2.1.2, as high/ low pressure interface valves.
These four valves, plus two
additional valves associated with steam condensing mode of residual heat
removal
(RHR), did have power removed during initial startup.
Two of these
valves, IE12*MOVF009 and 1E12*MOVF040, were subsequently reenergized.
VALVE 1E12*MOVF009
Valves 1E12*MOVF009 and IE12*M0VF008 are the containment isolation valves for
the RHR shutdown cooling mode suction line.
This is a high/ low pressure
interface between the recirculation system and the RHR shutdown cooling mode
piping.
The FSAR required one of the pair to have electrical power removed.
Meeting this requirement of the FSAR also met the assumption for power
removal in the FHA.
In Supplement 3 to the Safety Evaluation Report for River Bend Station,
August 1985, and a GSU letter dated August 6, 1985, a commitment was made to
add a keylock switch in the control circuitry of IE12*MOVF008 to " lock out
(block) control of the valve (E12*F008) from both the control
room and the
remote shutdown panel." The switch was to be installed in the motor control
center (MCC) located in the auxiliary building and was not to disable the
valve position indication in the control room or the remote shutdown panels.
Modification request (MR) 85-0956 was initiated and installed in November
1985 to add the keylock switch to the control circuitry for 1E12*MOVF008.
Both valves were then energized.
During the design for the MR, the engineer
perceived concerns with locating the switch in the auxiliary building at the
MCC.
There were no keylock switches available that could be qualified for
the harsh post accident environment in this area. There was also a concern
about operator access to the switch during a post accident environment. Due
to these concerns, the keylock switch was relocated to the remote shutdown
panel in the control building.
In this location, the keylock switch provided
easy operator access yet still prevented inadvertent opening of IE12*MOVF008
during a transfer of control
from the main control room to the remote
shutdown panel.
The design in the MR and the 10CFR50.59 safety evaluation failed to recognize
the concerns associated with fire exposure and subsequent spurious actuation
of both the IE12*MOVF008 and 1E12*MOVF009 velves. With the keylock switch
located in the remote shutdown panel and the manner in which it was installed
in the control, circuitry, a single fire in either the remote shutdown panel
or the main c6ntroT room could cause spurious actuation of both valves. With
the electrical power restored to both valves after the MR, the assumptions in
.
the FHA were violated.
An inadequate design analysis for MR 85-0956 is considered to be the root
!
cause for violating the assumptions of the FHA as related to 1E12*M0VF009.
Several factors contributed to the inadequate design analysis. An inadequate
i
depth of investigation as part of the design development failed to reveal the
FHA assumptions.
A lack of familiarity with the FHA and no formal training
in the requirements of the FHA on the part of individual
system engineers
Page 2 of 6
i
'
I
e
.
- i
.,
.
,
.
contributed to this oversight of the FHA. Coupled with this was a deviation
-
from the modification procedure.
The modification procedure required a fire
protection checklist to be completed if fire protection issues were affected.
The fire protection checklist was not prepared, and no review by the design
fire protection engineer was performed.
The lack in depth of documentation
in the methods and assumptions used in the FHA contributed not only in the
initial oversight but also in the delay in discovery of the problem.
The
lack of maturity in the engineering organization during the trans.. ion of
responsibility from the architect / engineer to GSV also contributed to the
oversight.
At the time of installation of the MR,
fire
protection
engineering responsibility was divided among GSV Nuclear Plant Engineering,
GSV Technical Staff, the architect / engineer design office, and the architect /
engineer Site Engineering Group.
VALVE 1E12*M0VF040
.j
Valves 1E12*MOVF040 and 1E12*M0VF049 are the system interface isolation
valves between the RHR system and the radwaste system.
This is considered a
high/ low pressure interface only during the steam condensing mode of RHR. A
license condition prohibits use of the steam condensing mode of RHR at River
!
Bend Station.
Due to this, Engineering Evaluation and Assistance Request
,
(EEAR) 87E-0216 was
initiated in May 1987 to evaluate
re-energizing
IE12*MOVF040 since it is not a high/ low pressure interface valve with steam
condensing mode of RHR disabled.
The EEAR was answered in June 1987 with the
required changes to the FSAR and operating procedures to allow energizing the
r
valve.
Included in these operational procedure changes was a revision to
" Shutdown from Outside the Main Control Room". This revision
.<
required verification that 1E12*M0VF040 was in the closed position if the 'A'
.
division of RHR was in shutdown cooling prior to transfer of control from the
main control room to the remote shutdown room.
Although subsequent reviews for separation showed that this situation was
acceptable, the FHA was not revised at the time to delete the assumption of
removing power on this valve.
It is not clear that the FHA and its
assumptions were considered in the evaluation process.
The oversight
associated with EEAR 87E0216 can be attributed to the same root causes as
i
associated with MR 85-0956.
,
REMAINING VALVES
The remaining thirteen valves that had not had electrical power removed as
assumed in the FHA remained energized due to oversight during the original
preparation of operational
procedures
in 1985.
This oversight was most
probably caused by a lack of awareness by the developers of the procedures of
the FHA and* its assumptions.
The FSAR listed only those valves that were
required to have power removed due to
high/ low
pressure
interface
considerations.
The valves listed in the FSAR were proceduralized to have
power removed but those that were only contained in the FHA were overlooked.
Valve 1821*MOVF019 is not a high/ low pressure interface valve but does
represent a potential
loss of coolant path.
Valve IB21*M0VF019 is an
isolation valve for the main steam drain lines. A fire in the main control
room could cause spurious actuation of this valve and the other valves in
series with this valve.
This would allow reactor coolant to bypass the main
,
Page 3 of 6
!..'.
,
.,.
.
.
steam isolation valves directly to the condenser.
Coupled with the Appendix
'-
R-required assumed loss of offsite power, the condenser could be pressurized
causing the rupture of the air relief diaphragms on the low pressure turbine.
Although of minor safety significance, this would represent an uncontrolled
discharge to the turbine building atmosphere.
CORRECTIVE ACTIONS WHICH HAVE BEEN TAKEN AND THE RESULTS ACHIEVED
The immediate corrective action that was taken in October 1989 upon discovery
of the problem was to remove the electrical power from valves
IE12*M0VF009
and
1821*MOVF019.
Power was removed from these two valves due to
inaccessible locations in the drywell and main steam tunnel.
A fire wa tch
was
initiated for the other valves along their control circuitry until the
separation required by 10CFR50, Appendix R could be verified.
By November
13, 1989, the review for adequate separation for those valves was completed
verifying that the necessary separation did exist. During that time period,
the requirenent for removal of power from 1E12*MOVF009 and IB21*MOVF019 was
verified since adequate separation did not exist-for these potential loss of
coolant paths. Adequate Appendix R separation does not exist in the main
control
room for either valve and does not exist in the remote shutdown room
for IE12*MOVF009.
The verification of divisional separation for thirteen of
-
the valves and removal of power for two of the valves, along with the four
valves which have had power removed since 1985, put the plant in a condition
that was
in compliance with the basis of the FHA for these valves.
90-0003 was issued on January 25, 1990 to revise the FHA to reflect the
current status of the valves in the plant.
As pa rt of the corrective action,
Engineering Analysis performed safety
-
assessments of the spurious opening of 1821*MOVF021 and 1E12*M0VF009 due to
fires
in the main control room and the remote shutdown panel.
(Note that
IB21*MOVF019 was open with the downstream valve IB21*H0VF021
closed.
"
Therefore,
the safety assessment for containment bypass via 1821*MOVF019
focused on the probability of spurious actuation of 1821*H0VF021 due to fire,
to create an open bypass pathway.) Details of these assessments are provided
in the referenced Licensee Event Report.
The probabilistic risk assessment (PRA) for 1821*MOVF021 indicated that the
probability for a steam release from the condenser was approximately 1.9 E-04
over the time the valve was energized.
The radioactivity releases from this
event
were
determined to remain below 10CFR20 and 10CFR100 limits.
Therefore, the safety significance of this event is low.
The PRA for IE12*M0VF009 examined the likelihood of an interfacing system
LOCA and estimated the core damage frequency (CDF) for this event as 5.8E-08.
,
l
This is a factor of 100 below the total CDF of 5.0E-06 for RBS.
Therefore,
~
j
the safety significance of this event is also low.
Due to the heightened awareness of the FHA and the lack of incorporation of
specific requirements associated with the valves, GSU Quality Assurance
Februa ry
7,
1990 a Safety System Functional
performed from January 1
-
Inspection (SSFI) of the FHA as related to the energized valves.
The SSFI
l
identified several
recommendations for operator actions from the FHA that
l
l
Page 4 of 6
i
- '
.
,,....,
'
were not reflected in the plant Prefire Strategies. The SSFI also identified
two instances where the necessary electrical jumpers were not available for
potential fire-induced repairs required for equipment necessary to
achieve
cold shutdown.
The affected Prefire Strategies were revised by March 8, 1990, to add the
reconcendations for operator action from the FHA.
The electrical jumpers and
work packages for the repairs necessary to equipment for cold shutdown were
fabricated and staged by January 26, 1990.
An initial review of the FHA by Design Engineering was completed in January
1990 to verify the consistency of the existing design and operational
procedures. This review was done in conjunction with review of the Prefire
Strategies to ensure all actions or plant conditions assumed in the FHA were
contained in the Prefire Strategies or other plant procedures.
No other
inconsistencies other than those already detailed were identified.
In addition to the actions taken to correct the sp;cific condition with the
valves and FHA, additional progrannatic actions have been taken over the last
few years.
In 1987, responsibility for fire protection engineering was
consolidated in Design Engineering.
This minimized the potential for errors
due to confusion over engineering responsibility.
Procedural compliance has
improved throughout River Bend. The need for procedural compliance has been
emphasized
to
all managers and supervisors.
The Design Engineering
supervisors review and evaluate each QA unsatisfactory finding (unsats) and
Quality Assurance Finding Report assigned to Design Engineering.
The results
of those evaluations are discussed in Design Engineering staff meetings to
determine if trends in unsats are developing and to correct those trends
early.
This has resulted in a significant decrease in the number of unsats
generated against Design Engineering documents.
The modification procedure has been revised to require increased depth of
design bases evaluation and documentation.
This will help preclude an
oversight of the FHA and its requirements in the future.
CORRECTIVE ACTIONS WHICH WILL BE TAKEN TO AVOID FURTHER VIOLATIONS
Although the corrective actions that have been taken to date bring the plant
into a state of full compliance with the operating license, additional
corrective actions are necessary to ensure a similar situation does not occur
in the future.
The corrective actions are separated into three areas:
the
FHA and associated procedures, modification requests, and training.
As stated above, an initial review of the FHA has been performed. A final
review and veH fication of the FHA will
be performed by an independent
contractor.
In addition,
the independent contractor is to provide fully
detailed documentation of the design bases and assumptions of the FHA.
-Additional
verification of the consistency between the FHA and plant
procedures will be perfonned by the independent contractor.
This will be
followed by another SSFI performed by GSU Quality Engineering to evaluate
implementation and effectiveness as outlined in the FHA.
To ensure that no additional modification requests with similar oversights
exist, a review of MRs engineered from the time GSU assumed control of the
Page 5 of 6
e
.....
......
....
,_
f.,;... '.,,
'
.
.-
,
.
.
.
' ' '
L
design modification process to January 20,
1987 will be perfonned.
This
review will be done to ensure that adequate documentation exists for
potential impact on the FHA.
After January 20, 1987,the fire protection
f
checklist was required to be completed for all MRs.
If the review of MRs
engineered prior to that date indicates the problem may extend beyond January
20, 1987, the scope of the review will be increased.
In order to increase the general awareness of the Fire Hazards Analysis and
t
its requirements, a training program on the FHA is to be developed.
The
training program will be provided to all engineers who perform modification
requests and safety evaluations.
In addition to the engineers, the members
of the Facility Review Committee and appropriate operations personnel will be
given training on the FHA.
The training for the operations personnel will
include the recommended operator actions that are included in the FHA.
In addition to the corrective actions that are being done to prevent a
recurrence, an investigation is being pursued to allow operations to energize
1821*MOVF019 during startup phases of plant, without continuous opera tor
attendance at the valve's MCC.
This investigation is evaluating the amount
of time that would be required to pressurize the condenser and rupture the
air relief diaphragms with the reactor at various power levels and pressures.
These times will be evaluated to determine at what pressure level or power
level adequate time is available for ensuring isolation of the main steam
drain lines
in the event of a main control
room fire.
Until
the
investigation is completed, this valve will remain under current controls.
DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED
As of Ma rch
8,
1990, with the issuance of the revised Prefire Strategies,
River Bend Station was in compliance with the Fire Protection Program as
required by its operating license.
Further corrective actions will be
accomplished per the following schedule:
The contract has been awarded to NUS Corporation for the FHA review and
-
documentation,
the proposed schedule requires the work to be complete by
January 15, 1991.
Review of the MRs will be complete by February 28, 1991.
-
The follow-up SSFI to evaluate the implementation and effectiveness of
-
revised procedures regarding the FHA will be performed by July 1991.
Implementat, ion of the training program will be complete during the second
-
quarter 1991'.
'
i
l
Page 6 of 6
l
I
\\
l
l
_