IR 05000269/1974001
| ML19308A640 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 02/20/1974 |
| From: | Jape F, Murphy C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19308A639 | List: |
| References | |
| 50-269-74-01, 50-269-74-1, NUDOCS 7911270815 | |
| Download: ML19308A640 (21) | |
Text
6
.
.
.
- - -
-
v RO Inspection Report No. 50-269/Th-1 pgg g y ggy
<
DUKE PO'4ER COMPANY Oconee 1 DISTRIBUTION:
H. D. Thornburg, RO RO:HQ (5)
D-
- y Central Files Regulatory Standards (3)
Directorate of Licensing (13)
cc encl. only:
,
- PDR
- Local PDR
- NSIC
- DTIE, OR
- State
- To be dispatched at a later date.
.
,
_
-.
F'M' 1 UNITED STATES
-
.
[
d. 't ATOMIC ENERGY COMMISSION
,
' * C"'E'E I.) *
-
DIRE *.1'%*J."M OF Rr;GULiTCPJ( OPERATICL3
'
acc,ios si - suit E e's
,, ",] k. ]
'i 2 ). In t A f. pe f R S, E $f A L E T, NOR T te gL %T
,
,,,
A T t. A e. r A G E QN G e A DJ 303
<
RO Inspection Report No. 50-269/74-1 Licensee: Duke Power Co=pany Power Building 422 South Giurch Street Charlotte, North Carolina 28201 Fde: Oconee Unit 1
< Docket No.: 50-269
'
License No. : DPR-38 Catego ry : B2 Location:
Seneca, South Carolina Type of License:
Type of Inspection: Routine, Unannounced Datea cf Inspection: January 22-25, 1974 Dates of Previous Inspection: Novecber 27-30, 1973 and December 3-7, 1973
.
Principal Inspector:
F. Jape, Reactor Inspector Facilities Test and Startup Branch Accompanying Inspectors :
K. W. Whitt, Reactor Inspector Facilities Test and Startup Branch T. N. Epps, Reactor Inspector Facilities Test and Startup Branch 9W 2'/d'7Y Principal Inspector:
ff G M.
>
F. Jape, Reactor Insp(ctor'
Date Facilitie Test and Startup Branch Reviewed 3y:
<E r
2
C. E. Murphy ( Chd6f [
' Date
.
4
,. _, _
p.
,e,.-w.
- " *
.
..
.
.
l
,RO Rpt. No. 50-269/74-1-2-
!{^3
-
'
.
..
L SUMMARY OF FINDINGS
,
I.
Enforcement Action A.
Violations The follcwing apparent violations are considered to be Category II severity :
a.
Removal of RC-9 Removal of RC-9, " Pressurizer Spray Line Manual Block Valve,"
from service was partially completed before the Station Review Committee (SRC) reviewed the station modification as required by Administrative Procedure No.10 " Station Modifications."
Failure to adhere to established procedures for accomplishing station modifications appears to be in violation of Criterion V of Appendix B to 10 CFR 50.
This violation was reported by the Station Quality Assurance i
Group on January 5, 1974.
(Details II, paragraph 3.e)
'
b.
Incomplete Power Runh-ack Following a Failed Rod On November 20, 1973, reactor power was run back to 59%
,
rather than 55% as required by EP 3800/21, " Inoperable Control Rod."
l Failure to follow established procedures appears to be in violation of Criterion V of Appendix B to 10 CFR 50.
(Details III, paragraph 2)
c.
Withdrawal of Rods During an Unscheduled Power Reduction On November 20, 1973, control rods were withdrawn during an unscheduled power reduction in an attempt to increase reactor power.
Operating Procedure 1102/04, " Operation at Power,"
prohibits such action.
Failure to adhere to an established
,
procedure appears to be in violation of Criterion V of i
Appendix B to 10 CFR 50.
(Details III, paragraph 3)
d.
Loss of Level Instrumentation for the BWST
.
On December 22, 1973, both channels of level instrumentation for the BWST became inoperable.
The loss of this instrumentation
.
is a violation of Technical Specification 3.3.1(f).
!
!
\\
l
.
.
.
. _ _~~
,..n...-_,-.
.n.-
-~ ~. - -~
'~'~~~r
' * ' ' * * ' ' * ' ' '
'*' ***
" '
~~#
,.
.
_,, -
--
-
r *
_
_
_
__ _
.
.
IK) Rpt. No. 50-269/74-1-3-
,
- -
.
This event was reported to RO:II on December 22, 1973, and DPC has issued Abnormal Qccurrence Report No. A0-269/73-8, dated January 4,1974, describing the event and corrective actions taken.
(Details I, paragraph 5)
e.
Momentary Inoperability of Reactor Building Spray System On Dececher 8,1973, both reactor building spray pumps were removed from service at a time when Technical Specificatiens require at least one crerable.
The removal of botn Reactor Building Spray pumps from service is a violation of Technical Specifications 3.3.1(a).
This event was reported to RO:II en Dece:ber 8,1973, and DPC has issue,d Abnormal Occurrence Report No. A0-269/73-9, dated December 20, 1973, describing the event and actions taken.
(Details I, paragraph 6)
B.
Safety Ite=s
!
None II.
Licensee Action on Previously Identified Enforcement Matters A.
Violations The licensee's response, dated January 25, 1974, regarding the violations identified in RO Inspection Report No. 50-269/73-13 is currently being reviewed by RO:II.
B.
Safety Ite=s None III.
New Unresolved Ite=s
)
None IV.
Status of Previously Reported Unresolved Ite=s 73-13/1 Wastewater collection Basin Modification No change. The enlargement of the basin to increase capacity and holding time has not yet been implemented.
'
.
'w'
~ _....- -.... - ~.
,q
.pp
,
_ %g s iu, g s-s
+-
eee
- --- esse - w - e
"~'*%'
- -..
_
_
-
,-
-
r,
__.
_
_
.
.
,RO Rpt. No. 50-269/74-1-4-l ~)
-
.
,
73-13/2 Control Rod Failures The licensee is reviewing the control rod failures and a
!
report is expected from the NSS.
73-12/1 Calibration of Ef fluent Monitors No change in status. The analysis of data to determine the correlation between monitor indication and sample analysis is continuing.
71-7/1 Thin Walled Valves (R0 Report No. 50-269/71-5,Section II, Paragraph 3 and 50-269/73-10, Details III, paragraph 2)
Region II has reviewed the licensee's reports concerning the power operated relief valve and comments have been submitted to DPC.
Supple = ental information from DPC, dated January 22, 1974, has been received by Region II and is currently under review by Region II inspectors.
V.
Unusual Occurrences A.
UE-269/73-12 " Engineered Safeguards Valve FDW-108" Corrective actions described in DPC's report, dated December 7, 1973, were verified during this inspection and the inspector has no further questions on this item.
(Details 1, paragraph 2)
B.
UE-269/73 13, " Engineered Safeguards Valve LP-17"
The corrective action described in DPC's report, dated January 4,
,
1974, was verified during this inspection and the inspector has no further questions on this item.
(Details I, paragraph 3)
C.
UE-269/73-14, " Engineered Safeguards Valve RC-7"
!
Corrective actions described in DPC's report, dated January 7, 1974, were verified during this inspection and the inspector has no further questions on this item.
(Details I, paragraph 4)
,
I D.
A0-269/73-8, " Borated Water Storage Tank Level and Temperature Instrumentation" Corrective actions described in DPC's report, dated January 4, 1974, were verified by the inspector.
There are no further
-
questions on this item.
(Details I, paragraph 5)
,
- .
,e-m, n
e'
W
.
.
RO Rpt. No. 50-269/74-1-5-h[') * *
.
.
I
E.
A0-269/73-9, "1bmentary Inoperability of Reactor Building Spray
.
System"
.
,
Corrective actions described in DPC's report, dated December 20, 1973, were verified by the inspector during this inspection.
There are no further questions regarding this item.
(Details I, paragraph 6)
VI.
Othi Significant Findines Personnel and Organizational Changes R. M. Koehler, previously Technical Support Engineer, is now a Staf f Engineer and is Chairman of the Station Review Committee.
O. S. Bradham, previously Instrument and Control Engineer, is now Technical Support Engineer.
R. C. Adams, previcusly Instrument and Control Supervisor, is now Instrument and Control Engineer.
J. M. Davis, previously a Staff Engineer, is now the Maintenance Engineer which is a new position in the organization.
The Technical Specifications have been revised and approved on January 24, 1974, to reflect the above changes.
VII.
Management Interview The management interview was held on January 25,1573, with J. E. Smith, R. M. Koehler, L. E. Schmid, O. S. Bradham, and J. W. Cox, in attendance.
During the meeting, enforcement matters, described in Section I, Su= mary of Findings, identified during this inspection were discussed.
The licensee was notified by telephone on January 31, 1974, of Region II's decision to identify item I. A.c., " Withdrawal of Rods During An Unscheduled Pcwer Reduction" as a violation.
This matter was discussed during the meeting but final decision was reached after further review.
(Details I, paragraphs 5 and 6, Details II, paragraph 3.e and Details III, paragraphs 2 and 3).
The inspector also reviewed his findings relative to three unusual events (Details I, paragraphs 2, 3 and 4) and the findings regarding the process radiation monitoring system.
(Details I, paragraph 8)
/
.
'
--
-
..._, _
,..-
"~w~~~~~-
+
m-wa
.,,., ___
_
_
..,. _ _
_ __-
- _ __._. _.
_ _ _ _. _ _ _ _ _ _
-_ -
.
.
,
Rd Rpt. No. 50-269/74-1 I-1
,
-
,,
s DETAILS I Prepared by:
4d4A
%
OY N
F. Japd, Reactor Inspector Date Facilities Test and Startup Branch Dates of Inspection: January 22-25, 1974
/
& 79'
Reviewed by:
/
C. E. Murphy, M isf (/
/ Date Facilities Test and Startup Branch 1.
Individuals Contacted Duke Power Company (DPC)
J. E. Smith - Plant Superintendent J. W. Hampton - Assistant Plant Superintendent R. M. Koehler - Staff Engineer D. Rogers - Junior Engineer E. E. Hite - Junior Engineer
'
O. S. Bradham - Instrument and Control Engineer D. J. Rains - Assistant Plant Engineer L. E. Schmid - Operating Engineer W. Morgan - Control Operator B. Smith - Control Operator 2.
UE-269 /73-12, " Engineered Safeguards Valve FDW-108" On November 8,1973, while performing the monthly periodic test, IP 310/13C, "RB Emergency Cooling and Isolation Logic Ch-6 On-line Test,"
engineered safeguards valve FDW-108 failed to close.
FDW-108 is a pneumatically operated steam generator sample line isolation valve which is normally maintained open. A electrically operated isolation valve, FDW-107, is provided as a redundant valve.
FDW-107 was tested on Nove=ber 8,1973, and operated as designed.
To correct the problem with FDW-108, the solenoid was replaced and the valve has performed sotisfactorily on subsequent periodic tests perfor=ed on December 6,1973, and January 7,1974.
The Station Review Committee reviewed this incident on December 19, 1973, and concluded that the corrective acticns taken were adequate. The inspector had no questions on this item.
_
W eam ee nw am e
.n w wwwe m,
,.we
.
-
h wy.t s4.-.-
'
, _ _,._ _,...
.
.
--
.
-
. -
--
- - - - -
-- - --
- "
i '
-
hi l F R0 Rpt. No. 50-269/74-1 I-2 3.
UE-269/73-13, " Engineered Safeguards Valve LP-17"
,
On December 3,1973, while performing the =onthly periodic test, IP 310/12B, "LPI Logic Ch-3 On-Line Test," the control power fuse blew causing 1-LP-17 to stop in travel.
An investigation of the incident revealed that the motor overloads had not tripped nor was there any evidence of loose, burned or shorted wires associated with the valve.
The fuse was replaced and the valve operated successfully on five successive tests. The valve also operated properly on a subsequent periodic test completed on January 7,1974.
.
This incident was reviewed by the Station Review Committee on January 3, 1974.
The committee concurred with the corrective actions taken.
The inspector has no further questions on this item.
4.
UE-269/73-14, " Engineered Safeguards Valve RC-7" on December 8,1973, while performing periodic test PT 150/15, "RB n
.
Isolation Valve Exercise Test," engineered safeguards valve RC-7 failed to open.
RC-7 is a pneumatically operated reactor building isolation valve in tne pressurizer sample line.
The valve closed as required but would not reopen when returning to normal position after completion of the tcst.
Investigation revealed that the valve was jammed in its ceat and hence would not open remotely.
Af ter manually unseating the valve, it opened j
properly.
An adjustment to the valve operator was made to prevent it from jamming.
The Station Review Committee reviewed this incident on January 3,1974, and recoc= ended that the valve be operated daily for five consecutive days to demonstrate operability.
The inspector reviewed the test results which indicated proper valve operation on each test.
The inspectar has no further questions on this incident.
5.
A0-269/73-9 " Borated Water Storage Tank Level and Temeerature Instrumentations" On December 22, 1973, the BWST level and temperature indications became inope rab le.
Investigation revealed moisture in a short, exposed instru-ment air supply line had frozen, causing the line to become blocked.
.
e,
/
~~~
j
_:_ _.
. - -.
-..
. -. - -
- - ~
- - - -
.
. - ~ - - ~ ~ "
.
.
i
'. ~ g R0[Rpt. No. 50-269/74-1 I-3
..
Corrective actions included thawing the line and clearing it of mositure, checking the instrumentation for prc'per operations and insulating the exposed line.
The loss of the BWST level instrumentation is a violation of Technical Specification 3.3.1 (g). RO:II was notified of the incident on Dece=b er 22, 1973, as required by Technical Specification 6.6.2.
The Station Review Committee reviewed this incident on January 3,1974, l
and recommended that a Station Modification Report (SMR) be prepared to enclose the exposed piping.
EMR 0-111-S was approved on January 24, 1974, by the station superintendent and is currently being reviewed by the engineering department.
The inspector verified the corrective actions and had no further questions on this item.
6.
A0-269/73-1, " Momentary Inoperability of Reactor Buildine Spray System" On December 8,1973, while performing PT 150/15A, "R3 Isolation Valve Exercise Test," both reactor building spray pumps were out of service for approximately 15 minutes.
l Immediate corrective action was taken to restore the equipment to se rvice.
In addition, a cautionary statement has been added to periodic test procedures to ensure against recurrence of this incident.
This incident was reported to R0:II on December 8, 1973, as required.by
Technical Specification 6.6.2.
The incident is considered to be a violation of Technical Specification 3.3.1(a).
The Station Review Committee reviewed this incident on December 19, 1973, and concurred with the corrective actions taken.
The inspector verified the corrective actions taken and had no further questions ou this item.
7.
Status of Preoperational Testing The inspector reviewed the status of all test procedures and found j
four tests that have not received final approval. These are:
a.
TP 115/5, "HEPA Filter Inspection Test" The test has been completed with no deficiencies noted and is
-
currently awaiting superintendent's approval.
The licensee's representative indicated that approval should be granted within two weeks.
'
,N I
i
.
L
_
'-
-
-
.
.
.-
-.. -
'
_._c-
_.
_ __
.. - _ _. _. _. - _ _.
~. ~ _
..._
._.
_, -
. _. _.
. _ _ _ _. -
t
.
RD Rpt. No. 50+269/74-1 I-4-3
.
,,
- !.
I b.
TP 170/35A, " Turbine Bypass Valve Test" and TP 270/36, '._' Turbine
.
}
Sypass Valve Booster Test"
"
'
l
-
These two tests have been completed with no deficiencies noted and are currently being reviewed by the plant superintendent for final
4 approval.
Final approval is expected within several weeks.-
i
c.
TP 330/6, " Testing CRD Stators" Test has been coupleted and the QA Group is currently auditing the test. Final approval should be granted within several weeks.
,
8.
Process Radiation Monitoring System
RO:II has received information II that at least one third of the I
radiation monitors were " tripped" during a tour through the Control
Room on Saturday evening October 27, 1973.
Furthe-it was alleged
that these monitors are "... reset before the AEC comes around."
l
'
The inspector revieved the process radiation =onitoring system by:
Examining results of periodic tests, a.
b.
Discussing the equipment performance with operators and
supervisors,
i Discussing the maintenance and t(sting of the equipment with'
c.
the instrument personnel, and d.
Reviewing records to determine if gaseous or liquid waste was being processed at the time of the tour.
(a) Periodic Tests The licensee has provided 'IP-360/15, " Functional Test of RIA-31 through RIA-51," to ' periodically check all process radiation monitors.
. The test was performed on the following dates and the inspector reviewed the test results:
February 2,19'73 August 6, 1973
March 5, 1973 Sep tember 10, 1973
April 4, 1973 October 2, 1973 May 5, 1973 November 8,.1973
.
1/-Letter from C. M. Parrott, III, to N. C. Moseley, December 20,.1973.
.
. w.
.
S
~mo
-
'
-
-. -.. -... _. -..
.
...
.
.-
-. -
.
.-
.- ~
,
RO Rpt. No. 50-269/74-1 I-5
, '3
.
,
,
)
June 8, 1973 December 1, 1973
<
July 9, 1973 In each case the test results indicated proper response of the nonitor to the radioactive check source.
The inspector had no questions or comments on these tests.
The licensee has also provided the following calibration procedures.
Each of these have been performed on the date indicated and the l
"
results reviewed by the inspector.
IP O A 0360 01 A
"RIA-43 Vent Particulate Monitor" 2-1-73 IP 0 A 0360 01 B
"RIA-44 Vent Iodine Monitor" 7-24-73 IP 0 A 0360 01 C
"RIA-45 Vent Gas Monitor Low" 6-15-73 IP 0 A 0360 01 D
"RIA-45 Vent Gas Monitor High" 6-15-73 IP 0 A 0360 02
"RIA-50 Component Cooling System Monitor" 6-16-73 IP 0 A 036'O 03 A
"RIA-47 RB Particulate Monitor" 9-28-73 IP 0 A 0360 03 B
"RIA-48 RB Iodine Monitor" 10-2-73 IP 0 A 0360 03 C
"RIA-49 RB Gas Monitor" 9-27-73 IP 0 A 0360 04 A
"RIA-33 LW Disposal Monitor" 10-3-73 IP 0 A 0360 04 B
"RIA-34 LW Disposal Monitor" 10-3-73 IP 0 A 0360 04 C
"RIA-37 Waste Gas Disposal Monitor 10-5-73 IP 0 A 0360 04 D
"RIA-38 Waste Gas Disposal Monitor" 11-2-73 IP 0 A 0360 05
"RIA-40 Air Ejector Vent Monitor 10-24-73 IP 0 A 0360 06
"RIA-36 Reactor Coolant Monitor" 11-6-73 IP 0 A 0360 08
"RIA-31 LPSW Monitor" 11-1-73
.
'
e
-
.
e
--
-_;
k L--.
.
._. -. -.. A
.
_,. -.
-.
,. - -. -. - -
., -
- - _.
.
i l~'T
.( J :'
R0 Rpt. No. 50-269/74-1
.I-6 t
IP 0 A 0360 09
"RIA-32< Aux Bldg Gas Monitor" 11-1-73
'
IP 0 A 0360 10
"RIA-35 LPSW Disch Monitor" 11-7-73 IP 0 A 0360 11
"RIA-39 Control Room Monitor" 11-29-73 IP 0 A 0360 12
"RIA-41 Spent Fuel Bldg Gas Monitor" 11-30-73 IP 0 A 0350 13
"RIA-42 Recirc Cooling Water Monitor" 11-30-73 IP 0 A 0360 14
"RIA-51 Pent Room Monitor" 8-22-73 The inspector had no com=ent or questions on these procedures or the results.
!
(b) Equipment Performance The operation and performance of the RIA's was discussed with reactor operators and supervisors. They indicated that the equipment has performed without major difficulty.
Spurious alarms have not been a problem. A suggestion from operations personnel to revise the Statalarm system associated with the RIA's is currently being con-sidered.
This idea would provide an annunciation reflash unit to i
permit an alarm to be received, acknowledged and would reset the Statalarm so that another alarm could audibly be signaled with the first one still in the alarm condition. This suggestion was submitted on Station Problem Report 190, dated July 30, 1973, and on January 10, 1974, the necessary documentation was being processed to make the installation for Units 1, 2 and 3.
.
(c) Equipment Maintenance
,
iDie system design features were described by the cognizant instrument
.
and control engineer.
In particular, the indicating lights and their function were described.
i l
Each RIA has three indicating lights.
l l
1.
Green - This light is normally on and indicates proper operation.
An off light indicates power failure.
'
<
L
~
u
.
~
j
,
a
_..,
.>
-
w
+
,y.-
.1
--
-.~.
.
.
l
.
I-7
'
. ~g r1 Rpt. No. 50-269/74-1
.
,
fa
!
2.
Red - This light comes on at the high alarm setpoint. When-ever liquid or gaseous waste is being processe'd, this light most likely will be on.
3.
Amber - This light comes on at the high alert setpoint. When this light comes on certain interlock functions take place on RIA-33, 34, 37, 38, 45 and 49.
The licensee's representative discussed problems related to RIA-33, 34 and 37 which are currently under study.
The problem with RIA-33 and 34 has to do with the monitors being located in a low point of the liquid waste discharge piping.
This location prevents the liquid waste from completely draining from the monitor with the result that the monitor is continually exposed to the waste. The resolution being considered is to relocate the monitors to a high point of the piping and to provide a means for flushing the monitors.
~'
j t
RIA-37 (and RIA-38) monitors vaste gas effluent from Units 1 and 2.
l The holding time for gases in the decay tanks is less than originally planned and has resulted in a higher activity of the gases being monitored by RIA-37.
The readings are often beyond the range of RIA-37 and fall in the range of RIA-38, which is the "High" monitor.
And the red light on RIA-37 is usually "on" when waste gas is being processed. The licensee is currently considering the possibility of extending the range of RIA-37 by 2 decades so that readings can be obtained on the normal monitor.
(d) Records Review To determine the status of the RIA's during the time that the monitors were alleged to be " tripped" the computer printout record for the period of 5 to 7 P.M. on October 27, 1973, was examined.
Also the Reactor Control log was examined to determine if any waste was being processed at this time.
The review revealed the following:
(1) No liquid or gaseous waste was being released during this period.
(2) The reactor building was being purged during this period. RIA-43, -44, -45 and -46 monitors the vent for radioactive air particulates-gas and iodine. - The computer log for this period
. indicated an alarm on'RIA-45.
The red light was probably on.
(3) The ecmputer. log for this period also indicated an alarm on the
._
fo1. lowing :
&
.
O
'
a
-.
s o,,e.
,
-, ww p-
--a
- - - -
..,.
...-
,
. _ _.-
.
__
.
.
,
RO Rpt. No. 50-269/74-1-I-8
--
..
>
a,.
RIA-37, Waste Gas Effluent Monitors This monitor remains activated for a period of time folicwing
'
a gaseous release due to waste gas trapped in the piping.
Its
'
red light may have been ca.
b,.
RIA-48 Reactor Building Iodine Monitor.
- This monitor uses a fileer cartridge and when the filter needs to be replaced, a red alarm may be indicated due to the accumulated activity.
c.
RIA-32 Auxiliary Building Air Monitor.
This monitor may be activated by the reactor building purge, depending on weather conditions.
During this period a slight (one degree) temperature inversion was indicated on the weather tower temperature chart at about 5:00 p.m. and got progressively greater throughout the evening.
d.
RIA-36 Reactor Coolant Activity Reactor coolant letdown stream is monitored by RIA-36. The activity in this reactor coolant steam is sufficient to cause the red alarm light to come on.
In summary, it appears that at least five red lights were on during the period for e:cplainable reasons. 17pe inspector, on unannounced visits the control room, has observed various red lights being on, and there
{
te i
was no attempt by any operator or supervisor to conceal this condition.
The presence of red and yellow lights is considered to be routine operation and provides necessary information regarding plant status.
l l
'
.
s e
oh.mw.
-
+-.
.
.
.
.
'( ])
- '
-
RO Inspection Report No. 50-269/74-1 II-1 DETAILS 11 Prepared by:
d.k 2!S/' 7(4
,
K.' W. Whit t, Reactor Inspector Date i
Facilities Test and Startup Branch i
Dates of Inspection:
January 15-18, 1974 Reviewed,by: [ / u :/
- le/7V aief6/I
' Dste C.E.Murphk,[tandStartupBranch Facilities Tes 1.
Individuals Contacted Duke Power Company (DPC)
J. E. Smith - Plant Superintendent J. W. Hampton - Assistant Plant Superintendent R. M. Koehler - Staff Engineer J. W. Cox - Assistnat Plant Engineer R. J. Brackett - Junior Engineer J. M. Davis - Fbintenance Engineer L. E. Schmid - Operating Engineer
>
2.
Steam Production Quality Assurance Staff Four reports of audits relevant to Unit 1 that had been performed by the steam production quality assurance staff were reviewed.
The reports were:
a.
Audit Report No. 0-5, " Power Escalation of Unit 1 and Other Safety Related Activities," conducted August 23-24, 1973.
b.
Audit Report No. 0-7, " Station Security," conducted September 19, 1973.
c.
Audit Report No. 0-8, " Station Security (Reaudit)," conducted October 2, 1973.
d.
Audit Report No. 0-12, " Health Physics," conducted January 8-10, 1974.
.
.
r=ee.-e 4- _, mg m &
s.-
=e,.sm
...
4erw 7...
.
p
.%cc
.
..
N
,
..
-
_
.
.
.
-
_._
-
..
('} ;-
R0 Rpt. No. 50-269/74-1 II-2 l
The audits were conducted in accordance with written audit procedures and the results were documented by audit checklists, remark sheets, and performance sheets as' appropriate. Memorandums from various members of the station staff to the plant superintendent reporting action taken in response to the audit reports indicate that reported deficiencics are being resolved and corrected.
It appears that the requirements of Duke Power Company's Administrative Policy Manual for Operational Quality A'ssurance and of Criterion XVIII of Appendix B to 10 CFR 50 are being satisfied.
3.
Station Quality Assurance Group The inspector reviewed five station quality assurance group audit reports as follows:
,
a.
Audit of Engineered Safeguards Equipment Operability for Unit 1, conducted October 23, 1973.
b.
Audit of Procedural Implementation of Changes in Technical
!
Specifications from Unit 1 to Units 1 and 2, conducted November 28, 1973.
c.
Audit of Health Physics Records of Gaseous and Liquid Radioactive Wasta Releases, conducted November 28, 1973.
l d.
Audit of Instrumentatior. and Maintenance Spare Parts, conducted December 19, 1973.
e.
Audit of Station Modification 0-92-S, Replacement of RC-9 (Pressurizer Spray Line Manual Block Valve) with a straight Run of Pipe, conducted January 5, 1974.
,
The audits were performed using checklists and the results were documented by the checklists, audit reports, and me=orandums.
It appears that the station quality assurance group is functioning in accordance with the require =ents of the Administrative Policy Manual for Operational Quality Assurance and the Station Technical Specifications.
,
!t
.
.
g.
,.4 m,e-
~
=
a s
'
_._
_
-
'
l
.
.
i i
.'l
'/
R0 Rpt. No. 50-269/74-1 II-3
'
The report of the audit of Station Modification 0-92-S indicated that a portion of the modification was performed be#cre the station review committee reviewed it as required by the Administrative Policy Manual for Quality Assurance and Administrative Procedure 10, Station Modifications.
Further the report indicated that part of the modification was made without written procedures.
This action appears to be in violation of Criterion V of Appendix B to 10 CFR 50.
The station review committee reviewed and approved the modification on January 5, 1974. A procedure was prepared for performance of the work. To prevent sinilar occurrences in the future, the station superintendent wrote a memorandum to the station supervisors centaining instructions that all safety related modifications must have the necessary reviews and approvals along with necessary instructions, procedures, and drawings prior to initiation of the work.
The memorandum stressed that the reivews and preparation of the work must be met regardless of the delays in unit operation that may rer210.
This corrective action and action taken to prevent recurrence were reviewed by the inspector while on site.
This matter is considered i
closed.
4.
Borated Water Storage Tank (BWST) Overflow System (Reference lettee dated July 19. 1973 frem R. C. DeYoung to Duke Power Company and letter dated September 6,1973 f rom A. C. Thies to R. C. DeYoung on aaove subject)
In the September 6, 1973, letter referenced above, Duke Ptaer Cc=pany stated tnat the BWST overflow system would be modified to route any overflow through the pipe trench drain to the low activity waste tanks.
This modification was co=pleted for Unit 1 on January 17, 1974 No furthc. inspection effort is planned for this item.
.
m )
.
.
].
'R0 Rpt No. 50-269/74-1 II-4
.
5.
Power Escalation Tests
<
The results of four power escalation tests were reviewed. The inspector had no comments as a result of this review.
The test procedures reviewed are listed below.
a.
TP/1/A/800/6-1, " Neutron Noise Analysis" b.
TP/1/A/800/18-4, " Induced Power Oscillation Test" c.
TP/1/A/800/28-1, "L' nit Load Steady State Test" d.
TP/1/B/800/30-4, " Loss of Control Room"
,
e-
-me,
.
,e,,v_,.
s Y
u._.____..
_
_
.)
- -
-
.
.
--.
-
,,!
,
RO Rpt. Ns. 50-269/74-1 III-l
"
.m
.
..
'
[.)
'
]
/j A-2 ~ 7 ~ 74~
DETAILS III Prepared by:
-
T. N. Epps, Reactor Inspector,
Date Facilfties Operations Branch
,
Dates of Inspection: January 22-26, 1974 e2-//- 7'/
Reviewed by:
-
'
W. C. Seidle [ Chief, Facilities Date Operations 3 ranch 1.
Persons Contacted Duke Power Company J. E. Smith - Plant Superintendent J. W. Hampton - Assistant Plant Superintendent
R. L. Wilson - Perforr:ance Engineer M. Sample - Junior Engineer i
O. S. Bradham - Technical Support Engineer R. M. Koehler - Staff Engineer
L. E. Schmid - Operating Engineer R. Knoerr - Assistant Plant Engineer D. Rogers - Junior Engineer 2.
Dropped Control Rod On November 20, 1974, Unit 1 was operating at 100% power.
At 2:47 p.m.
control rod No.1 in group 7 dropped into the core.
j Section 6.4 of the Technical Specifications requires that the station
'
be operated and maintained in accordance with approved procedures.
Criterion V of Appendix B to 10 CFR 50 requires that activities affecting
quality he prescribed by documented procedures and be acco plished in i
accordance with these procedures.
An emergency procedure for one assymetric rod, EP/0/A/1800/21, calls for verification.that an. automatic runback to 55% power occurs when this condition is. experienced.
If automatic runback does not occur, then the procedure requires that the reactor be run back to 55% power by the-operator.
Contrary to the above, after a 5% runback in power the reactor.was placed in manual control and reactor power was run back to only 59% power as in' ice.ed by heat balance data (66% as indicated by naclear instrumenta-d
tion). This power level was maintained for about 45 minutes'until the reactor protection-system tripped the reactor from another cause.
(Details III, Item 3)
<
_
-
J
.
-
e-eng+
,o b
6.m
-+gm
.
-
w
-
--
,- - --
-.
~
J
,
_
,
.
!
l
- RO Rpt. No. 50-269/74-1 III-2
- s
,,
-
}
,
3.
Spray Valve Failure ThereactortripinItem2abovehasaresultoflowpressureinthe primary system caused by the pressurizer spray valve and spray line block valve failing in the open position.
Section 6.4 of the Technical Specifications requires that the station be operated and maintained 1,n accordance with approved procedures.
Criterion V of Appendix B to 10 CFR 50 requires that activities affecting quality shall be prescribed by documented procedures and be accomplished in accordance with these procedures.
An operating procedure for operation at power, OP/1/A/1102/04, requires that in the event of an unscheduled power reduction, the power level shall not be increased until an investigation has been conducted and
any necessary corrective action taken.
Contrary to the above, after several attempts were made to clase the spray line block valve and spray valve an attempt was made to increase hot leg temperature to increase reactor coolant system pressure by withdrawing control rods to increase power.
At 3:51 p.m., November 20, 1973, a combination of reactor coolant system pressure and temperature tripped the reactor.
Shortly af ter the reactor trip the block valve torque overload circuit was jumpered and the valve closed.
The spray valve was repaired.
4.
Instrument Surveillance Recuirements a.
Heat Balance A review was conducted of the heat balance calibration procedure and the method of implementation of this procedure.
It was deter-nined that the licensee's list of periodic checks called for a heat balance check to be made daily and calibration to be conducted when a disparity between nuclear source instrumentation and heat balance data exceeds 2%.
b.
Reactor Protection System A review of the reactor protection system calibration procedures and the methods used to meet surveillance requirements was con-ducted.
On-line testing of reactor protective channels is required once every four weeks on a rotational basis such that one channel is tested each week.
Duke personnel indicated that this testing
.
is continued through shutdown periods.
.
.%3,
.,9
%..
-
s=--me.se--.
-a
, - - - - --r===
-+,s
-*%~
%--"
%---+eawn
- 8
--wh~+
~
-
>
im
'
-
_j
.
+
Letter to Duke Power Company from N. C. Moseley dated FEB 2 1 1974 50-269/74-1
<
DISThIBUTION:
H. D. Thornburg, R0 RO:HQ (h)
Directorate of Licensing (h)
D yR Central Files
- PDR
- Local PER
- NSIC
- DTIE, OR
- State
- To be dispatched at a later date.
>.
a
.
_~.
/
>