ML20214P268
| ML20214P268 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 11/25/1986 |
| From: | Knop R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20214P226 | List: |
| References | |
| 50-461-86-65, NUDOCS 8612040137 | |
| Download: ML20214P268 (34) | |
See also: IR 05000461/1986065
Text
.
'
=
.
' . -
.
'a
.
U. S. NUCLEAR REGULATORY COMMISSION
REGION III
-
Report No. 50-461/86065(DRP)
Docket No. 50-461
License No. NPF-55
Licensee:
Illinois Power Company
'
500 South 27th Street
Decatur, IL 62525
Facility Name: Clinton Power Station
Inspection At: Clinton Site, Clinton, IL
Inspection Conducted: October 6 through November 7, 1986
Inspectors:
T. P. Gwynn
P. L. Hiland
J. Holmes
<fc Q
,g,s
Approved By:
R. C. Knop, Chief
Projects Section IB
Date
Inspection Summary
Inspection on October 6 through November 7, 1986 (Report No. 50-461/86065(DRP))
Areas Inspected:
Routine, unannounced safety inspection by the resident
inspectors and a region-based inspector of licensee action on previous
inspection findings; IE Circular followup; licensee action on 10 CFR 50.55(e)
report; IE Bulletin followup; licensee event report review and followup;
followup on IE Information Notices; monthly maintenance observation; monthly
surveillance observation; operational safety verification; startup test
witnessing and observation; onsite followup of events at operating reactors;
and management meeting.
Results: Of the twelve areas inspected, no violations or deviations were
identified in nine of the areas.
Six violations were identified in the
remaining areas (paragraph 10 - failure to follow procedures for condition
reporting; paragraph 11.a - eight examples of failure to follow procedures
during initial fuel loading; paragraph 12.b - two examples of performance of
safety related activities without documented instructions or procedures;
paragraphs 9. and 12.b - three examples of inadequate surveillance test
procedures; paragraph 10.b - four examples of violations of the CPS Technical
Specifications; and paragraph 12.b - two examples of personnel errors during
the performance of testing). While each of the identified violations were of
minor safety significance when taken individually, the sum of the violations
over a relatively short period of time represents a trend in performance that
must be corrected.
F612040137 861126
ADOCK 05000461
0
.
..
-
.
t
.
DETAILS
'
.
.
1.
Personnel Contacted
Illinois Power Company (IP)
- @K. Baker, Supervisor - I&E Interface, Licensing and Safety (L&S)
- R. Campbell, Manager - QA
- W. Connell, Manager - Nuclear Planning & Support
- @J. Cook, Assistant Manager - Clinton Power Station (CPS)
- W. Gerstner, Executive Vice President
- J. Greene, Manager - Nuclear Station Engineering Department (NSED)
- D. Hall, Vice President, Nuclear
- E. Kant, Assistant Manager, NSED
- H. Lane, Manager, Scheduling and Outage Management
- J. Miller, Assistant Manager - Startup
- J. Palchak, Supervisor - Plant Support Services
- J. Peterson, Supervising Engineer - Licensing
- J. Perry, Manager - Nuclear Program Coordination
- F. Spangenberg, Manager - L&S
- @J. Weaver, Director - Licensing
- J. Wilson, Manager - CPS
Nuclear Regulatory Commission - Region III
- B. Davis, Deputy Regional Administrator
- C. Nore11us, Director, Division of Reactor Projects
- R. Knop, Chief, Projects Section 1B
@*#T. Gwynn, Senior Resident Inspector, Clinton
@*P. Hiland, Resident Inspector, Clinton
- Denotes those attending the monthly exit meeting on November 4, 1986.
9 Denotes those attending the exit meeting on November 7, 1986.
- Denotes those attending the management meeting on October 22, 1986.
The inspectors also contacted and interviewed other licensee and
contractor personnel.
2.
Licensee Action On Previous Inspection Findings (92701) (92702)
a.
(Closed) Open Item (461/85005-45): As identified in the Final Safety
Analysis Report and the Safety Evaluation Report, the licensee has
installed a Class "A" Fire Detection system. The preoperational test
demonstrated that during the ground minus test, all detectors for the
zone were indicating a trouble alarm. The licensee indicated that
work has been completed on the fire alarm system and that the system
will function as a Class "A" system. On September 17, 1986, the
inspector observed a ground plus and ground minus test on the fire
alarm system with satisfactory results.
This item is considered
closed.
2
. . . ._ -
_
. - . ..
-
--
_.
rc
,
.-
,
s
,
- ' '
-
s-
m ; e, .
_
w
-.
b.
(Closed) Open Item (461/86016-04):
Standby' Liquid Control (SC)
,
c
. System, operability verification. :During witnessing of a surveillance
' test conducted by-the licensee to verify SC operability, it.was
.
. . .
identified that'the' surveillance ~ procedure.(CPS No. 9015.01) could
not be performed.
.
This item was previously rev'iewed as documented in Inspection Report .
-
50-461/86048, paragraph 2.v. 'This item remained open pending NRC~
' inspector witnessi_ng performance of the procedure.
~
~
The inspector witnessed performance'of the SC quarterly operability
test, CPS No.'9015.01,'as documented in paragraph _7 of this report.
This item is closed.
c.
(Closed) Open. Item l(461/86028-08): The licensee provided the-
inspector with the-completed copy of the preoperational test results
'for the Emergency Lights (PTP-LL-01). The. review of the test results
found the results adequate. :This. item is considered closed.
d.
(Closed) Open Item (461/86028-12): The. reactor recirculation pumps
- at the Clinton Power Station are located in a non-inerted containment
-and are not provided with an oil collection system as required by
'
The'11censee's response package entitled "NRC I'.E. Issue'Open Item
86-28-12" (document.not dated) stated " Illinois Power submitted a
deviation ^ request to NRR for.the oil collection system of
recirculation pump. motors required by Appendix R,Section III.0".
In a telephone call on June 13, 1986, between NRR and Region III, NRR.
agreed to ~ review previously submitted material from the licensee
regarding the lack of an oil collection system and to document these
findings in a future . safety evaluation report.
In the Clinton Safety Evaluation Report Supplement No. 6, it states,
"In the SER the staff reported that an oil collection system is not
required for the reactor coolant pumps". This' action by NRR is
sufficient to close this item.
e.
(Closed) Violation (461/86054-04):
Unauthorized modification.
installed around the body of Low Pressure. Core Spray (LPCS) injection
isolation valve IE21-F005. A steel plate used as a temporary
security barrier was installed around the LPCS injection isolation
valve without proper controls. Although the licensee identified this
condition during a management walkdown,'no action had been taken to
determine the cause of the condition and to preclude recurrence.
As documented in Inspection Report (IR) 50-461/86060, paragraph 2.t.,
the inspector reviewed the results of the licensee's investigation
into the subject violation. The item remained open pending receipt
and review of the licensee's formal response.
3
'
.
-
em,
.,
LDuring..this report period, the licensee' responded to.the subject-
,
'.
.
iviolation in a timely manner. Additional corrective action taken by
ithe licensee to prevent furtheriviolations, not previously' documented
-
.
'in IR 50-461/86060, included the issuance of IP letter JW-1903-86.
Thisiletter from the Manager - CPS to all IP. Plant Staff Personnel
stressed the. requirements of adhering'to CPS procedures when
' personnel identify an indeterminate or adverse condition.
A
The inspector confirmed the licensee's corrective actions were
complete. -This item is closed.
'
f.
(0 pen) Openl Item (461/86028-13): . The licensee was requested to
verify the maximum depth of frost and if-necessary take corrective
action for.the underground firemain.
In'the licensee's. document entitled, " Depth Of Cover For Underground
-
Firemains",
indicates that the underground firemains were installed
in 1976 in accordance with the 1973 edition of NFPA-24 which states
in part, "The depth of cover over water pipes should be determined
-by the maximum depth of frost penetration in the locality".
In the
attachment 1 of the letter from R. Parson, Sargent and Lundy to
Illinois. Power Company, dated June 25, 1986, it indicates that
,
the depth of-frost at the Clinton site is 32 inches based on.
~
interrelating data from Foundation Engineering by Alfred R. Jumikis,
1971.
In addition, this letter states "The design basis of 42" depth
of cover is consistent with the practices of a local municipality and
a local water company. The local municipality and water company
recommend a minimum depth of cover of 3 1/2 feet for fire protection
lines (standing water)."
The licensee has identified several areas where the underground
firemain is not provided with a depth of cover of 42 inches. The
licensee indicated that a minimum depth of cover of 42 inches will
be provided for those areas by November 30, 1986.
This item will remain open pending verification of the licensee.'s
corrective actions regarding those areas provided with less than 42
inches of cover.
No violations or deviations were identified.
3.
IE Circular Followup (92701)
(Closed) IE Circular (461/78-18): UL Fire Tests.
IE Circular 78-18 identified three concerns as follows:
a.
Flamastic 77 fire retardant coatings may absorb flammable liquid from
an inadvertent spill inside of the barrier at the bottom of vertical
trays.
Attached to the licensee's internal memorandum dated August 6, 1984
from E. Haagar to D. Kerborn (L08-84[08-06]-6) is NSED review and
response which indicated that the fire barrier sealing subcontractor
4
--.-
-
.
.
,
.
anticipated! hat the proposed fire stops for. vertical cable. tray runs
t
- .
.
will be of different (other.than ceramic fiber) UL approved material.
4
The: licensee indicated to the' inspector that Bisco Silicone Foam'.
~
' SF-20, ~ SF-60, SF-150. NH 'is fire rated and is~ nonabsorbent.
The licensee:provided the inspector with Bisco detail No. 107 through
112 and No. 126. .The inspector noted that 9 inches.of silicone foam
,
is used above 1 inch of ceramic material in Bisco detail.No. 107 and
'110.
Based on the fact that there-is no gap between the silicone.-
foam'and the steel sleeve,-the' flammable liquid.from a potential
spill could not be absorbed by the 1 1ach of ceramic material. The
concern regarding the: absorption of flammable li_ quid into the fire
barrier is considered closed.
b.
Small fires may not actuate Sprinkler Heads.
/
To reduce ~ the possibility of small fires not actuating sprinkler
heads, fast response sprinkler heads should be considered.
In the
August 6, 1984 memorandum (L08-84 [08-06]-6), the licensee indicated
that all sprinkler heads installed by the CPS sub-contractor
" Automatic Sprinkler Corporation" are UL listed and meet the
requirements of UL Standard 199. < Based on the licensee's internal
response, it is not clear that'the licensee considered in the system
layout the concern for small fires not actuating sprinkler heads.
The licensee indicated to the inspector that the sprinkler systems
have been installed to meet NFPA-13 requirements, and that in
walkdowns consideration has been given to use of fast response
sprinkler heads where required. This. item is considered closed
based on the licensee's response.
c.
Path of air movement influences the actuation of detection devices.
The tentative conclusion from the test also identified the concern
that the location of fire detection devices is of great importance.
The path of air movement in the area influences the actuation of
such devices and should be considered in the system layouts.
In the August 6, 1984 memorandum (LO8-84[08-06]-6), the licensee
states, " fire detection devices are insta11ed'as required by NFPA
72E which establishes the rules for location and spacing in high air
moving areas." This item was previously closed, however, during the
9/10-12/86 inspection the licensee provided the inspector with the
document entitled " Fire / Smoke Detector Layout For Illinois
Power Company", Pyrotronics Job #PC-4920B which indicated that
consideration was given to the placement of detectors with respect to
ceiling tiles, lighting fixtures, air diffusers, egg crate ceiling
sections, soffits, and ceiling obstructions. A field walkdown was
also conducted to-ascertain that the drawings depicted actual
conditions and also to assure that influencing factors were picked
up during the drawing review. This document indicates that a
proposed detector layout took into full account the guidelines set
forth in NFPA 72E as well as good fire protection practices and
,
5
.
- -
.
.
.
engineering judgment and in no case was the area coverage per
.-
.
detector greater than that recommended by either the National Fire
Protection Association or Pyrotronics.
In many areas additional
detectors were located to compensate for increased fire loading or
the physical location of the fire load. The proposed detector
layout drawings were taken back to the. field and checked against
actual site conditions, and the detector layout drawing was
finalized, which was reviewed by Sargent and Lundy.
Based on the licensee's response, this item is considered closed.
No violations or deviations were identified.
4.
Licensee Action on 10 CFR 50.55(e) Report (92700)
.
(0 pen) 10 CFR 50.55(e) Report (461/84023-EE):
Ruskin Fire Dampers.
The letter dated August 28, 1985, from D. Hall, IPC, to J. G. Keppler,
NRC, indicated that " Illinois Power investigation of this matter is
complete". This letter also described the investigation results,
corrective action, and attached engineering change notices. The
corrective action taken consisted of the following:
Twenty horizontal dampers in the designated "VC" System (Control
Building) are not acceptable. Heat detectors are to be provided
within selected ducts so that administrative controls can be
implemented to manually shut the fan off to prevent transmission
of fire from one fire area to the other.
One vertical damper in the designated "VA" System (Auxiliary
Building) is acceptable per the test results with the recommended
change to larger springs. This change has been issued to the Zack
Company on Field Engineering Change Notice (FECN) 10503, dated
June 26, 1985.
Nine vertical dampers in the non-safety related ducting designated
"VF", "VJ", "V0", "VP", "VT", and "VW" Systems (Fuel Building,
Machine shop, Off gas, Drywell, Turbine, and Radwaste Building,
respectively) are not acceptable.
Heat detectors are to be provided
within the ducts to trip the fans off to prevent transmission of fire
from one fire area to the other.
The licensee provided the inspector with several documents which included
the Test Report Revision 1 NIBD23 entitled " Fire Damper Closure" and test
data sheets. The licensee identified in the August 28,4985 letter from
D. Hall to Keppler that 150 dampers (26 horizontal and 124 vertical) are
g
acceptable per the test results and no modification is required. The
information provided to the inspector did not clearly indicate and compare
the test results to actual physical conditions at the site. The licensee
was requested to organize the material to facilitate review by a person
not involved with the analysis.
6
. ~ , . -
-
. - .
-
-.
.
.
.
.
V
-
,
~~ c .'
.-
In addition, the licensee agreed (during the previous inspection), to test':
-
several dampers in the most severe configurations to insure that dampers:
.
3"
will close under air flow conditions.
.
- The organization of material to -facilitate for review and the testing of
i
dampers in their most severe configuration should be completed by the end
.
of.the first refueling outage. This item is open pending licensee actions
w
described above.-
,
No. violations or deviations were identified.
5.
IE-Bulletin Followup '(92703)(25581)
The bulletin listed below was. reviewed to verify that the written response
was within the time period stated in the bulletin, that the written
. response included the information required to be reported, and that the:
F
written. response included adequate corrective action commitments based
- on-information presented in the bulletin and the . licensee's response.
,
(Closed) IE Bulletin (461/86001-88): Minimum Flow Logic Problems
'
That Could Disable RHR Pumps.
-
J
This bulletin was previously reviewed as documented in Inspection
,
Report 50-461/86037. The bulletin remained open pending issuance of
.
a NRC Temporary Instruction for inspection of this bulletin and pending
completion of the licensee's reviews.
, <
!
The inspector reviewed the completed results of the licensee's actions
on this bulletin.. The bulletin was determined to be not applicable to
'
Clinton Power Station because independent logic systems were provided.
'
t
This bulletin is closed.
j
No violations.or deviations were identified.
j
6.
Licensee Event Report (LER) Review and Followup (90712&92700)
,
j
a.
In-Office Review Of Written Reports Of Nonroutine Events At Power
j
Reactor Facilities (90712)
1
l
For the LERs listed below, the inspector performed an in-office
!
review of each LER to determine that reporting requirements had
!
been met; that the corrective action discussed appeared appropriate;
that the information provided satisfied the applicable reporting
requirements; to determine if appropriate actions had been taken on
i
any generic issues present; and to Jetermine if any additional NRC
i
inspection, notification, or other response was appropriate. Where
determined appropriate, the LER was scheduled for onsite followup
!
inspection or other necessary action by cognizant NRC personnel.
!
(1)
(Closed)LERNo. 86-001-00 (461/86001-LL):
Isolatior, of Reactor
Water Cleanup (RT) System Due To Spurious Trip Of RT Fump Room
3 High Differential Temperature.
!
!
!
!
i
7
i
'
- . -
- - -
. - . - - . - - -
- . - - - - - . . -
.
..
.
.,
.
,.
(2) (Closed) LER No. 86-002-00 (461/86002-LL): Secondary
.
.
Containment Negative Pressure Lost Due To Personnel Defeating
Interlock Switch On Outer Airlock Doors.
-
(3) (Closed) LER No. 86-003-00 (461/86003-LL): Control Room
Ventilation Chlorine Mode Initiation Due To Chlorine Detector
Failure.
(4) (Closed) LER No. 86-005-00 (461/86005-LL): Personnel Failure To
Perform Hourly Firewatch Rounds.
This LER was reported pursuant to the requirements 10 CFR 50.73-
(a)(2)(i)B which requires that the licensee report any
~
operation or condition prohibited by the plant's technical
specifications.
It was not apparent to the inspector that the
event reported satisfied the referenced requirement.
This
matter was reviewed with the licensee as discussed in paragraph
6.c. of this report.
(5) (0 pen) LER No. 86-006-00 (461/86006-LL): Automatic Initiation
Of Essential Service Water Due To Transient Pressure Drop In
Nonessential Service Water.
The licensee is planning to provide a supplemental report by
December 13, 1986. This matter will be reviewed further with
the supplemental report.
No violations or deviations were identified.
b.
Onsite Followup Of Written Reports Of Nonroutine Events At Power
Reactor Facilities (92700)
(1) For the LERs listed below, the inspector performed an onsite
followup inspection of each LER to determine whether responses
to the events were adequate and met regulatory requirements,
license conditions, and commitments and to determine whether
the licensee had taken corrective actions as stated in the LER.
(0 pen) LER No. 86-004-00 (461/86004-LL): Unplanned Automatic
Initiation Of Standby Gas Treatment System Due To Inadequate
Procedures.
In-Office review of this LER raised a number of questions
concerning the adequacy of the report preparation and one
question concerning the basis for limitations on the corrective
actions taken. These matters were reviewed further by the
inspector through personnel interviews and discussion with
licensee management.
Concerning the adequacy of the report preparation, the inspector
noted that the report abstract contained information not
contained in the LER narrative description, as follows:
8
_
-
'
?
.
' :. . =
'
, - . .
--
,-
.'
-Plant conditions prior to the event.
Time of the event.
.
'
-
-
Operator actions as a result of the Engineered Safety-
Feature (ESF) actuation.
.
The results of a channel functional test.
kin addition, the narrative description did not clearly describe
the contribution made'to the event.by the-unknown status of.a
radiation monitor (IRIX-PR008C)' trip signal prior to the event -
and did not describe the ESF response'to the unplanned'
. automatic actuation signal.
Finally, .the inspector _ noted that -
"the narrative description was not sufficiently clear to' allow a
-
good understanding of the. event. These matters were discussed
.with responsible licensee management.
The inspector requested
that the licensee consider providing a revision to this LER.
Concerning the corrective actions being taken by the licensee,-
the LER stated that the procedure changes made would prevent
inadvertent initiation of ESFs associated with four radiation
monitors in addition to the radiation monitor that caused the
even t . -- The inspector noted that a significant contributing
cause of the event was the unawareness of plant operators and
'
test technicians of the actual' status of the radiation monitor
trip logic prior to commencement of the. surveillance test (i.e.,
the logic for radiation monitor 1R1X-PR008C was sealed in a
tripped condition prior to commencement of testing on
1R1X-PR008B with no indication that the tripped condition
existed) . The LER did not provide any information that
indicated the use of-a logic trip seal-in (without annunciation
or other status indication) was isolated to the five radiation
monitors mentioned in the report. The inspector requested that
the licensee provide the basis for limiting the scope of
corrective actions taken to the five radiation monitors
mentioned in the report.
The inspector reviewed CPS No. 9911.20, Safety Related PRM
Surveillance - Monthly Channel Functional Test, revision 23
dated October 17, 1986, to verify that procedure changes
identified in the licensee's report had been incorporated.
The inspector also reviewed Plant Managers Standing Order
(PMS0)-30, SS/ ASST. SS Notification During Surveillance Testing,
revision 3 dated October 20, 1986, and discussed its use with
several licensed senior reactor operators.
PMS0-30 appeared to
be an effective tool in assuring that the impact of initial
performance of mode 1, 2, or 3 surveillance tests was known to
shift operators prior to test performance. However, the
description of corrective action provided in the LER (i.e., "A
standing order was generated to evaluate all plant surveillances
prior to performance to ensure conditions will support the
test") did not appear to be consistent with the PMSO which
appeared to limit the evaluation to mode 1, 2, and 3 initial
9
.
-
.
.
.,
.
performance surveillance tests that required lifting leads,
-
.
installing jumpers, or installing test equipment that could
affect the circuit. This matter was also discussed with the
. licensee concerning the need for a revised LER as documented'
above.
This LER will be reviewed further during a subsequent
inspection.
No violations or deviations were identified.
c.
Event Reporting (92700)
Several questions regarding event reporting were discussed with the
licensee during the report period. Those questions resulted from
NRC review of LERs and followup on events that occurred during the
inspection. Those questions involved the following:
(1) LER No. 86-005-LL was reported pursuant to 10 CFR 50.73(a)(2)(1)B which usually applies to violations of Technical
Specification (TS) Action statements. The subject of LER No.
86-005-LL did not involve a violation of a TS action statement.
The licensee was requested to confirm the applicable reporting
requirements.
(2) The licensee did not report two matters involving actuations of
engineered safety features (ESFs) based on their determination
that the actuation was part of the " preplanned sequence during
testing or reactor operation" and that, in accordance with 10 CFR 50.72(b)(2)(11), the actuations were not reportable. The
actuations resulted from operator actions that were performed
without documented instructions or procedures. The licensee was
requested to confirm the applicable reporting requirements and
to provide clarification concerning their definition of a
" preplanned sequence during testing or reactor operation".
(3) The licensee reported a change to their operating organization
which was in conflict with the organization chart provided in
Section 6.1 of the CPS TS via the Emergency Notification System
(ENS).
This matter was clearly not ENS reportable. The basis
for their report was the CPS operating license paragraph 2.F.
The licensee was requested to clarify the applicability of
license paragraph 2.F for reporting violations of the TS and the
Environmental Protection Plan.
At the conclusion of the inspection period, the licensee was
developing interpretations for the information and use of their
,
operating organization concerning the above matters.
Items 1, 2,
and 3 above are considered to be unresolved pending issuance of the
interpretations, training of applicable personnel, and inspector
review (461/86065-01).
10
.
.
-_
- .
-
--- - -- -
.
-
-
._.
. - - . - -
"~
._
.
..
,
.
,
'
.
7.
Followup on IE Information Notices (92701)
-
.
For each of the IE Information Notices listed below, the inspector
verified that the licensee had received the Information Notice, had
distributed the Notice to appropriate personnel, had reviewed the Notice
for applicability, and, if applicable, had scheduled or completed
appropriate corrective actions.
Information Notice
Number
Subject
85-53
Performance Of NRC-Licensed
Individuals While On Duty.
85-59
Valve Stem Corrosion Failures.
85-62
Backup Telephone Numbers To The NRC
Operations Center.
85-77
Possible Loss of Emergency
Notification System Due To Loss Of AC
Power.
85-89
Potential Loss Of Solid State
Instrumentation Following Failure Of
Control Room Cooling.
85-91
Load Sequences For Emergency Diesel
Generators.85-101
Applicability Of 10 CFR 21 To
Consulting Firms Providing Training.
86-18
NRC On-Scene Response During A Major
Emergency.
86-27
Access Control At Nuclear Facilities.
86-30
Design Limitations Of Gaseous
Effluent Monitoring Systems.
86-37
Degradation Of Station Batteries.
86-39
Failures Of RHR Pump Motors And Pump
Internals.
86-44
Failure To Follow Procedures When
Working In High Radiation Areas.
86-48
Inadequate Testing Of Boron Solution
Concentration In The Standby Liquid
Control System.
11
-
--
. _
- -
- - _ .
.-.
- - - _
_
. . - . --
. -
=
-
.
y..
.
s..
.
.
.
'86-65
Malfunctions Of ITT Barton Model 580
-
Series Switches During Requalification
,
Testing.'
'
86-72'
Failure 17-7 PH' Stainless Steel
Springs In Valcor Valves Due To
Hydrogen Embrittlement.
-86-82--
Failures Of. Scram Discharge Volume
Vent And Drain-Valves.
Concerning Information Notice 86-72, the' inspector determined that'Valcor
- Engineering Inc. had notified the licensee of the potential deficiency by
letter dated April 25, 1986. _The licensee performed an initial evaluation
for appitcability/reportability under 10 CFR 21 referral No. 21PE36-from
Baldwin Associates (the plant constructor) dated May 5,1986. The results
of their evaluation, documented in memorandum Y-80932 dated June 4,~
1986,
indicated that the condition identified by Valcor did not apply to-Clinton
Power Station due to different chemistry. Subsequent correspondence,.
-
memorandum Y-81702 dated August 15, 1986, indicated that the Nuclear
Station Engineering Department was evaluating the desirability of
modifying plant chemistry controls to include hydrogen addition to the
reactor coolant system. The correspondence indicated that a part of
the modification approval process (if applicable) would include
reconsideration of the applicability of this problem. Information Notice 86-72 was still under review by the licensee at the time of this
inspection.
.
The inspector noted that, of the 17 information notices reviewed, only one
(Information Notice 85-91) had been completed with records transmitted to
the CPS Central Files for retention. The inspector observed that the
original records of IN 85-91 had been retained in the Licensing & Safety
(L&S) Department while copies had been transmitted to the CPS Central
Files. This practice appeared to deviate from the recommended practice
in CPS Records Management Standard (RMS) No. 2.01, Standard For the
Collection and Review of Records, which specifies that the original
documents should be transmitted for retention. Discussion with L&S
personnel indicated that the originals would be transmitted to the CPS
Central Files for Information Notice 85-91 and that future records
transmittals would be made in accordance with the RMS.
For the other 16 information notices reviewed, the inspector noted that
a number had exceeded their scheduled completion dates with no forecast
for completion and that the others were pending closure action by L&S
personnel. Discussion with the Manager - L&S indicated that management
was aware of the current program status and had implemented actions based
on recommendations of a management consultant to improve the program.
The results of actions taken by L&S to improve their industry experience
program will be reviewed during a future inspection. This is an open
inspection item (461/86065-02).
12
, ,
.
.
.
.
.
- _____ -_--____-_
-___________-_
.
..
.
.
.,
'
.
No violations or deviations were identified.
-
-
~
8.
Monthly Maintenance Observation (62703)
i
Station maintenance activities of safety related systems and components
listed below were observed / reviewed to ascertain that they were conducted
in accordance with approved procedures, regulatory guides and industry-
codes or standards and in conformance with Technical Specifications.
The following items were considered during this review:
limiting
conditions for operation were met while components or systems were removed
,
from service; approvals were obtained prior to initiating the work;
l
activities were accomplished using approved procedures and were inspected
as applicable; quality control records were maintained; activities were
accomplished by qualified personnel; parts and materials used were
properly certified; radiological controls were implemented; and fire
prevention controls were implemented.
The following maintenance activities were observed / reviewed:
a.
Maintenance Work Request (MWR) C-27926 was observed by the inspector
during its performance. This MWR was initiated to perform motor
operated valve testing (M0 VATS) on Reactor Core Isolation Cooling
system valve IE51-F078. Valve IE51-F078 had failed a local leak rate
test (LLRT) and the performance of M0 VATS testing was to verify
proper valve operation. During the testing sequence, maintenance
personnel identified that the stem follower (position indicator) was
interfering with a brass stem nut.
This interference was apparently
preventing the valve disc from seating properly. The Quality Control
(QC) inspector monitoring the maintenance activity requested an
evaluation from the Nuclear Station Engineering Department (NSED)
which was provided at the job site. Condition Report (CR)
1-86-10-252 was initiated to document the identified deficiency
and to provide for corrective action.
Review of the work package present at the job site indicated that
authorization to commence the maintenance activity had been obtained
from the Shift Supervisor; testing equipment was within calibration
,
and verified prior to commencement of the work; approved procedures
l
were present at the work location; and QC hold and witness points
were adhered to.
b.
MWR C-15430 was observed by the inspector during its performance.
This MWR was initiated to perform plant modification IS-7 on the
Main Steam Isolation Valve Leakage Control (IS) system.
The
inspector observed the electrical termination work activities in
progress at local panel 1PL100J. The inspector noted that the
applicable portions of the work package were present at the work
I
site; parts and materials used were properly certified; QC was
performing the inspections required by the MWR and applicable
procedures; the calibration of tools was current and verified by
QC.
13
__
_
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
.
..
,
.
.,
.'
No violations or deviations were identified.
.
-
9.
Monthly Surveillance Observation (61726)
The inspector observed Technical Specifications (TS) required surveillance-
testing on the Standby Liquid Control (SC) System and verified that
testing was performed in accordance with approved procedures in the time
frame required by the TS surveillance schedule; that test instrumentation
was calibrated; that limiting conditions for operation were met; that
removal from and restoration to service of the affected components was
properly accomplished; that documented test results were consistent with
inspector observations and conformed with Technical Specifications and
procedure requirements; that test results were reviewed and approved in
accordance with the licensee's procedures; and that any deficiencies
identified during the testing were properly reviewed and resolved by
appropriate management / supervision.
On October 9, 1986, the performance of surveillance test procedure CPS No.
9015.01, Standby Liquid Control System Operability, revision 22, step
8.2.6 resulted in closure of the reactor water cleanup (RT) system pump
suction isolation valve due to a logic interlock function. The licensee
determined that the surveillance test procedure was missing information
necessary to preclude operation of the interlock. A temporary procedure
change was written to complete test performance and a critique of the test
was held with the test crew and the shift supervisor at the completion of
the test.
10 CFR 50, Appendix B, Criterion V states, in part:
" Activities affecting
quality shall be prescribed by documented instructions, procedures, or
drawings, of a type appropriate to the circumstances and shall be
accomplished in accordance with these instructions, procedures, or
drawings...".
The failure to provide adequate instructions to prevent
actuation of the RT pump suction isolation interlock during performance
of surveillance test procedure CPS No. 9015.01 is a violation of
10 CFR 50, Appendix B, Criterion V (461/86065-04A(DRP)).
One violation was identified.
10. Operational Safety Verification (71707)
The inspectors observed control room operations, attended selected
pre-shift briefings, reviewed applicable logs, and conducted discussions
with control room operators during the inspection period.
The inspectors
verified the operability of selected emergency systems and verified
tracking of LCOs.
Routine tours of the auxiliary, fuel, containment,
control, diesel generator and turbine buildings and the screenhouse were
conducted to observe plant equipment conditions including potential for
fire hazards, fluid leaks, and operating conditions (i.e., vibration,
process parameters, operating temperatures, etc). The inspectors verified
that maintenance requests had been initiated for discrepant conditions
observed. The inspectors verified by direct observation and discussion
14
,
.
..
,
.
.,
.
with plant personnel that security procedures and radiation protection.
-
.
(RP) controls were being proper.ly implemented.
,
.The inspectors' observed plant housekeeping / cleanliness conditions. No
,
discrepancies were noted.
The'above ' reviews and observations were accomplished to verify that
facility operations'were conducted in conformance with the CPS technical
specifications and the conditions of the operating license.
The inspector reviewed the status and implementation of the licensee's
condition reporting (CR) program as defined in CPS No. 1016.01, CPS
Condition Reports, revision 11 dated September 18, 1986. About 300 CRs
were initiated during the month of October with a total of about 562 CRs
-
open at the end of the month. Three.CRs were reviewed in some detail
-(in process) and 17 CRs were reviewed for procedural compliance
(completed and in the CPS central files), as follows:
'CR STATUS
REVIEW TYPE
CR No.
C-COMPLETE I-INPROCESS
D-DETAILED N-NONDETAILED
1-86-08-075
C
N'
1-86-08-076
C
N
1-86-08-077
C
N
1-86-08-078
C
N
1-86-08-079
C
N
1-86-08-104
C
N
1-86-08-134
C
N
1-86-08-171
C
N
1-86-08-172
C
N
1-86-09-008
C
N
1-86-09-021
C
N
1-86-09-075
C
N
1-86-09-078
C
N
1-86-09-082
C
N
1-86-09-092
C
N
1-86-10-054
I
D
1-86-10-056
I
D
1-86-10-065
I
D
This review resulted in identification of the following violation of
CPS No. 1016.01:
CPS No. 1016.01, paragraph 2.2.9 requires a resolution for each
condition report as follows:
Resolution - An approved plan of corrective action (remedial and
generic) which includes a schedule with accountability assignments
for performing scheduled tasks, and a tracking and reporting system
to ensure adequate progress is being made. Resolution of remedial
and generic corrective action may occur simultaneously or separately.
For the purpose of this definition, an " approved" plan is one that
has been signed by the applicable Department Head.
15
-
.
..
,
.
.,
.'
CPS No. 1016.01, paragraph 8.6 requires the following:
-
-
Investigation and Resolution
The department / organization assigned investigation responsibility
shall complete the resolution and investigation and shall sign and
date BLOCK 9, unless a NCMR was initiated, except as noted in section
8.4.5.2.
Investigation responsibility may be reassigned per section
8.7 of this procedure.
CPS No. 1016.01, paragraph 8.2.1.2 identifies the requirements for
filling out block 2 of the CR form, as follows:
BLOCK 2 - Requirements Reference
Enter the references against which the CR is written.
Enter the
document title, number, revision, date, and applicable section, as
appropriate.
If there is no referencing document indicate "NONE".
CPS No. 1016.01, paragraph 8.4.3 states the requirements for the
Compliance Department to review the completed CR form, as follows:
Review
Compliance shall review the CR to ensure that it conforms to the
requirements of this procedure.
Nor,e of the 18 CRs reviewed had a documented corrective action plan that
had been reviewed and approved by the applicable department head prior to
corrective action implementation.
In each case, the information
documented in block 9 of the CR represented a historical summary of
actions taken and results achieved rather than a plan for corrective
action based on the results of an investigation.
In addition, several CRs
reviewed (e.g., CR Nos. 1-86-10-054,1-86-10-065), both in process and in
the CPS Central File, had no requirements referenced in block 2 of the CR.
These violations of CPS No. 1016.10 apparently were not identified during
reviews by the Compliance Department, the QA Department, and the Facility
Review Group. The above represent examples of failure to follow CPS No.
1016.01 in the processing of CRs. This is a violation of 10 CFR 50,
Appendix B, Criterion V and the IP Operational Quality Assurance Manual,
Chapter 5 (461/86065-03).
Several of the CRs reviewed were written during performance of activities
which had subsequently been completed with no plan of corrective action
prepared.
Failure to provide a documented corrective action plan based on
the results of an investigation represented a potentially serious lack of
management control of the licensee's corrective action program and
represented a violation of procedure that was either not recognized or
was recognized and accepted by plant management.
One violation was identified.
,
16
. . _ _
_
.
.,.
_ . _ .
~
-
,
r
..
,
.
.,
.'
-
-
11. Startup Test Witnessing And Observation (72302/72524)
The inspectors witnessed the conduct of portions of the following startup
"
test procedures to ascertain through direct observation and record review
,
that startup testing was performed in accordance with approved procedures-
and the requirements of the CPS operating license.
a.
STP-03, Initial Fuel Loading Witnessing (72524)
(1) Purpose
The inspectors witnessed initial fuel loading. activities being
performed by the licensee in accordance with test procedure
STP-03-0, Fuel Load.
For the portions of the test and fael load
activities witnessed, the inspectors verified that Clinton Power
Station (CPS) Technical Specification requirements and license
conditions were met; that nuclear instrumentation was calibrated
and operating with a measurable count rate; that prerequisites
and initial conditions were met; that staffing requirements and
communications were in accordance with Technical Specifications;
that proper procedures were in use and being followed; that
1/M plots were maintained in accordance with the procedure;
that shutdown margin and control blade operability were
verified at required frequencies; that shift turnovers were
conducted; that control of personnel access to refuel floors
was adequate; that refuel status boards were maintained; that
personnel at each refuel station understood responsibilities;
that overtime limits were observed; that the " master" copy of
procedure STP-03-0 was maintained; that changes to fuel load
procedures were technically adequate and in accordance with
approved CPS procedures; that corrective actions for
deficiencies or problems were adequate; that data sheet entries
were legible, traceable and permanent; and that problems or
deviations from the fuel load procedure were adequately
documented in the control room log.
(2) Discussion
During this report period, the licensee completed loading the
initial core of 624 fuel bundles. The inspectors monitored fuel
loading activities on a continuous basis during the first
one-third core load and on a daily basis thereafter until fuel
loading was completed on October 21, 1986.
Inspectior. results
for the first one-third core load were documented in Inspection
Report 50-461/86060.
The results documented below complete the
inspection activities for the initial fuel load.
(3) Results
Except as discussed below, fuel loading activities were
controlled in accordance with the plant Technical Specifications
and the governing procedures; CPS No. 3007.02, Preparation For
17
-
o
.
. . =
,
.
.,
.
and Recovery From Refueling Operations For Initial Fuel Load;
-
-
CPS No. 1898.00, Special Nuclear Materials Program; and startup
phase test procedure STP-03-0, Fuel Load. As documented in
paragraph 12 and in the paragraphs below, the licensee violated
the plant Technical Specifications for conducting core
'
alterations (fuel loading) on two occasions and made several
procedural errors during the conduct of initial fuel load.
Procedural errors made were as follows:
(a), On October 3, 1986, an incorrect fuel bundle was
transferred from the 755' elevation of the Fuel Building
to the 828' elevation of Containment. The Special Nuclear
Material (SNM) Transfer Checklist in use (86-106)
designated movement of fuel bundle No. LY3502 from 755'
Fuel Building to 828' Containment.
Fuel bundle No. LY3216
was leaded into the Inclined Fuel Transfer carriage and
moved from 755' Fuel Building to 828' Containment. The fuel
movement error was identified by the Senior Reactor
Operator (SRO) in charge of Fuel Handling. The incorrect
fuel bundle (LY3216) was returned to the 755' Fuel
Building and the correct fuel bundle (LY3502) was moved to
the 828' Containment in accordance with the approved SNM
transfer checklist. The licensee initiated Condition
Report (CR) 1-86-10-056 to document the procedural error
and to provide corrective action.
Criterion V states in part:
" Activities affecting quality
shall be prescribed by documented instructions, procedures,
or drawings, of a type appropriate to the circumstances
and shall be accomplished in accordance with these
instructions, procedures, or drawings...".
The failure
to follow the SNM checklist for movement of fuel bundles
between the 755' Fuel Building and the 828' Containment
as required by CPS No. 1898.00, paragraph 8.4.5, is a
violation of 10 CFR 50, Appendix B, Criterion V
(461/86065-05A(DRP)).
(b) On October 4, 1986, two fuel bundles of the wrong
enrichment were placed into the reactor.
Fuel bundles LY3213 and LY3246 of medium enrichment
(1.54%) were placed into reactor locations requiring high
enrichment (2.00%) type fuel . The licensee initiated
Condition Report (CR) 1-86-10-054 to docuraent the error
and provide corrective action. The licensee identified
the root cause of this event to be an error by the onshift
,
Nuclear Engineer who changed the SNM transfer checklist.
CPS No. 1898.00, paragraph 8.4.6.2, allowed changes to
an approved SNM transfer checklist if made by a member
of the Technicai Department Nuclear Group and approved
by an SRO. While making a change to the approved SNM
Transfer Checklist, the onshift Nuclear Engineer selected
the wrong enrichment type fuel resulting in its subsequent
18
_ _ , _
. _ .
.
.
.
.
.
..
,
.
.
.
.
placement into the reactor. This error was noted during
.
the previous report period and documented in Inspection
Report 50-461/86060, paragraph 6.e.
was not issued at that time based on the licensee's self
identification and immediate corrective action. However,
during this report period, the inspectors were unable to
verify the adequacy of corrective action implementation
due to the violation discussed in paragraph 10 above, and
a similar error (see subparagraph (e) below) which
occurred during this report period. This indicated the
licensee's corrective actions to this procedural violation
were.either inadequate or inadequately implemented.
10 CFR 50, Appendix B, Criterion V states, in part:
" Activities affecting quality shall be prescribed by
documented instructions, procedures, or drawings, of a type
appropriate to the circumstances and shall be accomplished
in accordance with these instructions, procedures, or
drawings...". The failure of the Nuclear Engineer to
prescribe the correct enrichment type fuel on the SNM
Transfer Checklist, is a violation of 10 CFR 50, Appendix
B, Criterion V (461/86065-05B(DRP)).
(c) On Octcber 5, 1986, after loading fuel into control cell
20-29, a subcriticality check of the just loaded cell
(20-29) was not performed.
Procedure STP-03-0, " Fuel
Load", paragraph 7.3.8 required the performance of a
subcriticality check for the just-loaded cell.
Paragraph 7.3.8 of STP-03-0 directed that the control rod
in the new fully loaded cell be fully withdrawn and
inserted. This action would perform a control rod
functional check; would perform a subcriticality check for
the just loaded cell; and would perform a subcriticality
check for the next cell to be loaded. Since the just
loaded control cell (20-29) was located in two quadrants,
the licensee determined that Source Range Monitors (SRM) A
and B were required to be on scale (above .7 counts) for
control cell 20-29. At that point in the fuel load process
SRM-B was not on scale; therefore, the licensee determined
that instead control rod 24-33 would be withdrawn (this
control-cell was located in only one quadrant) for the
subcriticality check required for the next control cell
(20-33) to be loaded.
The control rod functional check performed by paragraph
7.3.8 of STP-03-0 was a parallel test being conducted per
STP-05-0, " Control Rod Drive System". The conduct of the
control rod functional test (STP-05-0) was controlled by
its own procedure and its performance in parallel with the
subcriticality checks was not a requirement as defined in
STP-05-0, paragraph 7.1.
The licensee performed a subcriticality check for the next
control cell to be loaded (20-33) by withdrawing control
.
19
_
-
-
_.
.
~
.
.
.
.'
rod 24-33. A subcriticality check on the just loaded cell
-
-
(20-29) was not performed in accordance with STP-03-0,
paragraph 7.3.8 nor was a revision made to STP-03-0 that
allowed omission of this requirement. On October 8, 1986,
the licensee initiated Condition Report (CR) 1-86-10-096 to
document this procedural error and provide corrective
action.
10 CFR 50, Appendix B, Criterion V states in part:
" Activities affecting quality shall be prescribed by
documented instructions, procedures, or drawings, of a type
appropriate to the circumstances and shall be accomplished
in accordance with these instructions, procedures, or
drawings...".
The failure to perform a subcriticality
check on the just loaded cell 20-29 in accordance with
procedure STP-03-0, paragraph 7.3.8 is a violation of 10 CFR 50, Appendix B, Criterion V (461/86065-05C(DRP)).
(d) On October 7, 1986, two fuel bundles were loaded into
control cell 20-41 without performing a subcriticality
check before loading the cell.
Procedure STP-03-0,
" Fuel Load", paragraph 7.3.8.2 required performance of a
subcriticality check for the next cell to be loaded. The
itcensee initiated Condition Report (CR) 1-86-10-063 to
document the procedural error and provide corrective
action. The licensee's iraediate action upon identifying
this error was to suspend core alterations, remove the two
bundles that were loaded, and the required subcriticality
check was performed. The licensee determined the root
cause of this event was personnel error by the Shift Test
Director responsible for the performance of STP-03-0. 10 CFR 50, Appendix B, Criterion V states, in part:
" Activities affecting quality shall be prescribed by
documented instructions, procedures, or drawings, of a type
appropriate to the circumstances and shall be accomplished
in accordance with these instructions, procedures, or
drawings...".
The failure to perform a subcriticality
check prior to loading fuel into control cell 20-41 as
required by STP-03-0, paragraph 7.3 is a violation of 10 CFR 50, Appendix B, Criterion V (461/86065-05D(DRP)).
(e) On October 8, 1986, two fuel bundles of the wrong
enrichment were designated on SNM Transfer Checklist 86-24
for placement into the reactor.
The inspector was present
in the control room when the error was identified by the
Shift Test Director during the positioning of fuel bundle
LY3213 over the reactor for insertion into the core. The
Shift Test Director advised the Shift Supervisor of the
fuel enrichment error and core alterations were suspended.
A review of approved SNM Transfer Checklist 86-24 in use
identified that two fuel bundles of medium enrichment had
been designated in steps 16 and 17 for placement into
reactor locat. ions requiring high enrichment type fuel. The
licensee initiated Condition Report (CR) 1-86-10-065 to
20
,
.
_ __
_
._
_
_
..
.
.
,-
'
.
-
-
.
wm
document this procedural error and provide corrective
-=
-
-
action. The licensee determined the cause of this event
24
was multiple personnel errors on the part of the
--
individuals who had initiated, checked, and approved SNM
3
Transfer Checklist 86-24 in accordance with CPS No.
I
1898.00, paragraph 8.4.
10 CFR 50, Appendix B, Criterion
,p
V states, in part:
" Activities affecting quality shell
M
be prescribed by documented instructions, procedures, or
A,
drawings, of a type appropriate to the circumstances and
M
shall be accomplished in accordance with these
'l-
instructions, procedures, or drawings...".
The failure
@
of personnel who had initiated, checked, and approved SNM
-
Transfer Checklist 86-24 to prescribe the correct fuel
A
,
bundle enrichment type is a violation of 10 CFR 50,
@
Appendix B, Criterion V (461/86065-05E(DRP)).
' ;
-
m
(f) On October 11, 1986, two fuel bundles were loaded into the
ji:
reactor without verification by the control room SRO, Shift
+
Test Engineer, and Shift Test Director that the fuel loaded
was the proper enrichment type. As part of the licensee's
corrective action for previous fuel loading errors.
memorandum DLH-86-312, dated October 8, 1986, was
2-
incorporated into STP-03-0 (Fuel Load Procedure).
This
%1
memorandum directed that the control room SRO, Shi*t Test
R_
Engineer, and Shift Test Director verify and approve that
y
the fuel to be moved was the proper enrichment type prior
'"
to movement of fuel into each centrol cell.
Fuel bundles
Sg
LY3596 and LY3711 were loaded into the reactor before
igit
verification of fuel enrichment was performed in accordance
f
with memorandum DLH-86-312. The licensee initiated
==
Condition Report (CR) 1-86-10-095 to document this
Q
procedural error and provide corrective action.
Fuel
-;_;
bundles LY3596 and LY3711 were verified to be of the
E2
correct enrichment in accordance with the DLH-86-312
1.
memorandum before the licensee proceeded to load the next
e
fuel cell.
10 CFR 50, Appendix B, Criterion V states, in
Z
part:
" Activities affecting quality shall be prescribed by
p
documented instructions, procedures, or drawings, of a type
-
appropriate to the circumstances and shall be accomplished
- 7
in accordance with these instructions, procedures, or
1.
drawings...".
The failure of the control room SRO, the
=
Shift Test Engineer, and the Shift Test Director to verify
$
proper fuel bundle enrichments, in accordance with
4-
memorandum DLH-86-312, prior to placement of fuel bundles
2
LY3596 and LY3711 into the reactor is a violation of
'f
-
10 CFR 50, Appendix B, Criterion V (461/86065-05F(DRP)).
gr
(g) On October 14, 1986, two fuel bundles were again loaded
into the reactor without prior verification by the control
d
room SRO, Shift Test Engineer, and Shift Test Director.
A
The two bundles were verified to be of the correct
E
enrichment type after the licensee identified the failure
i
Y
l
T
.
21
g=4
. _ . . _
.
.
.,
-
.
,
.'
,
-to-follow the directions of memorandum DLH-86-312
.l
~
.
(described'in' item f. above) and before continuing fuel-
'
'
. loading of the next' control cell. . The licensee' initiated
i
LCondition Report.(CR) 1-86-10-135_to document the
~: procedural error' and provide corrective action.
10-CFR 50, Appendix _B, Criterion.V states, in part:
" Activities affecting -quality shall .be prescribed by
documented instructions, procedures, or drawings, of a
type appropriate to the circumstances and shall be
accompitshed in accordance with these instructions,
procedures, or drawings...".
The failure of the control
room SRO, the Shift Test Engineer, and the Shift Test
Director to verify proper enrichments, in accordance
with memorandum DLH-86-312, prior to placement of fuel
bundles LY3189 and LY3320'into the Lreactor is a violation
of 10CFR50, Appendix B, Criterion V (461/86065-05G(DRP)).
The inspector noted that the Shift Test Director's'(STD)
log entries for this procedural error did not properly
reflect the sequence of events. Specifically, the log.
entries on page 103 of _the STD log for-STP-03-0 for time
0231,.0232, and 0235 indicated that all procedural-
'
-requirements nad been met prior to insertion of fuel
bundles LY3189 and LY3320 into the reactor. The inspector
did note that the Shift Test Engineer's log entries for
the procedural error accurately reflected the sequence'of
'
events.
In addition, the licensee had initiated Condition
Report (CR) 1-86-10-135 when the procedural error was
initially identified. The inspector _ discussed the need for
i
accurate records with cognizant licensee personnel. A
subsequent log entry to the STD log for STP-03-0 was made
to accurately detail the sequence of events.
In addition,
memorandum DLH-86-316, dated October 20, 1986, was
,
distributed to all Startup Test Directors emphasizing the
need for accurate log keeping. Based on the review of the
Shift Test Engineers log entries for October 14, 1986, the
i
licensee's initiation of Condition Report (CR) 1-86-10-135,
subsequent log entries made into the STD log on October 14,
1986, and memorandum DLH 86-316, the inspector concluded-
that the licensee had taken. appropriate' action to resolve
the inspector's concerns.
,
'
(h) On October 16, 1986,- an incorrect fuel bundle was
transferred from the 755' elevation of the Fuel Building
to the 828' elevation of Containment. The Special Nuclear
~
Material (SNM) Transfer Checklist in use (86-82) designated
movement of fuel bundle No. LY3409 from 755' Fuel Building
to 828' Containment.
Fuel bundle No. LY3445 was loaded
into the Inclined Fuel Transfer carriage in error and moved
from 755' Fuel Building to 828' Containment. The fuel
movement error was identified by the Senior Reactor
Operator (SRO) in charge of Fuel Handling. The incorrect
-
22
,
-
._ _ _ _ _ . , _ . _ _ . _ _ _ _ - _ . _ __ _ _
-
'
- .
-
.
-
.
'
.
"
'
fuel bundle (LY3445) was placed in storage at an 828'.
Containment storage location and the correct fuel bundle
+
(LY3409) was subsequently moved from the 755' Fuel Building
to 828' Containment in accordance.with approved SNM
Transfer Checklist 86-82. The licensee initiated Condition-
Report-(CR) 1-86-10-142.to document the procedural error
and to provide corrective action. 10CFR50,' Appendix B,
Criterion V states, in part: " Activities affecting quality-
shall be_ prescribed by documented instructions, and shall
be accomplished in accordance with these instructions,
procedures, or drawings...". The failure to follow the
SNM transfer checklist for movement of fuel bundles between
the 755' Fuel Building and the 828' Containment as required
by CPS No. 1898.00, paragraph'8.4.7, is a violation of
10 CFR 50, Appendix B, Criterion V-(461/86065-05H(DRP)).
The above. identified procedural violations that occurred during
initial fuel loading were generally minor in nature when
considered individually. However, the repetitive nature of-
several of the violations and the number of violations that
occurred over a relatively short period of time,. represent a
trend in performance that must be corrected. This matter was
.
discussed between NRC management and the licensee in a meeting
documented in paragraph 13 of this report.
Following completion of fuel loading, the inspector reviewed
the fully loaded core geometry and fuel bundle locations and
-
verified that the initial fuel core had been properly loaded.
This review was performed by comparing completed checklists
CPS No. 2209.01C001, " Orientation / Location Verification
Checklist" to completed checklist CPS 2209.01C002,"QA
Verification Checklist".
In addition, these checklists were
compared to video tapes No. 1, 4, 5, and 9 which recorded the
.
fuel bundle orientation, location in the core, and the serial
!
number of each fuel bundle. The inspector also reviewed proper
fuel enrichment location by comparison of NRC Form 741 reports
.
to the licensee's Plant Staff Nuclear Record Core Map. No
!
discrepancies were noted during this review.
b.
STP-05-0, Control Rod Drive System Test Witnessing (72302)
L
The inspectors witnessed portions of procedure step 7.1, Individual
CRD Functional Tests, performed in conjunction with STP-03. For
the portion of the test witnessed, the inspector verified that the
'
procedure in use was the most recent revision; that the test crew
was adequate, knowledgeable of the test requirements, and observed
the requirements of the procedure; that test data was recorded as
'
required using calibrated test equipment; and that preliminary test
results were evaluated against appropriate acceptance standards. The
inspector observed that coordination between the two tests (STP-03
and STP-05-0) was good and that the interface with the control room
operators was well established and properly controlled.
4
23
,.
,
,
-
. . . .
-. . - - - - . -
. .-
-
. . . -
-
-- ,
- -. - -- ,---.,-.- ----
_ _ _
,
.
-
.
j
-
.
.
.
One violation with eight examples was identified.
-
-
12 .- Onsite Followup of Events at Operating Reactors (93702)
a.
General
The inspector performed onsite followup activities for events which
occurred during the inspection period.
Followup inspection included
reviews of operating logs, procedures, condition reports, direct
observation of licensee actions, and interviews of licensee
personnel.
For each event, the inspector reviewed one or more of
the following:
the sequence of actions; the functioning of safety
systems required by plant conditions; licensee actions to verify
consistency.with plant procedures and license conditions; and
attempted to verify the nature of the event. Additionally, in some
cases, the inspector verified that licensee investigation had
identified root causes of equipment malfunctions and/or personnel
errors and were taking or had taken appropriate corrective actions.
Details of the events and licensee corrective actions noted during
the inspector's followup are provided in paragraph b. below.
b.
Details
(1) Reactor Water Cleanup Isolation (ENS 06695).
At approximately 3:30 a.m. on October 9, 1986, the reactor
water cleanup (RT) system automatically isolated due-to a high
differential flow signal in the Leak Detection system. At the
time of the isolation, the_ plant operators were restoring the
RT system to service after an interlock between the standby
liquid control (SC) system and the RT pump suction isolation
valve caused the suction valve to shut during performance of
an SC surveillance test.
Plant operators apparently did not
recognize that the RT system piping had partially drained to
the main condenser while the pump suction was isolated. When
the operators restored the RT pump to service, the leak
detection system saw a high flow rate into the RT system with
a low flow rate out of the system while the system piping was
refilling. This condition resulted in the isolation.
The licensee notified the NRC Operations Center of the above
event via the Emergency Notification System (ENS) at 6:45 a.m.
the same day. Condition report 1-86-10-085 and a licensee
event report (LER) were initiated to provide for appropriate
corrective action and reporting of this event.
Initial
investigation by the licensee indicated that the RT system
operating procedure was the primary cause of the event. The
LER will be reviewed upon receipt.
(2) Reactor Trip Due to High Main Steam Line Radiation Signal
(ENS Nos. 06567 an.d 06574).
24
,
-- - - - _ _ .
,
-
-
-
-
-
.
.
-
.
4
.
.
-
At approximately 12:30 a.m. and again at approximately 1:00 p.m.
on October 15. 1986, while operating in mode 5 (initial fuel
'
load), a reactor trip signal was received from the main steam
line (MSL) radiation monitors. At the time of each occurrence,
MSL radiation monitor channel D was inoperable and in a tripped
condition when a spurious trip signal was received on MSL
radiation monitor channel B.
This satisfied the 2 out of 4
coincident logic and resulted in a reactor trip signal.
The licensee notified the NRC Operations Center of the above
events via the Emergency Notification System (ENS) at
approximately 2:00 a.m. and 5:00 p.m. respectively. Condition
Reports (CRs) No. 1-86-10-126 and No. 1-86-10-137 and a
Licensee Event Report (LER) were initiated to provide
appropriate corrective action and reporting of the event.
The LER will be reviewed when it is issued.
Following the second reactor trip due to the MSL radiation
monitors, the licensee noted that Intermediate Range Monitors
(IRM) Channels A and G were tripped upscale.
In evaluating
the cause for the IRM' upscale trips, the licensee decided to
perform a manual reactor trip in mode 5 with all control rods
fully inserted as part of their troubleshooting effort for the
IRM readings. On October 15, 1986, at approximately 6:00 p.m.,
a manual reactor trip was performed. The performance of this
reactor trip was noted in the Shift Supervisor's Log as a
" Pre-Planned Manual Scram".
However, the Shift Supervisor
was unable to provide the inspector with either the procedure
that was followed to perform the manual trip or an approved
troubleshooting plan. The inspector identified through
discussions with the licensee that the manual trip was conducted
for testing under the verbal direction of the Manager - CPS.
The performance of a manual trip for testing without an approved
procedure or troubleshooting plan is a violation of CPS
No. 1011.01, Test Programs and Control, paragraph 2.1 which
requires that performance tests, modification tests, and
special tests be conducted using approved procedures and is a
violation of 10 CFR 50, Appendix B, Criterion V (461/86065-06A).
This is more significant due to the direct involvement of the
Manager - CPS.
(3) Containment Isolation Valve Closure (ENS Nos. 06568 and 06552).
At approximately 5:15 a.m. on October 14, 1986, while operating
in mode 5 (initial fuel loading), two isolation valves were
inadvertently actuated during the performance of a surveillance
test. Containment and Drywell isolation valves IIA 005 and
IIA 008 actuated closed when surveillance procedure CPS No.
9056.01, " Automatic Actuation of ADS (DIV 1), revision 20,
dated May 3, 1986, was attempted to be performed.
.
25
.. , . . - .
.
..
-
- - . ..
.-
.-
-
..
.
.
-
.
- .a
-
,
.
.
'
'
-
The purpose of surveillance CPS No. 9056.01 was to simulate
- -
automatic actuation and logic system functional test for-
" "
. Division 1.of the Aut'omatic Depressurization System (ADS).
~
. Satisfactory completion of this surveillance was to satisfy
CPS Technical Specifications 4.5.1.e.2 and 4.3.3.2 for
Division 1 of ADS. This surveillance had a frequency
requirement of at least once per 18 months.
,
Performance of surveillance procedure CPS No. . 9056.01.was'
1
authorized by the Shift Supervisor at about 1:30 a.m. on
October 14, 1986. .During the installation of transmitter
simulators per paragraph 8.1.39 at about 5:15 a.m., containment
isolation valve IIA 005 and drywell isolation valve IIA 008
actuated closed. The prior performance of paragraph.8.1.15
l.
was supposed:to have prevented both of these valves-from
actuating. Paragraph 8.1.15 directed the lifting of a wire
at terminal PPP69 in panel H13-P862 to prevent closure of
valves IIA 005, IIA 008, ICC072,'1CC073, and opening of valves
ISM 001A and ISM 002A. The wire at terminal PPP69 was lifted
-
in accordance with paragraph 8.1.15; however, this action
prevented only four of the six valves identified from actuating.
,;
10 CFR 50, Appendix B, Criterion V states, in part:
" Activities affecting quality shall be' prescribed by
documented instructions, procedures, or drawings, of a type
appropriate to the circumstances and shall be accomplished
in accordance with these. instructions, procedures, or
.
' drawings...". The failure to prescribe the proper wiro(s)
to be lifted during the performance of surveillance CPS
i .
No. 9056.01, paragraph 8.1.15 is a violation of 10 CFR 50,
- Appendix B, Criterion V (461/86065-04B(DRP)).
,
As described above, the signal that closed isolation valves
IIA 005 and IIA 008 occurred during the installation of
transmitter simulators in accordance with paragraph 8.1.39 of
CPS No. 9056.01. More specifically, the transmitter simulators
installed to Analogue Trip Modules (ATM) 821-N691A and ATM
B21-691E generated the closure signal. These two ATMs are the
<
two channels capable of providing a signal to the trip system
" Reactor Vessel Water Level - Level 1" (Division 1). When
paragraph 8.1.39 was performed at about 5:15 a.m. on October 14,
,
1986, both channelslof the Reactor Vessel Water Level - Level 1
(Division 1) trip system were made inoperable.
CPS Technical Spacification Table 3.3.2-1, Instrument 1.j.,
" Reactor Vessel Water Level - Level 1" requires a minimum
operable channels per trip system of 2.
The performance of
paragraph 8.l.39 as written was a violation of Technical
,
Specification Table 3.3.2-1, Instrument 1.3. minimum operable
i
channel requirement (461/86065-07A(DRP)).
3
Following the isolation valve closure of 5:15 a.m., the
licensee's immediate action was to stop the performance of
26
. . -
. --
- - _ - - . _ , . - . , . , . - . - .
-
. . . . - . . - .
. .
..
.-. ..
.-.--
.
. __
__
._
_ _ _
_ _
.
.
_
,-
.
-
.
,
-
.
'
.
surveillance: CPS No. 9056.01; placed ATM_1821-N691E-in-a
- -
-
.
pretest mode and reset-the trip-logic; and rcopened valves-
,
,
.
~1IA005 and IIA 008. The inspector was informed of the isolation
. valve closures and the licensee's intent.to inform.the'NRC
-
Operations-Center of the event-via the ENS at about.7:00 a.m.
,
on October'14. 'The licensee notified the NRC Operations Center
of the above event via the ENS at approximately 8:00 a.m. on
'
!
October 14, 1986.
Condition Report (CR) 1-86-10-117 and an
.LER were initiated _to provide appropriate corrective action
and reporting of the event. The LER will be reviewed when it
is issued.
At-8:11 a.m. on October 14, 1986, the-licensee suspended
specified condition CORE ALTER /" IONS to meet Action Statement-
No. 29 of; Technical Specification 'able 3.3.2-1, Instrument 1.~j.
.
This action was taken by the.. licensee-when it was recognized
'
that channel IB21-N691E was not meeting Table 3.3.2-1, note
"k" which7al10wed a channel to be placed in an inoperable.
i
status for up to_2 hours for required surveillance. At this
i
,'
time in the event, the following misapplication of the
'
Technical Specification occurred.
.
.
- (a) Recognition by the licensee that channel 1821-N691E had
I
,
been inoperable for greater than two hours was based-on the
'
isolation valve closure time of 5:15 a.m.
The. inspector
,
noted that paragraph 8.1.15 of surveillance-CPS No. 9056.01
directed the lifting of electrical leads to prevent
actuation of six isolation valves.
Hence, the-correct
'
performance of this paragraph. effectively makes the trip
channels inoperable. .The licensee authorized. performance
of surveillance CPS No. 9056.01 at about 1:30 a.m.
1
'
i
Sometime between the authorized start time (1:30 a.m.) and
'
the time when the isolation valves went closed (5:15 a.m.),
paragraph 8.1.15 -(lifting of leads) was performed.
,
Technical Specification Table 3.3.2-1, Note "k"
required
that a channel may be placed in an inoperable status for
up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for-required surveillance. 'The licensee
performed CORE ALTERATIONS with both channels (1821-N691A
and 1B21-N691E) for Reactor Vesse1~ Water Level - Low (Table
3.3.2-1, Instrument 1.J.) in an inoperable status
between 1:30 a.m. and 8:11 a.m. on October 14, 1986. This
is a violation of Technical Specification Table 3.3.2-1,
Note "k" (461/86065-07B(DRP)).
(b) As described in Condition Report (CR) 1-86-10-118, the
<
licensee had recognized that channel 1821-N691E was not
'
restored to an operable condition within the required two
hours. The licensee performed the action required by
. Action Statement 29 of Technical Specification Table
3.3.2-1.
The correct Action Statement indicated in Table
3.3.2-1 for instrument 1.j. in specified condition CORE
<
( -
[
27
-. 1
~ _____,_.._._ _ _ _._ _-.
-
.
- __
- _ _ _
'
L
,,
-
.
.
f
.
~
ALTERATIONS was Action Statement No. _25.
The performance-
l
-
-
-
,-
of Action Statement 29 (suspend CORE ALTERATIONS) was
.
considered _ appropriate by the licensee based on the
- instructions contained within _ the action statement. When-
'
. core alterations'were suspended, Action Statement:25
(establish 1 SECONDARY CONTAINMENT. INTEGRITY with the standby
.gasLtreatment system operating within-1 hour) was no longer
requi red.'
The-inspector noted-through discussions with
cognizant licensee personnel that a technical specification
revision was_being requested to clarify the applicability
of Action Statement No. 29 in.-Technical. Specification Table
3.3.2-1.
This will1 remain an Open. Item pending the-
inspector's_ review of the licensee's revised technical
specifications (461/86065-08(DRP)).
The licensee revised surveillance procedure CPS No. 9056.01 by
- use of Temporary Procedure Deviation (TPD) 86-1261. . This TPD
,
provided the necessary steps to restore channel B21-N691E to
licensee to identify that the other channel-(1821-691A)
still had a transmitter simulator installed. With the
transmitter simulator still installed on ATM 1821-691A, the
licensee reentered specified condition CORE ALTERATIONS _at ,
11:55 a.m. on October. 14, 1986. Technical Specification 3.0.4
states in part: ." Entry . into an Operational Condition or other -
specified condition shall not be made unless the conditions for
the . Limiting Condition for Operation are met without reliance
on provisions contained in the Action requirements." Entry
into specified condition CORE ALTERATIONS while not meeting
Technical Specification Table 3.3.2-1, Instrument 1.J., minimum
operable channel requirement of 2 is a violation of~ Technical Specification 3.0.4 (461/86065-07C(DRP)).
The licensee identified that the transmitter simulator was still
installed on ATM IB21-N691A at about 11:00 p.m. on October 14,
,
1986, and suspended core alterations. When the licensee removed
'
the transmitter simulator from ATM 1821-N691A at 11:35 p.m. on
October 14, 1986, isolation valves IIA 005 and IIA 008
-
automatically actuated shut. This was an unnecessary ESF
actuation since a temporary Procedure Deviation could have been
used similar to that used for channel B21-N691E. The-inspector
noted that a control room log entry, which may have been made
after the transmitter simulator was removed, identifiec; the
" anticipated" closure of isolation valves IIA 005 and IIA 008.
The licensee removed the transmitter simulator fron ATM
1821-N691A under the verbal direction of the Shift Supervisor.
When the transmitter simulator had been removed from ATM
IB21-N691E earlier that day, the licensee had prepared a
temporary procedure deviation (TPD 86-1261) to preclude the
unnecessary actuation of isolation valves IIA 005 and IIA 008,
10 CFR 50, Appendix B, Criterion V states, in part: " Activities
affecting quality shall be prescribed by documented instructions,
procedures, or drawings, of a type appropriate to the
i-
28
- -
i.
!'
_.
.
..m
._, ~,_- . . . . . . , , _ _ , _ _ . . . _ , _ _ . . _ . . . . _ , .
. . . _ _ _ , , . - , . , . , _ , _ .
. . _ , . . _
.-
- . . _ .
.
c
,
-
circumstances and shall be accomplished in accordance with
-
-
these instructions, procedures, or drawings...".
The failure
'
of the licensee to implement written instructions to remove
the transmitter simulator from ATM 1821-N691A when returning
the ATM to service on October 14, 1986, is a violation of
-10 CFR 50, Appendix B, Criterion V (461/86065-06B(DRP)).
The' licensee notified the NRC Operation Center of the violation
of Technical Specification Table 3.3.2-1 via the ENS at about
2:00 a.m. on October 15, 1986.
Condition Reports (CRs)
1-86-10-118, CR 1-86-10-125, and an LER were initiated to
provide appropriate corrective action and reporting of the
event. The LER will be reviewed when it is' issued.
The inspector noted that the-licensee reported the violation of
Technical Specification Table 3.3.2-1 via the ENS as a 24-hour
report required by paragraph 2.F. of Facility Operating License
NPF-55. -The required reporting requirements for Technical
Specification violations were discussed with the licensee as
documented in paragraph 6.c of this report.
(4) Shutdown Service Water Pump Auto Start (ENS. No. 06588).
At approximately 10:00 p.m. on October 16, 1980, Division 1
Shutdown Service Water Pump started unexpectedly during the
performance of a routine surveillance.
Surveillance procedure
CPS No. 9463.02. " Suppress. ion Pool Water Level LT"CM030(31),
LT-SM013(DG) Channel Calibration", revision 21, dated July.14,
1986, directed in Appendix A the lifting of leads from
Termination Module (TCM) 107. When this step was performed the
Division 1 Shutdown Service Water pump sensed a low pressure
signal from its associated strainer output pressure transmitter
(IPT-SX028). The automatic start was initiated due to this low
pressure signal.
The licensee notified the NRC Operations Center of the above
events via the ENS at approximately 11:30 p.m. on October 16,
1986. Condition Report (CR) 1-86-10-150 and an LER were
initiated to provide appropriate corrective action and reporting
of the event.
The LER will be reviewed when it is issued.
The licensee determined the cause for the event was the
procedure in use (CPS No. 9463.02) was in error. Appendix A to
the procedure incorrectly identified Termination Module 107 for
the lifting of leads. The leads that were to be lifted for
correct performance of this surveillance were located on
4
Tercination Module 104. The root cause for the procedural error
was the incorrect use of electrical drawing E02-1SM013, revision
A in the preparation of the procedure (procedural reference
No. 11.20).
10 CFR 50, Appendix B, Criterion V states in part:
" Activities affect.ing quality shall be prescribed by documented
instructions, procedures, or drawings, of a type appropriate to
29
.-
.
- . -
--
-.
.,
. = - .
_.
..-
_.
.
P
a
.,-
.
.
f.
-
x
.
.
-
W
'
the circumstances and'shall be accomplished.in accordance with?
-
'these instructions, procedures, or drawings...".
The failure-
-
,
- to-prescribe the' correct Termination Module in-surveillance
1
procedure CPS No. 9463.02, revision 21, is a violation of
10 CFR 50, Appendix B, Criterion _V (461/86065-04C(DRP)).
_
- (5) Reactor Trip Due To Personnel ' Error ~ (ENS =No. 06615).
b
At approximately 2:30 a.m. on October 19, 1986, while in mode 5
and performing core alterations (initial fuel;1oad), a reactor-
trip signal was initiated by personnel performing surveillanc'e
procedure CPS No. 9031.14, "IRM Channel-Functional", revision.
20, dated May 17, 1986. Procedural-Step 8.1.1 directed the
placement of Intermediate Range Monitor (IRM)-C drawer function
switch to " Standby".
Personnel performing the surveillance
mistakenly placed the Source Range Monitor (SRM)-C drawer .
function switch to " Standby".
Since-the RPS shorting links had
been removed for initial fuel loading, placement of SRM-C drawer
function switch to " Standby" satisfied the noncoincident
circuitry resulting in a reactor trip. At the time of
occurrence, all control rods were already fully inserted; no
control rod motion resulted from the reactor trip. The licensee-
returned the SRM-C drawer function switch to operate and' reset
the reactor trip.
,
The licensee notified the NRC Operations Center of the above-
event via the ENS at approximately 5:00 a.m. on October 19,
1986. Condition Report (CR) 1-86-10-171 and an LER were
initiated.to provide appropriate corrective action and reporting
of the event. The LER will be reviewed when it is issued.
.
10 CFR 50,' Appendix B,- Criterion V~ states, in parts
" Activities
affecting quality shall be prescribed by documented
instructions, procedures, or drawings, of a type appropriate to-
the circumstances and shall be accomplished-in accordance with-
these instructions, procedures, or drawings...".
The failure of
personnel performing surveillance procedure CPS No. 9031.14 to
operate the IRM-C drawer function switch in accordance with
procedural step 8.1.1 is a violation of 10 CFR 50, Appendix B,
Criterion V (461/86065-09A).
The inspector noted that the
licensee's immediate respons- to this event adequately
identified the cause (persennel error); the licensee documented
the event-(CR No. 1-86-10-171); and the licensee made
appropriate NRC notifications. No response from the licensee
to this violation is required.
(6) Standby Gas Treatment System Exhaust Fan Start (ENS V-
06640).
At approximately 11:30 p.m. on Octobe-
1, 1986, the licensee
identified that the Division 2 Standby Gas Treatment System (VG)
'
Exhaust Fan had started. The licensee determined that the cause
i
30
,
.--,--4
,
.-mm,--m,w---,=m
- , ,
,,,..-vi,-
,..-%,----,-,3-,,w---*-ee.-,
,e-.-e--
% -- ,, .
--,-,--=w----+-
3--r----+---,,
--- = - - - -
r
.
.
.
-
.
of_the fan start was the installation of a jumper and start
-
-
signal onto the Division 2 VG fan start relay. Preoperational
,
Test Procedure (PTP)-VQ-01, step 7.6.96 directed installation
of a jumper and signal onto relay IVAY-VG515D (VR/VQ LOCA relay).
Personnel performing this procedural step mistakenly placed a
jumper and signal onto relay 10AY-VG511F (VG Division 2 fan
start relay). When the expected actions of the PTP did not
occur, the licensee identified the running exhaust fan and the
improperly connected jumpers. _The licensee removed the jumpers
from the Division 2 VG exhaust fan relay (IUAY-VG511F) and
restored the Division 2 Standby Gas Treatment train to a standby
status. At the time of occurrence, the plant was in mode 5.
The licensee notified the NRC Operations Center of the above
event via the ENS at approximately 2:00 a.m. on October 22,
1986. Condition Report (CR) 1-86-10-201 and an LER were
initiated to provide appropriate corrective action and reporting
of the event. The LER will be reviewed when it is issued.
10 CFR 50, Appendix B, Criterion V states, in part: " Activities
affecting quality shall be prescribed by documented
instructions, procedures, or drawings, of a type appropriate to
the_ circumstances and shall be accomplished in accordance with
these instructions, procedures, or drawings...".
The failure _of
personnel performing Preoperational Test Procedure (PTP)-VQ-01
to install a jumper and signal in accordance with procedural
step 7.6.96 is a violatier, of 10 CFR 50, Appendix B, Criterion V
-(461/86065-09B).
The inspector noted that the licensee's
immediate response to this event adequately identified the cause
(personnel error); the licensee documented the event (CR No.
1-86-10-201 and PTP-VQ-01 log entry); and the licensee made
appropriate NRC notifications. No response from the licensee
to this violation is required.
(7) Loss of ENS Due to Power Supply Breaker Accidental Opening
(ENS 06634).
At about 10:30 a.m. on October 21, 1986, an electrician
accidentally tripped 18 circuit breakers while performing
maintenance on Auxiliary Building Motor Control Center 1A1 120
volt AC distribution panel. While removing the 120 volt AC
distribution panel cover, the electrician accidentally tripped
18 circuit breakers interrupting the 120 volt power supplies to
various plant equipment. The licensee determined that one of
the breakers tripped interrupted power to the emergency
notification system (ENS).
In addition, the licensee identified
that two containment isolation valves (1RE022 and 1RF022) had
actuated in the " failed closed" position due to a loss of power
to their associated air solenoids.
The inspector obse.rved the licensee's critique and noted that
the sequence of events were accurately reconstructed. The
31
g
.
<
.
.
licensee notified the NRC Operations Center via the ENS at
-
approximately 12:00 p.m. on October 21 of the loss of ENS.
,
An additional notification was made at approximately 1:00 p.m.
on October 21, 1986, on the closure of the two containment
isolation valves. Condition Report 1-86-10-182 and an LER were
initiated to provide appropriate corrective action and reporting
of the event. The LER will be reviewed when it is issued.
(8) Entry Into Specified Condition CORE ALTERATIONS While Relying
On TS Action Statement [ ENS No. 06652]
At approximately 4:00 p.m. on October 22, 1986, the licensee
identified that a violation of the CPS TS had occurred the
previous day. On October 21, 1986, the licensee was operating
in specified condition CORE ALTERATIONS while one train of the
Standby Gas Treatment (VG). System was inoperable.
[TS 3.6.6.3
requires that two independent standby gas treatment subsystems
be OPERABLE during CORE ALTERATIONS. With one train of VG
inoperable, TS action statement 3.6.6.3.a.2 (which requires that
the inoperable train be restored to OPERABLE status within 7
days or suspend CORE ALTERATIONS) applies.] The plant operators
were tracking the inoperable VG train as limiting condition for
operation (LCO) No. 86-10-22. At about 4:10 p.m. on October 21,
the plant operators reentered specified condition CORE
ALTERATIONS while relying on the action statement of the LC0
for the VG system. TS 3.0.4 states, in part, that entry into
an OPERATIONAL CONDITION or other specified condition shall not
be made unless the conditions for the Limiting Conditions for
Operation are met without reliance on the action requirements.
No exception to this TS was stated in the TS for the VG system.
The licensee's reentry into operational condition CORE
ALTERATIONS on October 21 at about 4:10 p.m. was in violation
of TS 3.0.4 (461/86065-07D(DRP)).
The licensee notified the NRC Operations Center of the above
event via the ENS at 3:55 p.m. on October 22. Condition Report
1-86-10-197 and an LER were initiated to provide appropriate
corrective action and reporting of this matter. The LER will
be reviewed when it is issued.
Four violations were identified.
13. Management Meeting (30702)
On October 22, 1986, NRC management met with IP management at the
Region III Office in Glen Ellyn, IL to discuss NRC concerns related to
Clinton fuel loading procedure violations and other events occurring since
issuance of the low power license. Key personnel attending the meetings
are identified by (#) in paragraph 1 of this report.
NRC concerns included the re.latively high number of reportable events that
occurred during the initial fuel loading process; the number of procedural
32
F
'
-
~ . .
- .
-
-
-
violations that had occurred duringEthe. fuel loading. process; and the
-~
-
adequacy of implementation of the . licensee's condition reporting system
,
,
to provide'for good root cause determinations and clear corrective
actions to preclude recurrence of the condition.
.The licensee presented information concerning each of the fuel loading
procedure violations and several of the events reported during the
-
initial fuel' loading process. The information' discussed by-the li_censee
' involved personnel errors, equipment malfunctions, design deficiencies,
,
and procedure; inadequacies that contributed to each event, as applicable.
<
Much of .the information discussed by the licensee was preliminary
information. Corrective actions identified by the licensee during the
meeting included.the following:
- a.
A Plant Manager's Standing Order was initiated to provide a detailed
evaluation of the impact of performance of certain surveillance tests
prior to their implementation.
b.
. Expanded and improved pre-shift briefings.
c.
Improved vertical communications.
d.
Review and revision.of the CPS Condition Reporting system by the
onsite- safety review committee (Facility Review. Group).
e.
_ Reduced reliance on personnel overtime.
f.
Management change in the'0perations Department.
g.
Improved QA interface with the operating and maintenance
organizations.
h.
Established a plant goal of zero personnel errors.
During the course of the meeting, there was an apparent difference in the
sequence, timing, and numbers of procedure violations as discussed by the
licensee as opposed to information available to the Region III staff.
Those differences were resolved during subsequent discussions between the
NRC resident inspector and licensee management.
At the conclusion of the meeting, Region III management reiterated the
need for timely identification of root causes and completion of corrective
action to prevent recurrence. The licensee acknowledged this.
14. Open Items
Open items are matters which have been discussed with the licensee, which
will be reviewed further by the inspector, and which will involve some
action on the part of the NRC or licensee or both. Two open items
disclosed during the inspection were discussed in paragraphs 7 and
12.b.(3)(b).
-
,
33
,
-
..
.
15. Unresolved Items
-
-
'
Unresolved items are matters about which more information is required in
order to ascertain whether they are acceptable items, violations, or
deviations. One unresolved item disclosed during this inspection was
discussed in paragraph 6.c.
16. . Exit Meetings (30703)
The inspector met with licensee representatives (denoted in paragraph 1)
throughout the inspection and at the conclusion of the inspectio1 on
November 4 and 7, 1986. The inspector stemmarized the scope and findings
of the inspection activities. The licensee acknowledged the inspection
findings. The licensee stated that a detailed analysis which included
many of the findings of this inspection had been performed and that a plan
of corrective action was being developed.
The inspectors also discussed the likely informational content of the
inspection report with regard to documents or processes reviewed by the
inspectors during the inspection.
The licensee did not identify any such
documents / processes as proprietary.
The resident irspectors attended exit meetings held between Region III
based inspectors and the licensee as follows:
Inspector (s)
Date
Falevits
10/30/86
Maura
11/4/86
34