IR 05000315/1988002
| ML17325A607 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 02/09/1988 |
| From: | Burgess B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17325A606 | List: |
| References | |
| 50-315-88-02, 50-315-88-2, 50-316-88-03, 50-316-88-3, IEB-87-002, IEB-87-2, NUDOCS 8802180425 | |
| Download: ML17325A607 (23) | |
Text
U.
S.
NUCLEAR REGULATORY COMMISSION
REGION III
Repor ts No. 50-315/88002(DRP);
50-316/88003(DRP)
Docket Nos.
50-315; 50-316 Licensee:
Indiana Michigan Power Company 1 Riverside Plaza Columbus, OH 43216 Licenses No.
Donald C.
Cook Nuclear Power Plant, Units 1 and
Inspection At:
Donald C.
Cook Site, Bridgman, Michigan Inspection Conducted:
December 15, 1987 through January 26, 1988 Inspectors:
B.
L. Jorgensen J.
K.
H lier Approved By:
.
L. Burgess, C ief Projects Section 2A Date Ins ection Summar Ins ection on December
1987 throu h Januar
1987 Re orts No.
50-315 88002 DRP 50-316 88003 DRP Areas Ins ected:
Routine unannounced inspection by the resident inspectors of: actions on previously identified items; plant operations; reactor trips; radiological controls; maintenance; surveillance; fire protection; security; outages; management programs; reportable events; and Bulletins, Notices and Generic Letters.
Results:
Of the twelve areas inspected, no violations or deviations were identified in. any areas.
88021804~5 8802 PDR ADOCK 05000315 PDR
DETAILS 1.
Persons Contacted
"W. Smith, Jr., Plant Manager
"A. Blind, Assistant Plant Manager - Administration J.
Rutkowski, Assistant Plant Manager - Production L. Gibson, Assistant Plant Manager - Technical Support
"B. Svensson, Licensing Activity Coordinator
"K. Baker, Operations Superintendent
"J.
Sampson, Safety and Assessment Superintendent E. Morse, equality Control Supervisor T. Bei lman, I8C/Planning Superintendent
"J. Droste, Maintenance Superintendent
"T. Postlewait, Technical Superintendent
- Engineering L. Matthias, Administrative Superintendent
- M. Horvath, equality Assurance Supervisor D.
Loope, Radiation Protection Supervisor
- J. Kauffman, Construction Manager The inspector also contacted a number of other licensee and contract employees and informally interviewed operations, maintenance, and technical personnel.
- Denotes some of the personnel attending the Management Interview on January 27, 1988.
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2.
Actions on Previousl Identified Items a ~
(Open) Violation (316/83004-01)
and Open Item (316/83004-04):
As previously updated in Inspection Reports 50-316/85016(DRP),
50-316/85022(DRP),
and 50-316/86008(DRP),
the test program could not demonstrate containment spray additive system operability.
By letter dated March 4, 1987, (Wigginton to Dolan)
NRC requested additional information on the spray additive system.
By letter (AEP:NRC:0914D) dated April 2, 1987,- the licensee provided the requested information.
This item will be closed when NRC acts on the Technical Specification change request.
b.
(Closed) Violation (316/83004-02):
Failure to take corrective action for~an identified test deficiency.
The problems identified occurred between 1978 and 1983.
At that time, the administrative procedures for,'!writing and performance of surveillance tests did not provide personnel with adequate guidance.
Since that time the administrative procedures have been rewritten and notes define the steps necessary to write, perform and review/evaluate the test data.
Because of the age of this violation and revisions made to the administrative procedures, the inspector did not attempt to'verify the licensee response (AEP:NRC:0831) dated August 24, 1983, but instead attempted to determine if the current administrative procedures would have prevented the violation; no problems were identifie C.
(Closed)
Open Item (316/83004-03):
The licensee's program for check valve testing may not assure that the differential pressure testing requirements of ASME Section XI (1974 Edition)
Paragraph IWV-'520 (B)(2) are performed.
The licensee has subsequently committed to the 1983 Edition of Section XI, which restates the requirements of Paragraph IN/-3520 (B)(2).
The D.
C.
Cook Nuclear Plant IST valve program (Revision 2, dated August 31, 1987) lists in Paragraph 1.c approved alternative testing performed on check valves (as indicated under relief request).
This item is closed since the current program describes acceptable approved alternative testing methods.
(Open) Unresolved Item (316/85029-03):
Ventilation systems are not testable per ANSI N510-1975 as referenced in various sections of the Technical Specifications but are testable per ANSI N510-1980.
The licensee has submitted a Technical Specification change request (AEP:NRC:0959) dated May 28, 1987.
Additional information was provided by AEP: NRC: 0959A on November ll, 1987.
This item will be closed when NRC acts on the licensee's Technical Specification change request.
(Closed) Violation (316/86022-01):
Four Heise gauges were left installed on Unit 1 because the procedure did not have controls governing restoration.
This violation was closed in Inspection Report 50-315/87014(DRP)
for Unit 1.
Since the same corrective actions also apply to Unit 2, this item is closed.
g.
)
(Closed) Violation (316/86025-03):
The time limit for operation with one accumulator having low level was exceeded.
The installed level instrument had failed and as a compensatory measure, the licensee had verified tank level by ultrasonic inspection.
Due to a communication error, technicians failed to detect a valid low level.
The licensee response (AEP:NRC: 1006) dated September 6, 1986, describes the corrective action and the results achieved.
In addition, the defective level transmitter was repaired during a
subsequent outage.
(Closed) Violation (316/86029-01):
Failures to follow approved procedures.
The examples involved failure to implement controls for maintenance and temporary modifications and to report discovery of adverse conditions.
Two of the examples were incorporated in.'.Enforcement Action EA 86-150, which was transmitted to the 1;icensee on November 18, 1986 as Violations I.A. 1 and I.A.2.
The their'd'xample (not. reporting adverse conditions as prescribed)
was incorporated in EA 86-150 as Violation II.
A civil penalty of
$25,000.00 was proposed for Violation I (which included I.A.1, I.A.2 and - as discussed below - I.B) but no penalty was proposed for Violation II.
The licensee responded to the Violation and paid the civil penalty via his letter (AEP: NRC: 1009) dated December 18, 1986.
The corrective and preventive steps described therein have been verified and have proven effectiv (Closed) Violation (316/86029-02):
Limiting Conditions for Operation and Technical Specification "Action Statements" not met.
These items were a consequence of the procedural violations above, which led to a Unit startup with two of four power range nuclear instruments incorrectly wired.
They were incorporated in EA 86-150 as Violation I.B and constituted part of the basis for the
$25,000.00 civil penalty.
As was the case for the previous item, licensee corrective and preventive actions, as described in AEP:NRC: 1009, have been verified and have proven effective.
(Closed) Violation (316/86029-03):
Failure to notify NRC, pursuant to 10 CFR 50.72, when the incorrect wiring of two of four power range nuclear instruments was "discovered".
This item was not incorporated in EA 86-150.
No Notice of Violation was issued because the "discovery" was by an instrument technician who would not be considered to be at a level of responsibility that would oblige him to report to NRC.
However, the instrument technician also failed to report to other "responsible" licensee personnel, a
failure which was cited in the Notice of Violation (Item II) issued with EA 86-150.
When "responsible" licensee personnel (e.g.,
those at a level obliged to report to NRC) learned of the details of the event, reporting to NRC occurred within the specified time limits.
Since a separate Violation on this matter proved inappropriate, this item is being withdrawn to correct the record.
(Closed) Violation (316/86030-01):
A Licensee Event Report was not issued when main steam relief valve setpoints were found outside the Technical Specification range.
The surveillance procedure provided instructions on performance of the test, evaluation of the results, and resolution of out-of-specification setpoints, but was silent on reportability requirements.
The procedure has been revised and a
Licensee Event Report has been issued.
The licensee's response (AEP:NRC: 1013) dated November 7, 1986 provided the information in response to the Notice of Violation by reference to the Licensee Event Report.
In a letter dated November 16, 1986, Region III requested that the licensee consolidate future responses into one document containing the complete written statement or explanation in reply.
(Closed)
Unresolved Item (315/86030-02; 316/86030-02):
Pump suction pressure gauge installation for the containment spray pumps are inc'onsistent between the two units.
The two items identified (gael'ded versus threaded pump casing pipe connections and questionable installation of pressure gauges)
were reviewed by both corporate and plant personnel and documented on Condition Report 12-09-86-1052.
The welded versus threaded casing pipe connections are consistent with construction practices at the time of installation; both are acceptable.
The questionable gauges were removed.
Current practice requires installation of the gauges prior to each use and removal subsequent to us h e
(Closed) Violation (315/86042-01):
Maintenance activities on the high pressure safety injection system reduced the injection flow path from four loop injection (assumed in the safety analysis)
to two loop injection.
The licensee response letter (AEP:NRC: 1022A)
dated April 10, 1987 acknowledged the violation, paid the associated
$50,000.00 civil penalty and provided the corrective action to resolve and prevent a repeat violation.
The licensee is working with NRC to provide a safety analysis that would address less than four loop injection.
Meanwhile, the licensee has placed operator aids at the appropriate controls, warning of the consequences of flow path isolation.
In the response letter, the licensee acknowledged that this problem apples to the low pressure safety injection (LPSI) system and has taken similar steps to assure that LPSI flow paths are not reduced.
No violations, deviations, unresolved or open items were identified.
3.
0 erational Safet Verification Routine facility operating activities were observed as conducted in the plant and from the main control rooms.
Plant startup, steady power operation, plant shutdown, and system(s)
lineup and operation were observed as applicable.
The performance of licensed Reactor Operators and Senior Reactor Operators, of Shift Technical Advisors, and of auxiliary equipment operators was observed and evaluated including procedure use and adherence, records and logs, communications, shift/duty turnover, and the degree of professionalism of control room activities.
Evaluation, corrective action, and response for off normal conditions or events, if any, were examined.
This included compliance to any reporting requirements.
Observations of the control room monitors, indicators, and recorders were made to verify the operability of emergency systems, radiation monitoring systems and nuclear reactor protection systems, as applicable.
Reviews of surveillance, equipment condition, and tagout logs were conducted.
Proper return to service of selected components was verified.
a ~
Unit-1 operated at its 90-percent power administrative level throughout the inspection period with two significant exceptions:
f)',.'<i"'On January ll, 1988 the No.
2 main turbine control valve unexpectedly closed - this is discussed further in Paragraph 3.c below.
ii)
On January 13, 1988 the reactor tripped on inadvertent opening of the "B" reactor trip breaker - this is discussed further in Paragraph 4 belo Unit 2 operated at its 80-percent administrative level throughout the inspection period with a single significant exception when the licensee made a planned power reduction to about 10-percent power on December 23, 1987 to permit inspection of and addition of oil to the No.
23 reactor coolant pump.
The inspector observed portions of this evolution, which was uneventful.
At about 11: 18 a.m.
on January ll, 1988, Unit 1 main turbine control valve No.
2 unexpectedly "drifted" closed.
Control rods were immediately taken from "Manual" to "Auto", and they responded to reduce reactor power.
An approximate 160 MWe load decrease occurred as the control valve went full-closed.
During the valve closure transient, the steam dump system actuated to balance primary and secondary system energy transfer until the control rods stabilized reactor power at the reduced load.
Steam generator level control remained in "Auto" and responded smoothly.
Inspector review of this event identified no adverse operator or equipment information.
The No.
2 control valve was intentionally disabled in the closed position and the Unit was returned to 90-percent power on "partial arc admission",
pending investigation and a decision on further course of action.
The problem was traced to a mal-functioning EHC servo-valve, suspected of having a plugged nozzle strainer.
When the Unit subsequently tripped on January 13 from an unrelated cause (see Paragraph 4 below) the EHC servo-valve was replaced.
No further problems were experienced in subsequent operations.
The inspector continued the practice of performing weekly facility tours with licensee management representatives at the Assistant Plant Manager or Superintendent level.
During this inspection period, these tours identified only minor housekeeping needs which were noted for correction by the licensee representatives and which the inspector subsequently verified to have been corrected.
Examples of items noted on independent inspector tours of the facility and referred to the licensee were:
i)
The security door for the Technical Support Center would not relatch (December 31, 1987);
ii) shift turnover from the afternoon to the midnight shift (January 14, 1988)
was unnecessarily affected by the noise
-,,""--'nd the moving of the power cord underfoot of the operators while turnover was in progress.
I The licensee responded positively to address these items.
The inspector reviewed selected operating and alarm procedures associated with the spent fuel systems.
i)
""12 OHP 4021.018.002
"Placing in Service and Operating the Spent Fuel Pit Cooling System",
Revision 5 dated October 22, 198 In addition to the objective stated in the title, this procedure also contains instructions for operating the cleanup system (filter and demineralizer)
for the spent fuel pit, and for increasing or decreasing spent fuel pit, water level as desired.
The procedure does not stipulate that steps within each section must be performed in sequence, but some evolutions will not work unless they are.
No cases involving potential damage were apparent should steps be performed out of sequence.
This was discussed with the Operations Department procedure group, who stated a separate administrative requirement exists (Procedure PMI-2010, Section 3. 1. 1)
mandating double asterisk ("") procedures must always be performed
"step by step".
i)
2-OHP 4024.205
"Annunciator No.
5 Response
- Containment Spray" Revision 2 dated April 27, 1982 through Change Sheet
dated October 9, 1984.
ii)
1-OHP 4024. 105 "Annunciator No.
5 Response
- Containment Spray" Revision 2 dated April 20, 1982 through Change Sheet
dated August 28, 1987.
Each of the above procedures addresses three "shared" spent fuel pit alarms (system abnormal, level low, temperature high), the instructions for which were reviewed in detail.
The inspector noted the annunciator response instructions do not identify applicable procedures to be used to accomplish the stated objective; i.e., the operator is instructed to raise water level (annunciator 027) or to adjust cooling (annunciator 028) without reference to the approved procedure (""12-OHP 4021.018.002 noted above) which exists for this purpose.
This was also discussed with the Operations procedure group.
They stated their practice is to limit cross-referencing from annunciator response instructions.
Only Emergency or Abnormal procedures are normally cross-referenced; routine system operating procedures are not.
g.
The inspector examined the spent fuel pool cooling system using system description CH-109, procedure
""12-OHP 4021.018.002
"Placing in Service and Operating the Spent Fuel Pit Cooling System", print OP-5136
"Spent Fuel Pit Cooling and Clean-up",
and valve lineup sheets
""12-OHP 4021.018.002 Data Sheets 1 and 2.
No problems were identi fied.
No violations, deviations, unresolved or open items were identified.
Reactor Tri s
Unit 1 tripped from 90-percent power at 8:21 a.m.
on January 13, 1988, due to a personnel error that caused the "B" reactor trip breaker to open.
Monthly surveillance test
- ~1-THP 4030 STP.411,
"Train B Reactor Trip SSPS and Trip Breaker Train B" was in progress.
The first steps of STP.411 require racking-in the "B" bypass breaker.
However, the assigned
licensed operator attempted to rack-in the already closed
"B" Reactor Trip Breaker.
The operator's first attempt was unsuccessful.
During a second attempt he bumped/depressed the breaker locking device, which caused the breaker to open and trip the reactor.
The resident inspectors were onsite and responded to the control room.
Activities there were busy, but orderly and unconfused, such that the Unit was rapidly stabilized and required system performance verified.
A preliminary review of the post-trip information showed all systems responded properly.
The licensee's tour of the containment and review of a pre-planned forced outage schedule did not identify any item that prevented reactor startup.
The reactor was made critical at 4:35 a.m.
and returned to service at 10:32 a.m.
on January 14, 1988.
Additional reviews will be performed during close out of the associated Licensee Event Report.
No violations, deviations, unresolved or open items were identified.
5.
Radiolo ical Controls During routine tours of radiologically controlled plant facilities or areas, the inspector observed occupational radiation safety practices by the radiation protection staff and other workers.
Effluent releases were routinely checked, including examination of on-line recorder traces and proper operation of automatic monitoring equipment.
Independent surveys were performed in various radiologically controlled areas.
Using a licensee controlled E-130 Radiation Monitor, the inspector routinely confirmed posting requirements.
a 0 A high-sensitivity personnel contamination detector
"alarmed" on December 31, 1987 as the inspector was exiting the auxiliary building following-an extended tour.
A security guard exiting at the same time also received an "alarm".
He had been in the building only a few moments.
Whole-body frisking with a hand-held detector failed to locate a source of contamination on either individual.
Subsequently (within 15 minutes) the apparently generally distributed source had decayed/disappeared.
Consultation with Radiation Protection personnel at the time and thereafter suggested aspic'ollows:
i)
occasional very small, transient sources of short-lived radiogases develop about the building, sometimes associated with Containment-Building entries/exits (both individuals in this case had been in the area of the Unit 1 lower containment personnel hatch immediately prior to exiting);
ii) slight radiogas
"contamination" has become a too-familiar experience for security personnel, to the point they have requested uniforms made of cotton or some material other than the current polyester that would reduce the likelihood of gaseous activity clinging to the uniform.
The inspector reviewed licensee practices with respect to "securing" high radiation and extreme high radiation areas
"from unauthorized entry".
A steel mesh enclosure was erected to secure a radwaste evaporator room from entry (see 5.d below) because part of a wall had been tom out to gain access for maintenance involving evaporator disassembly.
The enclosure appeared to provide substantial security to this high radiation area - more so than original design provides in the case of an extreme high radiation area nearby, which would be accessible by climbing over an approximate six-foot wall.
The Technical Specifications (TS) were referred to, as NRC has issued TS in lieu of 10 CFR 20 for D.
C.
Cook.
Region III radiological protection staff were also consulted.
The culmination was that no prescribed industry-wide standard exists concerning
"how secure is secure enough?" for radiological control.
The question is reviewed on an individual plant basis; it has been reviewed before at D.
C.
Cook, and will be again, by a Region III expert.
During tours of the auxiliary building the inspector attempted to
.
enter an extreme high radiation (EHR) area by use of the card reader system.
EHR areas as defined by D.
C. Cook's Administrative Procedure PMP-6010 RAD.002 are secured by a lock and key system independent of the card reader system.
A couple of attempts to the Unit 1 and Unit 2 seal injection rooms were made within a five minute period.
In addition to being denied access the attempts were alarmed on a computer terminal located at the radiation access control facility.
These alarms prompted a search by Radiation Protection personnel to identify if anyone had entered the room and interview the person attempting to enter the room.
When the card reader is used the name and badge number is available.
During the interview the inspector was informed it was common practice to confirm attempted entries to any EHR.
Condition'eport 12-01-88-0016 (Problem Report 88-0005) identified that a door watch to a High Radiation (HR) area was inattentive.
High'.,Radiation areas as defined by D.
C.
Cook s Administrative Procedure PMP-6010 Rad.002 are secured by a lock and key system (wliich may be the card reader system) or by posting of a door watch.
The area involved was to the south radioactive waste evaporator, noted in 5.b above.
The door watch was replaced by construction of the steel mesh enclosure.
The inspector has reviewed this Condition Report and submitted it to Region III Radiation Protection personnel for review during a subsequent inspection.
The inspector interviewed a contractor and confirmed that he was cognizant of radiation work practices (see Paragraph 6.f-Maintenance).
No violations, deviations, unresolved or open items were identified.
6.
Maintenance Maintenance activities in the plant were routinely inspected, including both corrective maintenance (repairs)
and preventive maintenance.
Mechanical, electrical, and instrument and control group maintenance activities were included as available.
The focus of the inspection was to assure the maintenance activities reviewed 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: the Limiting Conditions for Operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures; and post maintenance testing was performed as applicable.
The following activities were inspected:
a.
Spent fuel pool heat exchanger pump "N" overhaul; b.
"S" (radwaste)
evaporator tube replacement and evaporator reassembly; C.
d.
e.
continuing auxiliary building structural steel and crane rail upgrading-i)
installation of a temporary work platform atop the spent fuel pool bridge to permit access to portions of the North wall immediately adjacent to the pool, ii)
temporary installation of a steel mesh net over the pool, iii) removal of old crane rail segments and their replacement, and iv)
continued large-scale welding of support steel.
Construction of a redesigned auxiliary building access control area (ref: Job Order No.
726953, Plant Modification 12-PM-626);
Lubrication of the Unit 2 North and South control rod drive mechanism motor generator set (ref: Job Order No.
LO-0180).
This job required entry into a carbon dioxide (COq) protected area.
Plant administrative procedures require that the CO> be isolated (for personnel protection) prior to entry into the room and posting of a fire watch or if the activity will take less than twenty-minutes, a member of the crew may be designated as a fire watch.
For this activity, the second option was used.
The inspector entered the COq protected area after the work had started and left the work area (but stayed in the CO> protected area)
before the work was complete.
When the crew exited the CO< area, the sign-in board was reviewed and it was identified that the inspector was still in the CO> protected area.
A member of the crew found the
inspector and informed him that the twenty-minutes time limit would expire and that a fire watch was required if he desired to stay in the room.
The crews activities were discussed with the fire protection coordinator and reviewed against the fire protection administrative procedures; the crew's actions were found to be appropriate and correct.
f.
Restoration of a double wall to the South radioactive waste evaporator.
The inspector interviewed the mason (contractor employee)
and confirmed that he was working to an RWP and wearing the proper personnel radiation monitoring devices.
While the first wall was being restored, the inspector attempted to enter the room through an open door without using the card reader.
The inspector was stopped by the mason and informed that he was required to verify entry into the room.
Subsequent reviews of PMP-6010 RAD.002 showed that the mason's actions were appropriate.
No violations, deviations, unresolved or open items were identified.
7.
Surveillance The inspector reviewed Technical Specifications required surveillance testing as described below and verified that testing was performed in accordance with adequate procedures, that test instrumentation was calibrated, that Limiting Conditions for Operation were met, that removal and restoration of the affected components were properly accomplished, that test results conformed with Technical Specifications and procedure requirements and were reviewed by personnel other than the individual directing the test, and that deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel.
The following activities were inspected:
a ~
""2-OHP 4030.027AB and ""2-OHP 4030 STP.027CD.
b.
These items were inspected after the "AB" diesel failed a start test on January 18, 1988 under the routine monthly surveillance test STP.027AB.
The Technical Specifications require the opposite diesel (in this case, the
"CD" diesel)
be verified OPERABLE, which the licensee did utilizing test STP.027CD.
The "AB" failure was documented for followup in Condition Report 2-1-88-0059.
~gg',
""1 THP 6030 IMP.035 and ~"1-THP 6030 IMP.036.
These items were reviewed to verify gA Surveillance findings regarding failure to perform selected seismic monitoring instrumentation calibrations within specified intervals in the pre-1987 time period.
The focus of the review was on whether current controls would prevent a recurrence.
More followup is anticipated upon receipt of the associated Licensee Event Repor C.
""1-THP 4030 STP.411 "Reactor Trip SSPS logic and Reactor Trip Breaker Train "B" Surveillance Test (Monthly).
d.
e.
The procedure was reviewed subsequent to the reactor trip of January 13, 1988 (See Paragraph 4, Reactor Trips).
The inspector found that the procedure identified the correct breaker and that local labeling and the language in this procedure were consistent.
- "2-THP 4030 STP.511
"Reactor Trip SSPS Logic and Reactor Trip Breaker Train "B" Surveillance Test (Monthly)".
The inspector observed performance of this Unit 2 test (subsequent to the Unit 1 reactor trip of January 13);
no problems were noted.
"*2-THP 4030 STP.101 "Reactor Coolant Flow Protection Set I Survei l lance Test (Monthly)".
On two occasions the inspector met with Instrument and Control (I8C)
group management.
Both meetings were on licensee initiative.
The first, on Oecember 15, 1987, involved a discussion of the status of 18C crew performance evaluations and possible options for performance enhancement.
The second, on January 7, 1988, involved additional details on performance enhancement plans.
NRC has previously expressed concerns in the area of error-free completion of I8C group surveillance testing and maintenance activities, and will remain interested in the licensee's enhancement efforts in this area.
No violations, deviations, unresolved or open items were identified.
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8.
Fire Protection Fire protection program activities, including fire prevention and other activities associated with maintaining capability for early detection and suppression of postulated fires, were examined.
Plant cleanliness, with a focus on control of combustibles and on maintaining continuous ready access to fire fighting equipment and materials, was included in the items evaluated.
a ~
b.
The inspector interviewed selected maintenance personnel and found they were knowledgeable of the fire protection program (see Paragraph G.e, maintenance).
An<'apparent design discrepancy was identified by the licensee and reported; it involved Local Shutdown Indicating (LSI) cabinets placed in the plant, remote from the main control rooms to meet
CFR 50 Appendix R safe shutdown requirements for postulated control room fires.
When the capabilities of the LSI cabinets themselves (to withstand and isolate fire effects)
came into question, they were declared inoperable and fire watches assigned.
An associated Licensee Event Report is being covered under Inspection Report No. 50-315/88003(ORS);
50-316/88004(ORS).
The inspector verified that fire watch coverage was established and ascertained adequate instructions were given to fire watch
personnel.
The licensee initially established an hourly tour coverage, then changed to "continuous" (inspected at least each 15 minutes)
on further consultation with NRC Region III)
No violations, deviations, unresolved or open items were identified.
~Securi t Routine facility security measures, including control of access for vehicles, packages and personnel, were observed.
Performance of dedicated physical security equipment was verified during inspections in various plant areas.
The activities of the professional security force in maintaining facility security protection were occasionally examined or reviewed, and interviews were occasionally conducted with security force members.
An example of an inattentive security guard was identified by the licensee and reported as required.
As this has been a recur rent problem, it prompted a special security inspection by NRC Region III.
The results are documented in Inspection Report 50-315/88005(ORSS);
50-316/88006(ORSS)
.
No violations, deviations, unresolved or open items were identified.
Outa<ues Subsequent to the reactor trip of January 13, 1988 (see Paragraph 4,
Reactor Trip) the outage planning group was instrumental in planning the approximate 18-hour window that occurred between the reactor trip and completion of the required pre-criticality surveillance activities.
The group conducted a minimum of three status meetings which assured the optimum number of noncritical items were performed without impacting the startup and confirmed that the forced outage schedule (licensee program for scheduling maintenance activities requiring pre-established plant conditions) did not list any items that would prohibit startup.
No violations, deviations, unresolved or open items were identified.
Mana ement Pro rams The effectiveness of management controls, verification and oversight activiti'es, in the conduct of jobs observed during this inspection, was eval uated'<"-.
The inspector frequently attended management and supervisory meetings involving plant status and plans and focusing on proper co-ordination among Oepartments.
The results of licensee auditing and corrective action programs were routinely monitored by attendance at Problem Assessment Group (PAG)
meetings and by review of Condition Reports, and Problem Reports.
As
applicable, corrective action program documents were forwarded to NRC Region III technical specialists for information and possible followup evaluation.
The inspection included a general review of initiatives by the licensee onsite to realign his organization in ways which should enhance quality of performance.
Three specific organizational changes were included.
The most substantive change was a mid-1986 realignment among departments associated with the creation of three Assistant Plant Manager (APM)
positions.
The three APMs were assigned responsibilities for production, technical support, and organization/administration, respectively.
The Production Department combined, for the first time, operating, maintenance (including the Instrument and Control group)
and planning functions.
The Technical Support APM acquired physical sciences (chemistry and radiation protection),
computer sciences, and technical
"engineering" (two testing sections and the nuclear engineering group).
Organization and Administration was comprised of all the "non-technical" support functions: training, security, personnel, accounting, quality control; and the Shift Technical Advisor (STA) group.
Concurrent with this change, a Licensing Activities Coordinator position, reporting directly to the Plant Manager, was created.
The inspector has observed the functioning of the APM structure as described above and considers it to have resulted in improved understanding of assigned responsibilities and authorities, better communications, stronger focus on problem prevention and problem solving, and increased teamwork.
A second change in mid-1987 involved a realignment within Organization and Administration which consolidated several sections into a new Safety and Assessment Department.
The effect has been to bring together (geographically as well as figuratively) all the plant-assigned staff with independent safety or assessment responsibilities.
Included are the STA's (who perform various operating experience reviews and administer the corrective action and trending programs) fire prevention/protection functions, and quality control (the guality Assurance organization remains fully independent of the plant organization and management).
The inspector has observed this refinement to be beneficial in escalating independent assessment to a level (a Department, headed by a Superintendent)
of equivalent stature as Operations or Maintenance.
The third change (at the beginning of 1988) reviewed as part of this inspection involved sub-tier realignments within Technical Support, such that the>-"Performance Engineering Section was split into two sections-one emphasizing mechanical performance (i.e., performance testing and engineering)
and one emphasizing electrical/controls performance.
The inspector considers this realignment a small positive step in the development of "system engineers" onsite.
Ho violations, deviations, unresolved or open items were identifie.
Re ortable Events The inspector reviewed the following Licensee Event Reports (LERs) by means of direct observation, discussions with licensee personnel, and review of records.
The review addressed compliance to reporting requirements and, as applicable, that immediate corrective action and appropriate action to prevent recurrence had been accomplished.
(Closed)
LERs 315/83079 Revision 0 with Supplement, and LER 316/83077 Revision 0: Control Room Fire Detectors not Smoke Tested.
Testing conducted in March 1982 failed, so the licensee decided to install additional detectors (AEP:NRC:0670 dated April 7, 1982).
Performance of the new installation was not demonstrated by successful smoke testing as of August 1983, when NRC challenged operability of the systems.
The licensee declared the detectors for each control room inoperable and issued the subject LERs.
Successful Unit 1 testing was completed by September 15, 1983 (AEP: NRC:0670F dated October ll, 1985) which was also deemed applicable for Unit 2 as. determined by the Corporate Office Mechanical Engineering Division.
Individual detectors continue to be routinely tested.
Further, both control rooms are continuously occupied, making an undetected fire extremely unlikely regardless of the status of the electronic detection systems.
b.
(Closed)
LER 316/82098 Revision 0:
KV Fire Detection System Inoperable.
A detector failed and alarmed, negating further alarm capability of other detectors in the same zone.
Compensatory fire watches were established until a
new detector had been installed and verified to be operating properly and the system was returned to service.
(Closed)
LER 316/83004 Revision 0 with Supplement:
Containment Spray System Eductor Flow below Technical Specification Minimum.
Eductor valve positions were established during a preoperational test, using the containment sump (vice the spray additive tank)
as the suction source, at a flow of 20 + gpm.
Amendment
subsequently changed the required flow rate to greater than
gpm bdt4~I'ess than 50.
The next 5-year test using the spray additive tank'4for suction then found flow to be below 20 gpm.
The eductor va/ves were opened fully and flow was demonstrated within limits in a final test.
An "as-found" test of Unit 1 showed flows to be within limits for that Unit.
Licensee evaluation of the significance of educted flow below 20 gpm has concluded automatic eduction is not vital, so the educted flow rate is not safety significant in the initial containment spray phase.
Pending Technical Specification revisions, as well as
"open" questions concerning containment spray, are discussed in Paragraph 2. a, above.
(Closed)
LER 316/83054 Revision 0: Fire Ooor Inoperable.
Latch component failures on two occasions a week apart rendered fire door No.
386 incapable of latching properly, though it would close.
The failures were repaired and the door tested satisfactorily.
Routine hardware failures of this type no longer require reporting to NRC.
(Closed)
LER 316/84005 Revision 0: Containment Type B and C Leak Test.
The report discussed an occasion in early 1984 involving cumulative local leak rate test results apparently in excess of 0.60 L(a).
The licensee subsequently withdrew the LER (letter dated Oecember 12, 1984)
on the basis that penetration leakage through two series valves will be no greater than the lower individual valve leakage rate.
NRC Region III prescribes the
"Maximum Pathway Leakage" methodology for local leak testing of this kind, where leakage is conservatively equated to the larger leakage rate of a pair of individually quantified series valves.
It is not known whether application of the
"Maximum Pathway Leakage" methodology would have yielded total leak results above or below 0.60 L(a).
That question, however, bears only on reportability, since the licensee uses much more conservative individual valve and penetration leakage limits as his basis for requiring repair.
In the subject case, penetrations/
valves exhibiting leakage in excess of these conservative criteria were repaired and retested, resulting in "as-left" leakage known to be well below 0.60 L(a).
Further, the licensee tracks and trends penetration/valves individually, such that the 0.60 L(a) criterion is substantially less limiting.
Though the situation has not recurred where methodology has potentially affected pass/fail against the 0.60 L(a) criteria, use of the
"Maximum Pathway Leakage" methodology in NRC Region III was discussed at the Management Interview.
(Closed)
LER 315/84006, Revision 0: Blocked Fire Ooor.
Fire door No.
230 was found obstructed by an air hose strung to support maintenance by a contractor; yet no firewatch was present.
The firewatch had been released at the end of the work day without the obstruction being cleared also.
The subject door is normally open but held by fusible links.
Appropriate administrative measures were; taken with respect to the work foreman who erred, and this has not+been a recurrent problem.
This event is representative of a violation of Technical Specification requirements which the licensee identified, reported and corrected, which has not recurred, and which lacked safety significance.
As such, no NRC Notice of Violation is being issued for this item.
No violations, deviations, unresolved or open items were identified.
13.
NRC Com liance Bulletins Notices and Generic Letters (Closed)
NRC Compliance Bulletin 87-02:
"Fastener Testing to Determine Conformance with Applicable Material Specifications".
This Bulletin described circumstances involving a potential for misapplication of fasteners because of their failure to conform to specifications.
The licensee was required to address:
i)
a description of his receipt inspection program/activities relating to fasteners, and how he controls issue from and return to stock internally after receipt; ii)
the results of prescribed testing of a sample of fasteners from stock - the sample to be as outlined in the Bulletin - along with support information such as names and addresses of suppliers; and, iii) a description of any discrepancies arising from the testing, including safety significance and, if appropriate, corrective actions.
The licensee responded by letter (AEP:NRC: 1045) dated January 12, 1988.
The inspector verified the information contained in the description responsive to i) above, participated in the selection of the samples and audited the identifying data associated with ii) above against his notes, and reviewed the description of further action described pursuant to iii).
The inspector had no further questions concerning the matter.
No violations, deviations, unresolved or open items were identified.
20'ana ement Interview The inspectors met with licensee representatives (denoted in Paragraph 1)
on January 27, 1988 to discuss the scope and findings of the inspection.
In addition, the inspector asked those in attendance whether they considered any of the items discussed to contain information exempt from disclosure.
No items were identified.
The following items were specifically discussed:
a.
the orderly and effective response of plant operators to a Unit 1 tur'bine control valve closure (Paragraph 3.c)
and to a Unit 1 re'ichor trip (Paragraph 4);
J b.
potentially chronic problems with security guard clothing retention of radioactive gases (Paragraph 5.a);
C.
approaches for NRC review of a reported (LER 315/87023)
Local Shutdown Indicating (LSI) cabinet design deficiency (Paragraph 8);
and, NRC Region-III technical position for the
"Maximum Pathway Leakage" method in certain applications of local containment penetration leakrate testing (Paragraph 12.e).
17