IR 05000220/1993001
| ML20059M035 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 11/12/1993 |
| From: | Boyle M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20059M030 | List: |
| References | |
| 50-220-93-21, 50-410-93-20, NUDOCS 9311180125 | |
| Download: ML20059M035 (16) | |
Text
.
.-
-.
-
.
.
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report Nos.:
93-21 93-20 Docket Nos.:
50-220
'
50-410 License Nos.:
DPR-63 l
'
Licensee:
Niagara Mohawk Power Corporation 301 Plainfield Road Syracuse, New York 13212 Facility:
Nine Mile Point, Units I and 2
,
'
Location:
Scriba, New York Dates:
September 19 - October 30,1993 Inspectors:
B. S. Norris, Senior Resident Inspector R. A. Plasse, Resident inspector
_:
W. F. Mattingly, Resident Inspector
!
Approved by:
{ulu
.
//-td -93 hiich5el L. Boyle', @ing@hief
- Date #
Reactor Projects Section No. l A
-
Division of Reactor Projects
!
!
,
t
,
931118012b 931110
PDR ADOCK 05000220 b
G pop j
-
-
.-
... - -.
.
...
.
.
.--
.
._
Y
.
.
.
,
.
SUMMARY i
Nine Mile Point Units I and 2
,
50-220/93-21 & 50-410/93-20 i
PLANT OPERATIONS Unit 1 operated at essentially full power throughout the period. Unit 2 commenced a refueling
,
outage, ending a continuous run of 326 days of safe power operations. Unit 2 established a goal of 60 days for completion of the outage; as of the end of the report period, they were on schedule with no major delays. A review of the Unit 1 Station Operations Review Committee
determined that the meetings were well controlled, comprehensive, and focused on safety.
'
L MAINTENANCE Maintenance and surveillance activities observed and reviewed by-the inspectors were l
'
satisfactory. Due to previous problems in the maintenance area, Niagara Mohawk Power Corporation (NMPC) developed maintenance performance principles to provide standard expectations and objectives for maintenance personnel. To reinforce the principles, coaching teams were developed to accompany craft personnel in the 5cid; this appeared to foster teamwork and improve performance. Unit 2 performed an in-vessel visual inspection (IVVI)
.
on a representative sample of the shroud weld population. The inspection found no indications of cracking. NMPC has established an excellent temporary modification program with proper
,
-
controls and resources in place to mmirnize plant impact.
i ENGINEERING Prior to starting the outage, Unit 2 removed the Erst of two layers of reactor cavity shield plugs
while at power. This allowed for better management of critical path activities and potentially
will reduce critical path outage time by eight hours. The safety evaluation adequately analyzed the activity for 'ooth structural and radiological considerations.
l
\\
PLANT SUPPORT
>
On a routine basis, performance in the areas of radiological controls, security, and emergency
preparedness was acceptable. Two issues were of note for exceptional performance. In the first
,
issue, a small quantity of natural uranium yellowcake was spilled in the corporate of6ces of j
'
Niagara Mohawk. Although this is not under the license of either Nine Mile unit, Nine Mile
!
Point radiological protection personnel responded to contain and clean up the spill in a timely
,
f i
manner. In the second issue, in response to a General Electric communication regarding the potential for high radiation levels in the drywell during fuel movements, Un'it 2 implemented
'
controls to minimize exposures not only for fuel movements but also for similar activities, such as removal of reactor control blades, local power range monitors, and other non-fuel irradiated components. The inspectors noted that the offsite emergency response facilities were well maintained and appropriately supplied, with radiation monitoring equipment in calibration.
,
i iii
-.
.
-
.
_
.~.~
_ _, _
.-
.
,
__
. _ _
.
.
_
_.
.j
.
.
DETAILS i
1.0 SU3151ARY OF FACILITY ACTIVITIES The Niagara Mohawk Power Corporation (NMPC) safely operated Nine Mile Point Unit 1 at
essentially full power throughout the period. Unit 2 started the period at full power; a refueling
'l
,
outaga began on October 1, ending a continuous run of 326 days of safe power operations.
l Resident inspectors conducted inspection activities during normal, backshift and weekend hours
over this period. There were 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> of backshift (evening shift) and 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> of deep
backshift (weekend, holiday, and midnight shift) inspection during this period. In addition,
specialist inspections were conducted in the areas of physical security, environmental monitoring,
!
and occupational exposure. The results of the specialist inspections will be reported separately.
2.0 PLANT OPERATIONS (60710,71707,90712,93702)*
l 2.1 Operational Safety Verification The inspectors observed overall plant operation and verified that the licensee operated the plant I
safely and in accordance with procedures and regulatory requirements. The inspectors conducted l
regular tours of the following plant areas
,
Control Room
- Reactor Building
- Control Building
Turbine Building
- Switchgear Rooms
- Access Control Points
- Refuel Floor
- Service Water Bays
- Diesel Generator Rooms j
Unit 2 Drywell
- Protected Area Perimeter i
-
!
,
The inspectors observed plant conditions through control room tours to verify proper alignment
of engineered safety features; to verify that operator response to alarm conditions was in
,
accordance with plant operating procedures; to verify compliance with Technical Specifications j
(TSs), including implementation of appropriate action statements for equipment out of service;
-
and to review logs and records to determine if entries were accurate and identify equipment l
status or deficiencies. These records included operating logs, turnover sheets, and system safety
{
tags.
.
The inspectors conducted detailed walkdowns of accessible areas :o inspect major components and systems for leakage, alignment, lubrication, cooling water supply, and general conditions
that might prevent fulfillment of their safety function.
The inspector observed plant i
housekeeping controls including control and storage of flammable material and other potential
!
!
safety hazards.
j
?
The inspector found that shift turnovers were comprehensive and accurate, and adequately i
"
reflected plant activities and status.
Control room operators effectively monitored plant l
operating conditions and made necessary adjustments. Housekeeping was commensurate with
{
I
The NRC inspection manual procedure or temporary instruction that was used as
inspection guidance is listed for each applicable report section.
)
_
m.
. -. _
-
.
.
.. _ -
.
.
2 ongoing work. The inspector concluded that NMPC conducted overall plant operations in a safe
)
and conservative manner.
'
2.2 Station Operations Review Committee
The inspector attended Unit 1 Station Operations Review Committee (SORC) meetings on l
September 21, October 5, and 19. Presentations on NMPC's review of NRC Bulletin 93-03, j
" Reactor Vessel Water Level Instrumentation," root cause analysis on the fire in the auxiliary t
control room, and various deviation / event report (DER) resolutions were well presented. The
inspector found the SORC meetings to be well controlled, comprehensive, and focused on safety.
- The inspector also verified that committee member composition and quorum requirements as
,
specified in the TSs were satisfied.
>
3.0 MAINTEN ANCE (60710, 61726, 62703, 90700, 90712)
l 3.1 Maintenance Observations j
i I
The inspectors observed maintenance activities to ascertain if safety related activities were being.
conducted according to approved procedures, TSs, and appropriate industry codes and standards.
Observation of activities and review of records verified that:
required administrative i
authorizations and tag outs were obtained, procedures were adequate, certified parts and
materials were used, test equipment was calibrated, radiological requirements were implemented, system prints and wire removal documentation were used, and quality control hold points were established. Maintenance activities observed included:
!
>
N2-MPM-GEN-R901 Disassembly of the reactor pressure vessel (wet-lift)
l N2-FHP-13.1 Complete core offload N2-FHP-10 LPRM removal and installation N2-FHP-21-Control rod uncoupling, removal, and installation N2-MMP-CSH-212 High pressure core spray (HPCS) injection valve overhaul WO 93-00358 2BYS* BAT 2A intercell connectors resistance checks following the
'
Division I battery replacement
!
'
WO 93-01092 Redundant reactivity control system troubleshooting WO 93-01773 Outboard main steam isolation valve 2 MSS *AOV7C machining i
WR 12-05007 Repair of flex conduit to valve 39-07 WO 93-04608 Replace relay 80-115-4-128
'
WO 93-01850 Repair CS pipe support base plate Breach Perm.it 1316 Excavation of electrical penetration to support cable routmg The above activities were effective with respect to meeting the. safety objectives.
{
3.1.1 Unit 2 Core Shroud Inspection In 1990, crack indications were reported at core shroud welds located in the beltline region of -
,
h nn-.
-- -,,
.,,,..
>-- -
- -,
., -,
,_
-
--
_
_
.
_
_
.
-
,
.
j
1 t
an overseas boiling-water reactor (BWR). As a result of this discovery, General Electric (GE)
issued Rapid Information Communication Services Information Letter (RICSIL) 054, " Core
Support Shroud Crack Indications," on October 3,1990, to all owners of GE BWRs. The
RICSIL summarized the cracking found in the overseas reactor and recommended that at the
~;
next refueling outage plants with high carbon type 304 stainless steel shrouds perform a visual examination of the accessible areas of the seam welds and associated heat affected zones on the
inside and outside surfaces of the shroud.
'
Based on the GE recommendations, a domestic BWR performed visual inspections of their shroud in July 1993 and discovered cracks in the weld regions. GE subsequently issued
.
Revision 1 to RICSIL 054 on July.21,1993, to update the information on the core support
.;
shroud cracks and to provide interim recommendations to perform visual examinations of
,
'
accessible areas of the shroud at all GE BWRs during the next scheduled outage. The Quality Assurance (QA) department reviewed the revision and considered it to be not applicable to Unit
i 2 "since both occurrences pertain to type 304 stainless steel in a GE BWR/4." Unit 2 is a GE-
BWR/5 with type 304L (low carbon content) stainless steel. Therefore, no provisions were
made to perform a visual examination of the shroud during the next outage scheduled to start
on October 1.
,
On September 30, the NRC issued Information Notice (IN) 93-79, " Core Shroud Cracking at
'
I Beltline Region Welds in Boiling-Water Reactors," to inform the holders of BWR operating licenses of ite cracking observed in the weld regions of the core shroud.
j
,
On October 1, GE issued Services Information Letter (SIL) 572, " Core Shroud Cracks," which
!
superseded RICSIL 054, to provide an overview of the situation and recommendations 'on l
suitable inspection techniques and frequency to detect cracking that could lead to structural j
integrity concerns. Revision 1 to the SIL was issued on October 4. GE concluded in the SIL
that the cracks in the shroud were typically intergranular and caused by either intergranular
stress corrosion cracking, irradiation assisted stress corrosion cracking, or a combination of
'
both. In addition, the cracking is believed to be aggravate.d by the following: high neutron j
fluence and long hot operating time; fabrication related surface cold work; high material carbon
content; unfavorably oriented material inclusions; weld residual stresses; and elevated coolant conductivity. GE recommended a visual examination of the accessible areas of the shroud.
.l during the first refueling outage after eight or more years of effective full power operations for j
all plants with type 304L stainless steel shrouds; Unit 2 has 3.4 effective full power years of l
operation.
i Although a shroud inspection was not required, based on the above information and the current ~
j plant conditions (core off-loaded), Unit 2 management decided to perform an in-vessel visual.
l inspection (IVVI) on a representative sample of the shroud weld population. At the end of this
inspection period, NMPC completed the partial IVVI consistent with GE's inspection j
recommendations and found no indications of cracking.
l
!
'
i
!
.
--.
.
.. -. -.
-. -
.
.
_.
_
(
,
.
i l
'
.i The inspector reviewed NMPC's response to the RICSIL, SIL, and IN, and considered the i
disposition to RICSIL 054 revision 1 to be inappropriate based on the information available at the time. QA supervision agreed with the inspector's assessment and initiated a DER to evaluate their review ofindustry operating experiences. The inspector considered this action appropriate.
l The inspector also reviewed the IVVI results and found them satisfactory. NMPC's final
decision to inspect the shroud was prudent and demonstrated the proper safety perspective.
3.1.2 Maintenance Performance Principles Review
.
In response to past performance errors in the maintenance area, NMPC developed Maintenance Performance Principics. The principles were developed and reviewed by maintenance craft and management personnel. The intent of this guidance was to provide standard expectations and
,
objectives to be communicated and emphasized through training, observation, and feedback with
'
maintenance personnel. Major areas included: maintenance standards and practices; pre-job briefings / post-job critiques; procedure adherence; self and peer verification; safety; radiological controls; proper use of tools; communication; documentation; training and qualification; plant
,
'
material condition; professionalism; and self-assessment.
The inspector reviewed the
'
maintenance performance principles and determined the document provided detailed fundamental guidance. The inspector considered this effort to be a significant initiative in improving plant performance in the maintenance area.
t The maintenance department has periodic group meetings on major topic areas to reinforce the maintenance principles. The inspector observed a discussion on " radiological controls." The
,
inspector assessed the discussion to be an open dialogue addressing various radiological controls
"
-
problems in the field. A radiological control supervisor facilitated the meeting and responded
well to the concerns addressed by the craft personnel.
To reinforce the maintenance principles reviewed and discussed, NMPC developed self.
assessment " coaching teams." The coaching team usually consists of one or two experienced
!
maintenance or training personnel who accompany craft personnel in the field. The inspector
!
observed a coaching team and concluded this initiative to be positive in fostering teamwork and performance improvement. In summary, the inspector determined the maintenance performance i
principles, open discussions, and coaching teams to be worthwhile initiatives with respect to j
improving overall maintenarice performance.
3.1.3 Unit 1 Temporary Modifications Review
,
The inspector reviewed administrative procedure GAP-DES-03, " Control of Temporary q
Modifications."
The purpose of this procedure is to establish controls for the review,
'
implementation, and clearance of temporary modifications.
The inspector discussed the temporary modification program with the Unit 1 technical support manager and the operations '
-f general manager. The manager of technical support maintains administrative control of the.
'
temporary modifications program and provides implementation coordination of temporary modifications utilizing system engineers. The inspector reviewed the outstanding temporary -
!
l t
,
!
.
_
_
,
.-
....
,
i
!
l i
.
i
modifications and verified 10 CFR 50.59 evaluations were completed when required. At the time of the review Unit I had 28 temporary modifications installed, with 12 of these temporary
'
modifications planned to be installed permanently. The majority of the remaining temporary modifications were installed in radwaste systems. The inspector determined all temporary l
modifications had a responsible system engineer and an estimated completion date. The j
inspector reviewed all existing temporary modifications and identified no adverse impact to'
l safety systems. It appeared to the inspector that NMPC had established an exec 11ent temporary i
modification program with proper controls and resources in place to minimize plant impact due l
'
i i
3.2 Surveillance Observations
!
Through observation of safety-related surveillance activities, interviews, and review of records,
.
j the inspectors verified: use of proper administrative approved, personnel adherence to procedure i
precautions and limitations, accurate and timely review of test data, conformance of surveillances j
to TSs, including required frequencies, and use of good radiological controls. Surveillance l
activities observed included those listed and discussed below.
!
N1-ISP-002-002 Turbine Anticipatory Trip Test Low Oil Pressure Instrument
Channel Calibration N1-ISP-036-006 Emergency Cooling System - High Steam Flow Instrument Trip l
Channel Test N1-ISP-002-013 Turbine Anticipatory Trip-Bypass Reactor Scram Instrument j
Channel Calibration
!
N1-ST-Ql A Core Spray Loop II Pump and Valve Operability Test
,
N2-MSP-EGS-R002 High pressure core spray (HPCS) diesel generator (EDG)
!
inspection
_
N2-OSP-CSL-R002 Low pressure core spray (LPCS) valve position indication l
opembility test-l N2-OSP-CSL-R001 LPCS pressure isolation valve leakage test
.;
N2-OSP-EGS-R002 Division I EDG 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> run and load reject test
!
N2-OSP-EGS-R004 Division I EDG simulated loss of offsite power with ECCS l
initiation N2-CSP-EGS-R008 HPCS EDG simulated loss of offsite power with an ECCS initiation The above activities were effective with respect to meeting the safety objectives.
,
!
4.0 ENG IN EERING (37700, 40500, 71707, 90700, 90712)
.i
_!
t 4.1 Reactor Cavity Shield Plug Removal at Power
i As part of the shutdown to enter the refueling outage, Unit 2 personnel removed the first of two layers of reactor cavity shield plugs at about 30% power. This was performed at power
,
i i
l
!
.-
- -
- -
- >
-
-
--
__..
-= -
-_
..
.
.
..
i
.
J
.
,
,
(procedurally less than 40% power) to better manage critical path activities and reduce critical path outage time by about eight hours. This activity is unique to Unit 2 because of its
!
interlocked and tapered shield plug design and is not practical for Unit 1, however, it has been
,
performed at several other utilities.
i The shield plugs are installed to comply with 10 CFR 20 radiation protection requirements during routine plant operations. No credit is taken for the shield plugs to limit radiological conditions below the limits of 10 CFR 100 or 10 CFR 50 Appendix A GDC 19. The removal of the first layer of shield plugs at 40% reactor power does not affect the structural integrity of the shielding because the remaining bottom layer of shield plugs can withstand the design seismic and missile loads. No other cr(dit for accident mitigation is considered in the design basis analysis. The shield plugs are physically located over the drywell head and are part of the
.
I refueling floor when installed.
.
Prior to the outage, Unit 2 engineering personnel perfor'aed a safety evaluation per 10 CFR
!
50.59 for removal of reactor cavity shield plugs A, B, C, # D at 40% or less reactor power.
l The safety evaluation analyzed the activity for both structural and radiological considerations and was reviewed and approved by the on-site SORC. In addition, an as low as is reasonably achievable (ALARA) benefit and impact review was prepared. The ALARA review evaluated f
'
the expected impact upon worker and public exposures, examined the benefits relative to exposure impact, and identified ALARA issues that could be encountered.
The inspector reviewed the USAR, TSs, safety evaluation, ALARA benefit and impact report,
,
and the off-site Safety Review and Audit Board's (SRAB) review of the safety evaluation. The inspector also interviewed the engineer and several health physicists involved with the shield plug removal and discussed the SRAB comments on the safety evaluation with the SRAB technical coordinator. Based on the above the inspector considered all aspects of the reactor
>
cavity shield plug removal at power satisfactory.
}
-
l 4.2 Review of Licensee Event Reports and Special Reports i
The inspector reviewed the followMg licensee event reports (LER) and special reports to verify
[
that they conformed to the requirements specified in 10 CFR 50.73 and TSs.
These requirements include a proper nrrative description of the event, the cause of the event, an
)
assessment of the safety consequeces, and corrective actions.
.
I ER 50-220/93-007. Loss of Offsite Power due to Lichting Strike
,
Inspection report 50-220/93-18 reviewed and assessed the event.
The inspector
.
determined that the LER contained the necessary information.
'
i LER 50-410/93-001-01. Actuation of Encineered Safety Feature due to Eauipment
Failure
,
j
l
_
-
_ _
,
.
_
__ _,
.
.
.-
.
.
P
,
The LER corrected the root cause for an ESF actuation, based upon updated information obtained during subsequent troubleshooting. The LER also detailed additional corrective -
actions. This event was discussed in detail in inspection report 50-410/93-18. The
!
inspector determined that the supplement provided the information.
l
+
Unit 2 Special Report (dated October 7.1993) Concerning the Inonerability of the
,
Radwaste/ Reactor Buildine Ventilation Gaseous Effluent Monitorine System (GEMS)
[
NMPC submitted this special report in accordance with TSs Table 3.3.7.10-1; i.e., the Vent GEMS was inoperable in excess of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. On September 22, the Vent GEMS system was declared inoperable due to the failure of the daily surveillance test for
detector signal resolution. Troubleshooting identified that the liquid nitrogen (LN:)
transfer collar was cracked. Subsequent investigation identified a ground in the system caused by the cracked collar. Replacement of the LN collar and repair of the system i
ground allowed the Vent GEMS to be returned to service on September 28. Inspector review of the special report determined that the report contained the necessary
-
information, identified the cause of the event, and detailed the corrective actions.
5.0 PLANT SUPPORT (60710, 71707, 90700, 90712)
-
5.1 Radiological and Chemistry Controls j
!
During routine tours of the accessible areas at both units, the inspectors observed the implementation of selected portions of NMPC's radiological controls program to ensure: the utilization and compliance with mdiological work permits (RWPs), detailed descriptions of radiological conditions, and personnel adherence to RWP requirements. The inspectors observed l
adequate controls for access to various radiologically controlled areas (RCA) and use of
)
personnel monitors and frisking methods upon exit from these areas. Posting and control of l
radiation areas, contaminated areas and hot spots, and labelling and control of containers holding
'
radioactive materials were verified to be in accordance with NMPC procedures. Radiation j
protection technician control and monitoring of these activities was satisfactory. Overall, the inspector observed an acceptable level of performance and implementation of the radiological controls program.
5.1.1 Vials of Uranium Found at Corporate IIcadquarters
On October 27, about 12:00 noon, the Unit I control room was notified of two broken 100 milliliter vials of " yellow powder" (potentially uranium) at the Niagara Mohawk corporate offices in Syracuse. The vials were in the auditorium, in a walk-in closet used for storage of files and displays, and were labeled " Niagara Mohawk Mining Company "
A radiation protection (RP) team from the Nine Mile site was sent to the corporate offices to contain and evaluate the spill. The spill was confined to the closet and the immediate area around the closet.
Direct frisk readings indicated 35,000 cpm # with detectable a. Airborne contamination samples were short half-life, indicating radon.
There were no personnel contaminations.
Initial d
..
-
,
-
_
.
- _ _ _ _ _ _ _ _
'
.
-
.
evaluations identified the powder as yellowcake (processed natural uranium), probably part of a public affairs "show and tell" to describe the Niagara Mohawk uranium mining operation. The spill was cleaned up the next day. The yellowcake, contaminated materials from the closet and surrounding area, and the waste generated during the clean up was moved from the corporate headquarters to the NMP site for later disposal with other radioactive waste.
The yellowcake is not covered under the Unit 1 or Unit 2 license, but under a license issued to Niagara Mohawk for the mining operation. Nonetheless, Unit 1 management responded as if the spill was within the bounds of the Nine Mile facility and directed cleanup activities accordingly. The inspectors monitored the response to the event and determined that the spill was contained in a timely manner, and that the licensee responded appropriately.
l
5.1.2 Radiation Exposure Controls in the Drywell Upper Elevations During the Outage GE SIL 354, revision 1, discussed the potential for high radiation levels in the upper drywell -
areas during specific fuel movement activities. To protect personnel in the drywell from potentially high radiation exposure during these activities, GE recommended that radiation monitors with audible alarms and other safety means be employed. NMPC addressed GE's recommendations at Unit 2 with the following measures. Installation of two portable area radiation monitors with local alarm and readout capability under the fue1 transfer shield bridge
,
(FTSB) to warn drywell personnel in the event of a dropped fuel bundle. The FTSB is the temporary shield installed between the reactor pressure vessel and the spent fuel pool for vessel component transfer. Should the radiation monitor alarm, all drywell personnel are trained to immediately exit the drywell and notify the control room and radiation protection personnel
controlling drywell access. Issue all drywell personnel self-alarming dosimeters to alert them i
to rapidly changing radiation fields. Restrict access to drywell elevations above 288 feet (bottom of the FTSB is at 330 feet) without special permission from the Station Shift Supervisor (SSS)
and radiation protection department (ladders leading to the area are locked and the boundary is identified by flashing yellow lights). In addition, brief each worker immediately prior to a
drywell entry on evacuation routes and potential radiation hazards in the upper drywell.
I GE SIL 354 specifically addressed fuel movements; however, Unit 2 also implemented similar controls to allow work in the upper elevations of the drywell while removing reactor control blades, local power range monitors, and other non-fuel irradiated components. A safety evaluation was performed per 10 CFR 50.59 and the activities were evaluated with respect to 10 CFR 20 considerations.
Based on a review of the Unit 2 response to the SIL and the safety evaluation for non-fuel component movement and direct observation at the drywell control point and in the drywell, the inspectors considered the controls for radiation exposure in the upper drywell elevations satisfactory.
- - - _ - - _ _ - _ - _ _ -
_ _ _ _ _ _ _ -
_
.
9-
.
~
5.2 Security and Safeguards Implementation of the physical security plan was observed in various plant areas with regard to the following: protected area and vital area barriers were well maintained and not compromised; isolation zones were clear; personnel and vehicles entering and packages being delivered to the protected area were properly searched and access control was in accordance with approved licensee procedures; persons granted access to the site were badged to indicate whether they have unescorted access or escort authorization; security access controls to vital areas were maintained and persons in vital areas were authorized; security posts were adequately staffed and equipped, security personnel were alert and knowledgeable regarding position requirements, and written procedures were available; and adequate illumination was maintained. Licensee personnel were observed to be properly implementing and following the physical security plan.
5.3 Emergency Preparedness The inspector accompanied the Director, Emergency Preparedness, and her staff on a tour of the NMPC emergency response facilities: including thejoint news center GNC), the emergency offsite facility (EOF), the technical support center (TSC), the operations support center (OSC),
the James A. Fitzpatrick EOF (a backup for Nine Mile), and the Oswego County emergency response center. All of the facilities were well maintained and appropriately supplied Radiation -
monitoring equipments was randomly verified to be in calibration. There were no additional questions.
6.0 PREVIOUSLY IDENTIFIED ITEMS (92701)
(Closed) Unresolved Item (50-220/92-12-01):
Review Response to NRC Bulletin 8848 NMPC identified portions of three systems (reactor head spray, feedwater, and emergency cooling system) potentially susceptible to the thermal cycling fatigue phenomena described in NRC Bulletin 88-08, " Thermal Stresses in Piping Connected to Reactor Coolant Systems."
NMPC instituted modifications and operational changes to resolve concerns for the reactor head spray line and the feedwater system. Specifically, the unisolable portion of the reactor head spray line was removed from service and changes were made to provide more stable control of feedwater flow under low flow conditions, thereby significantly reducing trie number of thermal cycles experienced. By letter dated July 20,1992, NMPC provided information which modified their response to NRC Bulletin 88-08. The submittal provided an evaluation of thermal fatigue cracks discovered in the bodies of emergency cooling condensate return isolation valves and provided a commitment to implement a temperature monitoring program for the unisolable portions of the system in accordance with the recommendations of NRC Bulletin 88-08. By letter dated November 13,1992 the NRC staffissued a safety evaluatbn which concluded that NMPC responses and commitments associated with NRC Bulletin 88-08 were consistent with the position of the Bulletin and therefore were acceptable. This item is close _.
.
-
- -.
..
-
l
.
.
,.
,
(Closed ) Unresolved Items (50-220/92-16-01 and 50-410/92-18-01):
i Falsification of Plant Records i
Combined inspection report 50-220/92-16 and 50-410/92-18 reviewed operator rounds at both
units as follow-up to NRC Information Notice (IN) 92-30 which discussed falsification of plant
'
,
records. NMPC thoroughly reviewed the issues discussed in the 1N and determined that several
records were incomplete or inaccurate. The NRC assessed your findings and issued a Notice i
'
of Violation (NOV) on October 15, 1993. However, after considering the facts of this case, including your identification of the violation and the corrective actions taken (both disciplinary
!
action against the offending individuals and program upgrades to improve the log keeping process), the NRC decided not to assign a severity level or require a response to the NOV.
,
Based on the above, unresolved items 50-220/92-16-01 and 50-410/92-18-01 and the NOV are closed.
i (Closed) Unresolved Item (50-220/92-24-01):
Less than Reauired SROs in the Control Room
.
On October 19,1992, the Unit I station shift supervisor (SSS), a licensed senior reactor operator l
(SRO), left the control room for approximately five minutes. At the time, this item was
!
identified as an apparent violation of TSs since the Assistant SSS, the other SRO on-shift, was i
also out of the control room.
f i
NMPC took prompt corrective actions including removal of the individual from licensed duties,
,
clarifying the duties and responsibilities of shift personnel, and reenforcing with the operations
crews of both units the expectations of management. The root cause for the event was the SSS's narrow focus on a specific on-going activity. This was an isolated occurrence.
l This unresolved item is closed. However, this was a violation of TSs 6.2.2.e, shift staffing l
requirements. But because the licensee identified the problem and took immediate and adequate
corrective action, the NRC is not citing this violation in accordance with 10 CFR 50, l
Appendix C, VII.B.2. This also closes LER 92-220/92-12 concerning the same subject.
i (Closed) Unresolved Item: (50-220/92-26-01):
Inadeouate Evaluation of Seismic Imoact for Loads Attached to Safety Related Structures
NRC review of design changes and modifications program implementation identified several
)
weaknesses in NMPC's program addressing the impact of seismic loads on safety related j
structures. Specifically, NMPC issued a design change to the field to fabricate and install a j
monorail and trolley to rig a containment isolation valve. The inspector determined NMPC's.
10 CFR 50.59 evaluation failed to consider the weight of the trolley rig and the valve when
calculating the seismic forces. NMPC reperformed the calculations to determine if the seismic
,
forces for the rigging unit were within the design capacity of the attached safety-related y
-
structural beam.
The inspector reviewed NMPC's corrective actions and root cause
.j determination.
NMPC concluded that engineering personnel lacked clear guidelines for
i
!
,
--
-
.. --
.
.-
-
-.
..
.--.
i l
.
l
-
i
'
- <.
'
!
I documenting seismic evaluations for attachments made to structures.
Engineering issued
engineering design standard -IS-EDS-001, "Ergineering Specification for the Design of
!
Temporary Supports and Rigging," which provides clear guidelines for evaluating seismic loads.
'
The inspector reviewed the calculation, updated 10 CFR 50.59 evaluation, and the engineering.
design standard and determined this issue was resolved. This item is closed.
l n
(Closed) Unresolved Item (50-410/92-27-02k Use of Procedures During Emergencies
!
This issue dealt with the use of recovery procedures during the event described in NUREG-
1455, " Transformer Failure and Common-Mode Loss ofInstrument Power at Nine Mile Point Unit 2 on August 13, 1991." Specifically, whether all applicable procedures were followed when the damage control team restored several nonsafety-related uninterruptible power supplies (UPSs) to normal operation, j
i NMPC reported the results of their review of this issue in a letter to Mr. Marvin W. Hodges
,
of the NRC, NMP2L1360, dated October 27,1992. They concluded the following: based on l
the situation at the time, the UPS restoration was not necessary on an emergency basis; restoration actions were well thought out, planned, appropriate, and correct; the actions taken
'
by the damage control team were consistent with the applicable procedure, however, additional precautions followed by the team were not documented in the procedure; prior to implementing l
the additional precautions the damage control team received verbal permission from a control
room senior reactor operator. Additionally, NMPC concluded that before proceeding with the
,
i additional precautions, the verbal approval of the procedure change should have been documented and reviewed in accordance with the " Damage Control Summary Form" of Emergency Plan Implementing Procedure, EPIP-EPP-22 (formerly S-EPP-22), " Damage
>
Control." A DER was initiated to address the issue. The DER disposition required a lessons -
,
learned transmittal and additional training for appropriate members of the emergency response
!
staff. The inspector agreed with NMPC's conclusion and based on their completed corrective
,
actions this item is closed.
!
l i
(Closed) Violation (50-410/92-29-01h
>
Failure to Use Proper Test Eauinment During Performance of Standby Liauid Control l
Surveillance Test
.
While observing performance of a surveillance test (ST) the NRC, inspector noticed that the
!
personnel were utilizing an improper test gauge (0-200 inch water vice a 0-100 inch water I
gauge). The test procedure required inservice test personnel to be notified and a procedure-change issued when a different gauge was utilized. NMPC replaced the gauge with a 0-100 inch
gauge and reperformed the test satisfactorily. The inspector reviewed the corrective actions and
,
determined NMPC took appropriate actions to prevent reoccurrence. This item is closed.
'
,
I w--r,-+
w a
-
r--<
w,
--
-
-*-
'ew w
~~
~
-n
.
..
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.
.
_
.
-
.
(Closed) Violation (50-220/93-01-03):
Failure to Follow Surveillance Procedure Unit 1 experienced a reactor scram caused by I&C technician failure to properly follow a surveillance test. NMPC continues to perform maintenance self-assessment activities to verify adherence to the maintenance performance principles and to determine the effectiveness of these corrective actions in reducing personnel errors. The inspector reviewed these self-assessment activities as discussed in section 3.1.2 of this report. This activity appears to be providing positive results within the maintenance staff. No personnel errors affecting plant equipment during maintenance or testing have been identified recently. This item is closed.
(Closed) Unresolved Item (50-220/93-18-01):
Fire in the Unit 1 Auxiliary Control Room A fire in the Unit 1 auxiliary control room resulted in the loss of some indicating lights and alarms, and the total loss of manual rod control. Temporary modifications provided power for the indicating lights and alarms until the repairs were completed. The NMPC actions to minirnize the potential for similar fires in the future was unresolved pending NRC review of the l
final root cause analysis and associated corrective actions.
NMPC identified the root cause of the fire as the failure of a normally energized GE CR-120 relay. At the request of NMPC, GE analyzed the relay and determined that it failed due to age and service conditions. In addition, GE specified two parameters that NMPC could use to evaluate the condition of the relays, while stillin service: coil temperature and insulation color.
NMPC inspected all of the normally energized CR-120 relays to identify any other relays needing replacement and to obtain baseline data as to temperature and color; no relays needed to be replaced. Based on the NRC review of the root cause analysis and the thetmographic j
inspections, this item is closed.
l u
(Closed) Unresolved Item (50-410/93-18-02):
Hvoass Leakage Path in Both Standby Gas Treatment System Charcoal Adsorbers I
Following the failure of a TSs required surveillance, NMPC determined that both standby gas treatment system charcoal adsorbers had inadequate reserve bed levels.
This condition eventually led to the erosion of the charcoal filter beds to the point of unsatisfactory bypass leakage. This issue was unresolved pending completion of NMPC's root cause evaluation, off-i site dose assessment, and determination of generic implications for other charcoal adsorbers used on site.
NMPC completed their root cause evaluation and concluded that inadequate written communication, i.e., the lack of vendor manual information for minimum fill heights, was the root cause of the event. Also, the failure to recognize low charcoal levels as a precursor to failure was a contributing factor. The inspector agreed with these findings, however, he additionally concluded that the proper questioning attitude by individuals involved with the initial
_
.-
-
,
i
.
..
~
fill and subseanent inspections could have precluded the event. The inspector also reviewed LER 50-41F-06, which addressed the event, and the off-site dose assessment calculations and discussed with site licensing the measures taken to identify any similar concems for other charcoal adsorbers on site. Based on the atave reviews and discussions, this item and the LER are closed.
7.0 MANAGEMENT MEETINGS At periodic intervals and at the conclusion of the inspection, meetings were held with senior station management to discuss the scope and findings of this inspection. Based on the NRC Region I review of this report and discussions held with Niagara Mohawk representatives, it was determined that this repon does not contain safeguards or proprietary information. NMPC did not object to any of the findings or observations presented at the exit meeting.
!
,