ML18152A195
| ML18152A195 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 08/26/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18152A196 | List: |
| References | |
| 50-280-96-07, 50-280-96-7, 50-281-96-07, 50-281-96-7, NUDOCS 9609100167 | |
| Download: ML18152A195 (26) | |
See also: IR 05000280/1996007
Text
U.S. NUCLEAR REGULATORY COMMISSION
REGION I I
Docket Nos:
License Nos:
Report Nos:
Licensee:
Facility:
Location:
Dates:
Inspectors:
Approved by:
9609100167 960826
ADOCK 05000280
G
50-280, 50-281
50-280/96-07, 50-281/96-07
Virginia Electric and Power Company (VEPCO)
Surry Power Station, Units 1 & 2
5570 Hog Island Road
Surry, Va 23883
June 16 - July 27, 1996
M. Branch, Senior Resident Inspector
D. Kern, Resident Inspector
W. Poertner, Resident Inspector
R. Chou, Reactor Inspector (Sections El.I and
El. 2)
L. Garner, Project Engineer (Sections M8.2,
M8.3, M8.4, M8.5, M8.6 and EB.I)
R. Gibbs, Reactor Inspector (Section Ml.3)
E. Testa, Senior Radiation Specialist (Sections
Rl. 1, R2. 1, R2. 2, R3 . 1 and RB. 1)
G. Belisle, Chief, Projects Branch 5
Division of Reactor Projects
Enclosure
EXECUTIVE SUMMARY
Surry Power Station, Units 1 & 2
NRC Inspection Report Nos. 50-280/96-07, 50-281/96-07
This integrated inspection included aspects of licensee operations,
engineering, maintenance, and plant support.
The report covers a six-week
period of resident inspection; in addition, it includes the results of
announced inspections by four regional inspectors.
Operations
Licensee actions with regard to Hurricane Bertha demonstrated a good
safety perspective and ensured that important systems were fully
operable and that the site area was prepared for severe winds
(Section 01.2).
A Unit 1 power reduction to less than 50 percent power was initiated on
July 17 due to failure of a moisture separator reheater stop valve to
reopen following testing.
The valve was repaired/retested and the unit
was returned to full power on July 18 (Section 01.3).
Maintenance
Failure to apply anti-corrosion coatings to the station and emergency
diesel generator battery intercell connections during refueling outage
maintenance had minor safety significance and was treated as a Non-Cited
Violation.
The licensee initially considered the condition acceptable
until the inspectors informed them that the Technical Specifications
(TSs) required the coating to be applied at a refueling outage frequency
(Section Ml.l).
Maintenance personnel properly evaluated and repaired a leak on valve
2-FW-10.
Procedure revisions were initiated to strengthen oversight
controls on the amount of injection material used.
The leak repair crew
was knowledgeable and the prejob briefing effectively addressed
communications, radiation protection escort coverage, stay time, and
radiation work permit requirements (Section Ml.2).
The preventive maintenance program for the low head safety injection
system was found to be comprehensive and was adequately implemented.
No
deficiencies in the program or implementation of the program were
observed (Section Ml.3).
Preventive maintenance activities performed on the Unit 1 C charging
pump were performed in accordance with the procedure requirements
(Section Ml.4).
Number 1 emergency diesel generator testing was satisfactorily
accomplished (Section Ml.5).
Performing preventive maintenance activities on the wrong containment
spray pump had low safety significance and was treated as a Non-Cited
Violation (Section Ml.6).
2
LER 50-280, 281/96006 description scope was incomplete.
It did not
address that TS 4.0.3 was entered for not completing TS surveillance
requirements on station battery 2B (Section MS.I)~
Engineering
The procedures and calculations generated to support heavy load
operations including spent fuel cask operations were adequate to provide
the details for conducting heavy load operation.
The engineers were
knowledgeable in preparing the procedures, calculations, analyses, and
dispositioning the deviation reports.
The crane operator training and
qualification were adequate (Sections El.I and El.2)
Design Change Package 95-019, Normal Switchgear Air Conditioning
Upgrade/Surry/Unit 1&2, was effectively implemented to establish
adequate environmental conditions for control rod drive cabinet
circuitry. Station management's decision to augment normal switchgear
room heating, ventilation and air conditioning with temporary spot
coolers through the summer was prudent (Section E2.l).
Emergency diesel generator fuel oil configuration and existing
operational practices assured that required fuel oil inventories were
being maintained (Section E2.2).
Plant Support
The program for shipping radioactive materials was effectively
implemented (Section Rl.1).
No concerns were identified during review of radiation monitor
calibration and alarm setpoints (Section R2.1).
Housekeeping and the control of contaminated and radioactive material
within the Auxiliary Building, radioactive waste warehouse, scrap
storage areas, and Fuel Handling Building were acceptable (Section
R2.2).
The inspectors concluded that the licensee had implemented and
maintained an effective program to monitor and control liquid and
gaseous radioactive effluents. The projected offsite doses resulting
from those effluents were well within the limits specified in the TSs,
Offsite Dose Calculation Manual, and 40 CFR 190 (Section R3.1).
The licensee had implemented water chemistry control programs in
accordance with the TS requirements.
The licensee was aggressively
sampling and monitoring the primary and secondary water chemistry
parameters (Section RS.I).
Report Details
Summary of Plant Status
Unit 1 operated at approximately 100 percent power until July 17 when power
was reduced to 49 percent to repair the 1 C reheat stop valve (Section 01.3).
The unit was returned to 100 percent power on July 18.
Unit 2 operated at approximately 100 percent power the entire reporting
period.
I. Operations
01
Conduct of Operations (40500, 71707)
01.1 General Comments
The inspectors conducted frequent control room tours to verify proper
staffing, operator attentiveness, and adherence to approved procedures.
The inspectors attended. daily plant status meetings to maintain
awareness of overall facility operations and reviewed operator logs to
verify operational safety and compliance with Technical Specifications
(TSs).
Instrumentation and safety system lineups were periodically
reviewed from control room indications to assess operability.
Frequent
plant tours were conducted to observe equipment status and housekeeping.
Deviation Reports (DRs) were reviewed to assure that potential safety
concerns were properly reported and resolved.
The inspectors found that
daily operations were generally conducted in accordance with regulatory
requirements and plant procedures.
01.2 Hurricane Bertha Preparations
a.
Inspection Scope
During the week of July 8, the inspectors monitored the licensee's
actions regarding preparations for Hurricane Bertha.
The inspectors
reviewed the Updated Final Safety Analysis Report (UFSAR) and the
Virginia Power Hurricane Response Plan, revision 2.
Additionally, on
July 12, the inspectors toured the outside and roof areas of the plant
looking for loose or unsecured items.
b.
Observations and Findings
During the week of July 8, Hurricane Bertha approached the U.S. coast.
The licensee monitored the storm track and implemented severe weather
preparations in accordance with Operations Checklist (DC) - 21, Severe
Weather.
The licensee's hurricane response is triggered by the
prediction of hurricane force winds on-site by the Virginia Power
Weather Center.
Hurricane force winds were not predicted during the
approach of Hurricane Bertha; however, the licensee did implement the
plan on July 12 as a precautionary measure when the storm was predicted
to pass through the area.
The inspectors reviewed the licensee's
c.
01.3
2
preparations, reviewed the status of important systems, and monitored
the storm progress.
The outside and roof areas inspected were clean and
the minor unsecured articles noted were immediately corrected by the
licensee. Maintenance and surveillance activities performed by the
licensee were minimized during the approach of the storm, the shift
operating crew was augmented prior to the arrival of the storm, and all
safety systems were fully operable.
Hurricane force winds were not
observed onsite and the Surry hurricane response plan was terminated at
5:00 a.m. on July 13 when the storm exited the area.
Conclusions
The licensee's actions with regard to Hurricane Bertha demonstrated a
good safety perspective and ensured that important systems were fully
operable and that the site area was prepared for severe winds.
Unit I Power Reduction
On July 17 at 3:12 p.m., during performance of procedure l-OSP-TM-001,
Turbine Inlet Valve Freedom Test, revision 7, the IC Moisture Separator
Reheater (MSR) stop valve did not reopen following close testing.
The
procedure required that power be reduced to less than 50 percent if a
reheat stop or intercept valve failed to reopen.
The required power
reduction reduces stresses on the MSR caused by increased flow rates and
moisture loading. A power reduction was commenced and power was
stabilized at 49 percent at 2:46 p.m., on July 17.
The valve was
repaired, retested and the unit was returned to 100 percent power on
July 18 at 8:20 p.m.
II. Maintenance
Ml
Conduct of Maintenance
Ml.I
Missed Emergency Diesel Generator and Station Battery Surveillance Tests
a.
Inspection Scope (92902)
On June 19 station management determined that anti-corrosion coating had
not been applied to the 18 and 28 station batteries and the No. 1, 2,
and 3 Emergency Diesel Generator (EOG) battery interconnectors within
the period specified in TS.
Operators entered a 24-hour Limiting
Condition of Operation (LCO) in accordance with TS 4.0.3 for failing to
perform the TS required surveillance. Priority I Work Orders (WOs)
were issued, anti-corrosion coating was applied to the battery
interconnections listed above, and the LCO was exited at 5:15 p.m.
The
inspectors reviewed this event to determine the cause and safety
significance of the missed surveillance, and to evaluate the licensee's
corrective actions.
3
b.
Observations and Findings
TS 4.6.C.l.f requires that an anti-corrosion coating be applied to
station battery cell interconnections during refueling outages.
Station
procedures require that the existing anti-corrosion coating be
inspected, but do not specify coating application unless bare metal is
exposed.
While processing a Preventive Maintenance (PM) deferral
request for the 28 station battery, maintenance personnel noted that
refueling outage battery inspection procedures did not require a coating
application.
DR S~96-1379 was initiated to determine whether current
maintenance practices satisfied TS requirements for the station
batteries.
Initial licensee DR review determined that the TSs were
satisfied.
The NRC reviewed the issue and informed station management that failure
to apply the anti-corrosion coating violated the Surry Station TSs.
In
addition, the inspectors informed the licensee that TS 4.6.D.l.e, which
requires anti-corrosion coating be applied to the EOG batteries each
refueling outage, was not satisfied. A TS surveillance requirement
validation project, performed in 1993-1994, had overlooked this same
discrepancy due to personnel error.
The inspectors noted that the TS
surveillance validation project was effectively implemented and
considered this an isolated discrepancy.
Station management directed
appropriate immediate corrective action for this event as discussed in
paragraph (a) above.
The station and EOG batteries remained able to perform their function
during this event.
Battery intercell connection resistance and anti-
corrosion coating condition were consistent with vendor's
recommendations and industry standards.
The inspectors reviewed
ANSI/IEEE STD 450-1987, IEEE Recommended Practice for Maintenance,
Testing, and Replacement of Large Lead Storage Batteries for Generating
Stations and Substations, and EXIDE 58.00, Instructions for Installing
and Operating Stationary Batteries, and concluded that the failure to
reapply the anti-corrosion coating did not adversely impact battery
operability.
The event was reported in Licensee Event Report (LER) 50-280, 281/96006
(Section MS.I) as required by 10 CFR 50.73(a)(2)(i)(B).
The licensee
intends to submit a TS amendment request to revise EOG battery and
station battery refueling cycle surveillance requirements to be
consistent with standard TSs.
Part of this change would remove the
requirement to reapply anti-corrosion coating when the existing coating
was satisfactory.
The revision would also implement a cell
tightness/resistance check similar to that described in standard TS.
The inspectors discussed the intended TS amendment request with
licensing and system engineers.
In addition, schedulers intend to add
anti-corrosion application battery PMs to the upcoming Unit I work scope
pending approval of the TS amendment request.
The inspectors concluded
that the planned corrective actions were appropriate .
':ii
4
c.
Conclusions
Ml.2
a.
b.
Failure to apply anti-corrosion coatings to the station and EOG battery
intercell connections during refueling outage maintenance is a violation
of TS 4.6.C.l.f and 4.6.0.1.e.
The discrepancy had low safety
significance and was initially identified by the licensee in DR S-96-
1379.
This failure constitutes a violation of minor significance and is
being treated as a Non-Cited Violation (NCV), consistent with Section IV
of the NRC Enforcement Policy. This item is identified as NCV 50-280,
281/96007-01.
Feedwater Injection Check Valve Seal Injection Leak Repair
Inspection Scope (62703)
On July 18, operators identified leakage from the Unit 2 feedwater check
valve 2-FW-10 during the monthly containment walkdown.
Feedwater check
valve 2-FW-10 is a containment isolation valve on the C feedwater line
inside containment.
The inspectors monitored repair activities to
determine whether appropriate reactor and personnel safety
considerations were addressed.
Observations and Findings
Mechanics promptly determined the leak was from the valve body to bonnet
at approximately 100 cubic centimeters per minute.
Engineering and
maintenance personnel developed a good repair plan in a timely manner.
The inspectors reviewed Engineering Transmittal (ET) S96-0217, FW Check
Valve Leak Sealing (2-FW-10) Surry Power Station-Unit 2, revision 0, and
associated WO 344933-01.
ET S96-0217 was technically sound.
The WO
provided generally good work instructions.
However, controls to limit
the amount of injection material used were very limited.
The inspectors
determined that a vulnerability for sealant over injection existed.
The inspectors observed that maintenance personnel primarily relied on
vendor experience to limit the maximum amount of sealant injected.
Although the void volume was calculated, neither implementing procedure
O-MCM-1918-01, On-Line Leak Repairs, revision 7, nor the WO specified
the maximum amount of sealant material to be injected.
The inspectors
discussed this observation with maintenance and engineering personnel
who initiated action to strengthen controls on the amount of injection
material to be used.
The inspectors attended the prejob briefing. Operations, maintenance,
and radiological aspects were briefed in excellent detail.
Previous
communications, radiation protection escort coverage, stay time, and
radiation work permit briefing weaknesses were completely addressed.
The inspectors discussed this repair evolution with the work crew and
determined that the crew was knowledgeable and knew the correct amount
of injection material to use for this repair.
The leak repair was
successfully completed on July 25.
5
c.
Conclusions
Maintenance personnel properly evaluated and repaired a leak on 2-FW-10.
Procedure revisions were initiated to strengthen oversight controls on
the amount of injection material used.
The leak repair crew was
knowledgeable.
The prejob briefing effectively addressed
communications, Radiation Protection (RP) escort coverage, stay time,
and Radiation Work Permit (RWP) requirements.
Ml.3
Preventive Maintenance (PM)
a.
Inspection Scope (62700)
This portion of the inspection was conducted to assess the licensee's
preventive maintenance program.
In order to accomplish this effort the
PMs conducted on the Unit 2, Train A, Low Head Safety Injection (LHSI)
system were selected for review.
The inspection included: verification
that all major equipment in the system was subjected to some kind of PM,
review of a sample of all of the latest PMs completed on the system for
technical adequacy (including comparison of the PMs to the applicable
vendor technical manuals for the equipment), review of the adequacy of
the technical justifications for all PM deferrals written on the system
since 1994, and review of the adequacy of the scheduling of future PMs
based on the specified frequency and the latest PM completion date.
In
addition, information in the applicable portions of section 6.2 of the
UFSAR were reviewed for inconsistency with other information reviewed
during the inspection.
The specific components reviewed during the
inspection were as follows:
Flow Transmitter 2-SI-FT-2945: WO 321113-01 (18 month calibration)
and vendor manual 38-R711-00001
Motor Operated Valve 2-SI-MOV-2860A: WOs 332985-01 & 04, 293332-
01, and 294124-01 (electrical and mechanical inspection); vendor
manuals 38-L553-01 (motor), 38-L553-01 (operator), and 38-C473-04
(valve)
Motor Operated Valve 2-SI-MOV-2862A: WOs 332863-01, 333510-01, and
333509-01 (mechanical inspection, electrical inspection, VOTES
testing, respectively), and vendor manual 38-L553-0l (motor), 38-
L553-02 (operator), 38-T459-02 (valve)
Motor Operated Valve 2-SI-MOV-2863A: Inspector verified scheduling
only
Motor Operated Valve 2-SI-MOV-2864A: Inspector verified scheduling
only
Motor Operated Valve 2-SI-MOV-2890C: Inspector verified scheduling
only
6
LHSI Pump 2-SI-P-lA: WO 303989-01 (electrical inspection), vendor
manual 38-W893-49 (motor), and 38-8940-02 (pump)
Pressure Indicator 2-SI-PI-2943: WO 335777-01 (18 month
calibration); no vendor manual available
Relief Valve 2-SI-RV-28458: WO 295137-01 (RV set point check) and
vendor manual 38-C515-00010
Check Valves 2-Sl-243, and 2-Sl-85: WOs 332163-01, 332157-01, and
341826-01 (acoustic testing, full flow testing, and seat leak
check); vendor technical manual not reviewed
Motor Control Center for Valve 2860A, 2-EP-BKR-2Hl-2NlB: WO 293797-01 (Clean, inspect and electrical test of MCC electrical
components); vendor technical manual not reviewed
The inspectors also reviewed 12 PM deferrals.
b.
Observations and Findings
The PM program on the LHSI system was found to be comprehensive and
provided adequate coverage for the important components in the system
that were reviewed by the inspectors.
No technical deficiencies were
noted during the review of the PM work packages, and no deviations from
vendor recommendations concerning periodic maintenance were observed.
PM deferrals were found to be based on sound engineering principles, and
the scheduling of future PMs was accurately reflected in the scheduling
computer data base.
c.
Conclusions
The licensee's PM program for the LHSI system was found to be
comprehensive and was adequately implemented.
No deficiencies in the
program or implementation of the program were observed.
No deficiencies
were noted during the UFSAR review.
Ml.4 Charging Pump 1 C Motor Service and Inspection (62703)
On July 23, the licensee performed a scheduled PM on the 1 C charging
pump.
The inspectors monitored activities in progress to verify proper
implementation of the maintenance program.
The maintenance activity was
accomplished in accordance with WO 00337622 and procedure O-ECM-1412-01,
Charging Pump Motor Maintenance, revision I.
The maintenance activity
consisted of a motor inspection and cleaning, oil flush, and coupled run
to monitor vibration, motor amps, and bearing temperatures.
The
inspectors monitored activities in progress, verified proper isolation,
and verified adequate post maintenance testing was implemented following
the maintenance activity.
The 1 C charging pump maintenance activity
was accomplished in accordance with the WO and maintenance procedure
requirements.
7
Ml.5
Emergency Diesel Generator (EOG) Testing (61726)
The inspectors observed portions of the performance of procedure l-OPT-
EG-001, Number 1 Emergency Diesel Generator Monthly Start Exercise Test,
revision 7, conducted on July 9.
The inspectors reviewed the procedure
and verified that the procedure acceptance criteria was satisfied.
No
discrepancies were noted during the portions of the procedure observed.
Ml.6
Work Performed on Wrong Component (62703)
a.
Inspection Scope
On July 16, a PM was commenced on the Unit 2 B containment spray pump
instead of the Unit 1 B containment spray pump as required by the WO.
The inspectors reviewed the circumstances surrounding the work activity
and the licensee root cause analysis.
b.
Observations and Findings
At 6:00 a.m., on July 16 containment spray pump l-CS-P-18 was tagged out
to perform a lubrication PM activity per WO 340622.
At 8:40 a.m. an
operator performing normal rounds observed a maintenance mechanic
working on containment spray pump 2-CS-P-18.
The pump coupling guard
had been removed and a grease gun was installed. The operator
recognized that work was being performed on the wrong component, stopped
further work activities by the mechanic, and then notified the control
room.
The pump was placed in pull-to-lock at 8:44 a.m., the coupling
guard was reinstalled and the pump was returned to service at 9:04 a.m.
The maintenance activity performed would not have prevented the pump
from operating if a start signal had been received.
The licensee
performed a category 2 root cause analysis.
The inspectors reviewed the
root cause analysis and agreed with the conclusions reached.
c.
Conclusions
Failure to implement the requirements of WO 340622 is a violation of
station administrative procedures.
The discrepancy had minor safety
significance and was identified by the licensee using the DR process.
This.licensee identified and corrected violation is being treated as a
NCV, consistent with Section VII.8.1 of the NRC Enforcement Policy.
This item is identified as NCV 50-280, 281/96007-02.
MB
Miscellaneous Maintenance Issues (92700, 92902)
MB.I
(Closed) LER 50-280, 281/96006:
Failure to Apply Anti-Corrosion Coating
to EOG and Station Batteries. This event is discussed in paragraph
Ml.I.
The inspectors noted that the event description scope was
incomplete in that TS 4.0.3 was entered for not completing TS
surveillance requirements on station battery 28 in addition to station
battery 18 and EOG batteries l, 2, and 3 listed in the LER.
The
licensee intends to submit an LER update to correct this discrepancy.
8
M8.2
(Closed) LER 50-280/94003: Hole in Recirculating Spray Heat Exchanger
Service Water Outlet Piping.
The hole resulted from corrosion at a
point on the outlet piping elbow where the original coal tar epoxy
coating failed.
Immediate corrective actions included replacing the
elbow and verifying that minimum wall thickness existed on similar
piping on the other trains.
Long term corrective actions included:
cleaning by sandblasting, inspections, weld repairs as needed, and
recoating with a new type epoxy.
The inspectors reviewed the work
documentation for the activities completed during the recent Unit 2
outage.
No concerns were identified.
The inspectors verified that the
pipe sections discussed in the LER and which remain to be recoated were
planned to be completed during each unit's next scheduled refueling
outage.
M8.3
(Closed) LER 50-280/94002: Three of Fifteen Main Steam Safety Relief
Valves (SRVs) out of Tolerance due to Minor Setpoint Drift.
(Closed) LER 50-281/95002: Two Main Steam and Two Pressurizer (PZR) SRVs
as Found Lift Setting Out of Tolerance.
The two PZR SRVs as found lift
settings were within+/- 3 percent tolerance as specified in TS
Amendment 207, dated December 28, 1995 and thus would no longer be
required to be reported.
However, PZR SRVs with as found setting
outside the+/- 3 percent tolerance band are discussed in Section M8.5.
Records document that the main steam SRVs, which were found out of
tolerance, were repaired and tested three times to verify that they
would lift within+/- 1 percent of thecTS specified value.
SRV as found and as left test data for the period 1988 to 1996 was
reviewed.
During this interval, testing was either performed in place
using Trevitest equipment or the main steam SRVs were sent to Wyle
Laboratory for testing. Testing using Trevitest equipment was
discontinued in 1993.
On Unit 1 there were 9 as found failures out of
55 tests conducted and on Unit 2 there were 3 as found failures out of
74 tests conducted.
The same main steam SRV never experienced two
consecutive as found test failures.
The as found main steam SRV
failures appeared to be random except that all as found failures
occurred after a change in testing equipment, e.g., Trevitest testing
performed after Wyle Laboratory testing. Whether this was coincidence
or indicative of a minor problem with one or both of the test methods
could not be determined.
Due to wear, whenever main steam SRVs were tested at Wyle Laboratory,
they are refurbished.
The inspectors concluded that present testing
practices and the refurbishment program was adequate.
M8.4
(Closed) Violation 50-280, 281/94028-01:
Failure to Stop Work When Work
Instructions Could not be Followed.
The response to the Notice of
Violation (NOV), dated December 27, 1994, identified that additional
guidelines would be provided to ensure adequate implementation
instructions are provided prior to implementing Design Change Packages
(DCPs) and guidelines would include management's expectations when work
instructions cannot be followed.
The inspectors verified that
M8.5
M8.6
9
guidelines were appropriately incorporated into procedures and training
was provided to maintenance and outage and planning personnel as
committed in their NOV response.
(Open) LER 50-281/96003: One PZR SRV as Found Lift Setting Out of
Tolerance.
(Closed) LER 50-280/95008:
One PZR SRV as Found Lift Setting Out of
Tolerance.
This item is being closed for administrative purposes since
it is similar to LER 50-281/96003 which remains open to track this
issue.
The inspector reviewed the documentation associated with the PZR SRVs
repair and subsequent satisfactory tests.
The inspectors had no
questions concerning the repairs performed.
However, an abnormality in
the test data was noted.
For six valves tested, the difference between
the second as found lift setting and the first as found lift setting was
approximately 2 percent greater (three valves), approximately 4 percent
greater (one valve) and less than 1 percent lower (two valves).
The
second as found test was performed immediately after the first test
without any adjustments to the SRVs.
Having four out of six SRVs open
at approximately 2 percent or higher on their second opening seems
unusual.
This observation was discussed with the maintenance engineer
who indicated that the subject had already been discussed with the
vendor.
The cause for this behavior is not known .
There is little industry operational experience with these new style
valves.
The SRVs were procured from Crosby and shop tested by Crosby in
1993.
The Unit 1 SRVs were removed and tested in September 1995.
The
Unit 2 SRVs were removed and tested in February 1995 and May 1996.
Pending review of additional test data associated with these new style
SRVs, this LER will remain open.
(Open) Violation 50-280, 281/94017-02:
Failure to Implement Corrective
Actions to Preclude Repetition of Foreign Material Exclusion (FME)
Deficiencies.
The inspectors verified that the corrective actions
described in the NOV response dated August 17, 1994, were implemented.
Review of audits and quarterly DR trend reports revealed that problems
in the FME area continue to exist.
The quality assurance Foreign Material Exclusion Follow-Up Assessment,
dated August 30, 1995, concluded that FME program requirements were not
consistently applied for work activities for open systems and
contributors to industry events exist at Surry.
The inspectors verified
.that the issues were addressed by the Maintenance Department.
The Third Quarter 1995 Station Deviation Trend Report recommended that a
FME task team be formed to review activities and past assessments and
report the results to management.
Although no task team was formed,
actions were taken to improve FME controls .
10
Nuclear Oversight finding 96-06-02S, Operations/Refueling Activities,
was issued concerning FME problems encountered during Unit 2 refueling
activities in June 1996.
The inspectors observed that the number of DRs associated with FME
problems decreased from six per quarter in mid-1995 to only one or two
during recent quarters. There appears to be no one single cause for the
continuing FME deficiencies.
Based upon documentation reviews and
interviews with station personnel, FME problems areas include personnel
that infrequently implement FME controls and a sensitivity to adversely
affecting surrounding equipment while performing work.
An example from
finding 96-06-02S was greasing the Fuel Building Crane drive wheel
assembly while it was positioned over the Spent Fuel Pool.
Although recognizing the improving trend in this area, problems do
continue. This violation will remain open pending additional inspection
activities to monitor progress in this area.
III. Engineering
El
Conduct of Engineering
El.I Heavy Load Program
a.
Inspection Scope (37700)
The inspectors examined the licensee's responses to the Generic Letters
(unnumbered) dated December 22, 1980, Control of Heavy Loads and 81-07,
Control of Heavy Loads; reviewed the adequacy of the heavy load program
to assure that it was prepared in accordance with regulatory
requirements, appropriate industrial codes and standards; and verified
through record review that the crane operators received proper training
and were qualified to perform the heavy load lifting operations.
The inspectors reviewed a Nuclear Standard for heavy loads, the
licensee's Phase I and Phase 2 responses to NUREG 0612, Control of Heavy
Loads at Nuclear Power Plants, and eight procedures which implement the
licensee's program.
The inspectors reviewed the codes and standards
that were listed as references in the licensee's heavy load program.
The inspectors examined Westinghouse Technical Report WCAP-9198, Reactor
Vessel Head Drop Analysis, revision 0, and three calculations that were
completed to support the responses sent to the NRC for Generic Letter 81-07.
The inspectors' review included the assumptions, theory, and
codes used.
The inspectors also reviewed the records for the training,
qualification, and requalification for 12 crane operators in order to
verify compliance with ANSI codes and the licensee procedure
b.
11
requirements.
The inspectors also reviewed licensee DRs in order to
determine the adequacy of resolution and corrective actions.
Observations and Findings
Inspection of the licensee's responses to the NRC and application of the
requirements and standards provided by the documents listed above
including the associated licensee's procedures indicated that the
licensee was committed to NUREG 0612.
The inspectors verified through
review of programs, procedures, calculations, and records that the
requirements of Phase I based on NUREG 0612 were met.
The seven
requirements are listed below:
Define the safe load paths
Develop load handling procedures
Establish periodic inspection and testing of cranes
Establish training and qualification for crane operators
Special lifting devices should satisfy the guidelines of ANSI
N14.6.
Standard lifting devices should be installed and used in
accordance with the guidelines of ANSI 830.9
Design cranes to ANSI 830.2 or CMAA-70.
Procedures GMP-001 and O-MCM-1150-01 were specified to be used to meet
NUREG 0612 for handling heavy load movements over the reactor core and
the spent fuel pool.
The safe load paths were contained in these
procedures.
Procedures GMP-010 and GMP-C-107 were used in the area of
balance of plant. There are approximately 10 additional existing
procedures that are used for crane maintenance and inspection.
The licensee classified the crane operators into four categories as:
1) mobile crane operators; 2) floor operated overhead crane operators;
3) cab operated overhead crane operators; and 4) cab operated polar
crane operators. During the records review for 12 crane operators, the
inspectors questioned training on several operators for overhead and
mobile crane operations.
The licensee presented the course materials
and the inspectors verified that they covered the training for both
overhead and mobile crane operations, even though the certificates only
stated one type of crane operations.
The licensee issued 21 DRs for heavy load movements during the past two
years.
The inspector randomly selected and reviewed five DRs: S-94-
1558, S-94-1928, S-94-2089, S-95-0379, and S-95-0577.
The inspectors
concluded that the resolution and corrective actions were adequate .
12
In the Phase II response to Generic Letter 81-07, the licensee stated
that the load drop analyses would be performed and submitted to the NRC
for review.
The NRC issued Generic Letter 85-11 and canceled the
requirements contained in the Phase II response.
Therefore, the load
drop analyses were not performed by the licensee except for the reactor
vessel head drop.
c.
Conclusions
The inspectors concluded that the program, procedures, and calculations
for heavy load movements were adequate.
The engineers were
knowledgeable in preparing the procedures, calculations, analyses, and
dispositioning the deviation reports.
El.2 Review of Spent Fuel Pool Cask Programs
a.
Inspection Scope (60855)
The inspectors reviewed five licensee procedures and calculations to
determine Spent Fuel Pool Cask Program compliance with the licensee's
commitments, regulatory requirements, and appropriate industry codes and
standards.
b.
Observations and Findings
The licensee generated calculations to support safe operations around
the spent fuel pool area during cask operations in case of an accidental
cask drop.
The licensee did not have any cask drop analysis for the
transfer from the Crane Enclosure (Fuel Handling Building) to the cask
storage area (ISFSI) even though the cask-carrying truck passes over
diesel generator fuel supply pipes during the transfer.
The di~sel fuel
pipes consist of six pipes which are part of a safety related piping
system.
The licensee's engineers gave the inspectors the following
reasons for not generating a load drop analysis:
The cask is being carried not greater than six inches above the
ground.
If the cask drops, the impact load due to the six-inch
drop will be very small.
The chance to hit all six diesel fuel pipes is very small unless
the cask drops directly into the center of the pipe group.
Every
two pipes supplies a diesel generator and each reactor only
requires one diesel generator to supply its electricity.
A station blackout due to the loss of outside (external) power
will require one diesel generator for safe shutdown or removal of
decay heat if both of the reactors are being shut down or both
units are already in shutdown conditions.
The inspectors accepted the licensee's explanation and considered that
the cask drop analysis is not required if the six diesel fuel lines are
the only safety-rela~ed piping or equipment in the outside cask transfer
13
route.
However, the licensee did perform a soil pressure impact check
- for the cask truck running over the six diesel fuel pipes.
The inspectors informed the licensee's engineers about the hydrogen
explosion on May 28, 1996, at the Point Beach Nuclear Plant, during a
welding operation to seal a cask cover lid.
The licensee's engineers
said that an explosion is unlikely at Surry, in that, the casks they use
are tightened with bolts and there is no welding involved.
c.
Conclusions
E2
E2.l
a .
b.
The inspectors considered that the procedures and calculations were
adequate.
The fact that no cask drop analysis for the transfer of the
cask from the crane enclosure area to the storage facility was
determined to be acceptable. A hydrogen explosion, similar to the May
1996 Point Beach event, is highly unlikelj to occur during the cask
drying process because no welding is involved.
Bolts are used to
tighten the cask cover lid.
Engineering Support of Facilities and Equipment
Normal Switchgear Room Ventilation Upgrade
Inspection Scope (92903)
In May 1995 Surry Unit 2 experienced two reactor trips due to Control
Rod Drive (CRD) circuit card failures.
The licensee determined that the
CRD circuit cards had become degraded from prolonged exposure to
elevated temperatures.
Short term corrective actions were previously
documented in NRC Inspection Reports Nos. 50-280, 281/95-08 and 95-17.
In May 1996 the Unit 1 and Unit 2 normal switchgear room ventilation
systems were upgraded as a long term corrective action.
DCP 95-019,
Normal Switchgear Air Conditioning Upgrade/Surry/Unit 1&2, was developed
to increase the Heating Ventilation and Air Conditioning (HVAC) cooling
capacity sufficiently to assure the normal switchgear rooms can be
maintained below 83°F.
Prior to this DCP, the HVAC system was sized to
maintain the rooms below 104°F.
The inspectors reviewed the ventilation
modification to determine whether corrective actions had been properly
implemented to preclude event recurrence.
Observations and Findings
DCP 95-019 installed larger capacity evaporator coils which nearly
doubled the normal switchgear room cooling capacity, removed ventilation
damper motor operators which had been unreliable, and reduced room heat
load by deenergizing switchgear space heaters.
The inspectors reviewed
DCP 95-019, associated WOs, and toured the normal switchgear rooms
during periods of elevated outside air temperature.
Room temperature
remained below 83°F when outside air temperature approached 100°F.
The
inspectors determined that DCP 95-019, through revision 13, was properly
implemented .
C.
14
Spot coolers, installed as a Temporary Modification in June 1995,
remained in service providing supplemental cooling to the CRD cabinets
following installation of DCP 95-19.
The inspectors questioned
management's intention regarding whether to make the spot coolers a
permanent design change or to close out the TM.
Station management
decided to continue augmenting normal switchgear room HVAC with the
temporary spot coolers through the summer.
Management intends to
reevaluate the TM for closure after the normal switchgear room HVAC
upgrade has demonstrated reliability through the hot summer period.
Conclusions
The inspectors concluded that the HVAC upgrade was an effective long
term corrective action to establish adequate environmental conditions
for CRD cabinet circuitry. Station management's decision to augment
normal switchgear room HVAC with temporary spot coolers through the
summer was prudent.
E2.2
Emergency Diesel Generator (EOG) Fuel Oil Tank Capacity
a.
Inspection Scope (37551)
TSs require each EOG fuel oil day tank to contain at least 290 gallons
and a total on-site fuel supply of at least 35,000 gallons.
The total
on-site fuel oil supply is required to provide sufficient fuel for one
EOG to run fully loaded for a week.
The UFSAR further specifies that
each EOG fuel oil day tank will have the capacity to support
approximately three hours full load operation.
The day tank is further
defined in the UFSAR to include both the EOG base tank and the attached
auxiliary wall tank located within the EOG room.
The inspectors
reviewed the UFSAR, TSs, engineering calculations, operator logs, and
local fuel oil tank indication to verify that a sufficient EOG fuel oil
supply was maintained.
b.
Observations and Findings
Calculation 01039.3410-M-2, EOG Underground Fuel Oil Storage Tank 7 Day
Usable Fuel Oil Supply, revision 0, states that 33,264 gallons are
required to support full load operation for 7 days.
Calculation
01039.3410-M-3, EOG Base and Day Tank 3 Hour Usable Fuel Oil Supply,
revision 0, states that 643 gallons are required to support full load
operation for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
Both calculations used the correct EOG fuel oil
consumption rate (198 gph) as independently derived by the inspectors
from Surry specification NUS-74 and correctly accounted for unusable
volumes within the tanks. Operator logs assign minimum acceptable fuel
oil tank levels which are above the TS required values.
In addition,
operator logs require a minimum EOG day tank level which is
approximately 100 gallons above the capacity needed for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of
15
operation.
The inspectors visually verified that the EDG day tanks were
filled to required levels.
The inspectors observed that operator logs
used different terms to identify the EDG day tanks than contained on the
tank labels, the UFSAR, and the two referenced calculations.
The system
engineer initiated appropriate action to revise the operator logs and
correct this discrepancy.
c.
Conclusions
The inspectors concluded that the EDG fuel oil configuration and
existing operational practices assured that required EDG fuel oil
inventories were being maintained.
EB
Miscellaneous Engineering Issues (92700)
E8.l
(Closed) LER 50-280/95005:
Error in Calculation to Convert TS NaOH
Volume to Level for the Chemical Addition Tank (CAT).
This error was
also addressed as NCV 50-280, 281/95009-01.
The NCV discussed the
acceptability of the immediate and proposed long term corrective
actions.
In this inspection period, the inspectors verified that
applicable engineering and operating procedures were revised and
implemented.
Specifically, a family of curves was provided in the Curve
Book to correlate percent NaOH concentration at various temperatures to
the indicated level needed to ensure that the TS required minimum volume
was met.
While reviewing 01/02-PT-36, Log Readings, revision 20, the inspectors
noted that the minimum allowed indicated level in the CAT was 98
percent, the value corresponding to the TS maximum allowed NaOH
concentration at 40°F.
However, no notation was provided to inform
Operations that with CAT temperatures less than 40°F, a 98 percent
indicated level would not ensure that the TS required minimum volume
would be available, i.e., at low temperatures further evaluation would
be necessary to verify that TS minimum volume requirements were met.
The Operations Manager indicated that the data logging procedure would
be revised to warn operators of this potential problem.
A review of Unit 1 and 2 CAT temperatures and indicated levels during
the period from December 1995 to March 1996 revealed that the minimum
temperature recorded was 42°F (Unit 1 CAT) and indicated level always
exceeded 98 percent. Thus, the TS required minimum CAT volume level
restriction was met during this cold weather period.
In addition, the
inspectors verified that the Unit 2 hand held computerized data logger
acceptance criteria was consistent with 02-PT-36.
RI
Rl.1
a.
16
IV. Plant Support
Radiological Protection and Chemistry (RP&C) Controls
Transportation of Radioactive Material
Inspection Scope (86750, TI 2515/133)
The inspectors evaluated the licensee's transportation and radioactive
materials programs for implementation of revised Department of
Transportation (DOT) and NRC transportation regulations for shipment of
radioactive materials as required by Title 10 Code of Federal
Regulations (CFR) Part 71.5 and 49 CFR Parts 170 through 189.
b.
Observations and Findings
C.
The inspectors selectively reviewed site transportation procedures and
determined that they adequately addressed the loading, shoring and
bracing of radioactive waste shipments to waste processors; placarding
of radioactive material loads; marking, labeling and placarding for
radioactive waste shipments to disposal facilities; radioactive material
shipment documentation; and radioactive waste surveys for shipment to
disposal facilities.
The inspectors witnessed the radioactive waste
shipment no.: 096-16 made on July 25, 1996 .
The inspectors reviewed the licensee's records for the four most recent
shipments that were made in 1996.
The inspector determined the shipping
papers contained the required information.
In addition, the inspectors
determined that the licensee had maintained adequate records of
shipments of licensed material for a period of three years after
shipment as required by 10 CFR 71.9l(a).
Conclusions
The licensee had effectively implemented a program for shipping
radioactive materials.
R2 Status of Radiation Protection Facilities and Equipment
R2.1
Radiation Monitor Calibrations
a.
Inspection Scope (84750, 82701)
b.
The inspectors reviewed selected radiation monitors for calibration and
alarm set points.
Observations and Findings
The inspectors reviewed selected alarm setpoints and calibration data
and determined that radiation monitors were within their calibration
interval and alarm setpoints were correctly set .
17
c.
Conclusions
No concerns with licensee facilities or equipment or analysis were
identified during the inspection.
R2.2 Tours of Licensee Radiological Control Areas CRCAs) (71750. 83750)
During tours of the licensee's facilities; the inspectors selectively
verified that radiological postings and labels were appropriate for the
radiological hazard.
The inspectors also observed that the housekeeping and the control of
contaminated and radioactive material within the licensee's Auxiliary
Building, radioactive waste warehouse, scrap storage areas, and Fuel
Handling Building were acceptable.
R3
RP&C Procedures and Documentation
R3.l
Control of Radioactive Effluents
a.
Inspection Scope (84750)
TS 6.4 for both units required the licensee to establish, implement, and
maintain a program for the control of radioactive effluents
The
program was required to include limitations on the annual and quarterly
radiation doses from radioactive materials in liquid and gaseous
effluents released to unrestricted areas.
TSs 6.6 and 6.8 for both
units described the reporting schedule and content requirements for the
Annual Radioactive Effluent Release Reports.
The reports were required
to be submitted prior to May 1 of each year and to cover the operation
of the facility during the previous calendar year. Summaries of the
quantities of radioactive materials in liquid and gaseous effluents
released from the facility and an assessment of the radiation doses due
to those releases were required to be included in the reports and an
assessment be made to ensure that doses are below
40 CFR 190 limits.
b.
Observations and Findings
The inspectors reviewed the effluent data compiled from the licensee's
effluent release report for the year 1995.
The values used for 1995
annual dose estimates were taken from the licensee's 1995 Annual
Radiological Environmental Operating Report.
The inspectors also
reviewed the supporting data for the effluent release report covering
the year 1995 and discussed the data presented in reports with the
licensee.
The inspectors determined that one effluent monitor instrument was
inoperable for more than 30 days.
The monitor was 02-SW-RM-220, Unit 2
Discharge Tunnel Radiation Monitor.
Coaxial and high voltage cables
were replaced.
During this period once-per-twelve-hour grab samples
18
were obtained in accordance with the Offsite Dose Calculation Manual
(ODCM).
There were no unplanned liquid or gaseous effluent releases classified
according to the criteria in the ODCM.
There were no major changes to
radioactive liquid, gaseous, and solid waste treatment system during
this reporting period.
No gas storage tanks exceeded the limits allowed by TS 3.7 during the
reporting period. Three minor changes were made to the ODCM during 1995
and were implemented in revision 7 dated October 31, 1995 (Procedure
VPAP-2103).
There were no continuous liquid effluent releases above the
lower limit of detection for either Surry Unit 1 or 2 during this
period.
c.
Conclusions
The inspectors concluded that the licensee had implemented and
maintained an effective program to monitor and control liquid and
gaseous radioactive effluents.
The projected offsite doses resulting
from those effluents were well within the limits specified in the TSs,
ODCM, and 40 CFR 190.
RS Miscellaneous Radiation Protection and Chemistry Issues
RB.I
Primary and Secondary Water Chemistry Review
a.
Inspection Scope (86750, TI 2515/133)
The inspectors reviewed and discussed the results of the licensee's
primary and secondary chemistry program.
b.
Observations and Findings
The inspectors reviewed the chemistry results for the TS data associated
with the primary and secondary water chemistry parameters for the period
from January l, 1996 through July 1, 1996, and determined that all
required TS chemistry results were maintained at small percentages of
limits.
c.
Conclusions
The licensee was aggressively sampling and monitoring the primary and
secondary water chemistry parameters.
F8
Miscellaneous Fire Protection Issues (92904)
F8.l
(Closed) Violation 50-281/96003-06:
Failure to Provide a Continuous
Firewatch.
This violation resulted from the failure of the diesel
generator rear exit door to close following exit from the room by plant
personnel due to welding cables blocking the door open.
Corrective
. *
19
actions for this violation included posting the rear exit door as an
emergency exit only and activating alarms on the doors.
The inspectors
verified that the corrective actions had been completed.
20
V. Management Meetings
XI
Exit Meeting Summary
The inspectors presented the inspection results to members of licensee
management at the conclusion of the inspection on August 1, 1996.
The
licensee acknowledged the findings presented.
The inspectors asked the licensee whether any materials examined during the
inspection should be considered proprietary.
No proprietary information was
identified .
.
.
21
PARTIAL LIST OF PERSONS CONTACTED
Licensee
R. Blount, Maintenance Superintendent
D. Christian, Station Manager
J. McCarthy, Assistant Station Manager, Operations & Maintenance
B. Stanley, Director Nuclear Oversight
J. Swientoniewski, Supervisor Station Nuclear Safety
W. Thorton, Superintendent, Radiological Protection
' *
IP 37551:
IP 37700:
IP 40500:
IP 60855:
IP 61726:
IP 62700:
IP 62703:
IP 71707:
IP 71750:
IP 82701:
IP 83750:
IP 84750:
22
INSPECTION PROCEDURES USED
Onsite Engineering
Design Changes, and Modifications
Effectiveness of Licensee Controls in Identifying, Resolving, and
Preventing Problems
Operation of an ISFSI
Surveillance Observation
Maintenance Implementation
Maintenance Observation
Plant Operations
Plant Support Activities
Operational Status of the Emergency Preparedness Program
Occupational Exposure
Radioactive Waste Treatment, And Effluent And Environmental
Monitoring
IP 86750:
Solid Radioactive Waste Management And Transportation Of
Radioactive Materials
IP 92700:
Onsite Followup of Written Reports of Nonroutine Events at Power
Reactor Facilities
IP 92902:
Followup - Maintenance
IP 92903:
Followup - Engineering
IP 92904:
Followup - Plant Support
TI 2515/133:lmplementation of Revised 49 CFR Parts 100-179 AND 10 CFR Part 71
Opened
50-280, 281/96007-01
50-280, 281/96007-02
Closed
50-280, 281/96006
50-280/94003
50-280/94002
50-281/95002
ITEMS OPENED, CLOSED, AND DISCUSSED
Failure to Apply Anti-Corrosion Coating to EOG
and Station Batteries (Section Ml.l).
Performing PM on the Wrong Component
(Section Ml.6).
LER
Station and EOG Battery Connections Not Coated
With Anti-corrosion Material Due to Procedural
Error (Section M8.l).
LER
Hole in Recirculating Spray Heat Exchanger
Service Water Outlet Piping (Section M8.2).
LER
Three of Fifteen Main Steam Safety Relief Valves
Out of Tolerance due to Minor Setpoint Drift
(Section M8.3).
LER
Two Main Steam and Two Pressurizer SRVs as Found
Lift Setting Out of Tolerance (Section M8.3).
l
"! .
50-280, 281/94028-01
50-280/95008
50-280/95005
50-281/96003-06
Discussed
50-281/96003
50-280, 281/94017-02
23
Failure to Stop Work When Work Instructions
Could not be Followed (Section M8.4).
LER
One PZR SRV as Found Lift Setting Out of
Tolerance (Section M8.5).
LER
Error in Calculation to Convert TS NaOH Volume
to Level for the Chemical Addition Tank
(Section E8.l).
Failure to Provide a Continuous Firewatch
(Section F8.l).
LER
One PZR SRV as Found Lift Setting Out of
Tolerance Section M8.5).
Failure to Implement Corrective Actions to
Preclude Repetition of Foreign Material
Exclusion Deficiencies (Section M8.6) .