ML20202E709
| ML20202E709 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee  |
| Issue date: | 11/28/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20202E698 | List: |
| References | |
| 50-271-97-08, 50-271-97-8, NUDOCS 9712080126 | |
| Download: ML20202E709 (38) | |
See also: IR 05000271/1997008
Text
. - - _ .- .. .. .
.:
..-
)
- U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No. 50 271
Licensee No. - DPR 28
Report No. 97 08
- Licensee: Vermont Yankee Nuclear Power Corporation
Facility: Vermont Yankee Nuclear Power Station.
Location: Vernon, Vermont
,
Dates: September 7 October 31,1997 .
Inspectors: William A. Cook, Senior Resident inspector -
Edward C. Knutson, Resident inspector
Jason C. Jang, Engineer, Division of Reactor Safety (DRS)
Robert J. Summers, Division of Reactor Projects
William A. Maler, Engineer, DRS .
Kenneth S. Kolaczyk, Engineer, DRS
Timothy L. Hoeg, Engineer, DRS
Mark Holbrook, Contractor, INEL
Approved by: Curtis J. Cowgill, Ill, Chief, Projects Branch 5
Division of Reactor Projects
.
<-
i
9712080126 971128
PDR ADOCK 05000271
-G PDR ,
_
.._
,
j '
l
l
l
l
l
EXECUTIVE SUMMARY
Vermont Yankee Nuclear Power Station
NRC Inspection Report 50-271/97-08
This integrated inspection included aspects of licensee operations, engineering,
maintenance, and plant support. The report covers a seven week period of resident
inspection and includes results of announced inspections by regional specialist inspectors.
Operations
The inspector catermined that the VY staff had responded appropriately to the September
27 seismic monitor alarm and that the declaration of an Unusual Event had been in
accordance with their emergency procedurx. However, the inspector considered that the
timeliness of licensee action to address this previously observed overly conservative
Emergency Plan entry condition, (single, unconfirmed seismic monitor alarm), to have been
slow,
inspector review of the October 10 ENS call, involving the Alternate Cooling System (ACS)
cable separation issue, identified an appropriate immediate response to the ACS operability
concern and appropriate follow-up corrective actions.
Maintenance
Based upon observation of a variety of maintenance and surveillance testing items,
appropriate control and execution of these activities was noted,
The licensee identified and corrected reactor building ventilation radiation monitor testing
discrepancy (LER 96-23) was not cited. The procedural non-compliance which contributed
to the fuel oil sampling and analysis events discussed in LER 96-29 was not cited. The
low pressure coolant injection surveillance testing discrepancy discussed in LER 96-27 was
not cited.
Enaineerina
At the end of the inspection period, the VY staff had completed formal emergency diesel
generator (EDG) support piping stress analyses and had completed a metallurgical analysis
which VY believes supports their initial EDG operability determination. These items were
under review by the NRC staff. Pending the results of further NRC staff review, the EDG
support piping welds issue is being tracked as an inspector follow-up item (IFI 97-08-01).
The licensee's response to this issue, to date, has been consistent with the guidance of
VY established a program that met their commitments to GL 89-10, " Safety-Related Motor-
Operated Valve Testing and Surveillance." Final val;dation of switch settings is currently
scheduled to be completed by January 30,1998. Use of the Electric Power Research
Institute (EPRI) motor operated valve performance prediction program to validate switch
ii
.. _. . . .. _ ._ .- . _. _. _
fi .
L,
,- .
.
.
-. settings for essentially all MOV's was exceptional and considered to be a program
strength.
. The failure to have included and tested a number of keep fill system check valves in the
- VY Inservice Testing Program (reference LER 9611)was not cited.
Plant Supoort
.,.
'
The radioractive liquid and gaseous effluent control programs were wellimplemented.
' '
- The licensee implemented good management control and oversight of the quhlity of
4 .the radioactive liquid and gaseous effluent control programs,
- * The effluent radiation monitoring s/ stem calibration program, including trending
analysis, was well-implemented.
'
s The ventilation system surveillance program was well implemented. However, the
l plant air balance measured in 1971 might be invalid, as described in Section R.2.3
of this inspection report. (IFl 97-08-02)
.
o Very good quality control for the chemistry laboratory and quality assurance audit -
programs were established.
The failure to have appropriately controlled the movement of reactor vessel shield blocks
preceding the 1990 and 1992 refueling outages (reference LER 96-03 and URI 96-03-05)
. was not cited. ,
4
i .
T'
'
iii
. .-. - , - , .- .
-
,- . - . - -- - . . - .
.
j
TABLE OF CONTENTS -
,
- EXEC UTIVE S U M M ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . il
TA B LE O F C O NT ENT S 1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - iv
Summary of Plant Status ...........,................................1 -
1. Operations ....................................................1
01 Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
01.1 Unusual Event Declared Due to indication of Possible Seismic Event
................................................l '
01.2 10 CFR 50.72 Notification involving inadequate Cable Separation of
the Alternate Cooling System . . . . . , . . . . . . . . . . . . . . . . . . . . 9
08 Miscellaneous Operations issues ...... ..................... 3 -!
08.1 (Closed) LER 9 7 1 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
. 08.2 (Closed) LER 9 7-14 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
08.3 (Closed) LER 97-17 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
08.4 (Closed) LER 9 7 12 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
ll . M aint e n a nc e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
M1 Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
M1.1 Maintenance Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
M1.2 Surveillance Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
M8 Miscellaneous Maintenance issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
M8.1 (Closed) LER 96 23 and NCV 9 7-0 8-0 3 . . . . . . . . . ,. . . . . . . . . . 5
M8.2 (Closed) LER 96 29 and NCV 97-08-04 . . . . . . . . . . . . . . . . . . . . 6
- M8.3 (Closed) LER 96 24 and IFl 9 6-0 9 -01 . . . . . . . . . . . . . . . . . . . . . 7
M8.4 (Closed) LER 96-27 and NCV 97 08-05 . . . . . . . . . . . . . . . . . . . . 8
,
Ill . Engine e rin g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
E1 Engineering Support of Facilities and Equipment . . . . . . . . . . . . . . . . . . . 9
E1,1 Safety Grade Qualification of Welds in Emergency Diesel Generator
S up p ort Syste m s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
E2 Motor-Operated Valve Program Review . . . . . . . . . . . . . . . . . . . . . . . 11
E2.1 I ntrodu ctio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1
E2.2 Evaluation of High Risk MOV Dynamic Test Results . . . . . . . . . . 12
E2.3 Use of the Electric Power Research Institute Thrust Calculation
Pr og r a m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 4
E2.4 Valve G rouping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
E8 Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
E8.1 (Open) IFl 96-11-01: Emergency Diesel Generator (EDG) Tornado
Pro t e c t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
E8.2 (Closed) LER 96-11 and NCV 9 7-0 8-0 6 . . . . . . . . . . . . . . . . . . 17
E8.3 (Closed) LER 96-14, Supplement 1. . . , . . . . . . . . . . . . . . . . . . 17
E8.4 (Closed) LER 96 21 ................................18
.
iv
.
--
__ _ . _ _ _ . _ _ . . . . . _ _ _ - . _ .. ._
-
-.--
.
4
,
)
E8.5 (Open) URl 9 7 -0 3 -0 2 . . . . . . . . . . . . . . . . . . . . . . . . -. . . . . . . . 1 8 .
E8.6 '(Closed) URI 9316-01: Pressure Locking / Thermal Binding (PLTB) of l
Gate Valves .....................................18
E8.7 - (Closed) VIO 96-05-03: Update and Control of MOV Program Manual
..............................................19
E8.8 - (Closed) EA 95-070,VIO 01013: Fa" * e to Correct a Condition '
Adverse to Quality . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
IV. Plant Support ................................................20
- P3 EP Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . . . . . . 20
R1 Radiological Protection and Chemistry (RP&C) Controls ... . . . . . . . . . . 20
R1.1 (Closed) URI 97 06-02: Implementation of the Radioactive Liquid and
Gaseous Ef fluent Control Programs . . . . . . . . . . . . . . . . . . . . . 20 -
R2 Status of RP&C Facilities and Equipment ..............,.......21
H2.1 Calibration of Effluent / Process / Area /Acciocet Radiation Monitoring
Systems (RMS) ...................................21 +
'
' R2.2 Air Jiean!ng Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22
R2.3 Plant Air Bala nce . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 3
R3 RP&C Procedures and Docuinentation' . . . . . . . . . . . . . . . . . . . . . . . . 24
- R6 RP&C Organization and Administration . . . . . . . . . . . . . . . . . . . . . . . . 25
,
R7 Quality Assurance (OA) in RP&C Activities . . . . . . . . . . . . . . . . . . . . . 26
R8 Miscellaneous issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 6
R8.1 Training ........................................26
R8.2 (Closed) URI 96-03-05, LER 96-03 Sup 1, and NCV 97-08-07 .. 27
.
V. M a nagem e nt Mee ting s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 8
X1 Exit Meeting Summary . . . . . . . . . .......................,..28 ;
X2 Review of Updated Final Safety Analysis Report (UFSAR) . . . . . . . . . . . 28 ,
INSPECTION PROCEDURES USED . . . . . . . . . . . . . . . . . . . . . . . , , . . . . . . . . . . . . 29
ITEMS OPENED, CLOSED, AND DISCUSSEO . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 i
.
PARTIAL LIST OF PERSONS CONTACTED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31
LIST OF ACRONYMS USED . . ..................................... . 32 ,
,
ATT A C H M E NT A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 3
3
l
v
s
4
l
l
'
-
__
..
.
.
Report Details
Summerv of Plant Status
During this inspection period, Vermont Yankee (VY) operated at full power with the
exception of power reductions to conduct planned surveillance testing.
A region based specialist inspector was on site the week of September 22 to examine VY's
radioactive liquid and gaseous effluent control programs. The results of that inspection
have been integrated into this report.
A region based specialist inspector was on site the week of September 29 to conduct a
follow-up inspection of the Architect / Engineering Design Inspection (report No. 50-
271/97-201) findings. The results of that follow-up inspection will be documented in
inspection report No. 50-271/97 10.
During the week of October 13, region based specialist inspectors conducted a follow-up
inspection of the motor operated valve program developed in accordance with Generic
Letter 8910.
On October 6, the inspectors were provided an overview of the licensee's Human
Performance improvement Program which has the goals of: achieving excellence in human
performance; achieve a reduction in error rate; and achieve an improved rating in human
error probability index. This program was initiated, in part, in response to recent Notices of
Violation (refer to inspection reports 97-04 and 97-05) citing poor human performance and,
in part, to a licensee recognized adverse trend in this area. Training sessions 'with small
groups of the plant staff were scheduled to commence later in the month,
l. Operations
.
01 Conduct of Operations' (93702)
01.1 Unusual Event Declared Due to Indication of Possible Seismic Event
I
a. Insoection Scone (71707)
The inspector examined the licensee's response to a seismic monitor alarm and the
basis of their decision to declare an Unusual Event.
b. Observations and Findinas
At 9:08 pm on September 27, the plant seismic monitor alarmed. Control room
operators were alerted to the event via the seismic monitor main control board
annunciator. There were no other indications that a seismic event (earthquake) had
occurred. The licensee declared an Unusual Event (UE) based on emergency
procedure AP-3125, " Emergency Plan Classification and Action Level Scheme,"
.
' Topical headings such as 01, M8, etc., are used in accordance with the NRC standardized
reactor inspection report outline. Individual reports are not expected to address all outline topics.
!
l .
\'
i
2
entry criteria U 5-c, "Any earthquake sensed on-site as recognized by observation or
detection." The states of Vermont and New Hampshire and the Commonwealth of
Massachusetts were notified and a one-hour emergency notification system (ENS 33001) call was made to the NRC, as required by 10 CFR 50.72.
The inspector observed that the VY staff properly responded in accordance with
operating procedure OP-3127, " Natural Phenomena." This included visual
inspection of selected plant structures for possible damage and the completion of a
seismic damage indicator walkdown. No evidence of earthquake damage was
observed. Civen that no other monitoring stations had detected an earthquake and
that a preliminary investigation of the seismic monitor ideatified an internal failure,
the licer.see declared the seismic monitor inoperable and terminated the UE at 11:10
p.m. bubsequent troubleshooting of the seismic monitor identified that the monitor
battery had f ailed, which caused the electrical transient that resulted in the monitor
alarm.
On May 31,1997, a malfunction of the seismic monitor had also resulted in the
declaration of an UE. As discussed in inspection report 50 271/97-04,the
inspector observed that the procedural requirement to declare an UE based upon a
single indicator was overly conservative. Accordingly, allowance to verify that a
seismic event has actually occurred, prior to making an Emergency Plan event
declaration, would potentially avoid the unnecessary mobilization of state and NRC
emergency response organizations, in light of ths September 27 occurrence, ute
inspector considered that the licensee has been slow to address this procedural
requirement. The inspector determined that procedure revisions were being
processed at the time of the Nptember 27 event, which were designed to provide
for a seismic event verification,if appropriate, prior to Emergency Plan entry,
c. Conclusions
The inspector determ;ned that the VY staff had responded appropriately to the
September 27 seismic monitor alarm and that the declaration of an Unusual Event
had been in accordance with their emergency procedures. However, tile inspector
considered that the timeliness of licensee action to address the previously identified
Emergency Plan entry condition problem, (single, unconfirmed seismic monitor
alarrn), to have been slow.
01.2 10 CFR 50.72 Notification involvina inadeauate Cable Separation of the Alternate
Coolina System
At 7:57 pm on October 10, the control room operators notified the Headquarters
Duty Officer (Event No. 33070)in accordance with 10 CFR 50.72, that a condition
outside the plant's Updated Final Safety Analysis Report (UFSAR) had been
. identified involving power cable separation of Alternate Cooling Systum (ACS)
cooling tower fan No. 21. Spechically, the two emergency power feeds, one from
motcr control center (MCC) 8C (safety related Division 1) and one from MC' 7C
(Division 11), were not properly separated per the UFSAR and Vermont how
Specification VYS-027 electrical separation criteria. The No. 2-1 fan is normally
.- - - - . - . _ . _ - .- .- - - --. . . . . . . .
'
c
..
. 3 l
powered via a non safety related MCC (MCC-582A).- To address the immediate-
operability concern, the licensee tagged open both safety related power supply
breakers (MCC 8C, breaker 2C is normally closed) and the seven-day limiting j
condition for operation (LCO) was entered, in accordance with ACS Technical !
Specification (TS) 3.5.D.3, pending further review,
inspector follow-up determined that the licensee revised the ACS operating
p. ocedure to maintain both of the cooling tower fan No. 2-1 safety related breakers
normally open and provided amplifying instructions for operators to closed the
breakers in the event that the alternate cooling tower fan was needed. The
inspector reviewed the safety evaluat;on (Safety Evaluation No. 97 28) supporting
- the procedure changes to OP 2181, OP-2143, and OT 3122, and found the
licensee's assessment of the changes consistent with 10 CFR 50.59 requirements.
The inspector also observed the Plant Operations Review Committee's deliberation 1
and approval of SE No. 97-28 and concluded their safety review was appropriate. I
The licensee satisfactorily implemented the procedure changes and exited the 1S
LCO on October 16.
08 Miscellaneous Operations issues (92700) j
<
1
i 08.1 (Closed) LER 97-11: The crimary containment torus was not Lnert_ed to Technical !
Soecifications reauirements due to an inadeauate orocedure who resulted in an !
Lqntfficient nitroaen inertino ourae flowrate.
LER 97-11, dated June 1 i,1997, was previously reviewed by the inspectors, as
docurnented in inspectica report 97-C5, section 01.2. As a result of this event, a
- Notice of Violation (VIO 97-05-01) was issued citing the non-compliaace with
Tect:nical Specification 3.7.A.7.b. Inspector review of the licensee's response,
dated September 18,1997, and any additional corrective action verification will be
tracked via VIO 97 05 01. LER 97-11 is closed.
08.2 (Closed) LER 97-14: Lack of understandina of clant licensina and desian bases
results in an inadeauste resoonse to industry operatina exoerience which allowed
resumotion of olant operations inconsistent with its desian basis.
.
LER 97-14, dated September 5,1997, was previously reviewed by the inspectors,
as documented in inspection report 97 06, section E.8.2. As a result of this event,
a Notice of Violation (VIO 97 06-03)was issued citing ineffective corrective action,
inspector review of licensee's response, dated October 1,1997, and any additional
corrective actions verifications will be tracked via VIO 97-06 03. LER 97-14 is
closed.
.
0
k
__. . . _ . , _ ,, , - ,
. - .. . .-
4
.
4
08.3 (Closed) LER 07:,1,7: An eauioment malfunction remainina undetected by thq
poeratina crew renults in olant operation in excess of rated thermal power.
LER 9717, dated October 2,1997, documented the licensee's assessment and
corrective acti.ans for the violation of the reactor thermal power limit which occurred
on Septemb'.,r 2,1997, due to a plant process computer data acquisition system
componen*. f ailure. This event was previously reviewed by the inspectors and
documented in inspection ieport 97-06, section 01.2. As stated in report 97-06,
this non-compliance with the VY thermal power limit was non-cited, consistent with !
section Vll.B.1 of the NRC Enforcement Poliev. The inspector determined that LER
9717 clearly and concisely deceribed the circumstances involving this event and ,
that the action taken by the VY staff to correct the problem and preclude a
recurrence were appropriate and well documented. LER 9717 is closed.
08.4 (Closed) LER 97-12: Residual heat removal service water fk,w could be potentially
less than the desian. basis flow due to instrument inaccuracies.
This event was previously discussed in Inspection report 97-04, section E.7.1 and
assigned an inspection follow item (IFl 97-04-04). The root cause for this event
remains under investigation. However, a Basis for Maintaining Operation (BMO) No.
97-27, dated June 13,1997, was initiated to summarize the residual heat removal
(RHR) service water system operability assessment and document the correctM
action plan. The inspector reviewed the licensee's interim corrective actiont I!nd
found them to be appropriate. Adequate RHR service water system cooling
capacity wcs demonstrated, via analysis, provided river water temperature remained
equal to or less than 80 degrees F (revised from the May 2,1997 limit of 70
degrees F). As of the conclusion of this inspection period, BMO No. 97 27 was
still in of fect.
LER No. 97-12 is c;osed. However, the licensee's actions to resolve this issue will
continue to be tracked via IFl 97 04-04. The inspector notes that the broader issue
of instrumentriion accuracy was identified as a concern in inspectioni report 97-201
(reference sectioli E.2,2.2.f, URI 97 201-16)and will be tracked separately,
ll Maintenance
M1 Conduct of Maintenanc.e
M 1.1 Maintenance Observations
a. Insoection Scope (62707)
The inspectors ob.,erved portions of plant maintenance activities to verify that the
, correct parts and tools were utilized, the applicable industry code and Technical
Specification requirements were satisfied, adeauate measures were in place to
ensure personnel safety and prevent damat. o plant structures, cystems, and
components, and to ensure that equipmer gerability was verified upon completion
of post-maintenance testing,
. . . _ . - . . - . - - - - .- _ .. . --
l.
,
5 ,
b.- Qbservations. Findinas, and Conclusions
The inspector observed all or portions of the following maintenance activities:
- Preventive maintenance to MCC 10C, on September 30.
- Scram solenoid pilot valve replacements (18 31,22-43, and 38 27), on
September 12.
- 'A' emergency diesel generator lubricating oil piping replacement, on .
Octob6.' 23.
on October 20 and 21.
The inspectors observed proper adherence to procedure and appropriate control and
execution of the above activities.
'
M1.2 Surveillance Observations
a. Inspection Scope (61726)
The inspectors observed portions of surveillance tests to verify proper calibration of
test instrumentation, use of approved procedures, performance of work by qualified
personnel, conformance to limiting condition for operations (LCOs), and correct
post test system restoration.
b. Observations Findinas, and Conclusions
The inspector observed all or portions of the following surveillance tests:
- Core spray system quarterly surveillance test, observed October 7.
- 'B' emergency diesel generator monthly testing, observed on September 22.
.
The inspectors observed proper adherence to procedure and appropriate control and
execution of the above activities.
M8 Miscellaneous Maintenance issues (92700,92903)
M8.1 ' (Closed) LER 96-23 and NCV 97-08-03: Inadeouate surveillance orocedure results
in failure to meet Technical Soecification reauirements for radiation monitor
'
functional testino.
LER 96-23, dated October 15,1996, documented a licensee identi9sd logic system
functional testing deficiency discovered during the biennial review of procedure OP-
4326. " Reactor building ventilation and refueling floor radiation monitors
functienrWealibration." After identification of the testing oversight and revision of
the surveillance procedure, the radiation monitors' high alarm output contacts,
previously not verified to actuate, were tested satisf actorily. Consequently,
although OP-4326 did not satisfy the Technical Specification functional testing
- requirements (per TS Table 4.2.3), the radiation monitors were demonstrated to
i
_ _ _ ,
- - - - . -. - - . .- - - -
. -
.
6
function, as designed. This non-repetitive, licensee identified and corrected
violation was treated as a non-cited violation (NCV 97 08 03), consistent with
Section Vll.B.1 of the NRC Enforcement Poliev.
.
The inspector noted that concurrent with this event, the VY staff was conducting a
< re evaluation of their logic system functional test (LSFT) procedures in accordance
with their April 18,1996 response to Generic Letter (GL) 96-01. By letter dated
September 20,1997, VY revised their August 31,1997 commitment to complete
GL 96-01 actions by February 20,1998. Inspector review of the licensee's
completed LSFT actions is being tracked by IFl 97 06-01 (reference inspection
report 97-06, section M1.5). LER 96-23 is closed.
M8.2 (Closed) LER 96-29 and NCV 97-M-04: Process and communication inadeoue::les
result in the failure to analyze em. roency diesel oenerator fuel oil within time
4
allotted by Technical Soecification surveillance reouirements. .
This LER was previously discussed in Section R8.2 of NRC Inspection Report 50-
271/96 11. The NRC concluded in that inspection report that the licensee's
description of the reported violation of plant Technical Specification (TS)
surveillance requirements was incorrect, in that the licensee erroneously assumed
that the diesel generator fuel oil sampling and quality verificatica surveillance
requirement (TS 4.10.C.2) had two separate surve!!!ance intervals, one for the act
of sampling the fuel oil and another for the ana!ysis of the sample (quality
verification).
I in this inspection period, further NRC follow up of this LER and the associated
requirerrents identified additional findings and a necessary clarification to the prior
inspection report discussion. The prior inspection findings included a statement that
"this TS requires the fuel oil to be sampled every 30 days and implies that the
sample sheuld be analyzed prior to the next 30-day sample being taken." Upon
further review, the NRC recognized that the actual requirement of the VY TS was
based on a "once a month" requirement and not "30 days" as stated in inspection
, report 50 271/96 11. While this difference does not change :he overall NRC
conclusion that no violation of the TS occurred, the interpretation of the
requirement in the previous inspection report was not completely accurate. To
clarify, the NRC determined that-TS 4.10.C.2 requires the diesel generator fuel oil to
,
be sampled once a month. Implied with this requirement is that the sample analysis
be completed prior to the next monthly sampling activity. While noting that there
are various interpretations of "once a month," the NRC concludes this could be as
long as 31 days or as short as 28 days. For exampie, if the surveillance is
conducted on the 15th of the month, the next ',urveillance would be due on the
15th of the next month. In addition, the NRC noted that the surveillance interval
could be extended by a plus 25 percent, in accordance with the licensee's TS
definition for surveillance frequency.
_
The inspector reviewed the licenseo's internal event report documentation for this
issue and determined that once identified, the concern was appropriately handled by
p station personnel. Based on the fact that the analyses results were already known
!
,
I
.-, - ,m w,..- .-m_, , --e w-
- . - . .- . . - -. - - .
.
.
7
to be acceptable, alt'eit late, and since the fuel oil quality integrity had been
maintained appropriately throughout, the licensee concluded that the emergency
diesel generators were unaffected by this event. Based on the review of the
licensee's analysis at the time of the event, the inspector agreed that the '
emergency diesel generators remained operable.
.
As outlined in the discussion above, the NRC concluded that no technical
specification violation occurred as stated in the LER. Upon review of the timing of
the sampling and analysis of the diesel fuel oil contained in the LER, the inspector
determined that the surve:llance requirements were met. However, the NRC
concluded that the licensee failed to implement station procedures used to schedule
and track the timely completion of important-to-safety activities, like TS required
surveillance tests. The f ailure to properly implement the associated station
procedures was a violation. The licensee's corrective actions described in the LER
were determined appropriate to correct this procedure violation. This non-repetitive, !
licensee identified and corrected violation was treated as a non-cited violation (NCV
97-08-04), consistent with Section Vll.B.1 of the NRC Enforcement Poliev.
M8.3 (Closed) LER 96-24 and IFl 96-09-01: Incomplete desian bases documentation
results in a f ailure to clearly describe Anoendix J methodoloav in the oroaram
descriotion delivered to the NRC for evaluation.
On October 2,1996, the VY staff notified the NRC staff that an engineering
evaluation had concluded that the lack of closure capability of the motor operated
core spray minimum flow valves (CS-5A and SB) was a condition outsde the plant
design basis. The ikensee had determined that the valve wiring and logic prohibited
minimum flow valve closure unless the core spray pump was running with injection
flow. This valve logic and wiring condition resulted in the inability to close the
minimum flow valves for containment isolation purposes. At that time, the licensee
modified the core spray minimum flow valve logic to permit valve closure from the
'
control room.
The inspector reviewed the licensee's evaluation of this event and corrective
actions. The licensee's evaluation considered various aspects of the plant's design
i and licensing bases and resulted in clearly determir.ing that the original design basis
i for this system did not require the minimum flow valves to be containment isolation
,
valves. That portion of the system was required to open for accident purposes and
was considered an extension of the containment boundary. This position was also
clearly reflected in the licensee's response to the TMl Action Plan for containment
'
isolation dependability, as stated in a licensee letter to the NRC on January 8,
1980. At that time, the core spray system was identified as one of a number of
systems that communicated directly with the containment space without an
automatic isolation valve. The licensee implemented routine inspections of the
associated piping as a means for ensuring the integrity of the containment
boundary. The licensee's evaluation noted that the conflicting information between
the design and licensing bases rogarding containment isolation capability .or the CS
mini-flow valves resulted from an incomplete review of the FSAR requirements for
i these particular valves. That resulted in an error translation into the Appendix J
.
_
. . _
,
8
mini-flow valves resulted from an incomplete review of the FSAR requirements for
these particular valves. That resulted in an error translation into the Appendix J
program. The licensee's corrective actions appropriately addressed the causes of
the design bases documentation error. Further, as a result of the licensee's review,
they modified the controls design for these valves in order to enhance the
operator's capability to erisure containment isolation by installing a remote manual
isolation function.
The inspector concluded that the licensee's evaluation, root cause determination
and corrective actions were acceptable, in that the design bases of the plant was
accurate, and that the as-built configuration met the design basis, this condition
was not a violation of NRC requirements. The licensee's action to modify the plant
configuration to provide an enhanced operator control for containment isolation
function was viewed as a positive measure. LER 96-024is closed.
M8.4 (Closed) LER 96 27 and NCV 97-08-05: Lack of reouired yerificadons results in
inconsistency between technical specification instrument settino description and the
as-built confiouration of the low oressure coolant iniection (LPCI) oumo control
lonic.
This LER describes the licensee's discovery of use of a time de;ay relay with a
setpoint inconsistent with the TSs while testing the LPCI system actuation log c
during the refueling outage in October 1996.
The time delay relay minimum time delay cetting was 0.55 seconds for LPCI pump
start and the plant TS recuired no time delay for two affected LPCI pumps. The
licensne determined that the installed time delay relays were consistent with the
materials used since initial plant startup and that the plant TS requirements (TS
7,4.3.5.2) have also not changed since initial plant startup. Therefore, this
inconsistency between the as-built design and the TSs always been present. Due
to the age of the issue, the licensee was not able to determine an actual root cause.
However, the apparent cause was an inadequato verification of the license
requirements versus the system design specifications during the development of the
TSs. The licensee replaced the time delay relays with a modified design to permit
instantaneous starts of the affected LPCI pumps. This corrected the inconsistency.
Further, the licensee was already implementing a major Technical Specification
improvement project that would result in verifying that the TSs and as-built design '
criteria were consistent.
The inspector concluded that the licensee's assessment, root cause determint.aon,
and corrective actions for this event were appropriate. However, failing to ensure
that tha LPCI surveillance tests met the acceptance criteria stated ir. the TSs was a
violation of the TS. This non-repetitive, licensee identified and corrected violation
was treated as a non-cited violation (NCV 97-08-05), consistent with Section
Vll.B.1 of the NRC Enforcement Poliev.
.
O
$
.- - _~ . . . _ . _- - .. _ - - - -
,
.
9
lil. Engineering
E1 Engineering Support of Facilities and Equipment
i
E1.1 Safety Grade Qualification of Welds in Emeroency Diesel Generator Suocort !
Systems
-l
a. Backoround and insoection Scope (93702,92903,37551) 1
i
1
The inspector observed and assessed the licensee's response to an industry event
at Millstone, involving safety class support systems (such as the Jacket water
cooling and lubricating oil systems) that had been fabricated and installed as part of
the EDG unit by Fairbanks Morse which were apparently not welded to ANSI B31.1 ,
standards or an equivalent, i
-
b. Observations and Findinos
On September 4, the inspector discussed the EDG subsystems weld issue with VY
< systems engineering staff. The inspector was informed that VY had received
information about the problem and would be investigating. The inspector visually
examined piping welds in the EDG lubricating oil and jacket water cooling systems.
- The inspector observed that there was a strong possibility that the VY EDG
subsystem piping had likewise not been welded to ANSI B31.1. The inspector
.
observed that the welds had not been ground smooth to support any form of non-
destructive testing, and areas of concavity existed in some welds Due to the
'
i
potential operability impact on both EDGs, the inspector p:omptly discussed this
issue and his preliminary observations with the plant manager. .
l
On September 10, Event Report (ER) No. 97-1224 was generated which addressed )
the potential weld problem with the EDGs. The immediate operability determination
was that the EDGs were operable, bamd on the vendor's conclusion that the
Millstone EDGs' piping welds had been found to satisfy Northeast Utilities' Millstone
'
Unit 2 seismic analysis, and based on a walkdown by engineering personnel who
judged the welds to be satisfactory by visual examination. ER 97-1224 was
reviewed by plant management during the September 11 ER screening meeting.
Initially, VY did not consider the Millstone problem to be an immediate concern
because their procurement specifications had been different than Millstone.
Specifically, Millstone had purchased the EDGs and dona their own seismic analysis,
, whereas VY had specified in the procurement specifications that the EDGs were to
be fabricated and delivered seismically qualified. The inspector expressed concern
regarding discovery of the partial penetration welds and the implications uf this
discovery on the seismic qualification of the VY EDG welds. He discussed this
concern with both NRC regional management and VY station management.
Subsequently, VY initiated development of a Basis for Maintaining Operation (BMO)
for the EDG weld issue, to be completed by September 17.
.
s
, , - , .- -
.,.
. . - - -. . - - -. .-
t .
l ,
10
- BMO 97 39,"Possible Less than Full Penetreion Welds on Vendor Supplied Skid
Mounted Piping for the Emergency Diesel Generators," was reviewed by the plant
operations review committee (PORC) on September 17. The licensee determined
the EDGs were operable based on:
1. Visualinspection of the welds ' hat showed no obvious external defects.
.
2. The vendor's position that the w? ids were deemed acceptable based on many
'
years of successfulin-service operation of their equipment.
3. Succusful operation of Fairbanks Morse diesels in harsh environments, including
temperature, vibration, and shock.
4. Analyses that had been performed by ancther licensee, which indicated that less e
than full penetration welds (in the;r case,66 porcent) were acceptable. ;
5. VY EDGs had been evaluated as part of the seismic qualification upgrade
(SOUG); the SQUG data base included diesels of similar vintage that had gone
through seismic events of magnitudes in excess of VY's design basis and did not
fall.
6. Preliminary calculations indicated that the EDG piping of concern had significant
margin to ASME Code B31.1 limits, for both normalloading and seismic loading.
'
On September 22, VY started an LCO maintenance outage on the 'B' EDG. As a
result of the weld issue, a section of lube oil system piping was removed for
destructive examination. The piping is approximately 4-inch diameter and contains
6 welds. There are two material thicknesses,0.250 and 0.120-inch, and three
weld combinations,0.250 to 0.250,0.250 to 0.120, and 0.120 to 0.120. Weld
penetration was determined by an off-site lab (Massachusetts Materials Research)
to be 50% for the 0.250-to-0.250 weld, and 22% for the 0.120-to-0.12Oweld.
During the inspection, a crack was found in the 0.250-to-0.250 weld, through wall,
starting from the root. The crack was about 0.5-inch in length, or 4% of the
circumference. The crack was on the same weld (first weld downstream of the LO
pump discharge) and in tt.e same location on the weld as had developeo a leak at
Millstone Unit 2. The VY engineering staff concluded that the crack would self-
arrest at about 120* circumferential due to reaching the compressive side of the
weld, and would result in a leak rather than a catastrophic failure. During this
inspection period, VY had also conducted structural testing by applying tensile
- stress equivalent to the operating stress, and then applying a bending moment of 6
times the combined operating and seismic stress to the weld. The weld did not fail
this structural test,
c. Conclusions
At the end of the inspection period, the licensee had completed formal EDG support
piping stress analyses and had completed a metallurgical analysis which the licensee
._
._-_a
_ _
,
11
has concluded supports their initial operability determination. These items were
under review by the NRC staff. Pending the results of further NRC staff review, the
EDG support piping welds issue is being tracked as an inspector follow-up item (IFl
97-08-01). The licensee's response to this issue, to date, has been consistent with
the guidance of Generic Letter 91 18 for identification and resoluiion of a degraded
or non conforming condition.
E2 Motor Operated Valve Program Review (Tl 2515/109)
E 2.1 Introduction
On Jun9 28,1989, the NRC issued Generic Letter (GL) 8910, " Safety-
Related Motor-Operated Valve Testing and Sur" ': lance," wWeh requested
licensees to establish a program to ensure that switch settings for safety-
related motor operated valves (MOV)s were selected, set, and maintained
properly. Seven supplements to the GL have been issued to provide
additionalinivrmation and guidance on development of programs,. NRC
inspections at Vermont Yankee (VY) were conducted based on guidance
contained in NRC Temporary Instruction (TI) 2515/109," Inspection
Requirements for Generic Letter 8910."
On December 29,1995, VY notified the NRC that the GL 89-10 program was
corr.plete. The NRC had previously conducted an initial programmatic (Part 1)
inspection at VY in May 1991, cs documented in Inspection Report (IR) 91-80.
During October 1993, the NRC performed an implementation (Part 2) inspection, as
documented in IR 9316. A closure (Part 3) inspection for the purpose of verifying
that VY completed its commitments to develop and implement a safety-related MOV
program as described in GL 89-10 and its supplements was performed in May
1996,(IR 96-05). During that inspection, the NRC determined although the VY
staff had generally implemented an acceptable GL 8910 program, the following
items were noted:
- Design basis evaluations of non-dynamically tested MOV's in accordance
with Attachment 6 of " Engineering Guideline for Evaluation of Motor-
Operated Valve Design Basis Capability" were nct completed.
- The assumptions applied to grouped MOVs were not adequately cupported
by test data.
To address the above items, in letters dated April 18, and May 9,1996,
respectfully, VY agreed to use the Electric Power Research lnstituto (EPRI) thrust
performance preo.ction program on six MOVs identified in GL 8910 supplement 3
and two valves classified as "high risk" in the Individual Plant Examination (IPE)
report. Additionally, the design basis evaluations of non-dynamically tested MOV's
described in the MOV program manual would be completed by July 1,1996.
Finally, fifteen additional "high risk" valves would receive dynamic tests during the
1996 refuel outage.
4
1
l
'
12
The purpose of this fourth inspection was to examine the actions implemented at
VY to address the closure i, sues identified during the Part 3 inspection and
determine if those actions were sufficient to warrant " closure" of the NRC staff
review of the GL 89-10 MOV program.
E2.2 Evaluation of Hiah Risk MOV Dynamic Test Results
a. Insnection Scoce
Fifteen valves received dynamic tests durir g the previous refuel outage. Of those
va;ves, the inspectors selected the test results for the following MOVs for review:
V10-16A/B Residual Heat Removal (RHR) Pump Discharge Mini Flow
Returns to the Suppression Pool
V10-25 A/B RHR to Recirculation Loop Isolation Valves
V10-39 A/G RHR Containment Spray / Suppression Pool Cooling Supply
Valves
V14-5 A Core Spray Pump Minimum Flow Valve
V70-19B Service Water Supply Header Cross Connect Valve
The review consisted of examining data associated with: (1) valve factor, which
correlates differential pressure to the stem-thrust requirement; (2) stem friction
coefficient, which affects the conversion of actuator output torque to valve-stem
thrust; and (3) rate of loading or load sensitive behavior, which reflects the change
(usually a loss) in deliverable stem thrust under dynamic conditions as compared -
with the available thrust measured under static conditions. The inspectors also
reviewed, " Vermont Yankee Engineering Guideline for Evaluation of MOV Design
Basis Capability," Rev.1, dated March 12,1996, and calcu:ctions which evaluated
difierential pressure tests performed on the Residual Heat Removal (RHR) Service
Water (SW) and Core Spray (CS) systems.
b. Observations and Findinas
General
The " Vermont Yankee Engineering Guideline for Evaluation of MOV Design Basis
Capability," outlined the process used to establish MOV switch settings, evaluate
data and monitor valve performance. The document also contained the
assumptions used to determine valve factor, load sensitive behavior, stem friction
coefficient, and various capability margins. The engineering guideline also specified
the statistical methods used to ovaluate multiple test results.
When performing MOV testing (under static or dynamic conditions), valves were
stroked three times in each direction. This allowed personnel to assess the valve's
ability to perform in a consistent manner. Each performance parameter was
determined or evaluated using a " student's t~ statistical evaluation of the three test
results using a 95% confidence level. The inspectors noten VY completed the
- evaluations required in Attachment 6 of the MOV program manual. Therefore, tnis
closure item identified in NRC inspection report 50-271/96-05,was resolved.
. _ _ _ _ _
.
,
s
13
.
Test Results
VY used the standard industry equations and a statistical evaluation of three
dynamic tests (for each valve), to determine actuator capability margins, structural
margins, valve f actors, load sensitive behavior, and stem friction coefficients. MOV
. performance parameters (i.e. , valve f actor, load sensitive behavior, and stem
friction coefficient) were compared with previous dynamic test results to verify
program ascumptions we's valid.
Apparent valve factors and loari sensitive behavior values measured during dynamic
testing of the selected MOVs were bounded by the current program assumptions
(i.e.,0.60 for gate valve f actors,1.10 for globe valve f actors, and 10% margin for
load sensitive behavior). However, the test results for globe valveh V10 34B and
V13 27 had measured load sensitive behavior values of 17.9% and 13.8%,
respectfully. Further, the test results were not fed back into the individual valve's
component thrust calculation.
The inspectors noted this observation appeared to be restricted to a few valves. To
ensure future MOV thrust calculations ref.'ected the results of test data, VY
comm;tted to rev!so the existing calculations to reflect the results of test data for
each dynamically tested valve. Program documents would also be revised as
appropriat- +L 'eflect this expectation. The inspectors determined the corrective
action was :%,ropriate to resolve this obsvvation,
based on the load sensitive behavior performance noted for valves V10 34B and
V13 27, the inspectors reviewed the dynamic test data for the remaining population
of globe valves. Tnis review included performing a " student's t" statistical analysis
of all available in plant globe valve load sensitive behavior data. Based on this
review, the irspectors determinod an 18% margin for load sensitive behav;or should
be applied to notedynamically tested globe valves. The inspectors also performed a
similar review of globe valve dynamic stem friction coefficient performance and
determined that the assumeo value of 0.15 was non conservative as compared to a
0.16 value that resulted from analysis of the wailable in plant globe valve test data.
Although the VY staff's assumptions for load sensitive behavior and stem friction
coefficient did not bound the majority of the globo valve test data, the inspectors
noted this finding did not effect valve operability since the non-dynamically tested
globe valves had adequate design margin. At the conclusion of this inspection, VY
committed to perform a review of the globe valve dynamic test data and provide an
additional allowance to account for the offects of Icad sensibve behavior for non-
tested globe valves.
l
c. Conclusions l
l
'
Overall, the dynamic test results reaffirmed the design assumptions used to
establist h0V switch settings. The exception was the load sensitive behavior l
I
assumpthn (c.' globe valves, which did not appear to bound the majority of the test
data, in some instances, component calculations were not updated to reflect the
l
--
_ _ ____ _ _ _ . - _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ . ___
,
.
-
e
<
14
latest test data. Neither of these observations was significant, since the examined -
MOVs had adequate capability and the thrus( calculations were generally up to-
date. VY committed to resolve these items by January 30,1998, which was
acceptable for program closure.
E2.3 Use of the Electric Power Research Institute Thrust Calculation Procram
a. insoection Scope
The inspectors reviewed calculations performed on valves in the Reactor Water
Cleanup (RWCU), High Pressure Coolant injection (HPCl), and the Reactor Core
isolation Cooling (RCIC) systems to assess how VY used the Electric Power
Research Institute (EPRI) performance prediction methodology MOV thrust
calculation program. Additionally, the inspector reviewed a summary analysis,
which compared the thrust calculated by the EPRI program to the thrust produced >
by valves with their current switch settings.
'
b. Observations and Findinas
The VY staff had properly used the EPRI calculational program. Specifically, def ault ,
friction coefficients were used when necessary, the appropriate valve disc and
guide material comt,inations were specified, and the system model used blowdown
flow parameters to establish the design basis requirements. However, one
exception was noted, V( did not perform the calculations needed to estimate the
unwedging loads for valves V13 21 and V2319, which had open safety functions.
VY indicated these calculations would be completed by December 1997.
The six Supplement 3 MOVs were evaluatect for the closing safety function under
blowdown flow. The initial EPRI calculation for these valves indicated that the
thrust requirements were unpredictable. This result was caused by the software
inputs that specified sharp guide edges for the valve disc. VY revised the input
values to reflect a 0.04" chamfer for the guide edges, which resolved the
sof tware's unpredictable results. This change was based on valve internals
inspections performed on all of the affected valves.
The results of the EPRI program were reconciled in an analysis, dated March 10,
1997, which compared the EPRI predicted thrust requirements to the current thrust
requirements contained in the component calculations. The existing in-plar.t valve
switch settings exceeded the EPRI predicted thrust requirements for ull six
Supplement 3 MOVs. However, the in-plant open thrust requirements were non-
conservative for two non Supplement 3 valves, which had open safety functions.
The VY staff indicated both valves had adequate thrust capability to ensure proper
operation considering the higher EPRI values. Therefore, the EPRI results did not
affect valve operability.
VY is corrntly using the EPRI program to validate the current switch setting on all
applicable valves. The inspector considered this initiative to be a program strength.
--
_.
-- . - , - _ . - _ . , .- - - - - -_ . -.
_ _ . - . _
.
.
15
c. Conclusions
VY properly used the EPRI program to develop predicted thrust values. The
exception was the failure to complete and apply the unwedging hand calculations
for two valves. This omission will be corrected by December 1997, which would
be acceptable for program closure.
E2.4 Valve Grouoina
a. Insocction Scan
The inspectors reviewed the grouping methodology used to analyze the performance
characteristics of non dynamically tested MOVs. The review consisted of an
examination of dynamic test data and grouping criteria outlined in the MOV program
manual. !
b. Observations and Findinas
VY divided their MOVs into six valve groups based upon manuf acturer, type, and
stem orientation. However, the inspectors determined that the grouping criteria
were too broad to provide meaningful comparisons between valve types.
For example, one group contained Anchor Darling double disc gate valves which
ranged in size from four to 28 inches. VY was not able to dynamically test any of
the val /es, so a valve factor of 0.50 was assumed based on analysis of eleven
Anchor Darling valves tested at another nuclear station. However, the inspectors
noted that the largest valves tested at the other station were six inches in diameter.
It wac not evident that this data would be applicabl' to all valves in this group
population. The VY staff was also unsure if the data was obtained from valves
oriented in the " preferred" direction (i.e., with the lower wedge downstream).
Industry testing has shown that disc orientation in the "non-preferred" location can
increase the thrust requirements to close the valve.
Based upon this observation, the VY staff committed to improve upon the current
grouping methodology by anulyzing the performance of the non dynamically tested
valves using the EPRI PPM program, if the thrust predicted by the EPRI program !.,
greater than the current MOV setup, the licensee committed to revise the MOV
switch settings. The licensee agreed to complete the validation process by
January 30,1998,
c. Conclusions
Although the current MOV grouping criteria were questionable, VY intends to
address this observation by using the EPRI PPM program and adjusting MOV switch
settings as appropriate bs January 30,1998. The mspector concluded this
approach would be acceptable for program closure.
.-
__
l
1
- 1
l
16
E8 Miscellaneous Engineering issues (92700)
E8.1 IQuen) IFl 96-11 01: Ememency Diesel Generator (EDG) Tornado Protection
a. Backaround and Inspection Scone (92903)
Licensee staff follow up of an industry operating event report (NRC Information
Notice 96 06, involving plant structures' tornado pressure relief), identified that the
EDG enclosures did not contain the differential pressure relieving capability specified
by the construction drawings. EDG operability was promptly assessed by the VY
staff and the Basis for Maintaining Operation (BMO) process was initiated. As
documented in inspection report 9611, the inspector foand the compensatory
actions taken to have been appropriate, however, final resolution to this plant
design issue was pending. The inspector conducted a follow up inspection of this
issue to evaluate the licensee's corrective actions and their progress in resolving
this design concern.
b. Observations and Findinas
The inspector determined that the current revision to BMO No. 96-08,"Effect of
design basis tornado pressure load on the diesel / day tank room enclosure "
(revision 4, dated 9/11/97) maintains the compensatory measures to block open the
= EDG enclosurs access doors when tornado conditions are anticipated. These
manual actions have remained in effect, even though the EDG enclosures were
modified in April 1997 with automatic spring actuated differential pressure relieving
dampers. The inspector determined that the compensatory measures remain in
place to address a discrepancy found in the tornado loading accident analysis.
Specifically, the discovery by the VY staff that the turbine building does not have a
pressure relieving capability in the event of a high energy line break (HELB)
(reference inspection report 97-02, section 01.4), potentially invalidates the
analysis assumption that the turbine building pressure is essentially atmospheric at
time zero in the tornado loading ac-ident scenario response time line.
Consequently, the design basis tornedo load (300 mph winds whh an
accompanying 3 psig pressure change in 5 seconds) impacting on the current EDG
enclosure and turbine building would petontially result in the EDG to turbine building
concrete block wall being subjected to a differential pressure in excess of its
allowable design limit (1 psig).
The inspector determined that engineering design change request (EDCR) 97 419 is
under development to address the turbine building HELB pressure relieving concern
and is targeted for field installation by December 1997. Preliminary discussions
with the responsible design engineering staff and plant management identified that
EDCR 97-419 was not originally proposed as a vehicts to resolve the apparent
discrepancy with the tornado accident analysis time-line assumption. However, the
licensee acknowledged the inspector's observation that the two design issues were
connected and, at the conclusion of the inspection period, VY was examining
avenues to resolve this apparent design assumption conflict within the scope of
EDCR 97 419.
_ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _
.
.
17
c. Conclusions
The VY staff's pursuit of resolving the EDG enclosure tornado ioading differential
pressure vulnerability demonstrated a generally thorough examination of ro ited a
turbine building HELB structural design attributes. IFl 961101 remains open,
persding inspector review of the VY staff's final resolution of this EDG enclosure
differential pressure concern.
E8.2 (Closed) LER 9811 and NCV 97-08 06: Failure to perform Intervice Testina (IST)
on valves that should have been included in the IST Pronram.
These licensee identified IST Program deficiencies involved the f allure to include the
high pressure coolant injection (HPCll, reactor core isolation cooling (RCIC), and
core spray (CS) systems' alternate keep fill system (condensate transfer system)
check valves (V23 208, V13 20B,V14 22A/B, and V14 23A/B)in the IST Program ,
< for quarterly reverse flow cessation stroke testing. The identification of these
testing oversights was part of an ongoing comprehensive IST Program review
initiated in late 1995 (reference inspection reports 95 22 and 95 23 and associated
LERs 9517 and 96 01).
As stated in LER 9611, dated May 16,1996,the Aoril 25 radiography testing (and
subsequent testing) of the effected check valves identified proper valve seating to
prevent reverse flow. The VY staff revised the IST Program to include these valves
for future testing and continued their comprehensive IST program review with no
additional discrepancies noted. The inspector determined that VY appropriately ,
'
documented and reported this event. The inspector also verified that a dedicated
IST Prooram coordinator was assigned, as stated in LER 9611. The inspector -
assessed that corrective actions taken for prior violations in this area would not
have reasonably prevented this violation. This licensee identified and corrected
violation was treated as a non cited violation (NCV 97 08 06), consistent with
Section Vll.B.1 of the NRC Enforcement Policy. LER 9611 is closed.
E8.3 (Closed) LER 96-14 Sucolement 1: Tornado orotection not orovided for diesel
aenerator rooms as specified in the Final Safety Analysis Report due to f ailure to
implement olant construction /confiauration chanae documenth
Supplement 1, dated January 29,1997, documented the results of the licensee's
root cause evaluation for this event and the associated corrective actions. As
previously documented in inspection report 9611 and in Section E8.1 of this report,
this condition is being addressed via BMO No. 96 08, revision 4. As stated in
Section E8.1,IFl 961101 remains open to track licensee final resolution and
inspector review of this issue. LER 9614, Supplement 1 is closed.
.
,
18
'i94 IClh ) LER
1 96 21: Inadgquate orocenural controls of MOV limit switch settinos
$1ML 3 ootential commen cause f ailu.e mode with the capacity to affect multinie
NM unnificant componente.
.
LER 96 21, dated October 7,1996, was reviewed in congnetion with inspection
follow up item IFl 96-09 03, which was closed in inspection report 97 05, section
,
E8.2. As previously documented, this licensee identified and corrected violation
was treated as a non cited violation consistent with Section Vll B.1 of the N_flQ
Enforcement Poliev. The inspector concluded that LER 96 21 appropriately satisfied
the NRC reporting requirements of 10 CFR 50.73. LER 96 21 is closed.
E8.5 (Onen) URI 97-03-02: Cable separation does not satisfy UFSAR seoaration criteria.
This unresolved item portains to a number of licensee identified discrepancies
between the as built wiring at the plant and the cable separation criteria stated in
the UFSAR associated with safety related circuits. The following information
updates the unresolved item.
The licensee reported in LER 96-028, dated November 18,1996, another example
of inadequate cable separation for the LPCI outboard isolation valve actuating
circuits. This was identified and corrected during the October 1996 refueling
outage. The licensee determined that the root cause for this event was an
erroneous mis-labeling of the associated circuit i during a 1976 design change.
While the 1976 modification did not cause the circuits to violate the electrical
separation criteria in the FSAR at that time,it resulted in the engineers not
recognizing that the circuits were required to be separated since they were from
different electrical divisions. Subsequent modifications resulted in the circuts being
placed in a common, non nuclear safety panduct, in violation of the cable separation
criteria.
The inspector concluded that the licensee's immediate corrective actir<ns were
appropriate and brought the configuratior of the affected circuits bacx into
conformance with the FSAR criteria. Accordingly, LER 96 28 is closad, but the
overallissue regarding cable separation remains unresolved pending further review
of the licensee's evaluation and additional corrective actions, as necessary to
prevent recurrence. A determination of future enforcement action for this
unresolved item will include information from the licensee's efforta discussed in LER
96-28. URI 97 03-02 remains open.
E8.6 (Closed) URI 9316-01: Pressure Lockina/ Thermal Bindina (PLTB) of Gate Valves
This item was opened to track the status of VY's corrective actions for gate valves
determined to be susceptible to pressure locking, in a letter dated February 8,
1996, VY described the process used to evaluate valves for susceptibility to PLTB
and its program to modify valves that may be susceptible to these phenomena.
Currently only V13 20, the Reactor Core Isolation Cooling (RCIC) injection test
valve, is susceptible to pressure locking. To address the pressure locking concern
for this valve, a hole will be drilled in one side of the valve disc daring the 1998
refuel outage. The actions taken by the licensee to address PLTB are currently
_ _ _ - - - - - - -_- - - - _ _ - - - ..- - - -. - -.- . - .
.
. .
19
f
under the review of the Office of Nuclear Reactor Regulation (NRR) who will !
evaluate the acceptability of the program in a safety evaluation report. Since NRR is '
t
now tracking the status of the PLTB program, maintaining a separate redundant
unresolved item for the same purpose is no longer necessary. This item is closed. :
,
E8.7 (Closed) VIO 96-05 03: Uodate and Control of MOV Proaram Manual i
'
This violation was issued to document that the MOV program manual was not being ,
updated in accordance with VY station administrative guidelines. As a result, the
program manual contained outdated, conflicting, and contradictory information. To
address this violation, the VY staff revised the program manual so that it described
the current MOV practices. Further, the manual was placed under the document
i
control procedure AP 6805 to ensure future program changes are routed and
distributed in a timely manner. The inspector reviewed the program document and
verified it described the approach used to establish MOV switch settings. The i
inspector concluded the licensee's actions were appropriate and this item is closei
E8.8 (Closed) EA 95 070, VIO 01013: Failure to Correct a Condition Adverse to Quality ,
This violation documented weaknesses in the VY's corrective action processes
including a f ailure to correctly analyze the susceptibility of core spray injection
,
valves to pressure locking. As outlined in NRC inspection report 50 271/96-05,the
inspector concluded VY's corrective actions appeared to be acceptable. However,
1
two areas for improvement were noted and the violation was not closed.
Specifically, the Basis for Maintaining Operability (BMO) guideline did not limit the
time a degraded condition could remain in service before full equipment qualification
was restored. Further it was not evident, the Nuclear Safety Audit Review
Committee (NSARC) had performed a comprehensive assessment of the
eMectiveness of the corrective actions implemented in response to this violation.
>
Based upon a review of the N"; ARC charter, the inspector concluded such a review
was appropriate in exercising the full extent of the ommittee's responsibilities. ;
The inspector reviewed the BMO process and concluded adequate controls were
established to limit the time a degraded or non conforming condition could exist.
The procedure now requires that a condition adverse to quality be dispositioned
within specified time periods as defined in Administrative Procedure (AP) 0009,
Event Reports. For example, category 1 event reports should be dispositioned
within 45 days, category 2 within 90 days, and category 3 within 120 days. In
addition, the BMO instruction states a periodic review (semi annual) of all open
items to verify that the conditions and assumptions remain valid shall be performed.
The results of this review are then presented to the plant operations review
committee (PORC).
The inspector noted there were 44 BMOs in place. VY intended to resolve each
BMO before startup from the winter 1998 refuel outage. The VY staff had also
performed an assessment to ensure the cumulative effect of outstanding BMOs
would not have a deleterious effect on the operator's response to a transient.
--
Based upon the corrective action, the inspectors concluded this issue was resolved.
.
- -. - .- .. _= -- .. - . . __
_ - _ . . . . - . _ _ --
.
4
20
To improve the oversight of plant operations, it appeared oversight groups increased
the number of self assessments and independent audits performed on individual
program areas. For example, a MOV program audit was completed in March 1997.
The inspector noted such periodic reviews could provide valuable insight into the
effectiveness of plant operations and should f acilitate the evaluation of the
licensee's corrective action process and resolve this weakness. This issue is
closed.
IV. Plant Support
P3 EP Procedures and Docurnentation
a. inspection Scoce (82701)
Regional inspectors reviewed several changes the licensee made to the emergency
plan and implementi!q) procedures. The inspectors reviewed these changes in the
NRC Region I office. They conducted this review to verify that the changes made
to the Emergency Plan and implementing procedures were made in accordance with
Part 50.54(q) of NRC regulations, (i.e., that they did not decrease the ef fectiveness
of the Emergency Plan). The list of Emergency Plan sections and implementing
procedures reviewed is contained in Attachment A of this report.
c. Conclusions
Based on the licensee's determinations that the changes did not decrease the
overall effectiveness of the Emergency Plan, and that the Plan, as changed,
continues to meet the standards of 10 CFR 50.47(b) and the requirements of
Appendix E to Part 50, NRC approval of these changes is not required. The in-
office review of these changes indicated them to have been made in accordance
with 10 CA 50.54(q).
R1 Radiological Protection and Chemistry (RP&C) Controls
t
R 1.1 (Closed) URI 97 06-02: Imolementation of the Radioactive Liould and Gasaput
Etfluer.t Control Proarams
a. insocction Scone (84750 01)
The inspection consisted of: (1) tour of the plant, including the control room; (2)
review of liquid and gaseous effluent release permits; and (3) review of unplanned
and unmonitored release pathways,
b. Observations and Findinas
The inspector toured the control room and selected radioactive liquid and gas
processing f acilities and equipment, including effluent / process radiation monitoring
systems (RMS) and air cleaning systems. All equipment was operable at the time M
the tour. The inspector also noted that the licensee maintained air balances for
--- - _ _
. _- - . _ - - - - - _ _ .
- - . . - - - - - - - - _ - - - - . . _ - _
.
21 !
reactor, turbine, and radwaste buildings to assure conformance to Final Safety
Analysis Report (FSAR) specifications. The inspector noted that the licensee ,
monitored a negative pressure only for the reactor building. (see Section R.2.3 of
'
this inspection report for details.)
During the review of selected radioactive gaseous effluent discharge permits, the
inspector determined that the discharge permits were complete and met the
Technical Specification /Offsite Dose Calculation Manual (TS/ODCM) requirements
,
. for sampling and analyses at the frequencies and lower limits of detection
established in the TS/ODCM. The inspector noted that there had been no
radioactive liquid releases from the Vermont Yankee site for several years while
pursuing effluent ALARA and plant water conservation.
The inspector also noted that there was one unplanned /unmonitored radioactive ;
liquid or gas release since the previous inspection. The licensee found cracks in the i
radwaste building exhausting ventilation duct leading to the plant stack and the
licensee made a 10 CFR 50.72 report (ENS No. 32842) on August 29,1997. The
inspector reviewed the licensee's radiological and environmental assessment results.
The inspector determined that the licensee's actions were acceptable and that there
was no radiologicalimpact to the public safety and the environment. Unresolved
item URI 07 06 021s closed,
c. Conclusion
Based on the above reviews and observations, the inspector determined that the
licensee mainta!ned ar.d implemented effective radioactive liquid and gaseous
<
effluent control programs.
R2 Status of RP&C Facilities and Equipment
R2.1 Calibration of Effluent / Process / Area / Accident Radiation Monitorina Systems (RMS) ,
,
a. Inanection Scope (84750 01)
The inspector reviewed the most recent calibration results for the following
effluent / process / area RMS and associated flow rate monitors to determine the
implementation of the TS requirements and FSAR commitments-
- Steam Jet Air Ejector Offgas Monitors
- Main Stack Noble Gas Monitors (Normal and High Ranges)
- Main Stack Flow Rate Monitor
- Augmented Offgas (AOG) Building Noble Gas Monitors
- AOG Flow Rate Monitor
- Reactor Building Monitor
- Spent Fuel Pool Floor Monitor
- Liquid Radwaste Discharge Monitor
- Service Water Discharge Monitor
l
-- -,
l
, , . -- - , - . - . .--. - . - - _
-. -.-
.
22
b. Observations and Findinas
The l&C, Chemistry, and Radiation Protection departments had the responsibility to
perform electronic and radiological calibrations for the above radiation monitors. All
reviewed calibration results were within the licensee's acceptance criteria. The
Chemistry Department performed trending analyses for the effluent RMS, which
was considered a licensee strength.
During the review of the above RMS calibration documentation, the inspector
independently calculated and compared several calibration results, including linearity
tests and conversion f actors. The inspector determined that the licensee's results
were comparable to the independent calculations,
c. Conclusions
Based on the above reviews, the inspector determined that the licensee maintained
and implemented a good calibration program and good trending analyses for effluent
radiation monitoring systems.
R2.2 Air Cleanina Systems
a. insoection Scone (84750 01)
The inspector reviewed the licensee's most recent surveillance test results (in place
HEPA and charcoalleak tests, air capacity tests, pressure drop tests, and laboratory
tests for the iodine collection efficiencies) for the standby gas treatment system
(SBGT) required by TS. In-place HEPA and charcoal surveillance tests for the
Advanced Off-Gas (AOG) system and the radwaste building air cleaning system
were also reviewed.
b. Eb3myations and Findinas
All surveillance results were either within the TS acceptance criteria or the
administrative acceptance criteria.
Recently, the Office of Nuclear Reactor Regulation (NRR) identified that there was a
potential conflict regarding the charcoal testing methodology for the iodine
collection efficiency performed by the licensee / contractor laboratory. Normally, the
licensee's TS specifies Regulatory Position C,6.a of RG 1.52, Revision 2,
March 1978, as the requirement for the laboratory testing of the charcoal. RG 1.52
references ANSI N5091976," Nuclear Power Plant Air Cleaning Units and
Components." ANSI N5091970 specifies that testing is to be performed in
accordance with paragraph 4.5.3 of RDT M 161T," Gas Phase Adsorbents for
Trapping Radioactive lodine and lodine Components." The essential testing criteria
are: (1) 70% or 95% relative humidity (RH): (2) 5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> pre equilibration time, with 4
air at 25' C and plant specific RH; (3) 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> challenge, with gas at 80* C and
plant-specific RH; and (4) 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> elution time, with air at 25' C and plant specific
RH. The latest acceptable methodology for the laboratory testing of the charcoalis
.
23
ASTM D 38031989,which requires licensees to maintain 30* C during all testing
phases.
The inspector noted that the VY staff performed two separate surveillance tests for
the SBGT: (1) ANSI 5091980and ASTM D 3803 Method C-1979 (13CP C and
95% RH); and (2) ASTM D 38031989(30* C and 95% RH). The licensee
recognized that the:e was a potential problem for the iodine collection efficiency
test methodology. The licensee, therefore, added the ASTM D 38031989
methodology in May 1995 and trended iodine collection efficiencies obtained from
both methodologies. The inspector concluded that the licensee utilized an excellent
surveillance test methodology for the SBGT system, and met all regulatory
requirements.
c. Conclusions
Based on the above reviews, the inspector determined that the licensee maintained
the plant air cleaning systems in accordance with estabilshed design specification.
The licensee performed excellent iodine collection efficiency test methodologies for
the SBGT system.
R2.3 Plant Air Balance
a. Insoection Scope (84570 01)
The inspection consirted of: (1) review of the main stack exhaust flow rate; (2)
review of exhaust flo'n rate from various buildings; and (3) assessment of the plant
air balance.
b. Observations and Findinas
Procedure OP 2011, " Gaseous Radwaste" listed maximum exhausting f an
capacities of various buildings (e.g., reactor, turbine, and radwaste buildings). The
exhaust air from these buildings was released to the environment through the main
stack. The maximum exhaust air flow rate from these building was 181,900 cfm
while the main stack flow rate was about 150,000 cfm. The difference, of about
32,000 cim, potentially demonstrated that station ventilation systems were not
properly balanced. The original plant air balance data, measured in 1071(150,000
cfm), was no longer valid since the turbine building exhaust was connected to the
main stack in 1993. The licensee estimated that the main stack air flow exhaust
would be increased, from 150.000 cfm to 200,000 cim, due to addition of the
turbine building ventilation (sce inspection report Nos. 50 271/93 25and 50-
271/94 27 for details). Verification and corrections to exhaust fan capacities
(actual and maximum) listed in procedure OP 2611," Gaseous Radwaste," will be
reviewed during a subsequent inspection (IFl 97 08-02).
The licensee maintained a negative pressure in the reactor, turbine, and radwaste
buildings. The licensee monitored differential pressure daily in the control room for
the reactor building. However, the licensee maintained a negative pressure for the
24
turbine and radwaste building through damper position indications since there were
no installed delta-P gauges. The licensee planned to install delta P gauges for
turbine and radwaste buildings to assure the negative pressure was maintained.
This action will also be reviewed by the inspector during a subsequent inspection
(IFl 97 08-02),
c. Conclusions
Dased on the above reviews, the inspector made the following conclusions:
- the actual and maximum f an capacities listed in procedure OP 2611 should
be verified to avoid a potential inaccurate projected dose calculation to the
public;
- the licensee maintained the negative pressure for the reactor building verified
delta P daily using a installed gauge;
e the licensee maintained air balanco for turbine and radwaste buildings
through administrative means (e.g., damper position); and,
e the licensee planned to install delta P gauges for the turbine and radwaste
buildings.
R3 RP&C Procedures and Documentation
a. Inspection Scone (84570-01)
The inspection consists of a review of: (1) selected chemistry procedures to
conduct the effluent control programs; (2) secor:d hclf of the 1995 Semiannual
Report and the 1996 Annual Radioactive Effluent Release Reports;(3) the ODCM;
and (4) implementation of 40 CFR 190 requirements,
b. Observations and Findinas
The inspector noted that reviewed ef fluent control procedures were detailed, easy
to follow, and ODCM requirements were incorporated into the appropriate
procedures. The licensee had good procedures to satisfy the TS/ODCM
requirements for routine end emergency operations.
The inspector reviewed the 1995 Semiannual and the 1996 Annual Radioactive
Effluent Release Reports. These reports provided data indicating total released
radioactivity for liquid and gar.cous effluents. The assessment of the projected
maximum individual doses resulting from routine radioactive airborne and liquid
effluents were listed as required. Projected doses to the public were well below the
TS limits. The inspector determined that the e were no anomalous measurements,
omissions, or adverse trends in the reports.
.__ . - -
25
The ODCM provided descriptions of the sampling and analysis programs, which
were established for quantifying radioactive liquid and gaseous effluent
concentrations, and for calculating projected doses to the public. All necessary
parameters, such as effluent radiation monitor setpoint calculation methodologies,
and site specific dilution f actors, were listed in the ODCM. The licensee adopted
other necessary parameters (dose f actors) from Regulatory Guide 1.109.
Section 3/4.8.M of the TS requires that the licdasee shall comply with the 40 CFR
190 requirements,25 mrem / year to the total body to a member of the public with
occupancy rate at the monitoring location. The inspector reviewed the 1996
Annual Radiological Environmental Surveillance Report and Effluent Report including
projected dose calculation results to the public. The licensee has 14
thermoluminescent dosimeters (TLDs) around the site boundary and two control
TLDs ste.tions (about 15 km from the plant) to comply with 40 CFR 190
requirements. The mean measurement value at the site boundary (including
background) and control stations were 69.14 and 56.1 mrem / year, respectively,
during 1996. The difference value, which is about 13 mrem / year, would be
contributed from the plant operation. The licensee reported the maximum total
body dose from f acility direct radiation was about 14 mrem during 1996 at the west
site boundary. However, there were no residents present at that location. Total
body dose due to radioactive liquid and gaseous effluent releases was 0.05 mrem
during 1996. The total dose would be 13.05 mrom during 1996 and the licensee
met the TS requirements,
c. Conclusions
Based on the above reviews, the inspector made the following conclusions:
(1) effluerit control procedures were sufficiently detailed to f acilitato
performance of all necessary steps for routine and emergency operations;
(2) the licensee offectively implemented the TS/ODCM requirements for
reporting effluent releases and projected doses to the public; and,
(3) the licensee's ODCM contained sufficient specification, information, and
instructic,n to acceptably implernent and maintain the radioactive liquid and
gaseous effluent control programs.
R6 RP&C Organization and Administration
The inspector reviewed the organization and administration of the radioactive liquid
and gaseous effluent control programs and discussed with the licensee changes
made since the last inspection. The inspector determined that there were no
changes to the radioactive effluent control programs. The Chemistry Department
has primary responsibility for conducting the radioactive liquid and gaseous effluent
control programs. The System Engineering, Operations, Radiation Protection, and
instrumentation and Cor.trols (l&C) Departments also have responsibilities to
-support effluent control programs, such as air cleaning systems, redwaste
.
26
discharges, and radiation monitoring system calibrations (radiological and electronic
calibrations).
R7 Quality Assurance (QA)in RP&C Activities
a. laipection Scone (84750-01)
The inspection consisted of a review of: (1) the 1996 and 1997 QA audit reports;
and (2) implementation of the measurement laborato y quality control program for
radioactive liquid and gaseous effluent samples,
b. Observations and Findinas
The inspector reviewed the 1996 and 1997 QA Audit Reports (Report Nos.
VY 96 02 and VY 97 02). These audits were conducted by the QA Department
staff and covered the radioactive liquid and gaseous effluent control programs,
including the implementation of the ODCM. The inspector noted that the audits
were conducted by members of QA Department with assistance from other
technical personnel. The 1996 audit team identified one finding. The 1997 audit
team identified no findings. The inspector determined that the 1996 finding was not
safety significant, but was intended for the enhancement of the effluent control
programs Prompt corrective action was performed by the Chemistry Department
staf f.
The inspector reviewed the implementation of QC for the chemistry laboratory, l
'
including control charts, inter laboratory and intra laboratory comparisons. The
inspector considered the chemistry laboratory QC program to be very good,
c. Q.gnelusion
Based on the above review and interviews, the inspector determined that the
technical depths of the audits was good and met TS requirements. The chemistry
laboratory QC program was very good. !
R8 Miscellaneous issues
R8.1 Trair.ina
l
The inspector reviewed the training courses for the licensed operators, in the areas ,
of radioactive liquid and gaseous effluent controls and discussed this program with I
a training instructor Required training courses for operators appeared to be i
'
appropriate, and included subjects such as: solid and liquid radwaste processes;
ventilation; ODCM; HVAC; AOG; area / process / effluent RMS: and service water. To
complete these courses required about 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> in class and the use of simulator.
The licensee also had the " Response Training" and an annual re-qualification
training. The response training involved specific effluent events, such as Event
Reports or LERs.
-- - _ . . . _ _ . . - _ _ . _ - - - _ . - . . . - - _ - .- -_ - - . . - . _ - . . . . . - -
!
27
,
The inspector discussed with the plant staff the involvement of the Chemistry staff
in response training. Currently, there is no Chemistry staff involvement for the ;
response training and the annual re qualification training. The licensee stated that t
the/ would evaluate Chemistry staff involvement in this training with respect to the
effluent ALARA program. '
R8.2 (Closed) URI 96 03-05, LER 96 03 Suo.1, and NCV 97 08 07: Removal of Reactor l
Vessel Shield Blocks et Power .
a. Insoection Scoce (92904,92700)
i
This issue was previously examined in inspection report 96 03, section 4.1, and ;
inspection report 96-09, section R8.2 and remained open pending completion of the l
licensee's root cause evaluation. LER No. 96-03, Supplement 1, " Removed reactor ;
shield blocks during power operations to facilitate outage scheduling due to i
personnel error," dated June 12,1997, documented the VY staff's formal root
cause evaluation results and associated corrective actions. The inspector
conducted a review of LER 96 03, Supplement 1, and verified the implementation of l
the stated corrective actions. .
c ;servations and Findinas
b.
The licensee's formal root cause of this event was personnel error, in that there was l
a lack of awareness by plant personnelin 1990 and 1992 of the consequences of l
removing all three sets of reactor cavity shield blocks while at power. A ]
contributing cause identified by the VY staff was the lack of formal procedural
guidance governing the removal of shield blocks. The inspector confirmed the l
adequacy of the licensee's corrective actions which included a revision to the
refueling preparation procedure Operating Procedure (OP) 1200," Preparation of the l
Reactor Vessel for Refueling," revision 18, dated 9/5/96) section 1.0, "Drywe!! I
I
Shield Block and Drywell Head Removal," which added the requirement that "the
first layer of shield blocks can be removed at any power level prior to shutdown, the _,
second and third layers of shield blocks cannot be removed until the reactor vessel I
is < 212 degrees F and vented." In addition, the licensee revised Plant Operations
Review Committee (PORC) member training, engineering support staff continuing '
training, and the Safety Evaluation Training lesson plans to reflect the lessons ,
'
learned from this event.
The inspector considered this refueling preparation event to have been reflective of
past poor VY staff performance. This non repetitive, licensee identified and
corrected violation was treated as a non-cited violation (NCV 97 08 07), consistent
with Section Vll.B.1 of the NRC Enforcement Poliev.
c. Conclusions
Licensee actions to identify and implement corrective actions to prevent drywell
shield blocks removal during power operations were appropriate. This licensee
identified event involving the violation of regulatory requirements was not cited.
~. -- - ,, - - . , --. .----- - . - - - -.- --- -
-- - . - - _ . - . . . - . -- - - - - - - -- .. .
.
I
!
28
LER 96-03 and 96 03 Supplement 1 were closed and unresolved item URI 96 03 05
was closed.
V. Management Meetings
X1 Exit Meeting Summary
t
The inspector presented the inspection results of the radiological environmental
monitoring program to members of the licensee management at the conclusion of
the inspection on September 26,1997. The results of the MOV review were
presented to station management at the conclusion of the on site inspection on
! October 17,1997. The licensee acknowledged the findings presented. l
The resident inspectorc met with licensee representatives periodically throughout ;
'
4
the inspection and following the conclusion of the inspection on November 20,
1997. At that time, the purpose and scope of the inspection were reviewed, and
the preliminary findings were presented. The licensee acknowledged the preliminary
inspection findings.
X2 Review of Updated Final Safety Analysis Report (UFSAR)
A recent discovery of a licensee operating their f acility in a manner contrary to the
Updated Final Safety Analysis Report (UFSAR) description highlighted the need for a i
special focused review that compares plant practices, procedures and/or parameters
'
to the UFSAR description. While performing the inspections discussed in this
- report, the insp'.9 tors reviewed the applicable portions of the UFSAR that related to
the areas inspected. The inspectors verified that the UFSAR wording was
consistent with the observed practices and procedures and/or parameters.
However, the inspectors observed that the FSAR wording was questionable with
respect to the observed plant practices, procedures, and parameters involving
building air balance and flow (See Section R.2.3 of this inspection report). The
inspectors reviewed Sections 4.4, and 4.8 of the VY UFSAR to assess if VY had
,
incorporated correct UFSAR information into plant MOV procedures regarding *
reactor coolant and reactor water cleanup systems (RWCU), respectively. This
information was found to be correct and up to-date.
i
- - . , _ _ _ _ _ _ _ _ _ . . , _ _
.- . . .. .
.
29
INSPECTION PROCEDURES USED
61726 Surveillance Observations
71707 Plant Operations
92901 Follow Up Plant Operations
92903 Follow Up Engineering
84750 01 Radioactive Waste Treatment, and Effluent and Environmental Monitoring
82701 Operational Status of Emergency Preparedness Program
02700 LER review
62707 Maintenance Observations
92903 Engineering Follow up
92904 Plant Support Follow up
37551 Onsite Engineering Review
Tl 2515/109 Inspection requirements for Generic Letter 8910," Safety Related Motor-
Operated Valve Testing and Surveillance "
.
.
.
30
ITEMS OPENED, CLOSED, AND DISCUSSED
OPEN
IFl 97 08-01 Inspector follow up of the licensee's resolution of the EDG piping welds
issue.
IFl 97-08-02 Verification of exhausting actual and maximum f an capacities listed in
l'.v. vdure OP 2611, " Gaseous Redwaste."
CLOSED
LER 96-03, Sup.1
'
LER 9611
LER 9614, Sup.1
LER 96 21
LER 96 23
LER 9711
- LER 9712
LER 9714
! ER 9717
URi 96-03 05 Removal of reactor vessel shield blocks at power.
URI 97-06 02 Cracks in radwaste building vent ducting.
LER 96 29
LER 96 24 i
IFl 96 09 01 Appendix J testing deficiencies
LER 96 27
URI 9316-01 Pressure lo:: king and thermal binding
VIO 96-05 03 Failure to update and control of Program Manual
EA95 070 VIO 01013: Failure to correct a condition adverse to quality
NCV 97 08-03 LER 96 23
NCV 97 08-04 LER 96 29
NCV 97 08 05 LER 96 27
NCV 97-08-06 LER 9611
DISCUSSED
LER 96-03
IFl 961101 EDG tornado protection
VIO 97 06 03 Ineffective corrective action, containment inerting event.
IFl 97-04 04 RHR service water flow instrument accuracy
'
IFl 97-06-01 - Licensee's LSFT review follow up
URI 97 03-02 Cabie separation issues
<
- . _ _ , _ .. _ _ , - _ . -
-
,
)
i
!
31 I
, t
PARTIAL LIST OF PERSONS CONTACTED -
l
.
G. Meret, Plant Manager i
F. Helin, Tech. Services Superintendent ;
- M. Balduzzi, Superintendent of Operations l
E. Lindamood, Director of Erigineering l
K. Bronson, Operations Manager !
M. Watson, Maintenance Superintendent !
'
G. Morgan, Security Msinager
J. Chamberlin, Sistem Engineer, instrument and Controls ;
M. Desiletes, Radiation Protection Manager !
,
- R. Gardes, Chemistry Manager
F. Helin, Technical Services Superintendent l
S. Jefferson, Scheduling Manager, Operations !
i S. McAvoy, Chemistry Supervisor !
D. Voland, Radiological Environmental Supervisor l
C. Hansen, MOV Engineer . ;
J. Lynch, Fluids Design Engineering i
,
C. Nichols, Manager, E&C l
?
!
}
I
i
.
9
I
f
s
f
>
-_ _ - - . - - - . - .. .- _, - .
. . - . - ~ - - . . . - _ - - . . . - . -
> :
l
'
32
LIST OF ACRONYMS USED
{
!
ALARA As Low As is Reasonably Achievable !
ARMS Area Radiation Monitoring System
GMO Basis for Meintaining Operation
CFR Code of Federal Regulation
CR control room .
J
CS core spray '
EDG emergancy diesel generator
ER Even; Report l
FSAR Final Safety Analysis Report l
01. Gener;c Letter i
W! IPA High Efficiency Particulate
,4PCI high pressur6 coolant injection j
HVAG Heating, Ventilation, and Air Conditioning ,
ifi inspecto' fol!oc item ,
IN %rmeifn Notice i
'
LCO L%thg fondition for Operation
LER Licensec Event Report
LPCI low pressure coolent injection ,
MCC -motor control center :
NRC Nuclear Regulatory Commission l
NNS Non nuclear safety !
ODCM Offsite Dose Calculation Manual
PO3C Plant Operations Review Committee
OA - Quality Assurance ;
OC Ouality Cor.aol ;
RMS Radiation Monitoring System
RP&C Radiation Protection
SFP Spent Fuel Pool
TS Technice! Specifications ,
UFSAR Updated Final Safety Analysis Report
URI - unresolved item
i
.. _.
I
!
<
.. . _ -.-.____-_. _ .._ _ ___ _ _ _ _ . . . . . . _ . . , . . _ . , _ _ , . . . . . , . _ _ _ _ , _ . _ . , _
. . _ - - . _ . _ . _ . . _ . _ _ _ _ _ . _ _ . _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ . _ _ . - - _ _ . _ - _ _ _ - _
.
.
33
ATTACHMENT A
List of Emergerv;y Plan and implementing Procedures Reviewed
DOCUMENT TITLE REVISION
NO.
Emergency Plan Section 6.0 Emergency Facilities and Equipment 20
Emergency Plan Section 8.0 Organizatic,n 20
Emergency Plan Section 10.0 Radiological Assessment and Protectivs 20
j Measures
Emergency Plan Section 12.0 Maintaining Emergency Preparedness 19
Emergency Plan Appendix E Letters of Agreement 22
l
OP 3504 Emergency Communication DI i
97 133
OP 3509 Environmental Sample Collection During an 15 l
Emergency
OP 3535 Post Accident Sampling and Analysis of 2 i
l
i
i
l
a
l
!
I
. . . . . . _ __ _ . . . . ._.......u,_ -
. . _ . . . _ . . _ _ _ . _ . , _ _ _ . . . . , , . _ _ , _ . - . . _ _ . . . , . , _ , . _ _ , .