ML20129B376
| ML20129B376 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 09/13/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20129B351 | List: |
| References | |
| 50-289-96-05, 50-289-96-5, NUDOCS 9609230020 | |
| Download: ML20129B376 (77) | |
See also: IR 05000289/1996005
Text
'a
l
.
U. S. NUCLEAR REGULATORY COMMISSION j
'
REGION I
i
l
Docket No. 50 289
License No. DPR 50 {
Report No. 96-05 l
l
l
Licensee: GPU Nuclear Corporation
1
Facility: .Three Mile Island Station, Unit 1 !
!
Location: P.O. Box 480 .
'
Middletown, PA 17057
- Dates: June 17,1996 - August 3,1996
Inspectors: Michele G. Evans, Senior Resident inspector
Samuel L..Hansell, Resident inspector l
Douglas A. Dempsey, Reactor Engineer l
Mark Holbrook, INEL, NRC Contractor, I
Joseph Colaccino, Mechanical Engineer, NRR
Dan Billings, Resident inspector
Approved by: Peter W. Eselgroth, Chief
Reactor Projecta, Section No. 7
.
9609230020 960913
O ADOCK 05000289
, _
pop
..
.._ _ _ - _ _ _____ ._ __m_______ -.___m________W
.
.
EXECUTIVE SUMMARY
Three Mile Island Nuclear Power Station
Report No. 50-289/96-05
,
This integrated inspection included aspects of licensee operations, engineering,
maintenance, and plant support. The report covers a 7 week period of resident inspection;
in addition, it includes the results of announced inspections in the areas of Inservice
Surveillance Testing and Motor Opersted Valve programs for unit 1.
Plant Operations
The operator's excellent response to the fuse clip failure for the station blackout diesel was
an example of positive performance related to the application of good self checking
techniques (Section M1.1).
Operations did not identify a degradation of Auxiliary Building and Fuel Handling Building '
Ventilation (ABFHV) system flow. In addition, the shift senior reactor operators did not
document entry into the applicable TS Limiting Condition for Operation. The untimely
ABFHV system operability determination was similar to an AB ventilation issue that was
documented in NRC Inspection Report No. 50-289/96-02 (Section 01.2).
1
Operations performed and implemented multiple detailed on-line safety risk assessments for
the decay heat valve modifications. The applicable Technical Specification limiting
conditions for operation were entered and exited correctly for the safety related equipment
outage times (Section M1.1).
Maintenance
The maintenance and surveillance test activities observed during this inspection were
performed satisf actorily and demonstrated that the associated systems could perform their
design safety functions, lit particular, the station blackout (SBO) diesel air start solenoid
valve replacement work activities were performed satisfactority and should improve the i
SBO diesel reliability (Section M1.1).
Enaineerina
The inspectors were unable to close the TMI MOV program during this inspection.
Adequate justification had not been provided to demonstrate the design-basis capability of
a large number of your nontestable-valve groups (Section E1.1).
The design-basis capabiFty of certain motor-operated valves (MOVs) was not demonstrated
adequately for Generic Letter 89-10 program closure. For certain low-margin MOVs, the
licensee did not adequately justify the valve factors (derived for nontestable valves) as the
basis for demonstrating design-basis capability. Specifically, the approach used to apply
. certain industry and Electric Power Research Institute (EPRI) valve test data was informal,
ii
- - . _ _ _ _ - - _ - - _ _ _ _ _ _ . --
. . _ . .- . . . _ _ _ _ _ _ . _ - . _ _ _ .
- *
.
P
not well controlled with respect to design input to safety-related calculations, and not
technically sound (Sections E1.2 and E1.3).
.The methods used to include load-sensitive behavior for rising stem MOVs were
inadequate. The use of the MOVATS displacement measuring device (DMT) for
determining load-sensitive behavior was inappropriate (Section E2.1).
The engineering modifications for decay heat valves DH-V-4A/B were very well planned, ,
controlled, and implemented to address potential valve pressure locking issues (Section
E 2.1 ).
The TMI engineering and operations response to the Arkansas Nuclear One OTSG dryout
and stuck open safety relief valve event was comprehensive, thorough, and displayed a-
strong initiative to address generic safety issues at other plants in order to reduce the
potential for a similar impact at TMI (Section E4.1).
Plant SUDDOrt
The radiography procedure changes and boundary controls significantly improved the -
radiological control personnels' ability to alert all plant workers about the conduct of
radiography and ensure the exclusion boundaries were controlled properly. Procedure
6610-ADM-411.07 was revised satisfactorily to address the prior radiography problems t
(Section R8.1).
The Radiological Controls / Occupational Safety department customized self-checking
training was a positive initiative to reinforce managements' expectations for each
department (Section RS.1). i
I
!
I
iii
l
I
l
l
.e
TABLE OF CONTENTS
EXEC UTIVE SU M M A RY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii
TABLE O F C O NTENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv
1. Operations .................................................... 1
01 Conduct of Operations (71707) ............................. 1
01.2 Auxiliary Building Ventilation Filter Operability Determination
and Documentation . . . . . . . ..... .................. 1
11. Maintenance .................. ... ......................... 3
M1 - Conduct of Maintenance '82705,62707,61726) ................ 3
- M 1.1 General Commente .................................. 3
111. Engineering . . . . . . . . . . . . . . . . . . . ............................... 4
E1- Conduct of Engineering (37551, 92903) . . . . . . . . . . . . . . . . . . . . . . . 4
E1.1 Generic Letter 89-10 Motor-Operated Valve Program Review . . . . 4
E1.2 Operator Sizing and Switch Setting Assumptions . . . . . . . . . . . . 5
E1.3 Design-Basis Capability ............................. 13
E1.4 G L 8 9-10 Program Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
E4.1 TMl Response to the Arkansas One Unit 1 OTSG Blowdown .. 16
E8 Miscellaneous Engineering Issues (92903) . . . . . . . . . . . . . . . . . . . . . 18
E8.1 (Closed) Unresolved item 50-289/94-12-01: .........,.... 18
IV. Plant Support ................................................ 18
R5 Staff Training and Qualification for RP&C, Security and
Chemistry /Radwaste (71750) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18
R5.1 Personnel Self-Checking Training ...................... 18
.R8. Miscellaneous Radiological Control Program items ............... 19
R8.1 Previously identified items (92904) . . . . . . . . . . . . . . . . . . . . . 19
V. Management Meetings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
X1 Exit Meeting Summ ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
X2 Meeting With GPU Nuclear Corporation Regarding Motor Operated
Valve (MOV) Program Concerns . , , . . . . . . . . . . . . . . . . . . . . . . . . . 21
X3 GPU Nuclear Engineering Integration Meeting . . . . . . . . . . . . . . . . . . . 21
PARTIAL LIST OF PERSONS CONTACTED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22
ITEMS OPENED, CLOSED, AND DISCUSSED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23
- LI ST OF ACRONYMS U SED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24
iv
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _
.
.
Report Details
Summary of Plant Status
At the beginning of the period, Unit 1 was operating at 100% power. On June 29,1996,
the licensee reduced reactor power to 50% to repair main condenser tube leaks. The
repair was completed satisfactorily and the unit returned to 100% power on June 30th.
At the end of the period, the unit was operating at 100% power.
I. Operations
01 Conduct of Operations (71707)'
01.1 General Comments
Using inspection Procedure 71707, " Plant Operations," the inspectors conducted frequent
reviews of ongoing plant operations. In general, the conduct of operations was
professional and safety-conscious; specific events and noteworthy observations are
detailed in the sections below, in particular, the inspectors noted a repetitive concern
related to the recognition and documentation of a Technical Specification (TS) operability
determination that was documented in NRC Inspection Report No. 50-289/96-02.
01.2 Auxiliary Buildina Ventilation Filter Operebility Determination and Documentation
a Insoection Scoce (71707)
The Auxiliary Building and Fuel Handling Building Ventilation (ABFHBV) system is
designed to maintain suitable and safe ambient conditions for operating equipment
and personnel during normal plant operation. The system is also designed to
minimize the release of radionuclides to the environment under postulated accident
-
conditions by maintaining a negative building pressure. The required combined
exhaust flow of the ABFHBV systems is 100,580 cubic feet per minute (cfm) to
130,691 cfm per TS.
On July 9,1996, the inspector's noted in the Control Room log that the system
flow had been logged by the night shift Reactor Operator at 100,560 cfm, which
was below the TS minimum limit. The inspectors reviewed the ABFHBV system
TSs, updated final safety analysis report (UFSAR), Control Room logs, and ABFHBV
chart recorders. The inspectors also interviewed the system engineer and
appropriate operations personnel to determine the actions taken in response to the
low system flow.
l
l
1
1
' 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.
I
.__ ----_-_______------___--____-__-___-_-__-___A
. . - ._ - . _. -
e
.
2
b. Observations and Findinas
n
'
The ABFHBV fans were shifted following the recognition of the low flow with the
'A' and 'C' fans secured and the'B' and 'D' fans placed in service. Subsequent
readings were above the TS required minimum flow. A surveillance deficiency
a report (SDR) was written to document the ABFHBV low flow condition and the
- system was considered operable by operations. ABFHV system flow readings were
recorded from the digital readouts in the control room at 9:00 a.m. and 9:00 p.m.
each day to verify operability.
The system engineer presented a comprehensive plan to troubleshoot the ventilation
system low flow condition. From July 9th through July 12th, troubleshooting
- continued with the ABFHBV system being shutdown for fan inspection, filter
inspection, and filter change. On July 12th, at 4:30 p.m., the exhaust fans were
restarted with no change in recorded flow. Af ter operation management's review of ,
the ABFHBV degraded flow, the SS entered the correct TS LCO due to the digital
,
chart readings at the low end of the oscillation being less than the TS minimum limit
of 100,580 cfm and the frequency of the oscillations indicating a degraded
> condition. The low end of the total flow oscillations were recorded as 99,400 cfm.
T
,
On July 14th, additional inspections of the dampers and ductwork revealed a
- broken manual damper. A temporary modification was written and the broken
approximately 110,000 cfm. A Plant Review Group meeting was held on July 15th
i to review the ABFHBV system condition and concurred with the operation
'
management's derJsica to call the ABFHBV operable. As of the end of this '
4 inspection period the licensee continued to troubleshoot the system to identify
additional causes of the degraded system flow.
.
'
The inspectors obtained copies of the control room ABFHBV charts showing a low
flow condition on July 5th approximately 10,000 cfm below TS LCO minimum with
no explanation documented on the chart. Operations personnel were questioned
l
and the Operation's manager explained that the low flow was due to a 480 Volt
electrical bus outage which caused a loss of control building fans. The same
condition occurred again on July 12th when the same bus was de-energized to
complete the breaker work. The low flow conditions existed for approximately one
hour each. Operations personnel did not recognize the abnormal ventilation trends
- that resulted in the total ABFHBV flow that was less than the TS limit. In addition,
the shift SROs did not document the applicable TS LCO when flow dropped below
then recovered above the minimum value.
I Operations management backdated the shift log on July 31st to include the correct l
dates and times for the three occasions when ABFHBV decreased below the
minimum TS LCO value. The inspectors also obtained flow values from the daily
logs showing a decreasing trend of ABFHBV flow over the previous two week
period from 105,000 cfm to the low value of 100,560 cfm on July 9th. This flow
'
was below the value of 118,000 cfm that was documented in October 1995 at the
end of the last refueling outage.
4
. - . .. _ - - - - - - - _ - - - - . _ _ ..
O
I
- .
3
'
c. Conclusions
'
Operations did not identify a degradation of Auxiliary Building and Fuel Handling
Building Ventilation system flow, in addition, the shift senior reactor operators did
not document entry into the applicable TS Limiting Condition for Operation. The
untimely ABFHV system operability determination is a repeat problem that was
i similar to an AB ventilation issue documented in NRC Inspection Report No. 50-
289/96-02. l
11. Maintenance i
M1 Conduct of Maintenance (62703,62707,61726)
'
I
M 1.1 General Comments
' Insoection Scope
a.
< The inspectors observed all or portions of the following maintenance and
surveillance work activities:
i
,
o Job Order No. 106833, "'B' 125 VDC Ground Troubleshooting."
J
- Job Order No. 128218, " Station Blackout Diesel Air Start Valve
i Replacement."
- * Job Order No. 121608, "'C' River Water Travelling Screen Troubleshooting
- Shear Pin Problem."
e
, * Job Order No. 117416, "DH-V-4A/B Pressure Locking Modification."
2 * Job Order No. 122221, "BS-V-2B Limitorque Preventative Maintenance."
i
- Operations Procedure 1107-9, " Station Blackout Diesel Surveillance Test."
- Surveillance Procedure 1303-5.5, " Control Room Emergency Filter System
Operability Test."
- Surveillance Procedure 1303-11.13, " Control Room Emergency Filter DOP
and Halide Test."
b. Observations and Findinas
Two positive observations were noted related to the station blackout diesel
maintenance and associated post maintenance test run. First, the replacement of
both air start solenoid valves with a new and improved design should result in fewer i
diesel slow starts and improved reliability. Secondly, control room operators (CROs)
- demonstrated an excellent questioning attitude during the performance of
_ _ _ _ _ _ __
.. .. . - - . - . - . - - _ =. _
'.
.
4
,
4
Operations Procedure 1107-9, " Station Blackout Diesel Surveillance Test." The
operator response to a fuse clip f ailure for the station blackout (SBO) diesel was a
good example of positive performance related to the application of proper self
checking techniques. The control room indications for the fuse clip failure were
similar to a previous problem related to the 'A' emergency diesel. The improved-
procedure guidance that was incorporated since the previous problem aided the
CROs in the early r'ecognition of the SBO diesel trouble.
j Positive observations were also noted related to the planned modification work on
the decay heat injection valves DH-V-4A/B. The modification was performed to
address motor operated valve pressure locking concerns. Multiple on-line
maintenance risk assessments were completed for the Technical Specification (TS)
components that were removed from service during the modification work. The
4 applicable TSs were entered and exited correctly during the modification outage
time. The work activities were well coordinated and executed by maintenance,
3
4 engineering, operation, and radiological control personnel.
c. Conclusions
The maintenance and surveillance test activities observed during this inspection
,
were performed satisfactorily and demonstrated that the associated systems could
i perform their design safety functions. The operate' response to the fuse clip failure
for the station blackout diesel was excellent and 6n example of positive
~ l
performance related to the application of good self checking techniques. The
engineering modification for decay heat vsives DH-V-4A/B was very well planned,
controlled, and implemented to address potential valvo pressure locking issues.
.
Plant operations performed and implemented multiple detailed on-line safety risk
assessments for the decay heat valve modifications. The applicable Technical
Specification limiting conditions for operation were entered and exited correctly for
the safety related equipment outage times.
Ill. Enaineerina
E1 Conduct of Engineering (37551,92903)
I
E1d Generic Letter 89-10 Motor-Operated Valve Proaram Review
Following this inspection, a management meeting was held between the NRC and the
4
licensee on July 22,1996, in the Region I office, to discuss the MOV inspection findings.
The slides used by the licensee during this presentation are provided as an attachment to
this inspection report.
'
a. Inspection Scope (Tl 2515/109)
A sample of valves was selected for inspection that included examples of all
methods utilized in the TMI's GL 89-10 program to demonstrate design-basis
capability. The methods for demonstrating MOV design-basis capability included
verification by: (1) valve-specific dynamic test at, or near, design-basis conditions,
.. _ . . . - . . - - - - -
.
.
5
.
(2) valve-specific test, linearly extrapolated to design-basis conditions, and (3)
industry information provided oy the Electric Power.Research Institute's (EPRI)
performance prediction program (PPP) and other nuclear facilities. The inspectors
'
reviewed special test packages and engineering evaluations for the following
selected MOVs:
DH-V-3 Decay heat system drop line isolation valve
DH-V-7A Decay heat system pump discharge to makeup pump suction .
isolation valve
FW-V 92A Startup feedwater block valve
FW-V-928 Startup feedwater block valve
.
MU-V-36 Makeup pump minimum recirculation flow line isolation valve
MU-V-37 Makeup pump minimum recirculation flow line isolation valve
RC-V-2 PORV block valve
RR-V-5 Bypass valve for RR-V-6 pressure control valve
The GL 89-10 program scope consisted of 77 MOVs. A total of 36 MOVs were '
dynamically tested.
EL2 Operator Sizina and Switch Settina Assumotions
a. Insoection Scope
The inspectors reviewed valve packages that established the thrust requirements for
MOVs. The purpose of this review was to assess the licensee's justifications for
assumptions used in MOV thrust calculations that form the basis for determining the
design-basis requirements.
b. Observations and Findinas
General Methodoloav
The thrust calculations typically utilized the standard industry equations. Mean seat
diameter was used to calculate valve seat area. Valve factors were based on in-
plant test results or from other industry sources, as specified by the licensee's
grouping methodology. A stem friction coefficient of 0.20 was used for
determination of actuator output thrust capability. A 2.9% bias margin and a
10.3% random margin were used to address load-sensitive behavior (also known as
" rate of loading") for those rising stem MOVs that were not dynamically tested. A
'
bias margin of 5% was included to account for potential future valve degradations.
i
,
___. ,
e
.
6
,
Minimum thrust requirements for setting of actuator torque switches were adjusted
to account for diagnostic equipment inaccuracy and torque switch repeatability.
Valve Factors - Valve Groupina
The MOVs were divided into valve groups based on valve manufacturer, valve type,
and ANSI pressure class rating. In-plant test data was used, when available, for
- justification of valve f actors for nondynamically-tested MOVs. If in-plant data was
'
not available, then data from other utilities and the EPRI PPP were used to establish
valve factors. However, some of the in-plant dynamic test data were not
considered adequate to justify applied valve factors. This placed greater burden on
obtaining adequate industry information to justify program assumptions.
The following are examples where grouping methods did not meet the intent of
! GL 89-10, Supplement 6, guidance for developing an adequate bases for program
assumptions. The valves in these groups also had small apparent thrust margins.
Grouc #6
Valve Group #6 (FW-V-92A & 92B - close safety function) consisted of 6" Crane
900 # flex-wedge gate valves. The licensee had assigned a 0.42 valve factor for
this group, which had a design-basis close differential pressure of 580 psid. The
valve factor was based on a friction coefficient obtained from the ambient fluid
temperature, pumped flow closing test of Valve #14 from EPRl's PPP. Two other i
valve factors from similar valves at two other PWRs were also considered. No l
specific analysis method was applied to the industry data, except that more weight
was given to the EPRI test result. One utility had conducted two dynarric tests on
a single MOV, and obtained valve factors of 0.363 (Test 9) and 0.463 (Test 10).
The other utility's dynamic test obtained a valve factor of 0.365. The inspectors
noted that the tests were conducted at differential pressures that were more than
double the design-basis conditions specified for the TMI Group #6 MOVs. EPRI
testing has shown that valve factors tend to decrease as differential pressure ;
increases. Therefore, it is reasonable to expect that the industry valve factors I
would be higher if the dynamic tests had been conducted at conditions closer to l
580 psid. !
!
The inspectors noted that Valve Group #6 included an additional 12 data points
from industry (outside TMI) dynamic tests of six other MOVs. However, these
valve factors were excluded from consideration ("given less weight"), because the
valve / actuators were orientated differently than FW-V-92A & 928. No technical
justification was provided to support these exclusions. A review of this data
showed that every excluded data point, except for one, was higher than the applied i
0.42 valve factor. I
!
1
Grouc #9 l
l
Valve RC-V-2 (PORV Block Valve) in Valve Group #9 was a 2.5 "Velan 2500 # flex-
wedge gate valve. The licensee had assigned a 0.40 valve factor for this group,
1
l
l
l
.. - -. -- - --- -- - - . . .-._.
c
.
i
'
7
4 which had a design-basis close differential pressure of 2367 psid. This valve factor
was based on a friction coefficient obtained from the ambient fluid temperature,
4 high pumped flow closing test of Valve #13 from EPRI's PPP, and in-plant test
results obtained from two 1500 # Velan gate valves (MU-V-36 & 37). The close
,
valve factors for these two valves were less than 0.40. However, the inspectors
i noted that MU V-36 & 37 were of a different pressure class than RC V-2. Further,
- when selecting EPRI data, GPUN did not use the highest valve factor observed by
EPRI prior to hard-seat contact (0.452) for Valve #13. Instead, a valve factor that
- EPRI identified as the valve factor present at flow isolation was used in the -
!
calculation. The use of flow isolation valve factors when applying test data to l
- similar nondynamically-tested MOVs is not appropriate due to the valve specific !
i nature of determining flow isolation. )
1
The industry data acquired for Valve Group #9 were for 1500 # gate valves and did l
I
not include any valve factors for 2500 # gate valves similar to RC-V-2. A close
valve f actor of 0.424 was obtained from a similar 2.5" Velan 1500 # gate valve at
ANO. However, this data was excluded ("given less weight") on the basis that the j
'
data came from another nuclear facility (instead of from in-plant testing or EPRI l
testing). Valve factors obtained from 20 other 1500 # Velan gate valves were l
excluded on the basis that they were slightly larger in size (3" to 4") than RC-V-2.
d
The licensee stated that the average valve factor for this other industry data was
,
0.56.
1 In summary, the inspectors considered the valve factor justifications for Valve
j Groups #6 and #9 (RC-V-2 only) to be weak. The close thrust margins (based on
the current valve factor assumptions) were as follows: FW-V-92A - 2.2%; FW-V-
i 928 - 3%; and RC-V-2 - 20.1 %. Appropriate technical justifications for the
- selection of the Group #6 and #9 valve factors are necessary to establish the .
j valve's design-basis capability, prior to GL 89-10 program closure.
,
s
Valve Factors - EPRI Data
3
The licensee used disc friction coefficients obtained from EPRI's PPP to support
- valve factor justifications for several valve groups. The inspectors had several
concerns with the application of these data.
3 Some valve groups (e.g., Valve Groups #6 and #13) used an individual EPRI disc
friction coefficient as the primary basis for the selected valve factor. Other valve
i groups (e.0., Valve Group #14) listed additional industry data, but the selected
valve factor for the group closely matched the EPRI disc friction coefficient. The
j- EPRI disc friction coefficients were documented in EPRI report, "EPRI MOV PPP
- Update of Results and Specifications and Drawings for Flow Loop Test Valves,"
dated December 14,1993. Consistent with the guidance provided in Supplement
6, individual data points do not demonstrate similarity of valve performance and,
therefore, are not adequate justification for valve factors applied to nondynamically-
tested MOVs.
l *
.
8
The EPRI PPP determined apparent disc friction coefficients at several points during
a valve stroke. These friction coefficients are not equivalent to valve factors and
should be converted to valve factors by accounting for the angle of the valve's
seating surface. These conversions were not made, which resulted in the use of
nonconservative valve factors (for the closing direction).
EPRI PPP disc friction coefficients may not be reliable for use as an individual data
point due to the lack of preconditioning (in some cases) and the general practice of
removing apparent parasitic loads from the measured force requirements before <
calculating the apparent friction coefficient. Removal of parasitic loads results in a
nonconservative thrust iequirements if the disc friction coefficient is applied to a .
different valve and the parasitic loads are not added back into the minimum thrust
requirement. The EPRI PPP does not identify when parasitic loads were present.
Valve Group #13 consisted of MS-V-2A and 2B (12" Walworth 600 # flex-wedge
gate valves) that are main-steam system valves that isolate the line that provides
steam to the auxiliary feedwater turbine. A valve factor of 0.47 was selected
based on early PPP test results of EPRI Valve # 31, which was a 12" Walworth 150
- flex-wedge gate valve that was tested under ambient temperature, low pumped
flow conditions. The inspectors questioned the selection of this EPRI valve based
on the lower pressure class and the difference in the test conditions as compared to
the design-basis conditions identified for Valve Group #13. The inspectors
indicated that EPRI Valve #30 (6" Walworth 900 # flex-wedge gate valve) would
have been a better match because of the typical similarity of construction between
600 # and 900 # gate valves, and because the EPRI testing for Valve #30 was
done under high temperature fluid flow conditions. These test conditions may be
more applicable should MS-V- 2A and 2B be cal!ed upon to close against a
downstream steam line break.
The inspectors noted that Valve Group #9 used an EPRI disc friction coefficient
(0.287 - EPRI Valve #13) that was measured at flow isolation. This same EPRI test
had a maximum close friction coefficient of 0.452. Due to the arbitrary nature of
determining flow isolation, the observed flow isolation friction coefficients are valve
specific. As stated in the NRC's safety evaluation report (SER) by the Office of
Nuclear Reactor Regulation of Electric Power Research Institute Topical Report TR-
103237, "EPRI Motor-Operated Valve Performance Prediction Program," dated
March 15,1996, "EPRI states that the model output for flow isolation is
' theoretical' flow isolation position that is for information only and is not to be used
to establish thrust requirements in accordance with the EPRI methodology."
The NRC staff reviewed the EPRI performance prediction model (PPM) software as
documented in the above referenced SER. The staff's endorsement of the PPM
(with the conditions stated in the SER) only covers the use of the PPM software.
The SER does not accept use of the PPP's individual disc friction coefficients.
Because of the above concerns, the inspectors did not consider the use of EPRI's
PPP friction coefficients to be appropriate for justification of vc!ve factors used by
Q
.
9
their grouping methoo., logy. A sound technical justification for the use of EPRI PPP
dato in selecting valve iactors is necessary prior to GL 89-10 program closure.
Mlve Factors -Industrv hta
T te licensee obtained valve factors information from industry sources via phone
e,onversations with other licensees. The inspectors noted a lack of thorough
supporting documentation for the valve factors used from industry sources.
Specifically, the licensee had not thoroughly documented the following information:
- The test conditions (i.e., pressures, flows, fluid temperatures) and system
configurations should be documented. Pressure instrument locations and
methods used to record differential pressure should be clearly understood.
This is especially important for cases where test results appear to be
abnormally low.
- How the force point on the diagnostic trace was selected (e.g., at flow
isolation, hard seat contact, or highest force up to hard-seat contact).
- What was the valve seat diameter measurement used by the valve factor
calculation and whether this diameter was obtained by measuring the valve's
orifice diameter, the mean seat diameter, or by some other point of
reference.
- The general method used to calculate the valve factor. This would include
knowing how packing loads and stem rejection loads are determined and
removed from the force measurements. It would also include knowing if
parasitic loads were removed before calculating the valve factor. l
This inforrnation is necessary to ensure that reliable data is used to establish design-
basis requirements. Documentation for industry data, used to determine valve
f actors, is necessary prior to the closure of the GL 89-10 program. I
Valve Factors - Nondynamicallv-Tested MOVs
The MOV program included several valve groups where the valve factor
justifications were inadequate:
Group 1 - 3" & 4" Alovco 150 # Solit-Wedae Gate Valves - Valve Factor = 0.50
I
One MOV (BS-V-2B) was tested; however, the licensee was unable to determine a l
valve factor due to a lack of observable differential pressure effects in the !
diagnostic trace. One EPRI prototype test friction coefficient was considered.
Twelve industry valve f actors (from eight MOVs) were obtained over the telephone
or via telefax communications. Some industry valve factors were larger than 0.50
and others were less than 0.50. Therefore, it was not clear how these data were
l
.
10
analyzed and TMi had not determined the disc orientation of the valves tested by
the industry sources (see the next section addressing concerns related to disc
orientation of Aloyco split-wedge gate valves).
Group 2 - 14" Alovco 150 # Double-Disc Gate Valves - Valve Factor = 0.60
In-plant or industry data had not been collected to support this assumption.
Group 3 - 4" Alovco 300 # Solit-Wedae Gate Valves - Valve Factor = 0.60
One MOV (DH-V-78) had been tested, and the licensee measured a valve fact'o r of
0.66 for the open safety direction. This valve factor was not used (see Section 3.0
of this report). Four industry valve factors (from two MOVs) were calculated by
TMI by reviewing the diagnostic traces. Five other industry valve factors (from
three MOVs) were obtained over the telephone or via telefax communications.
Some industry valve factors were larger than 0.60, and others were less than 0.60.
Therefore, it was not clear how these data were analyzed to arrive at a 0.60 valve
factor.
Group 9 - 2" Velan 1500 # Flex-Wedae Gate Valves (MU-V-36/37-ooen oniv) -
Valve Factor = 0.40
in-plant open test data was not available for MU-V-3G/37. One EPRI prototype
open test friction coefficient was not used. Nine industry valve factors (from seven
MOVs) were calculated by TMI by reviewing the diagnostic traces. Ten other
industry valve factors were obtained over the telephone or via telef ax
communications. None of the industry data were used for the open safety
direction.
Groun 10 - 4"/6"/8" Walworth 150 # Solid Wedae Gate Valves - Valve
Factor = 0.40
One MOV (IC-V-2) was tested and the licensee measured a valve f actor of 0.44 for
the open safety direction. This valve factor was not used (see Section 3.0 of this
report). Three other MOVs were dynamically tested, but TMl was unable to
determine a valve factor due to a lack of observable differential pressure effects in
the diagnostic trace. One EPRI prototype test friction coefficient was considered.
Six industry valve factors (from two MOVs) were calculated by TMI by reviewing
the diagnostic traces. One industry valve factor was obtained over the telephone or
via telefax communications. Some of these industry valve factors were larger than
0.40, and others were less than 0.40. Therefore, it was not clear how these data
were analyzed to arrive at a 0.40 valve factor.
Group 13 - 12" Walworth 600 # Solid Wedae Gate Valves - Valve Factor = 0.47
No in-plant or industry data was available to support this assumption. One EPRI
prototype test disc friction coefficient was the basis for the selected valve factor
early in the program's development. However, EPRI increased this friction
l
, . - -. - . . - . _ . -
.
.
11
coefficient in a later PPP test report, and TMl did not update their assumptions to
account for this change.
Group 14 - 6" Walworth 900 # Flex-Wedae Gate Valve - Valve Factor = 0.40
No in-plant data was available to support this assumption. One EPRI prototype
open test friction coefficient was considered. Nine other industry valve factors
(from four MOVs) were obtained over the telephone or via telefax communications.
Some industry valve factors were larger than O.40, and others were less than O.40.
Therefore, it was not clear how these data were analyzed to arrive at a 0.40 valve
factor.
Group 16 - 10"/12"/14" Walworth 1500 # Flex-Wedae Gate Valves - Valve
Factor = 0.50
Three MOVs were dynamically tested; however, the licensee was unable to I
determine the valve factors due to a lack of observable differential pressure effects
in the diagnostic trace. Industry valve factors for three MOVs were obtained over
the telephone or via telefax communications. Some of these industry valve factors
I
were larger than 0.50, and others were less than 0.50. Therefore, it was not clear
how these data were analyzed to arrive at a 0.50 valve factor.
While the MOVs in the valve groups listed above appear to have adequate margin
based on current assumptions, the basis for the thrust / torque requirements have not
been well established. Additional information (e.g., results from EPRl's PPM or
other applicable industry data) is necessary to support current thrust / torque
requirements for these MOVs prior to GL 89-10 program closure.
Valve Orientation
industry experience has shown that Aloyco split-wedge gate valvo performance is
sensitive to valve disc orientation. The licensee obtained Aloyco valve factor data
from other industry sources for use in their grouping methodology. However, the
test valves' disc orientations were not provided by the industry sources. Further,
the engineering staff were not sure of the orientation of certain Aloyco split-wedge
gate valves. Therefore, some of the industry data may not be applicable to the
valves at TMI. Further evaluation is necessary to obtain orientation information for
the industry data that was acquired, remove nonconservative data points, and
reassess the assigned valve factors for Valve Groups #1 & #3.
Load-Sensitive Behavior
The load-sensitive behavior data was provided in Appendix J of TMI's program
description. The average of the data for rising stem MOVs was 2.29%, with two
standard deviations oi 10.29%. The average value was increased to 2.9% and was
added to the thrust requirements as a bias error. The two standrtrd deviation value
(10.3%) was combined with other errors and uncertainties in a square root sum-of-
the-squares methodology.
-
.
12
The MOVATS displacement measuring device (DMT) was used during dynamic
testing to determine thrust output. The DMT was calibrated to indicate thrust by
correlating the spring pack displacement to a MOVATS stem strain ring's (SSR)
thrust measurement during a static test. However, the MOVATS DMT will not
detect the presence of load-sensitive behavior that may occur during a dynamic
test. This is because load-sensitive behavior is the result of an increase in stem
friction coefficient caused by the higher stem load present during a dynamic test. If
the stem friction coefficient increases during the dynamic test from what was
present during calibration of the DMT, the DMT measurement will overestimate the
actual thrust that was applied to the stem. Therefore, the DMT provides unreliable
thrust values during dynamic testing, an is unable to detect the presence of load-
sensitive behavior. Appendix J analyzed data from 23 rising stem MOVs, of which
21 were tested using the MOVATS DMT. Based on the inability of the DMT to
identify the presence of load-sensitive behavior, only two of the data points (which
were measured using the VOTES diagnostic system) were valid. The inspectors did
not consider two data points adequate to justify the licensee's load-sensitive
behavior assumptions for rising stem MOVs and were unable to close this aspect of
TMI's generic letter program.
- TMl used the VOTES diagnostic system to diagnostically test rising rotating-stem
globe valves. Because the VOTES system directly measures the thrust applied to
the valve stem, it can detect load-sensitive behavior. TMl tested eight rising
rotating-stem globe valves. The VOTES equipment was used for four of these
dynamic tests and the MOVATS DMT was used for the remaining four tests.
Therefore, the four VOTES tests provided valid load-sensitive behavior results.
Appendix J identified the average of the rising rotating-stem globe valve data as
6.22%, with two standard deviations of 17.74%. The inspectors independently
analyzed the four VOTES data points and obtained slightly higher values. However,
given the uncertainty involved with analysis of this small number of data, the
inspectors considered the licensee's assumptions for rising rotating-stem globe
valves to be adequate. The licensee will need to collect additional load-sensitive
behavior data, to improve their justification, as part of their periodic verification
program.
S_lem Friction Coefficient
The thrust calculations used a 0.20 stem friction coefficient assumption. Test
results were analyzed in a draft document that was generated to support TMl's
MOV trending program. The inspectors noted that one gate valve and two globe
valves had stem friction coefficients significantly above 0.20. All remaining test
data points were 50.20. Therefore, the inspectors determined that the licensee's
basis for the use of a 0.20 stem friction coefficient assumption was adequate. The
licensee will need to increase their confidence level in this assumption by obtaining
additional stem friction coefficient data in the future, as part of their MOV periodic
verification program.
a
.
13
Diaanostic Eauipment Uncertainties
On September 21,1993, GPU Nuclear Corp. submitted a response to the NRC's
4 Generic Letter 89-10, Supplement 5, " inaccuracy of Motor-Operated Valve
Diagnostic Equipment." This submittal documented TMI's actions taken to resolve
issues related to the open calibration of the MOVATS DMT using a load cell. To
1
resolve this issue,28 MOVs were reviewed and found to have acceptable settings.
TMI also uses the VOTES diagnostic system. However, all test data was acquired
with a version of the VOTES software that included torque correction factors.
Therefore, no actions were required for existing VOTES tests.
During discussions with TMI personnel, the inspectors identified that the licensee's
MOV switch setting methodology did not account for the diagnostic equipment
uncertainty associated with use of the DMT to determine actuator torque output.
Accounting for these uncertainties is important to ensure that torque limits are not
exceeded. The licensee will need to conduct a screening to determine the impact of
this error, take action as necessary, and to correct diagnostic pucodures prior to
MOV program closure.
c. Conclusions
in general, TMI's did not implement a rigorous analysis methc,d for application of l
1
industry information when determining valva factors for nondynamically-tested
MOVs. Given the many concerns, the inspectors did not find TMI's methods for
justifying the valve factors and load-sensitive behavior assumptions used in design- .
'
basis thrust equations to be adequate for GL 8910 program closure.
i
10 CFR 50, Appendix B, Criterion 111, and the GPU Nuclear Operational quality
Assurance Plan require that design control measures be established for verifying and
checking design inputs. The failure to establish proper design control measures,
when determining the valve factors for certain safety-related valves, is a violation of
these requirements (50-289/96-05-01). ,
i
El 3 Desian-Basis Capability
a. Inspection Scone
The inspectors reviewed TMl's " Program Description for NRC Generic Letter 89-10
Motor-Operated Valve Program" valve test packages and associated test reports for
the selected MOVs. The purpose of this review was to assess TMI's efforts to
establish design-br. sis capability for all MOVs in their GL 89-10 program.
b. Observations and Findinas
Thrust Calculations
The dynamic test of DH-V-3 measured a close valve factor of 0.56 and an open
valve factor of 0.64. The thrust calculation for D H-V-3 used an inappropriate valve
,
!
a
.
14
factor of 0.50. The licensee revised the thrust calculation using the measured valve
factors. The revised calculation indicated that the current settings were adequate
because the design-basis differential pressure was being reduced to remove a
mispositioning scenario. The inspectors found other cases where measured open
valve f actors were not applied (i.e., DH-V-78 and IC-V-2). TMI indicated that these
valve factors were not valid and should not be applied, because: (1) DH-V-7B;
differential pressure effects were not detected and unwedging loads were used to
calculate the valve factor, and (2) IC-V-2; differential pressure effects were not
detected and the force equivalent to spring pack preload was used to calculate the
valve factor. The inspectors noted that TMi's grouping justifications did not contain
this information. The licensee will need to revise the appropriate program
documents and thrust calculations to incorporate tested valve factors prior to
program closure.
Toraue Switch Repeatability
The TMl Program Description, Data Sheet 2, Appendix H, was used to document
the post-test acceptance criteria for static and dynamic testing. Completion of this
checklist was required prior to returning an MOV to service. The inspectors noted
that the dynamic test margin assessment did not consider the added margin
required to account for torque switch repeatability. The licensee agreed with this
observation and will revise the checklist to include this margin. Further, TMI will
review the impact of this omission. These items will require resolution prior to MOV
program closure.
Test Data Extracolation
The dynamic test review checklist for the dynamic test of RR-V-5 (Bypass Valve for
RR-V-6 Pressure Control Valve - 10" Pratt 150 # Butterfly Valve), noted that the
dynamic test differential pressure was 128 psid, which was approximately 90% of
the design-basis differential pressure of 141 psid. However, the dynamic test
review checklist identified this as a full differential pressure test and did not
extrapolate the hydrodynamic test results to design-basis conditions. TMI personnel
were unable to document why test results were not properly extrapolated, but
postulated that the extrapolation was not done, because the test differential
pressure was unknown at the point of maximum load during the open stroke. While
the lack of differential pressure information may hinder accurate determination of
design-basis torque requirements, it does not justify the documented evaluation j
where the test was considered a full differential pressure test. Similar concerns
were identified for valves NR-V-4A and NR-V-4B. The inspectors also noted that, if J
the measured seating or unseating torque bounded the measured hydrodynamic 1
torque, TMl assumed that the seating / unseating torque was bounding and that no ,
further extrapolation to design-basis conditions was needed. However, the I
. hydrodynamic torque should be extrapolated before making a comparison to the
measured seating / unseating torque. The licensee will need to complete these
extrapolations and reassess the adequacy of the vendor's (or TMi's internal)
calculations.
l
l
i
l
-.
O
e
15
Post-Maintenance Testina
The licensee's post-maintenance requirements are contained in Corrective
Maintenance Procedure 1410-V-10, " Gate, Globe, and Needle Valve Maintenance,"
Rev. 29, dated June 6,1996. The inspectors noted that this procedure did not
clearly indicate when dynamic test conditions would be considered for a post-
maintenance test requirement. Licensee personnel stated that they would clarify
their procedures to indicate when dynamic testing would be considered.
c. Conclusions
Several deficiencies were identified with the MOV thrust calculations and test data
differential pressure extrapolations. Other weaknesses were identified in the
methods used to verify test margins for dynamically-tested valves and procedure for
establishing post-maintenance test criteria. The inspectors concluded that
additional attention is required to improve the current GL 89-10 program prior to
program closure.
E1.4 GL 89-10 Proaram Scoce
a. Inspection Scone
The inspectors reviewed the licensee's technical bases for removing the main steam
isolation valves, turbine bypass line, and steam bypass to the condenser isolation
valves, and the emergency feedwater turbine steam supply isolation valves from the
GL 89-10 program.
b. Observations and Findinas
Main Steam isolation Valves (MS-V1 A/B/C/D)
The Generic Letter 89-10 program description, Table 1, " Selection of Valves For GL 89-10 Program," states that the safety function of the MSIVs is only the check
valve function. The licensee concluded that the motor operator does not perform a
safety function. The licensee's MOV engineers stated that the MSIVs are not
credited in UFSAR accident analyses for isolating a faulted steam generator due to
the small steam generator secondary side water inventory, the magniturie of the
blowdown, and the relatively slow valve stroke time.
The inspectors noted that the licensee's basis for exclusion of the MSIVs from the
MOV program did not address how the " tight-closing" containment isolation i
function is performed without the assistance of the motor-operator; nor does it l
consider the apparent licensing basis function to close against saturated steam. i
.Therefore, the inspectors concluded that these valves may belong within the scope l
of the GL 89-10 program. !
l
.
e
i .
16
Turbine Bvoass Line and Steam Bvoass to the Condenser Isolation Valves (MS-
V2A/B and MS-V8A/B)
The UFSAR Table 10.3-1, states that the MS-V2 isolation valves have an
approximate closure time of one minute against a maximum differential pressure of
1000 psid. However, the licensee's Generic Letter 89-10 program description
states that their safety function is to close against approximately 51 psid and
provide containment isolation following a main steamline break in the reactor
building. These valves would not likely close against a differential pressure greater
than approximately 500 psid.
The licensee's GL 89-10 program description indicates that valves MS-V8A/B,
located downstream of MS-V2A/B, are not included in the GL 89-10 program.
Valves MS-V8A/B are normally maintained in their required (open) position, but the
operator does not provide an active safety function; the licensee has excluded the
8A/B valves by reliance, instead, on the upstream 2A/B valves. The inspectors
concluded that these safety-related valves are the class boundary between
nonsafety grade (and nonseismic) main steam line piping and may belong within the
scope of the GL 89-10 program.
,
Ememency Feedwater Turbine Steam Suoolv isolation Valves (MS-V10A/B)
The Generic Letter 89-10 program description provided a justification for removal of
'
the EFW supply isolation valves from the GL 89-10 program, stating that: (a) 1
'
automatic start of the EFW turbine is initiated by opening valves MS-V13A/B; (b)
flow-through valve MS-V13 is adequate at hot shutdown, which is the ESW design
basis for TMI-1; (c) valves MS-V10A/B are available during cooldown, when steam
pressure is reduced to the point that MS-V13A/B no longer provides sufficient
steam flow, and (d) a diverse motor-operated EFW pump will provide required flow,
or MS-V10A/B could be manually opened.
c. Conclusions
The inspectors concluded that the licensee did not thoroughly document the
technical bases for removing these valves from the GL 89-10 program. In addition,
the inspectors considered that the licensee's justifications may have constituted
changes in the design basis of TMI-1 that were not evaluated by the licensee per 10
CFR 50.59. This matter is unresolved pending additional NRC review of the design
basis of the main steam valves (URI 289/96-05-02).
E4.1 TMl Resoonse to the Arkansas One Unit 1 OTSG Blowdown
a. Insoection Scone (37551)
The inspectors reviewed the TMl response to the Arkansas Nuclear One (ANO) Unit l
1 once through steam generator (OTSG) dryout event. The ANO-1 event included a
stuck open main steam safety relief valve that did not re-close when steam pressure
returned to normal because of a valve cotter pin and release nut that were too close
_ _
m
.
.
17
to the valve top lever. The stuck open valve resulted in the loss of the OTSG steam
and water inventory. The review included the TMl response to previous OTSG
dryout events, emergency operating procedures (EOPs) related to refilling a dry
OTSG, plant walkdown of the TMl safety relief valves, and discussions with the
OTSG system engineer and control room operators.
b. Observations and Findinas
The inspectors noted that the responsible engineer was very knowledgeable about
the ANO-1 safety concerns and understood how they related to the TMl OTSG
safety relief valves. The engineer inspected the TMl main steam safety relief
valves, prior to the inspectors review, to verify that the cotter pin was properly
installed and to ensure that the release nut had at least a 1/16 inch clearance above
the valve top lever. The TMl steam relief valves are Dresser valves that are
identical to the ANO-1 safety relief valves. The engineer's examination confirmed
that the TMl safety relief valves met the 1/16 inch vendor recommended clearance
and the cotter pins were installed properly on the release nut.
The residents performed an iridependent walkdown to verify the physical
arrangement of the cotter pins and release nuts on the safety relief valves. The
inspection confirmed that the TMl safety relief valves had more than the minimum
required clearance between the release nut and top lever. In fact, all valves
inspected had at least one to two '.nches of clearance and the cotter pin was
properly installed to ensure the relecse nut would not vibrate down if the valve
lifted. The inspectors alt,o verified that TMI had taken actions in 1984 in response
to a similar problem at Davis Besse.
A second area reviewed for the ANO-1 event was the EOPs available to the
operators for re filling a dry OTSG. The inspectors discussed the ANO-1 event with
the control room operators, the operators were very knowledgeable of the ANO-1
and Oconee events related to the dryout and subsequent refilling of the OTSG. The
operators were very familiar with the applicable EOPs and knew exactly how to refill
a dry OTSG. The procedures provided the preferred water make-up source for a dry
OTSG, main feedwater (MFW) with steam pre-heating, or emergency feedwater if
MFW was not available. The procedure contained excellent caution statements
describing the OTSG tubesheet and shell temperature limitations and associated
heatup limits, in addition, the procedure provided a clear definition of the
indications a CRO would use to determine if a generator had reached a dryout
condition.
c. Conclusions
The TMI engineering and operations response to the Arkansas Nuclear One OTSG
dryout and stuck open safety relief valve event was comprehensive, thorough, and
displayed a strong initiative to address generic safety issues at other plants in order
to reduce the potential for a similar impact at TMI.
._ __ . . _ _ _ _ _ _ _ _ _ . _ _ _ _ . ._ _ _ _ _ _ . _ _ _.- _ __
e
l
a. !
l
4
- l
'
l
'
18 i
I
E8 Miscellaneous Engineering issues (92903)
l
fJk1_ (Closed) Unresolved item 50-289/94-12-01:
'
l
The NRC Inspection Report 94-12 (URI 50-289/94-12-01) laentified several concerns .
l
,
related to TMl's use of best-fit-straight-line calibrations with their VOTES diagr;ostic l
j system. In response to this concern, TMl initiated Engineering Evaluation Request EER-94- l
, 0325, " VOTES Data Re-analysis," on October 6,1994. This evaluation reviewed all of i
TMI's VOTES test data and utilized the vendor recommendations contained in Liberty- .
"
- Technologies' Customer Bulletin (CSB) 032, which provided the methods for applying a
,
curve fit calibration to existing test data. Engineering Evaluation Request (EER) 94-0325 l
j determined that operation of FW-V-92A/B should be restricted to 100 cycles because the
application of curve fit calibrations increased the diagnostic inaccuracy to the extent that ;
!
j the measured peak thrust values exceeded 120% of the published actuator thrust ratings
for these MOVs. All other reviewed MOVs were found to have acceptable torque switch
settings. The Engineering Evaluation Request EER-94-0325 recommended that actuator
j inspections be performed and that torque switch settings be adjusted for FW-V-92A/B
- during the next plant outage. The inspectors reviewed September 1995 work requests
l that documented completion of the actuator inspection activities, including adjustment of
f the torque switch settings for FW-V-92A/B and confirmatory diagnostic tests. Based on
- review of this information, this unresolved issue was considered closed.
IV. Plant Support
l R5 Staff Training and Qualification for RP&C, Security and Chemistry /Radwaste
,
(71750) l
R5.1 Personnel Self-Checkina Trainina
i
a. InsDeCtion SCoDe
, The inspectors observed the radiological controls, security and chemistry /radwaste
- departments customized self-checking training. The training was initiated to
reinforce department specific examples related to performance that did not meet
< managements' expectations and could have been prevented by the use of the "BE
l SURE" self-checking program.
'
b. Observations and Findinas
} The training was presented in two parts, the first segment was presented by the
1 training department and provided an overview of the TMI self-checking program.
2 The presentation included generic examples of cases when self-checking
4 contributed to a positive performance and a videotape that demonstrated numerous
- . examples of unacceptab!e performance due to the lack of self-checking attributes.
i
,
After the classroom preseatation, the department management conducted an open +
'
forum discussion with their respective work force. Good ideas were exchanged for
- all of the department discussions, the most effective exchange of information
i
a
f
$
-- . , --
_ _ _ _ _ _
_ - _
e
l
l
.
19
occurred in the radiological controls discussion. The Radiological Controls Director
and senior managers provided their management expectations and specific
,
department examples of less than adequate performance to the employees. The
management team also fielded questions and concerns of the technicians to collect
additional ideas that could result in improved plant performance. The involvement
of the Radiological Controls Director and senior managers reinforced the
management support and expectations to continue efforts that could lead to
improved plant performance,
c. _Qgnclusions
in conclusion, the department cu*- Red self-checking training was a positive
initiative to reinforce managemer metations for each department. The
presentations included generic e of cases where self-checking contributed
to a positive performance and a that demonstrated numerous examples of
unacceptable performance due to thu lack of self-checking attributes.
Ril Miscellaneous Radiological Control Program items
R8.1 Previousiv Identified items (92904)
(Closed) Violation (VIO, 50-289/95-13-02) Unauthorized Entry into a Radiography
Area
a. Inspection Scoce
<
On September 6,1995, an auxiliary operator (AO) violated a radiography posting
and barrier and entered an area in which radiography was being performed. The
source was not exposed at the time and the AO received no exposure from the
device. The violation was identified by a TMI contractor radiographer and a site
engineer. The TMl radiographer and engineer did not inform either the shift
supervisor or other responsible management of the radiography area boundary
violation, in order to ensure adequate immediate correction of the problem, before
resuming radiography operations.
!
b. Observations and Findinas
Initially, the long-term corrective action was to review the use of postings for
possible improvements. However, the licensee had not committed to any other
long-term or corrective measures designed to prevent recurrence of this type of
event, such as other controls and reinforcement of the requirement to adhere to
radiological controls procedures and postings, in summary, the licensee's problem
identification and correction process was considered inadequate, in this instance, ;
'
because: (1) the shift supervisor or other responsible licensee management was not
provided an opportunity to exercise management oversight and review of the
occurrence prior to the resumption of radiography operations, and (2) originally
determined long-term corrective actions were limited only to review of use of
postings for improvements.
O
.
20
The licensee subsequently took several additional corrective actions to improve the
radiography process. The inspectors reviewed procedure 6610-ADM-411.07,
" Radiography Operations," to determine if procedure changes included a
documented review of all controls for each radiography operation prior to the start
of the job. The procedure was revised to provide clear direction to the radiological
control personnel for proper control of radiography evolutions. The procedure was
revised to include: 1) incorporation of a separate pre-job briefing checklist for
radiography operations; 2) written approval from all parties in a radiography
operation stating that they understand and agree to the radiological controls; 3)-
detailed instructions for abnormal occurrences and management review prior to
resumption of radiography activities; and 4) an ALARA review by radiological
engineering.
In conjunction with the procedure revision, the following additional corrective
actions were performed: a review of the signs used to warn workers regarding
radiography operations; revision of the training program for workers regarding these
changes; and issuance of a memorandum from management that communicates
management expectations regarding the use of self-checking, observation, and
coaching practices.
The inspectors observed implementation of the revised radiography controls and
procedure implementation during the performance of a radiography evolution in the
Turbine Building on July 20,1996. Prior to the radiography a detailed pre-evolution
briefing was provided to the shift personnel by the shift supervisor (SS) and the
group radiological control supervisor (GRCS). The shift supervisor and GRCS
emphssized the importance of personnel accountability and reviewed some of the
problem:. and corrective actions related to the previous radiography evolution. Plant
management's decision to perform the radiogrephy on the weekend resulted in a
relatively small number of workers onsite and minimized the potential for an
individual to inadvertently cross the boundaries.
In addition to the procedure changes, improved controls were used to alert plant ;
workers to the radiography boundaries. The controls included the installation of
large hexagon caution signs,20 inches by 20 inches, and flashing strobe lights at
all of the main access points. The inspectors independently verified that all access
control points in the Turbine Building were posted properly prior to the start of the
radiography examinations. Also, the radiological control technicians monitored the
boundaries continuously throughout the evolution.
, c. Conclusions
The radiography procedure changes and boundary controls significantly improved )
the radiological controls personnels' ability to alert all plant workers abcut the
' conduct of radiography and ensure.the exclusion boundaries were contto!Ied
properly. The inspectors concluded that procedure 6610-ADM-411.07 was revised
appropriately to address the prior radiography problems. In addition, the licensee's
corrective actions were comprehensive and should reduce the probability of the
recurrence of similar events. This item is closed. ;
!
l
.
21
V. Manaoement Meetinas
l
X1 Exit Meeting Summary l
- At the conclusion of the reporting period, the resident inspector staff conducted an exit. l
meeting with TMI management on August 22,1996, summarizing Unit 1 inspection i
'
activities and findings for this report period. TMI staff comments concerning the issues in
this report were documented in the applicable report section. No proprietary information
-
was identified as being included in the report.
l
X2 Meeting With GPU Nuclear Corporation Regarding Motor Operated Valve (MOV)- l
Program Concerns l
On July 22,1996, a public meeting was held between the NRC and GPU Nuclear
Corporation at the NRC Region 1 Office in King of Prussia, Pennsylvania. The purpose of ,
the meeting was to discuss the TMI-1 MOV program deficiencies that were raised during l
an inspection held the week of June 17,1996. GPU committed to inform the NRC about
the planned corrective actions to address the MOV issues and also provide a closure date
for the Generic Letter No. 89-10, " Safety-Related Motor-Operated Valve Testing and
Surveillance," program. The handouts used at the meeting are attached to this report.
X3 GPU Nuclear Engineering Integration Meeting
4
On July 24,1996, a meeting was held between the NRC and GPU Nuclear Corporation at !
the NRC Region 1 Office in King of Prussia, Pennsylvania. The purpose of the meeting was
to discuss the GPU engineering integration process and outline the newly organized
,
engineering organization. The key elements were focused on GPU's ability to maintain an
emphasis on safety as they streamline the engineering department into a more efficient,.
Unit. the handouts used at this meeting are attached to this report.
i
l
'
I
d
,
.
.
PARTIAL LIST OF PERSONS CONTACTED
Licensee '
'J. Knubel, Vice President, TMI
M. Ross, Director, Operations and Maintenance
L. Noll, Plant Operations Director
D. Etheridge, Manager, Radiological Engineering
W. Potts, Radiological Controls / Occupational Safety Director
J. SchorL, Regulatory Affairs
J. Wetmore, Manager, Regulatory Affairs
G. Skillman, Technical Functions Site Director
P. Walsh, Engineering Director
- Senior licensee manager present at exit meeting on August 22,1996.
NEG
J. Norris, TMI Project Manager, NRR
t
t
l
. i
0-
)
l
'
l
l
I
1
22
, - .
.
!
.
INSPECTIDN PROCEDURES USED
IP 62707: Maintenance Observation
IP 62703: Maintenance Observation
IP 71707: Plant Operations
IP 71750: Plant Support
IP 92903: Followup - Engineering
IP 92904: Followup - Plant Support
ITEMS OPENED, CLOSED, AND DISCU!iSED ;
L
'
Opened
50-289/96-05-01, " Failure to Establish Proper Design Control Measures, when Determining
the Valve Factors for certain Safety-Related Valves" (VIO)
50-289/96-05-02, "NRC review of the Design Basis of the Main Steam Valves" (URI)
Closed ,
50-289/95-13-02, " Unauthorized Entry into a Radiography Area"
50-289/94-12-01, "TMl's use of best-fit-straight-line calibrations with their VOTES
diagnostic system"
i Uodated
None
f
l
l
i
!
l
l
1
l
23
i
l
. . - . ._ _ . .
,
.: 'l
. :
'
LIST OF ACRONYMS USED
AB_ Auxiliary Building
ALARA As low As Reasonably Achievable . l
ASME American Society of Mechanical Engineers ;
CDF- Core Damage Frequency l
CR Control Room i
CRO Control Room Operator !
CFR Code of Federal Regulations !
DBD' Design Basis Documents l
ECCS Emergency Core Cooling System ,
'
ED Emergency Director
EDG Emergency Diesel Generator - l
EOF. Emergency Operations Facility .
ENMCF Event or Near Miss Capture Form !
EPIP Emergency Plan and implementing Procedure !
.ESF Engineered Safety Feature ;
HEPA ' High Efficiency Particulate .!
HRA ' High Radiation Area _ l
IFl Inspection Followup Item
IPE Individual Plant Evaluation j
IR inspection Report i'
IST Inservice Testing Program
JO- Job Order f
'LCO Limiting Condition of Operation !
LER Licensee Event Report ;
t- MNCR Material Nonconformance Report l
. NCV Non-Cited Violation >
'
NRC' Nuclear Regulatory Commission
- NSA Nuclear Safety Assessment
i ODCM Offsite Dose Calculation Manual
'OSC Operations Support Center i
PAS Post Accident Sample
i PCR Procedure Change Request
PPB' Part per Billion !
4 PPM Part per Million I
PRA- Probabilistic Risk Assessment i
l PRG Plant Review Group l
QV : Quality Verification !
- .. RCA~ - Radiological Control Area
RP Radiation Protection - _
RSP Remote Shutdown Panel ,
RWP Radiation Work Permits -
SALP Systematic Assessment of Licensee Performance !
>
SF Shift Foreman
f
'
-
24
!
!
. . , . - - ._- . . . - . - - _ __. ._ _ _____. _ _.______.____
- . . . - . . . .. .. ... . . . . . . - .-. . . - . . . . . - . . . . . - - . . ~ . . ..
e
<
..
l
SRO Senior Reactor Operator l
SS ~ Shift Supervisor !
- Ti . ' Temporary Instruction ;
!
.TLD ~ Thermoluminescent Dosimeter
TS' Technical Specification l
TSC. Technical Support Center ;
UFSAR' Updated Final Safety Analysis Report ' !
. URI Unresolved item i
VIO. . Violation .;
!
!
!
5
I
i
l
t
4
I
i
f
I
I
E
$
i
'
t
b
I
,
E
25
- .. . _. .-_ - _-- - .. -_ .. . _ . . . - - . - . . - - _ . -
.
.
+
,
TMI-1
1
i
GENERIC LETTER 89-10 MOTOR OPERATED VALVE
PROGRAM
O
NRC / GPU NUCLEAR MANAGEMENT MEETING
.
JULY 22,1996
O
,_
..
-
-.
,
.
, ,
O .
AGENDA
I. INTRODUCTION R. W. Keaten ;
.
II. DEVELOPMENT OF VALVE FACTORS R. J. McGoey i
-
1,
III. CURRENT REVIEW OF FW-V-92A/B.
AND RC-V-2 T. Basso
IV. MAIN STEAM VALVES J. Link
.
.
O V. MOTOR OPERATED VALVE INDEPENDENT
REVIEW J. C. Fornicola
!
VI. SUMMARY R. W. Keaten .
.
I
!
i
l
c
!
!
!
' Ivfdean. doc
,o'
= . . .-_ .
- ;
. l
INTRODUCTION
,
MA.IOR ISSUES
e GL 89-10 Program inspection of TMI-l June 17-21,1996 identified
areas of concern
-
Valve factor development
-
Capability of FW-V-92 A/B and RC-V-2 l
-
Main steam valve design basis
ACTIONS TAKEN
e Reestablished basis of original evaluations and selection of valve
factors
- Engineering staff organized to reassess valve data, valve factor
selection, and available valve factor
. Engineering and Safety Analysis staff rereviewed main steam
system design basis.
- MOV independent review team established
O
im
.
.. . -. - . . . - . . . - - __ . - . - .-
- ,
!
. .
RESULTS TO DATE
o
e New higher valve factors for valves in Groups 6 & 9 have been t
i selected.
! * Based on rereview, additional conservatisms have been incorporated
! for Rate of Loading and Aging Degradation.
!
< .
I
e The valves remain capable of performing their intended safety
function.
- O l
!
i e Preliminary review of gate valves in other groups and the butterfly i
valves shows sufficient margin exists. l
l
i
9
e Main steam isolation valve design bases confirmed
i
e Based on f~mdings to date, a thorough technical review of the total i
I
,
program will be performed.
i
I
1
.O
,_
'
I
.
i
.
.
n
U
HISTORICAL DEVELOPMENT OF VALVE FACTORS
. Valve factors evolve as industry understanding and knowledge base
expanded.
. Original valve factors based on accepted industry / vendor
practice
. Valve factors upgraded based on early industry testing of valves
similar to TMI's.
- Extensive industry search for additional valve test data initiated
August 1994 to establish appropriate valve factor.
. Reviewed EPRI valve data
O
G
e Contacted over 90 plant / utilities (approximately 1200 man / hour
effort)
,
e Approximately 50 plants provided relevant data on over 500
'
valve tests.
. Information collected:
. Manufacturer
,
. Disc type
'
- Valve Size
- Pressure Class
e _ Material
e Valve Factor
- Orientation
- '"*'
O
Ivfdattdoc
.
_
.
.
O
HISTORICAL DEVELOPMENT OF VALVE FACTORS
0 Methodology Established for Valve Factor Selection
0 Hierarchy of test data significance
0 TMI valve specific test data
O GPUN test data
O EPRI test data
O Applicable Industry test data
0 [ PPM Model]
.
O Screening of Test Data ]
~225 test results screened and included on Tables
O
O Additional screening based on:
0 Data scatter concerns
0 Test methods and test accuracy
0 Differences between TMI operation and conditions of test
0 Consideration of valve / actuator orientation.
O Target Valve Factor Of 0.50 Established
i
! O MOV User Group and Industry Surveys
0 Used Graded Engineering Approach I
O Valves modified where margins were minimal
OO Additional Valve Factor Selection Changes
l
Ivfdata. doc
- I
a
- . . - - - - _ _ _ - - _ _ _ . _ _ _ - _ - _ _ _ _ . _ . _ _ _
. . . . ._. . . .
O O O' -
RESULTS OF ASSESSING VALVE FACTOR DATA FOR GATE VALVES
DECEMBER 1994
~
Valve Factor
SOURCE OF DATA <0.50 0.50 >0.50
GPUN GRP7-Iv GRP 8 - Iv (.7)
GRP 11 - 9v (.6)
GPUN & EPRI GRP 9 - 3v (.40)
GRP 10 - 6v (.40)
EPRI GRP 13 - 2v (.47)
EPRI & Industry GRP 6 - 2v (.42) GRP 1 - 3v
GRP 14 - 2v (.40)
Industry GRP4-4v ORP 3 - 2v (.6)
GRP5-IV
GRP 16 - 4v
PPM * GRP 15* - 2v (.555)
', No data available GRP 15 - 2v (.4) GRP2-2v GRP 2* - 2v (.6)
No. of Valves 17 15 12
- Changes made afte.r December 1994
Ivfdstadoc
_ _ _
-
. _ . .__ _ _
.
<
h
V .
,
.
'
CURRENT REVIEW OF FW-V-92 A/B AND RC-V-2
.
OVERVIEW OF VALVE EVALUATION
4 0 NRC questioned valve factor selection and available valve factors
.
O NRC pointed out misapplication of EPRI data j
! O We decided to take a rigorous look at valve factor selection process
O and rate ofloading correction for available valve factor.
O With higher valve factors and more conservatism in ROL, valves
remain capable of performing their intended safety function with
present setup.
O Preliminary review of balance of groups completed. Similar
rigorous rev;ew will be performed.
.
O
_
- i-
. . _ . -- . . - - .
.
<
O
CURRENT REVIEW OF FW-V-92 A/B AND RC-V-2
VALVE FACTOR
GENERIC TMI DATA APPLICATION HIERARCHY
e TMI Valve Specific Test Data
e GPUN Test Data l
l
1
O . seni Test Data l
. Applicable Industry Test Data
e PPM Model
O
_
._. -
. . _ . .. . .. .. . -. .-
l-
l
l
CURRENT REVIEW OF FW-V-92 A/B AND RC-V-2
GROUP 9 (2%",2500 # - Velan)
j VALVE FACTOR DETERMINATION
RC-V-2
i
i
e Previous Valve Factor (0.400)
. TMI Valve Test data available
- Same valve manufacturer and type
. Valve Factor (0.400) based on TMI test data
- TMI and EPRI #13 valve test data bounded
- ANO valve data used as reasonableness check
i
- PPM not available
O
Ivfdata. doc
.
i
l
,
O CDRRENT REVIEW OF FW-V-92 A/B AND RC-V-2
'
GROUP 9 (2%",2500 # - Velan)
VALVE FACTOR DETERMINATION !
RC-V-2
l
I
O Current Valve Factor (0.472)
J .
'
!i
i
l 0 Changed valve factor due to rereview ofIndustry and EPRI #13 valve
A data which included conversion of to valve factor and selection of a
i more appropriate EPRI valve factor.
i
i
applicable Industry data.
.,
O Calculations performed using 0.50 valve factor.
4 0 PPM test run on TMI valve not available.
.
O
u m aoe
a
-
-
- . - . . _ . .
.
.
O -
CURRENT REVIEW OF FW-V-92 A/B AND RC-V-2
GROUP 6 (6",900 # - Crane)
VAINE FACTOR DETERMINATION
FW-V-92 A/B !
. Previous Valve Factor (0.420)
- No TMI Valve Test Data available l
l
'
O - vaive Factor Data on eeni ui4 vaive
- Valve Factor (0.420) based on EPRI #14 Valve Data (0.419)
- EPRI Test Data is controlled and reliable
- Consistent with applicable industry data
,
. PPM not available
I
i
!
O ;
hha doc
. .
.
I
- .
&
V' CURRENT REVIEW OF FW-V-92 A/B AND RC-V-2
GROUP 6 (6",900 # - Crane)
VALVE FACTOR DETERMINATION
FW-V-92 A/B
0 Current Valve Factor (0.464)
l
0 Valve Factor changed due to conversion of EPRI #14 valve
'
l
(friction) value to valve factor. (.419 to .436)
0 Since orientation of EPRI valve (H/V) differs from TMI valves,
Industry data moved up in data hierarchy.
O O Selected valve factor (0.464) based on applicable Industry
data.
O Manufacturer, size, pressure rating, and orientation same as l
l
TMI valve.
0 Industry Valve Factor bounds EPRI #14 and applicable
Industry data.
0 EPRI #14 valve provides reasonableness check.
O PPM test data results on TMI valves yielded Valve Factor of ;
0.577. j
l
Q0 PPM tends to over-predict valve factor, for example, preliminary
PPM result for EPRI #14 is 0.674 compared to test value of 0.436. !
I
hidata. doc
-
O O O -
.
PREVIOUS VALVE CAPABILITY ,
_
Assumed Selected Equipment Torque Rate of Time Available
Valve PSID Line Valve Accuracy Switch Loading Related Valve
Pressure Factor (% Rand) Repeat (% Bias + Degrad Factor
(% Rand) % Rand) (% Bias)
FW-V-92A -580 580 0.42 9.45 5 (2.9+10.3) 0 0.64
~7%
FW-V-92B 580 580 0.42 9.45 5 (2.9+10.3) 0 0.61
~7%
RC-V-2 2367 2367 0.40 10.08 5 (2.9+10.3) 0 0.62
~7%
.
.
l
_1vfdata. doc
l
'
l
_
,
- - .-- .. . . .. .. . - . . . - .. . -. .. . .. ... ~ . .-. -. ..._~ .
l O O O' '
-
l
! CURRENT VALVE CAPABILITY .
l
+
'
Assumed Selected Equipment Torque Rate'of Time Available
Valve. PSID Line Valve Accuracy Switch Loading Related Valve
Pressure Factor (% Rand) Repeat (% Bias } Degrad Factor <
(% Rand) (% Bias)
FW-V-92A 580 980 0.464 9.5 5 15 2 0.487
FW-V-92B 580 980 0.464 7.9 5 15 2 0.468
.
RC-V-2 2155 2155 0.472* 10.1 5 15 2 0.587
.
-
- Calculations performed with Valve Factor = 0.500
Ivfdata. doc
.
__ _-___ ___r-2 _ _ _ --___- _u - _ - . + .- - - . - . - - - + . - . - . - -~_w_ -- ,.m.m, _ - .- .,._--w . # .
. - . - - . - - _ .
!*
i
,. t
O ;
MAIN STEAM VALVES
.
!
! ISSUE
1 .
. Design basis of the isolation valves in the TMI-l Main Steam System
TMI-I DESIGN BASIS
- * No main steam motor-operated isolation valves required to close to
mitigate design basis accidents or transients. !
!O - TMi-i uSLB mitigated by isoiation of feedwater
2 . Safety function for any isolation valves in main steam system
'
. Provide long term containment isolation accomplished by
remote manual (Control Room) operator action
e Containment peak accident pressure (LOCA) of 50.6 psig used
for MS-V-2 A/B design basis AP
EOP directs operator to close MS-V-2 A/B as a prudent action in
the event of an over-cooling transient - not a design basis
requirement.
- No design basis requirements for MS-V-8 A/B to close.
O
ivraata doc
,
.. - . . . - . - .. . .
.
.
'
O
'
MAIN STEAM VALVES
(Cont'd)
4
i
e Steam.Line Break Licensing Basis
,
,
e TMI-l FSAR Chapters 10,14,14A
i
,
e HELB Analyses
!
e TMI-l Restart Report -- NUREG 0680
4
. NRC Operating License SER l
- O
SUMMARY
9
- * Normally open valves-receive no ES automatic closure signal
,
e MS-V-2 A/B in program for long term containment isolation
following MSLB inside containment.
-
.
4
O
- .
wam e
.
=--
- . - _ _
.
.
O -
MOTOR OPERATED VALVE (MOV)
INDEPENDENT REVIEW
MOV Independent Review Team (IRT) formed week of July 1,1996.
Mission
.
O Review the GPU Nuclear MOV Program including its
,
implementation to assess its compliance with NRC requirements
and effectiveness.
.'
- O
4
Scope
This independent review includes, but is not limited to, the following:
i
,
O The specific concerns raised by the NRC, GPU Nuclear's response
, and associated corrective actions.
i
0 Specific concerns identified by the GPU Nuclear self-assessment
i
process and associate corrective actions.
"
O Any additional concerns identified by the team.
'O
i
hidata. doc .
l
<
~
.
.
O
MEMBERSHIP
Mr. John Fornicola, GPU Nuclear Corporation - Team Leader
Mr. Paul Damerell, MPR Associates
Dr. Thomas Gerber, Structural Integrity Associates ,
Mr. John Hosler, EPRI
o
U Mr. David Lewis, Shaw, Pittman, Potts, & Trowbridge
Mr. Philip Moor, GPU Nuclear Corporation
l
Mr. Julian Nichols, MPR Associates
.
i
Mr. Dann Smith, GPU Nuclear Corporation
,
l - Mr.-Henry-Stone,-H.-E.--Stone-Inc.,-as needed.
O
i,e. m
i
. l
--
, . .. _ . . .__ _ _ _. _ . . _ _ . __ - - - .
- .
t
.
O -
l
REVIEW TEAM PLAN :
i
i
.
!
I e R.eview applicable NRC Requirements
j-
4
e Review applicable GPUN program and related documents
i
l
4
e Review GPUN Self-Assessment Report of the MOV Program
!
e Interview Pers nnel iny lved in the program
O
i
e Assess program adequacy
l
i- . Identify potential non-compliance l
l
l e Identify unaddressed problems / recommend corrective action I
i
2
e Recommend corrective action to prevent recurrence of similar
!
l problems
.
6
1
i
iO
,
4
9
, - . ... _. -- -- .- . . _
- , . - . . -
.
s
-
.
O .
! ACTIVITIES TO DATE
4
e
i
- = Developed Charter
e Reviewing requirements / documents
e Generic Letter 89-10 and Supplements
i
NSA Assessment Report 95-03
'
f e Review NRC Inspection Reports
TMI Pr gr m escription 89-10 MOV Program j
'O *
- e Requested / received several presentations from GPUN Staff
i e Interviews in progress
j e Requested NSA review of Engineering response to Self-Assessment
i Recommendations
. Requested documentation of program changes for review
i
l
.
Ivraata &c
'
.
- r
,-
..
.
O- -
PRELIMINARY OBSERVATIONS
. Self-Assessment was good in many areas, did not evaluate valve l
factor development in detail.
l
l
. Some open issues
!
- Gaps in program documentation
e Lack of detail in documenting certain valve factor selection and
engineeringjudgments
e
Breakdown in Communication / Teamwork between HQ and Site
Engineering
.
O
,-
.
e
___ -_
. . . .. - - - . . = - . - . . - . - - _ . . _ . . . - - - .
-
.
.
4
4
8 -
- n
lU '
1 1
REMAINING ACTIONS
,
- Review basis for operability of FW-V-92 A/B and RC-V-2
1
i
- Review basis for other valve factor changes
,
4
I * Review responses to NSA assessment recommendations ;
4
4
e
i e Complete interviews
1
.
'
!O * Review OC MOV Program
l
!
i
- Issue final independent review report
,
,
l
1
!'
i
i
I
i
!
!O
,
_
.
-...a~ - _ , _ _ _ _ _ . _ . _ _ ,,
,. - -. - - - -- .. - .- . -. .- . - - . . . - - . - .
~
l
SUMMARY
'
1
!
Valve Status !
,
e Upgraded valve factors for FW-V-92-A/B and RC-V-2 and ;
additional conservatisms for Rate of Loading and Aging :
Degradation have been included in the program.
.
'
. Magnitude of Group 6 & 9 valve factor changes indicates other
valve groups have sufficient margin to accommodate similar ]
changes. l
!
i
e Main steam isolation valve design basis confirmed. :
e Review t date indicates that program valves are capable of !
O -performing their safety function. ;
\
l
Additional Actions
-
- Detailed design review of entire program to be performed to ensure
l adequacy in margin for all program valves.
4
~
! * Continue to evaluate methods to increase valve margins where
- appropriate.
- -
- e Independent review of the historical and current MOV program is
- intended.to prov.ide-adequate assurance.that problems are resolved
.
and to recommend corrective actions to prevent recurrence.
!
!O
.
$
hidata. doc
_ __ - - .- - _ - - .. -
4
.
O
AGENDA
July 24,1996
A. Introduction '
T. G. Broughton
B. Engineering Integration Process
R. W. Keaten
.
C. Materials and Services Organization Restructuring
! J. Langenbach
. 1
,
j D. Summary
T. G. Broughton
-
.
i
5
O
reorg. doc .
i
.
.
.- ..
.
.
O
Engineering Integration Process
O '
.
1
i
.
O
I
'
reorg. doc
. . . . .
.
.
.
Initiative for Eneineerine Integration Process
!
C, Process Focus i
.
Process focus started in 1994 to identify and define core
business processes with the goal of improving the
efficiency and effectiveness of processes.
Review looked broadly across GPU Nuclear processes
>
supporting plant operation and looked more closely at the
- engineering and technical services processes.
- Q Key Elements
'
I
. Identify opportunities to streamline and standardize workflow l
without compromising quality.
. Evaluate process and end products against identified desired
'
characteristics and benchmark against other companies. )
l
- . Identify and eliminate duplicate and overlapping capabilities. !
e Evaluate various engineering activities.
. Technical Functions
'
. Plant Engineering
. Plant Operations Engineers
'
. Plant Maintenance Assessment
.O . Radiologicai Engineering
teorg. doc
,
- - _. _ - . ..
.
.
.
.O Engineering Integration Process
Engineering Integration Team
4
i
. Need for integration was emphasized in revised
4 organizational principles forged by Senior Staff in
'
February 1996.
. . Engineering Integration Team (EIT) constituted
! March 1996.
I
'
. Identify and eliminate unneeded engineering-related work.
. Simplify and streamline necessary work.
- O Objectives:
4
.
'
. Focus on revised organization principle which
! commits to a strong integrated engineering function
!
! . Continue to provide day-to-day engineering support for
Operations and Maintenance.
. Continue to provide Long Range Plant System, and
Component Engineering Planning and Development.
I
.
. Reduce nuclear plant operating costs
'
reorg. doc
. ----- _ - - -- - -
1
.
.
O Mission statement
Considering the challenges of deregulation, broadly
examine the engineering and technical activities
performed in support of Oyster Creek and TMI.
Recommend changes to process, work activities and
structure which will reduce cost and align resources in :
a seamless manner, while maintaining high levels of 4
quality and safety. ;
l
l
l
O
l
O
'
reorg. doc
_ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _
_ - _ _ - . . .
.
.
Engineering Integration Process
O Approach
. Focus Groups were conducted to identify opportunities
and critical success factors to be used in restructuring
engineering.
. Key stake-holders were interviewed to confirm needs
and expectations.
. Electronic Mail system ~was set up to receive comments
and suggestions. !
. Two EITs were established to analyze engineering work
and map the work flow.
. Functional Methodology
. Process Methodology
. After internal reviews were completed, the teams visited
low cost /high performance utilities (SALP Ratings) to
benchmark with their engineering organizations.
O l
-
reorg doc .
,
-
.]
INPO facilitated development of Processes under industry's '
Stranetic Plan for Building New Nuclear Power Plants. I
,
Electricity
JL
Station
l
Operations
JL
Materials & Work Configuration
Services Control Control
JL
Equipment
Reliability
JL JL
Administrative
Training
Support
--
-- _-
__ _ _ - _ _ _ _ _ _ _ _ _ - - - - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - - - _ - _ _ _ - _ _ _ _ -
m
.
.
En_eineerine Inte_gration Team Results
_
After all internal and external reviews were
completed, the two sub-teams compared results - the
process evaluation was found to give the most useful
insights.
The team unanimously recommended that:
. All engineering functions be integrated into a new
Engineering Division.
O
. The new division be organized along the lines of
the industry's Equipment Reliability and
Configuration Control processes.
. Functional groupings be used as appropriate.
,
. The engineering " center of gravity" be shifted to
! the plant sites.
1
- . Some engineering functions remain centralized.
.
O
'
reorg. doc
r'; .
.
O Equipment Reliability (ER) Process
Key Features and New Ideas
. ER Process consolidated under single organization at
,
each site which includes System Engineers,
Component Engineers, and ER Program Owners.
'
. System Engineers become System " Owners" - provide
single point accountability for all system issues.
,
System Engineers:
O .
. Maintain existing System Engineer accountabilities.
. Resolve long-term and short-term issues affecting the
reliable performance of the system.
. Assume ownership of P. M. Program technical content for
their systems.
. Assume ownership of on-line maintenance work tecimical
content.
. Perform SU&T functions on their systems.
. Coordinate and chair system performance team.
. System Engineers draw on Component Engineers, site
based Design Engineers, and Corporate based
Specialists to resoive issues. Centers of Excellence
O maintained.
reorg. doc
,_ .- _ _- . .- .- . ._ -
.
.
O configuration controi (cc) Process
'
Key Features and New Ideas
. Most modifications performed by site based Design
Organization
. Design & Project Management Centers of Excellence
maintained on site.
. Fewer plant modifications and reduced engineering worldoad
have resulted in reduced expenditures.
O
'
. Configuration Maintenance group supports procurement
and document control activities.
i
l
l
l
i
4
O
- reorg. doc
- _ .. - _ _ _ _ - . _ _
_
GPU NUCLEAR CAPITAL SPENDING
($ IN MILLIONS)
'
$120
$100
-
w
$60 Budget
-$40
Forecast
.
$20
$0 ~ ~--i ---' - + - - - - - - * .
1990 1991 1992 1993 1994 1995 1996 1997-
_ _ _ . _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ - -_ _ _ _ _ _ _ . _ _ . _ _ .
. .
_
. _ _ . _
.
.
Center of Excellence
O
Objective:
Retain control of proficiency and assurance of work
standards.
- Accountabilities will be assigned to appropriate
departments / personnel to ensure processes and procedures meet all
required technical and regulatory standards.
- Personnel who perform work with these processes and procedures
are trained, qualified, and perform the technical and regulatory
work standards.
O l
.v
1
EXAMPLE: Startup and Test sub-process
l
e Current Start-up and Test function will be included in the
Equipment Reliability Process, and System Engineering sub-
process.
- System Engineering Managers will own Start-up and test process !
,
and assure work is performed.
t
- Qualified and certified Start-up and Test Engineers will perform the
work.
4
d
- O
.
reorg. doc
,-- - - - - - - - - - - - - - - - - - - -_ _
--
.
.
.
Selection Process for En_eineerine Division _
. Key importance in selection process is GPU Nuclear's
need to maintain excellence in its primary focus on
nuclear safety, power production, and business acumen.
l
l
. Senior Level Selection Committee established to select )
new Directors.
l
l
1
'
. Targeted Selection Process focused on " dimensions" -
The key characteristics required for leaders in the new
,O organization. l
l
. Selection based on:
i
. Results of interviews using behavior-based questions for each
dimension.
-
Education and Experience.
-
Prior Performance.
. New Directors met to finalize organization structure and
begin selection process for managers.
O
reorg. doc
- _ _ - _ _ _ _ _ - _ - _ _ _ _ - _ .
.- -_
,
. _. -
-
.
e
Selection Process for Engineering Divisio.n
O continued
. Individual managers were selected utilizing a Selection
Conunittee and the Targeted Selection Process.
o Selected managers met to finalize initial staffing levels
and plan remaining selection activities.
o Individual contributors are being reselected by selection
committees based on appropriate criteria including key
O denaviorai dimensions. ,
!
o Engineering Division staffmg is based on retaining the
best qualified personnel.
-
l
4
i
4
i
teorg. doc
. .
,
O, O
-
-
O.
Engineering Process Management Structure
Engineering
Process Owner
Bob Keaten
l l l l
. ')irector Director Director Director
'
.
Engineering Configuration Engineering
Reliability Control " "# '
Chemistry-
OC - Art Rone OC - Dave Slear
Support "[ ""
Materials
TMI- Pat Walsh TMI- Dick Skillman Don Croneberger Scott Giacobbe
Manager Manager
3 Managers System Mechanical / Mechanical / Manager Safety & Manager
Structural
-
Owners
-
-
Structural -
Risk Analyses
~
Engineering NDE/ISI
Engineering
_
Manager _ Manager EP&l _ Manager EP&l _
Manager Nuclear _
Manar,er
Components Engineering Engineering Fuels Chemistry
.
Manager
_
Manager ER _
Manager Components & _
Manager License Manager
_ ~
Programs Modifications Programs Renewal Materials
!
Manager Manager
_ Manager Shift _ Configuration _ Manager Projects -
Environmental &
Engineers Maintenance ChemistrySupport
Manager
Manager Process Design & Drafting Decontamination &
- _
Manager
-
Computers Decommissioning
__ ._. - _ - - _ - _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ - - _ _ - _ _ _ _ _ - _ _ - _ _ _ _ _ -
. . . _ _ _ . - - - - . _ ..
.
Engineering Division Staffing Reductions
O
. Approximate 20% staff reduction overall
l . Engineering totals today approx. 480
!
. New Engineering Division approx. 390
.
!
,
I
,
!
!
!
1
i
!
l
4
4
.
reorg doc
.
.
O Transition Plan
Objective:
Effect a complete, orderly, and systematic transition from
the current Engineering Organizations to the new
Engineering Division.
- Transition team formed to implement new GPU Nuclear
Engineering Organization.
O e Team is comprised of internal experts who have the lead
for key activities:
- Procedures and processes
l - Budget and cost
- Training
'
- Information Tecimology and Data Systems
- Process Mapping and Center of Excellence
! - Facilities
- - Nuclear Safety Assessment, Self-Assessment, and Process
Measures
'
- Organizational Document Revision l
- Regulatory Compliance Documents
I
O
1
'
- '
reorg. doc
1
-
.. 1
I
l
.
O Transition rian
Continued
e A certification process will be used to track, document,
validate, and certify all work has been accomplished.
1
- Process is similar to that used during the TMI Unit 2 merge l
with TMI-1; The Site Services Division merge into the Oyster l
Creek Division; and Restart Certification from plant refueling
outages.
,
- A database has been developed to inventory all work tasks i
b*i"8 performed by current Engineering personnel. !
O
- Work will be reassigned to appropriate organizations and
individuals.
!
i ,
- Discontinued work will be justified and documented. !
l
'
.
- e NSA will verify all Operational Quality Assurance Plan
'
requirements are met.
lO
,
$
4
'
rearg. doc
l
l
. _ _ _ _. _ _ _.__
- . . -
.
.
Self-Assessment Engineering Organization
4
O
.
\
l e Importance of effective self-assessment program
i
recognized and included in Corporate 1996 Safety Goals.
- e Company-wide standard under development.
'
- Includes departmental assessment and independent
oversight aspects.
!
i :
!O e Engineering Transition Team will develop specific
program for Engineering
l .
i
e Product to be a model for other departments. ,
,
.
>
'
reorg. doc
.. -_. . -
<-
!
, .
- O Implementation of Integrated Engineering
Organization
l
l . Completion of the new process organization is targeted for
'
early August 1996.
-
. Goal is full implementation of the new organization before
- the OCNGS outage begins in September 1996. l
l
1
,
O
-
reorg. doc
_ _ _ _ _ _ _ _ _ _ _ . __
.. . .- - - . - . - . _ . - - _ ._. - -
-
.
e
.
O .
t
Materials and Services Organization
i
i
Restructuring
O
1
O
reorg. doc
-- - . ._
.
.
Materials Management Reorganization
O to Materiais and services
e Reasons for Reorganization
e Declining Workload
. Fewer Capital Modifications !
. Longer Tenn Agreements (Partnering)
. Leveled Workload (More work done while operating)
l
e Materials Management Process Reengineering
Recommendations
. Process Focus
. Broader Span of Control
O
e Low Risk
. Minimal Nuclear Safety Implications
,
. Minimal Operational Impact Potential
! . Nuclear Safety Assessment oversight
,
- e Future efficiencies expected to be gained from process focus
!
! = Initial changes - Approximately 20 fewer positions out of 100
'
full time employees
'
. Went from 16 to 8 managers and supervisors
. Average span of control from 6 to 10
,
4
O
T
'
reorg. doc ,
_ . . _ . . . . . _ _ _ . . _ _ _ . _ . . _ .
i
O O o-
i
GPU NUCLEAR PURCHASE ORDER
AND CHANGE ORDER SUMMARY
, 2s,000 --
1
e 24,000
4
^
"
22,000 -
.
'
a
b
O
13 20,000
5
.=
@ 18,000 -
. t6
,
j .16,000
E
Z
14,000
.
12,000
10,000 -
_
,_
1990 1991 1992 1993 1994 1995
)
_ _ _________-_ ______________ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ - _ _ - . _ - - - . _ _ _ _ _ _ __. - .__. __ .
.
r .
~
.
1 ;I
4
- f
'
.
J
i
TMI & OYSTER CREEK NEW STOCK SYMBOL NUMBERS
?
8
9,000
j !
, ,
1
4
.
8,000
-
,
i
! 7,000
l
'!
6,000
-
- -
I
5,000
i
i !
i
- 4,000 *
.
!
e
3,000
i
t
2,000 .
.
i
.
1,000 I
i
0-- - - * - - - --
- - - + - - - - - - - -
,
1991 1992 1993 1994 1995
!
i
i
i
_ _ _ . . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ ___._______.___.__.____._.)
.- - - - - - . - _ . . . - . . - - - --
,
.
Summarv ~
4O
i . New Engineering Support and Materials & Services organizations are
j based on Process Model Methodologies.
!
- . These models provide more cost-effective support organizations while
-
maintaining primary focus on nuclear safety and safe plant operation.
. Elimination of low value, redundant work activities and organizational
J overlap as well as increased spans of control, single points of
{ accountability, and staffing at levelized requirements achieve substantial
i
efficiencies.
. Staff reductions supported by declining work loads and reduced plant
i modifications. l
. Nuclear Safety Oversight Groups will monitor the implementation of the
- O
- new organization.
!
!
. Strengthened and formalized the self-assessment program to monitor
i work quality and other performance measures. ,
i l
l
. Transition program established.
.
. Process review for other areas will begin later this year.
- . Maintenance / Work control l
. Operations
e Encourage NRC feedback
i
k
I
1
'
reorg. doc
1
- - .__ _ ____ _ _