ML20195J724
| ML20195J724 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 01/04/1988 |
| From: | Dyer J, Harper J, Haughney C, Howell A, Isom J, Norrholm L, Sharkey J NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD), Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20195J705 | List: |
| References | |
| 50-312-87-29, NUDOCS 8801280686 | |
| Download: ML20195J724 (41) | |
See also: IR 05000312/1987029
Text
. __ ._ _ _ _ _ _ - _ _ - _ _ ..
l
- .
l
'
.
ENCLOSURE 2
U.S. NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
Division of Reactor Inspection and Safeguards
Report No.: 50-312/87-29
Licensee: Sacramento Municipal Utility District
P.O. Box 15830
Sacramento, California 95812
Docket No.: 50-312
Facility Name: Rancho Seco Nuclear Generating Station
Inspection Conducted: September 28, 1987 - October 9, 1987
Inspectors: 8thsW [4[8%
Date Signed
- J. E. Byer, Team Leader, NRR
k *J.364w
d4/ss
C. Har er, Metallurgical Engineer, NRR Date Signed
.
1/t,tl88
A.i.jfowell,Rea@torOperationsEngineer,AEOD Dhte' Signed
,'Yb0 misn - /*b/ff7
(,*J. A. Fsom Reactor Operations Engineer, NRR Date Signed
/ }l /2/3/lC7
/,*J) M./Sharkey', @ctor Operations Engineer, NRl Date Signed
Consultants: *G. Morris, WESTEC; *D. Prevatte, WESTEC
4
Accompanying Personnel: L. Miller, RV; R. Zimerman, RV; *D. Kirsch, RV;
- Crews, RV; *C. Myers, RV, *G. Perez, RV; ,
- A D'Angelo, RV; *D Baxter, EG&G; *M. Johnson, OED0 1
Reviewed By: i. M A /
L.' U Norrholm, Chief, Team Inspection Appraisal Date' Signed
and Development S tion #1, NRR
Approved By: v 1A _
l
- C Haughney Chief 1 Inspection Branch, DttetSigned
- Attended Exit Meeting on October 9, 1987
8801280686 880125
PDR ADOCK 05000312
G PDR
- - _ _ _ _ _ _ .
-
. .
b
.
Scope:
An NRC headquarters team performed a special, announced inspectior to examine
the open and unresolved items from NRC's Augmented Systems Review and Test
Program (ASRTP) Inspection (50-312/86-41), assess the adequacy of the
licensee's expanded ASRTP (EASRTP) inspections, and evaluate the effectiveness
of the licensee's Engineering Action Plan (EAP) for improving the q'Jality of
engineering analyses. Additionally, the inspection team reviewed the
licensee's program for purchasing and controlling safety-related fasteners.
Results:
The inspection team closed 21 of the 44 open and unresolved itemt identified in
ASRTPinspection(50-312/86-41) dealing primarily with engineering, management,
and quality assurance areas. The remaining items concerning maiittenance,
operations, and surveillance testing will be reviewed during a fJture
inspection. The EAP appeared to improve the quality of design activities at
the station. The EASRTP inspections had been conducted in a manner comparable
with the NRC ASRTP inspection and identified significant safety issues. The
licensee's program for purchasing and controlling safety-related fasteners
appeared to have significant deficiencies; the NRC will review this program
again before restart. During this inspection, six new open items were
identified,
11
. o
m
.
TABLE OF CONTENTS
PAGE
1 INSPECTION OBJECTIVES .................................... 1
l
l
1
2 STATUS OF PREVIOUSLY IDENTIFIED ITEMS .................... 2
2.1 Items Closed During the Inspection ....................... 2
2.2 Items Remaining Open After the Inspection ................ 9
3 DETAILED INSPECTION FINDINGS ............................. 17
3.1 Evaluation of Expanded ASRTP Inspection Program . .. . ... . .. 17
3.1.1 Revi ew of EASRTP Inspection Fi ndings . . . . . . . . . . . . . . . . . . . . . 17
3.1.2 Review of EASRTP Methodology ............................. 18
3.1.3 Comparative Inspection Results ........................... 20
3.2 Assessment of Engineering Action Plan .................... 21
3.2.1 Review of EAP Design Change Control Procedures ........... 22
3.2.2 R e v i ew o f EAP Wo r k P ro du c t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22
3.2.3 EAP Independent System Reviews ... ....................... 25
3.3 Procurement and Control of Fasteners ..................... 26
3.3.1 Investigation of Counterfeit Fasteners ................... 26
3.3.2 Warehouse Material Controls .............................. 28
3.3.3 Purchase Order Review .................................... 28
3.3.4 Licensee-Identified Procurement Problems ................. 29
4 MANAGEMENT EXIT MEETING .................................. 31
Appendix A PERSONNEL CONTACTED
Appendix B DOCUMENTS REVIEWED
1
l
l
l
l
iii
_
- - . _.
l
t
.
n
'
1 INSPECTION OBJECTIVES
The objectives of this inspection were to examine the corrective actions
taken by the Sacramento Municipal Utilities District (SMUD) as a result of the J
NRC's Augmented System Review and Test Program (ASRTP) inspection conducted at
.
1
Rancho Seco Nuclear Generating Station in February 1987. This effort included:
j
(1) a review of the specific open and unresolved items identified during the
ASRTP inspection; (2) an evaluation of the licensee's Engineering Action Plan
developed to improve the quality of ongoing engineering analyses and calcula-
tions, and (3) an assessment of the licensee's completed pnrtions of the
Expanded ASRTP (EASRTP) inspections performed on 33 safety-related systems to
better ensure safe design and operation. Additionally, the team reviewed
SMUD's program for procuring and controlling safety-related fasteners.
2 STATUS OF PREVIOUSLY IDENTIFIED ITEMS
'
ihe inspection team reviewed the status of the open and unresolved items
identified during its ASRTP inspection (50-312/86-41). In general, those items
associated with engineering, quality assurance, and management areas were
closed, but items concerning maintenance, operations, and surveillance testing
remained open. These open items will be reviewed during future inspections.
2.1 Items Closed During the Inspection
Open Item 50-312/86-41-02: Auxilibry Feedwater Flow Above the Maximum
Design Rate
Problem 54 of the AFW (auxiliary feedwater) System Status Report (SSR) stated
that AFW flow to a once-through steam generator (OTSG) could exceed the maximum
design rate of 1800 gpm. The team disagreed with the licensee's plans to
correct this problem after restart.
The licensee installed a cavitating venturi in each OSTG line to limit flow to
1000 gpm. The team reviewed the design change package and found no deficien-
cies. This item is closed.
Open Item 50-312/86-41-03: AFW Full Flow Test Line Instrumentation
Problem 3 of the AFW SSR stated that the instrumentation on the AFW full-flow
test line was inaccurate because downstream piping was subjected to condenser
vacuum. This deficiency contributed to further problems with measuring AFW
pump capacity using alternate methods as described in unresolved item
50-312/86-41-28. At the ASRTP inspection exit meeting, the licensee committed
to correct this problem before restart.
The licensee installed a restricting orifice downstream of the flow element to
increase back pressure. The element was resized accordingly and sensing lines
were redesigned to eliminate air entrapment. The team reviewed the design
change packages for these modifications and found no deficiencies. This item
is closed.
Open Item 50-312/86-41-04: AFW Pump Runout Damage
Problem 43 of the AFW SSR outlined the resolution for verifying that the AFW
pumps had not been danaged by the pump runout condition that occurred during
-1-
. - - ,
.
. .
e
'
the December 26, 1985 transient. The licensee had not planned to contact the
vendor to verify that the internal components of the pump had not been damaged.
The licensee subsequently contacted the pump marufacturer who identified which
internal components should be checked or replaced. The licensee completed the
necessary work. This item is closed.
Open Item 50-312/86-41-05: AFW Pump Performance Calculation
Calculation Z-FWS-M2081, performed in response to AFW SSR Problem 55, was found
to have an error that would have identified an incorrect and nonconservative
acceptance value for pump performance testing.
The licensee replaced this calculation with Z-FWS-H2237. The team reviewed this
calculation and found no discrepancies. This item is closed.
Unresolved Item 50-312/86-41-06: Instrument Air System Design For AFW
Flow Control Valves
Several deficiencies were identified in a proposed modification to install
backup instrument air bottles to supply air to AFW flow control valves. The
licensee's resolution for these items is as follows:
(1) Inadequate Valve Actuator Overpressure Protection: The pressure relief
setpoint was reduced to less than the design pressure of the main and
startup feedwater contrt' valve actuator diaphragms. Additionally, an
airset was installed in v a air supply to the diaphragms providing addi-
tional protection.
(2) Lack of Seismic Qualification of Air Valves: The licensee performed
calculation Z-ZZZ-C0863 to confirm the seismic qualification of the
control valves, excess flow valves and adjustable check valves for the
backup air system.
(3) Incorrect Check Valve Design: The excess flow check valves were replaced
with adjustable check valves which were better suited to the design
application in the backup air system.
(4) Monitoring of Backup Air Supply Pressure: Operating logs were issued
requiring the recording of bottle pressures on a daily basis. However,
the acceptable pressures identified on the log sheets were at the 2 and
3-hour alarm points for the bottles. These pressures were well below the
minimum specified by Engineering in the D3 sign Basis Report for the
modifications. The licensee initiated a revision to the operating logs to
establish new minimum bottle pressures which will accomplish the intended
purpose.
(5) Backup Air Supply Test Procedure: The licensee developed a periodic test
procedure for the backup air system. At the time of the inspection, the
procedure was being reviewed for approva*.
.
(6) Incorrect Design of the Pressure Control Valve: The valve manufacturer
certified that the valve would give the tight zero demand performance
desired for the system. The combinatior, of the manufacturer's statement,
the proposed testing, and the daily logging of pressure convinced the
-2-
__
- _ -
- _
1
_
. .
,
!
.
-
inspection team that reasonable assurance of pressure control valve
performance will be provided.
(7) Incorrect Fabrication Drawings: The drawings were changed to show the
proper orientation of the valves.
The team concluded that the above actions were appropriate. This item is closed.
Open Item 50-312/86-41-08: Environmental Qualification of Emergency Feedwater i
Initiation and Control (EFIC) System Excess Flow Valves. 1
The excess flow valves used in the instrument lines for OSTG 1evel instruments !
installed as a part of the EFIC system modifications were not environmentally !
qualified. I
J
The licensee's analysis of calculation Z-EQP-E0689 qualified the critical
components (0-rings) for 10 years. However, the basis for the acceptance
criteria for thermal aging in a dynamic application was not included in the
analysis, and the 10 year replacement requirement for the 0-rings had not been
included into the appropriate maintenance procedures. During the inspection,
the licensee revised the calculation to include the basis for the acceptance
criteria and inserted the requirement for replacement of the 0-rings into the
appropriate maintenance procedures. This item is closed.
Unresolved Item 86-41-09: EFIC System Single Failure Susceptibility
l
The EFIC system sensing instrumentation was designed so that certain failures
'
would activate unnecessary protective systems. This susceptibility appeared to
be contrary to the updated safety analysis report and had not been reported to
the NRC as is required by 10 CFR 50.59.
The licensee revised the safety analysis for the EFIC system modification to
include responses that could be expected with spurious initiation. None of the
responses were outside the design bases of the plant. This item is closed for
the inspection report purposes. The NRC is currently reviewing the adequacy of
this design as part of the safety evaluation report (SER) for plant startup.
Open Item 50-312/86-41-10: Maintenance Bypass Testing of the EFIC System
The proposed test for the EFIC system did not verify that whenever any EFIC
channel was placed in the maintenance bypass position, the remaining channels
would not be inhibited or degraded.
The inspection team reviewed the revised special test procedure (STP) 666,
"EFIC Cold Functional Test," and determined that the maintenance bypass feature
was adequately tested. This item is closed.
Unresolved Item 50-312/86-41-15: Condensate Storage Tank Pressure Relief and
Vacuum Protection
The setpoint tolerances for the condensate storage tank (CST) pressure relief
valves were such that the valves could be set above the design pressure for the
tank (setpoint as high as 2.5 psig vs. 2.0 psig design pressure).
Additionally, the CST was designed for a 1.0-inch water vacuum, but a
-3-
. -
.-_ _ _
,
. .
l -
calculation in the file indicated that this condition could be exceeded without
assuming a single-valve failure.
The licensee changed the relief valve setpoint tolerances to ensure that the
design pressure would not be exceeded. Additionally, an analysis (Z-MCM-M2212)
was performed that showed the tank capable of withr, Landing 5.0 psig pressure
and 1.74 inches water vacuum. This analysis also snowed that for the maximum
outflow with one vacuum 'reaker failing to open, thi!, new design value for
vacuum would not be exceeded. This item is closed.
Unresolved Item 50-312/86-41-16: DC System Short-Circuit Calculations
Inconsistencies were identified with the calculation of the battery's contri-
bution to a short circuit. This matter included inconsistencies in the
manufacturer's data referenced in, and attached to, :he calculation. The
inspection team's rough calculations indicated that ;ome circuit breakers may
not have been properly sized for the de system.
The licensee replaced existing de cables with smallec cables to reduce short-
circuit current and revised the calculation (Z-DCS-E)612) to incorporate the
team's comments. Additionally, the team also verified that the design inputs i
for the smaller cable, added to limit the short circuit current, were correctly
factored back into the new voltige calculations. This item is closed.
Unresolved Item 50-312/86-41-17: Battery Sizing Calculation
The staff was concerned about the load profile and ninimum design temperature
used for sizing the new batteries installed in the ruclear services electrical
building (NSEB).
The licensee revised the calculation (Z-DCS-E0636) t.o respond to the team's l
concerns. The team reviewed this revised calculation and found no deficien-
'
cies. This item is closed.
Open Item 50-312/86-41-18: Drawing E-101 Error
A discrepancy between calculation Z-EDS-E0076 and the main single line diagram
E-101 was identified. The team confirmed that the drawing was wrong, not the
calculation.
The licensee issued a drawing change notice (DCN) to correct the drawing. The
DCN was reviewed by the team for the identified concerns and found acceptable.
This item is closed.
1
Unresolved Item 50-312/86-41-19: AC Short Circuit Calculatio_n '
An apparent discrepancy was discovered between the startup transformer
impedance given on the unit's nameplate and that derived from the unit's test
data. The licensee could not immediately confirm which value was correct.
The licensee revised its calculation (Z-EDS-0120) using the more conservative
value and demonstrated that the design was adequar.e. The discrepancy between
the nameplate and test data impedances wa,s resolved by the vendor's statement
that the European nameplate information contains design data only. This item
is closed.
-4-
- .
.._ .- .- . _. - _ ,_ - - .-
- _ _ _ _ _ _ __
. a
-
Unresolved Item 50-312/86-41-20: Battery Charger Cable Size
The ampacity of the d: power cable from battery charger H4BAC to de bus SOC
would not meet the commitment to size power cables at 125 percent of the full
load current that was made in the "Rancho Seco Nuclear Generating Station
Updated Safety Analysis Report" (USAR).
The licensee respondtd with calculation Z-EDS-E0744 and proposed administrative
procedure changes that would justify operating this particular cable in its
overload region for a limited number of times over the life of the plant. The
response sufficiently justified the overload condition of the cable for this
specific cable. The response did not address the USAR commitment to size cable
at 125 percent of the full load current. However, because this cable is in a
dedicated conduit and its use will be administratively controlled, the team
found the ampacity of cable acceptable for limited use. This item is closed.
Unresolved Item 50-312/86-41-21: IDADS Alarm Interface Problems
The required time delay on pump startup for the AFW pump runout alarm had not
been provided either by hardware or software design. Additionally, the team
found that the voltage for the NSEB de system buses was being monitored incor-
rectly by the Interim Data Acquisition and Display System (IDADS) as digital
inputs instead of aralog inputs.
The licensee made scftware changes to the IDADS alarn for the AFW pump runout
alarm and issued Engineering Change Notice (ECN) R151 to modify the inptt
circuit hardware for de bus voltage. The team reviewed these modifications and
found no discrepancies. This item is closed.
Open Item 50-312/86-41-22: Motor Operated Valves Overload Protection
No thermal overload protection or overload alarms existed for the safety-
related motor opera;ed valves (MOVs). The licensee's response during the
initial inspection was that it was more important that a safety-related MOV
fail attempting to perform its safety function than to trip because of an
erroneous response from an overload relay. No further work was performed in
this area because tie licensee planned to resolve this problem after restart.
The team agrees that this was not a restart item that needed immediate atten-
tion. The team was concerned, however, about undetected long-term degradation
of the motor's insulation and the resultant increased probability for valve
failure. This iter is closed.
Open Item 50-312/86-41-23: Safety Evaluations
The licensee's procedures for conducting safety evaluations in accordance with
10 CFR 50.59 lacked sufficient guidance for conducting the evaluations; also
there were no qualification requirements for personnel performing the
evaluations.
The licensee issued revised procedure RSAP-0901, "Safety Review of Proposed
Changes, Tests, and Experiments," for conducting safety-evaluation required by
10 CFR 50.59 and implemented a training program to qualify personnel performing
the evaluations. This item is closed.
-5-
l
_ - _ -__-
_-
.
l
)
.
Unresolved Item 50-312/86-41-24: Drawing Control
Deficiencies were identified with the procedure for and implementation of a )
1
drawing control program.
The licensee developed and implemented administrative procedures RSAP-0503,
i
"Design Change Document Control," and RSAP-0505, "Site Document Control (SDC)
Distribution Control," for drawing control. Additionally, the licensee {
conducted a thorough audit of their drawing control program to verify imple- l
mentation of the new procedures. The team verified that the specific deficien-
cies identified in inspection report 50-312/86-41 were corrected.
Also, a random sample of approxinately 20 controlled drawings were
reviewed in four locations, with no noted discrepancies. This item is closed.
Unresolved Item 50-312/86-41-25: Engineering Calculations
Several programmatic deficiencies were found with calculatinns performed by the
licensee in support of system design. As a result, the licen;ee implemented an
Engineering Action Plan (EAP) to improve all calculations. The EAP evaluation
is discussed in Section 3.2 of this report. This item is closed.
Unresolved Item 50-312/86-41-26: Expired Inservice Test (IST) Program
The first 10-year Inservice Test (IST) Program that the licensee submitted had
apparently expired on April 6, 1986 and no extension or new program had been
submitted to the NRC for approval.
On June 25, 1987, the licensee submitted its second 10-year IST Program to the l
NRC for approval. The NRC is currently reviewing this program. This item is l'
closed.
Unresolved Item 50-312/ 86-41-27: Pump and Valve Test Data Trending Program
The licensee was not entering test data into the Inservice Inspection Log used i'
for trending test data, although testing was conducted.
1
'
The inspection team reviewed the licensee's new program for trending IST data
which included both a data log and a graphic trending program. These appeared
adequate. Additionally, the licensee is planning to implement a computer aided
trending program after restart. This item is closed.
Open Item 50-312/86-41-29: Inconsistent Stroke Times for AFW Flow Control
Valves
Identical AFW flow control valves FV-20527 and FV-20528 had inconsistent stroke
time acceptance values. There was no technical justification for the different
times since these valves appeared to be identical valves in identical
applications.
The licensee verified that the acceptance values should be the same. A special
test procedure was being developed to establish consistent acceptance values
during AFW functional testing. This item is closed.
-6-
. _. _ _
.
.
.
Unresolved Item 50-312/86-41-32: Systems and Equipment
Numerous examples of administrative deficiencies were found involving the safe
clearance tag system, the abnormal tag system, and the Information Sticker Log.
It appeared that operations personnel did not always pay sufficient attention
to detail to ensure adequate control of plant and system status.
The licensee issued Special Order 87-19, "Operations Audits," that discusses
the Operations Department requirements for auditing each of the above
referenced tagging systems. The team reviewed the administrative control
systems and verified that the previously identified weaknesses were corrected.
The team conducted a detailed evaluation of the Information Sticker Log and the
Abnormal Tag Log and noted no deficiencies. Additionally, the licensee had
recently implemented Procedure AP.90, "Work and Test Authorization (WATA)
Program " Revision 1, that documented the authorization of work and testing
that affected the operational status of components and system; tracked the
operational status of components and systems; and prevented removing a
Technical Specification required system or component from service while the
redundant train was inoperable. The team concluded that the implementation of
the WATA program will significantly improve the ability of the Operations
Department to effectively control and track equipment status. This item is
closed.
Open Item 50-312/86-41-35: Operator Training Materials
System training manuals were neither controlled nor maintained current by the
Training Department. Uncontrolled copies of these training manuals were found
in the control room.
The licensee placed an index manual with the four volumes of system training
manuals in the control room. This index featured a list of working and c>m-
pleted ECNs sorted by system. This list alerted the user to other documents
that could modify the system. Additionally, the Operations Department it, sued
procedure AP-23.06, "Operation Procedures," Revision 0, which listed spe:ific
station procedures that were approved for use and prohibited the use of system
training manuals in lieu of approved procedures. Finally, the licensee planned
to revise the system training manuals in 1988.
The ASRTP inspection team was also concerned that several operating and
casualty procedure revisions were too complex to be adequately addrersed by the
Operator Reading Assignment Program.
The licensee improved its scheduled training for the new procedurer,. The team
reviewed the procedure revisions and the Training Department's methods for
selecting procedures for training. On the basis of this review, the team
concluded that the degree of procedure training fer each new pro:edure could be
appropriately addressed by the proposed classroom training, simulator training,
procedure walkthrough, or the Operator Reading Assignment Program. The inspec-
tion team did not review the technical adequacy of the planned training for the
new procedures. This item is closed.
Open Item 50-312/86-41-37: Maintenance Trending Program
The licensee had not implemented its maintenance trending program as described
in procedure AP-650, "Preventive Maintenance Program," Revision 5.
-7-
. .
.
.
Maintenance Administrative Procedure MAP-009, "Preventive Maintenance Program,"
was issued and superseded AP.650. The licensee subsequently implemented the
maintenance trending program described in MAP-009 that included, in part, a
vibration monitoring program for rotating equipment. Although only recently
implemented, there was objective evidence that the program was effectively
administered. For example, the licensee had projected the iminent failure of ;
a service water pump (P-9768) thrust bearing on the basis of the results of
routine vibration testing. The licensee was able to replace the pump bearing
before a catastrophic failure occurred. The licensee planned to implement the
use of trending techniques in other areas of the predictive maintenance
programs using recently acquired computer software programs. This item is
closed.
Open Item 50-312/86-41-38: Quality Assurance (QA) Audit Program Deficiencies
Significant deficiencies were found with the scheduling, performance,
responses, followup, and Management Safety Review Committee (MSRC) oversight of
the QA audit program.
The licensee reorganized the quality assurance (QA) organization, providing new
emphasis and direction to the audit program. The team reviewed the improved QA
audit program and found that the licensee currently had no audits overdue. In
addition, a requirement was added to Procedure QAIP-1, "Quality Assurance
Audit Procedure" to issue audits within 30 days after the inspections were com-
pleted. All audit findings were addressed to the department manager of the
area reviewed and a response was required within 30 days after the report was
issued. Also, at the time the responsible department replied to the audit '
finding, a completion date was submitted for QA concurrence. It was the job of
the QA organization to verify that the responsible organization had taken the
appropriate corrective action. The audit finding was then tracked using the
Management Sumary Report that highlights both overdue audit responses and
corrective actions. This report was distributed monthly to the Chief Executive
Officer, Nuclear and to both Assistant General Managers. Lastly, the licensee
had formed an MSRC Quality Oversight Subcommittee to advise the MSRC on matters
related to the QA program. Specifically, the subcomittee reviewed audit
program findings and closure reports to ensure the effectiveness and timeliness
of the QA program. The subcomittee was required to report the results of its
review to the MSRC. This item is closed.
Open Item 50-312/86-41-40: Corrective Action Programs
Deficiencies were identified with the trending program for nonconformances, J
numerous outstanding NRC open or unresolved items, and lack of a program that l
allowed significant conditions adverse to quality to be brought to the atten-
tion of the appropriate level of management.
1
'
The licensee implemented procedure QAIP-16. "Trend Analysis Program," Revision
0, which related deficiencies to a common cause, identified recurring problems,
and defined the extent and severity of the problems. Additionally, the licen-
see reviewed and closed out numerous NRC open items and is continuing to do so.
The licensee issued procedure QAIP-27, "Corrective Action," Revision 0, which
allowed the QA organization to elevate issues of a significant nature to the
appropriate level of management. This item is closed.
-8-
.
1
.
l
.
'
i
Open Item 50-312/86-41-41: LRS Management Appraisal Report Issues
The licensee did not include outstanding open items from the LRS consultants'
Management Appraisal Report in the ongoing Plant Performance and Management
ImprovementProgram(PP&MIP).
The licensee added the open issues from the LRS report in its new tracking
program as described in procedure RSAP-0215, "Restart Scope List (RSL)/Long
Range Scope List (LRSL) Development and Administration," Revistor. O. The
RSL/LRSL superseded the Plant Performance and Management Improvement Program
(PP&MIP). This item is closed.
Open Item 50-312/86-41-42: Validation of the 00I-12 Process
During the ASRTP inspection, the licensee committed to validate the adequacy of
the reviews conducted as part of the QCI-12 review process.
The ifcensee implemented the EASRTP inspection program for all 33 selected
systems. The EASRTP inspection program appeared to contribute to improving
safety at the plant as discussed more fully in Section 3.1 of this report.
This item is closed.
Open Item 50-312/86-41-43: Selected System Status Report (SSR) Document
Control
The licensee did not appear to be properly controlling SSRs considering their
use as a basis for the NRC Safety Evaluation Report (SER) for restart authori-
zation.
The licensee issued procedure AP-93, "System Status and Investigation Reports,"
Revision 0, which adequately described the control procedures for the SSRs.
This item is closed.
Open Item 50-312/86-41-44: Restart Organization
Because the licensee's ' organization had several vacancies, temporary contractor
personnel filled key positions in the restart organization. Additionally,
there were no plans for making the transition from contractor personnel to
permanent employees to support restart.
The licensee has reorganized and contracter personnel were replaced in key
man 6gement positions with permanent SMUD employees. The inspection team
reviewed the qualifications of management' personnel and interviewed selected
managers. New personnel filling key positions appeared to be well qualified I
and were working toward becoming an effective management team. This item is
closed.
2.2 Items Remaining Open After the Inspection
Open Item 50-312/86-41-01: AFW Turbine Driven Overspeed Issues
The team identified the following concerns associated with overspeed of the
turbine-driven AFW pump: (1) The overspeed trip point (as high as 4650 rpm) I
could be well above the overspeed rating of the electric-drive motor (4320 rpm) !
connected to the comon shaft, (2) the pump discharge piping was not analyzed !
I
9-
- .
O
.
.
for the overpressure condition that could result from an overspeed event, and
(3) the time required for the depressurization of the turbine governor control
oil after overspeed trip was not known but was required for operating procedure
guidance to prevent a subsequent overspeed trip upon restart.
The licensee obtained a certification from the motor vendor that the motor was
not overstrested up to 4500 RPM and the tolerance on the turbine overspeed
setting was changed so as not to exceed 4500 RPM. An analysis was perfonned of
the discharge piping. All but one section was found to be within allowable
stresses and that one section was only slightly above the allowable strasses
for worst case conditions. The licensee committed to replace this section of
piping before restart. An overspeed test is planned to detennine the governor
bleed time. The test results will be incorporated into the operating procedure
to prevent overs ed on subsequent restarting of the AFW pump. The team found
the licensee's actions in response to this item acceptable. This item will
remain open until the results of pump testing are available and procedures are
revised to reflect those results.
Open Item 50-312/86-41-07: Main Feedwater (MFW) System Problems
The inspection team identified eight SSR problems that should be resolved
before restart. The status of each of those problems follows.
Problem 6: Faulty Main Feedwater Pump Lovejoy Control Response
.
The licensee developed a permanent modification to jumper out the dead
band module. The licensee determined that this configuration would be
more reliable. The inspection team found this response adequate.
Problem 9: Main Feedwater Startup Flow Control Valves Stick Closed
Occasionally and have Slow Response to OTSG Level Change
The licensee refurbished the flow control valves to correct the observed
problems during this outage. The team found the licensee's response to i
this problem adequate.
Problem 12: Main Feedwater Flow Control Valve Positioning During
Transients (Overfeed)
The inspection team reviewed Casualty Procedure C.10 "Main Feedwater
Induced Transients," Revision 2, and found that the prescribed guidance
adequately addresses actions for an overfeed condition.
Problem 18: Correct Casualty Procedure C.26 for Main Feedwater
Pump Operation With Low Condenser Vacuum
This problem will be reviewed during a future inspection.
Problem 19: Correct Casualty Procedure C.10 for Action on Lo,sss
of One Main Feedwater Pump
The inspection team reviewed procedure C.10, "Main Feedwater Induced
Transients," Revision 2, and found the guidance for the loss of one main
feed pump to be adequate.
-10-
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
e 0
-
Problem 32: Update Piping and Instrumentation Drawing M580, Sheet 1
The team reviewed the revised drawing and found no discrepancies. The
inspection team found the licensee's response to this problem adequate.
Problem 31: Main Feedwater Pump Control From Lovejoy to Bailey Hand / Auto
Station Is Not Performing as Designed
The licensee plans to revise Procedure A.50, "Main Feedwater System," to
ensure that controller inputs are matched before the shift occurs from
manual to automatic. The inspection team will review procedure A.50 at a
future inspection to determine whether the problem has been resolved.
Problem 45: Main Feedwater Pump Governor Is Slow to Respond
The dead band module was found to be the failing component. A modifica-
tion was generated to remove the dead band module and install a jumper.
This, in effect, was the same modification as the temporary jumper that
had been installed during previous plant operations and found to be
satisfactory. This change required the controller to be continually in
service rather than being in service only when the pump is operating at a
speed that is outside the dead band. The licensee determined that this
configuration was more reliable. The team found this response to be
adequate.
Open item 86-41-07 will remain open pending the closeout of MFW SSR problems 18
and 31.
Open Item 50-312/86-41-11: 125 V de System Problems
The inspection team identified four SSR problems that should be resolved before
restart. The status of each of those problems follows:
Problem 8: Operating Procedure A.61 Deficiencies. The revised procedure
will be reviewed during a future inspection.
Problem 9: Battery Room Temperature Control. Discrepancies existed
between the temperatures assumed in the battery sizing calculations and
the referenced mechanical heating, ventilation, and air conditioning
(HVAC) calculations. The licensee issued new calculations to establish
the allowable temperatures for the auxiliary building and NSEB battery
rooms. The licensee's electrical group prepared new battery sizing
calculations that had minimum battery temperatures traceable to the new
HVAC calculations. Additionally, maintenance procedures were revised
accordingly to limit the battery minimum and maximum temperatures to the
design input valves. The inspection team found this response adequate.
Problem 20: Battery Charger Refurbishment. The licensee had identified
one of six original Class lE battery chargers for refurbishment before
restart, although the filter capacitor banks for all six chargers were
overdue for replacement. The licensee will replace the filter capacitor
banks for all six chargers before restart. The inspection team found this
response adequate.
-11-
. - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
.
. .
a
'
Problem 22: Damaged Battery Terminal Posts. The licensee prepared
nonconformance reports (NCRs) 55504 and 55548 to identify the deformed
terminal posts on two vital battery cells. The manufacturer's representa-
tive inspected the cells on-site and concluded that the damage to the
posts would not affect the safety function of the batteries. The licensee
planned no further action. The inspection team found this response 1
adequate. l
l
Open item 86-41-11 will remain open pending resolution of SSR problem 8. '
Open Item 50-312/86-41-12: 120-V ac System Problems
The inspection team identified five SSR problems that required further resolu-
tion before restart. The status of each of those problems follows:
Problem 2: Control Room Indication for Breakers. The licensee had i
identified a number of problems with inadequate control room indication of I
circuit breaker position in the ac vital electrical systems. The licensee
contracted to have Impell review the vital ac systems to determine the
adequacy and consistency of the alarms resulting from a trip of any l
electrical protective device. The summary report identified 15 generic
recommendations requiring both procedure and hardware changes. The inspec-
tion team reviewed the proposed changes and determined that the procedure
changes would provide adequate guidance for restart. The licensee j
initiated the required procedure changes. The inspection team found this
response adequate.
Problem 5: 120 V ac Systems Casualty Procedures. This item will be
reviewed during a future inspection. :
Problem 6: Missing Load Schedule in Procedure A.62. The licensee revised 1
Procedure A.62, "120 Vac Vital System," to include the missing load list.
The inspection team found this response adequate.
Problem 7: Response to IE Bulletin 79-27. NUREG-1195 identified a number
of problems involving an inadequate response to IE Bulletin 79-27, "Loss
of Non-Class 1E Instrumentation and Control Power System Bus During Opera-
tion." The licensee had planned to review and update its original IE
,
'
Bulletin response, but, before the original ASRTP inspection, had not
planned to impleme-t the recommendations resulting from the review before
restart. The licensee c.ontracted with Babcock & Wilcox (B&W) to review
the requirements of IE Bulletin 79-27. The B&W report listed 13 recom-
mendations including procedure changes and provisions for alternate power
supplies. These recommendations were incorporated into the licensee's
response for IE Bulletin 79-27. The inspection team reviewed the response
associated with the Class 1E control power systems and determined that
one recommendation was significant enough to require a hardware modifica-
tion before restart. A spare 120 Vac cable should be dedicated to provide
an alternate power source to the electrically actuated pressurizer relief
valve (PSV-21511). The NRC is reviewing the licensee's complete response
to IE Bulletin 79-27 and will include the results of that review in a
supplement to the SER for re; tart.
Problem 12: Local Indication of Circuit Breakers Position. The licensee
identified a number of problems with inadequate local indication of
-12-
. .
-
tripped circuit breakers. An Impell review indicated that this problem
was limited to Westinghouse circuit breakers. The team determined that
Westinghouse circuit breakers in panels S1A3, S1A4, S183, and S184 had
trip indicators that may be questionable. The licensee agreed that these
circuit breakers should be improved by adding a white dot to their handles
to help the operator detect a tripped circuit. The implementation of
this resolution will be reviewed during a future inspection.
Open item 86-41-12 will remain open pending resolution of 120-V ac SSR problems
5, 7, and 12.
Open Item 50-312/86-41-13: 480-V ac System Problems. The inspection team
identified nine SSR problems which required further resolution before restart.
The following is a status of those problems:
Problem 10: Loss of Power Alarms for 480 Vac MCC Loads. The licensee
questioned whether acceptable alarm or indication was available to detect
loss of power (by tripping the circuit protective device) to the 480 Vac
Motor Control Center (MCC) loads. An Impell study identified acceptable
alternate indication for all but three MCC loads. The team determined
that the three identified problem circuits (Breakers 2A317, 28317, and
2A208A) should have their circuit breaker handles modified to provide
clear indication of circuit breaker trip status for operators (See
problem 16).
Problem 11: 480-V ac System Casualty Procedures. This item will be
reviewed during a future inspection.
Problem 16: 480-V ac Circuit Breaker Handle Position. The licensee
identified a problem with the inability to visually determine when a motor
control center circuit breaker had tripped. The MCC manufacturer
(Westinghouse) made recommendations for modifying the circuit breaker
handles to give adequate visual indication of the state of the breaker
(closed or tripped). Two ECNs were prepared to perform this work. The
licensee determined that the three circuits identified in SSR Problem 10
as having no other means of identification should have their circuit
breaker handles modified before restart. The remaining circuit breakers
will be modified after restart. The inspection team found this response
adequate.
Problem 19: Inconsistent Alarms for loss of 480-V ac Systems. The I
licensee identified an inconsistency in the manner of alarming the feeder
and incoming breaker trips on the 480-V ac switchgear. The licensee ,
provided a point paper stating that the depth of the inconsistency was of I
minor concern. The team agreed that this minor inconsistency should not
be a restart item and found this response adequate.
.
Problems 25, 33, and 34: Procedure A.59 "480 Vac System Operating Procedure"
Deficiencies. This item will be reviewed during a future inspection.
Problem 26: Indication for loss of Control Power for 480-V ac Circuit
Breakers. The licensee questioned the potential lack of indication for
loss of 480 Vai switchgear control power. A study performed by Impell
indicated that all 480-V switchgear circuits contain one or more indicat-
ing lights that will turn off when control circuit power is lost. The
-13-
.
-
team reviewed random examples of 480-V switchgear circuits, observing the
referenced indicating lights on the control circuit schematics drawings,
and found no deficiencies. The team found this response adequate.
Problem 46: Drawing E-108, Sheet 30 Deficiencies. The licensee identi-
fied a potential conflict between the single line diagram for bus S2C9 and
the schematic drawing for breaker 2C915. On further review, the licensee
noted that both drawings were consistent and that breaker 2C915 was a
spare. Bus SIGB was a non-safety instrument bus fed from a battery-backed
inverter and did not require a backup power source. The inspection team '
agreed with this review and found the response adequate.
Open item 86-41-13 will remain open pending resolution of 480-V ac SSR problems
25, 33, and 34.
Open Item 50-312/86-44-14: 4160-V ac System Problems.
The inspection team identified four SSR problems that had to be resolved before
restart. The status of those problems follows.
Problem 8: Slotted Protective Relays. The licensee found that the
mounting holes for magnets used in the protective overcurrent relays
mounted on the 4160-V switchgear had been elongated and the seismic
effects of the modification had not been adequately evaluated. The
licensee performed a calculation (Z-EDS-C0929) to prove the seismic
acceptability of the magnet mountings. The inspection team found this
response adequate.
Problem 25: Indication of Loss of Control Power for 4160-V Switchgear.
The licensee identified a potential problem involving indication of loss
of control power to the 4160-V switchgear control circuits. An Impell
study on this subject identified five circuits that could lose control
power and not be detected. The team determined that the indicating lights
for the control circuits associated with AFW pump P-319 and the under-
voltage trip circuit logic associated with the four safety-related 4160-V
buses should be modified to provide direct indication in accordance with IE
Problem 32: Procedure for Operating the Startup Transformer. The
licensee found that no guidance was provided for operating the startup
transformer. The team reviewed the revised procedure A.54, "220 KV
Electrical Systems," Revision 9, and found that the addition of the
startup transformer system checks provided were adequate guidance.
Problem 33: Casualty Procedure for Loss of Cooling to Transformer.
Casualty procedure C.143, "Loss of 480 Volt MCC S2El," incorrectly
directed the operator to check the alternate power supplies on the loss of
power to both auxiliary transformers, ACB-2E148 and ACB-2E149, and both
startup transformers, ACB-2E150 and ACB-20151. Casualty procedure C.143 I
was revised to direct the operator to check the normal power supplies to I
these transformers upon loss of power. The team found this response l
adequate.
'
l
Open item 86-41-14 will remain open pending resolution of 4160-V ac SSR Problem I
25. 3
!
-14-
]
.
-
Unresolved Item 50-312/86-41-28: AFW Pump Operaoility
The team was concerned that AFW flow testing did not reveal the actual AFW flow l
to the OTSGs because there was no valid basis for the 60 gpm minimum flow j
bypass flow rate that was used in determining the flow available to the OSTG's -
and the CST level instrument used to determine the total flow out of the CST i
was found to be unreliable and not properly calibrated. '
With the installation of the modifications described in the response to ASRTP
item 86-41-03 to allow accurate measurement of pump flow directly tt. rough the
full-flow test line, the concerns listed above are no longer pertinent to
determining the flow of the AFW pumps. A test procedure was developed to
determine the flow in the minimum-flow test line during operating conditions.
The licensee has also committed to performing a multi point calibration of the
CST level instrument LI-35803. This item will remain open pending completion
of the minimum-flow test procedure and performance of a full-loop, multi point
calioration of the CST level instrument LI-35803.
Unresolved Item 50-312/86-41-30: AFW System Surveillance Procedures
This item will be reviewed during a future inspection.
.
Unresolved Item 50-312/86-41-31: 125-V de System Surveillance Testing
Deficiencies were identified with the observed cell temperatures and electro-
lyte levels in the station batteries and recorded values for the battery
charger periodic maintenance tests. These deficiencies were caused by
inadequate procedures.
The licensee revised procedures SP.300, "Weekly Nuclear Service Battery Pilot
Cell Test," and SP.301, "Monthly Nuclear Service Battery Test." The inspection
team reviewed the revised procedures for the battery testing and found them
acceptable. The revised procedures for battery charger testing will be
reviewed during a future inspection. This item will remain open. '
Open Item 50-312/86-41-33: AFW Operating Procedure Deficiencies
This item will be reviewed during a future inspection.
Open Item 50-312/86-41-34: AFW Pump Runout Recognition and Control
This item will be reviewed during a future inspection. l
Unresolved Item 50-312/86-41-36: Valve Maintenance Procedures
The licensee's maintenance procedures for MOVs provided inadequate guidance for
inspection of the brakes and the maintenance procedures for safety-related
air-operated valves were not developed.
The licensee's corrective action for instituting maintenance procedures for
MOVs with brakes and safety-related, air-operated valves were reviewed. The
applicable maintenance procedures for the MOVs were revised to include the
vendor recommendations for inspecting the brakes. The condition of the brakes
will be examined every refueling cycle. The licensee's corrective action for i
establishing maintenance procedures for safety-related, air-operated valves had l
l
l
-15-
i
. _ _ __
.- . -
. .
-
not been completed at the time of this inspection. This item will remain open
pending verification of the development and implementation of maintenance
procedures for safety-related air operated valves.
Open Item 50-312/86-41-39: QA Surveillance Program
The team found (1) deficiencies in the quality of past QA surveillances, (2)
lack of accountability for correcting identified deficiencies, and (3) lack of
a trending program.
The licensee revised procedure QAIP-2, "Quality Assurance Surveillance
Program," to require that responsible organizations respond to findings within
14 days of receiving a report. Additionally, procedure QAIP-16, "Trend
Analysis," has been developed to trend results of the QA Surveillance Program
with other inputs. These changes were only recently issued and had not been
implemented sufficiently for the inspectors to adequately review the improved QA
Surveillance Program. This item will remain open.
l
.
-16-
- - - - _ _ - _ . - , . _ _ _ . . _ _ _ _ _
.
.
'
3 DETAILED INSPECTION FINDINGS
3.1 Evaluation of Expanded ASRTP Inspection Program
In response to the NRC Augmented System Review and Test Program (ASRTP)
concerns about the adequacy of the engineering and technical reviews performed
as part of the SRTP process, the licensee decided to perform expanded ASRTP
(EASRTP) inspections on all 33 selected systems using 6 multi-discipline
teams. The EASRTP inspections were patterned after the NRC ASRTP inspection
(50-312/86-14) and were intended to ensure that pertinent areas of system
operation, design, testing, and maintenance were reviewed before plant restart.
The NRC inspection team conducted its evaluation of the EASRTP inspections in
three ways:
(1) Reviewed issued EASRTP inspection reports for significant findings.
(2) Reviewed inspection documentation to detennine thoroughness of
reviews.
(3) Conducted independent reviews of the same systems to compare inspec-
tion findings.
At the time of the inspection, the EASRTP inspection team had completed
its investigation of 18 selected systems and was progressing on schedule
to complete inspection of the 33 systems by November 1987.
3.1.1 EASRTP Inspection Findings Review
After completing 18 systems inspections, the EASRTP team had identified more
than 150 findings that had been accepted and scheduled for resolution. The
issues raised by the EASRTP inspections varied in their safety significance,
but were comparable to issues raised during the NRC ASRTP inspection. The team
selected the following significant EASRTP inspection findings (RI) to determine
whether they were correctly scheduled for resolution before restart:
(1) Emergency diesel generator fuse coordination was inadequate (RI 35).
(2) Instrument air lines to the turbine bypass and atmospheric dump valves did ,
not allow for thermal expansion (RI 97) I
i
(3) Integrated control system (ICS) calibration setpoints were not adequately l
controlled (RI 36)
'
(4) Decay heat removal (DHR) system reactor building sump isolation valves
were located outside the reactor building and had no covering jackets (RI
18).
(5) Nuclear raw water (NRW) system cooling capacity may not be adequate for
all system heat loads (RI 25).
(6) NRW system flow was not properly balanced (RI 32).
(7) The 120-V ac system cabinet was unqualified because unauthorized ventila-
tion modifications had been accomplished (RI 146).
-17-
i
. .
,
.
'
(8) Design pressure of MFW heaters was less than pump discharge pressure (RI l
108)
(9) Non-nuclear instrumentation system (NNI) calibration setpoints were
incorrect (RI 150).
(10) Nuclear service cooling water (NSCW) system flows to decay heat removal
(DHR) coolers were above the design maximum and could cause tube damage
(RI104).
(11) NSCW system parallel flow analysis was incomplete (RI 107).
(12) NSCW heat removal capabilities may not be adequate for DHR system loads
(RI 128).
(13) No technical bases were stated for response time acceptance values for the
anticipatory reactor trip system (ARTS) surveillance tests (RI 194). l
(14) Carbon steel components were used in the relief valves of a borated water
system (purification and letdown) (RI 190).
I
'
(15) Containment fire protection system capabilities appeared inadequate
(RI 210).
In each case, the licensee had scheduled the resolution of the selected findings
before restart. The inspection team did not review the licensee's corrective
actions taken to resolve the EASRTP inspection findings, as they were still
being implemented. The tracking and closeout of EASRTP findings was being
conducted by the licensee's Engineering Response Team.
3.1.2 Review of EASRTP Methods
The NRC team interviewed EASRTP team members and reviewed document sheets that
identified the areas inspected for each system. The NRC sampled the EASRTP
inspection documentation for the following systems:
(1) emergency diesel generator l
(2) ac electrical
(3) fire protection
(4) nuclear raw water
(5) nuclear service cooling water
(6) main feedwater
In general, the methods used to conduct the EASRTP inspection were comparable
to that used for the NRC ASRTP inspection. However, the NRC team identified
the following areas in which enhancements could improve the EASRTP process:
-18-
1
. . l
-
\
'
(1) Some aspects of the operating, testing and maintenance programs were not
ready for the EASRTP inspections. Several modifications, procedures, and
work items were incomplete, preventing these functional areas from being l
fully evaluated. The EASRTP inspectors often had only draft procedures or :
I
preliminary guidance available to assess these functional areas for
several key systems. In these cases, the inspectors provided valuable l
input to the procedural development process, but were not able to complete
the verification. The EASRTP program did not identify the missing proce-
dures for followup review.
1
(2) The reviews conducted during the EASRTP inspection in the f9' lowing areas
appeared to be less than the reviews performed during the NRC ASRTP
inspection: l
(a) Apparently, only 1 of approximately 50 calculations was reviewed
for the NFW system inspection and problems were identified with that
calculation. The NRC team determined that further calculation !
reviews should have been made to determine if other problems existed.
(b) The reviews of some surveillance procedures consisted only of
ensuring that major components of the system were covered by testing.
There was no line-by-line review of the surveillance procedure to
ensure that testing was adequate.
l
'
(c) Overload protection for the safety-related 4160-V ac pump motors was
checked by comparing the Maintenance Department records with the ,
specified setpoints identified on drawing E1011. No attempt was made j
to verify the adequacy of the specified relay setpoints. The EASRTP
review of electrical protection for MOVs appeared to only check that
the rated motor load would not trip the circuit breaker. The review
did not consider the overall component coordination or question the
protection for MOV HV20003 which apparently had a 20-amp breaker
protecting a 10-amp wire supplying a 1 amp motor.
(d) The scope of the review for the 125 Vdc system did not appear to
check that the auxiliary building battery charger high voltage alann
setpoint was adequate for the rated voltage range of the de equipment
connected to the bus.
It should also be noted that the EASRTP inspection program covered some
areas in more depth than the NRC ASRTP inspection. The team was favorably i
impressed with EASRTP reviews of the procurement area, walkdown of systems
and assessment of system material condition.
(3) The NRC disagreed with the final disposition of one EASRTP issue. The
EASRTP inspection identified an apparent discrepancy between the nuci t r ,
service cooling water (NSCW) System design-basis document and drawing i
E203, sheet 43. The design-basis document stated that an NSCW surge tank
low-low level alarm would trip the NSCW pump, except when a safety
features actuation system (SFAS) signal was present. Drawing E203 showed
that the NSCW surge tank low-low level alarm would trip the pump with an
SFAS signal present. The EASRTP evaluator, team leader and program
manager concluded that this apparent discrepancy was not significant
enough to require assistance from site engineering to resolve the problem
prior to restart.
-19-
_
_ _ _ _ _
- O
l
.
Q
The NRC reviewed this same issue further and determined that the NSCW
surgetanklow-lowlevelswitches(LSLL-48401andLSLL-40402)werenot
safety-related and drawing E203 was correct. However, IEEE-279, "Criteria
for Protection Systems of Nuclear Power Generating Stations," requires
that protective system components be safety-related. Failure of either of
these two level switches in the presence of a valid SFAS signal would
prevent the affected pump from completing its intended safety function.
The inspection team concluded that in this one isolated case, the EASRTP
team did not aggressively pursue their initial observation completely.
The resolution of the apparent problem with the non-safety-related level
switches defeating an SFAS signal to the NSCW system pumps will remain
open pending NRC followup. (50-312/87-29-01)
3.1.3 Comparative Inspection Results
The NRC team conducted an independent review of selected aspects of the systems
listed in section 3.1.2 to form a basis for comparison with the EASRTP find-
ings. For most of the systems reviewed, the NRC inspection team did not
identify any new safety issues. However, some specific concerns were raised
involving the emergency diesel generator system that warrant licensee followup:
(1) The fire door between the emergency diesel generator GEB room and the
east-west hallway did not appear to meet the National Fire Protection
Association Code requirements for separating fire areas. The door must
meet these requirements because cables associated with the emergency
diesel generator GEA run through the east-west hallway outside the door.
This issue will remain open pending NRC followup (50-312/87-29-02).
(2) The drain system for emergency diesel generator GEA and GEB rooms may not i
be adequate to remove the water from the sprinkler systems in the rooms. i
Each room had two 2-inch drain lines that all merged to a 3-inch coninon
drain line. The licensee did not have any analyses to (a) show that the
-
two 2-inch drain lines would adequately remove the water from the
sprinkler systems .in the rooms, or (b) that the 3-inch common header would
adequately drain the water from the 2-inch lines. The team was concerned
about the capacity of the 3-inch drain line because there were no check
valves in the drain line to prevent reverse flow. Consequently, water
could back up into the other emergency diesel generator rooms. At the
conclusion of the on-site inspection, the licensee was preparing an
analysis to demonstrate the adequacy of the diesel system for the
emergency diesel generator room. This item will remain open pending
review of the licensee's analyses (50-312/87-29-03).
(3) The design of the starting air system may not be adequate for the new l
emergency diesel generators GEA2 and GEB2. On each new emergency diesel I
generator, the starting air system was safety-related from the inlet
check-valves on the accumulators to the engine. The balance of the air
system was designed as non-safety-related because it was not thought to be
required to perform a safety function after the diesel engine starting
sequence. Upon further review, the licensee determined that air was
required continuously to control emergency diesel generator operations.
In SMUD letter JEW 87-358 to the NRC, dated April 1,1987, the licensee
committed to test the capability of the air receivers to maintain the
required pressure for diesel generator control over a prolonged period of
-20-
.- . - - . . .
. .
.
,
time. This item will remain open pending NRC review of the test results
(50-312/87-29-04), l
(4) The air compressor outlet safety valves appeared to be set above the
nameplateratingofthecompressors(290psigvs.250psig). The NRC team
determined that because of the design of the system, this setting could
cause the compressors to operate regularly above tneir nameplate rating.
This item will remain open pending NRC reivew of the resolution of the
discrepancy between the compressor rating and emergency diesel generator
air system operating pressure (50-312/87-29-05).
3.2 Engineering Action Plan Assessment
The Engineering Action Plan (EAP) was developed after the ASRTP inspection
(50-312/86-41) to accomplish the following objectives:
(1) Upgrade engineering practices to improve the quality of future design
work.
(2) Independently review the work perfonned in support of the current outage.
(3) Re-establish the system design bases for the as-built configuration of the
plant.
The EAP was still in the process of being implemented during the NRC inspec-
tion. Report D-0050, "Engineering Action Plan for Rancho Seco Nuclear i
Generating Station," Revision 1, was in the final stages of management review l
before issuance. This report outlined the various programs in progress to '
accomplish EAP objectives and provided the latest draft revisions for design 1
change control procedures. The NRC team assessed the EAP improvements by the
following methods:
(1) Conducted a limited review of EAP design change control procedure
revisions.
(2) Observed the engineering processes being carried out, interviewed SMUD
engineers, and reviewed a sample of engineering documents completed since
the EAP had been implemented. l
(3) Reviewed the results of the independent reviews conducted by Babcock &
Wilcox (B&W) and Bechtel inspectors.
The NRC inspection team noted several improvements in the quality of engineer-
ing activities at Rancho Seco since the initial ASRTP inspection in February l
1987. It appeared that implementation of the EAP was improving the quality of I
design change control activities at Rancho Seco.
3.2.1 Review of EAP Design Change Control Procedures
The inspection team performed a cursory review of the draft procedures and
program overview included as part of the EAP. In general, these procedures and
documents provided adequate guidance for conducting the design change control
process . However, the inspection team identified the following concerns during
its review:
-21-
..- -.
_ _
.
. .
.
(1) Draft Procedure RSAP-0301, "Configuration Management Program," Revision 0
'
Section 5.2.6, stated that work items may be incrementally closed out
before the entire change package has been completed in order to support
real-time operational needs. The team considered that this practice
should only be allowed after a thorough safety analysis by engineering
personnel takes place to ensure the plant is left in a safe condition.
(2) Draft Procedure NEP 4109, "Configuration Control Procedure," Revision 7,
Section 4.2, stated that engineering change notices (ECNs) were not
required for drawing changes if no actual field work was performed. The
team was concerned that there may be drawing changes that involve no
actual field work but that could have an effect on plant safety.
Potentially significant changes to drawing notes or valve positions would
not require an ECN and the appropriate reviews afforded by the ECN
,
process.
(3) Page 23 of the EAP overview document stated that obsolete calculations
would be deleted. The team was concerned that obsolete calculations
should be superseded, not deleted, in order that a record of the design
evolution is maintained.
(4) There did not appear to be a method for closing out an ECN package that
ensured all the various elements of the ECN were included in the package.
None of the procedures reviewed required an inventory of the required
elements of the package to be listed for package closecut.
3.2.2 Review of the EAP Work Products
The inspection team reviewed the analyses conducted to resolve previous NRC
open and unresolved items, sampled calculations performed by the licensee in
support of new design work, and interviewed engineering personnel to assess the
improvements made by the EAP. Overall, the quality of engineering activities
at Rancho Seco was improving, but the team identified some weaknesses in this
area. The team concluded that these weaknesses were isolated instances and
were attributable to the fact that certain aspects of the EAP had not been
completely implemented throughout the site. The following weaknesses were
identified during the team's review:
(1) During the initial ASRTP inspection, the team identified errors in
calculation Z-FWS-M1742, which justified the acceptability of utilizing a
worn stem nut in an M0V. The licensee replaced the worn stem nut and
repeated the calculation. This new calculation has numerous nonconserva-
tive errors which rendered invalid the conclusion that the worn stem nut
was acceptable. The new errors were as follows:
(a) The thrust load used in the calculation was the minimum thrust
required to allow the valve to operate under design conditions. This
value was less than the thrust generated by several normal modes of
operation that were not considered, such as manual operation, higher
than minimum torque switch settings, and initial valve operation when
the torque switch is bypassed. '
l
1
-22-
_ _ _ _ _ _
1
. . )
i
-
1
'
(b) The effective factor used in the calculation to account for the
uneven load distribution inherent in threaded members was less than
the values identified in the machine design texts referenced in the
calculation. In subsequent conversations with the licensee's
representative, an even lower factor was argued to be acceptable
based on a machine design text. However, the factor cited was for
shear stress whereas the parameter being considered was tensile
stress.
(c) The licensee stated that localized yielding was acceptable, based on
a machine design text. The team agreed that localized yielding was
acceptable for some stati:: applications of screw threads, but it is
not acce) table for a power screw where yielding would cause mismatch
of the t1 reads for subsequent operations.
(d) No consideration was given to the increase in friction and load to !
produce given thrust as a result of the change in thread angle with
wea r. The team considered that this would likely be the predominant
factor in determining the acceptability of the worn condition.
(e) The originator of the calculation deemed a vigorous analysis unneces-
sary because the operation of the valve had been observed. The team I
disagreed with that conclusion because the observation had not taken l
place at the valve's design pressure drop loading.
Since the stem nut had been replaced, there did not appear to be any
safety problems with the existing system configuration. The team was l
concerned, however, with the programatic implications that this j
calculation may indicate. The calculation appeared to be performed and
'
verified by engineering personnel who were unfamiliar with MOV operation.
Additionally, this calculation was still in the engineering files and
could be used as a model for future calculations of a similar nature. In i
response to the team's concerns, the licensee removed the calculation from I
the files and initiated corrective actions to ensure that similar
calculations will be performed properly in the future. The licensee
maintained that this calculation was an isolated example and not
indicative of the overall engineering program quality. On the basis of
the other engineering activities reviewed, the inspection team agreed with
the licensee's position.
(2) During the initial ASRTP review, deficiencies were identified with the de
short-circuit calculation for the auxiliary building batteries that could
result in the short-circuit current at de panel SOB being above the ;
manufacturer's interrupting rating for the enclosed circuit breakers
(UnresovedItem 50-312/86-41-16). The licensee replaced the panel feeder
cable with a smaller size conductor to limit the short-circuit current at
the panel. The short-circuit calculation was revised to incorporate the
higher resistance of the smaller cable and also to address other concerns
originally raised by the team. After reviewing this revised calculation,
the team identified mathematical errors that reduced the margin of rated
interrupting current over calculated short circuit by approximately 500
amperes (25 percent of the calculated margin). The calculation cover
sheet indicated that checker had verified the mathematics of the calcula-
tion. The licensee corrected the errors in the calculation and noted that
the breakers were still appropriately sized.
-23-
_ _
. - - . -
._
,
. .
.
4
(3) The NRC inspection team identified the following instances in which there
did not appear to be effective comunications between engineering
personnel and other organizations within the station:
(a) As part of the licensee's response to the original ASRTP concern with
the backup air supply for the EFIC modifications (0 pen Item
50-312/86-41-06), the operating procedure was modified to include
recording the bottle pressures in daily logs. The minimum allowable
pressures specified in the log sheets were at the 2- and 3-hour alarm
points for the bottles which appeared to defeat the purpose of
performing the daily checks. These values were well below the
pressure that had been recormended by engineering personnel in the i
design-basis report for the backup air bottle modifications and did
not appear to have any technical bases. In response to the NRC
concerns, the licensee is revising the operating logs.
(b) A calculation perfomed in response to NRC open item 50-312/86-41-08 l
'
to determine the qualified life of the 0-rings for the excess flow
check valves in the OSTG level instrument lines installed as part of
the EFIC modifications detemined that the expected life of the
components was 10 years. However, this calculated life for the ,
0-rings had not been communicated to the Maintenance Department
for incorporation into its replacement procedures. In response to
the NRC team's concerns, the licensee is changing the preventive
maintenance procedures.
(c) Operations Department personnel were unaware that compressed air was
required for continued operation of the new emergency diesel genera-
'
tor units (GEA2 and GEB2). As a result, the emergency diesel genera-
tor casualty procedures did not include provisions for providing
backup air in the event of loss of the installed non-safety-related
air compressors. This information had not been comunicated from the
Engineering Department to the Operations Department.
The licensee had previously identified similar concerns regarding the
comunication of infomation from the Engineering Department to other
groups at the plant. A group was being established within the Operations
Department to facilitate the implementation of design infomation into
pertinent operating and test guidance.
3.2.3 EAP Independent System Reviews
In addition to the EASRTP inspections and as part of the EAP, the licensee
conducted independent safety reviews of the design, testing, and operation of
the System Review and Test Program (SRTP) selected systems by the following
three programs:
(1) DesignCalculationReview(Bechtel) l
l
(2) Rancho Seco System Configuration and Restart Test Program Evaluation
(Babcock & Wilcox) l
(3) QA Vertical Audit (Stone and Webster)
l
I
-24-
.- - ..
. _ _ _ _ _ _ _ - - - - - - - - - - - - - - -
.
.
'
These programs duplicated some of the efforts made by the EASRTP teams, but
overall identified a number of significant issues. The NRC inspection team
reviewed the results of these inspections, but not the methods or qualifi-
cation of personnel conducting the H! views. i
(1) The Bechtel Design Calculation Review Program )
I
Those calculations performed in support of the modifications scheduled to
be made during the current outage were evaluated. Of 173 observations
made hy Bechtel teams during their review, 62 required correction before
restart. The licensee determined that the remaining 111 observations did
not require imediate corrective action, but should be considered for
areas of improvement in future calculation revisions. At the time of the
NRC inspection, the licensee had responded to Bechtel on 98 of 173 obser-
vations, outlining its intended corrective action and offering its
proposed schedule. The proposed completion dates appeared to support the
restart date in January 1988, but the team did not review the adequacy of
the planned corrective actions or their implementation.
i
(2) The Rancho Seco System Configuration and Restart Test Program l
1
Thirteen systems designed by B&W that were part of the 33 SRTp selected
systems were evaluated. The NRC inspection team reviewed 66 preliminary i
findings from the evaluation report and found many of these to be
significant issues that should be resolved before restart. Several of the
issues had been previously identified by the EASRTP inspection and actions i
were being implemented to resolve the design problems. The final l
evaluation report was scheduled for issuance from B&W in October 1987 and
final corrective action will be developed on the basis of those findings.
(3) The QA Vertical Audit 4
The system modifications for the EFIC system installation and the electri-
cal portion of the new emergency diesel generator installation were
evaluated. The vertical audit involved a thorough review of the proposed
system design, testing, procurement, and training for the two modifica-
tions. The audit identified several significant issues that must be
resolved before restart. These issues concerned the adequacy of the
design analyses for proposed functional testing of components, operating l
procedures, and quality of components installed in the system. At the
time of the inspection, the licensee's audited organizations had responded
to some of the audit findings. The NRC inspection did not review the
adequacy of these responses or their scheduling with respect to completion
before restart.
The NRC inspection team concluded that these independent reviews of the
selected systems identified issues that were safety significant and contributed
to the overall safe operation of the plant. The team was concerned that the
nature and depth of these findings required significant expenditure of
resources to resolve the issue before restart because many of the problems were
being identified very late in the outage schedule.
-25
. . _ - .-. . _ . - . .
_ --, -
. .
.
'
3.3 Procurement and Control of Fasteners
In February 1987, as part of an industry survey, the NRC obtained six fasteners i
from Rancho Seco maintenance personnel that were supposedly safety-related,
quality class I materials. These fasteners were then sent to Idaho National
Engineering Laboratory (INEL) for mechanical, chemical, and microstructural
analyses. As described in an NRC letter (D. M. Crutchfield to G. C. Andognini,
August 26,1987), two of the six fasteners failed to meet the ASTM A-193 Grade
- 87 standards that were stamped on the bolts. Mechanical testing revealed an
ultimate tensile strength of 54.65 ksi instead of the required 105 ksi;
chemical testing indicated out-of-specification readings for carbon, chromium,
manganese, and molybdenum; and microstructural analysis revealed that the bolts
had not received the proper heat treatment to yield a quench and tempered
martensitic structure as required by ASTM A-193 Grade B7 material. This result
raised concerns that counterfeit fasteners may have been installed in the
plant. The inspection team initially concentrated on the licensee's investiga-
tion into the specific fasteners that had the potential to be counterfeit, but
later expanded the scope of the inspection to include the licensee's program
for procurement and control of fasteners. This effort included further
sampling of fasteners for testing, evaluation of purchasing and control
practices for safety-related fasteners and review of procurement issues raised
during the EASRTP and QA Vertical Audit Programs.
3.3.1 Itvestigation of Counterfeit Fasteners
The two fasteners that failed INEL testing were marked "CB B7," indicating that
they were ASTM A-193 Grade B7 bolts supposedly manufactured by Clark Brothers
Bolt Co. The licensee documented the results of its investigation by SMUD
Office Memorandum MIC 87-186, dated October 3, 1987, entitled "Clark Brothers
t Bolts, ASTM A-193 Grade B7." It appeared that SMUD had purchased 64 "CB B7"
bolts from Westinghouse as non-safety-related, quality class II materials in
1977 and 1982; these were accounted for under stock code 26736. Additionally,
the licensee identified 24 "CB B7" bolts in the warehouse stored under stock
code 25436. The licensee's investigation accounted for all 24 bolts identified
by stock code 25436 and 40 of the 64 bolts identified by stock code 26736. The
licensee committed to locate and replace all the bolts with the "CB B7" head
]
markings.
, After completion of the on-site inspection, the NRC contacted Clark Brothers
Bolt Co. regarding the defective bolts. A preliminary investigation revealed
that Clark Brothers Bolt Co. does not have the capabilities to make the type of
bolts found to be defective at Rancho Seco. Since the bolts were actually
quality class II, the non-safety-related materials were not readily traceable
through Westinghouse.
The inspection team was concerned that the licensee provided quality class II
fasteners to an NRC inspector when quality class I fasteners were requested.
Additionally, the inspection team obtained the following fasteners from the
licensee for further INEL testing:
-26-
.. .. . - - . - _ _ - _ - . _ . - - - - - . _ - _ _ - - - - - ___ _
. .
NRC Material Meterial BIN Quality
ID No. Description h Stock Code Location Class
RSI ASME SA 193/B7 Bolt 042223 A39LO9 II
RS2 ASME SA 193/B7 Bolt 042223 A39LO9 11
RS3 ASME SA 193/B7 Bolt 042222 A39LO9 II
RS4 ASME SA 193/B7 Bolt 042222 A39LO9 II
RS5 ASTM A449 Bolt 026736 A41F21 II
RS6 ASTM A449 Bolt 026736 A41F21 II
RS7 SAE J429/gr 8 Nut 110340 G02806 I
RS8 SAE J429/gr 8 Bolt 009025 A39J05 II
RS9 SAE J429/gr 8 Bolt 009025 A39J05 II
RS10 SAE J429/gr 8 Bolt 009025 A39J05 II
RS11 304 SS Bolt 008932 A39G09 II
RS12 ASTM F593-85/304 Bolt 103118 A39F08 I
RS13 ASTM F593-84/304 Bolt 103103 A39C07 I
RS14 ASTM F593-84/304 Bolt 103103 A39007 I
RS15 ASTM F593-84/304 Bolt 103103 A39007 I
RS16 304SS Bolt 008932 A39G09 II
RS17 ASTM A193/B7 Bolt Quarantined II
RS18 ASTM A193/B7 Bolt Quarantined II
RS19 304 SS Bolt 008928 A39G09 II
RS20 ASTM F593-85/304 Bolt 103118 A39G09 I
These materials will be tested by an NRC laboratory and the results issued by
separate correspondence.
!
3.3.2 Warehouse Material Controls
The inspection team reviewed the licensee's controls of fasteners that were in
the warehouse while obtaining additional fasteners for testing. The identifi-
cation and segregation of safety-related quality class I fasteners and non-
safety-related quality clans II fasteners appeared to be inadequate. The team i
found the following instarces in which adequate controls did not appear to be
implemented:
(1) Quality class I cap screws, identified in Section 3.3.1 as RS-12 and
RS-20, were found together in the same bin under a single quality accept-
ance (green) tag, but appeared to have been manufactured by different
companies. Both appeared to be stainless steel, but one was polished and
stamped "SS 304" on the head, the other was pitted and not stamped.
(2) Quality class II cap screws, identified in Section 3.3.1 as RS-11 and 1
RS-16, were in the same drawer as quality class I cap screws RS-12 and )
RS-20, but were in separate bins. The two bins were separated by a thin !
setal divider in the drawer. When the inspectors opened the drawer, the l
loose quality acceptance tag was found in the wrong bin. The cap screws '
in both bins were approximately the same size making it possible to issue
one cap screw in place of the other.
(3) Quality class II cap screws, identified in Section 3.3.1 as RS-5 and RS-6,
were actually ASTM 449 material that was stored in a bin marked for ASTM
-27-
__ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _
4
.
.
A-193/B7 material. This arrangement presented a problem because ASTM
A-193/87 material has a nominal yield strength of 105 ksi and ultimate
tensile strength of 125 ksi but ASTM 449 yield strength is 92 ksi and an I
ultimate tensile strength in 120 ksi. It appeared that bolts of lower i
!
strength could have been issued from the warehouse for non-safety-related
use. .
l
(4) Quality class II cap screws, identified in Section 3.3.1 as RS-8, RS-9,
and RS-10, were SAE J429 Grade 8 material and were mixed in a drawer with
similarly sized cap screws of ASTM A-325 material. The ASTM A-325 cap l
screws had a tensile strength of 105 ksi and a yield strength of 81 ksi; I
the SAE J429 Grade 8 cap screws, however, had a tensile strength of 105 l
ksi and a yield strength of 130 ksi. The team was concerned that since
there were no markings or tags associated with either of the cap screws,
the wrong material could be issued in error and cap screws of insufficient
strength could be installed for non-safety-related application.
Although the licensee's program for controlling fasteners was not reviewed
thoroughly, the team was concerned that past practices may have resulted in
incorrect fasteners and other materials being installed in the plant for
applications inconsistent with their design.
3.3.3 Purchase Order Review !
l
The NRC inspection team reviewed the licensee's practices for purchasing 1
'
safety-related fasteners. It appeared that before 1986, the licensee did not
have a written program for the dedication of comercial grade items for !
safety-related applications or for purchasing quality class I materials or
services. Procedures were issued in 1986 after the licensee discovered this
problem. The inspection team did not review the procedures in sufficient
detail to assess their adequacy. The inspection team sampled the following
safety-related purchase orders:
No. Date Vendor Material I
46520 7/15/83 Cardinal Industrial ASTM A-193/B7 bolts .
Products Corporation ASTM A-194/2H nuts
l
Cardinal Industrial
'
44767 5/12/83 ASME SA 325 nuts and bolts
Products Corporation
16899 3/06/81 Babcock & Wilcox ASME SA 320/C43 studs
These purchase orders did not impose material traceability requirements,
require work to be perfomed under an approved QA program, or prohibit the
repair or rework as required by 10 CFR 21 and ASME Codes. The team did not l
review whether these purchase order deficiencies resulted in substandard j
material being provided to the licensee.
1
3.3.4 Licensee-Identified Procurement Problems
Several procurement and material control issues also appeared to be identified ,
by the licensee in its EASRTP and QA Vertical Audit inspections. The EASRTP l
inspection identified the following issues which may indicate problems with the
procurement or control of material at Rancho Seco:
-28- l
. -
~
!
!
. !
" I
_
Ref. No. Description j
RI 029, 110 Inadequate comercial grade procurement and
~
dedication methods
RI 068 Carbon steel bolting in systems containing boric
acid
RI 058 Unqualified lube oil coolers in NRW system
Additionally, the GA Vertical Audit Program identified the following concerns !
during review of the EFIC and emergency diesel generator electrical l
modifications: l
Audit Finding No. Description 1
l
87-20-13 Missing mild environment specification on
purchase orders for NSEB equipment l
l
87-20-17 Inadequate traceability for purchase order for i
target rock solenoid valves for EFIC
87-20-18 Heat number control log deficiencies allowing
for loss of traceability )
87-20-19 QA did not approve all safety-related purchase
orders
87-20-20 Receipt inspection deficiencies
87-20-21 Inadequate QA storage of procurement documents
87-20-22 Missing vendor audit
87-20-23 Inadequate followup for vendor audit findings
The NRC inspection team concluded that there was cause for concern about the
quality of fasteners being used at Rancho Seco Nuclear Generating Station. The
team determined that further review of the procurement area by the licensee and
NRC was necessary based on the EASRTP and QA Vertical Audit findings, problems
identified by the NRC team with the purchasing and warehouse control of safety-
related fasteners, and results of the fastener testing conducted in February
1987. The licensee had already initiated a review of their warehouse materials
to ensure that all quality class I materials were properly documented and
controlled. This item will remain open pending further review and inspection
by the NRC (50-312/87-29-06).
-29-
,
l
.
'
4 MANAGEMENT EXIT MEETING
An exit meeting was conducted at the conclusion of the on-site inspection on
October 9, 1987. The licensee's representatives at the exit meeting are
identified in Appendix A. Mr. Dennis F. Kirsch Director, Division of Reactor
Safety and Projects, Region V, and Mr. Charles J. Haughney . Chief. Special
Inspection Branch, NRR, represented NRC management at this meeting. The scope
!
of the inspection was discussed and the licensee was informed that the
inspection would continue with further in-office data review and analysis by
tean members. Team members presented their observations for each area
inspected and responded to questions from licensee representatives. The
licensee was informed that some of the observations could become potential
enforcement findings.
l
l
l
1
-30-
, _ - _ _ _ _ __
.
. 0
.
APPENDIX A l
Personnel Contacted
C. Andognini CEO, Nuclear l
J. Anderson Procurement
P. Anderson Electrical Maintenance Engineer
- J. Atwell Licensing
S. Bagga Design Engineer
S. Bajaj I&C Engineer
J. Baldwin EASRTP Evaluator
- M. Basu Lead Electrical Engineer
J. Bergstrom Operations
J. Bingham Civil-Design Engineer
N. Brown Buyer
C. Buchanan Document Control
S. Cannichael EASRTP Engineer
J. Chnastyk Operations
- D. Compton Operations Engineer
J. Cola Preventive Maintenance Manager
- G. Coward AGM, Technical and Administrative Services
- B. Croley Director, Technical Services l
- G. Cranston Manager, Nuclear Engineering
- S. Crunk Technical Assistant, AGM Technical and
Administrative Services
J. Delezinski Nuclear Licensing
Q. Diwan EASRTP Reviewer
J. Dowson QC Coordinator
- D. Falconer Licensing Engineer
- S. Farkas Licensing Engineer
K. Farris Mechanical PM Engineer
- T. Fetterman Supervisor, Electrical Engineering
- J. Firlit AGM Nuclear Power Production
E. Frobom System Engineer
J. Gibson EASRTP Evaluator
R. Gupta Electrical Engineer
R. Gwynn Operation
R. Gynn Operations Engineer
J. Hayes Electrical Engineer
T. Himber Assistant Shift Supervisor
J. Hoffman Electrical Maintenance PM Supervisor
R. Holler Procurefrent
- D. Humanansky EASRTP Program Manager
S. Jacobs Electrical Engineer
L. Jau EASRTP Evaluator
P. Johnson Plant Utilities Principal I&C Engineer
J. Kearns Document Control
- J. Kelly Engineering Response Team
- B. Kemper Manager, Operations Department
- D. Keuter Director, Nuclear Operations, and Maintenance
T. Kahn Supervisor, Mechanical Systems
D. Koontz EASRTP Evaluator
- Attended Exit Meeting on October 9, 1987
A-1
. . . _ _ - - - _ _ - _ - _
_ _
. .
l
'
,
APPENDIX A - (Continued) l
I
Personnel Contacted
- B. Kumar Supervisor-Environnental Qualification 4
F. Lopez Procurement )
R. Lucaro Maintenance i
R. Mayhugh MOV Project Engineer l
J. McColligan Director, Plant Support
D. McIntyre Nuclear Training Department
L. McGreyth Product Quality Engineer
S. Miller Lead Electrical I.ngineer
D. Morena Lead F.lectrical ingineer
D. Morgan Work Planning
J. Myer Procurement
- K. Nyers Manager, Nuclear Licensing
R. Nagel Nuclear Training Department
- M. Nakao Licensing (Bechtel)
T. Parker Quality Assurance Auditor
- M. Price Supervisor, Mechanical Maintenance
- T. Redican Material Management
D. Rice Systems Engineer (AFW)
M. Rojas IDADS
D. Satpathy Supervisor, Mechanical EngineerinD
P. Schwartz Nuclear Training Department
T. Shaw Supervisor-Civil / Structural
- F. Sheenan Lead Electrical Engineer
- J. Shetler Director, SRTP
- T. Shewski Quality Engineer ,
- T. Telford Engineering Response Team
D. Tipton Assistant Operations Superintendent
- R. Tomchuck Procurement
- B. Thomas Public Information
T. Tucker Shift Operations Superintendent
- P. Turner Manager, Nuclear Training
J. Vinquist Director, QA Department i
W. Weaver Procurerrent l
J. Wheeler Senior Electrical Maintenance l
D. Yount ISC/ Electrical Superintendent
- P. Zelmer EASRTP Evaluator ,
J. Zott Group Supervisor-Fire Protection l
R. Zucker Supervisor-0perations and Maintenance
W. Adachi System Design Engineer
J. Brunner System Design Engineer
- R. Nossardi B0P Group Leader
R. Patel I&C Engineer
M. Schlager MOV Engineer Project Manager ;
K. Srini-Vasiah Design Engineer
- M. Akins EA'iRTP Team Leader
J. Arevalo EA1RTP Reviewer
R. VanEschen EASRTP Reviewer
- Attended Exit Meeting on October 9, 1987
A-2
_ _ _ -
________ ___- _
.
'
. .
'
,
.
~
APPENDIX A - (Continued)
Personnel Contacted
D. Alemayehu Fire Protection Engineer
H. Grover Electrical Engineer
F. Kayas Electrical Engineer
A. Morales Engineer's Assistant ,
R. Wise Design Engineer-EQ
- F. Stock EASRTP Team Leader
R. Dobson Lead I&C Maintenance Engineer
B. Beebe NSS Principal I&C Engineering
D. Fedol System Design Engineer
- M. Lawrence Engineering Response Team
- E. Murphy Engineering Response Team
T. Marshal EASRTP Team Leader
- X. Prince EASRTP Team Leader
- Attended Exit Meeting on October 9, 1987
A-3
\ . .
.
_ _ _ _ _ _ - - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _
__ _ - _ _ _ _ _ _ _ -
.
. .
.
.
APPENDIX B
Documents Reviewed
Calculations
Z-VBS-E0523 Rev. 3 07/21/87 120 Vac Load Study
Z-DCS-E0544 Rev. 3 08/14/87 (Aux Bldg.)
7-DCS-E0600 Rev. 3 07/24/87 DC Voltage
Battery Drop (Aux, Bldg.)
Sizing
Z-DCS-E0612 Rev. 3 08/31/87 DC Short Circuit (Aux. Bldg.)
Z-DCS-E0612 Rev. 4 09/30/87 DC Short Circuit (Aux. Bldg.)
Z-DCS-E0636 Rev. 2 09/03/87
Z-DCS-E0732 Rev. 0 07/31/87
Battery
DC ShortSizing
Circuit (NSEB)
(NSEB )
Z-DCS-E0678 Rev. 0 01/04/87 DC Short Circuit (NSEB)
Z-DCS-E0683 Rev. 0 08/07/87 DC Voltage Drop (NSEB)
Z-FWS-E0718 Rev. 0 04/29/87 DC M0V Terminal Volts
Z-EDS-E0744 Rev. 0 09/28/87 Ampacity Derating for H4BAC Cable
Z-VBS-10166 Rev. 0 06/04/87 120 Vac Bus Loading
Z-FPP-E0736 Rev. 0 (Draft) Breaker Coordination
Z-HVS-M2160 Rev. 0 05/22/87 Aux. Bldg. Battery Room Temperatures
Z-HVS-M2129 Rev. O NSEB Battery Room Temperatures
Z-EDS-C0929 Rev. 0 08/07/87 Relay Magnet Mounting Evaluation
Z-FWS-10167 Rev. 0 06/05/87 Aase Calculation for Orifice Plates
Z-ZZZ-C0863 Rev 0 01/21/87 fiM smic Analysis of I&C Valves
Z-EQP-E0689 Rev. 0 01/19/87 Evironmental Analysis of Excess Flow Valves
Z-FWS-IO150 Rev. 1 07/28/87 Required AFW Flow Capability (B&W)
Z-FWS-M2237 Rev. 0 07/14/87 AFW System Head Requirements
Z-ZZZ-C0884 Rev. '2 05/04/87 Condensate Storage Tank Seismic Calculation
Z-MCM-M2212 Rev. 0 09/26/87 CST Overpressure and Vacuum Protection
Z-ZZZ-C0886 Rev. 0 03/03/87 Seismic Calculation for Field Tanks
Z-FWS-M1742 Rev. 1 09/27/87 MOV Stem Nut Strength
Z-DF0-M0119 Rev. 0 03/22/72 Diesel Generator Fuel Tank Sizing l
Z-NRW-M0245 Rev. 0 02/01/74 Spray Pond Level
Engineering Reports
ERPT-E0224 09/29/87 Overcurrent Trip Indication on Westinghouse MCCs
ERPT-E0222 09/20/87 Standby Battery Charger Cable
Studies
B&W Task 847 09/87 Rereview of Instrumentation and Control
Impell 01-0790-1619 09/09/87 4160 Yac and 480 Vac Switchgear
Impell 01-0790-1620 09/09/87 480 Vac Motor Control Centers
Impell 01-0790-1621 09/09/87 120 Vac Distribution Panels
Impell 01-0790-1622 09/09/87 Alarm and Annunciation Summary Report
B-1
,
l
'
a
. l
.
t
APPENDIX B - (Continued) l
Documents Reviewed l
EASRTP l
!
Nuclear Services Cooling Water System ;
Nuclear Raw Water System )
Emergency Die',el Generator System l
Fire Protection System l
Main Stear System 1
i
Purific& ion and Letdown System
React'.,r Building Ventilation and Hydrogen Recombiner
Au,iliary Steam System
125 Vdc System
Integrated Control System l
Control Room / Technical Support Center HVAC System
120 Vac Electrical System
Main Feedwater System ,
Non Nuclear Instrumentation System l
Seal Injection and Makeup System Radiation Monitoring System l
Procedures )
AP.4A Rev. 5 Safe Clearance Procedure Danger Tags 11/14/86
AP.23.00-23.14 Original Conduct of Operations 03/06/87
AP.26 Rev. 13 Abnormal Tag Procedure
AP.90 Rev. 1 Work and Test Authorization Program 09/21/87
MMP-0021 Rev. O Technical and Quality Determination for 03/24/87
Procurement Documents
MMP-0024 Rev. O Identification and Marking of Items 03/18/87
PEP-0025 Preservation, Storage and Maintenance of 09/14/87
Items in Storage
VEP-0026 Rev. O Four Level Procurement System 03/11/87 l
NEAP-4801 Rev. O MOV Design Process
QAP-Policy Rev. O Procurement Document Control 01/01/86
QAP-Policy Rev. O Control of Purchased Material. 01/01/87
Section VII Equipment, and Services
QAP-Policy Rev. O Nonconfonning Materials, Parts, or 01/01/86
Section XV Components
QAP 3 Rev. 2 Quality Assurance Classification 01/01/86
QAP 4 Rev. 3 Procurement Document Control 03/19/87
QAP 5 Rev. 1 Supplier Quality Assurance 01/01/86
QAP 17 Rev 4 Nonconforming Material Control 01/09/87
RSAP-0401 Rev. O Preparation and Processing of 03/06/87
Requisitions
RSAP-0704 Rev. O Stock Reclassification Requirements 06/01/87
Evaluation
RSAP-0706 Rev. O Material Control 09/01/87
RSAP-0810 Rev. O ASME Section XI Repair and Replacement 07/27/87
Program
B-2
.
. .
-
l
.
'
t
l
APPENDIX B - (Continued)
Documents Reviewed
RSAP-1201 Rev. O Procedure Change Training 04/24/87 I
TDI-3321 Rev. O Training Materials Development
A.31 Rev. 28 Diesel Generator System Operation 04/29/87
Procedure
A.31B TDI Diesel Generator System Operating 05/02/87 l
Procedure 1
A.54 Rev. 9 220 Volt AC Electrical System 11/23/87 l
A.62 Rev. 10 120 Volt AC Vital System
C.10 Rev. 2 Main Feedwater Induced Transients 08/31/87
C.143 Rev. 6 Loss of 480 Volt MCC S2E1 03/05/87
SP.300 Rev. 6 Weekly Nuclear Service Battery Pilot 08/26/87 ,
Cell Test I
SP.301 Rev. O Monthly Nuclear Service Battery Test 08/26/87
STP.1070 AFW Hot Shutdown Test 09/14/87
STP.1086 AFW Pump Flow Test with Condenser at 08/20/87 ,
'
Atmospheric Pressure
Drawings
S20E4GW-0710-000 Rev. 5 05/19/83 GM Diesel Air Start System ,
520E4GW-0425-000 Rev 5 04/03/84 GM Diesel Coolant Flow Diagram !
S20E4GW-0600-000 Rev. 3 08/13/81 GM Diesel Fuel Oil Schematic
S20E4GW-0300-000 Rev. 4 07/29/81 GM Diesel Lub Oil System i
'
M-582, Sheet 1 Rev. 24 03/14/79 GM Diesel Oil System
M-583, Sheet 1 Rev. 11 09/25/86 GM Diesel P&I Diagram
M-522, Sheet 1 Rev. 35 08/12/87 Decay Heat Removal P&ID
M-545 Rev. 18 08/20/87 Nuclear Service Cooling Water P&ID
M-544 Rev. 21 08/20/87 Nuc. Service Raw Water System P&ID
M-342, Sheet 1 Rev. 16 08/30/85 Aux. Bldg. Flow Drains
M-342, Sheet 2 Rev. 1 05/20/87 Aux. Bldg. Flow Drains
A-110, Sheet 1 Rev. 19 Aux. Bldg. Grade Level Plan
Conesco 5675-1 Rev. 2 04/08/81 Diesel Oil Storage Tank
M-585, Sheet 1 Rev. 0 10/31/86 TDI DG-Train A
M-585, Sheet 2 Rev. 0 10/31/86 TDI DG-Train A
M-585, Sheet 3 Rev. 0 10/31/86 TDI DG-Train B
M-585, Sheet 4 Rev 0 10/31/80 TDI DG-Train B
M-547 Rev. 0 10/31/86 New Diesel Fuel Oil System
E-1012, Sheets 101, 102, 114, 150, 156, 159, 171, 187, 200, and 268
Westinghouse Electric Corp. Final Assembly Drawing-Feedwater Heater
Material Requests
Bolt. Hex Head,1.00 Dia x 12.00 length 10/20/76
Bolt, Hex Head, 5/8" x 2-3/4", CB Material 02/28/77
B-3
_ _ _ _ _ _ __
..
.
'
i
APPENDIXB-(Continued)
Documents Reviewed
Documentation Concerning CB-B7 Bolt
Partial Receiving Record #55101 09/19/76
Partial Receiving Record #55117 10/09/76
Partial Receiving Record #54189 08/16/76
Partial Receiving Record #54175 08/05/76
Shipment Notice SF-81155-T3 07/30/76
Partial Receiving Record (illegible) 07/22/76
Shipment Notice SF-81155-T3 07/21/76
Partial Receiving Record #55047 07/12/76
Shipment Notice SF-81155-T3 07/10/76
Partial Receiving Record #54161 07/19/76
Shipment Notice SF-81155-T3 07/17/76
Purchase Orders
RS 35449 Nut, Hex, S/8" and Bolt, Hex Head 08/25/82
5/8" x 2-3/4", CB Material
RS 30269 Studs for Check Valve, Size 3/4" x 4-1/2" 04/21/82
RS 4476 5/8" dia x 3" long Hex Bolt w/ Heavy Hex Nuts 05/12/83
RS 46520 ASTM A193 6R B7 Stud Bolts 5/8" diameter 07/15/83
RS 16899 Studs (1-14"-8UN-2A), Nuts (1-1/4"-8UN-28) 03/06/81
Secondary Manway
RQ-87-08-55800 Nuts All Thread Rod, Cap Screws, and Bolts 06/03/87
RS 74248 Stud 4" x 8 thd x 32" long, and Nut 4" x 8 thd 06/21/87
Work Requests
- 134329 Number 1 & 2 Low Pressure Turbine
- 134330 Service Spare 550 MVA XFFR
- 134331 Service Desuperheater Inlet, Air Operated Valve
Miscellaneous
Special Order 87-19, "Operations Audits"
Training Department Memo PET 87-169
Memorandum JVV 87-055, Bulk Items Exempt from Green
Tagging Requirement (09/30/87)
Memorandum--Revised P0 Number Format. Inventory Account, and Initials
and Codes found in NUCLEUS (12/20/86)
Report No. D-0050 Engineering Action Plan plus Appendices A, B, C, and D
(09/04/87)
B-4
- _
. . . . _ . .
. . _ _ . _ __
, o
1
\
APPENDIXB-(Continued)
Documents Reviewed
Design Casis Report for ECN-R-1188, Rev. 0 (06/03/87)
USAR, Section 9.9, Fire Protection
SMUD Office Memorandum - Subject: Clark Brother Bolts ASTM A193,
Grade B-7 (10/03/87)
l
1
l
l
!
!
I
B-5