ML20148B432
| ML20148B432 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 03/02/1988 |
| From: | Ang W, Jim Melfi, Miller L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20148B418 | List: |
| References | |
| TASK-1.D.2, TASK-2.E.1.1, TASK-2.E.1.2, TASK-2.K.2, TASK-TM 50-312-88-06, 50-312-88-6, NUDOCS 8803210511 | |
| Download: ML20148B432 (31) | |
See also: IR 05000312/1988006
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d.S.NUCLEARREGULATORYCOMMISSIONL ~
REGION V
Report No. 50-312/88-06
Docket No. 50-312
License No. DPR-54
Licensee: Sacramento Municipal Utility District
P. O. Box 15830-
Sacramento, California 95813
Facility Name: Rancho Seco Nuclear Generating Station
Inspection Conducted: January 24 - February 19, 1988
Inspection by: M
W. P
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ng, Pr ec Inspector
J- /- 88
Date Signed
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- Inspettor
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Date Signed
Approved by: .
8 -[-N
I(./Pliller, Chi $' Project Section 2 Date Signed
Accompanying Personnel: D. Waters, D. Beckman, Consultants - Prisuta -
Beckman Associates
Summary:
Inspection on January 24 - February 19, 1988 (Report No. 50-312/88-06)
Areas Inspected: Routine, unannounced inspection by a region based inspector
of licensee action on previously identified inspector items, licensee event
reports, follow-up on items of noncompliance, follow-up on EASTRP corrective
action, and TMI action items follow-up. Inspection Procedures 25565, 30703,
92700, 92701, and 92702 were used during this inspection.
Results: In the areas inspected, one violation, failure to remove expired
shelf life items from stock paragraph 5. A, was identified.
8803210511 880302
PDR ADOCK 05000312
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DETAILS i
1. personnel Contacted l
- B. Croley, Assistant General Manager, Technical and Administrative
Services
- J. Vinquist, Director, Nuclear Quality
G. Cranston, Nuclear Engineering Manager
S. Crunk, Licensing Manager
T. Redican, Material Control Department Manager $
J. Browning, Incident Analysis Group -
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E. Gough, Bechtel, Nuclear Engineering Department, Supervisor
T. Fetterman, Electrical Engineering Manager
NRC Resident Inspectors
- P. Qualls
- A. D'Angelo
- Attended the exit meeting.
The inspector also held discussions with other licensee and contract
personnel during the inspection. This included plant staff engineers,
technicians, QA/QC personnel, administrative and clerical assistants.
2. Onsite Follow-up of Written Reports of Nonroute Events (92700)
A. (Closed) Licensee Event Report (LER) 80-26-T0 - NSSS Analysis
Indicates FSAR Analysis for OTSG Tube Rupture May be Nenconservative
The report identified, in 1980, that analyses performed by B&W to
support Owner's Group Abnormal Transient Operating Guidelines
(AT0Gs) for emergency procedure development showed that offsite
doses following a steam generator tube rupture could be higher than
those reported in the SMUD Final Safety Analysis Report (FSAR). The
FSAR and Technical Specification (TS) 3.1.4, Reactor Coolant System
Activity - Basis postulate worst case leakage mechanisms and
maximum allowable coolant activities that will result in limiting
the offsite whole body dose to less than 0.5 rem (the limit of 10
CFR 20 for whole body dose in a restricted area).
Since that time, the AT0Gs have been superseded by the B&W Owner's
Group Technical Basis Document (TBD) and the plant response analyses
and emergency operating procedure guidelines have changed
substantially. As a result, the new analyses indicate that the
projected doses will not exceed either the FSAR or Technical
Specification Basis values. The licensee has prepared procedures in
accordance with the TBD and new calculations. The procedures were
approved and the licensee has since issued a letter to the NRC
retracting the LER on February 18, 1988.
The inspector reviewed the Technical Basis Document, B&W Calculation
86-115344-00 (SGTR Dose Calculations for Tube Rupture Alternate
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Control Criteria) and'a summary of the changes proposed to E.06,
SGTR, Revision 8. These changes indicate that, using current
methods and analyses, the projected offsite whole body doses will be
limited to 50% (250 mrem) of those FSAR and TS 3.1.4. The licensee
has prepared both real time and worst case criteria for termination
of faulted 0TSG steaming and has included them in the draft
procedures. The inspector confirmed these' criteria are consistent
with the TBD, the above calculation, the TS and the FSAR.
No violations or deviations were identified.
B. (Closed) LER 85-16-L1/L2/LS - Spurious Closure of DHR Dropline
Isolation Valve
The LER reported spurious closure of the Decay Heat Removal System
(DHS) suction block valve, HV-20002, on August 8, August 14, and
September 23, 1985, while the plant was in cold shutdown. Licensee
investigation of the events determined that voltage spikes on
pressure transmitter, PT 21099 and PT 21092, caused the block valves
to close. Further investigation by the licensee determined that
instrument cables for the pressure transmitters were mixed in with
power cable and, by performance of tests, determined that the
voltage spikes appeared when the power circuit for motor operated
valve HV-20002 was energized.
Nonconforming Report (NCR) S-5263 and 5968 were issued to identify
the problen reported and the misrouted cables. Engineering Change
Notice (ECN) R-0459 and R-1295 were issued to correct the misrouted
cables. The NCRs and ECNs had been satisfactorily closed out. The
root cause and generic issue of misrouted cables are addressed in
the licensee's Wire and Cable Program Report which has been
submitted to NRR. The report is being evaluated by NRR and the
results will be documented in a supplement to the Restart Safety
Evaluation Report. The LER was closed.
No violations or deviations were identified.
C. (Closed) LER 86-21-LO/1/2/3 - Failure to Implement Inservice Testing
of Safety-Related Valves
The report identified that certain safety related valves were not
included in the valve Inservice Testing (IST) Program as required by
Technical Specification 4.2.2.1. This resulted in the plant being
operated with systems which were not operable as defined in
Technical Specification 1.3. Several valves were inadvertently
omitted from the procedures which implement the IST Program, while
other valves were not tested in strict conformity with NRC approved
IST program requirements. Revisions 0, 1 and 2 of the LER reported
the above noted conditions. Violation 87-16-01 was issued for
failure to properly include valves in the IST program. This was
subsequently reported by the licensee in Revision 3 of the LER.
As resolution of the above noted conditions, the licensee has
submitted for approval by the NRC a new IST program that clarifies
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the testing requirements for the subject valves and any associated
code-relief requests. This' program is the result of extensive '
reviews of ASME Section XI requirements and application of these
requirements through a detailed review of plant systems and
equipment. The closure of Violation 87-16-01 based.on-the new
program and its implementing procedures is addressed in paragraph
5.B of this report. The closure of the. violation also serves to
4 close LER 86-21-L3. The specific program and implementing procedure
changes for the valves identified in LER 86-21-LO/1/2 are as
follows:
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PLS-045 - This valve has been removed from the IST Program
! since it was no longer within the scope of'the IST program.
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OHS-059 - This valve is required to be Full Stroke. Tested Open
(FSTO) under the IST Program Plan. The full stroke test will
now be accomplished via full flow testing in SP.31C.
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DHS-015/016 - These valves are tested per SP.318 for FSTO.
This will be accomplished by full flow testing. Supporting
calcu?ations are included in SP.67 for indirect leakage
measurement technique used in this test.
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Relief Valve #9 (VR9) was submitted for relief from the leakage
i testing requirements of Section XI, IWV. Test requirements ;
reflect the Technical. Specifications. ;
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DHS-017/018 - SP.67 replaced SP.203.11 and includes a change in
the testing method for leakage monitoring. SP.67 requires
DHS-017/018 to be closed, thereby isolating the Core Flood
Tanks from the piping downstream of DHS-015/016, respectively,
in order to accurately monitor leakage.
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RCS-001/002 - FS10 will be accomplished via full flow testing
in SP.18. Flow and delta P are recorded in this procedure.
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SIM-078,079,081 - These valves require Full Stroke Test Close
(FSTC) quarterly. This will be accomplished by manually
closing these valves. These valves will be tested in SP.28A.
Prior to completion of Revision 1 to the procedure to include
this testing, the valves will be tested in accordance with
Procedure Interim Change Notice 1.
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HV-20001,20002 - Testing of these valves includes the following
tests wita the associated procedures:
Gross Leakage Test (GLT) SP.19
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Full Stroke Test Open (FSTO) SP.31B
Full Stroke Test Closed (FSTC) SP.31B
Position Indication (VPI) SP.31B
The inspector reviewed these program revisions and the implementing
procedures and verified their compliance with the requirements of
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the proposed IST program. Based on these corrective actions, this
item is closed.
No violations or deviations were identified.
D. (Closed) LER 87-14-LG - Inadequate Inservice Inspection Procedures
for Testing Operability of Pumps and Valves
The report identified three occasions in February, 1987 when
deficiencies were discovered in Inservice Testing (IST) procedures
which were violations of Technical Specification 4.2.2.1, which
requires compliance with the ASME Boiler and Pressure Vessel Code,
Section XI, "Inservice Testing," for testing of pumps and valves.
Two of the deficiencies were identified through the licensee's
review and rewrite program for surveillance procedures-and involved
a lack of measurement of pump performance data to ensure adequate
pump performance, and use of criteria which were not definitive
enough to demonstrate compliance with Section XI requirements for
pump bearing capability. The third deficiency was discovered by the
system engineer in response to a question arising from a NRC
initiated review of the Auxiliary Feedwater System, and involved
missing a required five year inspection of a relief valve. The
licensee stated that the root cause of the first two deficiencies
was personnel error arising from insufficient understanding of ASME
Section XI by the procedure writers and insufficient technical
review of the procedures. The root cause of the third deficiency was
personnel error in employing an inadequate IST scheduling procedure
which did not provide the necessary link to the existing valve
inspection history file.
Specifically, Surveillance Procedure (SP) 210.11A, "Quarterly
Concentrated Boric Acid System Pump Surveillance," did not provide
for measuring pump flow; thus, degradation in the pump's performance -
could not be tracked. Additionally, the measurement of pump bearing l
temperatures employed temperature stability criteria that were too j
wide. The same deficiencies were subsequently found in SP.210.10, i
"Quarterly Spent Fuel Cooling Pump and Valve Test." Other pump .
tests (HPI, DHR, NSRW, NSCW, etc.) were found to be ratisfactory by l
the licensee. SP.214.02, "Inservice Testing of Relief / Safety l
Valves," did not contain sufficient instructions to ensure
coordination between the computerized valve history list and the
surveillance test scheduling procedure.
As part of the licensee's surveillance procedure upgrade program and
revision to the facility IST program, the deficiencies reported in
LER 87-14-L0 have been corrected. Surveillance Procedure 210.11A
has been replaced by SP.78A/B, Revision 0, "Quarterly Concentrated
Boric Acid Pump P-705A/B and Borated Water System Valves
Surveillance." The inspector reviewed the new procedures and found
that pump flow measurements were suitably incorporatad into the
procedures. The inspector also reviewed Relief Request No. PR-5,
which requested relief from the requirement of IWP-4310 (of the ASME
Code,Section XI) to measure pump bearing temperatures. The
licensee informed the inspector that the NRC indicated in a meeting
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on January-6-7, 1988, that this relief would be granted when
Revision 4 to the IST program was approved; thus the revised
procedures do not require the measurement of pump bearing
temperatures.
An additional improvement in the IST program surveillance procedures
has been to include pump parameter tables, as specified in ASME Sec-
tion XI, IWP, directly in the procedure. Previously, they had been
contained in a separate procedure, SP.213.01, applicable to all
pumps. With this improvement, SP.213.01 will be voided.
The revised IST program does not include the spent fuel cooling sys-
tem. Justification has been provided to the NRC and was accepted
based on the January 1988 meeting for not including the system in
the program. SP.210.10 has been deleted and will be replaced by
routine test procedure SFC001 for the spent fuel cooling system,
which is in the process of being written. ,
Tc correct the deficiency in the testing of safety and relief
valves, the licensee rnplaced SP.214.02 with SP.224, "Special
Frequency Inservice Testing of Class 2 and 3 Relief Devices,"
Revision 0. This procedure contains the instruction to review the
Pressure Relief Velve Test Matrix to determine the relief valves to
be tested and specified for setpoint verification in the procedure. ,
The inspector reviewed the valve test matrix and verified that it
was a suitable tool for ensuring that the test frequency of IST
program valves could be met.
As noted above, the IST program and the relief requests were being
reviewed by NRR. The corrective actions taken by the licensee for
the deficiencies reported in LER 87-14-LO provide adequate
corrective action for the specific items. The LER was closed.
No violations or deviations were identified.
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E. { Closed) LER 87-25-LO & L1 - Decay Heat Thermal Relief Valves Did l
Not Meet Acceptance Criteria During As Found Testing
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The report identified three overpressure protection relief valves in
the Decay Heat Removal System which failed their periodic lift tests
in the as found condition. Two valves lifted below the specified
pressura and one valve lifted above the specified pressure. The ,
remaining (two) similar valves were tested per the ASME Section XI !
Inservice Test Program expanded sample requiremente and were also i
found out of specification.
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The inspector reviewed both revisions of the LER, Work Requests i
1' 133374, 127426, and 133367 for valve testing and rework, Occurrence l
Description Report No. 87-1031, Engineering Action Request No. RC
87-002, associated vendor correspondence and procurement records,
and Nonconforming Report No. S-7087 documenting the licensee's
actions in response to the failures. The licensee's root cause
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determination concluded that the high lift pressure failure was
caused by boric acid crystal interference between the valve disc
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holder and -valve disc holder guide. The valve was overhauled to
correct the existing condition. The other valves displayed apparent I
boric acid corrosion of the carbon steel spring: which reduced the
springs' cross sectional area, thereby reducing the lift pressure.
No evidence was found that the previously set lift pressures were
incorrect.. Each of the valves was repaired and retested
satisfactorily with carbon steel springs and new stainless steel
disc guides. The valve manufacturer has advised that the newly
installed springs are suitable for at least two years of service in
a borated water environment and recommends that 31655 spring
assemblies be installed. NCR S-7087's disposition requires that the
valve springs be replaced as recommanded during the next refueling
outage; the NCR will remain open until this is completed. The
licensee appeared to have adequate measures in place to assure
correction and prevention of recurrence of the reported conditions.
The LER was closed.
No violattens or deviations were identified.
F. (Closed) LER 87-27-LO/L1 - Reactor Protection System Bypass Keys Used
During Reactor Power Operation
The report identified instances during reactor power operation where
shutdown bypass keys were used to test the "shutdown bypass" ,
function of the Reactor Protection System (RPS) in violation of '
Technical Specification 3.5.J.4, which prohibits use of the keys
during power operation. One event was specifically cited as an
example of the violatien, but it is believed that numerous
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occurrences may have occurred since initial operation of the plant,
based on the RPS calibration procedures having contained'the same
instructions for testing over this period. The plant was not
operated in an unsafe manner during these occurrences, since the
shutdown bypass function was only tested for a channel which was
already in a maintenance bypass condition (allowed for calibration,
maintenance or testing of a channel). Thus, the capability of the
plant to trip was maintained as allowed by the Technical
Specifications.
Procedure I-108, "RPS Channel Test,"-(subsequently revised to
I-108A-D for RPS Channels A-D), was written initially in 1974, and
the engineer writing the procedure included a check of the shutdown
bypass trip function in order to ensure a complete check of the
channel. The procedure writer did not realize that the Technical
Specifications specifically excluded use of the keys during power
operation, and thus did not specifically exclude their use. ;
Subsequent revisions to I-108 and I-108A-D, biennial reviews by j
Instrumentation and Control engineers and supervisors, PRC reviews i
and shif t supervisor practices did not identify the use of the keys
as a violation of Technical Specifications. This was based on an
understanding by the personnel involved that testing of the shutdown
bypass function had no effect on the RPS trip logic, since the i
channel was in maintenance bypass and, in effect, out of service. '
- The failure to correct the procedures to be in conformance with the
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Technical Specifications was personnel error compounded by
inadequate compliance review against the Technical Specifications.
The licensee has completed corrective actions specified in LER
87-27-LO. These actions consist of:
(1) Revisions to procedb.es I-108A-D (Revision 26) to include a
step in the Precautions and Limitations stating that the
Shutdown Bypass Trip section of the procedure shall not be
performed while the reactor is critical and to include a
caution in Section 6.8 that the section shall not be performed
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while the reactor is critical.
(2) Revision to OP B.4, "Plant Cooldown," to require test of the
RPS Shutdown Bypass pressure setpoint prior to decreasing
pressure below 1820 psig.
(3) Submission of an amendment to the Technical Specifications to
allow testing the shutdown bypass function on a channel already
in channel bypass.
The inspector reviewed these actions and determined that they are
appropriate to prevent recurrence of the Technical Specification
violation.
The inspector reviewed a r',ot cause report prepared by the licensee
to address the events rep',rted in LER 87-27-LO and discussed the
repore with a licensee representative. -One of the weaknesses
pointed out in the report was the failure of the biennial review of
procedures to identify the deficiency in the RPS channel test
procedures. This resulted from using individuals to perform the
reviews who were technically knowledgeable but did not havc an in-
depth knowledge of the Technical Specifications. Another weakness .
was the failure of the Operations personnel and PRC members to I
recognize that Technical Specifications were indeed being violated )
by the activities conducted during performance of the procedure at
power conditions. The corrective actions resulting from the ]
investigation were reported in Revision 1 to the LER, and consist
of:
The Biennial Procedure Review process will be evaluated to
ensure that the reviews include an in depth review of the
Technical Specifications and other governing documents.
l This incident will be included in the ongoing Licensed Operator
Reading Assignment training log and Plant Review Committee
training as an example of the need to follow the specific
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wording of the Technical Specifications or, when the specific
wording cannot be followed, to initiate a Technical
Specification revision.
The licensee representative stated that a comprehensive review of
the Technical Specifications was conducted to determine if other
specifications could result in violations similar to the one
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reported in.the LER. The review concluded that this was an isolated
occurrence. Based on the the corrective actions completed and the
corrective actions committed to in the LER revision, this item is
closed.
No violations or deviations were identified.
G. (Closed) LER 87-31-L0 - Visual Inspection Surveillance Test of
Hydrogen Purge System Not Performed
In May, 1987, the licensee found that visual inspections of the
Reactor Building Hydrogen Purge system had not been accomplished on
a refueling interval in accordance with TS 4.4.3.1. The Hydrogen
Purge System is planned to be retired in place prior to restart
pending NRC approval of Proposed License Amendment No.158 which ,
addresses new hydrogen recombiners installed by ECN R-1570 to
improve the containment hydrogen control capability. The LER
further noted that, should approval of Proposed Amendment (PA) No.
158 not be issued prior to restart, the original, defective
surveillance procedure would be corrected and reissued. However,
this PA has been identified as mandatory for issue prior to restart-
by both NRC and the licensee.
The inspector reviewed PA 158, existing TS 4.4.3.1, new SP.45A,
Semiannual Hydrogen Recombiner Train A Functional Test, and a draft
Incident Assessment Group (IAG) evaluation memo for the event. All
committed and required licensee action for this item appears.either-
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complete or incorporated into the Coordinated Commitment Tracking
System. This item was closed.
j No violations or deviations were identified.
H. (Closed) Special Report 83-09-X0 - Thermal Hanger Shields Not
Installed on Seismic Class I Pipe Supports !
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The report identified a nonconformance between the design and
installation of Seismic Class I pipe supports of the Hydrogen
Monitoring System. The design called for the installation of
thermal hanger shields between the system piping and the supports as
a means to minimize localized heat sinks along the the piping runs
but still provide Seismic Class I support canahi!!ty Instead, 45
of the supports were not installed per design, in that standard
fiberglass insulation was installed in place of the shields, thus
the system did not meet its seismic design requirements. The
nonconformance was caused by a lack of sufficient detail of this
specialty item in the design drawings and a lack of communication in
the field during construction.
The specification of the thermal hanger shields for this
installation was based on cost-effectiveness and specific
requirements for heat sink avoidance in combination with pipe
loadings for the hycrogen monitoring ystem, rather than a generic
requirenent to use thermal shields on all heat traced piping or
refrigecant piping. The thermal shields are just one of many
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methods employed by the licensee to minimize heat transfer between
piping and supports. Thus, the nonconformance was specific to this
installation.
The licensee corrected the nonconformance by installing 44 of the '
hanger shields by January 10, 1984 and the remaining hanger shield
by March 26, 1984. Locations in other systems where thermal hanger
shields were installed were verified by the licensee to be in
conformance with design requirements as a result'of the reviews
conducted in accordance witn IE Dulletin 79-14. This item is
closed.
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No violations or deviations were identified.
3. Generic Letters
A. (Closed) Generic Letter (GL) 85-22 - Potential for Loss of Post-LOCA
Recirculation Capability Due to Insulation Debris Blockage
The GL identified the potential for a primary coolant pipe break to
damage thermal insulation that could then be transported to the
Reactor Building Emergency Sump and cause blockage. The GL required
no response or specific action but recommended that licensees
consider it and the guidance of Regulatory Guide 1,82 while making
design changes involving thermal insulation within containment. The
GL also provided several sensitivity criteria for sump debris screen
size, ECCS pumping requirements, and ECCS NPSH margins.
The SMUD Nuclear Licensing and Nuclear Engineering Departments had
reviewed the GL and documented the following. The Performance
Analysis Group (PAG) had reviewed the item and assigned it a f
Priority 2 for completion of final actions af ter restart but prior
to the next refueling outage. CCTS Item T8611063800 has been issued
to Nuclear Engineering for the actions recommended by the Generic
Letter. Nuclear Licensing had yrepared a justification memo to
supplement the initial PAG justification for Priority 2 scheduling, ,
The Nuclear Licensing justification memo addressed the type of I
insulation predominant inside the reactor building, discussed
calculation Z-DHS-M2064 for sump performance with regard to the
sensitivity items in the GL, and concluded that deferral of the GL
acommendations was acceptable. The facility design fell outside
tne high sensitivity ranges identified by the GL. Nuclear
Engineering has assigned the revision of Design Criteria 5109.5,
Piping and Equipment Installation to a contractor for revision as
per the above schedule and the GL recommendations. The inspector
reviewed the licensee's schedule and justification finding it
reasonable and acceptable with respect to the GL. This item is
closed.
No violations or deviations were identified.
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B. (Closed) Generic Letter (GL) 86-05 - Implementation of TMI Action
Item II.K.3.5, Automatic Trip of Reactor Coolant Pumps
The licensee's f ritt41 response to the GL was reviewed during
Inspection 50-312/87 21 which founa discrepancies in the licensee's
internally prepared instrument uncertainty calculations, the NSSS
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vendor's calculations and emergency procedure operating curves.
These discrepancies affected the valfdity of the licensee's' prior GL 86-05 response to NRC and the accuracy of the Reactor Coolant Pump
Trip Criteria and operating curves.
Since that inspection, the licensee has obtained a new set of NSSS
developed instrument uncertainty data (B&W Calculation
32-1170997-01, P &.T-hot Errors of SMUD SPDS VSM Curves, Revision 1)
to replace the internally generated calculations previously
reviewed. The licensee has generated new emergency procedure
subcooling margin and pressurized thermal shock curves for
incorporation into the emergency operating procedures. The
inspector reviewed the above calculation, curves, and transmittal
letter (B&W Letter No. SMUD 88-024, dated January 20, 1988) and
discussed them with cognizant I&C and Operations personnel
confirming that the calculations represent as-built plant equipment
and post-accident conditions.
The remaining licensee actions to complete the disposition of GL- 86-05 were 1) incorporation of the new curves into the emergency
procedures and 2) submittal to NRC of a revised response reflecting
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the revisions to RCP Trip Criteria data submitted previously in
response to Section A.2 of the GL. These items were verified by the
inspector to be in progress and were being controlled by CCTS Item
T860910104 and were scheduled to be completed prior to restart.
This item was closed.
i No violations or deviations were identified.
4. Follow-up of Previous Inspection Items (92701)
A. (Closed) Inspector Follow-up Item ,86-13-02 - Lack of Proper
Corrective Actions when Identified Valves not on P& ids
The licensee has implemanted a program to walkdown sixteen selected
i balance of plant systems to identify and correct disagreements
between the as built system configuration and the piping and
instrumentation drawings (P& ids). Problems with the process
identified by QA surveillances resulted in reperformance of the
walkdowns for thirteen of the systems. As a result of these
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collective issues, the licensee was requested by NRC Region V to
! take a more generic approach to these configuration management
issues. Licensee progress on this program was previously reported ,
in NRC Inspection Report 50-312/87-13 (April - May, 1987). At that
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time the licensee's program scope and processes were reviewed, the
walkdowns were confirmed to be substantially complete, but the plant
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drawings and procedures had not yet been updated; the item was held
y open pending NRC verification of this last action. Since the prior ;
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inspection, the licensee has expanded the scope of their effort to
include additional balance of plant and NSSS systems and skid
mounted equipment. These additional scope items are scheduled for
completion over the next 22 months asipart_of the EASTRP effort.
The licensee's actions to correct drawings and procedures for the
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original sixteen systems were reviewed during this-inspection. The
inspector reviewed licensee status memos EMW 88-007, January 19,
1988; NL 88-017, January 7, 1988; NOM 87-122, May 27, 1987; and, NOM.
87-71,-March 23, 1987. Memo NOM 87-122, W. Kemper to J. Vinquist,
May 27, 1987, reported that all discrepancies for the initial
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walkdowns had been corrected. It further provided actions to be
taken after completion of the initial sixteen systems to treat the
plant wide issue of system configuration and identified the
procedure, lineup and drawing changes to be completed prior to
restart.
The inspector selected a sample of identified discrepancies from the
NSW, NRW, AFW, and MFW systems and confirmed that P& ids and plant
lineup procedures had been corrected. _The inspector also verified
that the omitted instrumentation valving noted in the unresolved ,
item was now included on P&IO M-599, Sheets 1 -12, Instrument Root '
Valves. The valve configurations in the drawings and procedures !
were confirmed by field inspection of .the installation. No
discrepancies were identified; the licensee's processes appear to be
functioning acceptably to provide reasonable assurance that the
balance of plant systems will be similarly completed. This item was
closed.
No violations or deviations were identified.
5. Follow-up on Items of Noncompliance (92702)
A. (Closed) Violation 87-06-05 - Quality Assurance Requirements for the
Control of Materials
The violation identified that the licensee had no system to
implement procedure AP 605 requirements regarding sorting of shelf
life items such that the oldest items were in front or on top, nor
had a manual or computerized tickler file system been in.plemented to -
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control special storage requirements. Fqrthermore, the violation ;
identified that expired shelf life items were not being reviewed or I
removed on a monthly basis. l
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The licensee responded to the violation on May 7, 1987 (GCA 87-001) l
and stated that the following corrective action had been taken or -)
was planned to be taken: '
(1) A computerized materials management information system (MMIS)
for support of the material management organization began
, addressing shelf life in April 1987, with pertiner.t data such
j as expiration date flagged in the database and displayed on the
label affixed to the material,
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(2) Routine report RSMM 0330, "Stock Items with Expiration Date
This Quarter," will be produced by MMIS and distributed monthly
to warehouse personnel. The report identifies material with
expired shelf life such that it can be removed from stock and
I
placed in the appropriate quarantine hold area for disposition. '
(3) To ensure proper rotation of stock, additional training in
procedure AP 605 would be provided to warehouse personnel.
Additionally, a color coded "chrono flag" system would be
implemented to require colored "flags" be fastened to the
material indicating how the material is to be rotated and thus 1
make those warehouse personnel responsible for putting the
material away more conscious of the need to rotate the material 4
in compliance with AP 605. Installing the flags to materials l
purchased before April, 1987 would be accomplished by
September, 1988. l
1
(4) A material reverification / stock baseline task was being '
implemented. Those materials with shelf life, or those 1
suspected of having short shelf life, would have hold tags
placed on them until materials analysts could verify the status
of the materials. Materials which had future shelf life would
be identified and entered into the MMIS for control. Materials
found to have expired shelf life would be quarantined and
dispositioned. Completion of the program was scheduled for
September 1988. In the interim, any requested stock items with
hold tags would be evaluated for shelf life exph oi, ion prior to
issuance.
The inspector discussed the Materials Management program and
particularly the implementation of the corrective actions described
above with the licensee. Significant improvements had been
implemented in the past year in many aspects of the program. All
aspects of the corrective actions were characterized by the licensee
as having been implemented as committed or on schedule to be ,
implemented by September 1988. As part of the program changes,
procedure AP 605 has been deleted. The licensee supplied the )
inspector with a copy of MMP-0025, "Preservation, Storage and 1
Maintenance of Items in Storage," which contained requirements for l
stock rotation and removal of items with expired shel; life. l
l
The inspector reviewed training records for training on Procedure AP l
605 and also training records for Procedure MMP-0025 and confirmed
that personnel were trained as required.
The inspector conducted several inspections of Warehouse A, I
Warehouse B and the GR Warehouse to observe the licensee's
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compliance with its commitments. The licensee supplied copies of
MMIS report RSMM 0330 dated December 1, 1987 for materials with
expiration dates between 0:tober 1,1987 and December 31, 1987 and a
report dated January 4, 1988 for materials with expiration dates
from January 1, 1988 to March 31, 1988. The report for the end of l
1987 was chosen for inspection since it showed the disposition of l
the expired items by the warehouse personnel.
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The first area inspected was the use of the colored ~"chrono flags."
The inspector observed these affixed to.most of the parts or parts
tags. The flags consisted of colored dots with different background
colors for different years as well as a year designation printed in
black for ease of differentiation. The use of this marking system
was implemented in mid-June, with newly purchased items being marked
as they were stocked in the warehouse. However, five shelf life
stock items were found with purchase dates later than' June 1987 but
without "chrono flags" attached. The items were rubber coated '
flexible hose, stock numbers 027521, 019921, 019922, 019923, 019924.
None of the items were safety related. All items had a proper MMIS
label affixed which included the expiration date. These
deficiencies were brought to the attention of the licensee, who
committed to apply thr. correct flags to the noced stock items, and
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to review remaining stock items and correct any additional
deficiencies. This review resulted in 20-30 additional stock items
which required the placement of flags.
The licensee indicated that no formal procedure existed for use of
the "chrono flags" as a means to implement the stock rotation
requirements of MMP-0025, and committed to provide a Department
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procedure for the use of warehouse personnel to allow a uniform and
comprehensive approach to the application of the flags. The
inspector also reviewed a material stocking and movement form that
warehouse material handlers'have recently begun to use which should
assist in more uniform implementation of this commitment.
The inspector noted that the licensee had not yet determined what
type of flag to attach to old items (pre-April 1987 purchase date)
or how to distinguish between different lots of shelf life material
purchased within the same year.
An inspection of the material quarantine portion of the GR Warehouse
was performed. Items which have been removed from stock due to ex-
pired shelf life or other requirements were observed. The use of
yellow hold tags on stock items with indeterminate shelf life which
were awaiting removal or reverification was also observed. No defi-
ciencies were noted in these areas.
Procedure MMP-0025, Step 7.6.5, required that "Items whose shelf
life has expired shall be removed from their storage location and
i placed in a ' Hold' location." Materials Management personnel issued
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quarterly reports listing items with expired shelf lives that were
in stock and was to be removed. The implementation of the program
to remove stock items with expired shelf life as indicated on report
RSMM 0330 was inspected. This report was issued by Materials
Management Department personnel based on information in the MMIS
.
system. No formal review of this report by Department management or
! by QA/QC for accuracy was apparent to the inspector. The inspector
i selected 16 stock items from the first quarter 1988 report (dated
1 January 4,1988) that had expiration dates up to and including the
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date of inspection (February 4,1988). Of these 16 items, 12 were
still in their bin locations, 2 had been removed to salvage eleven
days after their expiration date, and 2 could not be found in the
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reported bin location although they were listed as being available
for issue (i.e., not salvaged). One of the items was subsequently
found in another bin location, and the other item was confirmed as
being issued from stock prior to the expiration date. Three of the
twelve items with expired shelf lives were confirmed to be Class 1
(safety related) material. No hold tags were in place on any of the
items to prevent the issue of these items from stock.
Discussions with warehouse personnel disclosed that licensee
warehouse personnel were processing the removal of these items
"after the fact." Although the possibility existed for expired
items to be issued prior to its being removed to salvage, none of
the items noted by the inspector had been issued. Furthermore, the
licensee's QA Department management stated that QC personnel verify
that safety related parts meet all requirements prior to
installation, including shelf life.
When the above noted items were brought to the licensee's attention,
they immediately removed the items found by the inspector, and
committed to conduct a review of all expired stock items since May
1987 to determine if any items had been inadvertently issued for
use, and to instruct those responsible for removing expired items
from stock on the importance of clearing the items prior to the date
of expiration. The licensee's review of expired stock items covered
18 stock codes contained on various issues of RSMM 0330 between
June 30, 1987 and December 31, 1987. Eleven items were found to be
still in place, of which four were Class 1 safety grade. Of the
remaining seven items, two Class 1 items were removed to salvage 4
weeks after the date of expiration. The licensee confirmed that
none of the 18 items had been issued past the expiration date. The
eleven items that were found in the stock locations were removed for
disposition / reverification. Some of these items had questionable
expiration dates (i.e., expiration date was listed on the stock tag
as 6 days earlier that the purchase date), which suggest data entry
problems during the receiving process.
The failure to remove items with expired shelf lives from stock ap-
pears to be a violation of MMP-0025, Paragraph 7.6.5, and was
identified as Violation 88-06-01 - Failure to Remove Expired Shelf
Life Items from Stock. It is similar to violation 87-06-05.
Violation 87-06-05 will be administrative 1y closed since corrective
action for Violation 88-06-01 should provide for 87-06-05 also.
B. (Closed) Violation 87-16-01 - Failure to Implement Inservice Testing
of Safety-Related Valves
This violation involved failure to: 1) full stroke test Reactor
Coolant Pump Seal Injection isolation stop check valves SIM-019
through -022 and 2) measure, record and analyze stroke time data for
Main Turbine Throttle Stop (trip) valves TV-1 through -4 in
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accordance with ASME Section XI and the licensee's NRC approved
Inservice Testing Program.
The licensee's response letter, GCA 87-244, dated June 26, 1987,
identified the following corrective actions. LER 86-21, Revisions 1
through 3 were issued to report the conditions as required by 10 CFR
50.73(a). The licensee had submitted and was awaiting NRC approval
on a new Inservice Test Program for the'next 10 year operating
cycle; this Program included clarified code relief requests for the
subject valve testing. Unrelated prior testing of the SIM valves
was found to have satisfied the prior cited program requirements and
was determined by the licensee to be an acceptable alternative to
the testing not provided for by the defective procedure. New
surveillance procedures have been written for the TV valves for
performance when the plant returns to power.
The inspector reviewed the status of the licensee's IST submittal,
finding it to be as stated. Interim NRC approval for the program
(to support plant restart) is expected during February,1988. The
inspector reviewed the results of SP.422A performed on November 5,
1987 which incidentally exercised SIM-091 through -022 as part of
containment local leak rate tests and confirmed that it satisfied
the test requirements of the Program as stipulated by the licensee.
Further the licensee has documented the use of the alternative data
by Memo IST 88-108 to the test files for the valves. This memo and
the licensee's test matrix and schedule further indicated that a new
procedure will be written for testing of the valves prior to the
next scheduled test upon receipt of IST Program and relief request
approval from the NRC. SP.77, Special Turbine Throttle Stop Valve
Fail-Position Test, Revision 0, and SP.75, Monthly Turbine Throttle
Stop Valves Surveillance Test, Revision Original, were reviewed and
found to meet the requirements of the proposed IST Program and were
tcheduled for performance during plant startup. The licensee
actions were found to be in accordance with the response letter and
acceptably implemented to date. This violation was closed.
No new violations or deviations were identified.
C. (Closed) Violation 87-21-01 - Failure to Perform Cable Routing
Inspections
Violation 87-21-01 identified that independent inspection for proper
location and routing of approximately 1100 safety-related cables
installed during the 1983 to 1985 timeframa was not performed. On
August 28, 1987, the licensee submitted its response to the Notice
of Violation. The licensee admitted the violation and provided its
corrective action.
Licensee corrective action to preclude recurrence included several
changes to its engineering, inspection, and routing procedures.
These changes had been previously inspected (Inspection Report
50-312/87-21) and determined to have provided reasonable assurance
to preclude recurrence of the identified violation. In addition,
the licensee also stated, in its response to the violation, that
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completion'of the cable walkdown portion of the Wire and Cable
Corrective Action Program will complete the corrective action for
this violation. The cable walkdown was performed by the licensee on
100% of cables that had been installed and rerouted since original
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construction and on 142 of 1559 newly routed cables. NRR reviewed
the licensee's Wire and Cable Corrective Action Program including
specifically the cable walkdown portion of the program. By letter
dated January 27, 1988 (G. Holahan, NRR to D. Kirsch, RV), NRR
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concluded that based on the results of the cable >;alkdowns, "the
likelihood that newly installed cables are routed unsafely is
sufficiently low and that inspection of the remaining 1417 cables is
unwarranted." The violation was closed.
1
No new violations or deviations were identified.
D. (Closed) Deviation 87-11-01 - Failure to Implement Corrective Action
for NRC Violation Regarding Performance of Tests with Instrumen+.s
Not Calibrated
1
NRC Inspection 50-312/85-23 found that surveillance test acceptance l
criteria data was being obtained from instruments which had not been '
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calibrated as part of a controlled calibration program. The i
licensee's actions in response to that violation were reviewed '
during Inspection 87-11 and were found incompletely implemented in i
that a licensee's commitment to revise, prior to January 1, 1987, l
all surveillance procedures to require documenting of calibration I
data for instrumentation (except control room) used for testing had )
not been completed.
The initial licensee resnonse to this Deviation (SMUD letter GCA
87-033, May 18, 1987) provided corrective actions and preventive
actions which addressed revision of surveillance procedures prior to
plant restart and actions to prevent missing future commitments to
NRC. NRC Region V responded via letter dated June 3, 1987 noting
that the licensee's initial response did not address surveillance
tests that would be performed currently, prior to restart. A
licensee supplemental response was subsequently issued on
September 8, 1987 which stated that a special data sheet would be
inserted into those surveillance procedures performed prior to their
, revision to include the long term corrective actions.
1
l' Since submittal of the deviation response, the licensee has
implemented and subsequently upgraded the actions committed therein
to improve commitment control and status information regarding NRC
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commitments provided to senior management. The inspector reviewed
CCTS Monthly Management Commitment Statistical Reports, Weekly
4 Commitment Restraint Matrix Reports, semiweekly Project Managers
Meetings Reports, and various memoranda from Nuclear Licensing to
the CEO Nuclear reporting the status and followup actions for
internal commitments and commitments to NRC. These aspects of the
licensee's actions appear to exceed the licensee's initial
commitments and appear to be consistently implemented.
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The inspector reviewed a sample of six "new" revised surveillance
procedures and six "old" format procedures, including performance
data, to confirm that the provisions of the licensee's response had
been implemented. The "new" procedures were confirmed to have the )
calibration data provisions required by AP.2.26, Surveillance Proce- l
dures - Description and Format, Revision 0. Section 4.7. The proce- I
dures further conformed to Guidance Memo DPS87-159, dated l
September 15, 1987, which provided amplification of the requirements l
of AP.2.26 to all procedure writers. The "old" procedures were !
confirmed to have the interim calibration data sheets included and
completed. The licensee is continuing to revise surveillance
procedures in support of new/ proposed Technical Specifications,
pending IST Program changes, etc. The licensee's program to assure
completion of these activities in accordance with the provisions of
the above response appeared adequate.
During this review the inspector noted that SP.68, "Refueling
Interval DHRS and RBS System Leakage Rate Test (Revision Original),"
involved intersystem leakage testing. Permanently installed pump
suction gages PIs 29201, 29202, 26143, and 26144-were used to
establish test pressure conditions but were not included in the
procedure's list of instruments for which calibration data
verification was required. Although these instruments were not
specifically used to rieasure acceptance criteria and therefore fell
outside the guidance of AP 2.26 and DFS87-159, the gages were used
to establish requirea conditions for satisfactory completion of the
test.
Upon identification by the inspector, the licensee reviewed thirty
one additional SPs which had the potential for similar test
configurations and, therefore, similar omissions, finding no further
examples. The licensee further initiated corrections to the
guidance information to clarify the need to include instruments such
as the above in the procedure calibration data sheets. The noted
condition appeared 7.o be an isolated case.
The licensee actions in response to the deviation and the additional
inspector observations was found to be adequate. This item was
cicsed.
No new violaticns or deviations were identified.
6. Review of Expanded Augmented Systems Review and Test Program (EASTRP)
Items
In late 1987, the licensee conducted an EASTRP evaluation of selected
plant systems and activities. Review of SMUD action to disposition
findings from the EASTRP was initiated during Inspection 50-312/88-03 and
completed during this inspection period.
The licensee has categorized the EASTRP findings as either Restart Scope
List (RSL - to be complete prior to plant restart) or Long Range Scope
List (LRSL - to be done after restart). The licensee's actions and
dispositions were reviewed, on a sampling basis, with respect to the
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requirements of RSAP 0215, RSL/LRSL Development and Administration,
Revision 0, PCN 2, and RSAP 0218, Long Ranch Schedule Work Item
Justification, Revision 0, which jointly' provide the criteria for '
- characterizing and justifying the disposition of the items.-
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Restart Scope List Items
An initial sample of EASTRP finding Requests. for Information (RIs) ,
identified for completion prior-to restart (RSL items) -were selected for
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general review to determine that the licensee's proposed corrective
actions and schedules appeared reasonable and would be accomplished prior
to restart. Each RI was comprised of several RSL or LRSL items and a
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sample of individual items was further selected for more detailed review
of actions in progress and discussions with licensee personnel. ,
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RI No. Subject
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005 Fuse Control !
079 Human Factors
153 HPI/ Makeup Pump'NPSH
003 DHS Pressurizer Auxiliary Spray Testing
128 NSW Design Capacity
099 NSRW Heat Exchanger Plugging ;
027 NSRW Pond Recirculation and Filtration l
NSCW Radiation Monitor Piping Classification
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132 NSCW Pump Casing Overtorque
001 EDG Jack Water NSRW Flow Measurement !
056 NSCW Heat Exchanger / Spray Fozzle Differential i
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Pressure
109 NSCW Hydro Pressure Requirements
Long Range Scope List Items
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Items scheduled for completion after plant restart (LRSL items) were
selected for review of the adequacy of the licensee's justification for ;
deferral of action until after plant restart. Initially, twenty three i
items were selected for sampling review. Of those initial LRSLs, the i
licensee had previously completed all actions or had upgraded ten of the
LRSL items to be completed prior to restart. Another five of the items
i
below had been addressed by interim licensee actions not reflected in the
l RSL/LRSL data.
,
RSL Items Downgraded to LRSL Items
Additionally, the licensee had previously reviewed and downgrad2d some 1
I RSL items to LRSL to be completed after restart. Thirty such LRSL items '
downgraded since December 9, 1987 were selected for review to confirm
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that the licensee's justification to be in accordance with RSAP 0215 and
0218. The status of these items was found to be generally consistent
with RSAPs 0215 and 0218 and appeared reasonable, on that basis, to <
adequately support plant restart. About 25% of the 65 items were l
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reviewed in some detail to confirm that the scope and schedule of
individual licensee response actions agreed with those proposed.
3
Specific observations or inspector clarifications are provided below:
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A. Inspector review of the RI 035, Fuse Control, found that the
licensee had completed the engineering review of Class 1 4160 vac
and 480 vac applications and had issued ECN R-2561 to correct
drawings and replace improperly installed fuses. One hundred twenty
three drawings were revised by the ECN to correct fuse applications.
The actual system walkdowas and fuse replacements per the drawing
revisions were planned for completion during the bus outages
resulting from STP 961, Loss of Offsite Power Test, currently
scheduled for late February, 1988.
Completion of the remaining Class 1 and non-Class 1 equipment is not
specifically scheduled but is planned to be an uninterrupted
continuation of the current Class 1 switchgear activities. The
inspector reviewed the ECN and the scope of the program segments,
finding the licensee's plans to be acceptable.
B. The licensee had previously advised NRC that LRSL items scheduled
for post restart would receive earlier action as resources were
available and that some were expected to be closed prior to restart
(although this was not a formal regulatory commitment). The
licensee appears to have made substantial progress in this regard
with a large portion of the LRSL items reviewed during this
inspection having been either closed or provided with positive
interim measur:s as discussed above. Most of these appear to have
been accomplished on the initiatives of the individual departments'
management and were found to be meaningful enhancements pending the
long term actions.
Other items had been reevaluated by the cognizant personnel and
found to be somewhat different than the problem scope or description
initially reported by the EASTRP project team. As a result, the in
process resolutions varied from those shown on the licensee's status
keeping and control documents for the program. Similarly, the
interim actions previously discussed above were not shown. Examples
include:
(1) RI 056 and RSL 753 involve revision of surveillance procedures
for NSCW system delta P measurements. The draft procedure
revisions were not responsive to the problem identified by RI !
056. The licensee advised that the RI was incorrect and would
be revised.
(2) LRSL Item 2620 identified problems with operator awareness of l
radiation monitor oporability status. Interim pre-startup l
compensatory measures were in place but were not reflected in '
the LRSL Item.
(3) LSRL Items 2308 and 2309 identified that Auxiliary Building
Supply Fans SF-A-1 and -2 were not electrically interlocked
with the corresponding exhaust fans to assure that an exhaust
fan failure / shutdown would not cause a positive building
pressure and exfiltration.
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Accompanying Item 2310 appeared to provide compensatory actions
by requiring that administrative controls be established to
ensure that operators manually shutdown.the supply fans on
stoppage of the exhaust fans. However, this item had been
downgraded from pre-restart to post-restart.
Operations Department management advised that Item 2310 was
misleading but was intended to address the procedure changes-
necessary to incorporate the long term design modifications
into the system operating procedures after installation and was
not intended to address the compensatory procedures. Although
not addressed by Item 2310, the licensee had prepared draft
revisions to A.14, Auxiliary Building Ventilation Systems,
which will be issued prior to restert and will provide the
compensatory actions noted above.
(4) LSRL Item 2284 involved the need for emergency grab sample
procedures and periodic verification of grab sample capability
for the Auxiliary Building and Reactor Building ventilation
stacks. The LRSL item was prioritized for completion after
, restart but the procedures __had been prepared and initial tests
scheduled and was not reflected in the LRSL status.
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(5) LRSL Items 3861 and 3971 identified improper removal of
deficiency tags from equipment after work request completion.
These items were listed as post restart priorities. Interim
actions had been initiated for plant restart including an audit
of control room deficiency tags (in progress on January 26,
1988), additional emphasis and direction given to shift
supervision, and plans for additional interim changes in
written guidance. More substantial program changs are being
planned for post restart. This status was not reflected in the
- current LRSL documentation.
On January 28, 1988, the inspector was provided with Memo FS88-004
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documenting actions taken to correct the RSL and LRSL data base
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information for those items discussed above which were of a
different status or content than shown in the data bases. In
addition, the licensee indicated that this matter would be further
reviewed to ensure that the data bases were accurate with respect to j
the above. :
!
C. RI 132 identified nonconforming work performed on both NSCW pumps in
that a) the pump casing bolts were overtorqued by as much as a
factor of three, b) a casing gasket two times thicker than specified
had been used with a potential for pump casing deformation (per a
pump vendor representative), and, c) a hand written step had been
added to the work requests to repair pump mechanical shaft seals but
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had not been signed off.
The inspector was initially presented with NCR S-7014 and work l
request 01397620 which appeared to adequately disposition issue a) l
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above regarding bolt overtorquing. These documents did not,
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however, address the gasket and shaft seal issues of b) and c) above
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but'had apparently been submitted as the closure documents for RSL
Item 3855.
During a meeting with the cognizant Maintenance Department
supervisor and engineer on January 26, 1988, the inspector reviewed .
MAP-17 Root Cause Evaluation No.87-009, dated August 31, 1987,
prepared by the Maintenance Department to document their evaluation ,
of RI 132. The evaluation appeared to be extensive and detailed but
was silent with respect to the hand written step for shaft seal
repair. The licensee representatives advised the inspector that
their investigation could not identify the source of the RI finding,
that no such note was found in any of the referenced work request
packages, and that the seals had, in fact, been overhauled using
other work requests unrelated to those reviewed by RI 132. This ,
finding was therefore not addressed in the department's evaluation
or response.
The evaluation also noted that the on duty maintenance engineer on :
July 17, 1987 had contacted the pump vendor engineering
representative and had obtained telephone confirmation that
installation of a 1/16 inch gasket in lieu of the specified 1/32
inch gasket was acceptable and would not result in casing
misalignment or deformation. The licensee's evaluation further
acknowledged that the vendor telephone communication had not been i
formally documented but the engineer's hand written, plain paper
notes of the conversation were available. The evaluation further
opined that the EASTRP team member, may have obtained erroneous
information from onsite vendor representatives who were not
authorized or engineering qualified to comment on or disposition i
such matters.
The inspector reviewed work requests 1352030-1 and 1334420-0 which 4
documented the subject work on NSCW Pumps P482A and B respectively. ]
The work on both pumps was accomplished in July - August, 1987 and i
was categorized as Safety Class 1, Project Class 1 by the work j
requests. Both work requests included steps to install the 1/16 ;
inch gasketing material, including Stores Issue / Reservations '
Requests for three different types of Garlock brand 1/16 inch
material. In both cases, the installation was verified to meet the
provisions of the work request steps by signed off Quality Control
inspection points. The inspector also confirmed that B&W Pumps
Instruction Manual No. M29.03-IM10, Revision 2, Pump Data Sheet, for
the subject pumps specifies that Crane No. 888, 1/32~ inch thick
gasket material be used for the casing gasket.
On January 27, 1988, the inspector requested additional information
from Maintenance Department management regarding the acceptability
of the resolution of the EASTRP finding in that it did not appear to
adequately address the configuration management and material control
issues involved, i.e., the casing gasket was replaced with a
material of different form, fit and substance. NCR 57221 was issued
by the licensee to address this concern. Nuclear Engineering
disposition of the NCR was that the material substitution for the
pump casing gasket was acceptable. In addition, the licensee was
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evaluating its process for identifying when Nuclear Engineering
evaluation / acceptance would he required for material replacements. .
Pending completion of licensee actica, this item was identified as
follow-up item 88-06-02.
The QA Surveillance Group performed a surveillance of the RSL/LRSL
closure processes. At the close of this inspection, surveillances
had been completed for the. Nuclear Engineering and Operations
Departments, the SRTP Project, and a comparison of EASTRP Vs. RSL.
Items. Surveillance of Maintenance Department EASTRP items had been
in progress for about one and one half weeks and was continuing.
The inspector reviewed the findings from the completed surveillances
finding that the percentage of RSL/LRSL items sampled was relatively
small (about an average of 5% of each group's assigned items) but
was yielding a reasonable number of findings and observations.
Additional RSL/LRSL surveillances are planned for the other plant
departments.
In conclusion, the processes of licensee prioritization, followup
and closure of EASTRP findings appear, from the sample reviewed, to
be proceeding acceptably.
No violations or deviations were identified.
7. Follow-up on TMI Action Items
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The areas inspected in this report were modifications to the licensee's
Safety Parameter Display System (SPDS), and Auxiliary Feedwater (AFW) i
system. ;
The inspector reviewed previous inspection reports, NRC documents
specifying inspection requirements,-and licensee documentation on these
items. The statas of these THI items are not2d below.
A. (0 pen) Item I.D.2.2.3, "Plant Safety Parameter Display Console,
Fully Implemented"
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This item required the addition of a Safety Parameter Display System
(SPDS) in the control room to display a minimum set of plant
parameters for use by operating personnel (NUREG-0737). This
provides a redundant system to help rapidly assess how the plant is 4
performing. The important plant functions that this system displays
include: (1) Reactivity Control, (2) Reactor Core Cooling and heat
removal from the primary system, (3) RCS integrity, (4) Radio-
activity Control, and (5) Containment Conditions. The critical ,
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functions of the SPDS (Regulatory Guide 1.97 Category 1 parameters)
shall be provided during and following design basis events
(NUREG-0696). The functional criteria of the SPDS was described in
NUREG-0696 and NUREG-0835.
This TMI item was previously inspected in NRC Inspection Reports
50-312/85-09, 50-312/86-39, and 50-312/87-14. From a review of
these reports, the following items remained to be inspected.
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(1) Examination of the final turnover to operations of the SPDS.
(2) A review of procedura and Technical Specification Changes due
to the SPDS modification. ,
(3) Evaluation of NRR audit results of the upgraded SPDS performed
September 19, 1986.
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(4) Inclusion of certain variables included in the SPDS as noted in
the report. The variables were coolant inventory, neutron
flux, RHR heat exchanger outlet temperature, quench tank
temperature, containment atmosphere temperature, CCW
temperature to ESF system and pressurizer heater status.
(5) Resolution of certain issues of an NRR inspection at Rancho
Seco (April 7, 1987). These issues were software validation
and qualification of electrical ?solators.
Additional inspection items for completion of this review are: i
(6) On the SPDS installation,
a. Verify that equipment changes are properly approved and l
controlled.
b. Verify that as-built drawings are changed to show l
equipment changes.
(7) Ar. evaluation of the training of the operators on the SPDS. l
Items (1) through (5) were not reviewed during this inspection.
The proper installation of the SPDS was verified in Inspection
Report 50-312/85-09 and considered closed. Since there have been
modifications after that Inspection Report, further inspection on
the SPDS installation will focus on the recent modifications.
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Item (7) was evaluated by the inspector by reviewing a recent lesson j
plan (OD 24 , "')) and the attendees for the SPDS. The licensee ;
has trained ti- perators on the use of the SPDS, possible failure i
mechanistns of toe SPDS, hardware and software modifications. In !
talking with the operators, the inspector concluded that they
appeared knowledgeable on the SPDS. The training of the operators ,
was also noted in Inspection Report 50-312/85-09. Thi e. sub-item is !
closed.
This item will remain open pending closure of items (1) through (6)
mentioned above.
B. (0 pen) TMI Item II.E.1.1.2. "Auxiliary Feedwater System Evaluation -
Long-Term System Modifications"
This item required licensees to evaluate their Auxiliary Feedwater
(AFW) system and to upgrade the system where needed based on the
evaluation. The evaluation was intended to increase AFW
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reliability. The previous NRC inspection reports on this item were
50-312/82-06 and 50-312/85-21. These reports left closure of this
item pending installation of the Emergency Feedwater Initiation and
Control (EFIC) system.
Inspection Report 50-312/85-21 noted that the AFW system was
protected against internally generated missiles and that the AFW
pumps are automatically loaded onto the diesel buses. The specifics
left open by Inspection Report 50-312/85-21 were to verify that:
(1) The AFW piping was changed so that the failure to keep manual
valve FWS-055 closed would not disable the AFW system. This
was accomplished by the installation of an automatically
controlled EFIC valve (FV-31855).
(2) Installation of ~ 4mically qualified valves.
In addition, the NR 9rns are:
(3) Regarding inst .sion of the EFIC system,
a. That it meets NRC requirements and licensee commitments
b. That equipment changes are properly moroved and
controlled,
c. Verify that as-built drawings are cha,, s to show
equipment changes,
d. Necessary procedure changes have been made and that
training has been accomplished.
(4) An assessment and verification of the EFIC training. !
(5) Review of the new TSs on the AFW system. ;
(6) Verification that EFIC preoperational testing is complete.
(7) Verification that the EFIC instruments were in calibration.
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(8) A review of the new EFIC operating procedures.
The concern with item (1) was that a failure to keep manual valve
FWS-055 closed could result in a failure of the AFW system to in, lect l
water into the steam generators and instead pass the water through !
the valve to the atmosphere. Item (1) of was addressed by the
licensee by changing the piping configuration. The licensee moved
valve FWS-055 and installed the automatic valve FV-31855 in series
with valve FWS-055. The automatic controls on valve FV-31855 were
installed under ECN 5415J. The automatic controls on valve FV-31855
receive redundant close signals when EFIC is initiated. The new
installation was verified by the inspector in a walkdown and
appeared satisfactory. This item is closed.
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For item (2), the inspector was informed by the licensee that the
system is seismically classified. Item (2) was verified by the
inspector by selectively reviewing the seismic qualification of the
AFW valves and that the valves were classified seismic on the master
aquipment list. This item is closed. .
Item (3) concerned the installation of the EFIC system. The NRC staff
performed a Safety Evaluation of the system (Knighton to Andognini,
December 24, 1987) and found the design acceptable- The EFIC system
modification on the AFW system was completed under major ECN-5415,
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(with sub ECNs A through AH) which was turned over to operations
during the inspection (February 9, 1988). The ECNs were documented'
closed on February 10, 1988. The major ECN and Sub ECNs A, B, C, D,
E, F, J, M, N, Q, U, V, AB, and AE were reviewed by the inspector
and appeared to be properly approved and controlled. Items (3)a,
(3)b, and (3)d are closed.
Training of operators on the EFIC system was verified in discussions
with the licensee. The inspector reviewed the lesson plans and
attendance records on the EFIC system. The lessons reviewed were 00
23 K 0500, 00 24 0 3200, and OD 24 K 1000. These lessons were the
introduction, description and operation of the EFIC system,
respectively. The training appeared adequate. The operators
underwent the training on the EFIC system and satisfactorily' passed.
Item (4) is closed.
The TSs on the EFIC system were issued under TS Amendment 93, issued
Januiry 5,1988, after review by NRR. Item (5) is closed.
The inspector was informed that the cold EFIC system testing under
STP 666 is complete. The system hot functional tests are yet to be
performed (STP 1113) to verify system operability,at temperature and
to verify the density compensation in the steam generator which is
performed by EFIC. The_ system is also to be tested when decay heat
is available to provide steam to the system.
The instruments on the EFIC system were verified to be in
calibration, with an 18 month recalibration schedule. The licensee
performed the functional calibration in concert with the performance
of Special Test Procedure (STP) 666. The calibration was done with
a general calibration procedure, I-011, under work requests 121693,
121964, 121965, 121696, 121967, 121699,.121700, 128972, 128985, and
128986. Three of these work requests, 121693, 121694, and 121986,
which covered the majority of the calibration, were reviewed. No
problems were identified. Item (7) is closed.
The new operating procedures were not reviewed.
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This item will remain open pending satisfactory inspection of items 1
(3)c, (3)d, (6), and (8) above.
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C. (0 pen) Item II.E.1.2.1.8.2, "Auxiliary Feedwater System Initiation
and Flow," AFW Safety Grade Initiation
This item required the licensee to provid'e a reliable, automatic
initiation feature for the AFW system consistent with GDC 20.
This. item was previously inspected in NRC Inspection. Reports
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50-312/81-21, 50-312/82-06, and 50-312/85-21. The items remaining
to be closed from these inspection reports were-
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(1) Verify that the licensee has provided for. auto loading of AFW l
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pump P319 upon a Safety Features Actuation 3fqqal (SFAS). This
was still under NRC staff review as of the last inspection l
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reprt.
Additional NUREG-0737 requirements for this item are:
(2) The design shall provide for automatic initiation of the AFWS.
(3) The automatic initiation signals and circuits shall be designed ;
so that a single failure will not result in the loss of AFWS l
function.
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(4) Testability of the initiating signals and circuits shall be a
feature of the design.
(5) The initiating signals and circuits shall be powered from the
emergency buses. ,
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(6) Manual capability to initiate the AFWS from the control room
shall be retained and shall be implemented,so that a single
failure in the manual circuits will not result in loss of j
system function. l
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(7) The ac motor driven pumps and valves in the AFWS shall be
included in the automatic actuation (simultaneous and/or
sequential) of the loads onto the emergency buses.
(8) The automatic initiating signals and circuits shall be designed
so that their failure will not result.in the loss of manual
capability to initiate the AFWS from the control room.
Additional items identified for the equipment. inspection were to:
(9) Equipment Installation / Modification
a. Verify that the installation / modification of equipment
meets licensee commitments and NRC requirements
b. Verify that equipment changes are properly approved and
controlled.
c. Verify that.as-built drawings are changed to show the
equipment changes.
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d. Verify that the necessary procedure changes are made and
that the necessary personnel training has been
accomplished.
(10) Operation
a. Verify that preoperational testing is complete.
b. Verify that equipment is calibrated,
c. Verify that equipment is operable, and that operational
procedures are being used.
Item (1) was performed under Engineering Change Notice (ECN)
A-5415M. This ECN loaded the pumps to the new TDI diesels (GE
A2 and GE B2). The previous loading of the pumps was performed
under ECN A-3653 which was tied to the GM diesels. The SFAS
initiation and cor trols for the AFW pumps (P-318 and P-319)
were deleted in EU4 A-5415M. The auto loading of the pumps are
onto buses providet by diesel generators GE A2 and GE 82. Item
(1) is closed.
The NRC staff has completed their review of the EFIC system and
documented this review in a Safety Evaluation dated December
14, 1987 (Knighton to Andognini). At the conclusion of the
Safety Evaluation, the NRC staff concluded that the
requiremerts of NUREG-0737, item II.E.1.2 regarding safety
grade automatic initiation of the AFW system were met. The
EFIC system design was found to be acceptable. Item (2) is
closed.
The inspector also performed a review of the requirements of
NUREG-0737 for this item. The following is a description of
these line items.
The EFIC system will initiate on water level decreasing below a
desired setpoint in any steam generator, or by a SFAS signal.
The prior SFAS initiation of AFW pumps and valves'was deleted
under ECN A-5415F. The actuation of AFW following an SFAS
signal is now to have the SFAS signal start the EFIC system.
There is also a Reactor Protection System (RPS) signal into the j
EFIC system. This was installed under sub ECN A-54150. This !
ECN was reviewed and no problems were identified. This ECN I
also implemented NUREG-0737 item II.K.10.B, safety grade !
anticipatory reactor trip.
The items (3), (4), (5), (6), (7) and (8) are discussed in the
safety evaluation and in ECN A-5415. This major ECN was
reviewed, and the design was discussed with the licensee. The
system was also toured to verify these items. Based on this
inspection, and the design acceptance by the staff, these
sub-items are considered closed.
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Items (9) and (10) on the. equipment installation and operation
of the EFIC system were performed in conjunction with
NUREG-0737 item II.E.1.1.2. The equipment was verified
installed and calibrated. The equipment changes were verified
to be properly approved and controlled. The equipment was
determined to be. operable at the time of the inspection. The
preoperational testing was determined to be complete, except
for hot functional tests. Items (9)a and (9)b are closed.
This item will remain open pending verification of items (9)c,
(9)d, (10)a, and (10)c noted abova.
D. (0 pen) Item II.E.1.2.2.C.2, "Auxiliary Feedwater System Initiation ,
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and Flow" AFW Safety Grade Flow Indication
This item required the licensee to provide a safety grade indication
of the auxiliary feedwater flow to each steam generator in the I
control room. This is for the control room operators to ascertain
the actual performance of the AFW system when it is called to
perform its intended function.
This item was inspected previously in NRC Inspection Reports
50-312/81-21, 50-312/82-06, and 50-312/85-21. A violation on this
item (86-21-05) identified that the flow rate indicators in the
control room were neither powered by a class 1E power supply, nor '
were they built to class 1E requirements. This was closed out in
Inspection Report 50-312/86-33, noting that the flow rate indicators
will be installed concurrently with the EFIC syste:n. 1
The other item that remained open from the previous inspection
reports was:
(1) The AFW flow and condensate storage tank level needed to be
verified available on the SPOS.
Additional NUREG-0737 requirements for this item are:
(2) Safety grade indication of AFW flow to each S/G shall be i
provided in the control room, with a minimum of 2 AFW flow rate l
indicators for each S/G. The flow indication system should j
conform to the salient paragraphs of IEEE 279-1971 (as noted in
(3) The AFW instrument channels shall be powered from the emergency
buses consistent with satisfying the emergency power diversity
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requirements of the AFW system set forth in the Auxiliary
Systems Branch Technical Position 10-1 of the Standard Review
Plan, Section 10.4.9.
Additional inspection items identified for the equipment inspection
were:
(4) Equipment Installation / Modification
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a. Verify that the installation / modification of equipment
meets licensee commitments and NRC requirements
b. Verify that equipment changes are properly approved and
controlled.
c. Verify that as-built drawings are changed to show the
equipment changes.
d. Verify that the necessary procedure changes are made and
that the necessary personnel training has been
accomplished.
(5) Operation
6. Verify that preoperational testing is complete.
b, Verify that equipment is calibrated.
c. Verify that equipment is operable, and that operational
procedures are being used.
The AFW flow and condensate storage level were verified to be
available on the SPDS in the control room by the inspector. Item
(1) is closed.
The inspector observed the EFIC panel in the control room and
verified that there were two AFW flow meters per steam generator on
the EFIC control station. An inspection of the back panel revealed
proper separation and redundancy. The AFW flow is not used to
control the EFIC system, but the flow indication is provided on the
EFIC control station. The EFIC control station and instruments were
classified IE. Item (2) is closed.
The equipment was verified to be properly installed in the previous
inspection report. It was also noted that the installation was
properly approved and met NRC requirements. Items (4)a and (4)b were,
therefore, previously closed.
Items (3), (4)c, (4)d, and (5) above remain open.
E. (0 pen) Item II.K.2.8, "Orders on B&W Plants, Upgrade of the AFW
System"
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This item was previously reviewed in NRC inspection report I
50-312/85-21. I
The last inspection repert stated that all AFWS upgrade ,
modifications are being reviewed as part of TMI items II.E.1.1 and '
II.E.1.2. It remained open pending the closure of these two items.
The work on the AFW system is substantially complete. This will
remain open pending closure of these two items.
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D. (0 pen) Item II.K.2.10.8, "Orders on B&W Plants, Safety Grade Trip"
This item required B&W licensees to install a safety grade
Anticipatory Reactor Trip (ART) on loss-of-feedwater and turbine
trip. This provides an additional measure of protection by tripping
the reactor on a loss of heat sink. The previous NRC inspection
reports in which this item was inspected were 50-312/81-21,
50-312/82-06, and 50-312/85-21.
This item remained open pending:
(1) A review of the modification to upgrade the main feedwater trip
portion of the ART system.
(2) A review of the logic and electrical schematic diagrams for the
ART.
(3) A review of TS changes to assure channel functional checks are
tested on a basis commensurate with existing RPS channel
functional tests.
The modification was installed with the EFIC system in Engineering
Change Notice (ECN) A-5415D, which was completed December 28, 1987.
This ECN was reviewed by the inspector and found to have adequate
sign-offs and review. The ART occurs on a loss of both Main
Feedwater (MFW) pumps and reactor power greater than 20%.
The logic and electrical schematic diagrams were not reviewed by the
inspector.
This item remains open pending a review of the diagrams and the
TS changes made to the RPS system.
8. Exit Interview
The inspection scope and findings were summarized on June 25, 1987, with
those persons indicated in paragraph 1 above. The inspector described
the areas inspected and discussed in detail the inspection findings. The
licensee acknowledged the inspectors' findings and observations. The
following new items were identified during this inspection:
Violation 88-06-01 - Failure to Remove Expired Shelf Life Items
from Stock (paragraph SA).
Follow-up Item 88-06-02 - Design Engineering Review of Material
Substitutions (paragraph 6)