ML17157A803
| ML17157A803 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 07/24/1991 |
| From: | Eapen P, James Trapp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17157A801 | List: |
| References | |
| 50-387-91-06, 50-387-91-6, 50-388-91-06, 50-388-91-6, NUDOCS 9108260009 | |
| Download: ML17157A803 (9) | |
See also: IR 05000387/1991006
Text
U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Report Nos.
0- 87/ 1-06
nd
-
8/ 1-
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Licensee:
enn
lvania P w r
nd Li ht
m an
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Allentown Penns
lvani
1
101
Facility Name:
us uehanna
earn Electric
i n
I
Inspection At:
Salem Townshi
Penns
Ivania
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Inspector:
J. M. T
p, Sr. React r Engineer
System Sections
Engineering Branch, DRS
date
Approved by:
Dr. P. K. Eapen, Chi f
Systems Section, DRS
Ins ection Summa:
Routine Unannounced Inspection on May 6-10, 1991.
Inspection Report Nos. 50-387/91-06 and 50-388/91-06.
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bypass control.
Also included in the scope was engineering/technical
support follow-up to
identified deficiencies and a review of inservice testing.
The control of design changes
is
important to assure that modified systems continue to be capable of performing their intended
safety function after the design changes are implemented.
~Re uit: in general, the modifications reviewed were technically sound.
Piant modification
had been thoroughly "closed out" prior to the system being declared operable.
The Inservice Test /ST) results reviewed were acceptable, well organized and trended.
However, a check valve in the containment instrument gas system "Q" boundary, was not
910826000tyt
910813
ADOCH, 05000387
O
included in the IST program.
Consequently this valve was not tested in accordance with
ASME BEcPV Code,Section XI as discussed in Section 4.0.
This inspection found that Engineering/Technical Support follow-up for the deficiencies
reviewed, was thorough.
However, one weakness
was identified in the timeliness of repairs
being made to the Radiation Monitoring System as discussed in Section 3.0.
DETAIL/
1.0
De i n
h n es
nd M
ificati n
Review
(NRC IM 37700)
The objective of reviewing design changes (DCP's) was to ascertain that changes
to
the station's safety systems as described in the Safety Analysis Report were made in
accordance with regulatory requirements,
Technical Specifications, Station Procedures,
and the licensee's Quality Assurance Program.
This objective was accomplished by
performing a detailed review of the following modifications:
l.
88-3032,
"RHRSW Modification-Isolation Valves for RHR Heat Exchangers."
2.
88-3050A,
"Replacement of RHR Pump Room Cooling Coils 2E230 A/C."
3.
89-3030,
"ESW Loop "B" Supply and Return Piping Modification-
Reactor Building Cooling Coils."
These design change packages
(DCPs) were detailed and technically sound.
In
general, the DCPs reviewed incorporated requirements and standards
commensurate
with the safety significance of the design change involved.
The thoroughness of the
DCPs was also evident by the limited number of changes (PCNs) that were made to
the DCPs during installation.
The inspector observed:
~
All design packages were adequately reviewed and approved by the plant
operations review committee prior to release for installation.
~
Post modification testing was performed in accordance'with previously
established
acceptance criteria and the test acceptance criteria were met during
post modification testing.
~
Control room drawings had been modified in accordance with station
procedures prior to declaring the modified system operable.
~
Training of licensed operators on the modified plant systems was conducted
prior to declaring the modified system operable.
~
Station procedures were updated prior to declaring the modified system
~
ISI and IST Programs were revised to reflect changes in the system design.
However, a formal process for assuring design changes are reflected in the IST
program was not included in modification close out procedures.
One of the modifications reviewed, DCP 88-3032, incorporated the changes
to
facilitate a temporary fire water connection to the RHR service water system in the
event of a station blackout.
This was a licensee initiative to enhance the performance
of installed safety systems.
The systems engineers'ctive
involvement in the modification process was perceived
as an additional licensee strength.
The system engineers provide a focal point for the
modifications from inception.
The system engineers interviewed exhibited detailed
knowledge of the modifications performed on their assigned
systems and the overall
modification process.
Design kickoffand installation kickoffmeetings were observed
to be useful efforts to enhance cooperation and communication between responsible
groups.
The system engineers performed an effective modification close-out process.
One deficiency was identified in the establishment of testing criteria for modification
88-3032,
The RHR service water system is a ASME, Class 3 system.
Code,Section XI, Article IWD-5210 (b), states in part that "The contained fluid in
the system shall serve as the pressurizing medium."
PPAL Specification M-1040,
states in section 3.8.1.3 (a), pneumatic pressure
test requirements for Class 3 systems,
that Class 3 water systems must be pressure
tested only by means of water.
For the
RHR service water system the pressurizing medium is water.
A one time exception
was approved by engineering to perform a pneumatic pressure test in lieu of a
hydrostatic test for the "hot-tapped,"3" drain connections.
The failure to perform a
hydrostatic test is not a plant safety issue because the pneumatic test conducted was
comparable to a hydrostatic test in detecting leakage.
However, the engineering
assessment
which allowed the pneumatic test was not thorough because it did not
consider ASME BkPV Code requirements to hydrostatic test these welds.
The licensee's quality assurance
audit group recently completed an audit of the plant
modification system (Audit Report 91-024).
This was a three week audit by a team of
four auditors.
The audit report concluded that the plant modifications program
remains effective and satisfactorily implemented.
No significant technical or design
issues were identified.
The audit report was reviewed and it was found to be an
effective assessment
of the plant modification program.
2.0
Electric
1 and Mechanical B
s
Control (NRC IM 37700)
The inspector reviewed the temporary modifications (bypasses)
to assure that such
modifications were completed in accordance with NRC requirements.
Electrical and
Mechanical Bypass Control procedure AD-QA-484, Rev. 3, was reviewed to verify
that administrative controls for bypass control were adequate.
In addition, the
following two bypasses
were selected to verify that the licensee was implementing
these bypasses
in accordance with the bypass control procedure.
~
2-91-022, "Temporary 250VDC Battery"
~
2-89-005, "Temporary Pressure Transmitter on HPCI"
The following observations were made during the review of the two bypasses
listed
above:
~
The bypasses
were reviewed and approved in accordance with the bypass
control procedure.
The bypasses
were installed in accordance with the installation package.
The records for the installation of bypasses
were complete.
Periodic reviews of the bypasses
were being performed in accordance with the
administrative procedures.
The total number of active bypasses for both units was less than twenty at the time of
this inspection.
This is a reduction from over one hundred bypasses
installed one year
ago.
This reduction of the active bypasses
indicates a strong management involvement
in the control and close-out of bypasses.
The reduction of active bypasses
required a
large engineering effort to transform the long standing bypasses into permanent station
modifications.
This effort provided additional reviews and incorporation of the
modification into plant design documentation.
A review of AD-QA-484, Rev. 3, "Electrical and Mechanical Bypass Control," found
that certain bypasses
are not reviewed by PORC.
To satisfy section 6.5.1,6 of the
technical specifications which requires PORC be responsible for review of all
proposed changes or modifications to unit systems or equipment that affect nuclear
safety, the bypasses
are screened by the technical section.
The screening criteria are
as follows:
Ifthe bypass constitutes a change to the facility or procedures
as described in
the FSAR or it is a test or experiment not described in the FSAR, then a safety
evaluation is written and PORC review is required prior to installation.
Ifthe bypass does not meet the above screening criteria, it may be installed
after approval by a technical supervisor and this bypass does not require PORC
review.
The inspector noted that the above screening criteria were established in a revision of
AD-QA-484. The previous revision of AD-QA-484 screening criteria did not provide
similar guidelines for the PORC review of bypasses.
The inspector observed that a
small number of the older installed bypasses
were not screened for review by PORC
using the present screening criteria.
The inspector discussed
this matter with the
licensee's
representatives.
The licensee stated at the exit meeting that the installed old
bypasses
would be screened
using the latest edition of AD-QA-484 and corrective
actions would be taken, ifneeded.
The inspector found this action to be acceptable.
3.0
En ineerin
rr
tive Action Foll w-u
Two licensee identified deficiencies were reviewed by the inspector to assure that
engineering/technical
resolutions
to deficiencies were performed in a thorough and
timely manner.
The Unit 2 automatic depressurization
system was declared inoperable when the
containment instrument gas header pressure dropped below 135 psig due to a pressure
relief valve (PRV) lifting. Licensee Event Report LER 90-003 was written for the
event which occurred on February 28, 1990.
The LER was updated by the licensee
on June 29, 1990.
The licensee performed bench testing of the PRV on
April 23, 1990; but could not positively identify the root cause for the PRV opening
and failing to reseat.
The internals of the valve were replaced and the valve was re-
installed.
As a long term corrective action the licensee stated that a PRV designed
specifically for gas applications would be installed.
The licensee has completed the
installation of the improved PRV for Unit 2 and is scheduled to install a similar valve
in Unit 1 during the 1992 refueling outage.
The inspector found that the technical
support provided to correct these deficiencies was thorough and timely.
The second licensee identified deficiency reviewed related to the corrective actions
taken to enhance the vent stack radiation monitoring system (SPING).
This radiation
monitoring system experienced
a high rate of failure.
The failure of the SPING places
the plant in a Technical Specification Limiting Condition of Operation (LCO).
Problems with the.SPING were first identified by the licensee in 1983.
An NRC
unresolved item regarding this issue was opened in 1985.
An update of the NRC
unresolved item was provided in NRC Inspection Report 50-387/90-20, at which time
the schedule to start installation of the phase
1 of the SPING upgrade project was
given as December
1990.
At the time of this inspection, the date for phase
1
installation had been delayed until September of 1991.
The engineering which has
gone into the phase
1 effort appears to be extensive and addresses
the problems
presently experienced by the SPING.
However, the delays encountered
in the attempt
to install this modification indicate a potential weakness in the engineering process.
7
Inservice Te t Pr
ram (NRC IM 73756)
The scope of the IST program review was to assure that all valves and pumps which
are required to be tested in accordance with ASME BEcPV code Section XI are
included in the IST program.
In addition, the results of selected IST pump and valve
tests were reviewed to assure the following:
~
Acceptance criteria are met or proper corrective actions were taken.
~
Tests adequately verified the design functions of the components.
~
Test records are maintained.
The 1990 and 1991 quarterly core spray valve exercise data for core spray valve HV-
151-F015A were reviewed.
The test records reviewed were well organized and
complete.
The valve stroke time data were reviewed and trended by the technical
support organization.
The pre-establish
acceptance criteria for valve stroke times were
'et for the data reviewed.
The quarterly core spray flow verification test data were
reviewed for core spray
pump 1P206A.
The test procedure was found to be acceptable.
The test results met
pre-established
acceptance criteria.
The test data were'well organized and trended by
the technical staff.
The overall IST program implementation was effective.
A review to assure that required valves and pumps were included in the IST program,
was conducted for the containment instrument gas system (CIG) and for the core spray
system.
The valves and pumps in the core spray system, selected for review, were adequately
incorporated in the IST program.
No deficiencies were identified with regard to the
core spray system IST program.
The CIG system is divided into a safety related (Q) and a non-safety related section
(non-Q).
The safety related section of the CIG system has
instrument gas bottles to
provide backup pneumatic energy actuation of the main steam system safety relief
valves (SRVs). The SRVs are required during accident conditions as part of the
automatic depressurization
system (ADS). A one inch check valve and a one inch
solenoid operated valve in the CIG system provide double valve isolation between the
Q and non Q portions of the CIG system.
Both valves are included within the Q
boundary.
The solenoid operated valve is provided to close the valve on a
containment isolation signal.
The solenoid valve is included in the IST program and
is routinely leak checked.
The check valves (1-26-018, 1-26-029, Unit 1; and 2-26-
018, 2-26-029 Unit 2), which are located in a ASME, B&PV Code,Section III, class
2 piping system, were not included in the IST program and were not leak tested.
The
licensee has not applied for relief from testing this valve.
The 10 CFR 50.55a
requires that inservice testing be performed on certain ASME Code Class 1, 2, and 3
valves.
Section XI Subsections IWV-1100 defines the scope of valves to be tested
in'erms
of plant shutdown" and accident mitigation.
Since this check valve is required
for accident mitigation it is required to be included in the IST program.
The omission
of this valve from the IST program is an apparent violation (Violation 50-387/91-06-
01).
5.0
nre
Ived Item
-
7
/
-
-
I
ed
Inspection Report 90-08 discussed
a licensee identified deficiency in the main steam
pipe tunnel fan wiring.. The reactor building main steam pipe tunnel coolers are a
non-safety related system.
An error in the wiring of the main steam pipe tunnel fan
resulted in the opposite fan's cooling coils functioning when the fan was in service.
It
also appeared
that the post installation testing was inadequate to identify this concern.
The licensee's immediate corrective action was to open the cooling coils service water
return bypass valve and verify that one fan in service could maintain tunnel
temperature.
Additional corrective actions were to re-wire the fan controls to the
service water valves and to survey the operators for similar type problems.
The
inspector reviewed Engineering Change Order (ECO) 90-6060 which re-wired the
valve controls and corrected plant drawings.
The ECO adequately corrected this
'iring
error.
The results of the operators survey and the technical section's response
to the survey were reviewed and found to be satisfactory.
Based on the above
corrective actions, this unresolved item is closed.
6.~
E~i
The inspector met with those individuals denoted in Attachment I, on May 10, 1991,
to discuss the inspection findings as detailed in this report.
I
P~d
ATTA HMENT I
A.
L~ien ee
- E. Bragger, Sr. Proj. Eng.
'. Creasy, Pwr. Prod. Eng.
M. Golden, Plant Eng. Supv.
- R. Harris, Sr. Results Eng.
D. McGann, Pwr. Prod. Eng.
G. Maertz, Pwr. Prod. Eng.
A. Nargoski, QA
T. Nork, Plant Eng. Supv.
- L. O'eil, Supv. Eng.- Nuclear
R. Paley, Pwr. Prod. Eng.
- D. Roth, Sr. Compliance Eng.
R. Rarig, Tech. Asst.
D. Ritter, Pwr. Prod. Eng.
T. Sweeny, Pwr. Prod. Eng.
B. Saccone, NPE
- G. Stanley, Supt. of Plant
- R. Wehry, Compliance Eng.
B.
NRC
- S. Barber, Senior Resident Inspector
B. McDermott, Reactor Engineer
- J. White, Chief, RPS 2A
- Denotes those present at the exit meeting'conducted
onsite on May 10, 1991.