ML18102B256
| ML18102B256 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 04/29/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML18102B253 | List: |
| References | |
| 50-272-97-02, 50-272-97-2, 50-311-97-02, 50-311-97-2, NUDOCS 9705070034 | |
| Download: ML18102B256 (25) | |
See also: IR 05000272/1997002
Text
U. S. NUCLEAR REGULATORY COMMISSION
Docket Nos:
License Nos:
Report Nos:
Licensee:
Facility:
Location:
Dates:
Inspectors:
Approved by:
9705070034 970429
ADOCK 05000272
G
REGION I
50-272; 50-311
50-272/97-02; 50-311197-02
Public Service Electric and Gas Company
Salem Nuclear Generating Station, Units 1 and 2
Hancocks Bridge, NJ
January 2 - February 7, 1997
A. Della Greca, Sr. Reactor Engineer, EEB, DRS
R. Bhatia, Reactor Engineer, EEB, DRS
L. Cheung, Sr .. Reactor Engineer, EEB, DRS
T. Kenny, Sr. Reactor Engineer, SEB, DRS
G. Morris, Reactor Engineer, EEB, DRS
B. Smith, NRC Contract Engineer *
William H. Ruland, Chief, Electrical Engineering Branch
Division of Reactor Safety
TABLE OF CONTENTS
EXECUTIVE SUMMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii
Ill. Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
1
E1
Conduct of Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
1
E1 .1
Introduction
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . .
1
E1 .2
NRC Salem Restart Item 111.a.18 - Parts Availability and Bill of Material
Accuracy (Closed) ............. * . . . . . . . . . . . . . . . . . . . . . . . . . .
1
E1 .3
NRC Restart Item 11.21 - Wiring Separation and Redundancy Concerns
with RG 1.97 Instruments and Cable Separation (Open) ............ * 3
E1 .4
NRC Restart Item 11.20 - Pressurizer Overpressure Protection System.
(POPS) Ability to Mitigate Over pressure Events. (Closed) . . . . . . . . . . .
5
E1 .5
NRC Restart Item 11-14, Hagan Module Replacement Project (Closed) . . .
6
E1 .6
NRC Restart Item 11-12 - Review Adequacy of Fuse Control Program
(Closed) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
9
E8
Miscellaneous Engineering Issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
11
E8.1
(Closed) Violation EA94-112-06014, 2.8, Inspection Reports 50-272;
311 /94-80 and 94-13 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . * 11
E8.2
(Open) Unresolved Item 50-272/94-04-01 and 50-311194-04-01 . . . . .
13
E8.3
(Closed) Unresolved Item 50-272/89-13-07 and 50-311189-12-07 . . . .
16
E8 .4
(Closed) Unresolved Item 50-271 ; 311 /93-82-13 . . . . . . . . . . . . . . . .
1 6
E8.5
(Closed) Unresolved Item 50-311193-82-04 . . . . . . . . . . . . . . . . . . . .
19
E8.6
(Closed) Unresolved Item 50-272; 311/94-32-05 . . . . . . . . . . . . . . . .
21
E8.7
Review of UFSAR Commitments . . . . . . . . . . . . . . . . . . . . . . . . . . . .
21
V. Management Meetings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
21
XI. Exit Meeting Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
21
PARTIAL LIST OF ATTENDEES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
21
LIST OF ACRONYMS USED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
22
ii
..
EXECUTIVE SUMMARY
This inspection included aspects of licensee engineering and plant support. The report
covers an 8-week inspection related to equipment and engineering performance issues that
require resolution prior to Salem restart. These issues are included in Checklists II and Ill.a
of the NRC restart action plan.
Engineering
Based on their review of five closure packages and six unresolved items and violations, the
inspectors concluded that:
PSE&G took appropriate corrective actions to address NRC concerns regarding
material availability and accuracy of Bills of Material. As a result, considerable
improvements were made in these areas and in the area of material management.
Further improvements were sought by fostering better communication. (E1 .2)
Corrective actions to address RG 1.97 separation and redundancy concerns were
extensive. Modification packages and safety evaluations were acceptable. (E1 .3)
Engineering developed good analytical bases to address some NRC concerns about
the Hagan modules, but some other concerns were not clearly understood. As a
result, rework was necessary to reach acceptable resolutions. (E1 .5)
Proposed revision to the 50.59 process to evaluate equivalent replacements should
provide a better understanding of the replacement process and ensure a better
control on design and system configurations. (E1 .5)
The licensee's evaluation of the oversized fuse in the control power transformer
secondary circuit was deficient in that it failed to recognized that the fuse did not
adequately protect #20 AWG conductors. (E1 .6)
Acceptable actions were taken by PSE&G to demonstrate qua,lification of the power
range monitor. (E8.2)
Inadequate implementation of test procedures resulted in violations of the
Appendix B test control program. (E8.4)
. *
The licensee's failure to address identified discrepancies between the molded case
test procedure and the supporting calculation resulted in a violation of the corrective
action program. (E8.4)
iii
E1
E1 .1
Report Details
Ill. Engineering
Conduct of Engineering
Introduction
On February 23, 1996, the NRC issued the restart action plan for Salem Units 1
and 2. Restart Issue Checklists II and Ill.a include the technical and programmatic
issues that require resolution. These issues, related to NRC concerns regarding
equipment performance problems and plant personnel issues, involved previously
identified unresolved items and violations. as well as generic concerns. The purpose
of the current inspection was to review the closure packages prepared by the
licensee to address these issues. Except as noted, the review was conducted in
accordance with inspection procedure 92903.
E1 .2
NRC Salem Restart Item 111.a.18 - Parts Availability and Bill of Material Accuracy
(Closed)
a.
Inspection Scope
The NRC partially reviewed the availability of parts to the maintenance department
and the accuracy of the bill of material {BOM)_ database in December 1996, as
detailed in NRC inspection reports 50-272; 311 /96-20. At that time, the NRC noted
improvement in both areas. However, licensee process improvement activities were
still ongoing. Therefore, the inspectors were unable to reach definite conclusions
regarding the acceptability of the program. During this inspection, the NRC
continued to assess the adequacy of these areas by reviewing additional material
availability performance data, SOM-related material problem requests and their
resolutions by the station, and selected BOMs.
b.
Observations and Findings
Material Availability and Material Management Processes
In the above noted inspection, the NRC observed that PSE&G haq improved the
overall availability of materials and the material management processes. They had
accomplished this by clarifying individual responsibilities and management
expectations. In addition, PSE&G had streamlined the nuclear procurement and
material management (NP&MM) department, enhanced communications between
the department and the end users, and initiated a material prioritization process.
During the current review, the inspectors determined that the work orders on hold
due to parts unavailability had been reduced further since the previous inspection
and that NP&MM was ordering parts well in advance of the scheduled work
implementation dates. Discussion with maintenance personnel i_ndicated that the
material support of maintenance activities had been significantly improved. They
attributed this improvement to good communication among departments, timely_ *
identification of material needs, and good support from NP&MM. The inspectors'
----
2
review of planned work list showed no material restraints at that time. Further, the
inspectors observed that NP&MM representatives met with maintenance department
planners and their supervisor on a daily basis to review material needs and
concerns.
To improve the material availability of the station, during the last year, PSE&G had
increased the on-hand inventory of key critical system component. This increase
was evident in the NP&MM's department performance indicators .. In addition, the
licensee had developed a means to identify items that require restocking and was
emph~sizing the use of blanket purchase orders for stock replenishment.
Based on the above review, the inspectors concluded that the PSE&G had
adequately addressed the material concerns at Salem and Hope Creek station.
BOM Accuracy and Documentation Concerns
To properly address the concerns with MMIS database updating and to improve the
accuracy of the BOMs, the licensee conducted three self-assessments. During the
1996 inspection, the inspectors reviewed the results of these assessments and
sampled the licensee's resolution of the identified BOM deficiencies. He found the
BOMs acceptable. However, the review of this area was limited.
During the current review, the inspectors conducted additional reviews of existing
BOMs. He randomly selected five BOM data records (6590 - reactor coolant vent
line air operated globe valve; 7490 - diesel generator control interlock relay; 7650 -
chemical volume control safety relief valve; 0380 - diesel generator starting air
motor solenoid valve; and 1030 - 21 auxiliary building vent containment spray
pump room cooling unit motor) and evaluated their accuracy. He found that for the
selected components the database was accurate; the licensee ha~ accurately
identified the sub-components, material characteristics, and applicable class code.
The sub-component requirements were consistent with the vendor manual and
design drawing specified requirements.
As per procedure SC.SA-SD.ZZ-11 (Z)), to add a part to an existing component
BOM, the planner is required to issue a material requisition (E-REQ) to procurement
engineering. To evaluate the type of BOM concerns identified by the licensee, the
inspectors reviewed a sample of E-REQs dispositioned by procurement engineering
during the last year. He determined that the majority of the sub-component
requests generated were to replace the damaged sub-components or to add new
components to the BOM catalog. The inspectors identified no accuracy issues with
the parts added to the BOM database.
3
c.
Conclusions
Based upon the above review, the inspectors concluded that the licensee had taken
appropriate corrective actions to address NRC concerns regarding material
availability and accuracy of BOMs. The inspectors also concluded that PSE&G had
made considerable improvements in the area of material management, material
availability and BOM database accuracy. The licensee continued to foster better
communications to further improve the material management processes.
E1 .3
NRC Restart Item 11.21 - Wiring Separation and Redundancy Concerns with RG 1.97
Instruments and Cable Separation (Open)
a.
Inspection Scope
Wiring separation and redundancy concerns with post-accident monitoring
equipment were identified during a 1989 NRC inspection of the licensee's
implementation of the Regulatory Guide (RG) 1.97 program. The results of that
inspection were documented in inspection report 50-272/89-13; 50-311 /89-12.
During a 1990 integrated performance assessment team (IPA T) inspection, the NRC
identified generic cable separation concerns, as documented in inspection report No.
50-272; 311 /90-81. The generic cable separation issues were not addressed by
the licensee in the subject closure package, only the RG 1.97 issues were
addressed. Therefore, the scope of this inspection was limited to a review of the
licensee's resolution of RG 1.97 concerns. The generic cable separation issues will
be reviewed separately when a package is prepared.
The 1989 RG 1.97 inspection identified eight deviations and unresolved issues. All
. but one of those issues. w~re resolved and closed during subsequent inspections.
The remaining issue, unresolved items 50-272/89-13-07 and 50-311189-12-07,
pertaining to inadequate cable separation and electrical isolation for various RG 1 .97
instruments, was the subject of this review.
b.
Observations and Findings
Following the 1989 inspection, the licensee issued, in October 1990, Technical
Standards DE-TS-ZZ-1023 and -2023 which were used as the separation criteria for
a RG 1 .97 instrumentation walkdown. This walkdown, conducted during the
March 1991 . and April 1992 refueling outages, identified additional cable separation
and electrical isolation deficiencies.
In early 1993, the licensee hired Proto-Power Corporation of Groton,. Connecticut,
to conduct an extensive evaluation and walkdown of all RG 1.97 instrumentation to
address these deficiencies. The results of this evaluation and walkdown were
documented in Salem Document S-C-GSC-CEE-0470, "Engineering Evaluation of*
SGS 1 &2 Re.gulatory Guide 1 .97 Instrumentation Compli~nce with Physical
Separation and Electrical Isolation Criteria," Revision 0, dated August 31, 1993.
Subsequently, the licensee issued 14 plant modification packages to resolve the
4
electrical separation and isolation deficiencies identified in this evaluation and
walkdown. Each modification package contained an appropriate safety evaluation
in accordance with 10 CFR 50.59. These modification packages:
For Unit 2
Installed a Leeds and Northrup Model 125 dual pen recorder in the main
control room panel IRP1 to provide recording capability to the Channel C
Gamma-Metrics neutron flux monitor; * .
Installed signal isolators (manufactured by NUS Inc.) to 31 instrument loops
for isolating safety-related indication circuits from nonsafety-related circuits.
Additional 16 NUS isolators were also installed to isolate safety-related
steam generator level signals from nonsafety-related circuits;
Installed 17 molded case circuit breakers (Heinemann Model AM2S) in
various. process cabinets and separated power supplies for Class 1 E loads
from nonclass 1 E loads;
Relocated 6 RG 1.97 instrument loop components from process group 3
racks to process group 2 racks;
Installed aluminized cable sheaths (Siltemp WT 65 thermal barrier material) to
10 affected cable sections to provide suitable barrier between cables which
did not maintain adequate physical separation; and reroute various sections
of pressure instrument impulse lines to provide mechanical separation of
impulse lines for different channel instrumentation;
Rearranged wiring and terl'!linations for 10 containment isolation valves to
separate safety-related position indication circuits from nonsafety-related
- circuits.
The modifications for Unit 1 were similar.
The inspectors reviewed the modification packages, including safety evaluations,
and found them acceptable. The inspectors also examined the installed condition of
NUS signal isolators, Heinemann molded case circuit breakers, Leeds and Northrup
dual pen recorders, and cable wrapped with aluminized cable sheaths. The
inspectors did not identify any unacceptable conditions.
c.
Conclusions
The inspectors concluded that the licensee's corrective actions for this item were
extensive. The modification packages and safety evaluations were acceptable. The
RG 1.97 part of this issue is closed. As stated above, the generic cable separation
issue remains open pending NRC review.
.***,
'
5
E1 .4
NRC Restart Item 11.20 - Pressurizer Overpressure Protection System (POPS) Ability
to Mitigate Over pressure Events. (Closed)
The NRC reviewed the POPS ability to mitigate overpressure events below 31 2 ° F in
June 1996 (NRC IR 50-272; 311 /96-07). During that review, the inspectors
concluded that the licensee had taken acceptable actions to address the POPS
issue. PSE&G, however,* had proposed new design-basis transient limits for POPS.
Therefore, the issue remained open pending the NRC's Office of Nuclear Reactor
Regulation completing their review of the adequacy of the POPS new design basis.
The li~ensee had requested the change to the POPS Technical Specifications (TS)
basis in a letter dated May 31, 1996.
On September 11, 1996, in a letter to the Director, Division of Reactor Projects, the
NRC's Office of Nuclear Reactor Regulation informed Region I, that they had found
the licensee's supporting analysis acceptable. This analysis showed that the
additional margin approved by the use of ASME code case N-514 allowed operation
of a high head SI pump with an operating PD pump, or operation of one
intermediate head SI pump in mode 5, without exceeding. the pressure/temperature
(PIT) limits. The licensee considered the combined flow from the PD and the. high
head SI pumps to be the most limiting mass addition transient. The NRC
acknowledged the change to the TS basis in a letter to the licensee, dated
November 1, 1996 .
The inspectors verified that the changes to the TS and FSAR had been made. In
addition, he confirmed that:
Procedures OP-10.22-0006, "Hot Standby to Cold Shutdown," and IAW SI.
OP-SO.PZR-0006, "RCS Venting," had been revised to reflect the TS change
and require the establishment of a reactor coolant system vent path if" the
POPS became inoperable. *
The use of GOTHIC code as a means for calculating mass addition had been
rescinded, as requested in LER 94-017 safety evaluations. This was done in
via an internal licensee memorandum (NE-96-0220) dated February 2, 1996.
Calculation SGS/M-DM-042 (September 1, 1977) did show that one PORV
was sufficient to mitigate pressure transients from plant heat-up or pump
starts as described above.
The management review committee (MRC) had reviewed and approved the
POPS closure package prior to its release to the NRC for closure.
Based on the above review, item 11-20 is closed
6
E1 .5
NRC Restart Item 11-14, Hagan Module Replacement Project (Closed)
a.
Inspection Scope
b.
In NRC Inspection Report 50-272, 311 /96-06, the inspectors documented four
issues with the Hagan module project that required resolution by PSE&G and further
review by the NRC. The four issues were:
1.
Instrument loop accuracy calculation concern with regards to minimum
calibration temperature versus maximum operating temperature;
2.
additional information required for the upgraded Hagan modules in regards to
design change and 10 CFR 50.59 evaluation;
3.
operating temperature concerns associated with the Hagan and NUS
modules; and,
4.
concerns with electromagnetic and radio frequency interference.
In NRC Inspection Report 50-272, 311 /96-20, the inspectors determined that
PSE&G had provided a~ceptable justification for concluding that module operating
temperature limits would not be exceeded during normal operating conditions, but
additional justification was required to resolve Hagan and NUS module accuracy
concerns during a station blackout (SBO) event.
The purpose of this inspection was to review PSE&G's resolution of the open issues
and evaluate their acceptability.
Observations and Findings *
Issue 1 - Minimum Calibration Temperature versus Maximum Operating Temperature
To address this issue, PSE&G revised Engineering Evaluation S-C-RCP-CEE-1037.
This evaluation assumed a 20° F temperature differential between the minimum
calibration temperature and maximum operating temperature. Considering that
current procedures require that loop calibrations be made with closed instrument
rack doors and that, during a station blackout, procedure require opening of the rack
doors, the inspectors *concluded that the stated 20° F temperature differential was
sufficient to ensure loop accuracy.
The inspectors reviewed the engineering evaluation and concluded that PSE&G's
actions acceptably resolved the inspectors' concern. This issue is closed.
Issue 2 - Additional Information for Design Changes and 10 CFR 50.59 Evaluation
During the original review of the Hagan module refurbishment program, the
inspectors determined that the safety evaluation in accordance with 10 CFR 50.59
for the changes made to the modules was done generically, under the Workbook 3
process. This process is used for equivalent replacements. At that time, the
inspectors expressed a concern that the physical changes made to the modules
might induce subtle changes to their performance that may not be easily detectable
I
7
by the licensee testing, e.g., EMl/RFI susceptibility and response time, unless the
response characteristics were clearly understood by the licensee. As an example,
the inspectors cited the upgrading of the manual/automatic setpoint station modules
which involved among other things the addition of two capacitors and the
replacement of three transistors, three relays and one diode.
In response to the inspectors' concern, the licensee stated that the module supplier,
with an approved Appendix B program, maintained control for the internal design of
the Hagan modules and that the module design changes were reviewed and
approved as part of the supplier's quality assurance program. The licensee also
stated that, under the Workbook 3 process, replacement components must meet or
exceed the specifications of the replaced components, and that an evaluation form
must list and compare all pertinent component specifications.
The inspectors disagreed with PSE&G's position for several reasons: (1) the
supplier's specifications available to the licensee were not sufficiently detailed to
ensure that the replacement (refurbished) modules met or exceeded the
performance requirements of the replaced modules; (2) it was not immediately
evident that the supplier's Appendix B program had maintained control on the
Hagan modules supplied to Salem; and (3) once PSE&G began the refurbishment of
the modules, the responsibility of the module design no longer rested on the
supplier's Appendix B program.
The inspectors discussed these concerns with the licensee who initiated a revision
of procedure NC.DE-WB.ZZ-003(0), "Engineering Workbook for Equivalent
Replacement." A draft revision of this procedure provided for information purposes
on March 4, 1997, requires: (1) a documented assessment for identified differences
between the original and the replacement item that could impact design; (2) a
discussion of how the identified differences of the replacement item do not affect
the design basis of the original item; and !?) an evaluation of vendor supplied
analytical and test data as a proof of acceptability of the replacement item.
The inspectors' review of the proposed procedure changes concluded that, if
properly implemented, they should result in acceptable evaluations of equivalent
replacement items and acceptable determinations of whether the proposed design
change constitutes an umeviewed safety question. The inspectors had no further
concerns regarding equivalent replacements.
Regarding the inspectors' original concern with the Hagan module changes and their
potential impact on the performance of the module, discussions with responsible
engineering personnel showed that, although not adequately documented, the
changes had been reviewed, discussed with the m*anufacturer, and understood. In
addition, the licensee had obtained from Westinghouse the documentation
necessary to effectively maintain the modules~ Therefore, the inspectors' concerns
regarding the module change impact on the systems supported by the modules are
resolved and the issue is closed.
..
- --~
8
Issue 3 - Hagan and NUS Module Accuracy During SBO Event
PSE&G determined that Hagan and NUS modules would operate satisfactorily during
a SBO event and based their conclusion on the following information:
During a SBO event, the control equipment room worst-case temperature
was calculated to be 114.8° F.
The Hagan and NUS rack doors are opened by procedure during this event.
The opening of the doors will increase rack ventilation and the transfer of
- heat to the equipment room atmosphere and, hence, prevent heat buildup
. within the racks. Based on the differential temperature measured with doors
closed, the rack ambient temperature can be expected to be below the
120° F desigri limit for the modules.
Westinghouse and NUS qualification reports provide- documented*-evidence
that the Hagan and NUS modules can perform their design functions and
operate at the elevated temperatures postulated during a SBO.
A decrease in instrument loop indication accuracy due to temperature effects
can be expected during a SBO event. However, test showed that the error is
not sufficient. to impair the operator's primary concern for trending
parameters.
PSE&G concluded the Hagan and NUS modules are capable of performing their
intended safety function as designed under the installed service conditions with
adequate margin for abnormal conditions.
The .inspectors reviewed PSE&G's revised closure package and supporting
.documents and concluded that the licensee had provided reasonable assurance that
the Hagan and NUS modules would be operated within design limits and would be
capable of providing operator information required during a SBO event. This issue is
closed.
Issue 4 - Electromagnetic and Radio Interference
The NRC concerns associated with electromagnetic and radio interference (EMl/RFI)
involved the use of switching power supplies in the NUS modules; the replacement
or installation of new, different speed solid-state components in Hagan modules;
and the relocation of a choking circuitry installed by Westinghouse.
PSE&G evaluated the NRC concerns, but concluded that the EMl/RFI was not a
problem. They based their conclusion on the following considerations: (1 l the
original Hagan design did not consider EMl/RFI, therefore, there was no baseline for
comparison; (2) the refurbishment of the modules, especially the replacement of the
electrolytic capacitors, combined with a good preventive maintenance program,
should result in less internal module noise and ensure acceptable' performance of the
module, even in the presence of the more noisy NUS switching power supplies;
9
(3) the replacement solid state components had been approved by Westinghouse;
(4) the RC snubber circuit in the comparator modules, used to suppress inductive
transients due to the deenergization of downstream relays, had been discussed with
and agreed upon by Westinghouse; (5) there was no evidence that previous Hagan
module problems were due to EMl/RFI; and (6) recent experience with refurbished
modules did not indicate any susceptibility to noise.
In light of past and recent experience with the module performance, the inspectors
concurred that there was no reasonable basis for concern in this area. To address
aging of the modules and the potential increase in their susceptibility to noise, he
discussed planned preventive maintenance with the licensee. He determined that
the licensee had issued an action request (No. 960623093) to use the Hagan
module failure data history to develop a reasonable preventive maintenance _program
and to replace electrolytic capacitors after a combined shelf and operation life of 10
years.
c.
Conclusions
E1 .6
b.
PSE&G engineering developed good analytical bases to address some of the NRC
concerns. However, some other concerns had not been clearly understood and
rework was necessary to reach an acceptable resolutions.
The failure to evaluate module changes to ensure that an unreviewed safety
question did not exist resulted in a violation of the 10 CFR 50.59 requirements.
However, the posed revisions to the 50.59 process to evaluate equivalent
replacement component should provide the licensee with a better understanding of
the replacement process and ensure better controls on design and system
configurations.
NRC Restart Item 11-12 - Review Adequacy of Fuse Control Program (Closed)
The adequacy of fuse control program was .evaluated previously as documented in
Inspection Report (IR) 50-272; 311 /96-16. In a previous inspection
(50-272;31 _1/96-01 ), the NRC expressed concerns regarding oversized fuses in the*
secondaries of the motor control center (MCC) control power transformers (CPTs).
The NRC had questioned the licensee's evaluation of the acceptability of the
15 Amp fuses in circuits with 300 Volt-Ampere (VA) CPTs. NRC Restart Item 11.12
remained open pending the licensee resolution of the oversized fuse concern.
Observations and Findings
I
The inspectors' review of the oversized fuse concern determined that the licensee
had initiated two change requests, on February 12, 1996, to replace the 15 Amp
fuses with 5 Amp dual element (time delay) fuses in both the vital and ilonvital
MCC _control circuits (2EE-00203 and 2EE-00204 respectively). A design change
package (DCP) to comply with these requests, however, had not yet been prepared
because the project had not heen released for engineering.
..
10
The licensee recognizing the need for better control circuit protection, when they
issued DCP 2EC-3396, which added control circuit transfer switches for Appendix R
alternate shutdown, they replaced the existing 1 5 Amp control circuit fuses with
5 Amp time delay fuses. However, as stated above, the safety evaluation,
approved on April 15, 1996, indicated that only the fuses associated with this DCP
would be replaced; the remaining MCC control circuits would be evaluated
separately by the Electrical Engineering Group.
To evaluate the existence of potential safety concerns associated with this issue,
the inspectors reviewed time-current curve 2131 X from calculation ES.-13.008,
Revision 2, Breaker Coordination, dated October 8, 1996. This curve confirmed
that the 15 Amp fuse would adequately protect the No. 14 AWG control circuit
conductors. However, based on information added to the curve, the 15 Amp fuse
would not protect the No. 20 AWG CPT winding. The licensee stated that the lack
of protection for CPT was riot a concern because each MCC pan was a separate
metal enclosed compartment which would contain any damage to the CPT due to
potential local failures.
The inspectors also reviewed calculation ES-13.005(0), Revision 5, "Salem Unit 2
Penetration Overcurrent Protection.
11
Page 2 of Appendix 9 to this calculation,
"Conax 90°C l
2T Thermal Limit Curves, II showed the thermal capability of No. 10
AW.G containment penetration control circuit conductors. The inspectors concluded
that either the 1 5 Amp fuse or the proposed 5 Amp fuse provided sufficient
protection for the No. 10 AWG penetration conductors.
During the Safety System Functional Inspection (SSFI) of the component cooling
water (CC) system (IR 50-311/96-81), the inspectors determined that sections of
the control circuits in the relay panels were wired with No. 20 AWG conductors.
Based on his review of time-current curve 2131 X and a comparison of the trip
characteristics of the 5 and 1 5 Amp fuses, the inspectors expressed a concern that
the 1 5 Amp fuses would not adequately protect the relay panel circuits wired with
No. 20 AWG conductors. The inspectors also concluded that the licensee's
previous evaluation of PIR 951230143, which originally identified the issue, had
failed to recognize the existence of control circuits with No. 20 AWG conductors.
The licensee documented this discrepancy with Action Request (AR) 961212235.
Discussions with the licensee also revealed that the circuits in question were not
grounded and that field wiring was primarily interconnection of control devices.
Therefore, (1) two grounds on both sided of the load would be required to initiate a
protective action from the fuse; (2) short circuits of field wiring would typically not
require fuse protective action; and (3) short circuits within the relay cabinet,
although unlikely, would generate sufficient energy to blow the oversized fuse.
..
,
11
c.
Conclusion
The inspectors concluded that the proposed 5 Amp time delay fuses would provide
the required component protection while still providing adequate assurance of
freedom from nuisance tripping caused by circuit inrush currents. Although the
currently installed 15 Amp fuses provide limited protection f~r the No. 20 AWG
wiring, their replacement with 5 Amp fu~es was not considered to be an immediate
safety issue. Therefore, this restart item is closed.
The inspectors also concluded that the licensee's evaluation of PIR 951230143 was
deficient in that it failed to recognize that the 15 Amp fuse did not provide the
required protection for the No. 20 AWG control circuit conductors. The licensee's
failure to take adequate corrective actions to address the PIR concern is a v.iolation
of 10 CFR, Appendix B, Criterion XVI, Corrective Action. (VIO 50-272;311/
97-02-01)
Miscellaneous Engineering Issues
E8.1
(Closed) Violation EA94-112-06014, 2.8, Inspection Reports 50-272; 311 /94-80
and 94-13: Inadequate controls to ensure that parts were being correctly identified
and installed.
This violation involved the identification of two examples of inadequate
configuration controls to ensure that correct materials were used during
maintenance and in the installation of design change packages (DCPs) at Salem. In
the first example, two power-operated relief valves (PORVs) internals were found to
be made of 17-4PH rather than 420 stainless steel, as recommended by the vendor
and required by the DCP. In the second example, a Hagan summator module, used
for high steam flow measurement, was found to be not the modified variant
.(special) type required for the application.
In their response to the NRC, letter dated November 1, 1994, PSE&G stated that
- the cause for the material deficiency was lack of attention to details by the staff. In
the first case, the work order planning process, a shared activity between the
project installation group and the maintenance department, did not assure that the
proper valve parts were prestaged, verified, and installed. In addition, the
installation and test engineer and the station planner did not compare the DCP and
the preventive maintenance (PM) work order requirements. In the second case,
PSE&G concluded that the incorrect Hagan module was used as a result of
inadequate configuration control. They found that the two types of module (normal
and special) had identical assembly numbers and arrangement drawings. The only
exception was in the interconnecting loop drawings.
. "
12
The adequacy of the internal PORV material was verified by the NRC during their
review of restart item 11.22, as documented in Section E2.1 of NRC Inspection
Report 50-272;311 /96-12. The inspectors also noted that PSE&G had implemented
the changes described below in the DCP installation and preventive maintenance
process, All responsible personnel had received appropriate process change
training.
When work is shared by multiple departments, a meeting must be held by
project team to assure that all organizations involved understand their
responsibilities and interfaces.
All DCP material will be prestaged within the Nuclear Engineering material
staging areas.
The installation and test engineer assigned to the DCP is responsible for
verifying that all materials issued for installation are consistent with those
specified in the DCP.
The project team, including the installation and test engineer, will review
completed job packages to validate the accuracy of materials and MMIS data
base.
The project manager will assure that, upon completion of the DCP, remaining
material is properly accounted for, entered to folio, or placed in surplus.
The inspectors randomly selected four modifications packages and verified that
materials selected were accurate. In addition, he conducted interviews with
appropriate staff members, including contractors and prestaging area staff, to
ensure that the process used conformed with the guidance provided by the licensee.
The inspectors found that the staff were cognizant of and had used the new
process.
Regarding the second issue, the inspectors noted that the PSE&G had implemented
a Hagan module reconfiguration program. This program involved review and
verification of module arrangements and interconnect drawings; updating vendor
manuals, bills of material and history cards; and labeling front and rear of modules.
The Hagan module configuration control issues were review~d in detail as a part of
restart issue II. 14, as documented in NRC IRs. 50-272; 311-96-06.
Interviews of l&C staff confirmed that technician involved in the removal and
installation of Hagan modules had been instructed and were informed of the need
for verifying the correct identification of the modules.
Based on the above, the inspectors concluded that the actions taken by the licensee
to address material deficiencies were acceptable. This item is closed.
13
E8.2
(Open) Unresolved Item 50-272/94-04-01 and 50-311 /94-04-01: Power Range
Neutron Detector Qualification Issue
This issue was originally identified during. a 1993 Quality Assurance (QA) audit
(93-101 ). The QA auditor raised a concern that pertained to the qualification of the
power range neutron detectors at Salem. The concern was that, following a small
size steam line break (0.1 ft2 to 0.25 ft2) inside the reactor containment, the harsh
environment in the containment could cause the power range neutron detectors to
malfunction and the control rod to move in an outward direction, before the reactor
trip.
This item was reviewed and closed during a December 1994 inspection, IR 94-33.
During this inspection, the NRC reviewed again engineering evaluation No. EE-S-C-
NIS-CEE-0702, Revision 1, entitled "Evaluation of Applicability of Salem
Environmental Qualification Program Requirements for Nuclear Instrumentation
System Flux Detectors," which was used for the December 1994 closure of this
item. The NRC concluded that the document did not provide sufficient basis for the
closure of this item.
Each Salem unit uses four power range neutron detector assemblies supplied by
Westinghouse. Each assembly, consisting of two neutron detectors (top and
bottom detectors), has three electrical connections: one for the power supplies (the
power supply can be varied from 300 Vdc to 1500 Vdc, but at Salem it was set at
800 Vdc); one for the top detector signal; and one for the bottom detector signal.
The signal range for 100% full power could be from 100 microamperes to
3:6 milliamperes. The cables connecting the three detector terminals and the field
cables at the nearby junction box are aluminum-insulated coaxial cables,
hermetically sealed to the detector by a series of metallic joints and terminated with
ceramic-to-metal seals. The connections between the aluminum-insulated coaxial
cables and the field cables (Rockbestos RSS-6-108, XLPE insulated and polyolefin
jacketed triaxial cables) used Westinghouse crimp-on Kynar triaxial connectors,
sealed with Raychem type WCSF seals, making all connections moisture and
condensate resistant. The lengths of the field cables from the junction boxes to the
containment penetrations varies from 78 feet to 180 feet. The junction boxes were
in concrete wells above the neutron detector assemblies. The concrete wall
between the junction boxes and the reactor vessel provided substantial shielding
from neutron and gamma radiation.
The functional capabilities of the *affected components following a postulated small
size steam line break are discussed as follows:
Field cables and Raychem Seals Due to Temperature Increase
According to the licensee, the original field cables were Westinghouse-supplied
coaxial cables. In a Technical Bulletin dated February 27, 1986, Westinghouse
recommended to replace the field cables with better quality cables. One of the
cables recommended by Westinghouse was Rockbestos RSS-6-108 triaxial cable.
The licensee completed the cable replacement for both units in 1991 .
..
14
The licensee stated that the new field cable had been environmentally qualified by
the cable manufacturer (Rockbestos). The qualification tests were documented in
Rockbestos Test Report QR-6802, "Report on Qualification Tests for Rockbestos
Adverse Service Coaxial, Twinaxial, and Triaxial Cable Generic Nuclear Incident for
Class 1 E Service in Nuclear Generating Stations," Revision 1, dated July 2, 1987.
The inspectors' review of this report determined that the Rockbe*stos tests had
qualified the cables for general application. For special applications, where current
signals are low (100 microamperes), the impact of leakage currents (caused by low
insulation resistance) on the accuracy of the instruments should be evaluated.
Rockbestos tested the cables to a maximum temperature of 346.2° F and a
maximum pressure of 122.1 psig. The test report shows that, after .eight hours, the
measured insulation resistance (IR) dropped to 0.54 x 106 ohms. The measured IR
recovered to 36 x 106 after 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />, when the temperature dropped to 25*1.3 ° F
and the pressure to 18.9 psig, and* remained at a much higher resistance
afterwards.
To address the inspectors' concern, the licensee calculated the leakage current of
the field cable using a minimum IR of 36 x 106 ohms and a cable length of 180 feet.
The licensee's bases for choosing the above IR value were: 1) the steam lines are
located at a much higher elevation than that of the triaxial cables, therefore, steam
impingement following a small size steam line break was improbable; 2) because the
triaxial cables are inside electrical conduits, the thermal lag effect would limit cable
temperature rise during the thermal transient before the reactor trip, estimated by
Westinghouse to be several minutes. The licensee estimated the cable temperature
would be not higher than 250 ° F and the containment pressure 4 psig (containment
high pressure trip set point). The inspectors considered the licensee's engineering
reasoning and the calculated leakage current acceptable.
The Raychem seals are located inside junction boxes in concrete wells, The *
temperature increase inside the junction boxes. during the short duration following a
small size steam line break would be insignificant. Therefore, the seals IR would
not be affected by the temperature increase.
Field cables and Raychem Seals Due to Radiation
The triaxial cable used in the Rockbestos tests was preaged to 40 years and
irradiated to a total integrated dose (TIO) of 200 x 106 rads before the LOCA (loss
of coolant accident) test.
The Raychem seals had also been qualified for a
radiation TIO of 200 x 106 rads as indicated in Salem's EO data sheets. To
demonstrate that the expected Salem total integrated dose (TIO) would be less than
the qualification test dose, the licensee prepared Calculation DS2.6-0063,
"Estimates of Gamma and Neutron Fluxes at the Top of the Reactor," dated
January 22, 1 997. The inspectors' review of this calculation determined that the
licensee took credit for the shielding effect of various concrete thickness between
the junction boxes and the reactor cavity. Based on 16 inches concrete thickness,
I
15
the licensee calculated that it would take 96 years for the TID (gamma plus neutron)
to reach the qualified value of 200 x 106 rads (accident dose was negligible in this
case). The inspectors' review of construction drawing determined that the concrete
thickness between the reactor core and the junction boxes was greater than
17 inches.*
Based on his review of the above calculation, the inspectors concluded that the
licensee had adequately demonstrated the radiation capabilities of the field cable
and Raychem seals in the postulated small steam line break accident.
Design Life of Power Range Neutron Detectors
The power range neutron detectors supplied by Westinghouse are boron depletion
type uncompensated ionization chambers. Westinghouse technical manual for
11W-23686 Power Range Uncompensated Ionization Chamber,
11 dated
September 1, 1972, specifies that the neutron exposure required to cause a 10%
decrease in sensitivity (each section) is 2 x 1019 nvt (neutron-density velocity time),
which is the same as N/cm 2 (neutrons per square centimeter). In Salem QA Audit 93-101, the auditor stated that the total neutron dose design limit was 2 x 1018
N/cm2 and estimated that the detector design life was 8 years. Because the power
range nuclear instrumentation is required to be calibrated frequently per Technical
Specification requirements, 10% decrease in the detector sensitivity does not affect
the calibrated accuracy of the nuclear instrumentation.
During this inspection; the licensee calculated the neutron fluence of both units at
the reactor cavity inner concrete surface where the neutron detectors are located.
The calculation showed that the 32 EFPY (effective full power year) neutron fluence
at the highest flux plane was 4.89 x 1017 N/cm2 * The licensee stated that 32 EFPY
is equivalent to a neutron fluence of 40 normal plant operation years. The results of
this calculation imply that the design life of the power range neutron detectors is
much higher than the expected neutron fluence.
The inspectors also reviewed Salem Calculation DS1 .6-0132,
11PTS (Pressure
Thermal Shock) Calculations with New Chemistry Compositions,
11 dated
November 8, 1995. This calculation showed that the 32 EFPY fluence for the inner
wall of the reactor vessel at the maximum flux plane was 1.63 x 1019 N/cm2, which
was also lower than the specified design fluence of the neu~ron detectors. Based
on the above, the inspectors concluded. that the design life of the power range
neutron detectors was sufficient'to cover 40 years normal plant operation.
The inspectors concluded that the licensee had provided sufficient additional
documentation to demonstrate the ability of the powe.r range neutron detectors to
perform their safety function in the postulated Salem environment. The restart
issue of this item is resolved.
16
In the enclosure entitled "Master List of Electrical Equipment for Environmental
Qualification, Salem Generating Station, Units 1 and 2" to a letter to the NRC dated
June 8, 1984, the licensee requested environmental qualification (EQ) exemption for
the neutron detectors for both units (not in the scope of 10 CFR 50.49). This
exemption request was accepted by the NRC in the subsequent Safety Evaluation
Report dated January 14, 1985. Exemption from qualification of this equipment
(equipment item no. 59) was also discussed in the Technical Evaluation Report
(TER) issued by Franklin Research Center on July 15, 1982.
However, based on the conditions discussed in the above paragraphs, the power
range* neutron detectors could be subject to a brief harsh environment following a
certain size high energy line break. Region I will refer this .issue to NRR for
resolution. This item remains open pending the result of additional NRR review to
confirm that this equipment, although demonstrated qualifiable as discussed above,
does not require qualification.
E8.3
(Closed) Unresolved Item 50-272/89-13-07 and 50-311 /89-12-07: Inadequate
Cable Separations and Electrical Isolation for Various Regulatory Guide (RG) 1.97
Instrumentation
This item pertains to cable separation and electrical isolation deficiencies identified
by the licensee in 1989 associated with RG 1 .97 instrumentation. These
deficiencies were documented .in the 1989 NRC RG 1.97 inspection report (89-13;
89-12).
The licensee's resolution of these issues was evaluated during the review of item
11.21 of the NRC restart plan for Salem, as documented in Section E1 .3 of this
report. Based o*n the satisfactory closure of that item, this item is also closed.
E8.4
(Closed) Unresolved Item 50-271: 311/93-82-13: Molded Case Circuit Breaker
Testing
a.
Inspection Scope
b.
Inspection reports 93-82 and 96-16 documented the NRC's concerns regarding the
acceptability of the licensee's testing of the molded case circuit breakers (MCCBs.)
The licensee's test program was limited to those circuit breakers that provided
protection for the containment electrical penetrations and did not include other
safety-related MCCBs such as those used for 1 E to non-1 E isolation. The
inspectors reviewed the licensee's progress to expand the test program to other
Observations and Findings
The inspectors reviewed the licensee's evaluation No. S-C-VAR-EEE-1057,
Revision 0, "Tabulation of Molded Case Circuit Breakers and Parameters, dated
April 30, 1996." The inspectors observed that the scope of this engineering
evaluation was limited to safety-related MCCBs and those nonsafety-related MCCBs
that provided protection for the containment electrical penetrations. The inspectors
noted that the eval.uation made the following recommendations:
17
All safety-related MCCBs should be included in the circuit breaker test
program.
The NEMA AB-4 acceptance criteria range is too broad to provide assurance
that the results of the coordination studies are not affected.
A procedure acceptance criteria range of +/- 10% would provide reasonable
assurance that the results of the coordination studies would remain
acceptable.
Test results outside the procedure limit but inside the NEMA limit must be
evaluated by the System Engineer for acceptability.
The inspectors reviewed maintenance te.st procedure SC.MD-ST.ZZ-0004(0),
Revision 7, "Containment Penetration Molded Case Circuit Breaker Test," dated
September 26, 1996. A note on page 6 of the procedure indicated that MCCBs
with thermal magnetic trip elements require trip testing in both the overcurrent
(300%) and instantaneous regions. The procedure referenced calculations
ES-13.010 and ES-13.005, Salem Unit 1 and 2 (respectively) Penetration
Overcurrent Protection, for the electrical protective device settings. The inspectors'
review of calculation ES-13.005, Revision 5, dated March 27, 1996, determined
that Attachment 3 to this calculation was a summary of the protective devices,
including their trip setpoint and required response time. The inspectors found that:
(1) for some MCCBs, Attachment 3 did not list the required response time (e.g.,
breakers 21 RM-1 and 21 RM-3); (2) for some MCCBs, the required response times in
Appendix 3 were different than those listed in the test procedure (e.g., the response
times specified for breaker 2CCDC-20 were 9-45 seconds and 8-35 seconds,
respectively); and (3) Attachment 3 did not contain setpoints for the adjustable or
non-adjustable instantaneous MCCB trips associated with the thermal-magnetic
breakers.
The inspectors' review of Attachment 3 also noted discrepancies between the
required test currents and response times for the MCCBs associated with different
pressurizer heater circuits (e.g. 2GP15 and 2EP1 X.)
In response to the inspectors' concerns regarding the identified discrepancies
between calculation ES-13.005 and procedure SC.MD-ST.ZZ-0004(0), the licensee
stated that they had already had identified similar problems and presented ten
problem identification reports (PIRs) and an action request (AR) that addressed the
licensee-identified discrepancies. PIR 960612148 specifically addressed
discrepancies between the calculation and the procedure. Based on this PIR, the
corrective actions had been completed on July 26, 1996. Yet, a review of the
procedure showed that the procedure had been revised t_wice since July 1996
(Revisions 6 and 7 were dated August 30 and September 26, 1996, respectively),
but the discrepancies remained .
.
- .,
!-
18
Attachment 2 to calculation ES-13.006, Revision 2, "Breaker and Relay
Coordination Calculation of Safety-Related AC System," dated October 18, 1995,
provided coordination plots for the 240 Volt vital ac system motor control centers
(MCCs.) The inspectors confirmed that the coordination plots for the adjustable trip
MCCBs used the trip setpoint with a +/- 10% tolerance band. The inspectors also
noted that the nonadjustable instantaneous region of the thermal magnetic MCCBs
used the manufacturer's time-current curve which included a manufacturing
tolerance band. It did not appear that any other tolerances were used for the
instantaneous region of the MCCBs in the coordination studies.
The inspectors reviewed the test results of selected MCCBs that had been
performed to demonstrate the functional capability of the containment electrical
penetration protective devices. Section 4.8.3.1.a.2 of the Salem TS require_d that
the protective devices for the low voltage electrical penetrations be included in the
surveillance program. The TS also required that every 18 months, at least 10% of
the MCCBs providing penetration protection be demonstrated operable by a
functional test. The inspectors observed that some of the electrical penetrations
were protected by the instantaneous portion of the MCCB trip devices and that test
procedure SC.MD-ST.ZZ-0004 required the adjustable instantaneous trip for both
thermal-magnetic and instantaneous only breakers to be verified. The inspectors
reviewed the test data for selected thermal-magnetic breakers with adjustable trips
(2GP14X, 2GP1 OX, 2EP2X, 2EP3X and 2G5X) and found that none of the
instantaneous trips had been tested. The inspectors confirmed that some of those
breakers required the breaker to operate in the instantaneous region to assure
containment penetration protection. Therefore, in these cases, the licensee failed to
demonstrate the functional capability of those MCCBs to maintain containment
integrity.
The inspectors reviewed maintenance test procedure SC.MD-ST.ZZ-0005(0),
"Molded Case Circuit Breaker Maintenance." This procedure was in the review
stage as Revision 3 DRAFT, undated. The inspectors found that neither Revision 2
nor the draft in review included a test of the instantaneous region of the thermal-
magnetic MCCBs with non-adjustable magnetic trips.
The inspectors reviewed the results of selected MCCB testing between June 1991
and January 1996. The inspectors determined that the testing of thermal-magnetic
breakers with the nonadjustable instantaneous trips had not tested the breakers
instantaneous trip mechanism.* In response to the NRC concerns in this area, the
licensee contacted the manufactcirer who stated that the thermal trip curve could be
extended into the instantaneous region and that they expected the breaker to
operate on the thermal mechanism even if the instantaneous mechanism was not
functional. MCCBs operating in this manner, however, should be replaced.
..
19
c.
Conclusion
The inspectors concluded that the licensee's corrective actions for PIR 960612148,
issued to correct discrepancies between calculation ES-13.005 and the procedure
SC.MD-ST.ZZ-0004(0), were inadequate in that, based on the PIR, the actions_-were
completed on July 26, 1996, yet two subsequent revisions of the procedure failed
to correct the discrepancies. The inspectors also concluded that the licensee's
resolution of the_ above PIR was insufficient to identify all discrepancies between the
two documents. This is another example of inadequate corrective ~ction.
(VIO 50-272;311 /97-02-01)
The inspectors further concluded that the licensee's implementation of the MCCB
surveillance test procedure requirements was inadequate, in that they failed to test
some thermal-magnetic MCCBs in the instantaneous region as required by the test
procedure. (VIO 50-272;311/97-02-02)
Future test plans for all safety-related MCCBs were generally acceptable, except
that they did not provide sufficient guidance for resolution of test results found
outside the nominal +/- 10% acceptance criteria. This item will remain open pending
the NRC review of the licensee's actions to resolve the above deficiencies and to
ensure that sufficient guidance is provided regarding inadequate test results in the
newly revised maintenance procedure.
-- E8.5
(Closed) Unresolved Item 50-311193-82-04: Overvoltage Effect on Safety-Related
Motors and Control Relays
a.
Scope
b.
The NRC EDSFI inspection report (IR 93-82) identified a potential overvoltage
concern of up to 4500V. This item was last reviewed in IR 96-13. PSE&G
modified the power distribution system to reduce *the potential for overvoltage.
Calculation ES-15.004, Load Flow and Motor Starting Calculation, Revision 1, dated
October 9~ 1996, established an analytical limit of 4400V and a maximum voltage
with the load tap changer in operation at 4368V. The closure package for this item
was presented to the management review committee at meeting 96-111 on
December 6, 1996. This closure package identified the Unit 2 procedures that were
revised and issued incorporating the li_mits. The inspectors reviewed selected
procedures for incorporation of the procedure limit of 4360V identified in the
closure package. The value of 4360 includes a 10% allowance for the meter
inaccuracies.
Observations and Findings
The inspectors reviewed the procedure that had been identified in the closure
package as requiring revision. In general, he found that the procedures had been
correctly revised. -He also found some minor discrepancies as described below:
-.
.
. " .
20
Procedure S2.0P-ST.DG-0001, Revision 27, "2A Diesel Generator Surveillance
Test," had been revised to specify an acceptance criteria for voltage of 4.36kV in
Section 5.2.11. However, Attachment 5, "2A Diesel Generator Failure Report,
Acceptance Criteria Not Satisfied," of the same procedure continued to show a limit
of 4520V.
Procedure S2.0P.SSP-0002(Q), Revision 15, "Engineered Safety Features Manual
Safety Injection 2A Vital Bus," simulates an accident loading of the emergency
diesel generator. The inspectors confirmed that Section *5 and Appendices 5, 7 .8
and 12 of the procedure documented the correct maximum voltage limit of 4360V.
However, Section 1, "Purpose," incorrectly identified the maximum voltage limit as
4580V, and Appendix 9, "Failure Report," did not record the voltage limit. The
licensee stated that the purpose of Section 1 was to demonstrate compliance with
the TS 4.8.1. 1 in e"ffect and indicated that when license change request (LCR)
95-36, which revised the TS limits for voltage and frequency, had been approved by
the NRC, they would also revise Section 1.
In procedure "S2.0P-AR.ZZ-0009(Q), Revision 16, "Overhead Annunciator
Window J," window J38 for the group buses had been revised to include the new
limit. The group buses feed nonsafety-related loads. However, windows J17, J18,
and J19 for the vital buses did not address overvoltage.
Drawing 206061, Revision 30, "No. 2 Unit 4160V, Vital Buses One Line," indicates*
that station power transformers (ST A) Nos. 23 and 24 are the preferred power
supplies to the vital 4kV buses. The inspectors reviewed procedure S2.0P-AR.ZZ-
001 O(Q), Revision 7, "Overhead Annunciator Window K," and found that
Section 1.0, "Causes," for windows K27 (23 Station Power Transformer) and* K35
(24 Station Power Transformer) indicated a possible cause for the alarm to be
voltage outside the range of 4.15kV to 4.43kV. This statement would imply that
the alarm setpoint is outside the normal range of 4.22kV to 4.36kV as given in
Section 3.0, Operator Actions, for these same windows.
The inspectors consider the above inconsistencies minor in that the more important
sections of the procedures had been correctly revised and the alarm issues would
not prevent the operator to correctly respond to the alarm. He discussed the
inconsistencies with the licensee. They stated that the documents involved would
be revised to reflect the inspectors' observations.
c.
Conclusions
The inspectors concluded that the majority of the affected procedures had been
correctly revised to incorporate the lower acceptabl.e voltage of 4360V. Some
observed inconsistencies were minor and did not affect the safety function of the
vital power supply. Therefore, this item is closed.
>*
E8.6
21
(Closed) Unresolved Item. 50-272: 311 /94-32-05: Proposed Limiting Design-Basis
Transient for POPS
This issue pertains to new design-basis transient limits proposed by PSE&G for the
pressurizer overpressure protection system. The adequacy of the new proposed
limits was reviewed as part of Salem Restart Item 11.20 and approved by NRR, as
detailed in,NRC in~pection report 50-272; 311/96-07 and Section E1 .4, above.
This item is closed.
E8.7
Review of UFSAR Commitments
A recent discovery of a licensee operating their facility in a manner contrary to the
Updated Final Safety Analysis Report (UFSAR) description highlighted the need for a
special focused review that compares plant practices, procedures and/or parameters
to the UFSAR descriptions. While performing the inspections discussed in this
report, the. inspectors reviewed the applicable portions of the UFSAR that related to
the areas inspected. The inspectors verified that the UFSAR wording was
consistent with the observed plant practices, procedures and/or parameters.
V. Management Meetings
XI
Exit Meeting Summary
The inspectors presented the inspection results to members of licensee management
at the conclusion of the inspection on March 7; 1997. The licensee acknowledged
the findings presented.
The inspectors asked the licensee whether any material examined during the
inspection should be considered proprietary. No proprietary information was
identified.
PARTIAL LIST OF ATTENDEES
Public Service Electric and Gas Company
G. Boerschig
G. Cranston
N. Conicella
S. Funsten
L. Hajos
P. Moeller
G. Nagy
M. Rencheck
E. Villar
Manager, Nuclear Electrical Engineering
Manager, Nuclear Design Engineering
Salem Projects
Engineering
Supervisor, Design Engineering
Principal Staff Analyst
System Engineering
Manager Salem System Engineering
Licensing* Engineer
U. S. Nuclear Regulatory Commission
R. Fuhrmeister
R. Quil'.k
L. Thonus
Senior Reactor Engineer
NRC Contract Engineer
.
-,
.., .
CAG
CA/QS
CCHX
CROM
CRs
eve
ECAC
EOG
l&C
INPO-
LER
MS IVs
N/A
NBU
NRC
NTOC
OEF
OTSC
PSE&G
SERT
.SI
SIRA
SNSS
SORC
SRG
TOR
TR Gs
TRIS
TS
- .,
- ...
22
LIST OF ACRONYMS USED
Action Request
Corrective Action Group
Corrective Action Program
Corrective Action and Quality Services
Component Cooling Heat Exchanger
Control Rod Drive Mechanisms
Condition Reports
Centrifugal Charging
Emergency Control Air Compressor
Emergency Operating Procedures
Emergency Response Guideline
Hilti Drop-In
Instrumentation and Controls
Institute of Nuclear Power Operations
lnservice Inspection
Licensee Event Report
Management Review Committee
Not Applicable
Nuclear Business Unit *
- Nuclear Regulatory Commission
Nuclear Training Oversight Committee
Operating Experience Feedback
On-The-Spot Change
Public Document Room
Post-Maintenance Testing
Public Service Electric and Gas
Primary Water Stress Corrosion Cracking
Reactor Coolant Pump
Reactor Vessel Level Indicating System
Significant Event Response Team
Safety Injection
Salem Integrated Readiness Assessment
Senior Nuclear Shift Supervisor
Station Operations R_eview Committee
Safety Review Group
Senior Reactor Operator
Technical Document Room
Training Review Group
Tagging Request Inquiry. System
Technical Specification
Updated Final Safety Ar:ialyses Report