ML18102A778
| ML18102A778 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 01/21/1997 |
| From: | Wiggins J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Eliason L Public Service Enterprise Group |
| Shared Package | |
| ML18102A779 | List: |
| References | |
| NUDOCS 9701270181 | |
| Download: ML18102A778 (7) | |
See also: IR 05000311/1996081
Text
Mr. Leon R. Eliason
Chief Nuclear Officer & President
Nuclear Business Unit
January 21, 1997
Public Service Electric and Gas Company
P.O. Box 236
Hancocks Bridge, New Jersey 08038
SUBJECT:
SALEM SAFETY SYSTEM FUNCTIONAL INSPECTION REPORT 96-81
Dear Mr. Eliason:
On January 8, 1996, the NRC completed a Safety System Functional Inspection (SSFI) at
your Salem Unit 2 facility, examining the component cooling (CC) system. The enclosed
report presents the results of that inspection.
We conducted this SSFI inspection to independently assess the scope, depth and quality of
your efforts covered by your Updated Final Safety Analysis Report Project Plan. Other
NRC activities have included our overall review of the Project Plan and inspections
conducted in the May to October 1 996 time period to monitor and assess your
implementation of that Plan. We will finalize our views on the adequacy of your activities
after we meet with you. That meeting will provide you an opportunity to describe the
results of your overall efforts in the licensing and design bases areas. We anticipate
scheduling that meeting in February, 1997.
The team noted that significant improvements were made to the component cooling
system during the current outage. These improvements included the completion of a
system flow balance, resolution of instrument calibration errors, and the completion of a *
significant amount of corrective and preventive maintenance activities. Also, the team
noted that your Updated Final Safety Analysis Report Project Plan review effort identified
and resolved a number of licensing basis discrepancies. However, the team identified two
design basis issues that call into question the system's ability to perform its safety
function.
A single failure of the CC room ventilation could adversely affect 2 CC pumps, leaving only
one CC pump available for long term operation. For certain accident scenarios, your
Emergency Operating Procedures (EOPs) rely on having at least two CC pumps functional.
Therefore, the EOP need for two pumps appears inconsistent with the ventilation system
design. The second issue is related to the operation of the CC pumps during certain
postulated accident conditions. The EOPs require that operators manually start a single CC
pump following certain accidents. For a short period of time, after switching emergency
core cooling pump suction from the refueling water storage tank to the containment sump,
the CC pump may operate at flow rates beyond its current design limit. Operating a CC
pump at high flow rates is inconsistent with its design basis and our review of your prior
analysis of this condition raised several questions. In addition, the team also identified
weaknesses in the quality of several other engineering calculations and analyses associated
with the CC system.
9701270181 970121
ADOCK 05000311
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270012
. "*
_ ....
Mr. Leon R. Eliason
2
It also appears that a number of the design and analysis issues discussed in this report
have strong connections to the existing restart concern associated with emergency diesel
generator (EOG) loading. You should carefully review the results of this report and factor
the insights gained from that review into your short and long term approach to the EOG
loading issue. In addition, as discussed during our January 8, 1997 exit meeting, you
should consider the need for measuring the extent to which the identified design and
analysis issues affect the condition of other important plant systems, structures and
components, along with the need for a prioritized plan to address those conditions.
At the February 1997 meeting, along with your discussion of the integrated results of your
licensing and design bases conformance activities, you should also be prepared to discuss
the impact of the SSFI findings on any conclusions reached as a result of those activities.
In addition, you should address: (1) your plans for any "extent of condition" reviews; (2)
the impact of this report's results on the EOG loading issue; and, (3) whether any of the
report results constitute new or revised restart items.
Since the final resolution of the SSFI findings are pending further evaluation by your staff
and by the NRC, this inspection report does not address enforcement actions. Inspection
findings that are candidates for enforcement are identified in this report as unresolved
items . .You will be notified in future correspondence of our decision on any enforcement
actions.
In accordance with 10 CFR 2. 790 of the Commission's regulations, a copy of this letter
and its enclosure will be placed in the NRC Public Document Room.
Docket No. 50-311
Enclosures:
Sincerely,
.* James T. Wiggins, Director
Division of Reactor Safety
1 . NRC Region I SSFI Report No. 50-311 /96-81
2. Slides from Exit Meeting
o
I
Mr. Leon R. Eliason
3
cc w/encl:
L. Storz, Senior Vice President - Nuclear Operations
E. Simpson,. Senior Vice President - Nuclear Engineering
E. Salowitz, Director - Nuclear Business Support
C. Schaefer, External Operations - Nuclear, Delmarva Power & Light Co.
D. Garchow, General Manager - Salem Operations
J. Benjamin, Director - Quality Assurance & Nuclear Safety Review
D. Powell, Manager, Licensing and Regulation
R. Kankus, Joint Owner Affairs
A. Tapert, Program Administrator
R. E. Selover, Esquire
M. Wetterhahn, Esquire
P. MacFarland Goelz, Manager, Joint Generation
Atlantic Electric
Consumer Advocate, Office of Consumer Advocate
William Conklin, Public Safety Consultant, Lower Alloways Creek Township
Public Service Commission of Maryland
State of New Jersey
State of *Delaware
--*
Mr. Leon R. Eliason
Distribution w/encl:
Region I Docket Room (with concurrences)
J. Wiggins, DRS
Kay Gallagher, DRP
Nuclear Safety Information Center (NSIC)
L. Nicholson, DRP
S. Barber, DRP
R. DePriest, DRP
G. Kelly, DRS
N. Della Greca, DRS
G. Morris, DRS _
S. Klein, DRS
L. Prividy, DRS
J. Trapp, DRS
D. Screnci, PAO
NRC Resident Inspector
PUBLIC
DRS Files
Distribution w/encl: (Via E-Mail)
L. Olshan, NRR
W. Dean, OEDO
J. Stolz, PDl-2, NRR
M. Callahan, OCA
Inspection Program Branch, NRR (IPAS)
R. Correia, NRR
R. Frahm, Jr., NRR
DOCUMENT NAME: A:\\SAL9681.INS
4
To receive a copy of this document, indicate in the box: *c* = Copy w' out attachment/encl
OFFICE
RI/DRS
NAME
JTrapp
DATE
01/17/97
01 t- /97
"E" = Copy with attachment/enclosure "N" = No copy
RI/DRS
RI/
JWiggins
01121191
01 / /97
OFFICIAL RECORD COPY
EXECUTIVE SUMMARY
Salem Nuclear Generating Station, Unit 2
NRC Inspection Report 50-311 /96-81
The objective of this inspection was to conduct an independent inspection to determine if
the Salem Unit 2 component cooling (CC) system would perform its intended safety
function.
Operations
The team identified an operating procedure weakness in that under certain *
conditions, a single active failure of certain equipment is not supported by the
emergency operating procedures (EOPs). The single failures identified by the team
were the failure of the 22/23 CC pump room ventilation (2VHE-34) or the failure of
the Train C electrical power. The loss of Train C electrical power would prevent the
operation of the 21 CC pump room ventilation (2VHE-33) and the 23 CC pump. A
single failure of this equipment could adversely affect the performance of two CC
pumps leaving only one CC pump available. For certain accidents, the EOPs require
two CC pumps be operating. The licensee failed to evaluate this vulnerability in
1995 when administrative controls were developed requiring 3 operable. CC pumps
(Section 03.1 ).
'
The CC system normal and abnormal operating procedures were recently upgraded
and were of good quality (Section 03.2).
The procedures and lesson. plans used for CC training were of high quality and
appropriately complete for evaluation of operator knowledge and abilities on the
system (Section 05.1 ).
Maintenance
The team identified that Publip Service Electric and Gas (PSE&G) had no
documented calculations to support the CC flow acceptance criteria used in the
flow balance test. Final conclusions regarding the results of the flow balance test
could not be made pending completion of these documented calculations (Section
M1 .1 ).
In general, PSE&G was adequately implementing testing required by the surveillance
testing program. However, the team identified several valves where controls were
not in place to ensure that they would be periodically tested (Section M1 .2). Also,
an error was noted in the battery surveillance test procedure and the 1993 battery
test data was not properly evaluated to determine battery degradation (Section
M3.2) *
iv
-: .-,-- ..
PSE&G did not provide an adequate documented technical basis for justifying the
5 % instrument measurement uncertainty assumption used in CC heat exchanger
performance calculations. Acceptance criteria for assessing the as-found condition
of the CC room coolers had not been established and room cooler service water and
air flow rates were not being monitored for assuring equipment operability (Section
M1.3).
The team concluded that PSE&G failed to repair the CC radiation monitors in a
timely manner since they had been out of service for over a year (Section M2.1 ).
Good corrective actions were being taken to identified CC equipment problems
(Section M2.2).
The team identified several CC pump room dampers that were closed which was
inconsistent with information on applicable ventilation system drawings. These
deficiencies demonstrated inadequate configuration control of ventilation equipment
needed to support CC system operability (Section M3.1 ).
The CC heat exchanger thermal performance computer model is not conservative
(Section M1 .3).
Engineering
The team identified a c*ondition where the operation of the CC pumps appears
inconsistent with documented design limits. The team concluded that the CC
pumps will probably be at or near runout conditions when the residual heat removal
heat exchanger outlet valves are automatically opened on low refueling water
storage tank level during a postulated loss of coolant accident. Component cooling
pump operation at runout during these conditions has not been adequately analyzed
by the license~. Consequently, the CC pumps may be adversely affected if
sufficient net positive suction head (NPSH) is not available, and the pumps are
subjected to the effects of cavitation (Sections 03.1 and E1 .1 ).
In general, design basis information and calculations were available and retrievable.
However, the team identified weaknesses in several calculations. The most
significant weaknesses were noted in the CC pump NPSH, thermal overload heater
sizing, and molded case circuit breaker overcurrent setting calculations. In these
cases, inadequacies in calculation methodology or assumptions invalidate the
conclusion of the calculation. In addition, the team also noted that the licensee ~
failed to document a number of engineering judgements and assumptions. In these
cases, the missing engineering judgements and unsubstantiated assumptions did not
invalidate the results of the calculations (Sections E1.1, E1 .4, E3.1, and E3.4).
v
The design change to place the thermal overload (TOL) heaters in service resulted in
the installation of TOL heaters without an adequate documented design basis. The
team concluded that the licensee had not maintained document control of the TOL
relay heaters associated with the CC system and other safety-related systems. - The
team identified heaters existed in motor operated valve (MOV) circuits that were not
based on the existing calculated basis. The team also concluded that a change
document to the design calculation did not provide an adequate documented basis
for the installed TOL heaters for 30 safety-related MOVs (Section E1 .4).
The team concluded that the design basis documents for the CC system radiation
monitor setpoints were inconsistent. The team also determined that the CC ;, *. -
radiation monitor setpoints may be inappropriately set to high. These radiation* __
monitors are not safety-related and are not used to calculate off site radioactive*
releases (Section E1 .5).
The team concluded that the licensing basis descriptions for the CC system were,
with a few minor exceptions, consistent with the actual plant design. The team -
concluded that the CC UFSAR Macro-Review was a *good initiative and identified
and corrected several updated final safety analysis report (UFSAR) discrepancies
(Section E8.2).
The team noted that the EOPs provide instructions to isolate component cooll-ng
water from the post accident sampling system (PASS) heat exchangers during
accident conditions. The PASS heat exchangers are provided cooling water using
temporary hoses from the demineralized water system. Operation of PASS using
demineralized water during postulated accident conditions is inconsistent with the
design and licensing basis (Section 8.1-).
The team concluded that the development of the technical standards program was a
positive initiative. However, the team noted that the standards did not include a
technical justification for acceptance of existing conditions in the plant. The team
considered this to be a program weakness. In addition, the team identified one case
where work in progress, during the development of the technical standards, was
not coordinated with the developing standard (Section E3.2).
The team concluded that the CC system drawings were generally accurate. Jhe
licensee initiated actions to correct the minor discrepancies identified by the team
(Section E3.3) *
vi