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#REDIRECT [[IR 05000255/1997201]]
{{Adams
| number = ML18066A314
| issue date = 10/01/1998
| title = Provides Update to Design Insp Action Items Re Insp Rept 50-255/97-201 Conducted on 970916-1114.Util Recommends That NRC Consider Scheduling Efforts Early in 1999 to Review Insp Items for Closure Based on Completion Dates for Items
| author name = Haskell N
| author affiliation = CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.),
| addressee name =
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000255
| license number =
| contact person =
| document report number = 50-255-97-201, NUDOCS 9810070265
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE
| page count = 55
}}
See also: [[see also::IR 05000255/1997201]]
 
=Text=
{{#Wiki_filter:A CMS Energy Company October 1, 1998 U.S. Nuclear Regulatory
Commission
Attn: Document Control Desk Washington
D.C. 20555 Palisades
Nuclear Plant 27780 Blue Star Memorial Highway Covert. Ml 49043 DOCKET 50-255 -LICENSE DPR-20 -PALISADES
PLANT * Tel. 616 764 2276 Fax.* 616 764 2490 Nathan L. Ha.ks/I Director.
Licensing
OCTOBER 1, 1998 UPDATE TO DESIGN INSPECTION
ACTION ITEMS During the period from September
16 through November 14, 1997, the NRC conducted
a design inspection
at the Palisades
Nuclear Plant. By letter dated December 30, 1997, the NRC issued Inspection
Report No. 50-255/97-201, and requested
a response within 60 days detailing
our plans to complete the corrective
actions required to resolve the open items listed in Attachment
A of the inspection
report. Contained
within our March 2, 1998 response was a single commitment
to provide the NRC a status of our progress in completing
actions associated
with each open inspection
item. The purpose of this commitment, in part, was to assist the NRC in planning for follow-up
review and closeout of these items. Attachment
A of this letter contains the text of each open inspection
item from the December 30, 1997 inspection
report, followed by our 60 day response as submitted
in our March 2, 1998 letter, followed by the status of associated
action as of October 1, 1998. This status includes the results of our investigations
and corrective
actions, along with planned completion
dates for ongoing actions. Attachment
B contains similar information
for programmatic
issues related to inspection
findings.
_J Based on completion
dates for the remaining
open items, we recommend
that NRC consider scheduling
efforts early in 1999 to review inspection
items for closure. A review of completion
dates for open items indicates
that a majority of actions will be completed
by the end of 1998. 9810070265
981001 PDR ADOCK 05000255 G PDR 
-. . .:.; * * -.. -Sl:JMMAR¥-'-8F
COMMITMENTS
This letter closes the March 2, 1998 commitment
as .restated
below, and contains no new commitments. "By October 1, 1998, Consumers
Energy will provide NRC with a status of our progress in completing
all actions identified
in the attachments
to this letter.''
* Nathan L. Haskell . Director, Licensing
CC Administrator, Region Ill, USNRC Project Manager, NRR, USNRC NRC Resident Inspector
-Palisades
Attachments
2 
ATTACHMENT
A CONSUMERS
ENERGY COMPANY PALISADES
PLANT DOCKET 50-255 STATUS OF PLANS FOR CORRECTIVE
ACTIONS TO RESOLVE NRC DESIGN INSPECTION
OPEN ITEMS 45 Pages 
* ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS Unresolved
Item 50-255/97-201-01
The team questioned
whether the CCW system design met the vendor-recommended
minimum flow of 2000 gpm for the CCW pumps under all operating
conditions.
The team was concerned
that small differences
in the pump operating
characteristics
could cause significant
differences
in flow through each pump during parallel pump operation
due to the flatness of the pump operating
* curves at low flows. The licensee had no analysis available
to demonstrate
that the CCW pumps met the minimum flow requirements.
During the inspection, the licensee developed
a preliminary
system flow model, which showed that, when all three pumps were started upon receiving
a safety injection
system (SIS) signal, the minimum pump flow was through CCW pump P-52A at 1768 gpm. The licensee received a revised minimum flow requirement
of 1600 gpm from the pump manufacturer.
The team's review of the licensee's
completed
flow model calculation
will be an Inspection
Fol/owup Item 50-255197-201-01.
* Palisades
60 Day Response:
As a result of CCW system balancing, scheduled
for the 1998 refueling
outage, a reanalysis
of minimum predicted
CCW system flow rates will be performed.
This reanalysis
will verify that minimum flow rate requirements
will be met under a worst case scenario with appropriate
pump IST degradation
input. This action will be completed
by September
1, 1998. 10/1/98 Update: Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-01)
was identified
as open. Pump performance
data was obtained during the 98 refueling
outage. The completion
for the reanalysis
has been rescheduled
for August 1, 1999 to accommodate
emerging higher priority analytical
work. Unresolved
Item 50-255/97-201-02
The team verified the heat removal capability
of the CCW heat exchangers
by reviewing
the results of various accident analyses.
The licensee had performed
the following
LOCA analyses:
* EA-D-PAL-93-207-01, "LOCA Containment
Response Analysis With Reduced LPSI Flow Using CONTEMPT El-28 Code," Revision 0, * EA-D-PAL-93-272-03, "LOCA Containment
Response Analysis With Degraded Heat Removal System Using CONTEMPT El-28A Computer Code," Revision 0, *and * EA-GEJ-96-01, "A-PAL-94-324
Containment
Spray System (CSS) Sensitivity
on the Containment
Heat Removal During Recirculation (Post-RAS)," Revision 1. The team verified that the input assumptions
relating to the CCW system for the above analyses were correct. The above LOCA analyses demonstrated
that the heat exchangers
could remove sufficient
heat from containment
following
a LOCA to keep the containment
pressure and 1 
* ----------------------
-----ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS temperature
within the design limits. In each case, the analysis documented
a CCW temperature
exiting the shutdown coolers exceeding
the system design temperature
of 140 degrees Fahrenheit
(140 °F) as stated in FSAR table 9-6 and DBD 1.01, "Component
Cooling Water," Revision 3. The team noted that the licensee accepted the maximum CCW temperature
that resulted from the scenarios
analyzed in EA-D-PAL-207-01
and EA-D-PAL-93-272-03
by Corrective
Action D-PAL-93-272G, based primarily
on an evaluation
of the effects on pipe stress. However, the licensee had not considered
the other negative effects, such as any detrimental
effects from elevated CCW temperature
on pump seals. Also, the licensee had not determined
the maximum possible CCW temperature
under worst case conditions
and had not identified
that a change to the FSAR could be required.
The team reviewed the latest LOCA analysis, EA-GEJ-96-01, and determined
that it documented
a CCW temperature
exiting the shutdown cooling heat exchanger
was 184 °F. The licensee determined
the system was operable under this condition
and issued Condition
Report (CR) C-PAL-97-1363F
to determine
the most limiting CCWtemperature
for any condition
and to evaluate all the effects resulting
from that limiting temperature
on the CCW system. ' It appeared that the requirements
of 10 CFR 50, Appendix B, Criterion
111, "Design Control," were not met in this case in that the design basis for the CCW system, as defined in 10 CFR 50.2, did not encompass
the entire range of bounding temperatures.
The team identified
this item as Unresolved
Item 50-255197-201-02.
Palisades
60 Day Response:
Prior to the Design lnspection;.we
determined
that the CCW system is operable at a predicted
maximum system temperature
of 184°F. The CCW system will be analyzed to confirm the most limiting temperature
for any design basis condition, and to determine
the effects of this temperature
on system components
by October 1; 1998. The FSAR will be updated as appropriate.
The programmatic
design control aspects related to this issue will be addressed
as identified
in Attachment
B, Item 1. 10/1/98 Update: In June of 1998, Engineering
Analysis EA-LOCA-98-01
was performed
to determine
the limiting condition
CCW temperature.
The results show a maximum 180°F CCW temperature
out of the CCW heat exchanger.
The effects of this temperature
on system components
was then evaluated.
It was determined
that the CCW heat exchanger
outlet temperature
indication
range was too narrow and needed to be expanded to meet RG 1.91 requirements.
By December 15, 1998, these temperature
indicators
will be replaced and full compliance
with RG 1.97 requirements
will be achieved.
All other evaluated
CCW system component
peak temperature
ratings fall within the predicted
180°F temperature.
The FSAR was changed to clarify CCW system design temperature
and LOCA maximum temperatures.
The temperature
indicator
range issue (50-255/97201-02)
was identified
as open, and was the subject of a NOTICE OF DEVIATION
(50-255/98003-02), in NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION." Palisades
responded
with additional
information
to the NRC under correspondence
dated June 24, 1998, entitled "RESPONSE
TO NOTICE OF VIOLATION
AND NOTICE OF DEVIATION
FROM INSPECTION
REPORT 50-255/98003." 2 
* ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS Refer to Attachment
B, Item 1 for the programmatic "design control" aspects associated
with this issue. Unresolved
Item 50-255/97-201-03
The team reviewed C-PAL-96-1-63-01, "120 day response to GL 96-06, Assurance
of Equipment
Operability
and Containment
Integrity
during Design Basis Accident Condition," Revision 0, which was the licensee's
response to Nuclear Regulatory
Commission (NRG) Generic Letter 96-06, "Assurance
of Containment
Operability
and Containment
Integrity
During Design-Basis
Accident Conditions," and observed that the licensee took credit for relief valve RV-0939 to protect the CCW piping inside containment
from overpressurization
in the event of a LOCA. RV-0939 was not included in the /ST program. The team questioned
whether RV-0939 performed
a safety function and if it should have been included in the /ST program. The licensee issued CR C-PAL-97-1686
to evaluate this
discrepancy.
10 CFR 50.55a requires /ST in accordance
with ASME Section XI of valves that perform a safety function.
It appeared that the licensee did not fully implement
these requirements
for RV-0939. The team identified
this item as part of Unresolved
Item 50-255197-201-03.
Palisades
60 Day Response:
During the Design Inspection, it was determined
that sufficient
overpressure
protection
is provided for the CCW system without taking credit for relief valve RV-0939, and the CCW system is therefore
operable.
The CCW piping in containment
is not required during an accident and is classified
non-Q, safety related. As a result, the ISl/IST programs have classified
the CCW piping and related components, including
RV-0939, as non-class
and excluded the same from inspection/test
requirements
of the Code. The Palisades
response to GL 96-06 determined
acceptability
of systems by generally
taking credit for 1) steam/gas
service, 2) available
expansion
paths, or 3) relief valves as a means to provide *sufficient
protection
against thermally
induced over pressurization.
In the case of the CCW system, "available
relief valves" serves as the basis for acceptability.
Relief valve operation
is considered
important
but not a safety related function, and therefore, the classification
of the CCW system and its components
such as RV-0939 were not changed. Although RV-0939 is not in the IST program, it, along with RV-2108 and RV-0956, is inspected, maintained
and set point verified via
maintenance
activity PPAC CCS043 on a 10-year interval.
These are essentially
the same as the requirements
of the Code (ASME/ANSI
OM-1987, Part 1 ). Based on this evaluation, no further action is required.
RV-0939 is appropriately
classified, maintained
and tested. Our existing GL 96-06 submittal
is accurate.
3 
* ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS 10/1/98 Update: This response has not changed since the submittal
of our original 60-day inspection
report response.
Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-03)
was identified
as closed. No further actions on this item are planned. Unresolved
Item 50-255/97-201-04
FSAR Section 9.3.2.3 stated that the CCW pipingwithin
containment
was not vulnerable
to failure caused by a high energy line break (HELB) and referred to Deviation
Report (DR) D-PAL-89-061, "Post Accident Operation
of CCW System, 11 dated March 23, 1989, for the evaluation.
This DR referred to Engineering
Analysis (EA) EA GW0-7793-01, "CCW Piping Inside Containment
HELBA," Revision 0. This EA was reviewed by the team, and it concluded
that the CCW piping inside containment
was not affected by HELBs, but did not contain the analysis performed
or a reference
to the analysis.
The EA contained
an outline of the methodology, listed the drawings and walkdowns
used, and referenced
the source of the postulated
HELBs. Palisades
Administrative
Procedure
No. 9.11, "Engineering
Analysis, 11 Revision 9, stated that an EA shall present an argument which substantiates
the conclusion
of the EA. The EA also contained
an error in the identification
of the Systematic
Evaluation
Program (SEP) topic number for evaluation
of the effects of internally
generated
missiles.
The licensee initiated
Engineering
Assistance
Request (EAR) EAR-97-0632
to revise EA-GW0-7793-01.
During the inspection, the licensee issued Revision 1 of EA-GW0-7793-01, which included a discussion
of the walkdown analysis used and corrected
the SEP references.
This revised EA was acceptable
to the team. It appeared that the requirements
of 10 CFR Part 50, Appendix B, Criterion . Ill, "Design Control," regarding
verifying
the adequacy of designs were not adhered to in this case. Also, the requirements
of the licensee's
Administrative
Procedure
9. 11 were not fully met in that EA-GW0-7793-01, Revision 0, did not contain full substantiation
of the conclusion.
The team identified
this item as Unresolved
Item 50-255197-201-04.
Palisades
60 Day Response:
As a remedial action, EA-GW0-7793-01
was revised to provide justification
for its conclusion
and to correct references
to related NRC corresponqence.
The related programmatic
design control and calculation
control aspects will be addressed
as identified
in Attachment
B, Items 1 and 2. 10/1/98 Update: This response has not changed since the submittal
of our original 60-day inspection
report response.
Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-04)
was identified
as closed. No further actions are planned for this item . Refer to Attachment
B, Item 1 for the programmatic "design control" aspects associated
with this issu.e. 4 
ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS Unresolved
Item 50-255/97-201-05
The team reviewed the implementation
of the licensee's
commitment
to NRG Regulatory
Guide (RG) 1.97, "Instrumentation
for Light-Water-Cooled
Nuclear Power Plants To Assess Plant and Environs Conditions
During and Following
an Accident," Revision 3, as described
in FSAR Appendix 7C. The RG stated a range for CCW flow instrumentation
of 0-110 percent. Since
there was no instrument
to directly measure CCW flow, the licensee used a combination
of instruments, including
TE-0912 and TE-0913, which measure shutdown cooling heat exchanger
outlet temperature, to indicate flow. Use of instruments (other than flow indicators)
to monitor for CCW flow was determined
as acceptable
by the NRG (a letter from NRG to Consumers
Power Company, dated July 19, 1988, entitled "Palisades
Plant-Response to Generic Letter 82-33 Conformance
to Regulatory
Guide 1.97 "Instrumentation
for Light-Water-Cooled
Nuclear Power Plants To Assess Plant and Environs Conditions
During and Following
an Accident'').
The required range for these TEs in FSAR Appendix 7C was 0-180 °F. This range did not encompass
the temperature
determined
in EA-GEJ-96-01, "A-PAL-94-324
Containment
Spray System (CSS) Sensitivity
on Containment
Heat Removal During Recirculation (Post-RAS)," Revision 1. This analysis determined
an outlet temperature
of the CCW from the shutdown cooling heat exchanger
of 184 °F. The licensee issued CR C-PAL-97-1363E
to evaluate the process instrumentation
and controls associated
with the CCW system for the effects of the higher temperature
predicted
by the analysis.
The licensee did not appear to meet their commitment
to NRG RG 1.97, "Instrumentation
for Light-Water-Cooled
Nuclear Power Plants To Assess Plant and Environs Conditions
During and Following
an Accident," in that the installed
CCW temperature
indicators
were not capable of monitoring
the full temperature
range expected to be observed in the CCW system. The team identified
this item as part of Unresolved
Item 50-255197-201-05.
Palisades
60 Day Response:
Prior to the Design Inspection, we determined
that the COW system is operable at a predicted
maximum system temperature
of 184°F. The CCW system will be analyzed to confirm the most limiting temperature
for any design basis condition, and the effects of this temperature
on system components.
In response to this specific issue, process instrumentation
and controls associated
with the CCW system will be reviewed to identify the impact of the maximum predicted
temperature.
This action will be completed
by October 1, 1998. 10/1/98 Update: Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-05)
was identified
as closed. This item was also the subject of a NOTICE OF DEVIATION
(50-255/98003-02)
from the same letter. Palisades
responded
with additional
information
to the NRC under correspondence
dated June 24, 1998, entitled "RESPONSE
TO NOTICE OF VIOLATION
AND NOTICE OF DEVIATION
FROM INSPECTION
REPORT 50-255/98003." In summary, the range of the CCW heat exchanger
outlet temperature
indicators
will be changed to meet RG 1.97 requirements
by December 15, 1998. 5 
* ** * ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS Unresolved
Item 50-255/97-201-06
The team identified
a lack of closure verification
testing on SI system check valves that could potentially
result in an overpressure
condition
affecting
the low-pressure
piping on the suction of the HPSI pumps. The minimum flow recirculation
lines associated
with the two HPSI pumps and the two LPSI 'pumps were interconnected
upstream of the air-operated
minimum flow recirculation
isolation
valves. In the event that only one HPSI pump was operating
under post-accident
conditions
with the minimum flow recirculation
isolation
valves closed, back leakage through the minimum flow piping associated
with the idle HPS/ pump could over pressurize
the idle HPS/pump suction piping. Backflow between the HPS/ minimum flow lines should be prevented
by check valves CK-ES3339
or CK-ES3331, and CK-ES3340
or CK-ES3332.
However, EGAD-EP-01, "lnservice
Testing Program-Valve
Test Program," Revision 10 indicated
that closure verification
testing of these check valves was not included in the /ST program. *The team asked the licensee if closure of these check valves was considered
a safety function requiring
/ST. The licensee initiated
CR C-PAL-97-1660
to evaluate the testing requirements
of these check valves. On November 10, 1997, the operability
determination
concluded
that these system check valves had not been subject to closure verification
testing as required, and both HPSI pumps were declared inoperable.
In accordance
with TS Section 3.0.3, 3.3, and 4.0.3, the licensee entered a Limiting Condition
for Operation (LCO) action statement, performed
closure verification
testing of check valves CK-ES3339
and CK-ES3340, and verified the operability
of these valves. The licensee stated that closure verification
testing of these check valves would be added to the /ST program. The team also identified
a lack of closure verification
testing on SI system valves that could potentially
result in a Safety Injection
Tank (SIT) being degraded under post-accident
conditions.
The normally closed SIT vent valves, CV-3051, 3063, 3065, and 3067, could be opened in accordance
with SOP-3, "Safety Injection
and Shutdown Cooling System," Revision 28, to reduce SIT pressure.
SOP-3 did not require the affected SIT to be declared inoperable
when a vent was opened. When a vent valve was opened the SIT pressure boundary (250 psig design pressure)
was exposed to the SIT vent header piping (100 psig design pressure).
SOP-3 did not include d(rections
to isolate an open vent valve in the event of an accident.
EGAD-EP-01, lnservice
Testing Program -Valve Test Program," Revision 10, indicated
that closure verification
testing of these valves was not included in the /ST program. The team asked the licensee if the failure of a valve to close could result in a SIT being degraded under accident conditions, and if closure of these valves was considered
a safety function requiring
/ST testing. The licensee initiated
CR C-PAL-97-1592
to evaluate this item and placed caution tags on the control room switches for vent valves CV-3051, 3063, 3065, and 3067 to prevent the valves from being opened without entering an LCO for the SITs. The licensee also stated that these valves had been opened rarely during plant operation.
1 O CFR 50. 55a requires in-service
inspection
in accordance
with Section XI of the ASME Boiler and Pressure Vessel Code. This code requires testing of valves which perform a safety function.
It appeared that the licensee did not implement
these requirements
with regard to valves CK-ES3339, CK-ES3340, CV-3051, CV-3063, CV-3065, and CV-3067. The team identified
this item as part of Unresolved
Item 50-=255197-201-06.
6 
-------ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS Palisades
60 Day Response:
During the Design Inspection, high pressure safety injection
pump minimum flow recirculation
line check valves CK-ES3339
and CK-ES3340
were tested and the HPSI system was declared operable.
Action to include check valves CK-ES3339
and CK-ES3340
in the IST Program will be completed
by July 15, 1998. lri the interim, the check valves are tested to meet quarterly
testing requirements.
During the Design Inspection, the Safety Injection
Tank (SIT) vent valves CV-3051, CV-3063, CV-3065 and CV-3067 were closed and cautioned
tagged with the tanks declared operable.
Action to revise operating
procedures
to address opening the SIT vent valves will be completed
prior to removal of the caution tags. Prior to March 15, 1998, a representative
sample of check valves, AOVs and MOVs will be reviewed and verified to be incorporated
in the IST program as required.
10/1/98 Update: Check valves CK-ES3339
and CK-ES3340
have been included in the IST Program. Operating
procedures
have been revised to address opening of the SIT vent valves CV-3051, CV-3063, CV-3065 and CV-3067 and caution tags have been removed. A representative
sample of check valves, AOVs and MOVs have been sampled to determine
if they are included in the IST Program as required.
The sampling identified
additional
AOVs and one check valve that required inclusion
into the IST Program. These valves have been incorporated
into the IST Program and have been tested to confirm their safety related function.
In addition, several other actions associated
with the IST Program are underway to enhance databases, review ISi Program bases for IST Program impact, and revise IST Program and bases to enhance purpose, scope and program descriptions.
These actions are projected
to be complete by . May 1, 1999. Presently, Palisades
is in full c_ompliance
with the ISi and IST program requirements.
Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-06)
was identified
as closed. .This item was also the subject of a NOTICE OF VIOLATION
(50-255/98003-03)
from the same letter. Palisades
responded
with additional
information
to the NRC under correspondence
dated June 24, 1998, entitled "RESPONSE
TO NOTICE OF VIOLATION
AND NOTICE OF DEVIATION
FROM INSPECTION
REP.ORT 50-255/98003." Unresolved
Item 50-255/97-201-07
The team reviewed the HVAC system serving the cable spreading
room. The team observed that DR F-CG-91-072
was prepared in May 1991 when it was discovered
that the assumptions
in calculation
EA-FC-573-2, "Calculated
Required Air Flow for Inverter/Charger
Cabinet Cooling Fan," dated October 3, 1982, used an ambient temperature
of 94 °F instead of the correct design basis temperature
of 104 °F. The Safety System Design Confirmation (SSDC) Team that found this discrepancy
recommended
that the EA be updated. Procedure*9.11, "Engineering
Analysis," Revision 9, required all EAs to be revised if analytical
inputs or major assumptions
change. The 7 
ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS licensee aec1dedtiotl6
reVisetfie
EA-; and ffie alscrepaiicy
was recorded in DBD 4.02 (125-V de system) and DBD 4.03 (preferred
ac system). The fans were installed
in 1983 and were not safety related. DR F-CG-91-072
was closed in October 1994, when the decision was made not to revise the calculation.
The licensee stated that specifications
were being developed
for replacing
the inverters
and chargers during the time the discrepancy
was being evaluated
and that this knowledge
contributed
to the decision not to update the EA. The inverters
and chargers were scheduled
to be replaced in the near future by Specification
Change (SC) SC-96-033.
The new equipment
would have internal cooling fans designed for a 104 °F maximum ambient and SC-96-033
would supersede
EA-FC-573-2
upon installation.
The team had no other concerns about the cable spreading
room HVAC system. It appeared that the requirements
of 10 CFR Part 50, Appendix B, Criterion/I, "Quality Assurance
Program," were not followed in this case in that the requirements
of Procedure
9. 11 regarding
revising EAs were not fully implemented.
The team identified
this item as part of Unresolved
Item 50-255197-201-07.
Palisades
60 Day Response:
Prior to the Design Inspection, Design Basis Documents
were revised to address this discrepancy.
Analysis EA-FC-573-2
will be revised or superseded
by December 1, 1998. The calculation
control aspects related to this issue (in this case, the revision of all analyses whenever analytical
inputs or major assumptions
change) will be addressed
by the action described
in Attachment
B, Item 2. 10/1/98 Update: The schedule for resolving
remains as stated above. Per NRG correspondence
dated May 18, 1998, titled "NRG INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-07)
was identified
as closed. This item was also the subject of a NOTICE OF VIOLATION
(50-255/98003-04)
from the same letter. Palisades
responded
with additional
information
to the NRC under correspondence
dated June 24, 1998, entitled "RESPONSE
TO NOTICE OF VIOLATION
AND NOTICE OF DEVIATION
FROM INSPECTION
REPORT 50-255/98003." Unresolved
Item 50-255/97-201-08
The team identified
the following
discrepancies
in SJ system mechanical
calculations:
* EA-DBD-2.01-004, "Electrical
and Mechanical
Failure Analysis for the Low Pressure Safety Injection
System," Revision 0, pages 10 and 25, identified
a situation
in which a Joss of an emergency
diesel generator (EOG) during a large-break
LOCA would result in only one LPSI pump and two LPS/ injection
valves being operable.
The EA stated: "The acceptability
of this situation
could not be verified." The team asked if this statement
was correct. The licensee replied that the statement
was not current, and that the statement
appeared to be based on superseded
calculation
ANF-88-107, "Palisades
Large Break LOCA/ECCS
Analysis With Increased
Radial Peaking," Revision 1. Calculation
ANF-88-107
was superseded
by Seimens calculation
EMF-96-172, "Palisades
Large Break LOCA/ECCS
Analysis," Revision 0. The licensee initiated
Engineering
Assistance
Request (EAR) 97-0635 to revise EA-DBD-2.01-004.
8 
ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS * EA-A-NL-92-185-01, "Worst Case Operating
Conditions
for the LPSllSDC System MOVs," Revision 1, addressed
the most limiting conditions
under which the system motor-operated
valves (MOVs) were required to open and close. This analysis included MOVs M0-3015 and M0-3016. These valves were the isolation
valves installed
in the shutdown cooling inlet . piping from primary coolant system (PCS) loop 2. For all normal operations
-other than shutdown cooling being in service, -the valves were electrically
locked closed. Page 19 of EA-A-NL-92-185-01
stated that the scenario that could produce the most limiting differential
pressure was that these valves would be required to close in the event of a downstream
pipe break. The EA addressed
a potential
12-in. downstream
pipe break and determined
that complete depressurization
and blowdown of the PCS to the hot-leg elevation
would occur before operators
could enter the EOPs and attempt to isolate the break. Therefore, the analysis then established
a maximum flow rate of 4120 gpm through valves M0-3015 and M0-3016, based on a normal system flow rate of 3000 gpm and a calculated
leakage of 1120 gpm through a break of a 1-112-inch
branch line downstream
of the valves. The team asked the licensee to provide the basis of the postulated
1-112-inch
branch line failure, since it did not appear to be consistent
with the postulated
pipe crack used in the internal flooding analysis of the safeguards
areas (EA-C-PAL-95-1526-01, "Internal
Flooding Evaluation
for Plant Areas Outside of Containment," Revision 0). The licensee verified that the flooding analysis break flow was different
and that this difference
would not affect the conclusions
of EA-A-NL-92-185-01.
Assumptions
5.9 and 5.10 of EA-A-NL-92-185-01
stated that the HPS/ and LPSI injection
flows to the loops were approximately
equal under post-accident
conditions.
These assumptions
did not appear consistent
with the flow values calculated
in EA-SDW-95-001, "Generation
of Minimum and Maximum HPSllLPSI
System Performance
Curves Using Pipe-Flo," Revision 2. The team asked the licensee to provide the bases of these values. The licensee stated that the values were not current and verified that the difference
between these values and the current values would not affect the EA results. The licensee initiated
CR C-PAL-97-1670
to resolve the discrepancies
in EA-A-NL-92-185-01.
* EA-E-PAL-93-004E-01, "/ST Check Valve Minimum Flow Rate Requirements
to Support Chapter 14 Events," Revision 0, identified
1601 gpm as the required test flow for the LPS/ injection
check valves. The team observed that this value appeared to be less limiting than the values calculated
in EA-SDW-95-001, "Generation
of Minimum and Maximum HPS/ILPSI
System Performance
Curves Using Pipe-Flo," Revision 2. The licensee initiated
CR C-PAL-97-1603
to address this discrepancy.
The licensee determined
that the LPSI test flow presented
in EA-E-PAL-93-004E-01
was less than the current calculated
requirement.
However, the actual LPSI check valve flow acceptance
criterion
in /ST Procedure
Q0-88, "ESS Check Valve Operability
Test (Cold Shutdown)," Revision 17, was verified to be 1690 gpm, which was greater than the current calculated
requirement.
The licensee stated that the affected documentation
will be corrected.
Administrative
Procedure
9. 11, "Engineering
Analysis," Revision 9, Section 6. 1. 5. c stated that an analysis shall be revised if analytical
inputs changed. In the above instances, engineering
analyses were not updated to reflect analytical
input change. The licensee initiated
C-PAL-97-1636
to evaluate the overall issue of calculation
control. The team identified
this item as part of Unresolved
Item 50-255197-201-08.
9 
ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS Palisades
60 Day Response:
During the Design Inspection, it was determined
that the LPSI check valves are operable since IST acceptance
criteria and actual test flow rates exceeded the minimum required flow rates in analysis EMF-96-72
which had superseded
EA-E-PAL-93-004E-01.
By June 1, 1998, engineering
guideline
EGAD-EP-09
and IST procedure
Q0-8B basis document will be revised to assure that the increased
minimum design flow requirement
is met, and that design bases agree with IST acceptance
criteria.
Remedial actions to revise EA-DBD-2.01-004
to accurately
reflect electrical
system response to events will be completed
by August 15, 1998. EA-A-NL-92-185-01
and EA-SDW-95-001
are bounding analyses which will not be required to be revised or superseded.
Specifically, * the calculation
control process will be revised to allow bounding analyses to remain unchanged
when revisions
to inputs or assumptions
do not affect the analysis conclusions.
The calculation
control aspects related to this issue will be addressed
by the action described
in Attachment
B, Item 2. 10/1/98 Update: Engineering
guideline
EGAD-EP-09, IST procedure
Q0-8B Basis Document, and engineering
analysis EA-DBD-2.01-004
were revised as stated above. Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item
(50-255/97201-08)
was identified
as closed. This item was also the subject of a NOTICE OF VIOLATION
(50-255/98003-04)
from the same letter. Palisades
responded
with additional
information
to the NRC under correspondence
dated June 24, 1998, entitled "RESPONSE
TO NOTICE OF VIOLATION
AND NOTICE OF DEVIATION
FROM INSPECTION
REPORT 50-255/98003." Unresolved
Item 50-255/97-201-09
During an SI system walkdown on October 6, 1997, the team observed scaffolding
installed
adjacent to the SIRWT on the roof of the auxiliary
building.
The team questioned
how the installation
of scaffolding
in the vicinity of safety-related
equipment
was controlled
to prevent damage to the safety-related
equipment
during a seismic event. The licensee provided Procedure
MSM-M-43, "Scaffolding," Revision 2, for the team's review. Section 5. 3 of this procedure
required an engineering
review of scaffolding
installed
in the vicinity of safety related equipment.
However, the licensee determined
that the scaffolding
observed during the walkdown had not received engineering
review in accordance
with the procedure.
The licensee initiated
CR C-PAL-97-1417
to address the scaffolding
installation, and the scaffolding
was removed on October 8, 1997. EA-C-PAL-97-1417A-01, "Operability
Reassessment
of SIRWT Scaffolding," Revision 0, was completed
during the inspection.
Based on a structural
analysis of the maximum loading on the SIRWT due to seismic interaction
with the scaffolding
during a safe shutdown earthquake, this analysis concluded
that the SIRWT was not inoperable
due to this nonconforming
condition.
10 
* * ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS During another SI system walkdown on October 30, 1997, the team observed additional
scaffolding
installed
in the east ESG room adjacent to safety-related
piping. An evaluation
by the licensee determined
that this scaffolding
had not been installed
in accordance
with Procedure
MSM-M-43, "Scaffolding," Revision 2. The licensee initiated
CR C-PAL-97-1585
to address this scaffolding
installation
and, based on a visual inspection, concluded
that this nonconforming
scaffolding
would not render any safety-related
piping or components
inoperable.
The licensee removed the scaffolding.
In addition, the licensee performed
a walkdown of all plant scaffolding
during the inspection
and verified that there were no additional
nonconforming
conditions.
The licensee stated that all scaffolding
erections
would cease until appropriate
personnel
underwent
remedial training.
The team observed the following
three separate conditions
in the west ESG room involving
potential
seismic interactions
with safety-related
equipment.
The team noted that, during a seismic event, unrestrained
items could potentially
damage safety-related
piping and equipment.
The safety-related
piping and equipment
in the west ESG room were required for operation
of the HPSI, LPSI, and containment
spray systems in the event of an accident.
* The team observed an unsecured
operations
storage cabinet located adjacent to safety-related
piping and valves. The team asked the licensee if the condition
was in accordance
with plant procedures.
The licensee initiated
CR C-PAL-97-1587, which determined
that the cabinet was not placed in accordance
with the spacing requirements
of Administrative
Procedure
1.01, "Material
Condition
Standards
and Housekeeping
Responsibilities," Revision 11. The operability
evaluation
concluded
that the nonconforming
condition
did not result in any safety-related
equipment
being inoperable.
The cabinet was laid on its side to eliminate
the toppling concern. The licensee stated that the cabinet would be removed from the area. * The team observed an* unsecured
chainfall
located adjacent to and above the shutdown cooling heat exchangers.
A similar chainfall
in the east ESG room was secured. The team asked the licensee if the condition
was in accordance
with plant procedures.
The licensee determined
that the chainfall
location was not in accordance
with Administrative
Procedure
1.01, and initiated
CR C-PAL 97-1586. The operability
evaluation
concluded
that the nonconforming
condition
did not result in any safety-related
equipment
being inoperable.
The licensee stated that the chainfall
chains would be moved away from the heat exchanger.
* The team observed a ladder in the west ESG room that appeared to be improperly
stored. The ladder was lying on the floor under the installed
ladder rack. The team asked the licensee if the condition
was in accordance
with plant procedures.
The licensee initiated
CR C-PAL-97-1601
and determined
that the ladder location was not in accordance
with the "Palisades
Ladder Control Policy for Operating
Spaces," dated May 14, 1997. The CR concluded
that, although the ladder storage did not meet the ladder control policy, the nonconforming
condition
did not result in any safety-related
equipment
being inoperable.
The licensee stated that the ladder was removed from the area. Procedure
MSM-M-43 required an engineering
review of scaffolding
installed
in the vicinity of safety-related
equipment.
Procedure
1. 01 and the "Palisades
Ladder Control Policy for Operating
Spaces," dated May 14, 1997, contain requirements
for storing items in the vicinity of safety-related
equipment.
In these cases, the licensee did not comply with the procedural
requirements
for activities
affecting
quality as required by 1 O CFR Part 50, Appendix B, Criterion
V, "Instructions, 11 
ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS Procedures, and Drawings." The team identified
this item as Unresolved
Item 50-255197-201-09.
Palisades
60 Day Response:
Remedial actions consisted
of dispositioning
all scaffolding
and unrestrained
items near the SIRW Tank and in the East and West Safeguards
Rooms to assure operability
of safety-related
equipment.*
Subsequently, walkdowns
were conducted
in other areas containing
safety-related
equipment
and no conditions
similar to the scaffolding
conditions
identified
in this open item were observed.
Maintenance
and construction
crews were briefed on the lessons learned pertaining
to scaffolding
erection.
By July 15, 1998, we will revise procedures, provide training and reinforce
management
expectations
as necessary
to maintain compliance
with seismic interaction
requirements
for related equipment.
10/1/98 Update: Specific actions to revise procedures, provide training and reinforce
management
expectations
as necessary
to maintain compliance
with seismic interaction
requirements
for safety-related
equipment
have been completed.
Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003
* (DRS) AND NOTICE OF VIOLATION", this item
(50-255/97201-09)
was identified
as closed. This item was also the subject of NOTICES OF VIOLATION
(50-255/98003-05
and 50-255/98003-06)
from the same letter. Palisades
responded
with additional
information
to the NRG under correspondence
dated June 24, 1998, entitled "RESPONSE
TO NOTICE OF VIOLATION
AND NOTICE OF DEVIATION
FROM INSPECTION
REPORT 50-255/98003." This response is associated
with plans to enhance maintenance
personnel
scaffolding
training, and provide training for Auxiliary
Operators
to recognize
unrestrained
items for prompt identification.
Training will be completed
by March 1, 1999. Unresolved
Item 50-255/97-201-10
During the surrogate
tour, the team obseNed the ends of two vent pipes that connected
the containment
sump to the 590-ft elevation
of the containment.
The team asked the licensee to explain the design of these vent lines. During a review of the vent lines, the licensee determined
that the top of the vents were located inside the containment
at an elevation
of approximately
595-ft. The maximum calculated
post-accident
water elevation
was at elevation
597-ft. The vent pipes did not have screens on their inlets. The licensee also determined
that the two vent lines entered the containment
sump inside the sump screens, creating a potential
path for debris to enter the EGGS pump suction piping under post-accident
conditions.
The licensee initiated
CR C-PAL-97-1571, on October 29, 1997, to evaluate this condition
and determined
that the postulated
type and quantity of debris that could enter the vent pipes under post-accident
conditions
would not prevent the SI and containment
spray systems from performing
their safety function, and that these systems were operable under this condition.
The licensee also installed
Temporary
Modification
TM-97-046, on October 29, 1997, to add screens to the top of the vent pipes during the inspection.
These screens would prevent debris from entering the EGGS pump suctions in the event of an accident.
12 
* ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS It appeared that the requirements
of 10 CFR Part 50, Appendix B, Criterion
Ill, "Design Control," were not met in this instance in that the design basis of the containment
sump to exclude debris from the EGGS pump suction piping was not fully implemented.
The team identified
this item as part of Unresolved
Item 50-255197-201-10.
Palisades
60 Day Response:
As stated above, an operability
determination
concluded
the Engineered
Safeguards
Systems were operable in the as-found condition.
As additional
assurance
for continued
operability, temporary
screens were placed over the vent pipes. These screens will be permanently
installed
in the 1998 refueling
outage. The programmatic
design control aspects related to this issue will be addressed
as identified
in Attachment
B, Item 1. 10/1/98 Update: Containment
sump vent screens were permanently
installed
during the 1998 refueling
outage. Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-10)
was identified
as closed. This item was also the subject of a NOTICE OF VIOLATION
(50-255/98003-0?a)
from the same letter. Palisades
responded
with additional
information
to the NRC under correspondence
dated June 24, 1998, entitled "RESPONSE
TO NOTICE OF VIOLATION
AND.NOTICE
OF DEVIATION
FROM INSPECTION
REPORT 50-255/98003." As part of our annual design basis document update projected
for June 1999, the Containment
Spray Design Basis Document DBD-2.03 will be revised to address issues vital to the function of the Engineering
Safety Features following
a LOCA. Refer to Attachment
B, Item 1 for the programmatic "design control" aspects associated
with this issue. Inspection
Followup Item 50-255/97-201-11
The team also observed several piping penetrations
between the east and west ESG rooms which included rubber piping expansion
joints used as penetration
seals. The team questioned
the design of these piping penetration
seals. The licensee stated that the engineering
analyses that demonstrated
that these penetrations
met the design basis did not-specifically
address the use of rubber piping expansion
joints in the penetration
seals. The team reviewed EA-RJC-92-0508, * ''Analysis
of the Effect of a Fire on the Fire Barrier Penetration
Seal Number FZ-0508," Revision 0, and verified that the rubber piping expansion
joints were not addressed.
The licensee initiated
CR C-PAL-97-1627
and determined
that the failure to specifically
justify the presence of rubber expansion
joints did not invalidate
the conclusions
of the original engineering
analyses and that the penetration
seals were adequate.
The licensee also stated that the affected documentation
would be corrected, and that an "extent of condition" review would be performed.
The team identified
this item as Inspection
Fo/lowup Item 50-255197-201-11.
13 
* * ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS Palisades
60 Day Response:
An operability
determination
during the Design Inspection
concluded
that the safety function provided by the fire barriers separating
the East and West Safeguards
Rooms is not affected by the use of rubber expansion
pipe joints. By August 1, 1998, we will revise the design basis engineering
analysis to formally justify the installed
rubber expansion
pipe joints, and perform an investigation
of other area fire barriers for potential
unanalyzed
designs. 1011198 Update: Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-11)
was identified
as closed. The revision to the design basis engineering
analysis for rubber expansion
pipe joints is complete along with investigations
for other fire barriers for potential
unanalyzed
designs. No other unanalyzed
fire barrier design issues were discovered.
No further actions are planned for this inspection
item. Inspection
Followup Item 50-255197-201-12
The team reviewed 10 SI system calculations
and 1 pressurizer
pressure uncertainty
calculation;
these were identified
as "basis documents." Basis Document Rl-38, "SIRW Tank Level Instrument
Calibration," Revision 6, was reviewed for adequacy.
It provided the basis for calibration
of SIRWT level indicators
LT-0332A *and LT-0332B to enable their use to monitor the TS requirement
that the tank contain at least 250, 000 gallons of borated water. Rl-38 used a tank boron concentration
of 1720 parts per million (ppm) and did not consider the range of 1720 to 2500 ppm allowed by TS Section 3.3. Rl-38 was the basis document for the calibration
of the level indicator
that supported
manual actuation
of post-accident
recirculation
operation.
The team was concerned
that the increased
density of the tank water at higher boron concentrations
would increase the instrument
uncertainty.
The calculation
also did not account for variation
in boron concentration
density caused by temperature
changes; an effect which could also affect the total uncertainty.
The licensee recalculated
the total instrument
uncertainty
using the most conservative
boron concentrations
and temperature, and the *resulting
change to the total uncertainty
remained bounded by the original uncertainty
value. Bases Document Rl-69, "Subcooled
Margin Monitor Surveillance," Revision 6, was reviewed for adequacy.
The subcooled
margin monitor (SMM) provided the operator indication
of the PCS margin to .saturation
conditions.
Rl-69 evaluated
possible errors induced in the SMM. The team found that Rl-69 did not account for seismic uncertainty.
This was inconsistent
with RG 1.97 "Instrumentation
for Light-Water-Cooled
Nuclear Power Plants To Assess Plant and Environs Conditions
During and Following
an Accident," May 1983. This RG identifies
subcooled
margin as a Category/, Type A variable, which must continue to read within the required accuracy following, but not necessarily
during, a safe-shutdown
earthquake
event. The team was concerned
that the calculated
error was nonconservative
because it did not consider seismic uncertainty, and could provide misleading
information
to the operators.
The licensee reanalyzed
the potential
error in the SMM, including
seismic uncertainty, and the resulting
total uncertainty
remained bounded by the original uncertainty
value. The licensee assigned Procedure
Change Request (PCR) 5569 to revise Rl-69. 14 
* * ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS EA-RSW-94-001, "F/-0404 Instrumentation
Uncertainty
Calculation," Revision 2, was also reviewed for adequacy.
The analysis established
the recommended
uncertainties
of Fl-0404, which was used in flow testing of the SJ pumps. The instrument
was installed
in 1989, and has been calibrated
five times since then. Drift error was determined
using historical
calibration
data. For the first 4 years, the instrument
was calibrated
once a year. The team found that 24 months had transpired
between the fourth and fifth calibrations.
The licensee stated that the interval was
in 1993 from 11 months to 24 months. The team asked if the drift analysis was revised to account for this change in the calibration
interval.
The* team was concerned
that increasing
the calibration
interval to 24 months would increase the drift error and consequently
increase the total uncertainty
of the instrument.
The licensee reanalyzed
the Fl-0404 uncertainty
using appropriate
drift performance
data for the longer calibration
interval, and the resulting
change to the total uncertainty
remained bounded by the original uncertainty
value. The licensee issued EAR-97-0658
to revise EA-RSW-94-001.
The team also reviewed Basis Document Rl-15A, "Safety Injection
Tank Pressure Channel Calibration," Revision 7, for adequacy.
Rl-15A formed the bases for the pressure channel setpoints
for PIA-0363, 0367, 0369, and 0371, which defined low-and high-pressure
alarms for the S/Ts. The /ow-pressure
alarms warned the operators
of decreasing
nitrogen pressure in the tanks. The channel alarms were set to annunciate
earlier than the pressure limits of TS Section 3.3. 1 (b) so appropriate
action could be taken before pressure reached the setpoints
of pressure switches PS-03408, 03448, 03738, and 30508, which were set to alarm at the TS limits. The team was concerned
that Rl-15A did not consider uncertainties
such as stability
and temperature
effects and that the current total uncertainty
was not adequate.
Considering
the low alarm point of 207 psig, the calculated
uncertainty
allowance
of +/-6.85 psig could result in an alarm at close to 200 psig, which was the TS limit. If additional
uncertainties
were added, the channel pressure switches could alarm after the TS pressure switches.
The licensee reanalyzed
the setpoint for P/A-0363, 0367, 0369, and 0371 using additional
appropriate
uncertainty
inputs and determined
that the resulting
instrument
uncertainty
was bounded by Rl-15A. The team observed that the results of these basis documents
were determined
to encompass
specific additional
uncertainties
due to the assumed margins used in the documents
to account for unquantified
effects. The licensee had a guide entitled "Design & Maintenance
Guide on Instrument
Setpoint Methodology," EGAD-PROJ-16, Revision 0, and the team concluded
that it provided a satisfactory
methodology
for setpoint calculations
and was consistent
with industry standard S67-04, Part I, "Setpoints
for Nuclear Safety-Related
Instrumentation." The licensee stated that EGAD-PROJ-16
provided identical
guidance as EGAD-PROJ-08, Revision 0, of the same title, which was the current designation
of the guide. The instruments
that were re-analyzed
during the inspection
used the guidance of EGAD-PROJ-08.
This methodology
affirmed that margins remained bounded. The licensee stated that use of this guide was not required by plant procedures.
However, the licensee has previously
recognized
from past assessments
that its basis documents
were not as rigorous as required by the current /SA standards.
The licensee stated that EGAD-PROJ-08
was being revised and that the appropriate
procedures
would be revised to require its use. The team identified
this item as Inspection
Fol/owup Item 50-255197-201-12.
Palisades
60 Day Response:
None of the above calculational deficiencies
identified
during the Design Inspection
affected the operability
of any safety-related
equipment.
During the inspection, EGAD-ELEC-08
Rev 1 was approved and issued to provide
for instrument
setpoint methodology.
Our engineering
staff 15 
* ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS has been briefed as to the need to utilize this guidance.
Plant procedures
will be revised by August 15, 1998, to incorporate
EGAD-ELEC-08
for use when setpoint calculations
are required.
10/1/98 Update: Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-12)
was identified
as closed. Applicable
plant administrative
procedures
have been changed to reference
guidance document EGAD-ELEC-08
for use when performing
setpoint calculations, and enhanced to more clearly . describe the applicability
of EGAD-ELEC-08.
No further actions are planned
for this inspection
item. Unresolved
Item 50-255/97-201-13
During a walkdown of the SI system, the team observed that transmitters
for containment
spray flow, FT-0301 and FT-0302, and shutdown cooling heat exchanger
flow, FT-0306, were properly mounted below their flow elements, but the process tubing was observed to be inadequately
sloped back to the transmitters.
Additionally, a walkdown performed
by the licensee at the team's request during an * in-containment
inspection
revealed that the process lines to the HPSI cold-leg flow transmitters
FT-0308, FT-0310, FT-0312, and FT-0313, and the LPSI flow transmitters, FT-0307, FT-0309, FT-0311, and FT-0314, were also installed
with inadequate
slope. The team was concerned
that inadequate
slope in instrument
tubing could contribute
to significant
instrument
uncertainty
by entraining
unequal amounts of air in either leg of the transmitter, causing erroneous
readings.
This was shown to be a valid concern when an operator observed an erroneous
reading in the left channel containment
spray loop indicator, Fl-0301A.
The "below zero" reading was caused by air trapped in one of the process iines. The licensee issued CR C-PAL-97-1561
to vent the line. The lack of tubing slope was inconsistent
with original plant installation
specification
J-F020, Revision 0. This specification
stated: "Flow instruments (differential
tyP.e) in liquid and condensable
vapor service shall preferably
be mounted below the main line connection
so that the impulse lines will slope down to the instrument." The specification
also stated: "Impulse lines to flow instruments
shall slope (up or down) a minimum of one inch per foot." Plant drawings J-F133, Revision 1; * J-F134, Revision O; J-F140, Revision O; and J-F141, Revision 0, depict various acceptable
installation
configurations
for a differential
transmitter.
The current installations
of the flow instruments
identified
above were not consistent
with these drawings.
A later specification, J-465 (Q), "The Technical
Specification
for Installation
of Instrumentation
For Nuclear Service for CPCo Palisades," Revision 0, dated 1981 stated: "The installation
shall be neat in appearance, properly supported, and shall provide for proper slope for adequate drainage or venting of the instrument
lines." This specification
has since been incorporated
into specification
20557-J-59 (Q) under the same title, which requires that a "horizontal
tubing run is continually
sloped in accordance
with design drawings." The licensee issued CR C-PAL-97-1561
to evaluate these instrument
tubing sloping discrepancies.
According
to the operability
determination
of the CR, the instruments
have never shown any adverse effects of trapped air during the last 20 years of operation.
The HPSI and LPSI flow transmitters
were mounted as much as 8 ft above their flow elements.
To accommodate
instruments
mounted above flow elements, specification
J-F020 stated: "5 foot minimum "drop legs (equivalent
of a loop seal)" may be required before the tubing is sloped up the I 16 
* ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS meter." Plant drawings J-F152, Revision 1, and J-F153, Revision 0, depict these mounting configurations.
The licensee stated that the bottom and side tap locations
for the tubing would tend to limit the amount of air getting into the transmitters
and that air entrainment
would be minimal due to the ratio of the volume of the HPSI and LPSI pump suction piping to the tubing volume. EA-C-PAL-95-0877D, "Evaluation
of the Potential
for Excessive
Air Entrainment
Caused by Vortexing
SIRWT During a LOCA," Revision 0, evaluated
the potential
for excessive
air entrainment
in the lines of the pumps caused by vortexing
in the SIRWT during a LOCA, and determined
that the air f]ntrainment
would be a small percentage
of the flow volume. The licensee also stated that technicians
are required to vent the transmitters
during every 18 month surveillance.
However, the team was concerned
that, since the transmitters
sense low static pressure during normal standby operation, air may accumulate
between calibration
intervals
and between system tests. Additionally, the water circulated
through the SI lines from the containment
sump could contain significant
amounts of dissolved
gasses, which could enter the tubing up to the flow transmitters.
The team was concerned
that the effect of air entrapped
in the instrument
tubing could cause large and unquantifiable
errors in the flow indications.
EOP Supplement
4, "Loss of Coolant Accident Recovery Safety Function Status Check Sheet," contained
curves presenting
total SI flow ranges intended to help ensure that the minimum values utilized in the accident analyses (LOCA, MSLB, Steam Generator
Tube Rupture (SGTR)) were met. There was also a minimum total flow criterion
for the operators
to meet, which ensured the containment
sump check valves remained in a stable condition
in EOP-4. 0, "Loss of Coolant Accident Recovery," Revision 9. The operators
would use the HPSI and LPSI flow indication
from FT-0308, 0310, 0312, 0313, 0307, 0309, 0311, and 0314 to compare SI system performance
against the EOP requirements.
The team was concerned
that the potentially
large errors could confuse the operator and impair decision making. The licensee stated that the opetators
are trained to use all available
indications
and that alternate/additional
instrumentation
could be used to confirm trending of PCS conditions
such as that for pressurizer
level, subcooling
margin, reactor vessel level, and charging pump flows. The licensee issued EAR-97-0699
to evaluate this item. It appeared that the design basis for instrument
tubing installation
was not implemented
in the plant installation
as required by 10 CFR Part 50, Appendix B, Criterion
Ill, "Design Control." The team identified
this item as Unresolved
Item 50-255197-0201-13.
Palisades
60 Day Response:
During the Design Inspection, an operability
determination
was made concluding
the HPSI and LPSI flow indication
is operable based on plant operating
experience.
Since the inspection, a plant walkdown was conducted
which revealed that the HPSI and LPSI tubing configuration
met design requirements
but did not conform to associated
design drawings.
The existing tubing configurations
*were observed, and the tubing was determined
not to be susceptible
to air entrainment.
The * conclusions
reached from this walkdown review further justify the reliability
of the HPSI and LPSI flow indication, although configuration
discrepancies
exist. By August 15, 1998, we will resolve the HPSl/LPSI
flow indication
tubing discrepancies
and compare our design requirements
to additional
samples of safety related instrument
tubing to identify any additional
nonconformances
with design criteria.
The programmatic
design control aspects related to this issue will be addressed
as identified
in Attachment
8, Item 1. 17 
* * * ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS 10/1/98 Update: Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item
(50-255/97201-13)
was identified
as closed. This item was also the subject of a NOTICE OF VIOLATION
(50-255/98003-07b)
from the same letter. Palisades
responded
with additional
information
to the NRC under correspondence
dated June 24, 1998, entitled "RESPONSE
TO NOTICE OF VIOLATION
AND NOTICE OF DEVIATION
FROM INSPECTION
REPORT 50-255/98003." Subsequent
to the Design Inspection, Palisades
walked down these installations
during the 98 refueling
outage and confirmed
that the sensing lines for HPSI and LPSI flow transmitters
FT-0308, 0310, 0312, 0313, 0307, 0309, 0311 , and 0314 are appropriately
sloped -thus no deviations
from design requirements
exist. A sampling of other sensing lines associated
with safety-related
equipment
were also walked down and confirmed
to meet design requirements
for sensing line slope. NRC correspondence
dated August 3, 1998 rescinded
this cited potential
violation.
No further actions are planned for this inspection
item. Unresolved
Item 50-255/97-201-14
The team reviewed EA-ELEC-LDTAB-005, "Emergency
Diesel Generator
1-1 & 1-2 Steady State Loading," Revision 4, and verified that the analysis was consistent
with the design basis information
in the FSAR. All required accident loads for a LOCA and a LOOP were identified
and tabulated.
The electrical
loads exceeded the continuous
rating of the EOG during the first 32 minutes of operation
but were below the EOG maximum 2-hour rating. One of the inputs to this analysis was the electrical
toad estimate for LPSI pumps P-67 A and P-678. These electrical
load estimates
were based on the minimum hydraulic
LPS/ pump performance
used in EA-A-PAL-92-037, "Emergency
Diesel Generator
Loadings-First
Two.Hours," Revision 1, which determined
that LPSI pump flow would be* 3600 gpm. Although the LPS/ pump flow was conservative
for evaluating
LOCA mitigation, it was not conservative
for determining
the maximum load the EOG could experience
during a LOCA. The team determined
that the LPS/ pumps could pump 4500 gpm with one LPS/ pump discharging
into all four injection
loops as identified
in EA-SDW-95-001, "Generation
of Minimum and Maximum HPSllLPSI
System Performance
Curves Using Pipe-Flo," Revision
2. The team was concerned
that the licensee had not analyzed for the worst-case
electrical
load demand on the EDGs. Preliminary
evaluations
by the_ licensee using the correct maximum loads indicated
that the electrical
loading on one EOG could be higher than that determined
in EA-ELEC-LDTAB-005.
The licensee issued CR C-. PAL-97-1650
to review and correct all necessary
electrical
analyses and determined
the EDGs to be operable.
The team reviewed EA-ELEC-VOL
T-13, "Palisades
Loss of Coolant Accident With Off$ite Power Available," Revision 0, which evaluated
the ac voltage available
during normal operating, refueling, and accident conditions.
The team noted that the calculation
had not been revised since 1993 and . that the load magnitudes
identified
in EA-ELEC-LDTAB-005, Revision 4, and EA-SDW-95-001, Revision 2, had not been included.
The licensee reviewed the impact of the revised loads on EA-ELEC-VOL
T-13 and determined
that the changes had minimal effect on the analysis.
The team also noted that FSAR Section 8.3 stated that backfeeding
via the main and station power transformers
could be utilized;
however, EA-ELEC-VOL
T-13 had not analyzed this particular
operating
mode. The licensee stated that it had recognized
that an analysis for backfeeding
needed to be performed
in 1994 and had issued AIR A-PAL-94-223
to create an analysis in order to bound 18 
* * ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS this condition
of operation.
The licensee initiated
C-PAL-97-1619
to review and update EA-ELEC-VOLT-13
for load changes. It appeared that the requirements
of10 CFR Part 50, Appendix B, Criterion
Ill, "Design Control," had not been met for EA-ELEC-LDTAB-005
an*d EA-ELEC-VOLT-13
in that the design basis had not been updated to document the actual plant parameters.
The team identified
this item as part of Unresolved
Item 50-255197-201-14.
Palisades
60 Day Response:
During the Design Inspection, an operability
determination
was made which concluded, based on an evaluation
which bounded recent load changes, that the electrical
system is operable.
Mechanical
flow model analyses, which serve as input to the electrical
load flow analyses, will be completed
by December 15, 1998. The electrical
load flow analyses, which will assure plant loads are accounted
for and applicable
operating
scenarios
are addressed, will be completed
by August 15, 1999. A specific backfeed analysis will be completed
by Januar}t 15, 1999. The programmatic
design control aspects related to this issue will be addressed
as identified
in Attachment
8, Item 1. 1011/98 Update: Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", th.is item (50-255/97201-14)
was identified
as closed. The mechanical
and subsequent
electrical
flow model analyses are on target for completion
by December 15, 1998 and August 15, 1999, respectively, as stated above. Backfeed analysis EA-ELEC-FL
T-009, "GSU Short Circuit Analysis" was completed
with design attributes
captured in the applicable
Design Basis Document.
Refer to Attachment
8, Item 1 for the programmatic "design control" aspects associated
with this issue. * Inspection
Followup Item 50-255/97-201-15
FSAR Section 8.5.2 stated that cables would be sized in accordance
with the National Electric Code (NEC) or Insulated
Power Cable Engineers
Association
(/PCEAllCEA)
ampacity values and the cable ampacities
would be adjusted on the basis of actual field conditions
when possible.
The adjustments
included conductor
operating
temperature, ambient temperature, cable overall diameter, raceway fill, and fire stops. The licensee had recently initiated
a program to verify the adequacy of its cable ampacity sizing. EA-ELEC-AMP-032, "Ampacity
Evaluation
for Open Air Cable Trays With a Percent Fill Greater Than 30% of the Usable Cross Sectional
Area," Revision 1, was issued in 1997 to address cable sizing. While reviewing
the EA, the team noted the absence of fire stop derating and increased
cable temperatures
due to thermal radiation
from hot pipes. The licensee had initiated
AIR A-PAL-97-062
to evaluate the effects of local heat sources on fire stops; however, evaluation
of the effects on cable degradation
due to the close proximity
of hot piping systems had not been included.
The licensee stated that evaluation
of the effects of hot piping would be included under A-PAL-97-062.
The team identified
this item as Inspection
Followup Item 50-255197-201-15 . 19 
* * ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS Palisades
60 Day Response:
We will complete our Cable Ampacity Sizing Program by September
15, 1998 which will identify any cable degradation
due to the close proximity
of hot piping, and any degradation
of fire stops due to local heat sources. 10/1/98 Update: Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-15)
was identified
as open. Cable * degradation
due to the close proximity
of hot piping, and any degradation
of fire stops due to local heat sources has been evaluated.
Results confirm that the cable design is acceptable.
No further actions are planned for this inspection
item. Unresolved
Item 50-255/97-201-16
The 120-V ac safety-related
and non-safety-related
loads were powered from instrument
ac bus Y-01. Bus Y-01 was powered from either motor control center (MCC) 1or2 via automatic
transfer switch Y-50. MCCs 1 and 2 were redundant
safety-related
busses. The licensee stated in a January 24, 1978, letter to the NRG that it would. implement
the recommendation
of RG 1. 6 in that no . provision
would exist for automatically
transferring
loads between redundant
power sources. The NRG issued a safety evaluation
report, dated April 7, 1978, confirming
the licensee's
commitment.
FC-364, "Feeder Change for Instrument
Bus Y-01," Revision 0, implemented
this commitment
and powered bus Y-01 from MCC 1 and non-safety-related
MCC 3. However, FC-854, "Y-01 Power Supply Feed Modification," Re.vision
0, moved the backup power source from MCC 3 to the safety-related
MCC 2, and resulted in a departure
from the plant's licensing
basis. The modification
installed
fuses in series with the existing breakers, which provided an additional
level of protection
for the two safety-related
busses. The team observed that the safety evaluation
performed
for FC-854 did not identify that prior NRC approval was required.
The licensee issued CR C-PAL-97-1678
to document this deviation
from the licensing
basis. It appeared that this modification
was a USO in that the possibility
of a common-mode
failure of the redundant
safety-related
busses was created, which was not previously
evaluated
in the FSAR and, thus, the criterion
of 10 CFR 50.59(a)(2)(ii)
was satisfied.
The team identified
this item as Unresolved
Item 50-255197-201-16.
Palisades
60 Day Response:
During the Design Inspection, an operability
determination
was completed
which concluded
that the implemented
design meets the intent of RG 1.6 and provides a single failure proof method of preventir:ig
the transfer of a fault between redundant
load sources. The current configuration
was implemented
under FC-854 with the modification
safety evaluation
concluding
that an unreviewed
safety question does not exist. Prior NRC approval of the change was not required.
A description
of the implemented
modification
was transmitted
to the NRC in our Annual Report of Facility Changes, Tests and Experiments
dated April 2, 1991. This 1989 modification
resulted in a change to a prior NRC commitment.
In accordance
with NEI guidelines, we will submit by November 1, 1998, a revised commitment
which reflects the existing plant configuration
and governing
design basis. 20 
* * ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN iNSPECTION
OPEN ITEMS 10/1/98 Update: Per NRG correspondence
dated May 18, 1998, titled "NRG INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item
(50-255/97201-16)
was identified
as closed. This item was also the subject of a NOTICE OF DEVIATION
(50-255/98003-08)
from the same letter. Palisades
responded
with additional
information
to the NRG \ . .mder correspondence
dated June 24, 1998, entitled "RESPONSE
TO NOTICE OF VIOLATION
AND NOTICE OF DEVIATION
FROM INSPECTION
REPORT 50-255/98003." In summary, Palisades
concludes
that our commitment
to assure that redundant
safety related power sources cannot be both affected by a fault on the instrument
bus has been maintained.
NRG correspondence
dated August 3, 1998 concluded
that a USQ does not exist, and that Consumers
appropriately
notified the NRG of past design changes, and rescinded
this cited potential
deviation.
No further actions are planned for this inspection
item. Inspection
Followup Item 50-255/97-201-17
The team observed that no system analysis existed to show that all the Class 1 E 120-V ac loads had *adequate
voltages.
The licensee demonstrated
during the inspection
that adequate voltages did exist for selected loads. For example, EA-ELEC-VOLT-24, "Voltage Drop From Preferred
AC Power Source Y10 Breaker 2 and Y40 Breaker 2 Out to the 5U12 Relays," Revision 0, showed that adequate ac voltage for those selected components
was available
at the minimum.inverter
voltage. The licensee initiated
CR C-PAL-97-1621
to evaluate and resolve this concern. The team identified
this item as part of Inspection
Fol/owup Item 50-255197-201-17.
Palisades
60 Day Response:.
During the Design Inspection, an operability
determination
was made concluding
the Class 1 E 120 V * ac loads are operable based on past plant operating
experience
and the expected minimal change in supplied voltage between normal and accident plant conditions.
By August 15, 1998,. we will perform a bounding analysis to confirm that Class 1 E 120 V ac loads have adequate voltage during accident conditions.
10/1/98 Update: Per NRG correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item
(50-255/97201-17)
is identified
as open. A bounding calculation
was performed
under EA-C-PAL-97-1621A-01
that developed
worst case voltage levels for the Preferred
AC System and confirmed
adequate available
voltage during accident conditions . These analysis results will be incorporated
into Design Basis Document DBD-4.03, "Preferred
AC System" and tracked under change request number 4.03-12-R3-0728.
No further actions are planned for this inspection
item . 21 
* * ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS Unresolved
Item 50-255/97-201-18
The team reviewed relay settings for protective
relays associated
with LPSI pump P-67 A, HPSI pump P-66A, SW pump P-7A, CCW pump P-52A, EOG 1-1 differential
protection, bus 1C undervoltage
protection, and Bus 1 C second-level
undervoltage
protection.
The settings were consistent
with the design parameters
of the devices being protected.
However, during the review, the licensee determined
that the overcurrent
relays for supply breakers 152-105 and 152-106 to bus 1C had not been calibration
tested during the last refueling
outage (1995) as required by Periodic and Predetermined
Activity (PPAC) SPS025, "Bus 1 C Relay Testing." The licensee stated that these relays would be calibrated
during the 1998 refueling
outage. The licensee reviewed past calibration
data for this type of relay and determined
that negligible
drift had previously
been documented.
The licensee initiated
CR C-PAL-97-1568
to resolve this discrepancy.
It appeared that the requirements
of 10 CFR Part 50, Appendix B, Criterion
XI, "Test Control," had not been implemented
in this case in that certain relays had not been tested as required by the test program. The team identified
this item as Unresolved
Item 50-255197-201-18.
* Palisades
60 Day Response:
During the Design Inspection, an operability
determination
was made concluding
that past calibrations
of overcurrent
relays for breakers 152-105 and 152-106 revealed insignificant
drift and the relays are operable.
We will perform maintenance
activity PPAC SPS025 to calibrate
the overcurrent
relays during the 1998 refueling
outage. Our corrective
action history identified
no other examples of failure to perform scheduled
relay testing. 10/1/98 Update: Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item
(50-255/97201-18)
was identified-as
closed. This item was also the subject of a NOTICE OF VIOLATION
(50-255/98003-09)
from the same letter. Palisades
responded
with additional
information
to the NRC under correspondence
dated June 24, 1998, entitled "RESPONSE
TO NOTICE OF VIOLATION
AND NOTICE OF DEVIATION
FROM INSPECTION
REPORT 50-255/98003." In summary, the overcurrent
relays for breakers 152-105 and 152-106 will be tested/calibrated
by December 31, 1998. The requirements
for PPAC SPS025 have been revised to allow performance
of the testing and calibration
while the plant is at power operation.
Unresolved
Item 50-255/97-201-19
The team questioned
the replacement
schedule for Agastat E7000 series relays. The team was aware that the manufacturer, in correspondence
to other utilities, had recommended
a 10-year replacement
schedule for these relays. The licensee stated that 52 E7000 series relays were installed
and that 7000 series Agastats were also installed
in Class 1 E applications.
Some circuits containing
7000 series relays included the 2400-V bus 1C and*1D supply breakers, time delay relays associated
with charging pumps. P-55A, B, and C, and auto transfer failure alarms for 2400-V busses 1C and 10. The manufacturer's
stated qualified
life forthe E7000 relays was 10 years. The licensee stated that the
qualified
life applied if the relays were located in a harsh environment
and, 22 
* * * ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS since the E7000 relays were located in a mild environment, no qualified
life determination
was required.
Based upon this justification, the licensee issued PPAC Deletion Form MSE 034, dated March 3, 1995, which stated that the relays would not require replacement
at 10-year intervals.
The team believed that the qualified
life stated by the manufacturer
applied to any environment.
The team verified with the manufacturer
that the projected
qualified
life of 10 years was the operating
life of the E7000 series relay as long as the device did not exceed the equipment
ratings, and that the life of 10 years was applicable
to either a mild or harsh environment.
The licensee had not evaluated
the qualified
life ofthe 7000 series relays. The manufacturer
of Agastat relays issued a 10 CFR Part 21 notification
concerning
the inability
of the E7000 series relays to switch a 1-amp load at rated voltage. The licensee evaluated
the installed
E7000 series relays and identified
no concerns.
The team observed that this evaluation
did not review those 7000 series relays dedicated
by the licensee to safety-related
use. The licensee issued CR C-PAL-97-1663
to resolve the issues concerning
Agastat relays and determined
that all the relays were operable.
It appeared that the requirements
of 10 CFR Part 50, Appendix B, Criterion
Ill, "Design Control," had not been met in this instance in that the design basis lifetime for Agastat relays as stated by the manufacturer
had not been correctly
implemented
in the facility.
The team identified
this item as Unresolved
Item 50-255197-201-19.
Palisades
60 Day Response:
During the Design Inspection, an operability
determination
was made concluding
that the 7000 series relays are operable based on their similarity
in application
and design to E7000 relays. By July 15, 1998, we will complete our analysis of both 7000 and E7000 series relays dedicated
for safety related use to confirm their ability to perform safety-related
functions
during their installed
life and their conformance
with applicable
design requirements.
10/1/98 Update: Per NRG correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-19)
was identified
as open. A review of both 7000 and E7000 relay age-sensitive
components
was performed
that indicates
that all relay materials
will last for greater than 40 years without significant
degradation
when installed
in mild environments.
Based on this review, a 10 year replacement
interval is not justified
and the relays can be expected to perform their design function for greater than 40 years. No further
actions are planned for this inspection
item. Unresolved
Item 50-255/97-201-20
The 125-V de system was divided into two independent
systems. Each system consisted
of a battery, switchgear, distribution
panel, and two chargers.
Station battery 1, battery charger 1, and battery charger 3 supplied 125-V de bus 1. Battery charger 1 was supplied from MCC 1 and battery charger 3 was supplied from MCC 2. Administrative
controls limited the operation
so that only one charger per battery was in service. This prevented
a common-mode
failure from affecting
both * 23 
ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS emergency
busses. The supply to 125-V de bus 2 was similar, with battery charger 2 fed from MCC 2 and battery charger 4 fed from MCC 1. Operating
Procedure
SOP-30, "Station Power," Revision 20, required the battery chargers to be operated in pairs (1 and 2 or 3 and 4). The licensee stated that the battery chargers were swapped monthly to provide equal operating
time for each battery charger. During swapping of the battery chargers in accordance
with Section 7. 7. 2 of SOP-30, the 125-V de breaker on the in-service
battery charger was opened and then the 125-V de breaker for the battery charger to be placed in service was closed. During this evolution, both battery chargers were disconnected
from the station battery and 125-V de switchgear
bus. Although temporary
disconnecting
the battery charger from the de bus had minimal safety impact on the plant, the team observed that TS 3. 7. 1 h required two station batteries
and the de systems (including
at least one battery charger on each bus) to be operable when the PCS was above 300 °F. The licensee stated that an LCO was not entered when no battery chargers were connected
to the de busses. The licensee initiated
CR C-PAL-97-1537
to resolve this discrepancy.
The team identified
the licensee's
failure to enter an LCO during battery charger switching
evolution
as Unresolved
Item 50-255197-201-20.
Palisades
60 Day Response:
Prior to the Design Inspection, we concluded
that our design bases were met and an LCO would not entered when realigning
battery chargers.
This conclusion
was based on no appreciable
battery discharge
occurring
during the short realignment
period when neither
charger was connected
to the 125 Vdc bus. In response to this Design Inspection
item, however, operating
procedure
SOP-30 was revised in anticipation
of an amendment
approving
our December 27, 1995 technical
specifications
change request. Although the requested
change does not require a connected
charger, the change defines 125 Vdc bus operability
in terms of applied bus voltage. SOP-30 now requires entry into an LCO whenever performing
charger realignment.
On January 26, 1998, a technical
specification
change request was resubmitted
as part of the Improved Technical
Specifications
Program. An amendment
in response to this latest change request will resolve this open item. 10/1/98 Update: Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-20)
was identified
as closed. In July 1998, Amendment
180 of the Palisades
Electrical
Technical
Specifications
was implemented
that clarifies
the 125 Vdc system operational
requirements.
With the issuance and implementation
of Amendment
180, no further actions are planned for this inspection
item. Inspection
Followi.Jp
Item 50-255197-201-21
The team reviewed the 125-V de battery loading during the normal and alternate
battery charger alignment.
During the normal battery charger alignment, battery charger 1 was powered from EOG 1-1 and battery charger 2 was powered from EOG 1-2. During a LOCA combined with a LOOP in this normal alignment, the batteries
would be without ac power for approximately
1 O seconds until the EDGs restored power. The team reviewed EA-ELEC-LDTAB-009, "Battery Sizing for the Palisades
Class 1 E Station Batteries
ED-01 and ED-02," Revision 2, which verified that the battery was sized to provide adequate power during the 10 second interval until the EDGs provided ac power to battery 24 
* * ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGNINSPECTION
OPEN ITEMS chargers 1 and 2. During the alternate
battery charger alignment
with battery charger 3 powered from EOG 1-2 and battery charger 4 powered from EOG 1-1, the station batteries
would be required to carry the de loads for more than 10 seconds in the event of a LOCA combined with a LOOP and a single failure of ac power. EA-ELEC-LDTAB-009
did not analyze the battery loading for station batteries
ED-01 and ED-02 during this condition.
When questioned
by the team the licensee stated that the de loading during this scenario would be greater than the worst-case
loading assumed in ELEC-LDTAB-009.
The licensee issued CR C-PAL-97-1596
to resolve this discrepancy.
Additionally
the team had concerns on whether the licensee met the single failure criterion
when the alternate
battery charger alignment
was in effect. The team identified
the question with respect to the single failure criterion
and the additional
loading on the battery as an Inspection
Followup Item 50-255197-201-21.
Palisades
60 Day Response:
During the Design Inspection, an operability
determination
was made concluding
that the station batteries
are operable.
Operability
was based on a preliminary
analysis where additional
* conservative
loads were included in the battery load analysis showing that the battery terminal voltage would be greater than the required minimum output of 105 Vdc throughout
the exp.ected
load duration until an operable charger would be connected
to the bus. Operating
procedures
control alternate
charger alignment
but do not restrict this practice which is allowed by technical
specifications.
By January 15; 1999, we will complete a formal analysis of battery loading considering
the battery chargers are in their alternate
alignment, and a combined event of a LOCA, LOOP and single failure of ac power occurs. 10/1/98 Update: Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-21)
was identified
as open. As stated above, by January 15, 1999, the formal battery loading analysis will be completed.
Inspection
Followup Item 50-255/97-201-22
The team identified
that TS Section 4. 7.2c required that each station battery be demonstrated
operable by verifying
that the battery capacity was adequate to supply and maintain in an operable status all of the actual emergency
loads for 2 hours when the battery was subjected
to a battery service test. The battery service tests performed
on station batteries
ED-01 and ED-02 were performed
for a duration of 4 hours. The 4-hour duration and loading was based on the design basis station blackout (SBO) coping time. The team noted that the 2-hour requirement
of TS 4. 7.2c was non-conservative
with respect to the design basis, which required the station batteries
to be available
for4 hours. The design basis duration of 4 hours was included in FSAR Section 8.4.2; DBD 4.01, "Station Batteries," Revision 3; RE-83A, "Service/Modified
Performance
Test-Battery
No. ED-01," Revision 9, and RE-838, "Service/Modified
Performance
Test-Battery
No. ED-02," Revision 9. Testing the batteries
in accordance
with RE-83A and B has ensured that batteries
ED-01and02
have met the 4-hour design basis requirement.
The licensee has submitted
TS changes to correct the non-conservative
TS Section 4. 7.2c and issued CR C-PAL-97-1551
to resolve this discrepancy.
The team identified
this item as Inspection
Followup Item 50-255197-201-22.
25 
* ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS Palisades
60 Day Response:
During the Design Inspection, an operability
determination
was made concluding
that the 4-hour SBO station battery load profile envelops the 2-hour OBA load profile. By January 15, 1999, we will complete a formal analysis of battery loading considering
the battery chargers are in their allowed alignment
configurations
with a combined event of a LOCA, LOOP and.single
failure of ac power. We submitted
a technical
specification
change request on December 27, 1995 to describe the test profile as the design basis profile without stipulating
a specific period for the profile. On January 26, 1998, a technical
specification
change request was resubmitted
as part of the Improved Technical
Specifications
Program which identifies
a four hour load profile for the service test. An amendment
in response to this latest technical
specifications
request will resolve this open item. 10/1/98 Update: Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item
(50-255/97201-22)
was identified
as open. As identified
above, by January 15, 1999, the formal battery loading analysis will be completed.
In July 1998, Amendment
180 of the Palisades
Electrical.Technical
Specifications
was implemented.
Amendment
180 does not specify a duty cycle (profile)
duration in units of time. Therefore, the design basis requirements
found in the FSAR can be used. Inspection
Followup Item 50-255/97-201-23
EA-ELEC-FL
T-005, "Short-Circuit
for the Palisades
Class 1 E Station Batteries
ED-01 and ED-02," . Revision 0, was submitted
to the team as the short-circuit
analysis for the Class 1E 125-V de system. The following
discrepancies
with the assumptions, methodology, and conclusions
were identified:
* Section 4. 4 and 4. 5 assumed various breaker and fuse impedances, which had not been verified against the installed
facility.
* Section 5. 2 utilized the battery charger current limit of 220 amps as the maximum short-circuit
contribution
without supporting
documentation.
* Section 5.2 stated that the open-circuit
voltage was 2.06 V per cell, whereas the EA utilized an open-circuit
voltage of 2. 0 V per cell. * Section 8. 0 stated that the results were to be further reviewed by the licensee;
however, the team found no evidence of this review. Section 8. O also contained
no conclusion
about the de system acceptability.
The licensee issued A/Rs A-PAL-97-108, 109, and 110 to resolve these discrepancies.
The licensee stated that the* analyses would be reviewed and the conclusions
revised. During the 1995 refueling
outage, FES-95-206
replaced existing batteries
ED-01 and ED-02. The team questioned
if the sh9rt-circuit
current provided by the new battery was analyzed and if there were any effects on the de distribution
panel breakers, since the team noted that EA-ELEC-FL
T-005 26 
* * ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS . had not been updated since 1994. The team also noted that the design basis for the evaluation
of fault current contributions
on de circuits was in FSAR Section 8.5.2, which stated "The 125 volt de protection
design considers
the fault current available
at the source side of the feeder protective
device." However, the licensee stated that the short-circuit
contribution
value for de circuits was taken at the electrical
load terminals
and not at the breaker load terminals (de short-circuit
current value would be less when calculated
at the load terminal vice the source side of the feeder protection
device because voltage available
at the load terminal would be less than at the source breaker).
The licensee determined
that the short-circuit
contribution
at 8 breakers (breakers
72-101, 72-105, 72-106, 72-121, 72-127, 72-133, and 72-135) on distribution
panels 011-1and011-2
could exceed the short-circuit
interrupting
ratings when evaluated
in accordance
with the design basis method in the FSAR. Also, when the team questioned
the assumed breaker fault ratings on de busses 010, 020, 011-1, and 011-2of13,000
amps in EA-ELEC-FLT-005, the licensee was unable to show manufacturer
or testing documentation
to support this assumption.
The team believed that this assumption
was inconsistent
with its experience.
The licensee performed
an operability
review and issued CR C-PAL-97-1652
to resolve these discrepancies.
The maximum short-circuit
current of the battery installed
by FES-95-:206, as provided by the manufacturer, was 17094 amps. Calculation
EA-ELEC-FL
T-005 did not reflect this new short:..circuit
current. Upon questioning
by th.e team, the licensee stated that an evaluation
was performed
to ensure that the system short-circuits
were acceptable.
During the team's review of this evaluation
it was determined
that the maximum battery short-circuit
current was not utilized.
The.licensee
stated that the short-circuit
current utilized, 12,821 amps, was provided by the manufacturer
as a more realistic
value than 17,094 amps. However, the licensee could not document a basis for the 12,821 amps and stated that they would verify it with the manufacturer.
The team identified
these discrepancies
concerning
EA-ELEC-FL
T-005 as part of Inspection
Followup Item 50-255197-201-23.
Palisades
60 Day Response:
During the Design Inspection, an operability
determination
was made concluding
that a fault would more likely occur at the load rather than at the breaker terminals.
A fault at the load (esults in a reduced value of fault current which falls within the breaker interrupting
rating. We have since obtained vendor specifications
which envelop our calculated
peak short circuit currents assumed to occur at the breaker terminals.
These specifications
confirm our earlier conclusion
that the breakers are suitable for their intended service, and resolve any concerns with respect to breaker short circuit interrupting
capability.
Revisions
to analysis EA-ELEC-FL
T-005, to correct the plant-identified
deficiencies
described
in the Design Inspection
report, will be complete by January 15, 1999. 10/1/98 Update: Per NRG correspondence
dated May 18, 1998, titled "NRG INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-23)
was identified
as open. Revisions
to analysis EA-ELEC-FL
T-005, to correct the plant-identified
deficiencies
described
in the Design Inspection
report, remains scheduled
for completion
by January 15, 1999 . 27 
* ** * ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS Inspection
Followup Item 50-255197-201-24
FSAR Section 8.4.3.3 stated that the batteries
were designed to furnish their maximum load down to an operating
temperature
of 70 °F without dropping below 105 V de, and that the equipment
supplied by the batteries
was capable of operating
satisfactorily
at this voltage rating. EA-ELEC-VOL
T-026, "Voltage Drop Model of the Palisades
Class 1 E Station Batteries
D01 and D02," Revision 0, evaluated
the de voltages at the distribution
panels based upon a battery voltage of 105 V de, but did not evaluate the voltages that would be available
at the load device terminals.
The team was concerned
that the additional
voltage drop from the distribution
panel to the loads could result in voltages less than the design basis of the loads, and that no analysis was performed
to evaluate this situation.
For example, the deign-basis
minimum input voltage for the inverters
was 105 V de and the licensee could not show any vendor documentation
to support operating
at a value Jess than 105 V de. The team noted that the inverters
could be subjected
to an input voltage of approximately
102 V de if the battery voltage were 105 V de. The licensee stated that battery surveillance
testing has shown that battery voltage, when subjected
to an SBO duty cycle, did not decrease below 108 V de. During the inspection, the licensee evaluated
several safety-related
loads and verified that adequate voltages would exist at 105 V de battery voltage. The licensee issued CR C-PAL-97-1620
to evaluate the lack of an EA to ensure that adequate voltages would exist at the load terminals.
The team identified
this item as part of Inspection
Followup Item 50-255197-201-24.
Palisades
60 Day Response:
During the Design Inspection, an operability
determination
was made concluding
that the 125 Vdc system is operable based on an evaluation
of several safety related loads, in which adequate load voltage was found to exist with a 105 Vdc battery terminal voltage. By November 15, 1998, we will perform a bounding analysis to identify the worst-case
minimum voltage levels at the load assuring that minimum load voltage req.uirements
are met. * 1011198 Update: Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-24)
was identified
as open. As stated above, this issue is scheduled
for completion
by November 15, 1998. Unresolved
Item 50-255197-201-25
The team also questioned
the capability
of solenoid valves to operate at voltages of 87 V de as stated in DBD 1. 01,
Cooling Water System," Revision 4. The licensee determined
that the DBD was incorrectly
worded and that the correct solenoid capability
was 90-140 V de. Upon further review, the licensee identified
that improperly
rated coils, rated 102-126 V de, were installed
in solenoid valves SV-0918 and SV-09778.
The licensee initiated
Engineering
Assistance
Request (EAR) 97-0652 to replace the coils. It appeared that the requirements
of 10 CFR Part 50, Appendix B, Criterion
Ill, "Design Control," were not followed in that the design basis for the solenoid valve coils was not implemented
in the plant. The team identified
this item as Unresolved
Item 50-255197-201-25 . 28 
** ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS Palisades
60 Day Response:
Since the design inspection, further evaluation
identified
that there is no impact on the mitigation
of an accident if solenoid valves SV-0918 and SV-09778 fail to open due to low voltage since the close position is both the failed position and the required safety position.
Based on this review, the design basis is met by the existing solenoid valve installation.
The actions in response to Inspection
Followup Item 50-255/97-201-24
will identify any other minimum voltage problems.
10/1/98 Update: Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item
(50-255/97201-25)
was identified
as closed. No further actions are planned for this inspection
item. Inspection
Followup Item 50-255/97-201-26
The team identified
other discrepancies
in calculations
as follows: * Assumptions
4. 6 and 4. 7 of EA-ELEC-VOL
T-26, Revision 0, and assumptions
4. 8 and 4. 9 of EA-ELEC-M/SC-022, "Electrical
Systems Model of the Palisades
Class 1 E Safety Re/a.fed 125 V de System," Revision 1, assumed various fuse and breaker impedances
which had not been verified against the installed
equipment.
* Section 7. 0 of EA-ELEC-VOL
T-26, Revision 0, "Conclusion," stated that the results were to be further reviewed by the licensee;
however, the team found no indication
that this review had been performed.
The "Conclusion" section also contained
no statement
concerning
the de system acceptability.
* EA-ELEC-VOL
T-26, Revision 0, utilized a correction
factor for battery temperature
of 77 °F instead of the correction
factor for 70 °F, which was the minimum design basis temperature
for the battery. The number utilized is less conservative
and the licensee evaluated
that the overall effect on voltages in the calculation
would be less than 0. 5 percent. * EA-ELEC-LDTAB-029, Revision 2, stated the type of battery constant as 1.0 in Attachment
A and 1.4 on Sheet 4. The constant to be utilized depended on the type of battery. 1. 0 referred to a lead acid battery; 1.4 referred to a nickel-cadmium
battery. The licensee reviewed the EA and determined
that the correct constant was utilized in the EA and that the reference
to 1. 4 was an editorial
error. The licensee issued CR C-PAL-97-1656
to address the battery temperature
correction
factor and stated that the other discrepancies
would be corrected
in future revisions
to the calculations.
The team identified
this item as part of Inspection
Fo//owup Item 50-255197-201-26.
29 
* * ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS Palisades
60 Day Response:
During the Design Inspection, an operability
determination
was made concluding
that the calculation
deficiencies
identified
had no affect on the analyses conclusions;
ie, supplied voltages remain within equipment
ratings and the station batteries
are not affected.
By January 15, 1999, EA-ELEC-VOLT-26, EA-ELEC-MISC-022
and EA-ELEC-LDTAB-029
will be revised to resolve the deficiencies
noted above. 10/1/98 Update: Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item
(50-255/97201-26)
was identified
a:s closed. Analyses EA-ELEC-VOLT-26, EA-ELEC-MISC-022
and EA-ELEC-LDTAB-029
will be revised by January 15, 1999 as projected
above. Inspection
Followup Item 50-255/97-201-27
The team noted that TS Section 4. 7.1.b required testing to be performed
at every. refueling
to demonstrate
the overall automatic
operation
of the emergency
power system. Proper operation
was verified by bus load shedding and automatic
starting of selected motors and equipment
to establish
that emergency
power had been restored within 30 seconds. FSAR Tables 8-6 and 8-:-7 stated that sequencing
would occur in 65 seconds. Technical
Surveillance
Procedure
RT-BC, "Engineered
Safeguards
System -Left Channel," Revision 8, and RT-8D, "Engineered
Safeguards
System -Right Channel," Revision 8, required performance
testing to be within the 65-second
requirement.
The team questioned
the use of a 30-second
test duration in the TS instead of a 65-second
duration, which would demonstrate
that all required equipment
would start. The licensee stated that the TS did not specifically
require full testing of the entire diesel load sequence but only required testing of selected loads. The team noted that the licensee was testing the diesel loading to the full accident loading sequence and has submitted
a proposed TS change which would be more consistent
with the current design. The team reviewed Test Procedures
R0-128-1, "Diesel Generator
1-1 24 Hour Load Run," Revision 2, and R0-128-2, "Diesel GeneratOr
1-2 24 Hour Load Run," Revision 2. The team noted that Section 3. O of the Acceptance
Criteria and Operability
Sheet for Procedure
R0-128-2 referred to TS Section 3. 7. 1 and 4. 7. 1. 11, and that these references
would only be correct when the proposed improved TS, which have been submitted
to NRG for approval, became effective.
The licensee issued CR C-PAL-97-1566
to resolve these discrepancies.
The team identified
this item as Inspection
Followup Item 50-255197-201-27.
Palisades
60 Day Response:
Several issues identified
in the Design Inspection
are associated
with interpretation
of existing Technical
Specifications.
On December 27, 1995 we submitted
an electrical
technical
specifications
change* request which served to resolve the discrepancy
noted above pertaining
to the Emergency
Diesel Generator (EOG) load sequence test. On January 26, 1998, we submitted
a request for improved technical
specifications
which specifies
testing the EOG to the load 30 
ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS intervals
programmed
by the sequencer;
eliminating
any specific reference
to the sequence time. It is expected that the amendment
resulting
from the most recent .technical
specification
change request will serve to resolve this and other technical
specification
related open items. 1011198 Update: Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item
(50-255/97201-27)
was identified
as closed. In July 1998, Amendment
180 of the Palisades
Electrical
Technical
Specifications
was implemented.
Amendment
180 specifies
testing the EOG to the load intervals
programmed
by the sequencer;
eliminating
specific reference
to the sequence time. No further actions are planned for this item. Inspection
Followup Item 50-255197-201-28
The team identified
the following
discrepancies
when reviewing
station battery Test Procedures
RE-83A, "Service/Modified
Performance
Test-Battery
No. ED-01," Revision 9, and RE-83B, "Service/Modified
Performance
Test-Battery
No. ED-02," Revision 9: * The tests evaluated
whether the final discharge
voltage (105 V de) of station batteries
ED-01and02 was met at the end of the test (4 hours). Load parameters (amps) at 1 and 239 minutes were not verified during the test. These load parameters
were design requirements
of EA-ELEC-LDTAB-009, Revision 2. The licensee demonstrated
that the 1-and 239-minute data were recorded elsewhere
and that the duty cycle was* tested in accordance
with the design requirements.
The licensee stated that the battery testing procedures
would be revised to include verification
of these design parameters.
* The procedures
did not require any calibration
tolerances
for the discharge
testing shunt and control unit. The licensee stated that the tolerance
was removed from the procedure
before testing during the 1996 refueling
outage and issued PCRs 5422 and 5423 to change the
procedures
to include these tolerances.
* The battery charging data in Procedure
RE-83B for the 1996 refueling
outage did not meet Step 5. 2. 2, which required the battery charging rate to be decreasing
and to remain within 5 percent over the last 8 hours before stopping the equalization
process, in that the process was stopped before the end of the 8-hour period. The licensee stated that the nearly steady state voltage operation
of the charger gave adequate assurance
that the battery was operable before exiting the test and issued CR C-PAL-97-1460
to resolve this discrepancy.
* During the performance
of procedure
RE-83B at the 1996 refueling
outage, the elapsed time recorded manually did not agree with the testing control unit time. The licensee stated that because the testing unit did not have the capability
to record the time, the test start and stop times were recorded manually.
The inconsistencies
were minor and had no effect on the test results. The licensee issued C-PAL-97-1460
to evaluated
this discrepancy.
The team identified
this item as Inspection
Followup Item 50-255197-201-28.
31 
* * ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS Palisades
60 Day Response:
Note: Inspection
Followup Item 50-255/97-201-28, Unresolved
Item 97-201-30
bullets 7, 8, 9, 10, 11 and 12, and Unresolved
Item 97-201-31
bullets 6 and 13 are completed
under this action due to their subject similarity.
Surveillance
tests RE-83A and RE-838 will be revised as appropriate
to eliminate
the identified
deficiencies
to support 1998 refueling
outage performance.
By December 15, 1998, we will review DC system requirements, FSAR Chapter 8 and surveillance
tests RE-83A and RE-838 for consistency, and resolve the deficiencies
identified
in this open item and the following:
* Reconcile
FSAR section 8.2.3 concerning
the battery supplying
safe shutdown loads for 4 hours with the requirement
to strip loads. (Inspection
report item #30-7.) * * Disposition
battery shunt and de tie breakers which are not consistent
with FSAR section 8.3.5.2. (Inspection
report item #30-8.) * Reconcile
one battery charger capability
to supply normal loads and recharge battery in less than 9 hours with FSAR section 8.3.5.3. (Inspection
report item #30-9.) * Reconcile
alternate
alignment
of battery chargers with FSAR section 8.4 .. 2.2. (Inspection
report item #30-10.) * Reconcile
battery chargers cross connection
with FSAR section 8.5.2. (Inspection
report item #30-11.) * Reconcile
design of system 1, 2, 3, 4 circuits and their separation requirements
with FSAR section 8.5.3.2. (Inspection
report item #30-12.) * Add battery discharge
restriction
to the D8D. (Inspection
report item #31-6.) * Disposition
battery cell specific gravities. (Inspection
report item #31-13.) 10/1/98 Update: Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION'', this item (50-255/97201-28)
was identified
as open. Surveillance
tests RE-83A and RE-838 were revised and satisfactorily
performed
during the 1998 refueling
outage. The June 30, 1998 FSAR revision resolved inspection
report items #30-8, #30-9, and #30-12. The above remaining
items are scheduled
to be complete by December 15, 1998 .. Inspection
Followup Item 50-255/97-201-29
The team reviewed the following
electrical
modification
packages and found them consistent
with the plant design basis: 32 
* * * * * * * * ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS Temporary
Modification
TM-96-027, "lnsta/1152-Spare
#5 Breaker in 152-113 Cubicle," dated April 10, 1996 FES-95-206, "ED-01 and ED-02 Station Battery Replacement," Revision O FC-364, "Feeder Change for Instrument
Bus Y-01," Revision O FC-854, "Y-01 Power Supply Feed Modification," Revision 0 FC-638, ''Add Component
Cooling Water Pumps to the Normal Shutdown Sequencer," Revision 0 FC-798, "Battery Room Temperature
Indication
and Alarm," Revision O FC-683, "Removal of Pressurizer
Heaters from SIS Trip," Revision O Except as previously
discussed, all these modifications
were adequately
prepared, provided the necessary
technical
basis for the changes, and contained
adequate installation
instructions
and testing requirements.
The 10 CFR 50. 59 safety evaluations
were adequate, except for the two listed below: = Safety Reviews 95-1431and95-1432, dated July 7, 1995, for FES-95-206
stated that the battery duty cycle service test duration for station .batteries
ED-01 and ED-02 was changed from 2 hours to 4 hours. The licensee noted that TS Section 4. 7.2.c was affected by this design change. However, the USQ evaluation, Question 2 of Section II, was not checked "Yes" for a TS change. TS 4. 7.2.c required that a 2-hour battery test be performed;
while design analysis ELEC-LDTAB-009
and FSAR Section 8.4.2 required a 4-hour battery duty cycle. The licensee has submitted
a proposed TS change to reflect the proper battery test duration and issued CR C-PAL-97-1551
to address this discrepancy.
* The safety review documentation
for TM-96-027
stated that the FSAR was not reviewed.
Administrative
Procedure
3. 07, "Safety Evaluations," page 12, required that the FSAR be reviewed and that thos*e sections reviewed be noted on the safety review sheet. The licensee initiated .C-PAL-97-1493
to evaluate this discrepancy.
The team identified
these safety review discrepancies
as Inspection
Fol/owup Item 50-255197-
201-29. Palisades
60 Day Response:
It was not documented
in the safety evaluation
for FES-95-206
that a technical
specification
change would be required to change the battery duty cycle service test duration from 2 to 4 hours. An FES-95-206-specific
technical
specifications
change was not considered
necessary
by the preparer of the safety evaluation
since a technical
specifications
change request eliminating
reference
to a specific duty cycle time was to be submitted
under the Improved Technical
Specifications
Program in the near term. Since completion
of the FES-95-206
safety evaluation, Palisades
has implemented
a Safety & Design Review Group which reviews and approves all design changes and safety evaluations.
The purpose for forming and employing
this group is to provide consistent
oversight
The quality of safety evaluations
and their reviews has significantly
improved over the recent years. It is unlikely that a safety evaluation
deficiency, similar to that associated
with FES-95-206, would have occurred
since deployment
of the Safety & Design Review Group. 33 
* * ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS The original safety review for TM-96-027
inappropriately
indicated
that FSAR sections had not been reviewed.
In reality, the FSAR was reviewed during safety review preparation
and the FSAR was found to contain description
at a level of detail that the TM would not affect. The review of the TM-96-027 safety review was performed
by telecon (an infrequent
practice)
with no follow-up
review performed
by the Safety & Design Review telecon reviewer.
By April 15, 1998, design control procedures
will be revised to require a follow-up
review whenever a review is performed
by telecon. 1011198 Update: Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND.NOTICE
OF VIOLATION", this item (50-255/97201-29)
was identified
as closed. Administrative
Procedure
AP 3.07, "SAFETY EVALUATIONS" was revised to require follow-up
reviews as stated above. No further actions are planned for this inspection
item. Note: Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-30)
was identified
as open. FSAR changes identified
in Unresolved
Item 50-255/97201-30
are identified
below. Some of these bullets are grouped and evaluated
with other URl's or IFl's. For clarity, each bullet's actions will be separately
addressed.
Unresolved
Item 50-255197-201-30
The team identified
the following
discrepancies
in the FSAR: * Page 6. 7-4 stated that 'containment
isolation
valves fail closed with loss of voltage or control air except for the CCW return isolation
valves. However, the CCW supply isolation
valve (CV-0910)
is also a fail-open
valve and should have *been noted as an exception
to fail-closed
containment
isolation
valves. The licensee issued FSAR Change Request 6-142-R20-1426
to correct the FSAR. Palisades
60 Day Response:
The next FSAR annual update revision will incorporate
this change. 1011198 Update: Annual FSAR update issued June 30, 1998, included this change. * Section 6. 7 classified
the CCW penetrations
as Class C-2, which was defined as penetrations
with lines not missile protected.
However, EA-GW0-7793-01
stated that the entire CCW system (both inside and outside containment)
was missile protected.
The licensee issued FSAR Change Request 6-143-R20-1427
to state that the CCW penetrations
were not vulnerable
to internally
generated
missiles . 34 
* * ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS Palisades
60 Day Response:
The next FSAR annual update revision will incorporate
this change. 10/1/98 Update: Annual FSAR update issued June 30, 1998, included this change. * Table 9-10 stated that valves 3029 and 3030, containment
sump suction valves, failed closed upon loss of air and were equipped with an accumulator.
The valves actually failed as is and had no accumulator.
The licensee issued FSAR Change Request 9-293-R20-1431
to correct *the FSAR and CR C-PAL-97-1559
to evaluate and trend the FSAR discrepancies
being identified
at the plant. Palisades
60 Day Response:
The next FSAR annual update revision will incorporate
this change. 10/1/98 Update: Annual FSAR update issued in June 30, 1998, included this change. * Table 9-9 correctly
stated that the high-pressure
air piping was seismic Class I from the receivers
to the valve operators.
However, FSAR Table 5.2-3 stated that the entire system was seismic Class I. The licensee issued FSAR Change Request 5-155-R20-1432
to correct the FSAR 5. 2-3. Palisades
60 Day Response:
The next FSAR annual update revision will incorporate
this change. 10/1/98 Update: Annual FSAR update issued in June 30, 1998, included this change. * Section 8.4.2.2 stated that the station batteries
would be tested to Institute
of Electrical
and Electronics
Engineers (IEEE) 450-1975.
However, battery testing procedures
RE-83A, Revision 9, and RE-838, Revision 9, referred to IEEE 450-1995.
FSAR Change Request 8-126-R20-1249
had been initiated, but the licensee did not intend to act on this change until approval was received from NRG of a related proposed TS change. Palisades
60 Day Response:
This FSAR change is on hold until the license amendment
responding
to our improved electrical
technical
speeification
change request, submitted
January 26, 1998, is received.
This change cites IEEE 450-1995 for the battery testing . 35 
* ** ATTACHMENT
A * STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS 1011198 Update: Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-30)
was identified
as open. In July 1998, Amendment
180 of the Palisades
Electrical
Technical
Specifications
was implemented
with IEEE 450-1995 as a reference.
FSAR change 8-126-R21-1249
will be implemented
as part of the next annual FSAR update. to reflect the use of this IEEE standard.
* Table 5. 7-8 listed the seismic design value for the station batteries
and racks as "later" instead of including
the actual values of the batteries
installed
by FES-95-206.
The licensee issued EAR-97-0636
to evaluate this discrepancy
and revise the FSAR. Palisades
60 Day Response:
The table in the FSAR is designated
as containing
the original seismic design values for the plant. The term "later" was an original FSAR description
which acknowledged
that an impending
upgrade to install a second redundant
electrical
train would be made and the applicable
seismic criteria would not be available
until then. Since we have chosen to keep this table for historical
record, the word "later" will be removed and the table maintained
as original seismic criteria.
The next FSAR annual update will incorporate
this change requested
by FSAR Change Request 5-157-R20-1456.
1011198 Update: Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this portion of unresolved
item 50-255/97201-30
was identified
as closed. The annual FSAR update issued June 30, 1998, included this change. No further actions are planned for this inspection
item. * Section 8.2.3 stated the "The de battery system is designed to supply the required shutdown loads, with a total loss of ac power, for at ieast 4 hours." This statement
did not reflect the fact that load stripping
was required during the 4 hours for the battery to perform its intended function during a loss of ac power. * Palisades
60 Day Response:
Refer to our response to Inspector
Followup Item 50-255/97-201-28.
1011198 Update: The resolution
of this issue is addressed
in Inspection
Followup Item 50-255/97201-28
due to subject similarity.
This item is projected
to be complete by December 15, 1998. Section 8. 3. 5. 2 stated that "Operation
of all circuit breakers in the de and the preferred
ac systems is manual with automatic
trip for fault isolation." The battery shunt trip breakers and the de bus tie breakers do not comply with this statement.
36 
* ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS Palisades
60 Day Response:
Refer to our response to Inspector
Followup Item 50-255/97-201-28.
10/1/98 Update: Revision 20 of FSAR Chapter 8 incorporates
the exclusion
of the battery isolation
shunt trip breakers and tie breakers between the left and right sections of each switchgear
bus that do not have an automatic
trip for fault isolation.
Our June 30, 1998, annual FSAR update includes this change. * Section 8. 3. 5. 3 stated that "Each of the two battery chargers provided on the. de bus is capable of supplying
the normal de loads on the bus and simultaneously
recharging
the battery in a reasonable
time. A fully discharged
battery can be recharged
in less than nine hours." Contrary to the statement, one battery charger could not supply the normal loads and recharge a fully discharged
battery in less than 9 hours. Palisades
60 Day Response:
Refer to our response to Inspector
Followup Item 50-255/97-201-28.
10/1/98 Update: Revision 20 of FSAR Chapter 8 now states that two battery chargers are needed to recharge a fully discharged
battery in less than nine hours. Our June 30, 1998, annual FSAR update includes this change. * Section 8.4.2.2 stated that "Emergencv
Operation
-.On loss of normal and standby ac power, the batteries
will supply power to all preferred
ac and de loads, until one of the (diesel generators)
DGs has started and can supply power for the chargers." This statement
was not correct if the battery chargers were in their alternate
alignment
and did not reflect load shedding during the 4-hour duration.
Palisades
60 Day Response:
Refer to our response to Inspector
Followup Item 50-255/97-201-28.
10/1/98 Update: The resolution
of this issue is addressed
in Inspection
Followup Item 50-255/97201-
28 due to subject similarity.
We plan to complete this item by December 15, 1998. * Section 8.5.2 stated that ''The power source for the driven equipment
and the control power for that system are supplied from the sources in one channel." This statement
would not be correct if the battery chargers were cross-connected . 37 
* ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS Palisades
60 Day Response:
Refer to our response to Inspector
Followup Item 50-255/97-201-28.
10/1/98 Update: The resolution
of this issue is addressed
in Inspection
Followup Item 50-255/97201-
28 due to subject similarity.
We plan to complete this item
by December 15, 1998. * * Section 8.5.3.2 referred to "System 1, 2, 3, 4 Circuits" and separation
requirements
for those circuits.
The licensee was not able to identify these circuits.
* Palisades
60 Day Response:
Refer to our response to Inspector
Followup Item 50-255/97-201-28.
10/1/98 Update: Revision 20 of FSAR Chapter 8 expands the definition
along with providing
routing and isolation
requirements
for 'left', 'right' and channel '1 ', '2', '3', and '4' circuits.
Our June 30, 1998, annual FSAR update includes this change. * Section 8.4.1.3 required clarification
as to whether the reserve capability
margin referred to the capability
of the overall EDG and engine or if it referred to the capability
of the EOG to handle an increase loading due to a control circuit ma/function
during the loading sequence.
The licensee issued C-PAL-97-1309
to resolve this discrepancy.
Palisades
60 Day Response:
Prior to the Design Inspection, an operability
determination
was made concluding
that the EDGs are operable.
This conclusion
was reached based on the capability
of the EDGs to provide the required design function
in the event of a control. circuit malfunction
or delayed containment
high pressure signal; but not both concurrently.
The design basis accident analysis does not require that these two events occur simultaneously.
Due to the change being descriptive
in nature, rather than licensing
basis information, we have elected to use the Design Basis Documents
rather than the FSAR to make the clarification.
Design Basis Document Change 5.03-11-R3-
0617 was initiated
and the revision will be made by December 15, 1998. 10/1/98 Update: Revision 4 of DBD 5.03 incorporates
the requested
change which evaluated
the system functional
requirements
of the EOG starting and carrying the largest load due to a control circuit malfunction.
Revision 4 also includes discussion
regarding
the EOG control circuit malfunction
and starting a containment
spray pump during a delayed containment
high pressure scenario;
*concluding
that the malfunction
and the pump start are mutually
exclusive.
No further actions are planned for this item. 38 
* .ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS * Section 6.1.2.3 stated that ''The RAS ... provides a permissive
to manually close the valves in the pump minimum flow lines." EOP-4. 0, "Loss of Coolant Accident Recovery," Revision 9, Step 23, directed the operators
to place the hand switches for these valves in the pump minimum flow lines (CV-3027 and CV-3056) to CLOSE when SIRWT level lowered to between 25 percent and 15 percent. Per EOP-4.0, Step 51, the RAS occurred when the SIRWT level reached 2 percent. The FSAR appeared to conflict with EOP-4.0. The licensee initiated
FSAR Change Request 6-141-R20-1425
to update the FSAR. Palisades
60 Day Response:
The next FSAR annual update revision will incorporate
this change. 10/1/98 Update: Annual FSAR update issued in June 30, 1998, included this change. * The footnote for Table 14.17.1-1
implied that a containment
building temperature
of 90 °F was used as input to the large-break
LOCA analysis because it is the limiting temperature
during normal operation.
The 90 °F value did not appear to be limiting.
The licensee stated that the 90 °F value was the nominal containment
building temperature, not the limiting temperature, and was used in the accident analysis in accordance
with Seimens Power Corporation's
large-break
LOCA methodology
guidelines.
The licensee initiated
FSAR Change Request 14-95-R20-1441
to update the FSAR. * Palisades
60 Day Response:
The next. FSAR annual update revision will incorporate
this change. 10/1/98 Update: Annual FSAR update issued in June 30, 1998, included this change. The above discrepancies
had not been corrected
and the FSAR had not been updated to ensure that the material in the FSAR contained
the latest material
as required by 10 CFR 50. 71(e). The team identified
this item as Unresolved
Item 50-255197-201-30.
Palisades
60 Day Response:
10 CFR 50.71(e) requires that the FSAR be updated to contain the latest material developed
and that it includes the effects of all changes made in the facility or procedures
described
in the FSAR. Although several of the identified
FSAR discrepancies
were clear errors, most were cases where statements
in the FSAR were misleading
or unclear and not cases where the FSAR was not updated per 10 CFR 50.71 (e). Our ongoing FSAR verification
and validation
effort should provide identification
and correction
of similar conditions
which may exist in the FSAR. Our processes
were also changed a few years ago to require a safety review (1 O CFR 50.59 screening)
for 39 
ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS all analyses, modifications, etc which have the potential
to affect the design basis of the facility.
This widespread
10 CFR 50.59 screening
will prevent failures to update the FSAR in accordance
with 10 CFR 50.71(e).
In addition, a license basis self assessment
performed
in accordance
with NEI 96-05, "Guidelines
for Assessing
Programs for Maintaining
the Licensing
Basis," found few discrepancies
in the FSAR sections sampled which
had not been previously
identified
for correction
by other plant processes.
Therefore, we feel that the current efforts underway will correct other errors which may exist in the FSAR and the current plant processes
will ensure that the FSAR is updated properly.
10/1/98 Update: The above response remains unchanged
from our 60-day response.
Note: Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-31)
was identified
as open. DBD changes identified
in Unresolved
Item 50-255/97-201-31
are identified
below. Some of these bullets are grouped and evaluated
with other UR l's or IFl's. For clarity, each bullet's actions will be separately
addressed.
Unresolved
Item 50-255/97-201-31
The team identified
the following
discrepancies
in the DBDs: * DBD 1.07, ''Auxiliary
Building HVAC Systems," Revision 1, Table 3.2.1, incorrectly
stated that the design basis temperature
for Room 123, which contains the CCW pumps, was 125 °F. The correct temperature
was 104 °F as stated in 080 7.01, "Electrical
Equipment
Qualification
Program," Revision 1, Appendix A. The 125 °F temperature
was a conservative
assumption
used to size the outside air supply fans. Table 3.2.1 also contained
a typographical
error in a reference
number. The licensee issued 080 Change Requests 1.07-71-R1-0512
and 1.07-72-R1-0532
to correct the 080. Palisades
60 Day Response:
The identified
Design Basis Document Change Request will be incorporated
into the DBD by July 1, 1998. 10/1/98 Update: Revision 2 of DBD 1.07, "Auxiliary
Building HVAC Systems" incorporated
the above changes. The basis for the 125 ° F CCW room temperature
was clarified
and references
were corrected.
* 080 1.07, Revision 1, Section 3.2.1.3, listed maximum room temperatures
for the west ESF room from an outdated analysis.
The latest analysis, EA-O-PAL-93-272F-01, "Engineering
40 
* ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS Safeguards
Room Heatup Following
LOCA in Conjunction
With a Loop," Revision 0, determined
lower maximum room temperatures
for various SW flows through the air coolers. The 080 also required clarification
of the normal design temperature
of the ESG room. The licensee issued 080 Change Request 1.07-73-R1-0543
to correct the 080. Palisades
60 Day Response:
The identified
Design Basis Document Change Request will be incorporated
into the DBD by July 1, 1998. 10/1/98 Update: Revision 2 of DBD 1.07, "Auxiliary
Building HVAC Systems" incorporated
the above change. The basis for the 135°F Engineering
Safeguards
Room temperature
was clarified.
* 080 7. 08, "Plant Protection
Against Flooding, 77 Revision 1, incorrectly
stated that the EOG would be inoperable
before a flood reached the EOG windings because the lube oil heaters were located below the windings at 7 inches above the floor. EA-C-PAL-95-1526-01, "Internal
Flooding Evaluation
for Plant Areas Outside of Containment, 77 Revision 0, stated that the minimum flood level at which the EOG could become inoperable
was 10 inches due to the exciter cubicle bus bars and that the lube oil heaters were not needed for EOG * operability.
The licensee issued CR C-PAL-97-1557
to initiate a 080 change and evaluate the item. Palisades
60 Day Response:
During the Design Inspection, an operability
determination
concluded
that the EDGs * are operable based on other indications
available
to inform operations
that water level in the rooms is increasing.
DBD change request 7.08-40-R1-0561
was initiated
to state that the limiting component
is not lube oil heaters but the exciter cubicle bus bars located ten inches above the EOG room floor. The identified
Design Basis Document Change Request will be incorporated
into the DBD by December 15, 1998. 10/1/98 Update: This DBD change is on target for completion
by December 15, 1998 as identified
above. * 080 2. 03, "Containment
Spray System, 77 Revision 2, stated that the air supply to the sump outlet valves, CV-3029 and 3030, was backed by an accumulator.
There were no accumulators
for these valves. The licensee identified
this error while evaluating
an FSAR statement
that these valves had an accumulator
backup that was questioned
by the team, and issued 080 Change Request 2.03-22-R2-0531
to correct the 080 . 41 
* ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS Palisades
60 Day Response:
The identified
Design Basis Document Change Request will be incorporated
into the DBD by July 1, 1998. 10/1/98 Update: Revision 3 of DBD 2.03, "Containment
Spray System" corrected
the terminology
from "accumulator" to "high pressure air receivers".
No further action is planned. * DBD 1.01, "Component
Cooling Water System," Revision 3, Section 3.3. 7, incorrectly
indicated
that Class 1 E and non-Class 1 E breakers were installed
in the same distribution
panels. The licensee initiated
DBD Change Request 1.01-14-R3-0518
to correct the DBD. Section 3. 3. 7 of this DBD also stated that solenoid valves had been tested to operate at 87 V de instead of 90 V de. The licensee stated that the DBD would be corrected.
Palisades
60 Day Response:
The identified
Design Basis Document Change Request will be incorporated
into the DBD by July 1, 1998. 10/1/98 Update: Due to competing
priorities, this DBD change has been rescheduled
to be completed
by December 15, 1998. * * * During the teain's review of FES-95-206, it was noted that the battery manufacturer
had imposed a limit of 40 battery discharges
for the 20-year life of the battery. This restriction
had not been identified
in any DBD. The licensee stated that the requirement
would be added to DBD4.01. . Palisades
60 Day Response:
A Design Basis Document Request will be incorporated
into the DBD by December 15, 1998. Refer to our response to Inspector
Followup Item 50-255/97-201-28.
10/1/98 Update: This DBD change is on target for completion
by December 15, 1998, as
above. * Appendix A of DBD 7. 02, "Palisades
Design Basis Document EQ Master Equipment
List," Revision 2, incorrectly listed
the location for L T-0383; referred to EIP 0343 instead of E/P 0346; and did not include SV-32138 in Table A-1. The licensee issued DBD Change Requests 7. 02-4-R2-0522, 7. 02-6-R2-0527, and 7.D2-4-R2-0523
to correct the DBD . 42 
* * * ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS Palisades
60 Day Response:
The identified
Design Basis Document Change Request will be incorporated
into the DBD by December 15, 1998. 10/1/98 Update: These DBD changes are on target for completion
by December 15, 1998. * DBD 2.01, "Low Pressure Safety Injection
System," Revision 3, and DBD 2.02, "High Pressure Safety Injection
System," Revision 3, both contained
references
to ANF-88-107, "Palisades
Large Break LOCNECCS Analysis With Increased
Radial Peaking," Revision 1. ANF-88-107
was superseded
by Seimens Calculation
EMF-96-172, "Palisades
Large Break LOCNECCS Analysis," Revision 0. The licensee Initiated
DBD Change Requests 2. 01-30-R3-0519 and 2.02-27-R3-0520
to update the DBDs. * Palisades
60 Day Response:
The identified
Design Basis Document Change Request will be incorporated
into the DBD by July 1, 1998. 10/1/98 Update: Revision 4 of DBD 2.01, "Low Pressure Safety Injection
System," and Revision 4 of DBD 2.02, "High Pressure Safety Injection
System," incorporated
reference
to the most current LOCA analysis.
No further action is planned for this item. DBD 2.01, "Low Pressure Safety Injection
System," Revision 3, Section 3.3.1.3, stated that the SIRWT must maintain a minimum of 20,000 gallons at the time of a RAS to limit the radiological
consequences
of an accident.
The DBD reference
for this statement
was TAM-95-05, "Radiological
Consequences
for the Palisades
Maximum Hypothetical
Accident & Loss of Coolant Accident," Revision 0. A review of EA-TAM-95-05
indicated
that this analysis did not take credit for the 20,000 gallons at the time of RAS to limit the radiological
consequences
of an accident.
The licensee issued DBD Change Request 2.01-31-R3-0524
to update the DBD. Palisades
60 Day Response:
The identified
Design Basis Document Change Request wlll be incorporated
into the DBD by July 1, 1998. 10/1/98 Update: Revision 4 of DBD 2.01, "Low Pressure Safety Injection
System," clarifies
the SIRW tank minimum volume design requirements.
No further action is planned for this item . 43 
* * * ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS The team also identified
the following
discrepancies
in other documentation:
* P&ID M-232, Sheet 2A, incorrectly
identified
L T-0383 as connected
to penetration
#54 instead of#56. The licensee issued Document Change Request (OCR) 97-0856lo
correct the drawing.
Palisades
60 Day Response:
P&ID M-232, Sheet 2A has been reviseo to incorporate
OCR 97-0856. 10/1/98 Update: No further update necessary.
* Documents
E-33, Revision 46, and E-37, Revision 46, were not revised to reflect the installed
condi(ion
of the battery charger cabling that was rerouted by SC-89-284.
The licensee issued CR C-PAL-97-1495
to resolve this discrepancy.
Palisades
60 Day Response:
E-33, Rev 46 and E-37, Rev 46 have been revised to reflect the correct battery charger cable routing installed
by SC-89-284 . 10/1/98 Update: * No further
necessary.
* * P&ID M-209, Sheet 3 (Revision
34), incorrectly
depicted valves SV-0918 and SV-09778 as normally deenergized.
The licensee issued EAR 97-0652 to revise the drawing. * Palisades
60 Day Response:
P&ID M-209, Sheet 3, Revision 35 has been issued to depict SV-09778 as normally energized.
Further evaluation
of SV-0918 identified
that the normally deenergized
state as depicted on M-209 Sheet 3 is appropriate
per FSAR Table 9-10. 10/1/98 Update: No further update necessary.
* Vendor drawing E-12A, Sheet 39, Revision 0, indicated
that the battery discharge
characteristics
were based upon battery cell specific gravities
of 1.215 +/- 0.005. However, the batteries
were being maintained
to a criterion
of 1.215 +/- 0.010. The licensee issued EAR 97-0669 to update the drawing. Palisades
60 Day Response:
E-12 A, Sheet 39, Rev O will be updated by December 15, 1998. Refer to our response 44 
ATTACHMENT
A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION
OPEN ITEMS to Inspector
Followup Item 50-255/97-201-28.
10/1/98 Update: This item is on target for completion
by December 15, 1998. These documentation
discrepancies
were not consistent
with 1 O CFR Part 50, Appendix B, Criterion
Ill, "Design Control," which requires that the design basis be correctly
translated
into drawings.
The team identified
this item as Unresolved
Item 50-255197-201-31.
The programmatic
design control aspects related to this issue will be addressed
as identified
in Attachment
B, Item 1. 'i 45 
----* * ATTACHMENT
B CONSUMERS
ENERGY COMPANY PALISADES
PLANT DOCKET 50-255 STATUS OF PLANS TO RESOLVE RELATED PROGRAMMATIC
ISSUES AND EVALUATE FINDINGS AGAINST THE 10 CFR 50.54(f) RESPONSE 6 Pages 
* * ATTACHMENT
8 STATUS OF PLANS TO RESOLVE RELATED PROGRAMMATIC
ISSUES AND EVALUATE FINDINGS AGAINST THE 10 CFR 50.54(f) RESPONSE Per NRC correspondence
dated May 18, 1998, titled "NRC INSPECTION
REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", inspection
item 50-255/98003-01
was identified
as open. As stated in the report, this item
will remain open
pending NRC review of the results of the collective
significance
of individual
inspection
items and planned programmatic
improvements.
The following
summarizes
of our programmatic
improvements.
1. DESIGN CONTROL ISSUES: The following
issues were identified
in the Design Inspection
report as potentially
not meeting requirements
of 10 CFR 50, Appendix B, Criterion
Ill, "Design Control." Our design control program provides assurance
that the plant as-built configuration
conforms to design requirements, and the configuration
is operated, tested and maintained
within required design parameters.
The deficiencies
identified
during the Design Inspection
relate to these design control program objectives.
Design Objective
For Operating
Systems Within Design Parameters:
* Loss-Of-Coolant
Accident analysis identified
the maximum CCW temperature
of 184°F yet the effects of this temperature
on CCW system components
was not performed. (Unresolved
Item 50-255/97-201-02.)
* Incomplete
analysis (inadequate
justification
for conclusion
and incorrect
references
to related NRC correspondence)
for CCW piping for High Energy Line Break. (Unresolved
Item 50-255/97-201-04.)
* Some AC Load calculations
have not been updated to reflect current design. (Unresolved
Item 50-255/97-201-14.)
Design Objective
For As-Built Conditions
Conforming
To Design Requirements:
* * * Unscreened
Emergency
Core Cooling System Suction piping vent. (Unresolved
Item 50-255/97-201-10.)
Some instrument
tubing is not sloped consistent
with design requirements . (Unresolved
Item 50-255/97-201-13.)
Design Basis Document I design documentation
discrepancies. (Unresolved
Item 50-255/97-201-31.)
1 
* ATTACHMENT
B STATUS OF PLANS TO RESOLVE RELATED PROGRAMMATIC
ISSUES AND EVALUATE FINDINGS AGAINST THE 10 CFR 50.54(f) RESPONSE Palisades
60 Day Response:
Elements comprising
and supporting
our.design
control program consist of our calculation
control program, instrument
setpoint program, FSAR verification
and validation (V&V), design basis documents (DBDs) with associated
safety system design confirmations, and as-built confirmation
through drawing review or field walkdown.
These elements will be revised as appropriate
by December 15, 1998 to prevent the recurrence
of conditions
similar to those identified
in the Design Inspection
and cited above. Resolution
of any nonconforming
conditions
identified
will be implemented
through our corrective
action program. 10/1/98 Update: Programs exist at Palisades
that ensure proper station design attributes
are considered, evaluated, changed and documented.
These programs makeup our overall "Design Control" program. In past months, several programs have been reviewed in various inspections
and routine assessments
such as: * NRC INFORMATION
NOTICE 98-22:"DEFICIENCIES
IDENTIFIED
DURING NRC DESIGN INSPECTIONS" was evaluated
by comparing
the adequacy of our program design controls against other station Design Inspection
identified
concerns.
* Self assessments
were performed
in areas such as design document control and modification
programs.
* NRC inspections
and internal NPAD audits in the areas of Engineering
and Technical
Support were performed
in mid 1998 that evaluated
several Palisades
design and configuration
program attributes.
As a result of these and other efforts, "Design Control" Program enhancements
have been identified
and incorporated
into the appropriate
programs.
For example, several changes have been made to design change processes
to better define the applicability
of each distinct process, and to ensure that design change inpuUoutput
requirements
are adequately
addressed.
2 
* ATTACHMENT
B STATUS OF PLANS TO RESOLVE RELATED PROGRAMMATIC
ISSUES AND EVALUATE FINDINGS AGAINST THE 10 CFR 50.54(f) RESPONSE No major programmatic
weaknesses
were identified
in these reviews and program enhancements
are now complete.
To conclude, the Palisades "Design Control" Program is considered
effective.
2. CALCULATION
CONTROL ISSUES: The Design Inspection
issues identified
below reflect weaknesses
in our calculation
control program. Improvements
in our calculation
control program will serve to prevent recurrence
of these conditions.
Inspection
Report Issues: * Required justification
for conclusion
and correct references
to related NRC correspondence
not provided in analysis. (Unresolved
Item 50-255/97-201-04.)
* Not all analyses revised whenever analytical
inputs or major assumptions
change. (Unresolved
Item 50-255/97-201-07.)
* Analyses not reflecting
accurate as-built configuration
and system operation, not all interdependent
analyses have been revised together in response to changes, and analytical
design bases do nofagreewith
test acceptance
criteria. (Unresolved
Item 50-255/97-201-08.)
Palisades
60 Day Response:
Prior to the Design Inspection, calculation
control weaknesses
were recognized
and an improvement
plan was implemented.
Over 19,000 calculations
have .been indexed to provide for improved retrievability.
A cross-index
between selected calculations
of record and the documents
that use the results of the calculations
is being developed.
These and other improvements
to our calculation
program serving to prevent recurrence
of the deficiencies
cited above will be made by December 15, 1998. 10/1/98 Update: The identification
of calculations
referenced
in the major design documents
has been completed.
The Calculation
Control Improvement
Project is on target for 3 
* * * ATTACHMENT
8 STATUS OF PLANS TO RESOLVE RELATED PROGRAMMATIC
ISSUES AND EVALUATE FINDINGS AGAINST THE 10 CFR 50.54(f) RESPONSE completion
of the detailed calculation
cross-index
by December 15, 1998. Development
of the computerized
calculation
retrieval
application
and completion
of associated
engineer training will follow in early 1999. 3. SETPOINT CONTROL ISSUES: Station procedures
and guidance to require the use of established
uncertainty
methodology
need to be implemented.
The plan for implementation
should be validated
against weaknesses
identified
in* Unresolved
Item 50-255/97-201-12.
Palisades
60 Day Response:
An instrument
uncertainty
evaluation
methodology
manual has been developed.
Uncertainty
calculations
for Reactor Protection
System and Engineered
Safety Features Actuation
System setpoints
have been performed
Ul?ing .the methodology
manual. Incorporation
of instrument
uncertainty
evaluation
requirements
in procedures, and training select engineers
to perform uncertainty
calculations, will be completed
by December 15, 1998. 10/1/98 Update: As stated in Inspector
Follow-up
Item 50-255/97201-12, station procedures
have been revised to consider use of established
instrument
uncertainty
guidance when developing
test acceptance
criteria and determining
errors for operating
instrument
loops. In addition, a self assessment
of the Setpoint Control Process was performed
with potential
areas for improvement
being evaluated.
4. 10 CFR 50.54(F} RESPONSE:
Evaluate inspection
findings, both specific and programmatic, against the Palisades
response to NRC's October 9, 1996 request for information
pursuant to 1 O CFR 50.54(f) regarding
adequacy and availability
of design bases information.
Palisades
60 Day Response:
After review of the inspection
findings and comparison
to our response to the 1 O CFR 50.54(f) letter regarding
the adequacy and availability
of design basis .4 
.. * ATTACHMENT
B STATUS OF PLANS TO RESOLVE RELATED PROGRAMMATIC
ISSUES AND EVALUATE FINDINGS AGAINST THE 10 CFR 50.54(f) RESPONSE information, we have determined
that our response to the 10 CFR 50.54 (f) letter remains complete and accurate.
Improvements
to our design programs, initiated
through our response, will be directly responsible
for resolution
of issues * identified
within the Design Inspection
report. The programs and projects being improved include our Calculation
Control Program, Setpoint Methodology
and Control Program, FSAR Verification
and. Validation
Project, and our Fuse Control Program. * Beyond programmatic
improvements, design basis knowledge
will be further enhanced by the development
of 1 O additional
DB Os and the performance
of. three safety system design confirmations
similar to the NRC's safety system functional
inspections.
To date, four of the new DBDs have been issued and one design confirmation
has been completed.
No additional
programmatic
improvement
efforts have initiated
as a result of actions being taken
to satisfy our 10 CFR 50.54(f) response.
A final review of the adequacy of our response will be completed
by December 15, 1998. 10/1/98 Update: Some of the initiatives
noted in our 60-day response to the Des_ign Inspection
were not part of Palisades
formal response to the NRC's October 9, 1996 request for information
pursuant to 10 CFR 50.54(f) regarding
adequacy and availability
of design bases information.
Our February 6, 1997, 50.54(f) response coneluded
that the Palisades'
design bases information
was adequate, and reasonabie
assurance
exists that: 1) design bases information
has been translated
into operating, maintenance, and testing procedures, and 2) system, structures, and component
configuration
and performance
are consistent
with the design bases. Our 50.54(f) response also referred to specific initiatives
to further strengthen
plant processes
and design basis documentation.
Specifically
noted as * commitments
in the 50.54(f) response were: 1) performing
an FSAR Verification
Project, 2) completing
ten new Design Basis Documents, 3) conducting
one Safety System Functional
Type inspection
per fuel cycle, and 4) updating and re-instituting
use of a Quality Assurance
Requirements
Matrix database.
Other initiatives
to strengthen
plant processes
and design basis documentation
were also undertaken
that were not specifically
included ln the 50.54(f) response 5 
* ATTACHMENT
B STATUS OF PLANS TO RESOLVE RELATED PROGRAMMATIC
ISSUES AND EVALUATE FINDINGS AGAINST THE 10 CFR 50.54(f) RESPONSE such as: 1) implementing
a calculation
control improvement
project, 2) implementing
improvements
in instrument
setpoint uncertainty
methodology, 3) performing
an assessment
of instrument
setpoint control, and 4) performing
an assessment
of the fuse control program. The 50.54(f) response remains complete and accurate.
The response to Attachment
B Item 1 relates to and supports this position.
It should be noted, however, that the 50.54(f) response and its committed
programmatic
initiatives, along with other initiatives
noted above, will not resolve all issues identified
within the Design Inspection
since it is more effective
to resolve certain issues on an individual,
basis. A formal review that evaluates
the Design Inspection
findings against the 50.54(f) response is on target for completion
by December 15, 1998. 6
}}

Revision as of 19:21, 17 June 2019

Provides Update to Design Insp Action Items Re Insp Rept 50-255/97-201 Conducted on 970916-1114.Util Recommends That NRC Consider Scheduling Efforts Early in 1999 to Review Insp Items for Closure Based on Completion Dates for Items
ML18066A314
Person / Time
Site: Palisades Entergy icon.png
Issue date: 10/01/1998
From: Haskell N
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
50-255-97-201, NUDOCS 9810070265
Download: ML18066A314 (55)
v • d • e


See also: IR 05000255/1997201

Text

A CMS Energy Company October 1, 1998 U.S. Nuclear Regulatory

Commission

Attn: Document Control Desk Washington

D.C. 20555 Palisades

Nuclear Plant 27780 Blue Star Memorial Highway Covert. Ml 49043 DOCKET 50-255 -LICENSE DPR-20 -PALISADES

PLANT * Tel. 616 764 2276 Fax.* 616 764 2490 Nathan L. Ha.ks/I Director.

Licensing

OCTOBER 1, 1998 UPDATE TO DESIGN INSPECTION

ACTION ITEMS During the period from September

16 through November 14, 1997, the NRC conducted

a design inspection

at the Palisades

Nuclear Plant. By letter dated December 30, 1997, the NRC issued Inspection

Report No. 50-255/97-201, and requested

a response within 60 days detailing

our plans to complete the corrective

actions required to resolve the open items listed in Attachment

A of the inspection

report. Contained

within our March 2, 1998 response was a single commitment

to provide the NRC a status of our progress in completing

actions associated

with each open inspection

item. The purpose of this commitment, in part, was to assist the NRC in planning for follow-up

review and closeout of these items. Attachment

A of this letter contains the text of each open inspection

item from the December 30, 1997 inspection

report, followed by our 60 day response as submitted

in our March 2, 1998 letter, followed by the status of associated

action as of October 1, 1998. This status includes the results of our investigations

and corrective

actions, along with planned completion

dates for ongoing actions. Attachment

B contains similar information

for programmatic

issues related to inspection

findings.

_J Based on completion

dates for the remaining

open items, we recommend

that NRC consider scheduling

efforts early in 1999 to review inspection

items for closure. A review of completion

dates for open items indicates

that a majority of actions will be completed

by the end of 1998. 9810070265

981001 PDR ADOCK 05000255 G PDR

-. . .:.; * * -.. -Sl:JMMAR¥-'-8F

COMMITMENTS

This letter closes the March 2, 1998 commitment

as .restated

below, and contains no new commitments. "By October 1, 1998, Consumers

Energy will provide NRC with a status of our progress in completing

all actions identified

in the attachments

to this letter.

  • Nathan L. Haskell . Director, Licensing

CC Administrator, Region Ill, USNRC Project Manager, NRR, USNRC NRC Resident Inspector

-Palisades

Attachments

2

ATTACHMENT

A CONSUMERS

ENERGY COMPANY PALISADES

PLANT DOCKET 50-255 STATUS OF PLANS FOR CORRECTIVE

ACTIONS TO RESOLVE NRC DESIGN INSPECTION

OPEN ITEMS 45 Pages

  • ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS Unresolved

Item 50-255/97-201-01

The team questioned

whether the CCW system design met the vendor-recommended

minimum flow of 2000 gpm for the CCW pumps under all operating

conditions.

The team was concerned

that small differences

in the pump operating

characteristics

could cause significant

differences

in flow through each pump during parallel pump operation

due to the flatness of the pump operating

  • curves at low flows. The licensee had no analysis available

to demonstrate

that the CCW pumps met the minimum flow requirements.

During the inspection, the licensee developed

a preliminary

system flow model, which showed that, when all three pumps were started upon receiving

a safety injection

system (SIS) signal, the minimum pump flow was through CCW pump P-52A at 1768 gpm. The licensee received a revised minimum flow requirement

of 1600 gpm from the pump manufacturer.

The team's review of the licensee's

completed

flow model calculation

will be an Inspection

Fol/owup Item 50-255197-201-01.

  • Palisades

60 Day Response:

As a result of CCW system balancing, scheduled

for the 1998 refueling

outage, a reanalysis

of minimum predicted

CCW system flow rates will be performed.

This reanalysis

will verify that minimum flow rate requirements

will be met under a worst case scenario with appropriate

pump IST degradation

input. This action will be completed

by September

1, 1998. 10/1/98 Update: Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-01)

was identified

as open. Pump performance

data was obtained during the 98 refueling

outage. The completion

for the reanalysis

has been rescheduled

for August 1, 1999 to accommodate

emerging higher priority analytical

work. Unresolved

Item 50-255/97-201-02

The team verified the heat removal capability

of the CCW heat exchangers

by reviewing

the results of various accident analyses.

The licensee had performed

the following

LOCA analyses:

  • EA-D-PAL-93-207-01, "LOCA Containment

Response Analysis With Reduced LPSI Flow Using CONTEMPT El-28 Code," Revision 0, * EA-D-PAL-93-272-03, "LOCA Containment

Response Analysis With Degraded Heat Removal System Using CONTEMPT El-28A Computer Code," Revision 0, *and * EA-GEJ-96-01, "A-PAL-94-324

Containment

Spray System (CSS) Sensitivity

on the Containment

Heat Removal During Recirculation (Post-RAS)," Revision 1. The team verified that the input assumptions

relating to the CCW system for the above analyses were correct. The above LOCA analyses demonstrated

that the heat exchangers

could remove sufficient

heat from containment

following

a LOCA to keep the containment

pressure and 1

  • ----------------------

ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS temperature

within the design limits. In each case, the analysis documented

a CCW temperature

exiting the shutdown coolers exceeding

the system design temperature

of 140 degrees Fahrenheit

(140 °F) as stated in FSAR table 9-6 and DBD 1.01, "Component

Cooling Water," Revision 3. The team noted that the licensee accepted the maximum CCW temperature

that resulted from the scenarios

analyzed in EA-D-PAL-207-01

and EA-D-PAL-93-272-03

by Corrective

Action D-PAL-93-272G, based primarily

on an evaluation

of the effects on pipe stress. However, the licensee had not considered

the other negative effects, such as any detrimental

effects from elevated CCW temperature

on pump seals. Also, the licensee had not determined

the maximum possible CCW temperature

under worst case conditions

and had not identified

that a change to the FSAR could be required.

The team reviewed the latest LOCA analysis, EA-GEJ-96-01, and determined

that it documented

a CCW temperature

exiting the shutdown cooling heat exchanger

was 184 °F. The licensee determined

the system was operable under this condition

and issued Condition

Report (CR) C-PAL-97-1363F

to determine

the most limiting CCWtemperature

for any condition

and to evaluate all the effects resulting

from that limiting temperature

on the CCW system. ' It appeared that the requirements

of 10 CFR 50, Appendix B, Criterion

111, "Design Control," were not met in this case in that the design basis for the CCW system, as defined in 10 CFR 50.2, did not encompass

the entire range of bounding temperatures.

The team identified

this item as Unresolved

Item 50-255197-201-02.

Palisades

60 Day Response:

Prior to the Design lnspection;.we

determined

that the CCW system is operable at a predicted

maximum system temperature

of 184°F. The CCW system will be analyzed to confirm the most limiting temperature

for any design basis condition, and to determine

the effects of this temperature

on system components

by October 1; 1998. The FSAR will be updated as appropriate.

The programmatic

design control aspects related to this issue will be addressed

as identified

in Attachment

B, Item 1. 10/1/98 Update: In June of 1998, Engineering

Analysis EA-LOCA-98-01

was performed

to determine

the limiting condition

CCW temperature.

The results show a maximum 180°F CCW temperature

out of the CCW heat exchanger.

The effects of this temperature

on system components

was then evaluated.

It was determined

that the CCW heat exchanger

outlet temperature

indication

range was too narrow and needed to be expanded to meet RG 1.91 requirements.

By December 15, 1998, these temperature

indicators

will be replaced and full compliance

with RG 1.97 requirements

will be achieved.

All other evaluated

CCW system component

peak temperature

ratings fall within the predicted

180°F temperature.

The FSAR was changed to clarify CCW system design temperature

and LOCA maximum temperatures.

The temperature

indicator

range issue (50-255/97201-02)

was identified

as open, and was the subject of a NOTICE OF DEVIATION

(50-255/98003-02), in NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION." Palisades

responded

with additional

information

to the NRC under correspondence

dated June 24, 1998, entitled "RESPONSE

TO NOTICE OF VIOLATION

AND NOTICE OF DEVIATION

FROM INSPECTION

REPORT 50-255/98003." 2

  • ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS Refer to Attachment

B, Item 1 for the programmatic "design control" aspects associated

with this issue. Unresolved

Item 50-255/97-201-03

The team reviewed C-PAL-96-1-63-01, "120 day response to GL 96-06, Assurance

of Equipment

Operability

and Containment

Integrity

during Design Basis Accident Condition," Revision 0, which was the licensee's

response to Nuclear Regulatory

Commission (NRG) Generic Letter 96-06, "Assurance

of Containment

Operability

and Containment

Integrity

During Design-Basis

Accident Conditions," and observed that the licensee took credit for relief valve RV-0939 to protect the CCW piping inside containment

from overpressurization

in the event of a LOCA. RV-0939 was not included in the /ST program. The team questioned

whether RV-0939 performed

a safety function and if it should have been included in the /ST program. The licensee issued CR C-PAL-97-1686

to evaluate this

discrepancy.

10 CFR 50.55a requires /ST in accordance

with ASME Section XI of valves that perform a safety function.

It appeared that the licensee did not fully implement

these requirements

for RV-0939. The team identified

this item as part of Unresolved

Item 50-255197-201-03.

Palisades

60 Day Response:

During the Design Inspection, it was determined

that sufficient

overpressure

protection

is provided for the CCW system without taking credit for relief valve RV-0939, and the CCW system is therefore

operable.

The CCW piping in containment

is not required during an accident and is classified

non-Q, safety related. As a result, the ISl/IST programs have classified

the CCW piping and related components, including

RV-0939, as non-class

and excluded the same from inspection/test

requirements

of the Code. The Palisades

response to GL 96-06 determined

acceptability

of systems by generally

taking credit for 1) steam/gas

service, 2) available

expansion

paths, or 3) relief valves as a means to provide *sufficient

protection

against thermally

induced over pressurization.

In the case of the CCW system, "available

relief valves" serves as the basis for acceptability.

Relief valve operation

is considered

important

but not a safety related function, and therefore, the classification

of the CCW system and its components

such as RV-0939 were not changed. Although RV-0939 is not in the IST program, it, along with RV-2108 and RV-0956, is inspected, maintained

and set point verified via

maintenance

activity PPAC CCS043 on a 10-year interval.

These are essentially

the same as the requirements

of the Code (ASME/ANSI

OM-1987, Part 1 ). Based on this evaluation, no further action is required.

RV-0939 is appropriately

classified, maintained

and tested. Our existing GL 96-06 submittal

is accurate.

3

  • ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS 10/1/98 Update: This response has not changed since the submittal

of our original 60-day inspection

report response.

Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-03)

was identified

as closed. No further actions on this item are planned. Unresolved

Item 50-255/97-201-04

FSAR Section 9.3.2.3 stated that the CCW pipingwithin

containment

was not vulnerable

to failure caused by a high energy line break (HELB) and referred to Deviation

Report (DR) D-PAL-89-061, "Post Accident Operation

of CCW System, 11 dated March 23, 1989, for the evaluation.

This DR referred to Engineering

Analysis (EA) EA GW0-7793-01, "CCW Piping Inside Containment

HELBA," Revision 0. This EA was reviewed by the team, and it concluded

that the CCW piping inside containment

was not affected by HELBs, but did not contain the analysis performed

or a reference

to the analysis.

The EA contained

an outline of the methodology, listed the drawings and walkdowns

used, and referenced

the source of the postulated

HELBs. Palisades

Administrative

Procedure

No. 9.11, "Engineering

Analysis, 11 Revision 9, stated that an EA shall present an argument which substantiates

the conclusion

of the EA. The EA also contained

an error in the identification

of the Systematic

Evaluation

Program (SEP) topic number for evaluation

of the effects of internally

generated

missiles.

The licensee initiated

Engineering

Assistance

Request (EAR) EAR-97-0632

to revise EA-GW0-7793-01.

During the inspection, the licensee issued Revision 1 of EA-GW0-7793-01, which included a discussion

of the walkdown analysis used and corrected

the SEP references.

This revised EA was acceptable

to the team. It appeared that the requirements

of 10 CFR Part 50, Appendix B, Criterion . Ill, "Design Control," regarding

verifying

the adequacy of designs were not adhered to in this case. Also, the requirements

of the licensee's

Administrative

Procedure

9. 11 were not fully met in that EA-GW0-7793-01, Revision 0, did not contain full substantiation

of the conclusion.

The team identified

this item as Unresolved

Item 50-255197-201-04.

Palisades

60 Day Response:

As a remedial action, EA-GW0-7793-01

was revised to provide justification

for its conclusion

and to correct references

to related NRC corresponqence.

The related programmatic

design control and calculation

control aspects will be addressed

as identified

in Attachment

B, Items 1 and 2. 10/1/98 Update: This response has not changed since the submittal

of our original 60-day inspection

report response.

Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-04)

was identified

as closed. No further actions are planned for this item . Refer to Attachment

B, Item 1 for the programmatic "design control" aspects associated

with this issu.e. 4

ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS Unresolved

Item 50-255/97-201-05

The team reviewed the implementation

of the licensee's

commitment

to NRG Regulatory

Guide (RG) 1.97, "Instrumentation

for Light-Water-Cooled

Nuclear Power Plants To Assess Plant and Environs Conditions

During and Following

an Accident," Revision 3, as described

in FSAR Appendix 7C. The RG stated a range for CCW flow instrumentation

of 0-110 percent. Since

there was no instrument

to directly measure CCW flow, the licensee used a combination

of instruments, including

TE-0912 and TE-0913, which measure shutdown cooling heat exchanger

outlet temperature, to indicate flow. Use of instruments (other than flow indicators)

to monitor for CCW flow was determined

as acceptable

by the NRG (a letter from NRG to Consumers

Power Company, dated July 19, 1988, entitled "Palisades

Plant-Response to Generic Letter 82-33 Conformance

to Regulatory

Guide 1.97 "Instrumentation

for Light-Water-Cooled

Nuclear Power Plants To Assess Plant and Environs Conditions

During and Following

an Accident).

The required range for these TEs in FSAR Appendix 7C was 0-180 °F. This range did not encompass

the temperature

determined

in EA-GEJ-96-01, "A-PAL-94-324

Containment

Spray System (CSS) Sensitivity

on Containment

Heat Removal During Recirculation (Post-RAS)," Revision 1. This analysis determined

an outlet temperature

of the CCW from the shutdown cooling heat exchanger

of 184 °F. The licensee issued CR C-PAL-97-1363E

to evaluate the process instrumentation

and controls associated

with the CCW system for the effects of the higher temperature

predicted

by the analysis.

The licensee did not appear to meet their commitment

to NRG RG 1.97, "Instrumentation

for Light-Water-Cooled

Nuclear Power Plants To Assess Plant and Environs Conditions

During and Following

an Accident," in that the installed

CCW temperature

indicators

were not capable of monitoring

the full temperature

range expected to be observed in the CCW system. The team identified

this item as part of Unresolved

Item 50-255197-201-05.

Palisades

60 Day Response:

Prior to the Design Inspection, we determined

that the COW system is operable at a predicted

maximum system temperature

of 184°F. The CCW system will be analyzed to confirm the most limiting temperature

for any design basis condition, and the effects of this temperature

on system components.

In response to this specific issue, process instrumentation

and controls associated

with the CCW system will be reviewed to identify the impact of the maximum predicted

temperature.

This action will be completed

by October 1, 1998. 10/1/98 Update: Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-05)

was identified

as closed. This item was also the subject of a NOTICE OF DEVIATION

(50-255/98003-02)

from the same letter. Palisades

responded

with additional

information

to the NRC under correspondence

dated June 24, 1998, entitled "RESPONSE

TO NOTICE OF VIOLATION

AND NOTICE OF DEVIATION

FROM INSPECTION

REPORT 50-255/98003." In summary, the range of the CCW heat exchanger

outlet temperature

indicators

will be changed to meet RG 1.97 requirements

by December 15, 1998. 5

  • ** * ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS Unresolved

Item 50-255/97-201-06

The team identified

a lack of closure verification

testing on SI system check valves that could potentially

result in an overpressure

condition

affecting

the low-pressure

piping on the suction of the HPSI pumps. The minimum flow recirculation

lines associated

with the two HPSI pumps and the two LPSI 'pumps were interconnected

upstream of the air-operated

minimum flow recirculation

isolation

valves. In the event that only one HPSI pump was operating

under post-accident

conditions

with the minimum flow recirculation

isolation

valves closed, back leakage through the minimum flow piping associated

with the idle HPS/ pump could over pressurize

the idle HPS/pump suction piping. Backflow between the HPS/ minimum flow lines should be prevented

by check valves CK-ES3339

or CK-ES3331, and CK-ES3340

or CK-ES3332.

However, EGAD-EP-01, "lnservice

Testing Program-Valve

Test Program," Revision 10 indicated

that closure verification

testing of these check valves was not included in the /ST program. *The team asked the licensee if closure of these check valves was considered

a safety function requiring

/ST. The licensee initiated

CR C-PAL-97-1660

to evaluate the testing requirements

of these check valves. On November 10, 1997, the operability

determination

concluded

that these system check valves had not been subject to closure verification

testing as required, and both HPSI pumps were declared inoperable.

In accordance

with TS Section 3.0.3, 3.3, and 4.0.3, the licensee entered a Limiting Condition

for Operation (LCO) action statement, performed

closure verification

testing of check valves CK-ES3339

and CK-ES3340, and verified the operability

of these valves. The licensee stated that closure verification

testing of these check valves would be added to the /ST program. The team also identified

a lack of closure verification

testing on SI system valves that could potentially

result in a Safety Injection

Tank (SIT) being degraded under post-accident

conditions.

The normally closed SIT vent valves, CV-3051, 3063, 3065, and 3067, could be opened in accordance

with SOP-3, "Safety Injection

and Shutdown Cooling System," Revision 28, to reduce SIT pressure.

SOP-3 did not require the affected SIT to be declared inoperable

when a vent was opened. When a vent valve was opened the SIT pressure boundary (250 psig design pressure)

was exposed to the SIT vent header piping (100 psig design pressure).

SOP-3 did not include d(rections

to isolate an open vent valve in the event of an accident.

EGAD-EP-01, lnservice

Testing Program -Valve Test Program," Revision 10, indicated

that closure verification

testing of these valves was not included in the /ST program. The team asked the licensee if the failure of a valve to close could result in a SIT being degraded under accident conditions, and if closure of these valves was considered

a safety function requiring

/ST testing. The licensee initiated

CR C-PAL-97-1592

to evaluate this item and placed caution tags on the control room switches for vent valves CV-3051, 3063, 3065, and 3067 to prevent the valves from being opened without entering an LCO for the SITs. The licensee also stated that these valves had been opened rarely during plant operation.

1 O CFR 50. 55a requires in-service

inspection

in accordance

with Section XI of the ASME Boiler and Pressure Vessel Code. This code requires testing of valves which perform a safety function.

It appeared that the licensee did not implement

these requirements

with regard to valves CK-ES3339, CK-ES3340, CV-3051, CV-3063, CV-3065, and CV-3067. The team identified

this item as part of Unresolved

Item 50-=255197-201-06.

6

ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS Palisades

60 Day Response:

During the Design Inspection, high pressure safety injection

pump minimum flow recirculation

line check valves CK-ES3339

and CK-ES3340

were tested and the HPSI system was declared operable.

Action to include check valves CK-ES3339

and CK-ES3340

in the IST Program will be completed

by July 15, 1998. lri the interim, the check valves are tested to meet quarterly

testing requirements.

During the Design Inspection, the Safety Injection

Tank (SIT) vent valves CV-3051, CV-3063, CV-3065 and CV-3067 were closed and cautioned

tagged with the tanks declared operable.

Action to revise operating

procedures

to address opening the SIT vent valves will be completed

prior to removal of the caution tags. Prior to March 15, 1998, a representative

sample of check valves, AOVs and MOVs will be reviewed and verified to be incorporated

in the IST program as required.

10/1/98 Update: Check valves CK-ES3339

and CK-ES3340

have been included in the IST Program. Operating

procedures

have been revised to address opening of the SIT vent valves CV-3051, CV-3063, CV-3065 and CV-3067 and caution tags have been removed. A representative

sample of check valves, AOVs and MOVs have been sampled to determine

if they are included in the IST Program as required.

The sampling identified

additional

AOVs and one check valve that required inclusion

into the IST Program. These valves have been incorporated

into the IST Program and have been tested to confirm their safety related function.

In addition, several other actions associated

with the IST Program are underway to enhance databases, review ISi Program bases for IST Program impact, and revise IST Program and bases to enhance purpose, scope and program descriptions.

These actions are projected

to be complete by . May 1, 1999. Presently, Palisades

is in full c_ompliance

with the ISi and IST program requirements.

Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-06)

was identified

as closed. .This item was also the subject of a NOTICE OF VIOLATION

(50-255/98003-03)

from the same letter. Palisades

responded

with additional

information

to the NRC under correspondence

dated June 24, 1998, entitled "RESPONSE

TO NOTICE OF VIOLATION

AND NOTICE OF DEVIATION

FROM INSPECTION

REP.ORT 50-255/98003." Unresolved

Item 50-255/97-201-07

The team reviewed the HVAC system serving the cable spreading

room. The team observed that DR F-CG-91-072

was prepared in May 1991 when it was discovered

that the assumptions

in calculation

EA-FC-573-2, "Calculated

Required Air Flow for Inverter/Charger

Cabinet Cooling Fan," dated October 3, 1982, used an ambient temperature

of 94 °F instead of the correct design basis temperature

of 104 °F. The Safety System Design Confirmation (SSDC) Team that found this discrepancy

recommended

that the EA be updated. Procedure*9.11, "Engineering

Analysis," Revision 9, required all EAs to be revised if analytical

inputs or major assumptions

change. The 7

ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS licensee aec1dedtiotl6

reVisetfie

EA-; and ffie alscrepaiicy

was recorded in DBD 4.02 (125-V de system) and DBD 4.03 (preferred

ac system). The fans were installed

in 1983 and were not safety related. DR F-CG-91-072

was closed in October 1994, when the decision was made not to revise the calculation.

The licensee stated that specifications

were being developed

for replacing

the inverters

and chargers during the time the discrepancy

was being evaluated

and that this knowledge

contributed

to the decision not to update the EA. The inverters

and chargers were scheduled

to be replaced in the near future by Specification

Change (SC) SC-96-033.

The new equipment

would have internal cooling fans designed for a 104 °F maximum ambient and SC-96-033

would supersede

EA-FC-573-2

upon installation.

The team had no other concerns about the cable spreading

room HVAC system. It appeared that the requirements

of 10 CFR Part 50, Appendix B, Criterion/I, "Quality Assurance

Program," were not followed in this case in that the requirements

of Procedure

9. 11 regarding

revising EAs were not fully implemented.

The team identified

this item as part of Unresolved

Item 50-255197-201-07.

Palisades

60 Day Response:

Prior to the Design Inspection, Design Basis Documents

were revised to address this discrepancy.

Analysis EA-FC-573-2

will be revised or superseded

by December 1, 1998. The calculation

control aspects related to this issue (in this case, the revision of all analyses whenever analytical

inputs or major assumptions

change) will be addressed

by the action described

in Attachment

B, Item 2. 10/1/98 Update: The schedule for resolving

remains as stated above. Per NRG correspondence

dated May 18, 1998, titled "NRG INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-07)

was identified

as closed. This item was also the subject of a NOTICE OF VIOLATION

(50-255/98003-04)

from the same letter. Palisades

responded

with additional

information

to the NRC under correspondence

dated June 24, 1998, entitled "RESPONSE

TO NOTICE OF VIOLATION

AND NOTICE OF DEVIATION

FROM INSPECTION

REPORT 50-255/98003." Unresolved

Item 50-255/97-201-08

The team identified

the following

discrepancies

in SJ system mechanical

calculations:

  • EA-DBD-2.01-004, "Electrical

and Mechanical

Failure Analysis for the Low Pressure Safety Injection

System," Revision 0, pages 10 and 25, identified

a situation

in which a Joss of an emergency

diesel generator (EOG) during a large-break

LOCA would result in only one LPSI pump and two LPS/ injection

valves being operable.

The EA stated: "The acceptability

of this situation

could not be verified." The team asked if this statement

was correct. The licensee replied that the statement

was not current, and that the statement

appeared to be based on superseded

calculation

ANF-88-107, "Palisades

Large Break LOCA/ECCS

Analysis With Increased

Radial Peaking," Revision 1. Calculation

ANF-88-107

was superseded

by Seimens calculation

EMF-96-172, "Palisades

Large Break LOCA/ECCS

Analysis," Revision 0. The licensee initiated

Engineering

Assistance

Request (EAR) 97-0635 to revise EA-DBD-2.01-004.

8

ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS * EA-A-NL-92-185-01, "Worst Case Operating

Conditions

for the LPSllSDC System MOVs," Revision 1, addressed

the most limiting conditions

under which the system motor-operated

valves (MOVs) were required to open and close. This analysis included MOVs M0-3015 and M0-3016. These valves were the isolation

valves installed

in the shutdown cooling inlet . piping from primary coolant system (PCS) loop 2. For all normal operations

-other than shutdown cooling being in service, -the valves were electrically

locked closed. Page 19 of EA-A-NL-92-185-01

stated that the scenario that could produce the most limiting differential

pressure was that these valves would be required to close in the event of a downstream

pipe break. The EA addressed

a potential

12-in. downstream

pipe break and determined

that complete depressurization

and blowdown of the PCS to the hot-leg elevation

would occur before operators

could enter the EOPs and attempt to isolate the break. Therefore, the analysis then established

a maximum flow rate of 4120 gpm through valves M0-3015 and M0-3016, based on a normal system flow rate of 3000 gpm and a calculated

leakage of 1120 gpm through a break of a 1-112-inch

branch line downstream

of the valves. The team asked the licensee to provide the basis of the postulated

1-112-inch

branch line failure, since it did not appear to be consistent

with the postulated

pipe crack used in the internal flooding analysis of the safeguards

areas (EA-C-PAL-95-1526-01, "Internal

Flooding Evaluation

for Plant Areas Outside of Containment," Revision 0). The licensee verified that the flooding analysis break flow was different

and that this difference

would not affect the conclusions

of EA-A-NL-92-185-01.

Assumptions

5.9 and 5.10 of EA-A-NL-92-185-01

stated that the HPS/ and LPSI injection

flows to the loops were approximately

equal under post-accident

conditions.

These assumptions

did not appear consistent

with the flow values calculated

in EA-SDW-95-001, "Generation

of Minimum and Maximum HPSllLPSI

System Performance

Curves Using Pipe-Flo," Revision 2. The team asked the licensee to provide the bases of these values. The licensee stated that the values were not current and verified that the difference

between these values and the current values would not affect the EA results. The licensee initiated

CR C-PAL-97-1670

to resolve the discrepancies

in EA-A-NL-92-185-01.

  • EA-E-PAL-93-004E-01, "/ST Check Valve Minimum Flow Rate Requirements

to Support Chapter 14 Events," Revision 0, identified

1601 gpm as the required test flow for the LPS/ injection

check valves. The team observed that this value appeared to be less limiting than the values calculated

in EA-SDW-95-001, "Generation

of Minimum and Maximum HPS/ILPSI

System Performance

Curves Using Pipe-Flo," Revision 2. The licensee initiated

CR C-PAL-97-1603

to address this discrepancy.

The licensee determined

that the LPSI test flow presented

in EA-E-PAL-93-004E-01

was less than the current calculated

requirement.

However, the actual LPSI check valve flow acceptance

criterion

in /ST Procedure

Q0-88, "ESS Check Valve Operability

Test (Cold Shutdown)," Revision 17, was verified to be 1690 gpm, which was greater than the current calculated

requirement.

The licensee stated that the affected documentation

will be corrected.

Administrative

Procedure

9. 11, "Engineering

Analysis," Revision 9, Section 6. 1. 5. c stated that an analysis shall be revised if analytical

inputs changed. In the above instances, engineering

analyses were not updated to reflect analytical

input change. The licensee initiated

C-PAL-97-1636

to evaluate the overall issue of calculation

control. The team identified

this item as part of Unresolved

Item 50-255197-201-08.

9

ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS Palisades

60 Day Response:

During the Design Inspection, it was determined

that the LPSI check valves are operable since IST acceptance

criteria and actual test flow rates exceeded the minimum required flow rates in analysis EMF-96-72

which had superseded

EA-E-PAL-93-004E-01.

By June 1, 1998, engineering

guideline

EGAD-EP-09

and IST procedure

Q0-8B basis document will be revised to assure that the increased

minimum design flow requirement

is met, and that design bases agree with IST acceptance

criteria.

Remedial actions to revise EA-DBD-2.01-004

to accurately

reflect electrical

system response to events will be completed

by August 15, 1998. EA-A-NL-92-185-01

and EA-SDW-95-001

are bounding analyses which will not be required to be revised or superseded.

Specifically, * the calculation

control process will be revised to allow bounding analyses to remain unchanged

when revisions

to inputs or assumptions

do not affect the analysis conclusions.

The calculation

control aspects related to this issue will be addressed

by the action described

in Attachment

B, Item 2. 10/1/98 Update: Engineering

guideline

EGAD-EP-09, IST procedure

Q0-8B Basis Document, and engineering

analysis EA-DBD-2.01-004

were revised as stated above. Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item

(50-255/97201-08)

was identified

as closed. This item was also the subject of a NOTICE OF VIOLATION

(50-255/98003-04)

from the same letter. Palisades

responded

with additional

information

to the NRC under correspondence

dated June 24, 1998, entitled "RESPONSE

TO NOTICE OF VIOLATION

AND NOTICE OF DEVIATION

FROM INSPECTION

REPORT 50-255/98003." Unresolved

Item 50-255/97-201-09

During an SI system walkdown on October 6, 1997, the team observed scaffolding

installed

adjacent to the SIRWT on the roof of the auxiliary

building.

The team questioned

how the installation

of scaffolding

in the vicinity of safety-related

equipment

was controlled

to prevent damage to the safety-related

equipment

during a seismic event. The licensee provided Procedure

MSM-M-43, "Scaffolding," Revision 2, for the team's review. Section 5. 3 of this procedure

required an engineering

review of scaffolding

installed

in the vicinity of safety related equipment.

However, the licensee determined

that the scaffolding

observed during the walkdown had not received engineering

review in accordance

with the procedure.

The licensee initiated

CR C-PAL-97-1417

to address the scaffolding

installation, and the scaffolding

was removed on October 8, 1997. EA-C-PAL-97-1417A-01, "Operability

Reassessment

of SIRWT Scaffolding," Revision 0, was completed

during the inspection.

Based on a structural

analysis of the maximum loading on the SIRWT due to seismic interaction

with the scaffolding

during a safe shutdown earthquake, this analysis concluded

that the SIRWT was not inoperable

due to this nonconforming

condition.

10

  • * ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS During another SI system walkdown on October 30, 1997, the team observed additional

scaffolding

installed

in the east ESG room adjacent to safety-related

piping. An evaluation

by the licensee determined

that this scaffolding

had not been installed

in accordance

with Procedure

MSM-M-43, "Scaffolding," Revision 2. The licensee initiated

CR C-PAL-97-1585

to address this scaffolding

installation

and, based on a visual inspection, concluded

that this nonconforming

scaffolding

would not render any safety-related

piping or components

inoperable.

The licensee removed the scaffolding.

In addition, the licensee performed

a walkdown of all plant scaffolding

during the inspection

and verified that there were no additional

nonconforming

conditions.

The licensee stated that all scaffolding

erections

would cease until appropriate

personnel

underwent

remedial training.

The team observed the following

three separate conditions

in the west ESG room involving

potential

seismic interactions

with safety-related

equipment.

The team noted that, during a seismic event, unrestrained

items could potentially

damage safety-related

piping and equipment.

The safety-related

piping and equipment

in the west ESG room were required for operation

of the HPSI, LPSI, and containment

spray systems in the event of an accident.

  • The team observed an unsecured

operations

storage cabinet located adjacent to safety-related

piping and valves. The team asked the licensee if the condition

was in accordance

with plant procedures.

The licensee initiated

CR C-PAL-97-1587, which determined

that the cabinet was not placed in accordance

with the spacing requirements

of Administrative

Procedure

1.01, "Material

Condition

Standards

and Housekeeping

Responsibilities," Revision 11. The operability

evaluation

concluded

that the nonconforming

condition

did not result in any safety-related

equipment

being inoperable.

The cabinet was laid on its side to eliminate

the toppling concern. The licensee stated that the cabinet would be removed from the area. * The team observed an* unsecured

chainfall

located adjacent to and above the shutdown cooling heat exchangers.

A similar chainfall

in the east ESG room was secured. The team asked the licensee if the condition

was in accordance

with plant procedures.

The licensee determined

that the chainfall

location was not in accordance

with Administrative

Procedure

1.01, and initiated

CR C-PAL 97-1586. The operability

evaluation

concluded

that the nonconforming

condition

did not result in any safety-related

equipment

being inoperable.

The licensee stated that the chainfall

chains would be moved away from the heat exchanger.

  • The team observed a ladder in the west ESG room that appeared to be improperly

stored. The ladder was lying on the floor under the installed

ladder rack. The team asked the licensee if the condition

was in accordance

with plant procedures.

The licensee initiated

CR C-PAL-97-1601

and determined

that the ladder location was not in accordance

with the "Palisades

Ladder Control Policy for Operating

Spaces," dated May 14, 1997. The CR concluded

that, although the ladder storage did not meet the ladder control policy, the nonconforming

condition

did not result in any safety-related

equipment

being inoperable.

The licensee stated that the ladder was removed from the area. Procedure

MSM-M-43 required an engineering

review of scaffolding

installed

in the vicinity of safety-related

equipment.

Procedure

1. 01 and the "Palisades

Ladder Control Policy for Operating

Spaces," dated May 14, 1997, contain requirements

for storing items in the vicinity of safety-related

equipment.

In these cases, the licensee did not comply with the procedural

requirements

for activities

affecting

quality as required by 1 O CFR Part 50, Appendix B, Criterion

V, "Instructions, 11

ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS Procedures, and Drawings." The team identified

this item as Unresolved

Item 50-255197-201-09.

Palisades

60 Day Response:

Remedial actions consisted

of dispositioning

all scaffolding

and unrestrained

items near the SIRW Tank and in the East and West Safeguards

Rooms to assure operability

of safety-related

equipment.*

Subsequently, walkdowns

were conducted

in other areas containing

safety-related

equipment

and no conditions

similar to the scaffolding

conditions

identified

in this open item were observed.

Maintenance

and construction

crews were briefed on the lessons learned pertaining

to scaffolding

erection.

By July 15, 1998, we will revise procedures, provide training and reinforce

management

expectations

as necessary

to maintain compliance

with seismic interaction

requirements

for related equipment.

10/1/98 Update: Specific actions to revise procedures, provide training and reinforce

management

expectations

as necessary

to maintain compliance

with seismic interaction

requirements

for safety-related

equipment

have been completed.

Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003

  • (DRS) AND NOTICE OF VIOLATION", this item

(50-255/97201-09)

was identified

as closed. This item was also the subject of NOTICES OF VIOLATION

(50-255/98003-05

and 50-255/98003-06)

from the same letter. Palisades

responded

with additional

information

to the NRG under correspondence

dated June 24, 1998, entitled "RESPONSE

TO NOTICE OF VIOLATION

AND NOTICE OF DEVIATION

FROM INSPECTION

REPORT 50-255/98003." This response is associated

with plans to enhance maintenance

personnel

scaffolding

training, and provide training for Auxiliary

Operators

to recognize

unrestrained

items for prompt identification.

Training will be completed

by March 1, 1999. Unresolved

Item 50-255/97-201-10

During the surrogate

tour, the team obseNed the ends of two vent pipes that connected

the containment

sump to the 590-ft elevation

of the containment.

The team asked the licensee to explain the design of these vent lines. During a review of the vent lines, the licensee determined

that the top of the vents were located inside the containment

at an elevation

of approximately

595-ft. The maximum calculated

post-accident

water elevation

was at elevation

597-ft. The vent pipes did not have screens on their inlets. The licensee also determined

that the two vent lines entered the containment

sump inside the sump screens, creating a potential

path for debris to enter the EGGS pump suction piping under post-accident

conditions.

The licensee initiated

CR C-PAL-97-1571, on October 29, 1997, to evaluate this condition

and determined

that the postulated

type and quantity of debris that could enter the vent pipes under post-accident

conditions

would not prevent the SI and containment

spray systems from performing

their safety function, and that these systems were operable under this condition.

The licensee also installed

Temporary

Modification

TM-97-046, on October 29, 1997, to add screens to the top of the vent pipes during the inspection.

These screens would prevent debris from entering the EGGS pump suctions in the event of an accident.

12

  • ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS It appeared that the requirements

of 10 CFR Part 50, Appendix B, Criterion

Ill, "Design Control," were not met in this instance in that the design basis of the containment

sump to exclude debris from the EGGS pump suction piping was not fully implemented.

The team identified

this item as part of Unresolved

Item 50-255197-201-10.

Palisades

60 Day Response:

As stated above, an operability

determination

concluded

the Engineered

Safeguards

Systems were operable in the as-found condition.

As additional

assurance

for continued

operability, temporary

screens were placed over the vent pipes. These screens will be permanently

installed

in the 1998 refueling

outage. The programmatic

design control aspects related to this issue will be addressed

as identified

in Attachment

B, Item 1. 10/1/98 Update: Containment

sump vent screens were permanently

installed

during the 1998 refueling

outage. Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-10)

was identified

as closed. This item was also the subject of a NOTICE OF VIOLATION

(50-255/98003-0?a)

from the same letter. Palisades

responded

with additional

information

to the NRC under correspondence

dated June 24, 1998, entitled "RESPONSE

TO NOTICE OF VIOLATION

AND.NOTICE

OF DEVIATION

FROM INSPECTION

REPORT 50-255/98003." As part of our annual design basis document update projected

for June 1999, the Containment

Spray Design Basis Document DBD-2.03 will be revised to address issues vital to the function of the Engineering

Safety Features following

a LOCA. Refer to Attachment

B, Item 1 for the programmatic "design control" aspects associated

with this issue. Inspection

Followup Item 50-255/97-201-11

The team also observed several piping penetrations

between the east and west ESG rooms which included rubber piping expansion

joints used as penetration

seals. The team questioned

the design of these piping penetration

seals. The licensee stated that the engineering

analyses that demonstrated

that these penetrations

met the design basis did not-specifically

address the use of rubber piping expansion

joints in the penetration

seals. The team reviewed EA-RJC-92-0508, * Analysis

of the Effect of a Fire on the Fire Barrier Penetration

Seal Number FZ-0508," Revision 0, and verified that the rubber piping expansion

joints were not addressed.

The licensee initiated

CR C-PAL-97-1627

and determined

that the failure to specifically

justify the presence of rubber expansion

joints did not invalidate

the conclusions

of the original engineering

analyses and that the penetration

seals were adequate.

The licensee also stated that the affected documentation

would be corrected, and that an "extent of condition" review would be performed.

The team identified

this item as Inspection

Fo/lowup Item 50-255197-201-11.

13

  • * ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS Palisades

60 Day Response:

An operability

determination

during the Design Inspection

concluded

that the safety function provided by the fire barriers separating

the East and West Safeguards

Rooms is not affected by the use of rubber expansion

pipe joints. By August 1, 1998, we will revise the design basis engineering

analysis to formally justify the installed

rubber expansion

pipe joints, and perform an investigation

of other area fire barriers for potential

unanalyzed

designs. 1011198 Update: Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-11)

was identified

as closed. The revision to the design basis engineering

analysis for rubber expansion

pipe joints is complete along with investigations

for other fire barriers for potential

unanalyzed

designs. No other unanalyzed

fire barrier design issues were discovered.

No further actions are planned for this inspection

item. Inspection

Followup Item 50-255197-201-12

The team reviewed 10 SI system calculations

and 1 pressurizer

pressure uncertainty

calculation;

these were identified

as "basis documents." Basis Document Rl-38, "SIRW Tank Level Instrument

Calibration," Revision 6, was reviewed for adequacy.

It provided the basis for calibration

of SIRWT level indicators

LT-0332A *and LT-0332B to enable their use to monitor the TS requirement

that the tank contain at least 250, 000 gallons of borated water. Rl-38 used a tank boron concentration

of 1720 parts per million (ppm) and did not consider the range of 1720 to 2500 ppm allowed by TS Section 3.3. Rl-38 was the basis document for the calibration

of the level indicator

that supported

manual actuation

of post-accident

recirculation

operation.

The team was concerned

that the increased

density of the tank water at higher boron concentrations

would increase the instrument

uncertainty.

The calculation

also did not account for variation

in boron concentration

density caused by temperature

changes; an effect which could also affect the total uncertainty.

The licensee recalculated

the total instrument

uncertainty

using the most conservative

boron concentrations

and temperature, and the *resulting

change to the total uncertainty

remained bounded by the original uncertainty

value. Bases Document Rl-69, "Subcooled

Margin Monitor Surveillance," Revision 6, was reviewed for adequacy.

The subcooled

margin monitor (SMM) provided the operator indication

of the PCS margin to .saturation

conditions.

Rl-69 evaluated

possible errors induced in the SMM. The team found that Rl-69 did not account for seismic uncertainty.

This was inconsistent

with RG 1.97 "Instrumentation

for Light-Water-Cooled

Nuclear Power Plants To Assess Plant and Environs Conditions

During and Following

an Accident," May 1983. This RG identifies

subcooled

margin as a Category/, Type A variable, which must continue to read within the required accuracy following, but not necessarily

during, a safe-shutdown

earthquake

event. The team was concerned

that the calculated

error was nonconservative

because it did not consider seismic uncertainty, and could provide misleading

information

to the operators.

The licensee reanalyzed

the potential

error in the SMM, including

seismic uncertainty, and the resulting

total uncertainty

remained bounded by the original uncertainty

value. The licensee assigned Procedure

Change Request (PCR) 5569 to revise Rl-69. 14

  • * ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS EA-RSW-94-001, "F/-0404 Instrumentation

Uncertainty

Calculation," Revision 2, was also reviewed for adequacy.

The analysis established

the recommended

uncertainties

of Fl-0404, which was used in flow testing of the SJ pumps. The instrument

was installed

in 1989, and has been calibrated

five times since then. Drift error was determined

using historical

calibration

data. For the first 4 years, the instrument

was calibrated

once a year. The team found that 24 months had transpired

between the fourth and fifth calibrations.

The licensee stated that the interval was

in 1993 from 11 months to 24 months. The team asked if the drift analysis was revised to account for this change in the calibration

interval.

The* team was concerned

that increasing

the calibration

interval to 24 months would increase the drift error and consequently

increase the total uncertainty

of the instrument.

The licensee reanalyzed

the Fl-0404 uncertainty

using appropriate

drift performance

data for the longer calibration

interval, and the resulting

change to the total uncertainty

remained bounded by the original uncertainty

value. The licensee issued EAR-97-0658

to revise EA-RSW-94-001.

The team also reviewed Basis Document Rl-15A, "Safety Injection

Tank Pressure Channel Calibration," Revision 7, for adequacy.

Rl-15A formed the bases for the pressure channel setpoints

for PIA-0363, 0367, 0369, and 0371, which defined low-and high-pressure

alarms for the S/Ts. The /ow-pressure

alarms warned the operators

of decreasing

nitrogen pressure in the tanks. The channel alarms were set to annunciate

earlier than the pressure limits of TS Section 3.3. 1 (b) so appropriate

action could be taken before pressure reached the setpoints

of pressure switches PS-03408, 03448, 03738, and 30508, which were set to alarm at the TS limits. The team was concerned

that Rl-15A did not consider uncertainties

such as stability

and temperature

effects and that the current total uncertainty

was not adequate.

Considering

the low alarm point of 207 psig, the calculated

uncertainty

allowance

of +/-6.85 psig could result in an alarm at close to 200 psig, which was the TS limit. If additional

uncertainties

were added, the channel pressure switches could alarm after the TS pressure switches.

The licensee reanalyzed

the setpoint for P/A-0363, 0367, 0369, and 0371 using additional

appropriate

uncertainty

inputs and determined

that the resulting

instrument

uncertainty

was bounded by Rl-15A. The team observed that the results of these basis documents

were determined

to encompass

specific additional

uncertainties

due to the assumed margins used in the documents

to account for unquantified

effects. The licensee had a guide entitled "Design & Maintenance

Guide on Instrument

Setpoint Methodology," EGAD-PROJ-16, Revision 0, and the team concluded

that it provided a satisfactory

methodology

for setpoint calculations

and was consistent

with industry standard S67-04, Part I, "Setpoints

for Nuclear Safety-Related

Instrumentation." The licensee stated that EGAD-PROJ-16

provided identical

guidance as EGAD-PROJ-08, Revision 0, of the same title, which was the current designation

of the guide. The instruments

that were re-analyzed

during the inspection

used the guidance of EGAD-PROJ-08.

This methodology

affirmed that margins remained bounded. The licensee stated that use of this guide was not required by plant procedures.

However, the licensee has previously

recognized

from past assessments

that its basis documents

were not as rigorous as required by the current /SA standards.

The licensee stated that EGAD-PROJ-08

was being revised and that the appropriate

procedures

would be revised to require its use. The team identified

this item as Inspection

Fol/owup Item 50-255197-201-12.

Palisades

60 Day Response:

None of the above calculational deficiencies

identified

during the Design Inspection

affected the operability

of any safety-related

equipment.

During the inspection, EGAD-ELEC-08

Rev 1 was approved and issued to provide

for instrument

setpoint methodology.

Our engineering

staff 15

  • ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS has been briefed as to the need to utilize this guidance.

Plant procedures

will be revised by August 15, 1998, to incorporate

EGAD-ELEC-08

for use when setpoint calculations

are required.

10/1/98 Update: Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-12)

was identified

as closed. Applicable

plant administrative

procedures

have been changed to reference

guidance document EGAD-ELEC-08

for use when performing

setpoint calculations, and enhanced to more clearly . describe the applicability

of EGAD-ELEC-08.

No further actions are planned

for this inspection

item. Unresolved

Item 50-255/97-201-13

During a walkdown of the SI system, the team observed that transmitters

for containment

spray flow, FT-0301 and FT-0302, and shutdown cooling heat exchanger

flow, FT-0306, were properly mounted below their flow elements, but the process tubing was observed to be inadequately

sloped back to the transmitters.

Additionally, a walkdown performed

by the licensee at the team's request during an * in-containment

inspection

revealed that the process lines to the HPSI cold-leg flow transmitters

FT-0308, FT-0310, FT-0312, and FT-0313, and the LPSI flow transmitters, FT-0307, FT-0309, FT-0311, and FT-0314, were also installed

with inadequate

slope. The team was concerned

that inadequate

slope in instrument

tubing could contribute

to significant

instrument

uncertainty

by entraining

unequal amounts of air in either leg of the transmitter, causing erroneous

readings.

This was shown to be a valid concern when an operator observed an erroneous

reading in the left channel containment

spray loop indicator, Fl-0301A.

The "below zero" reading was caused by air trapped in one of the process iines. The licensee issued CR C-PAL-97-1561

to vent the line. The lack of tubing slope was inconsistent

with original plant installation

specification

J-F020, Revision 0. This specification

stated: "Flow instruments (differential

tyP.e) in liquid and condensable

vapor service shall preferably

be mounted below the main line connection

so that the impulse lines will slope down to the instrument." The specification

also stated: "Impulse lines to flow instruments

shall slope (up or down) a minimum of one inch per foot." Plant drawings J-F133, Revision 1; * J-F134, Revision O; J-F140, Revision O; and J-F141, Revision 0, depict various acceptable

installation

configurations

for a differential

transmitter.

The current installations

of the flow instruments

identified

above were not consistent

with these drawings.

A later specification, J-465 (Q), "The Technical

Specification

for Installation

of Instrumentation

For Nuclear Service for CPCo Palisades," Revision 0, dated 1981 stated: "The installation

shall be neat in appearance, properly supported, and shall provide for proper slope for adequate drainage or venting of the instrument

lines." This specification

has since been incorporated

into specification

20557-J-59 (Q) under the same title, which requires that a "horizontal

tubing run is continually

sloped in accordance

with design drawings." The licensee issued CR C-PAL-97-1561

to evaluate these instrument

tubing sloping discrepancies.

According

to the operability

determination

of the CR, the instruments

have never shown any adverse effects of trapped air during the last 20 years of operation.

The HPSI and LPSI flow transmitters

were mounted as much as 8 ft above their flow elements.

To accommodate

instruments

mounted above flow elements, specification

J-F020 stated: "5 foot minimum "drop legs (equivalent

of a loop seal)" may be required before the tubing is sloped up the I 16

  • ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS meter." Plant drawings J-F152, Revision 1, and J-F153, Revision 0, depict these mounting configurations.

The licensee stated that the bottom and side tap locations

for the tubing would tend to limit the amount of air getting into the transmitters

and that air entrainment

would be minimal due to the ratio of the volume of the HPSI and LPSI pump suction piping to the tubing volume. EA-C-PAL-95-0877D, "Evaluation

of the Potential

for Excessive

Air Entrainment

Caused by Vortexing

SIRWT During a LOCA," Revision 0, evaluated

the potential

for excessive

air entrainment

in the lines of the pumps caused by vortexing

in the SIRWT during a LOCA, and determined

that the air f]ntrainment

would be a small percentage

of the flow volume. The licensee also stated that technicians

are required to vent the transmitters

during every 18 month surveillance.

However, the team was concerned

that, since the transmitters

sense low static pressure during normal standby operation, air may accumulate

between calibration

intervals

and between system tests. Additionally, the water circulated

through the SI lines from the containment

sump could contain significant

amounts of dissolved

gasses, which could enter the tubing up to the flow transmitters.

The team was concerned

that the effect of air entrapped

in the instrument

tubing could cause large and unquantifiable

errors in the flow indications.

EOP Supplement

4, "Loss of Coolant Accident Recovery Safety Function Status Check Sheet," contained

curves presenting

total SI flow ranges intended to help ensure that the minimum values utilized in the accident analyses (LOCA, MSLB, Steam Generator

Tube Rupture (SGTR)) were met. There was also a minimum total flow criterion

for the operators

to meet, which ensured the containment

sump check valves remained in a stable condition

in EOP-4. 0, "Loss of Coolant Accident Recovery," Revision 9. The operators

would use the HPSI and LPSI flow indication

from FT-0308, 0310, 0312, 0313, 0307, 0309, 0311, and 0314 to compare SI system performance

against the EOP requirements.

The team was concerned

that the potentially

large errors could confuse the operator and impair decision making. The licensee stated that the opetators

are trained to use all available

indications

and that alternate/additional

instrumentation

could be used to confirm trending of PCS conditions

such as that for pressurizer

level, subcooling

margin, reactor vessel level, and charging pump flows. The licensee issued EAR-97-0699

to evaluate this item. It appeared that the design basis for instrument

tubing installation

was not implemented

in the plant installation

as required by 10 CFR Part 50, Appendix B, Criterion

Ill, "Design Control." The team identified

this item as Unresolved

Item 50-255197-0201-13.

Palisades

60 Day Response:

During the Design Inspection, an operability

determination

was made concluding

the HPSI and LPSI flow indication

is operable based on plant operating

experience.

Since the inspection, a plant walkdown was conducted

which revealed that the HPSI and LPSI tubing configuration

met design requirements

but did not conform to associated

design drawings.

The existing tubing configurations

  • were observed, and the tubing was determined

not to be susceptible

to air entrainment.

The * conclusions

reached from this walkdown review further justify the reliability

of the HPSI and LPSI flow indication, although configuration

discrepancies

exist. By August 15, 1998, we will resolve the HPSl/LPSI

flow indication

tubing discrepancies

and compare our design requirements

to additional

samples of safety related instrument

tubing to identify any additional

nonconformances

with design criteria.

The programmatic

design control aspects related to this issue will be addressed

as identified

in Attachment

8, Item 1. 17

  • * * ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS 10/1/98 Update: Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item

(50-255/97201-13)

was identified

as closed. This item was also the subject of a NOTICE OF VIOLATION

(50-255/98003-07b)

from the same letter. Palisades

responded

with additional

information

to the NRC under correspondence

dated June 24, 1998, entitled "RESPONSE

TO NOTICE OF VIOLATION

AND NOTICE OF DEVIATION

FROM INSPECTION

REPORT 50-255/98003." Subsequent

to the Design Inspection, Palisades

walked down these installations

during the 98 refueling

outage and confirmed

that the sensing lines for HPSI and LPSI flow transmitters

FT-0308, 0310, 0312, 0313, 0307, 0309, 0311 , and 0314 are appropriately

sloped -thus no deviations

from design requirements

exist. A sampling of other sensing lines associated

with safety-related

equipment

were also walked down and confirmed

to meet design requirements

for sensing line slope. NRC correspondence

dated August 3, 1998 rescinded

this cited potential

violation.

No further actions are planned for this inspection

item. Unresolved

Item 50-255/97-201-14

The team reviewed EA-ELEC-LDTAB-005, "Emergency

Diesel Generator

1-1 & 1-2 Steady State Loading," Revision 4, and verified that the analysis was consistent

with the design basis information

in the FSAR. All required accident loads for a LOCA and a LOOP were identified

and tabulated.

The electrical

loads exceeded the continuous

rating of the EOG during the first 32 minutes of operation

but were below the EOG maximum 2-hour rating. One of the inputs to this analysis was the electrical

toad estimate for LPSI pumps P-67 A and P-678. These electrical

load estimates

were based on the minimum hydraulic

LPS/ pump performance

used in EA-A-PAL-92-037, "Emergency

Diesel Generator

Loadings-First

Two.Hours," Revision 1, which determined

that LPSI pump flow would be* 3600 gpm. Although the LPS/ pump flow was conservative

for evaluating

LOCA mitigation, it was not conservative

for determining

the maximum load the EOG could experience

during a LOCA. The team determined

that the LPS/ pumps could pump 4500 gpm with one LPS/ pump discharging

into all four injection

loops as identified

in EA-SDW-95-001, "Generation

of Minimum and Maximum HPSllLPSI

System Performance

Curves Using Pipe-Flo," Revision

2. The team was concerned

that the licensee had not analyzed for the worst-case

electrical

load demand on the EDGs. Preliminary

evaluations

by the_ licensee using the correct maximum loads indicated

that the electrical

loading on one EOG could be higher than that determined

in EA-ELEC-LDTAB-005.

The licensee issued CR C-. PAL-97-1650

to review and correct all necessary

electrical

analyses and determined

the EDGs to be operable.

The team reviewed EA-ELEC-VOL

T-13, "Palisades

Loss of Coolant Accident With Off$ite Power Available," Revision 0, which evaluated

the ac voltage available

during normal operating, refueling, and accident conditions.

The team noted that the calculation

had not been revised since 1993 and . that the load magnitudes

identified

in EA-ELEC-LDTAB-005, Revision 4, and EA-SDW-95-001, Revision 2, had not been included.

The licensee reviewed the impact of the revised loads on EA-ELEC-VOL

T-13 and determined

that the changes had minimal effect on the analysis.

The team also noted that FSAR Section 8.3 stated that backfeeding

via the main and station power transformers

could be utilized;

however, EA-ELEC-VOL

T-13 had not analyzed this particular

operating

mode. The licensee stated that it had recognized

that an analysis for backfeeding

needed to be performed

in 1994 and had issued AIR A-PAL-94-223

to create an analysis in order to bound 18

  • * ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS this condition

of operation.

The licensee initiated

C-PAL-97-1619

to review and update EA-ELEC-VOLT-13

for load changes. It appeared that the requirements

of10 CFR Part 50, Appendix B, Criterion

Ill, "Design Control," had not been met for EA-ELEC-LDTAB-005

an*d EA-ELEC-VOLT-13

in that the design basis had not been updated to document the actual plant parameters.

The team identified

this item as part of Unresolved

Item 50-255197-201-14.

Palisades

60 Day Response:

During the Design Inspection, an operability

determination

was made which concluded, based on an evaluation

which bounded recent load changes, that the electrical

system is operable.

Mechanical

flow model analyses, which serve as input to the electrical

load flow analyses, will be completed

by December 15, 1998. The electrical

load flow analyses, which will assure plant loads are accounted

for and applicable

operating

scenarios

are addressed, will be completed

by August 15, 1999. A specific backfeed analysis will be completed

by Januar}t 15, 1999. The programmatic

design control aspects related to this issue will be addressed

as identified

in Attachment

8, Item 1. 1011/98 Update: Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", th.is item (50-255/97201-14)

was identified

as closed. The mechanical

and subsequent

electrical

flow model analyses are on target for completion

by December 15, 1998 and August 15, 1999, respectively, as stated above. Backfeed analysis EA-ELEC-FL

T-009, "GSU Short Circuit Analysis" was completed

with design attributes

captured in the applicable

Design Basis Document.

Refer to Attachment

8, Item 1 for the programmatic "design control" aspects associated

with this issue. * Inspection

Followup Item 50-255/97-201-15

FSAR Section 8.5.2 stated that cables would be sized in accordance

with the National Electric Code (NEC) or Insulated

Power Cable Engineers

Association

(/PCEAllCEA)

ampacity values and the cable ampacities

would be adjusted on the basis of actual field conditions

when possible.

The adjustments

included conductor

operating

temperature, ambient temperature, cable overall diameter, raceway fill, and fire stops. The licensee had recently initiated

a program to verify the adequacy of its cable ampacity sizing. EA-ELEC-AMP-032, "Ampacity

Evaluation

for Open Air Cable Trays With a Percent Fill Greater Than 30% of the Usable Cross Sectional

Area," Revision 1, was issued in 1997 to address cable sizing. While reviewing

the EA, the team noted the absence of fire stop derating and increased

cable temperatures

due to thermal radiation

from hot pipes. The licensee had initiated

AIR A-PAL-97-062

to evaluate the effects of local heat sources on fire stops; however, evaluation

of the effects on cable degradation

due to the close proximity

of hot piping systems had not been included.

The licensee stated that evaluation

of the effects of hot piping would be included under A-PAL-97-062.

The team identified

this item as Inspection

Followup Item 50-255197-201-15 . 19

  • * ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS Palisades

60 Day Response:

We will complete our Cable Ampacity Sizing Program by September

15, 1998 which will identify any cable degradation

due to the close proximity

of hot piping, and any degradation

of fire stops due to local heat sources. 10/1/98 Update: Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-15)

was identified

as open. Cable * degradation

due to the close proximity

of hot piping, and any degradation

of fire stops due to local heat sources has been evaluated.

Results confirm that the cable design is acceptable.

No further actions are planned for this inspection

item. Unresolved

Item 50-255/97-201-16

The 120-V ac safety-related

and non-safety-related

loads were powered from instrument

ac bus Y-01. Bus Y-01 was powered from either motor control center (MCC) 1or2 via automatic

transfer switch Y-50. MCCs 1 and 2 were redundant

safety-related

busses. The licensee stated in a January 24, 1978, letter to the NRG that it would. implement

the recommendation

of RG 1. 6 in that no . provision

would exist for automatically

transferring

loads between redundant

power sources. The NRG issued a safety evaluation

report, dated April 7, 1978, confirming

the licensee's

commitment.

FC-364, "Feeder Change for Instrument

Bus Y-01," Revision 0, implemented

this commitment

and powered bus Y-01 from MCC 1 and non-safety-related

MCC 3. However, FC-854, "Y-01 Power Supply Feed Modification," Re.vision

0, moved the backup power source from MCC 3 to the safety-related

MCC 2, and resulted in a departure

from the plant's licensing

basis. The modification

installed

fuses in series with the existing breakers, which provided an additional

level of protection

for the two safety-related

busses. The team observed that the safety evaluation

performed

for FC-854 did not identify that prior NRC approval was required.

The licensee issued CR C-PAL-97-1678

to document this deviation

from the licensing

basis. It appeared that this modification

was a USO in that the possibility

of a common-mode

failure of the redundant

safety-related

busses was created, which was not previously

evaluated

in the FSAR and, thus, the criterion

of 10 CFR 50.59(a)(2)(ii)

was satisfied.

The team identified

this item as Unresolved

Item 50-255197-201-16.

Palisades

60 Day Response:

During the Design Inspection, an operability

determination

was completed

which concluded

that the implemented

design meets the intent of RG 1.6 and provides a single failure proof method of preventir:ig

the transfer of a fault between redundant

load sources. The current configuration

was implemented

under FC-854 with the modification

safety evaluation

concluding

that an unreviewed

safety question does not exist. Prior NRC approval of the change was not required.

A description

of the implemented

modification

was transmitted

to the NRC in our Annual Report of Facility Changes, Tests and Experiments

dated April 2, 1991. This 1989 modification

resulted in a change to a prior NRC commitment.

In accordance

with NEI guidelines, we will submit by November 1, 1998, a revised commitment

which reflects the existing plant configuration

and governing

design basis. 20

  • * ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN iNSPECTION

OPEN ITEMS 10/1/98 Update: Per NRG correspondence

dated May 18, 1998, titled "NRG INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item

(50-255/97201-16)

was identified

as closed. This item was also the subject of a NOTICE OF DEVIATION

(50-255/98003-08)

from the same letter. Palisades

responded

with additional

information

to the NRG \ . .mder correspondence

dated June 24, 1998, entitled "RESPONSE

TO NOTICE OF VIOLATION

AND NOTICE OF DEVIATION

FROM INSPECTION

REPORT 50-255/98003." In summary, Palisades

concludes

that our commitment

to assure that redundant

safety related power sources cannot be both affected by a fault on the instrument

bus has been maintained.

NRG correspondence

dated August 3, 1998 concluded

that a USQ does not exist, and that Consumers

appropriately

notified the NRG of past design changes, and rescinded

this cited potential

deviation.

No further actions are planned for this inspection

item. Inspection

Followup Item 50-255/97-201-17

The team observed that no system analysis existed to show that all the Class 1 E 120-V ac loads had *adequate

voltages.

The licensee demonstrated

during the inspection

that adequate voltages did exist for selected loads. For example, EA-ELEC-VOLT-24, "Voltage Drop From Preferred

AC Power Source Y10 Breaker 2 and Y40 Breaker 2 Out to the 5U12 Relays," Revision 0, showed that adequate ac voltage for those selected components

was available

at the minimum.inverter

voltage. The licensee initiated

CR C-PAL-97-1621

to evaluate and resolve this concern. The team identified

this item as part of Inspection

Fol/owup Item 50-255197-201-17.

Palisades

60 Day Response:.

During the Design Inspection, an operability

determination

was made concluding

the Class 1 E 120 V * ac loads are operable based on past plant operating

experience

and the expected minimal change in supplied voltage between normal and accident plant conditions.

By August 15, 1998,. we will perform a bounding analysis to confirm that Class 1 E 120 V ac loads have adequate voltage during accident conditions.

10/1/98 Update: Per NRG correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item

(50-255/97201-17)

is identified

as open. A bounding calculation

was performed

under EA-C-PAL-97-1621A-01

that developed

worst case voltage levels for the Preferred

AC System and confirmed

adequate available

voltage during accident conditions . These analysis results will be incorporated

into Design Basis Document DBD-4.03, "Preferred

AC System" and tracked under change request number 4.03-12-R3-0728.

No further actions are planned for this inspection

item . 21

  • * ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS Unresolved

Item 50-255/97-201-18

The team reviewed relay settings for protective

relays associated

with LPSI pump P-67 A, HPSI pump P-66A, SW pump P-7A, CCW pump P-52A, EOG 1-1 differential

protection, bus 1C undervoltage

protection, and Bus 1 C second-level

undervoltage

protection.

The settings were consistent

with the design parameters

of the devices being protected.

However, during the review, the licensee determined

that the overcurrent

relays for supply breakers 152-105 and 152-106 to bus 1C had not been calibration

tested during the last refueling

outage (1995) as required by Periodic and Predetermined

Activity (PPAC) SPS025, "Bus 1 C Relay Testing." The licensee stated that these relays would be calibrated

during the 1998 refueling

outage. The licensee reviewed past calibration

data for this type of relay and determined

that negligible

drift had previously

been documented.

The licensee initiated

CR C-PAL-97-1568

to resolve this discrepancy.

It appeared that the requirements

of 10 CFR Part 50, Appendix B, Criterion

XI, "Test Control," had not been implemented

in this case in that certain relays had not been tested as required by the test program. The team identified

this item as Unresolved

Item 50-255197-201-18.

  • Palisades

60 Day Response:

During the Design Inspection, an operability

determination

was made concluding

that past calibrations

of overcurrent

relays for breakers 152-105 and 152-106 revealed insignificant

drift and the relays are operable.

We will perform maintenance

activity PPAC SPS025 to calibrate

the overcurrent

relays during the 1998 refueling

outage. Our corrective

action history identified

no other examples of failure to perform scheduled

relay testing. 10/1/98 Update: Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item

(50-255/97201-18)

was identified-as

closed. This item was also the subject of a NOTICE OF VIOLATION

(50-255/98003-09)

from the same letter. Palisades

responded

with additional

information

to the NRC under correspondence

dated June 24, 1998, entitled "RESPONSE

TO NOTICE OF VIOLATION

AND NOTICE OF DEVIATION

FROM INSPECTION

REPORT 50-255/98003." In summary, the overcurrent

relays for breakers 152-105 and 152-106 will be tested/calibrated

by December 31, 1998. The requirements

for PPAC SPS025 have been revised to allow performance

of the testing and calibration

while the plant is at power operation.

Unresolved

Item 50-255/97-201-19

The team questioned

the replacement

schedule for Agastat E7000 series relays. The team was aware that the manufacturer, in correspondence

to other utilities, had recommended

a 10-year replacement

schedule for these relays. The licensee stated that 52 E7000 series relays were installed

and that 7000 series Agastats were also installed

in Class 1 E applications.

Some circuits containing

7000 series relays included the 2400-V bus 1C and*1D supply breakers, time delay relays associated

with charging pumps. P-55A, B, and C, and auto transfer failure alarms for 2400-V busses 1C and 10. The manufacturer's

stated qualified

life forthe E7000 relays was 10 years. The licensee stated that the

qualified

life applied if the relays were located in a harsh environment

and, 22

  • * * ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS since the E7000 relays were located in a mild environment, no qualified

life determination

was required.

Based upon this justification, the licensee issued PPAC Deletion Form MSE 034, dated March 3, 1995, which stated that the relays would not require replacement

at 10-year intervals.

The team believed that the qualified

life stated by the manufacturer

applied to any environment.

The team verified with the manufacturer

that the projected

qualified

life of 10 years was the operating

life of the E7000 series relay as long as the device did not exceed the equipment

ratings, and that the life of 10 years was applicable

to either a mild or harsh environment.

The licensee had not evaluated

the qualified

life ofthe 7000 series relays. The manufacturer

of Agastat relays issued a 10 CFR Part 21 notification

concerning

the inability

of the E7000 series relays to switch a 1-amp load at rated voltage. The licensee evaluated

the installed

E7000 series relays and identified

no concerns.

The team observed that this evaluation

did not review those 7000 series relays dedicated

by the licensee to safety-related

use. The licensee issued CR C-PAL-97-1663

to resolve the issues concerning

Agastat relays and determined

that all the relays were operable.

It appeared that the requirements

of 10 CFR Part 50, Appendix B, Criterion

Ill, "Design Control," had not been met in this instance in that the design basis lifetime for Agastat relays as stated by the manufacturer

had not been correctly

implemented

in the facility.

The team identified

this item as Unresolved

Item 50-255197-201-19.

Palisades

60 Day Response:

During the Design Inspection, an operability

determination

was made concluding

that the 7000 series relays are operable based on their similarity

in application

and design to E7000 relays. By July 15, 1998, we will complete our analysis of both 7000 and E7000 series relays dedicated

for safety related use to confirm their ability to perform safety-related

functions

during their installed

life and their conformance

with applicable

design requirements.

10/1/98 Update: Per NRG correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-19)

was identified

as open. A review of both 7000 and E7000 relay age-sensitive

components

was performed

that indicates

that all relay materials

will last for greater than 40 years without significant

degradation

when installed

in mild environments.

Based on this review, a 10 year replacement

interval is not justified

and the relays can be expected to perform their design function for greater than 40 years. No further

actions are planned for this inspection

item. Unresolved

Item 50-255/97-201-20

The 125-V de system was divided into two independent

systems. Each system consisted

of a battery, switchgear, distribution

panel, and two chargers.

Station battery 1, battery charger 1, and battery charger 3 supplied 125-V de bus 1. Battery charger 1 was supplied from MCC 1 and battery charger 3 was supplied from MCC 2. Administrative

controls limited the operation

so that only one charger per battery was in service. This prevented

a common-mode

failure from affecting

both * 23

ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS emergency

busses. The supply to 125-V de bus 2 was similar, with battery charger 2 fed from MCC 2 and battery charger 4 fed from MCC 1. Operating

Procedure

SOP-30, "Station Power," Revision 20, required the battery chargers to be operated in pairs (1 and 2 or 3 and 4). The licensee stated that the battery chargers were swapped monthly to provide equal operating

time for each battery charger. During swapping of the battery chargers in accordance

with Section 7. 7. 2 of SOP-30, the 125-V de breaker on the in-service

battery charger was opened and then the 125-V de breaker for the battery charger to be placed in service was closed. During this evolution, both battery chargers were disconnected

from the station battery and 125-V de switchgear

bus. Although temporary

disconnecting

the battery charger from the de bus had minimal safety impact on the plant, the team observed that TS 3. 7. 1 h required two station batteries

and the de systems (including

at least one battery charger on each bus) to be operable when the PCS was above 300 °F. The licensee stated that an LCO was not entered when no battery chargers were connected

to the de busses. The licensee initiated

CR C-PAL-97-1537

to resolve this discrepancy.

The team identified

the licensee's

failure to enter an LCO during battery charger switching

evolution

as Unresolved

Item 50-255197-201-20.

Palisades

60 Day Response:

Prior to the Design Inspection, we concluded

that our design bases were met and an LCO would not entered when realigning

battery chargers.

This conclusion

was based on no appreciable

battery discharge

occurring

during the short realignment

period when neither

charger was connected

to the 125 Vdc bus. In response to this Design Inspection

item, however, operating

procedure

SOP-30 was revised in anticipation

of an amendment

approving

our December 27, 1995 technical

specifications

change request. Although the requested

change does not require a connected

charger, the change defines 125 Vdc bus operability

in terms of applied bus voltage. SOP-30 now requires entry into an LCO whenever performing

charger realignment.

On January 26, 1998, a technical

specification

change request was resubmitted

as part of the Improved Technical

Specifications

Program. An amendment

in response to this latest change request will resolve this open item. 10/1/98 Update: Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-20)

was identified

as closed. In July 1998, Amendment

180 of the Palisades

Electrical

Technical

Specifications

was implemented

that clarifies

the 125 Vdc system operational

requirements.

With the issuance and implementation

of Amendment

180, no further actions are planned for this inspection

item. Inspection

Followi.Jp

Item 50-255197-201-21

The team reviewed the 125-V de battery loading during the normal and alternate

battery charger alignment.

During the normal battery charger alignment, battery charger 1 was powered from EOG 1-1 and battery charger 2 was powered from EOG 1-2. During a LOCA combined with a LOOP in this normal alignment, the batteries

would be without ac power for approximately

1 O seconds until the EDGs restored power. The team reviewed EA-ELEC-LDTAB-009, "Battery Sizing for the Palisades

Class 1 E Station Batteries

ED-01 and ED-02," Revision 2, which verified that the battery was sized to provide adequate power during the 10 second interval until the EDGs provided ac power to battery 24

  • * ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGNINSPECTION

OPEN ITEMS chargers 1 and 2. During the alternate

battery charger alignment

with battery charger 3 powered from EOG 1-2 and battery charger 4 powered from EOG 1-1, the station batteries

would be required to carry the de loads for more than 10 seconds in the event of a LOCA combined with a LOOP and a single failure of ac power. EA-ELEC-LDTAB-009

did not analyze the battery loading for station batteries

ED-01 and ED-02 during this condition.

When questioned

by the team the licensee stated that the de loading during this scenario would be greater than the worst-case

loading assumed in ELEC-LDTAB-009.

The licensee issued CR C-PAL-97-1596

to resolve this discrepancy.

Additionally

the team had concerns on whether the licensee met the single failure criterion

when the alternate

battery charger alignment

was in effect. The team identified

the question with respect to the single failure criterion

and the additional

loading on the battery as an Inspection

Followup Item 50-255197-201-21.

Palisades

60 Day Response:

During the Design Inspection, an operability

determination

was made concluding

that the station batteries

are operable.

Operability

was based on a preliminary

analysis where additional

  • conservative

loads were included in the battery load analysis showing that the battery terminal voltage would be greater than the required minimum output of 105 Vdc throughout

the exp.ected

load duration until an operable charger would be connected

to the bus. Operating

procedures

control alternate

charger alignment

but do not restrict this practice which is allowed by technical

specifications.

By January 15; 1999, we will complete a formal analysis of battery loading considering

the battery chargers are in their alternate

alignment, and a combined event of a LOCA, LOOP and single failure of ac power occurs. 10/1/98 Update: Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-21)

was identified

as open. As stated above, by January 15, 1999, the formal battery loading analysis will be completed.

Inspection

Followup Item 50-255/97-201-22

The team identified

that TS Section 4. 7.2c required that each station battery be demonstrated

operable by verifying

that the battery capacity was adequate to supply and maintain in an operable status all of the actual emergency

loads for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when the battery was subjected

to a battery service test. The battery service tests performed

on station batteries

ED-01 and ED-02 were performed

for a duration of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The 4-hour duration and loading was based on the design basis station blackout (SBO) coping time. The team noted that the 2-hour requirement

of TS 4. 7.2c was non-conservative

with respect to the design basis, which required the station batteries

to be available

for4 hours. The design basis duration of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> was included in FSAR Section 8.4.2; DBD 4.01, "Station Batteries," Revision 3; RE-83A, "Service/Modified

Performance

Test-Battery

No. ED-01," Revision 9, and RE-838, "Service/Modified

Performance

Test-Battery

No. ED-02," Revision 9. Testing the batteries

in accordance

with RE-83A and B has ensured that batteries

ED-01and02

have met the 4-hour design basis requirement.

The licensee has submitted

TS changes to correct the non-conservative

TS Section 4. 7.2c and issued CR C-PAL-97-1551

to resolve this discrepancy.

The team identified

this item as Inspection

Followup Item 50-255197-201-22.

25

  • ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS Palisades

60 Day Response:

During the Design Inspection, an operability

determination

was made concluding

that the 4-hour SBO station battery load profile envelops the 2-hour OBA load profile. By January 15, 1999, we will complete a formal analysis of battery loading considering

the battery chargers are in their allowed alignment

configurations

with a combined event of a LOCA, LOOP and.single

failure of ac power. We submitted

a technical

specification

change request on December 27, 1995 to describe the test profile as the design basis profile without stipulating

a specific period for the profile. On January 26, 1998, a technical

specification

change request was resubmitted

as part of the Improved Technical

Specifications

Program which identifies

a four hour load profile for the service test. An amendment

in response to this latest technical

specifications

request will resolve this open item. 10/1/98 Update: Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item

(50-255/97201-22)

was identified

as open. As identified

above, by January 15, 1999, the formal battery loading analysis will be completed.

In July 1998, Amendment

180 of the Palisades

Electrical.Technical

Specifications

was implemented.

Amendment

180 does not specify a duty cycle (profile)

duration in units of time. Therefore, the design basis requirements

found in the FSAR can be used. Inspection

Followup Item 50-255/97-201-23

EA-ELEC-FL

T-005, "Short-Circuit

for the Palisades

Class 1 E Station Batteries

ED-01 and ED-02," . Revision 0, was submitted

to the team as the short-circuit

analysis for the Class 1E 125-V de system. The following

discrepancies

with the assumptions, methodology, and conclusions

were identified:

  • Section 4. 4 and 4. 5 assumed various breaker and fuse impedances, which had not been verified against the installed

facility.

  • Section 5. 2 utilized the battery charger current limit of 220 amps as the maximum short-circuit

contribution

without supporting

documentation.

  • Section 5.2 stated that the open-circuit

voltage was 2.06 V per cell, whereas the EA utilized an open-circuit

voltage of 2. 0 V per cell. * Section 8. 0 stated that the results were to be further reviewed by the licensee;

however, the team found no evidence of this review. Section 8. O also contained

no conclusion

about the de system acceptability.

The licensee issued A/Rs A-PAL-97-108, 109, and 110 to resolve these discrepancies.

The licensee stated that the* analyses would be reviewed and the conclusions

revised. During the 1995 refueling

outage, FES-95-206

replaced existing batteries

ED-01 and ED-02. The team questioned

if the sh9rt-circuit

current provided by the new battery was analyzed and if there were any effects on the de distribution

panel breakers, since the team noted that EA-ELEC-FL

T-005 26

  • * ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS . had not been updated since 1994. The team also noted that the design basis for the evaluation

of fault current contributions

on de circuits was in FSAR Section 8.5.2, which stated "The 125 volt de protection

design considers

the fault current available

at the source side of the feeder protective

device." However, the licensee stated that the short-circuit

contribution

value for de circuits was taken at the electrical

load terminals

and not at the breaker load terminals (de short-circuit

current value would be less when calculated

at the load terminal vice the source side of the feeder protection

device because voltage available

at the load terminal would be less than at the source breaker).

The licensee determined

that the short-circuit

contribution

at 8 breakers (breakers

72-101, 72-105, 72-106, 72-121, 72-127, 72-133, and 72-135) on distribution

panels 011-1and011-2

could exceed the short-circuit

interrupting

ratings when evaluated

in accordance

with the design basis method in the FSAR. Also, when the team questioned

the assumed breaker fault ratings on de busses 010, 020, 011-1, and 011-2of13,000

amps in EA-ELEC-FLT-005, the licensee was unable to show manufacturer

or testing documentation

to support this assumption.

The team believed that this assumption

was inconsistent

with its experience.

The licensee performed

an operability

review and issued CR C-PAL-97-1652

to resolve these discrepancies.

The maximum short-circuit

current of the battery installed

by FES-95-:206, as provided by the manufacturer, was 17094 amps. Calculation

EA-ELEC-FL

T-005 did not reflect this new short:..circuit

current. Upon questioning

by th.e team, the licensee stated that an evaluation

was performed

to ensure that the system short-circuits

were acceptable.

During the team's review of this evaluation

it was determined

that the maximum battery short-circuit

current was not utilized.

The.licensee

stated that the short-circuit

current utilized, 12,821 amps, was provided by the manufacturer

as a more realistic

value than 17,094 amps. However, the licensee could not document a basis for the 12,821 amps and stated that they would verify it with the manufacturer.

The team identified

these discrepancies

concerning

EA-ELEC-FL

T-005 as part of Inspection

Followup Item 50-255197-201-23.

Palisades

60 Day Response:

During the Design Inspection, an operability

determination

was made concluding

that a fault would more likely occur at the load rather than at the breaker terminals.

A fault at the load (esults in a reduced value of fault current which falls within the breaker interrupting

rating. We have since obtained vendor specifications

which envelop our calculated

peak short circuit currents assumed to occur at the breaker terminals.

These specifications

confirm our earlier conclusion

that the breakers are suitable for their intended service, and resolve any concerns with respect to breaker short circuit interrupting

capability.

Revisions

to analysis EA-ELEC-FL

T-005, to correct the plant-identified

deficiencies

described

in the Design Inspection

report, will be complete by January 15, 1999. 10/1/98 Update: Per NRG correspondence

dated May 18, 1998, titled "NRG INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-23)

was identified

as open. Revisions

to analysis EA-ELEC-FL

T-005, to correct the plant-identified

deficiencies

described

in the Design Inspection

report, remains scheduled

for completion

by January 15, 1999 . 27

  • ** * ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS Inspection

Followup Item 50-255197-201-24

FSAR Section 8.4.3.3 stated that the batteries

were designed to furnish their maximum load down to an operating

temperature

of 70 °F without dropping below 105 V de, and that the equipment

supplied by the batteries

was capable of operating

satisfactorily

at this voltage rating. EA-ELEC-VOL

T-026, "Voltage Drop Model of the Palisades

Class 1 E Station Batteries

D01 and D02," Revision 0, evaluated

the de voltages at the distribution

panels based upon a battery voltage of 105 V de, but did not evaluate the voltages that would be available

at the load device terminals.

The team was concerned

that the additional

voltage drop from the distribution

panel to the loads could result in voltages less than the design basis of the loads, and that no analysis was performed

to evaluate this situation.

For example, the deign-basis

minimum input voltage for the inverters

was 105 V de and the licensee could not show any vendor documentation

to support operating

at a value Jess than 105 V de. The team noted that the inverters

could be subjected

to an input voltage of approximately

102 V de if the battery voltage were 105 V de. The licensee stated that battery surveillance

testing has shown that battery voltage, when subjected

to an SBO duty cycle, did not decrease below 108 V de. During the inspection, the licensee evaluated

several safety-related

loads and verified that adequate voltages would exist at 105 V de battery voltage. The licensee issued CR C-PAL-97-1620

to evaluate the lack of an EA to ensure that adequate voltages would exist at the load terminals.

The team identified

this item as part of Inspection

Followup Item 50-255197-201-24.

Palisades

60 Day Response:

During the Design Inspection, an operability

determination

was made concluding

that the 125 Vdc system is operable based on an evaluation

of several safety related loads, in which adequate load voltage was found to exist with a 105 Vdc battery terminal voltage. By November 15, 1998, we will perform a bounding analysis to identify the worst-case

minimum voltage levels at the load assuring that minimum load voltage req.uirements

are met. * 1011198 Update: Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-24)

was identified

as open. As stated above, this issue is scheduled

for completion

by November 15, 1998. Unresolved

Item 50-255197-201-25

The team also questioned

the capability

of solenoid valves to operate at voltages of 87 V de as stated in DBD 1. 01,

Cooling Water System," Revision 4. The licensee determined

that the DBD was incorrectly

worded and that the correct solenoid capability

was90-140 V de. Upon further review, the licensee identified

that improperly

rated coils, rated 102-126 V de, were installed

in solenoid valves SV-0918 and SV-09778.

The licensee initiated

Engineering

Assistance

Request (EAR) 97-0652 to replace the coils. It appeared that the requirements

of 10 CFR Part 50, Appendix B, Criterion

Ill, "Design Control," were not followed in that the design basis for the solenoid valve coils was not implemented

in the plant. The team identified

this item as Unresolved

Item 50-255197-201-25 . 28

    • ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS Palisades

60 Day Response:

Since the design inspection, further evaluation

identified

that there is no impact on the mitigation

of an accident if solenoid valves SV-0918 and SV-09778 fail to open due to low voltage since the close position is both the failed position and the required safety position.

Based on this review, the design basis is met by the existing solenoid valve installation.

The actions in response to Inspection

Followup Item 50-255/97-201-24

will identify any other minimum voltage problems.

10/1/98 Update: Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item

(50-255/97201-25)

was identified

as closed. No further actions are planned for this inspection

item. Inspection

Followup Item 50-255/97-201-26

The team identified

other discrepancies

in calculations

as follows: * Assumptions

4. 6 and 4. 7 of EA-ELEC-VOL

T-26, Revision 0, and assumptions

4. 8 and 4. 9 of EA-ELEC-M/SC-022, "Electrical

Systems Model of the Palisades

Class 1 E Safety Re/a.fed 125 V de System," Revision 1, assumed various fuse and breaker impedances

which had not been verified against the installed

equipment.

  • Section 7. 0 of EA-ELEC-VOL

T-26, Revision 0, "Conclusion," stated that the results were to be further reviewed by the licensee;

however, the team found no indication

that this review had been performed.

The "Conclusion" section also contained

no statement

concerning

the de system acceptability.

  • EA-ELEC-VOL

T-26, Revision 0, utilized a correction

factor for battery temperature

of 77 °F instead of the correction

factor for 70 °F, which was the minimum design basis temperature

for the battery. The number utilized is less conservative

and the licensee evaluated

that the overall effect on voltages in the calculation

would be less than 0. 5 percent. * EA-ELEC-LDTAB-029, Revision 2, stated the type of battery constant as 1.0 in Attachment

A and 1.4 on Sheet 4. The constant to be utilized depended on the type of battery. 1. 0 referred to a lead acid battery; 1.4 referred to a nickel-cadmium

battery. The licensee reviewed the EA and determined

that the correct constant was utilized in the EA and that the reference

to 1. 4 was an editorial

error. The licensee issued CR C-PAL-97-1656

to address the battery temperature

correction

factor and stated that the other discrepancies

would be corrected

in future revisions

to the calculations.

The team identified

this item as part of Inspection

Fo//owup Item 50-255197-201-26.

29

  • * ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS Palisades

60 Day Response:

During the Design Inspection, an operability

determination

was made concluding

that the calculation

deficiencies

identified

had no affect on the analyses conclusions;

ie, supplied voltages remain within equipment

ratings and the station batteries

are not affected.

By January 15, 1999, EA-ELEC-VOLT-26, EA-ELEC-MISC-022

and EA-ELEC-LDTAB-029

will be revised to resolve the deficiencies

noted above. 10/1/98 Update: Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item

(50-255/97201-26)

was identified

a:s closed. Analyses EA-ELEC-VOLT-26, EA-ELEC-MISC-022

and EA-ELEC-LDTAB-029

will be revised by January 15, 1999 as projected

above. Inspection

Followup Item 50-255/97-201-27

The team noted that TS Section 4. 7.1.b required testing to be performed

at every. refueling

to demonstrate

the overall automatic

operation

of the emergency

power system. Proper operation

was verified by bus load shedding and automatic

starting of selected motors and equipment

to establish

that emergency

power had been restored within 30 seconds. FSAR Tables 8-6 and 8-:-7 stated that sequencing

would occur in 65 seconds. Technical

Surveillance

Procedure

RT-BC, "Engineered

Safeguards

System -Left Channel," Revision 8, and RT-8D, "Engineered

Safeguards

System -Right Channel," Revision 8, required performance

testing to be within the 65-second

requirement.

The team questioned

the use of a 30-second

test duration in the TS instead of a 65-second

duration, which would demonstrate

that all required equipment

would start. The licensee stated that the TS did not specifically

require full testing of the entire diesel load sequence but only required testing of selected loads. The team noted that the licensee was testing the diesel loading to the full accident loading sequence and has submitted

a proposed TS change which would be more consistent

with the current design. The team reviewed Test Procedures

R0-128-1, "Diesel Generator

1-1 24 Hour Load Run," Revision 2, and R0-128-2, "Diesel GeneratOr

1-2 24 Hour Load Run," Revision 2. The team noted that Section 3. O of the Acceptance

Criteria and Operability

Sheet for Procedure

R0-128-2 referred to TS Section 3. 7. 1 and 4. 7. 1. 11, and that these references

would only be correct when the proposed improved TS, which have been submitted

to NRG for approval, became effective.

The licensee issued CR C-PAL-97-1566

to resolve these discrepancies.

The team identified

this item as Inspection

Followup Item 50-255197-201-27.

Palisades

60 Day Response:

Several issues identified

in the Design Inspection

are associated

with interpretation

of existing Technical

Specifications.

On December 27, 1995 we submitted

an electrical

technical

specifications

change* request which served to resolve the discrepancy

noted above pertaining

to the Emergency

Diesel Generator (EOG) load sequence test. On January 26, 1998, we submitted

a request for improved technical

specifications

which specifies

testing the EOG to the load 30

ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS intervals

programmed

by the sequencer;

eliminating

any specific reference

to the sequence time. It is expected that the amendment

resulting

from the most recent .technical

specification

change request will serve to resolve this and other technical

specification

related open items. 1011198 Update: Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item

(50-255/97201-27)

was identified

as closed. In July 1998, Amendment

180 of the Palisades

Electrical

Technical

Specifications

was implemented.

Amendment

180 specifies

testing the EOG to the load intervals

programmed

by the sequencer;

eliminating

specific reference

to the sequence time. No further actions are planned for this item. Inspection

Followup Item 50-255197-201-28

The team identified

the following

discrepancies

when reviewing

station battery Test Procedures

RE-83A, "Service/Modified

Performance

Test-Battery

No. ED-01," Revision 9, and RE-83B, "Service/Modified

Performance

Test-Battery

No. ED-02," Revision 9: * The tests evaluated

whether the final discharge

voltage (105 V de) of station batteries

ED-01and02 was met at the end of the test (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />). Load parameters (amps) at 1 and 239 minutes were not verified during the test. These load parameters

were design requirements

of EA-ELEC-LDTAB-009, Revision 2. The licensee demonstrated

that the 1-and 239-minute data were recorded elsewhere

and that the duty cycle was* tested in accordance

with the design requirements.

The licensee stated that the battery testing procedures

would be revised to include verification

of these design parameters.

  • The procedures

did not require any calibration

tolerances

for the discharge

testing shunt and control unit. The licensee stated that the tolerance

was removed from the procedure

before testing during the 1996 refueling

outage and issued PCRs 5422 and 5423 to change the

procedures

to include these tolerances.

  • The battery charging data in Procedure

RE-83B for the 1996 refueling

outage did not meet Step 5. 2. 2, which required the battery charging rate to be decreasing

and to remain within 5 percent over the last 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> before stopping the equalization

process, in that the process was stopped before the end of the 8-hour period. The licensee stated that the nearly steady state voltage operation

of the charger gave adequate assurance

that the battery was operable before exiting the test and issued CR C-PAL-97-1460

to resolve this discrepancy.

  • During the performance

of procedure

RE-83B at the 1996 refueling

outage, the elapsed time recorded manually did not agree with the testing control unit time. The licensee stated that because the testing unit did not have the capability

to record the time, the test start and stop times were recorded manually.

The inconsistencies

were minor and had no effect on the test results. The licensee issued C-PAL-97-1460

to evaluated

this discrepancy.

The team identified

this item as Inspection

Followup Item 50-255197-201-28.

31

  • * ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS Palisades

60 Day Response:

Note: Inspection

Followup Item 50-255/97-201-28, Unresolved

Item 97-201-30

bullets 7, 8, 9, 10, 11 and 12, and Unresolved

Item 97-201-31

bullets 6 and 13 are completed

under this action due to their subject similarity.

Surveillance

tests RE-83A and RE-838 will be revised as appropriate

to eliminate

the identified

deficiencies

to support 1998 refueling

outage performance.

By December 15, 1998, we will review DC system requirements, FSAR Chapter 8 and surveillance

tests RE-83A and RE-838 for consistency, and resolve the deficiencies

identified

in this open item and the following:

  • Reconcile

FSAR section 8.2.3 concerning

the battery supplying

safe shutdown loads for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> with the requirement

to strip loads. (Inspection

report item #30-7.) * * Disposition

battery shunt and de tie breakers which are not consistent

with FSAR section 8.3.5.2. (Inspection

report item #30-8.) * Reconcile

one battery charger capability

to supply normal loads and recharge battery in less than 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> with FSAR section 8.3.5.3. (Inspection

report item #30-9.) * Reconcile

alternate

alignment

of battery chargers with FSAR section 8.4 .. 2.2. (Inspection

report item #30-10.) * Reconcile

battery chargers cross connection

with FSAR section 8.5.2. (Inspection

report item #30-11.) * Reconcile

design of system 1, 2, 3, 4 circuits and their separation requirements

with FSAR section 8.5.3.2. (Inspection

report item #30-12.) * Add battery discharge

restriction

to the D8D. (Inspection

report item #31-6.) * Disposition

battery cell specific gravities. (Inspection

report item #31-13.) 10/1/98 Update: Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION, this item (50-255/97201-28)

was identified

as open. Surveillance

tests RE-83A and RE-838 were revised and satisfactorily

performed

during the 1998 refueling

outage. The June 30, 1998 FSAR revision resolved inspection

report items #30-8, #30-9, and #30-12. The above remaining

items are scheduled

to be complete by December 15, 1998 .. Inspection

Followup Item 50-255/97-201-29

The team reviewed the following

electrical

modification

packages and found them consistent

with the plant design basis: 32

  • * * * * * * * ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS Temporary

Modification

TM-96-027, "lnsta/1152-Spare

  1. 5 Breaker in 152-113 Cubicle," dated April 10, 1996 FES-95-206, "ED-01 and ED-02 Station Battery Replacement," Revision O FC-364, "Feeder Change for Instrument

Bus Y-01," Revision O FC-854, "Y-01 Power Supply Feed Modification," Revision 0 FC-638, Add Component

Cooling Water Pumps to the Normal Shutdown Sequencer," Revision 0 FC-798, "Battery Room Temperature

Indication

and Alarm," Revision O FC-683, "Removal of Pressurizer

Heaters from SIS Trip," Revision O Except as previously

discussed, all these modifications

were adequately

prepared, provided the necessary

technical

basis for the changes, and contained

adequate installation

instructions

and testing requirements.

The 10 CFR 50. 59 safety evaluations

were adequate, except for the two listed below: = Safety Reviews 95-1431and95-1432, dated July 7, 1995, for FES-95-206

stated that the battery duty cycle service test duration for station .batteries

ED-01 and ED-02 was changed from 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The licensee noted that TS Section 4. 7.2.c was affected by this design change. However, the USQ evaluation, Question 2 of Section II, was not checked "Yes" for a TS change. TS 4. 7.2.c required that a 2-hour battery test be performed;

while design analysis ELEC-LDTAB-009

and FSAR Section 8.4.2 required a 4-hour battery duty cycle. The licensee has submitted

a proposed TS change to reflect the proper battery test duration and issued CR C-PAL-97-1551

to address this discrepancy.

  • The safety review documentation

for TM-96-027

stated that the FSAR was not reviewed.

Administrative

Procedure

3. 07, "Safety Evaluations," page 12, required that the FSAR be reviewed and that thos*e sections reviewed be noted on the safety review sheet. The licensee initiated .C-PAL-97-1493

to evaluate this discrepancy.

The team identified

these safety review discrepancies

as Inspection

Fol/owup Item 50-255197-

201-29. Palisades

60 Day Response:

It was not documented

in the safety evaluation

for FES-95-206

that a technical

specification

change would be required to change the battery duty cycle service test duration from 2 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. An FES-95-206-specific

technical

specifications

change was not considered

necessary

by the preparer of the safety evaluation

since a technical

specifications

change request eliminating

reference

to a specific duty cycle time was to be submitted

under the Improved Technical

Specifications

Program in the near term. Since completion

of the FES-95-206

safety evaluation, Palisades

has implemented

a Safety & Design Review Group which reviews and approves all design changes and safety evaluations.

The purpose for forming and employing

this group is to provide consistent

oversight

The quality of safety evaluations

and their reviews has significantly

improved over the recent years. It is unlikely that a safety evaluation

deficiency, similar to that associated

with FES-95-206, would have occurred

since deployment

of the Safety & Design Review Group. 33

  • * ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS The original safety review for TM-96-027

inappropriately

indicated

that FSAR sections had not been reviewed.

In reality, the FSAR was reviewed during safety review preparation

and the FSAR was found to contain description

at a level of detail that the TM would not affect. The review of the TM-96-027 safety review was performed

by telecon (an infrequent

practice)

with no follow-up

review performed

by the Safety & Design Review telecon reviewer.

By April 15, 1998, design control procedures

will be revised to require a follow-up

review whenever a review is performed

by telecon. 1011198 Update: Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND.NOTICE

OF VIOLATION", this item (50-255/97201-29)

was identified

as closed. Administrative

Procedure

AP 3.07, "SAFETY EVALUATIONS" was revised to require follow-up

reviews as stated above. No further actions are planned for this inspection

item. Note: Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-30)

was identified

as open. FSAR changes identified

in Unresolved

Item 50-255/97201-30

are identified

below. Some of these bullets are grouped and evaluated

with other URl's or IFl's. For clarity, each bullet's actions will be separately

addressed.

Unresolved

Item 50-255197-201-30

The team identified

the following

discrepancies

in the FSAR: * Page 6. 7-4 stated that 'containment

isolation

valves fail closed with loss of voltage or control air except for the CCW return isolation

valves. However, the CCW supply isolation

valve (CV-0910)

is also a fail-open

valve and should have *been noted as an exception

to fail-closed

containment

isolation

valves. The licensee issued FSAR Change Request 6-142-R20-1426

to correct the FSAR. Palisades

60 Day Response:

The next FSAR annual update revision will incorporate

this change. 1011198 Update: Annual FSAR update issued June 30, 1998, included this change. * Section 6. 7 classified

the CCW penetrations

as Class C-2, which was defined as penetrations

with lines not missile protected.

However, EA-GW0-7793-01

stated that the entire CCW system (both inside and outside containment)

was missile protected.

The licensee issued FSAR Change Request 6-143-R20-1427

to state that the CCW penetrations

were not vulnerable

to internally

generated

missiles . 34

  • * ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS Palisades

60 Day Response:

The next FSAR annual update revision will incorporate

this change. 10/1/98 Update: Annual FSAR update issued June 30, 1998, included this change. * Table 9-10 stated that valves 3029 and 3030, containment

sump suction valves, failed closed upon loss of air and were equipped with an accumulator.

The valves actually failed as is and had no accumulator.

The licensee issued FSAR Change Request 9-293-R20-1431

to correct *the FSAR and CR C-PAL-97-1559

to evaluate and trend the FSAR discrepancies

being identified

at the plant. Palisades

60 Day Response:

The next FSAR annual update revision will incorporate

this change. 10/1/98 Update: Annual FSAR update issued in June 30, 1998, included this change. * Table 9-9 correctly

stated that the high-pressure

air piping was seismic Class I from the receivers

to the valve operators.

However, FSAR Table 5.2-3 stated that the entire system was seismic Class I. The licensee issued FSAR Change Request 5-155-R20-1432

to correct the FSAR 5. 2-3. Palisades

60 Day Response:

The next FSAR annual update revision will incorporate

this change. 10/1/98 Update: Annual FSAR update issued in June 30, 1998, included this change. * Section 8.4.2.2 stated that the station batteries

would be tested to Institute

of Electrical

and Electronics

Engineers (IEEE) 450-1975.

However, battery testing procedures

RE-83A, Revision 9, and RE-838, Revision 9, referred to IEEE 450-1995.

FSAR Change Request 8-126-R20-1249

had been initiated, but the licensee did not intend to act on this change until approval was received from NRG of a related proposed TS change. Palisades

60 Day Response:

This FSAR change is on hold until the license amendment

responding

to our improved electrical

technical

speeification

change request, submitted

January 26, 1998, is received.

This change cites IEEE 450-1995 for the battery testing . 35

  • ** ATTACHMENT

A * STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS 1011198 Update: Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-30)

was identified

as open. In July 1998, Amendment

180 of the Palisades

Electrical

Technical

Specifications

was implemented

with IEEE 450-1995 as a reference.

FSAR change 8-126-R21-1249

will be implemented

as part of the next annual FSAR update. to reflect the use of this IEEE standard.

  • Table 5. 7-8 listed the seismic design value for the station batteries

and racks as "later" instead of including

the actual values of the batteries

installed

by FES-95-206.

The licensee issued EAR-97-0636

to evaluate this discrepancy

and revise the FSAR. Palisades

60 Day Response:

The table in the FSAR is designated

as containing

the original seismic design values for the plant. The term "later" was an original FSAR description

which acknowledged

that an impending

upgrade to install a second redundant

electrical

train would be made and the applicable

seismic criteria would not be available

until then. Since we have chosen to keep this table for historical

record, the word "later" will be removed and the table maintained

as original seismic criteria.

The next FSAR annual update will incorporate

this change requested

by FSAR Change Request 5-157-R20-1456.

1011198 Update: Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this portion of unresolved

item 50-255/97201-30

was identified

as closed. The annual FSAR update issued June 30, 1998, included this change. No further actions are planned for this inspection

item. * Section 8.2.3 stated the "The de battery system is designed to supply the required shutdown loads, with a total loss of ac power, for at ieast 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />." This statement

did not reflect the fact that load stripping

was required during the 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for the battery to perform its intended function during a loss of ac power. * Palisades

60 Day Response:

Refer to our response to Inspector

Followup Item 50-255/97-201-28.

1011198 Update: The resolution

of this issue is addressed

in Inspection

Followup Item 50-255/97201-28

due to subject similarity.

This item is projected

to be complete by December 15, 1998. Section 8. 3. 5. 2 stated that "Operation

of all circuit breakers in the de and the preferred

ac systems is manual with automatic trip for fault isolation." The battery shunt trip breakers and the de bus tie breakers do not comply with this statement.

36

  • ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS Palisades

60 Day Response:

Refer to our response to Inspector

Followup Item 50-255/97-201-28.

10/1/98 Update: Revision 20 of FSAR Chapter 8 incorporates

the exclusion

of the battery isolation

shunt trip breakers and tie breakers between the left and right sections of each switchgear

bus that do not have an automatic trip for fault isolation.

Our June 30, 1998, annual FSAR update includes this change. * Section 8. 3. 5. 3 stated that "Each of the two battery chargers provided on the. de bus is capable of supplying

the normal de loads on the bus and simultaneously

recharging

the battery in a reasonable

time. A fully discharged

battery can be recharged

in less than nine hours." Contrary to the statement, one battery charger could not supply the normal loads and recharge a fully discharged

battery in less than 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />. Palisades

60 Day Response:

Refer to our response to Inspector

Followup Item 50-255/97-201-28.

10/1/98 Update: Revision 20 of FSAR Chapter 8 now states that two battery chargers are needed to recharge a fully discharged

battery in less than nine hours. Our June 30, 1998, annual FSAR update includes this change. * Section 8.4.2.2 stated that "Emergencv

Operation

-.On loss of normal and standby ac power, the batteries

will supply power to all preferred

ac and de loads, until one of the (diesel generators)

DGs has started and can supply power for the chargers." This statement

was not correct if the battery chargers were in their alternate

alignment

and did not reflect load shedding during the 4-hour duration.

Palisades

60 Day Response:

Refer to our response to Inspector

Followup Item 50-255/97-201-28.

10/1/98 Update: The resolution

of this issue is addressed

in Inspection

Followup Item 50-255/97201-

28 due to subject similarity.

We plan to complete this item by December 15, 1998. * Section 8.5.2 stated that The power source for the driven equipment

and the control power for that system are supplied from the sources in one channel." This statement

would not be correct if the battery chargers were cross-connected . 37

  • ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS Palisades

60 Day Response:

Refer to our response to Inspector

Followup Item 50-255/97-201-28.

10/1/98 Update: The resolution

of this issue is addressed

in Inspection

Followup Item 50-255/97201-

28 due to subject similarity.

We plan to complete this item

by December 15, 1998. * * Section 8.5.3.2 referred to "System 1, 2, 3, 4 Circuits" and separation

requirements

for those circuits.

The licensee was not able to identify these circuits.

  • Palisades

60 Day Response:

Refer to our response to Inspector

Followup Item 50-255/97-201-28.

10/1/98 Update: Revision 20 of FSAR Chapter 8 expands the definition

along with providing

routing and isolation

requirements

for 'left', 'right' and channel '1 ', '2', '3', and '4' circuits.

Our June 30, 1998, annual FSAR update includes this change. * Section 8.4.1.3 required clarification

as to whether the reserve capability

margin referred to the capability

of the overall EDG and engine or if it referred to the capability

of the EOG to handle an increase loading due to a control circuit ma/function

during the loading sequence.

The licensee issued C-PAL-97-1309

to resolve this discrepancy.

Palisades

60 Day Response:

Prior to the Design Inspection, an operability

determination

was made concluding

that the EDGs are operable.

This conclusion

was reached based on the capability

of the EDGs to provide the required design function

in the event of a control. circuit malfunction

or delayed containment

high pressure signal; but not both concurrently.

The design basis accident analysis does not require that these two events occur simultaneously.

Due to the change being descriptive

in nature, rather than licensing

basis information, we have elected to use the Design Basis Documents

rather than the FSAR to make the clarification.

Design Basis Document Change 5.03-11-R3-

0617 was initiated

and the revision will be made by December 15, 1998. 10/1/98 Update: Revision 4 of DBD 5.03 incorporates

the requested

change which evaluated

the system functional

requirements

of the EOG starting and carrying the largest load due to a control circuit malfunction.

Revision 4 also includes discussion

regarding

the EOG control circuit malfunction

and starting a containment

spray pump during a delayed containment

high pressure scenario;

  • concluding

that the malfunction

and the pump start are mutually

exclusive.

No further actions are planned for this item. 38

  • .ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS * Section 6.1.2.3 stated that The RAS ... provides a permissive

to manually close the valves in the pump minimum flow lines." EOP-4. 0, "Loss of Coolant Accident Recovery," Revision 9, Step 23, directed the operators

to place the hand switches for these valves in the pump minimum flow lines (CV-3027 and CV-3056) to CLOSE when SIRWT level lowered to between 25 percent and 15 percent. Per EOP-4.0, Step 51, the RAS occurred when the SIRWT level reached 2 percent. The FSAR appeared to conflict with EOP-4.0. The licensee initiated

FSAR Change Request 6-141-R20-1425

to update the FSAR. Palisades

60 Day Response:

The next FSAR annual update revision will incorporate

this change. 10/1/98 Update: Annual FSAR update issued in June 30, 1998, included this change. * The footnote for Table 14.17.1-1

implied that a containment

building temperature

of 90 °F was used as input to the large-break

LOCA analysis because it is the limiting temperature

during normal operation.

The 90 °F value did not appear to be limiting.

The licensee stated that the 90 °F value was the nominal containment

building temperature, not the limiting temperature, and was used in the accident analysis in accordance

with Seimens Power Corporation's

large-break

LOCA methodology

guidelines.

The licensee initiated

FSAR Change Request 14-95-R20-1441

to update the FSAR. * Palisades

60 Day Response:

The next. FSAR annual update revision will incorporate

this change. 10/1/98 Update: Annual FSAR update issued in June 30, 1998, included this change. The above discrepancies

had not been corrected

and the FSAR had not been updated to ensure that the material in the FSAR contained

the latest material

as required by 10 CFR 50. 71(e). The team identified

this item as Unresolved

Item 50-255197-201-30.

Palisades

60 Day Response:

10 CFR 50.71(e) requires that the FSAR be updated to contain the latest material developed

and that it includes the effects of all changes made in the facility or procedures

described

in the FSAR. Although several of the identified

FSAR discrepancies

were clear errors, most were cases where statements

in the FSAR were misleading

or unclear and not cases where the FSAR was not updated per 10 CFR 50.71 (e). Our ongoing FSAR verification

and validation

effort should provide identification

and correction

of similar conditions

which may exist in the FSAR. Our processes

were also changed a few years ago to require a safety review (1 O CFR 50.59 screening)

for 39

ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS all analyses, modifications, etc which have the potential

to affect the design basis of the facility.

This widespread

10 CFR 50.59 screening

will prevent failures to update the FSAR in accordance

with 10 CFR 50.71(e).

In addition, a license basis self assessment

performed

in accordance

with NEI 96-05, "Guidelines

for Assessing

Programs for Maintaining

the Licensing

Basis," found few discrepancies

in the FSAR sections sampled which

had not been previously

identified

for correction

by other plant processes.

Therefore, we feel that the current efforts underway will correct other errors which may exist in the FSAR and the current plant processes

will ensure that the FSAR is updated properly.

10/1/98 Update: The above response remains unchanged

from our 60-day response.

Note: Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", this item (50-255/97201-31)

was identified

as open. DBD changes identified

in Unresolved

Item 50-255/97-201-31

are identified

below. Some of these bullets are grouped and evaluated

with other UR l's or IFl's. For clarity, each bullet's actions will be separately

addressed.

Unresolved

Item 50-255/97-201-31

The team identified

the following

discrepancies

in the DBDs: * DBD 1.07, Auxiliary

Building HVAC Systems," Revision 1, Table 3.2.1, incorrectly

stated that the design basis temperature

for Room 123, which contains the CCW pumps, was 125 °F. The correct temperature

was 104 °F as stated in 080 7.01, "Electrical

Equipment

Qualification

Program," Revision 1, Appendix A. The 125 °F temperature

was a conservative

assumption

used to size the outside air supply fans. Table 3.2.1 also contained

a typographical

error in a reference

number. The licensee issued 080 Change Requests 1.07-71-R1-0512

and 1.07-72-R1-0532

to correct the 080. Palisades

60 Day Response:

The identified

Design Basis Document Change Request will be incorporated

into the DBD by July 1, 1998. 10/1/98 Update: Revision 2 of DBD 1.07, "Auxiliary

Building HVAC Systems" incorporated

the above changes. The basis for the 125 ° F CCW room temperature

was clarified

and references

were corrected.

  • 080 1.07, Revision 1, Section 3.2.1.3, listed maximum room temperatures

for the west ESF room from an outdated analysis.

The latest analysis, EA-O-PAL-93-272F-01, "Engineering

40

  • ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS Safeguards

Room Heatup Following

LOCA in Conjunction

With a Loop," Revision 0, determined

lower maximum room temperatures

for various SW flows through the air coolers. The 080 also required clarification

of the normal design temperature

of the ESG room. The licensee issued 080 Change Request 1.07-73-R1-0543

to correct the 080. Palisades

60 Day Response:

The identified

Design Basis Document Change Request will be incorporated

into the DBD by July 1, 1998. 10/1/98 Update: Revision 2 of DBD 1.07, "Auxiliary

Building HVAC Systems" incorporated

the above change. The basis for the 135°F Engineering

Safeguards

Room temperature

was clarified.

  • 080 7. 08, "Plant Protection

Against Flooding, 77 Revision 1, incorrectly

stated that the EOG would be inoperable

before a flood reached the EOG windings because the lube oil heaters were located below the windings at 7 inches above the floor. EA-C-PAL-95-1526-01, "Internal

Flooding Evaluation

for Plant Areas Outside of Containment, 77 Revision 0, stated that the minimum flood level at which the EOG could become inoperable

was 10 inches due to the exciter cubicle bus bars and that the lube oil heaters were not needed for EOG * operability.

The licensee issued CR C-PAL-97-1557

to initiate a 080 change and evaluate the item. Palisades

60 Day Response:

During the Design Inspection, an operability

determination

concluded

that the EDGs * are operable based on other indications

available

to inform operations

that water level in the rooms is increasing.

DBD change request 7.08-40-R1-0561

was initiated

to state that the limiting component

is not lube oil heaters but the exciter cubicle bus bars located ten inches above the EOG room floor. The identified

Design Basis Document Change Request will be incorporated

into the DBD by December 15, 1998. 10/1/98 Update: This DBD change is on target for completion

by December 15, 1998 as identified

above. * 080 2. 03, "Containment

Spray System, 77 Revision 2, stated that the air supply to the sump outlet valves, CV-3029 and 3030, was backed by an accumulator.

There were no accumulators

for these valves. The licensee identified

this error while evaluating

an FSAR statement

that these valves had an accumulator

backup that was questioned

by the team, and issued 080 Change Request 2.03-22-R2-0531

to correct the 080 . 41

  • ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS Palisades

60 Day Response:

The identified

Design Basis Document Change Request will be incorporated

into the DBD by July 1, 1998. 10/1/98 Update: Revision 3 of DBD 2.03, "Containment

Spray System" corrected

the terminology

from "accumulator" to "high pressure air receivers".

No further action is planned. * DBD 1.01, "Component

Cooling Water System," Revision 3, Section 3.3. 7, incorrectly

indicated

that Class 1 E and non-Class 1 E breakers were installed

in the same distribution

panels. The licensee initiated

DBD Change Request 1.01-14-R3-0518

to correct the DBD. Section 3. 3. 7 of this DBD also stated that solenoid valves had been tested to operate at 87 V de instead of 90 V de. The licensee stated that the DBD would be corrected.

Palisades

60 Day Response:

The identified

Design Basis Document Change Request will be incorporated

into the DBD by July 1, 1998. 10/1/98 Update: Due to competing

priorities, this DBD change has been rescheduled

to be completed

by December 15, 1998. * * * During the teain's review of FES-95-206, it was noted that the battery manufacturer

had imposed a limit of 40 battery discharges

for the 20-year life of the battery. This restriction

had not been identified

in any DBD. The licensee stated that the requirement

would be added to DBD4.01. . Palisades

60 Day Response:

A Design Basis Document Request will be incorporated

into the DBD by December 15, 1998. Refer to our response to Inspector

Followup Item 50-255/97-201-28.

10/1/98 Update: This DBD change is on target for completion

by December 15, 1998, as

above. * Appendix A of DBD 7. 02, "Palisades

Design Basis Document EQ Master Equipment

List," Revision 2, incorrectly listed

the location for L T-0383; referred to EIP 0343 instead of E/P 0346; and did not include SV-32138 in Table A-1. The licensee issued DBD Change Requests 7. 02-4-R2-0522, 7. 02-6-R2-0527, and 7.D2-4-R2-0523

to correct the DBD . 42

  • * * ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS Palisades

60 Day Response:

The identified

Design Basis Document Change Request will be incorporated

into the DBD by December 15, 1998. 10/1/98 Update: These DBD changes are on target for completion

by December 15, 1998. * DBD 2.01, "Low Pressure Safety Injection

System," Revision 3, and DBD 2.02, "High Pressure Safety Injection

System," Revision 3, both contained

references

to ANF-88-107, "Palisades

Large Break LOCNECCS Analysis With Increased

Radial Peaking," Revision 1. ANF-88-107

was superseded

by Seimens Calculation

EMF-96-172, "Palisades

Large Break LOCNECCS Analysis," Revision 0. The licensee Initiated

DBD Change Requests 2. 01-30-R3-0519 and 2.02-27-R3-0520

to update the DBDs. * Palisades

60 Day Response:

The identified

Design Basis Document Change Request will be incorporated

into the DBD by July 1, 1998. 10/1/98 Update: Revision 4 of DBD 2.01, "Low Pressure Safety Injection

System," and Revision 4 of DBD 2.02, "High Pressure Safety Injection

System," incorporated

reference

to the most current LOCA analysis.

No further action is planned for this item. DBD 2.01, "Low Pressure Safety Injection

System," Revision 3, Section 3.3.1.3, stated that the SIRWT must maintain a minimum of 20,000 gallons at the time of a RAS to limit the radiological

consequences

of an accident.

The DBD reference

for this statement

was TAM-95-05, "Radiological

Consequences

for the Palisades

Maximum Hypothetical

Accident & Loss of Coolant Accident," Revision 0. A review of EA-TAM-95-05

indicated

that this analysis did not take credit for the 20,000 gallons at the time of RAS to limit the radiological

consequences

of an accident.

The licensee issued DBD Change Request 2.01-31-R3-0524

to update the DBD. Palisades

60 Day Response:

The identified

Design Basis Document Change Request wlll be incorporated

into the DBD by July 1, 1998. 10/1/98 Update: Revision 4 of DBD 2.01, "Low Pressure Safety Injection

System," clarifies

the SIRW tank minimum volume design requirements.

No further action is planned for this item . 43

  • * * ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS The team also identified

the following

discrepancies

in other documentation:

  • P&ID M-232, Sheet 2A, incorrectly

identified

L T-0383 as connected

to penetration

  1. 54 instead of#56. The licensee issued Document Change Request (OCR) 97-0856lo

correct the drawing.

Palisades

60 Day Response:

P&ID M-232, Sheet 2A has been reviseo to incorporate

OCR 97-0856. 10/1/98 Update: No further update necessary.

  • Documents

E-33, Revision 46, and E-37, Revision 46, were not revised to reflect the installed

condi(ion

of the battery charger cabling that was rerouted by SC-89-284.

The licensee issued CR C-PAL-97-1495

to resolve this discrepancy.

Palisades

60 Day Response:

E-33, Rev 46 and E-37, Rev 46 have been revised to reflect the correct battery charger cable routing installed

by SC-89-284 . 10/1/98 Update: * No further

necessary.

  • * P&ID M-209, Sheet 3 (Revision

34), incorrectly

depicted valves SV-0918 and SV-09778 as normally deenergized.

The licensee issued EAR 97-0652 to revise the drawing. * Palisades

60 Day Response:

P&ID M-209, Sheet 3, Revision 35 has been issued to depict SV-09778 as normally energized.

Further evaluation

of SV-0918 identified

that the normally deenergized

state as depicted on M-209 Sheet 3 is appropriate

per FSAR Table 9-10. 10/1/98 Update: No further update necessary.

  • Vendor drawing E-12A, Sheet 39, Revision 0, indicated

that the battery discharge

characteristics

were based upon battery cell specific gravities

of 1.215 +/- 0.005. However, the batteries

were being maintained

to a criterion

of 1.215 +/- 0.010. The licensee issued EAR 97-0669 to update the drawing. Palisades

60 Day Response:

E-12 A, Sheet 39, Rev O will be updated by December 15, 1998. Refer to our response 44

ATTACHMENT

A STATUS OF ACTION PLANS TO RESOLVE DESIGN INSPECTION

OPEN ITEMS to Inspector

Followup Item 50-255/97-201-28.

10/1/98 Update: This item is on target for completion

by December 15, 1998. These documentation

discrepancies

were not consistent

with 1 O CFR Part 50, Appendix B, Criterion

Ill, "Design Control," which requires that the design basis be correctly

translated

into drawings.

The team identified

this item as Unresolved

Item 50-255197-201-31.

The programmatic

design control aspects related to this issue will be addressed

as identified

in Attachment

B, Item 1. 'i 45

* * ATTACHMENT

B CONSUMERS

ENERGY COMPANY PALISADES

PLANT DOCKET 50-255 STATUS OF PLANS TO RESOLVE RELATED PROGRAMMATIC

ISSUES AND EVALUATE FINDINGS AGAINST THE 10 CFR 50.54(f) RESPONSE 6 Pages

  • * ATTACHMENT

8 STATUS OF PLANS TO RESOLVE RELATED PROGRAMMATIC

ISSUES AND EVALUATE FINDINGS AGAINST THE 10 CFR 50.54(f) RESPONSE Per NRC correspondence

dated May 18, 1998, titled "NRC INSPECTION

REPORT 50-255/98003 (DRS) AND NOTICE OF VIOLATION", inspection

item 50-255/98003-01

was identified

as open. As stated in the report, this item

will remain open

pending NRC review of the results of the collective

significance

of individual

inspection

items and planned programmatic

improvements.

The following

summarizes

of our programmatic

improvements.

1. DESIGN CONTROL ISSUES: The following

issues were identified

in the Design Inspection

report as potentially

not meeting requirements

of 10 CFR 50, Appendix B, Criterion

Ill, "Design Control." Our design control program provides assurance

that the plant as-built configuration

conforms to design requirements, and the configuration

is operated, tested and maintained

within required design parameters.

The deficiencies

identified

during the Design Inspection

relate to these design control program objectives.

Design Objective

For Operating

Systems Within Design Parameters:

  • Loss-Of-Coolant

Accident analysis identified

the maximum CCW temperature

of 184°F yet the effects of this temperature

on CCW system components

was not performed. (Unresolved

Item 50-255/97-201-02.)

  • Incomplete

analysis (inadequate

justification

for conclusion

and incorrect

references

to related NRC correspondence)

for CCW piping for High Energy Line Break. (Unresolved

Item 50-255/97-201-04.)

  • Some AC Load calculations

have not been updated to reflect current design. (Unresolved

Item 50-255/97-201-14.)

Design Objective

For As-Built Conditions

Conforming

To Design Requirements:

  • * * Unscreened

Emergency

Core Cooling System Suction piping vent. (Unresolved

Item 50-255/97-201-10.)

Some instrument

tubing is not sloped consistent

with design requirements . (Unresolved

Item 50-255/97-201-13.)

Design Basis Document I design documentation

discrepancies. (Unresolved

Item 50-255/97-201-31.)

1

  • ATTACHMENT

B STATUS OF PLANS TO RESOLVE RELATED PROGRAMMATIC

ISSUES AND EVALUATE FINDINGS AGAINST THE 10 CFR 50.54(f) RESPONSE Palisades

60 Day Response:

Elements comprising

and supporting

our.design

control program consist of our calculation

control program, instrument

setpoint program, FSAR verification

and validation (V&V), design basis documents (DBDs) with associated

safety system design confirmations, and as-built confirmation

through drawing review or field walkdown.

These elements will be revised as appropriate

by December 15, 1998 to prevent the recurrence

of conditions

similar to those identified

in the Design Inspection

and cited above. Resolution

of any nonconforming

conditions

identified

will be implemented

through our corrective

action program. 10/1/98 Update: Programs exist at Palisades

that ensure proper station design attributes

are considered, evaluated, changed and documented.

These programs makeup our overall "Design Control" program. In past months, several programs have been reviewed in various inspections

and routine assessments

such as: * NRC INFORMATION

NOTICE 98-22:"DEFICIENCIES

IDENTIFIED

DURING NRC DESIGN INSPECTIONS" was evaluated

by comparing

the adequacy of our program design controls against other station Design Inspection

identified

concerns.

  • Self assessments

were performed

in areas such as design document control and modification

programs.

  • NRC inspections

and internal NPAD audits in the areas of Engineering

and Technical

Support were performed

in mid 1998 that evaluated

several Palisades

design and configuration

program attributes.

As a result of these and other efforts, "Design Control" Program enhancements

have been identified

and incorporated

into the appropriate

programs.

For example, several changes have been made to design change processes

to better define the applicability

of each distinct process, and to ensure that design change inpuUoutput

requirements

are adequately

addressed.

2

  • ATTACHMENT

B STATUS OF PLANS TO RESOLVE RELATED PROGRAMMATIC

ISSUES AND EVALUATE FINDINGS AGAINST THE 10 CFR 50.54(f) RESPONSE No major programmatic

weaknesses

were identified

in these reviews and program enhancements

are now complete.

To conclude, the Palisades "Design Control" Program is considered

effective.

2. CALCULATION

CONTROL ISSUES: The Design Inspection

issues identified

below reflect weaknesses

in our calculation

control program. Improvements

in our calculation

control program will serve to prevent recurrence

of these conditions.

Inspection

Report Issues: * Required justification

for conclusion

and correct references

to related NRC correspondence

not provided in analysis. (Unresolved

Item 50-255/97-201-04.)

  • Not all analyses revised whenever analytical

inputs or major assumptions

change. (Unresolved

Item 50-255/97-201-07.)

  • Analyses not reflecting

accurate as-built configuration

and system operation, not all interdependent

analyses have been revised together in response to changes, and analytical

design bases do nofagreewith

test acceptance

criteria. (Unresolved

Item 50-255/97-201-08.)

Palisades

60 Day Response:

Prior to the Design Inspection, calculation

control weaknesses

were recognized

and an improvement

plan was implemented.

Over 19,000 calculations

have .been indexed to provide for improved retrievability.

A cross-index

between selected calculations

of record and the documents

that use the results of the calculations

is being developed.

These and other improvements

to our calculation

program serving to prevent recurrence

of the deficiencies

cited above will be made by December 15, 1998. 10/1/98 Update: The identification

of calculations

referenced

in the major design documents

has been completed.

The Calculation

Control Improvement

Project is on target for 3

  • * * ATTACHMENT

8 STATUS OF PLANS TO RESOLVE RELATED PROGRAMMATIC

ISSUES AND EVALUATE FINDINGS AGAINST THE 10 CFR 50.54(f) RESPONSE completion

of the detailed calculation

cross-index

by December 15, 1998. Development

of the computerized

calculation

retrieval

application

and completion

of associated

engineer training will follow in early 1999. 3. SETPOINT CONTROL ISSUES: Station procedures

and guidance to require the use of established

uncertainty

methodology

need to be implemented.

The plan for implementation

should be validated

against weaknesses

identified

in* Unresolved

Item 50-255/97-201-12.

Palisades

60 Day Response:

An instrument

uncertainty

evaluation

methodology

manual has been developed.

Uncertainty

calculations

for Reactor Protection

System and Engineered

Safety Features Actuation

System setpoints

have been performed

Ul?ing .the methodology

manual. Incorporation

of instrument

uncertainty

evaluation

requirements

in procedures, and training select engineers

to perform uncertainty

calculations, will be completed

by December 15, 1998. 10/1/98 Update: As stated in Inspector

Follow-up

Item 50-255/97201-12, station procedures

have been revised to consider use of established

instrument

uncertainty

guidance when developing

test acceptance

criteria and determining

errors for operating

instrument

loops. In addition, a self assessment

of the Setpoint Control Process was performed

with potential

areas for improvement

being evaluated.

4. 10 CFR 50.54(F} RESPONSE:

Evaluate inspection

findings, both specific and programmatic, against the Palisades

response to NRC's October 9, 1996 request for information

pursuant to 1 O CFR 50.54(f) regarding

adequacy and availability

of design bases information.

Palisades

60 Day Response:

After review of the inspection

findings and comparison

to our response to the 1 O CFR 50.54(f) letter regarding

the adequacy and availability

of design basis .4

.. * ATTACHMENT

B STATUS OF PLANS TO RESOLVE RELATED PROGRAMMATIC

ISSUES AND EVALUATE FINDINGS AGAINST THE 10 CFR 50.54(f) RESPONSE information, we have determined

that our response to the 10 CFR 50.54 (f) letter remains complete and accurate.

Improvements

to our design programs, initiated

through our response, will be directly responsible

for resolution

of issues * identified

within the Design Inspection

report. The programs and projects being improved include our Calculation

Control Program, Setpoint Methodology

and Control Program, FSAR Verification

and. Validation

Project, and our Fuse Control Program. * Beyond programmatic

improvements, design basis knowledge

will be further enhanced by the development

of 1 O additional

DB Os and the performance

of. three safety system design confirmations

similar to the NRC's safety system functional

inspections.

To date, four of the new DBDs have been issued and one design confirmation

has been completed.

No additional

programmatic

improvement

efforts have initiated

as a result of actions being taken

to satisfy our 10 CFR 50.54(f) response.

A final review of the adequacy of our response will be completed

by December 15, 1998. 10/1/98 Update: Some of the initiatives

noted in our 60-day response to the Des_ign Inspection

were not part of Palisades

formal response to the NRC's October 9, 1996 request for information

pursuant to 10 CFR 50.54(f) regarding

adequacy and availability

of design bases information.

Our February 6, 1997, 50.54(f) response coneluded

that the Palisades'

design bases information

was adequate, and reasonabie

assurance

exists that: 1) design bases information

has been translated

into operating, maintenance, and testing procedures, and 2) system, structures, and component

configuration

and performance

are consistent

with the design bases. Our 50.54(f) response also referred to specific initiatives

to further strengthen

plant processes

and design basis documentation.

Specifically

noted as * commitments

in the 50.54(f) response were: 1) performing

an FSAR Verification

Project, 2) completing

ten new Design Basis Documents, 3) conducting

one Safety System Functional

Type inspection

per fuel cycle, and 4) updating and re-instituting

use of a Quality Assurance

Requirements

Matrix database.

Other initiatives

to strengthen

plant processes

and design basis documentation

were also undertaken

that were not specifically

included ln the 50.54(f) response 5

  • ATTACHMENT

B STATUS OF PLANS TO RESOLVE RELATED PROGRAMMATIC

ISSUES AND EVALUATE FINDINGS AGAINST THE 10 CFR 50.54(f) RESPONSE such as: 1) implementing

a calculation

control improvement

project, 2) implementing

improvements

in instrument

setpoint uncertainty

methodology, 3) performing

an assessment

of instrument

setpoint control, and 4) performing

an assessment

of the fuse control program. The 50.54(f) response remains complete and accurate.

The response to Attachment

B Item 1 relates to and supports this position.

It should be noted, however, that the 50.54(f) response and its committed

programmatic

initiatives, along with other initiatives

noted above, will not resolve all issues identified

within the Design Inspection

since it is more effective

to resolve certain issues on an individual,

basis. A formal review that evaluates

the Design Inspection

findings against the 50.54(f) response is on target for completion

by December 15, 1998. 6

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