ML17250B199: Difference between revisions
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| issue date = 06/08/1990 | | issue date = 06/08/1990 | ||
| title = Responds to NRC 900509 Ltr Re Violations Noted in Insp Rept 50-244/89-81.Corrective Actions:Improved Battery Load Profile Developed Incorporating Calculational Improvements Contained in Current Industry Std IEEE 485-1983 | | title = Responds to NRC 900509 Ltr Re Violations Noted in Insp Rept 50-244/89-81.Corrective Actions:Improved Battery Load Profile Developed Incorporating Calculational Improvements Contained in Current Industry Std IEEE 485-1983 | ||
| author name = | | author name = Mecredy R | ||
| author affiliation = ROCHESTER GAS & ELECTRIC CORP. | | author affiliation = ROCHESTER GAS & ELECTRIC CORP. | ||
| addressee name = | | addressee name = Martin T | ||
| addressee affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) | | addressee affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) | ||
| docket = 05000244 | | docket = 05000244 | ||
Line 14: | Line 14: | ||
| page count = 37 | | page count = 37 | ||
}} | }} | ||
See also: [[ | See also: [[see also::IR 05000244/1989081]] | ||
=Text= | =Text= |
Revision as of 05:51, 19 June 2019
ML17250B199 | |
Person / Time | |
---|---|
Site: | Ginna ![]() |
Issue date: | 06/08/1990 |
From: | Mecredy R ROCHESTER GAS & ELECTRIC CORP. |
To: | Martin T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
References | |
NUDOCS 9006200487 | |
Download: ML17250B199 (37) | |
See also: IR 05000244/1989081
Text
x REGULATOY INFORMATION
DISTRIBUTZOYSTEM (RIDE)x~x ACCESSION NBR:9006200487
DOC.DATE: 90/06/08 NOTARIZED:
NO;C FACIL:50-244
Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G AUTH.NAME~~~~AUTHOR AFFILIATION
CREDY,R,C.
Rochester Gas&Electric Corp.RECIP.NAME
RECIPIENT AFFILIATION
MARTINET.T.
Region 1, Ofc of the Director SUBJECT: Responds to NRC 900509 ltr re violations
noted in Insp Rept 50-244/89-81.
DISTRIBUTION
CODE: IEOZD COPIES RECEIVED:LTR
1 ENCL 3 SIZE: TITLE: General (50 Dkt)-Insp Rept/Notice
of Violation Response NOTES:License
Exp date in accordance
with 10CFR2,2.109(9/19/72).
DOCKET 05000244 05000244 c RECIPIENT ID CODE/NAME PD1-3 PD RGN1 ERNAL: LPDR NSIC INTERNAL: AEOD AEOD/TPAD NRR MORISSEAU,D
NRR/DLPQ/LPEB10
NRR/DREP/PEPB9D
NRR/DST/DIR
8E2 NUDOCS-ABSTRACT
OGC/HDS2 FILE 01 COPIES LTTR ENCL 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 RECIPIENT ID CODE/NAME JOHNSONFA AEOD/DEIIB
DEDRO NRR SHANKMAN,S
NRR/DOEA DIR 11 NRR/DRIS/DIR
NRR/PMAS/ILRB12
OE~I REG~02 NRC PDR COPIES LTTR ENCL 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1<1 1~~+NG: P3 fo)579+/NOTE TO ALL"RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASCONTACT THE DOCUMENT CONTROL DESK ROOM Pl-37 (EXT.20079)TO ELIMINATE YOUR NAME FROM DISTRIBUTION
LISTS FOR DOCUMENTS YOU DON'T NEED!A D D, TOTAL NUMBER OF COPIES REQUIRED: LTTR 22 ENCL 22
AE;f, Nf e~~~5OAA 55455 ROCHESTER GAS AND ELECTRIC CORPORATION
~89 EAST AVENUE, ROCHESTER, N.Y.14849-PPPI
June 8, 1990 TEEER<04C AREA CODE 7555 546 2700 Mr.Thomas T.Martin Regional Administrator
Region I U.S.Nuclear Regulatory
Commission
475 Allendale Road King of Prussia, PA 19406 Subject: Response to Inspection
Report 50-244/89-81
Safety System Functional
Inspect'ion
--RHR System R.E.Ginna Nuclear Power Plant NRC Docket 50-244 Dear Mr.Martin: This letter provides the initial 30-day response to the Safety System Functional
Inspection (SSFI)of the Residual Heat Removal (RHR)System at the R.E.Ginna Nuclear Power Plant, conducted, between November 6 and December 8, 1989.The NRC letter of May 09, 1990 from Marvin W.Hodges (NRC)to Robert C.Mecredy (RG&E)transmitted
the report for that inspection.
This letter provides the RG&E responses, pursuant to 10 CFR 2.201, to the two violations
issued in conjunction
with the SSFI report.In addition, we are providing schedule information
concerning
the unresolved
issues, including the postulated
flooding of the RHR room, identified
in the inspection
report.Additional
information
will be provided in the 120-day response to the SSFI report.The nuclear industry is going through major upgrade efforts involving configuration
management
and design basis documents.
RG&E is not alone in recognizing
the benefits of these improvements
and has been proceeding
with these efforts.On March 6, 1990 RG&E made a formal presentation
to NRC Region I staff and on March 27, 1990 made a presentation
to NRR regarding our configuration
management
program.We have completed three pilot system design basis documents and are reviewing them to determine the optimal specification
for the overall design basis document program for the remaining plant systems.In addition, RG&E has developed a separate program to provide further assurance that all design basis information
and, commitments
which may have been relied upon by the NRC are captured.The objective of the NRC SSFI of the RHR systems was to assess the capability
of that system to perform its design basis functions.
As part of that inspection, the SSFI team assessed the overall design control program and other work processes used by RG&E.The review of these programmatic
aspects was far broader than the RHR system.Special emphasis was placed upon the engineering
processes and their interfaces
with other activities.
900b200487
900b08 PDR ADOCK 05000244 8 PNU ggIJ t~fP(
2The primary result of the SSFI was that no situations
were identified
that would prohibit the RHR system from performing
its intended functions under normal and design basis accident conditions.
As would be expected from an SSFI of any nuclear power plant, and in particular
one of the early SEP plants, the SSFI identified.
areas where improve-ment is warranted.
Two Severity Level IV violations
were cited., and ten specific unresolved
items were documented.
The NRC letter of May 09, 1990 requires that the violations
be addressed, pursuant to 10 CFR 2.201, within 30 days.The letter also requests that RG&E provide its evaluation
of the specific unresolved
items and planned actions, within 120 days.In addition, the NRC letter requests that RG&E also provide schedule information
regarding the actions to address the unresolved
items, within 30 days.The schedules requested are exclusive of unresolved
item 89-81-11, Engineering
Assurance, for which a response was requested in 120 days.k Responses to two violations
identified
ar'e provided as Enclosures
A&B to this letter.The first violation involved.not maintaining
an up-to-date load profile for the batteries.
The actual capability
of the batteries was not an issue, only the adequacy of the testing.RG&E had already reached a state of full compliance
on this matter when the SSFI report was received.>>The second violation cited had two parts.The first part involves having not already developed a periodic testing program for the molded case circuit breakers.The second part involves not having an explicit acceptance
criterion in the test procedure for the setpoints of the dc undervoltage
alarm relays.Although a generally accepted periodic test method for molded case circuit breakers is not available in the industry today, we choose not to take issue with this violation.
The industry is currently examining the need for and/or requirements
for molded case circuit breaker testing.RG&E will implement, when.available, those testing methods and requirements
endorsed by the industry.With regard to the acceptance
criterion for the undervoltage
relay setpoints, we had already reached a state of full compliance
on this matter when the SSFI report was received.In addition, on our own initiative, we have expanded this concern to include the ac undervoltage
relays for the safety buses.In addition to these violations, NRC also identified
ten unresolved
items.The identification
of these items is contained in Enclosure C.Several of these unresolved
items have already been completed and several more are in process.During the RG&E review of the SSFI report, management
recognized
that many of the unresolved
items were examples of broader, underlying, programmatic
concerns.Many of these concerns focused on engineering
functions and, controls'.
Because RG&E understands
the importance
of resolving the programmatic
and management
issues as well as the specific items cited by the NRC, we are developing
a systematic
approach to address both types of concerns.This approach is a two-part, parallel effort.The first part focuses on the management
processes in a disciplined
manner, while the second.part focuses on the resolution
of the specific unresolved
items.
P
To begin the review of the broader concerns, we have re-reviewed
the SSFI report and the cited issues, and have categorized
them into general topical areas.For example, unresolved
item 89-81-05 involves not having a mechanism to assure that design calculations
are main-tained up-to-date.
We see this specific item as being part of a more general area called design control.Enclosure D is a preliminary
categorization
of the unresolved
items into the general topical areas.In addition, RG&E is initiating
a more detailed review of the work processes and their controls for each of the general areas which contain significant
identified, weaknesses.
This review will encompass identifying
the cause of the violations, as well as the unresolved
issues, identified
by the SSFI report.Enclosure E contains the schedular information
as requested by the staff.We have separated this schedule information
into two catego-ries: resolution
completed and scheduled for resolution.
RG&E has resolved items 89-81-04, 06, 07A, and 10 as identified
in Enclosure C.In particular, RG&E has promptly resolved the issue regarding flooding of the RHR pump room.The UFSAR has been updated, and the EOPs and training documents have been revised.A detailed account of those actions taken to resolve the items identified
above are con-tained in Enclosure E.RG&E believes that the approach outlined in this letter assures proper and complete resolution
of the specific issues identified
as well as the more programmatic
issues discussed.
Very truly yours, Robert C.Mec e y Division Manager Nuclear Production
GAHN108 Enclosures
xc: U.S.Nuclear Regulatory
Commission (original)
Document Control Desk Washington, D.C.20555 Allen R.Johnson (Mail Stop 14D1)Project Directorate
I-3 Washington, D.C.20555 Ginna NRC Senior Resident Inspector
0
ENCLOSURE A Response to Notice of Violation 50-244/89-81
Violation 1
.Ins ection Re ort 44/88-81 VIOLATION 1:STATEMENT OF VIOLATION-
10 CFR 50, Appendix B, Criterion III, requires in part that measures be established
to ensure that applicable
regulatory
requirements
and design bases are translated
into specifications
and procedures.
These measures shall provide for verifying the adequacy of design by performance
of design reviews.Ginna Station Quality Assurance Manual, Section No.11,"Test Con-trol," requires that engineering
establish design test requirements
and that testing be performed in accordance
with approved procedures
which incorporate
the requirements
and acceptance
criteria contained in applicable
Technical Specifications
and regulatory
requirements.
Contrary to the above, on November 15, 1989, the design reviews for Engineering
Work Request (EWR)3891 were inadequate
in that the EWR did not establish the battery load, requirements
thereby resulting in a battery load.profile used during the service test.not reflecting
the design basis load requirements.
This is a Severity Level IV Violation (Supplement
1).ACCEPTANCE
OF VIOLATION:
RG&E agrees that it did not update the battery load profile as part of EWR 3891.DISCUSSION:
The purpose of EWR 3891 was to replace the batteries because they were nearing the end of their service life and, while replacing them, to increase the capacity margin.EWR 3891 did not include an updat-ing of the battery test profile because it had been determined
that no large loads had been added.to the battery since the original load profile had been developed.
The battery load profile was based upon the original Westinghouse
design data.That information
was consistent
with industry practice at the time it was developed.
Analytical
techniques
were not as sophisticated
as those in use today.Rather than explicitly
quanti-fying such factors as momentary loads and.the load starting currents, it was general practice to provide additional
battery sizing based upon experience
and engineering
judgement.
Today's standards (such as IEEE standard 485)suggest a more refined, more precisely quanti-fied analysis.The actual battery capacity was sufficient
to provide its safety functions.
The battery has been shown to have adequate capacity as confirmed by a physical test.Although there is no requirement
for the Ginna Nuclear Power Plant to incorporate
all newly-developed
industry standards, we believe it prudent to use the current industry standards for developing
revised battery load profiles, and have done so.A-1
CORRECTIVE
STEPS T A preliminary
analysis, performed during the inspection, demonstrated
that the battery size is adequate.The revised battery size calculation
had, been finalized subsequent
to the NRC inspection
and prior to the receipt of the inspection
report, which confirms that the battery size is not a concern.An improved battery load profile has been developed which incorpo-rates calculational
improvements
contained in current industry standard IEEE 485-1983.The upgraded battery load profile (Design Analysis EWR 3341"Sizing of Vital Batteries", dated March 12, 1990)has been transmitted
by Engineering
to the plant staff, and the battery testing procedures'PT-10.2
and PT-10.3, Battery Service Tests)have been revised.The batteries were tested during the recent outage using the revised procedures.
The results demonstrated
the adequacy of the battery capacity.CORRECTIVE
STEPS TO BE TAKEN TO PREVEBVP REClJRRENCE:
The applicability
of this violation has been broadened by RG&E to assure that not only the important dc electrical
loads are analyzed and tested, but also that the important ac electrical
loads which may impact the operation of the plant emergency diesel generators
are identified
and tracked.We have implemented
an electrical
load'rowth
program as described under unresolved
item 89-81-05.DATE WHEN FULL COMPLIANCE
WILL BE ACHIEVED: Engineering
established
updated battery load requirements.
The battery test procedures
have been revised and the batteries have been tested using the new procedure.
These actions were completed prior to the receipt of the NRC inspection
report.RG&E is in full compli-ance.A-2
ENCLOSURE B Response to Notice of Violation 50-244/89-81
Violation 2
I P
~RG&E/Ginna
Ins ecti Re ort 50-244/89-81
VIO N 2: STATEMENT OF VIOLATION:
R.E.Ginna Technical Specifications Section 6.8.1 requires that written procedures
be established
and.implemented.
for activities
such as surveillance
and testing activities
of safety-related
equipment.
Ginna Station Quality Assurance Manual,Section II,"Test Control," establishes
the requirements
for establishing
and implementing
test programs to demonstrate
that safety-related
systems and components
will perform satisfactorily.
Furthermore, this section requires that testing shall be performed in accordance
with written procedures
which incorporate
acceptance
criteria.Contrary to the above, on December 9, 1989, Class 1E 480V ac molded case circuit breakers have not been subjected to scheduled periodic testing.Furthermore, there is no established
acceptance
criteria for testing the dc undervoltage
relay al'arms in Procedure PT-11,"60-Cell Battery Banks'AB'his is a Severity Level IV Violation (Supplement
1).ACCEPTANCE
OF VIOLATION:
RG&E agrees that the periodic testing program of safety-related
equipment at the Ginna.Nuclear
Power Plant does not currently include molded case circuit breakers.RG&E also agrees that the Ginna periodic test procedure PT-11"60-Cell Battery Banks'AB'" did not specify an acceptance
criterion for the setpoint of the dc undervoltage
relay alarms.This violation has two parts which are addressed separately
below: Part 1: Molded Case Circuit Breaker Testing DISCUSSION:
Molded case circuit breakers are designed for nuclear and non-nuclear
applications.
This type circuit breaker is sealed.and does not include design features to test all the capabilities
of the breaker beyond functional
tests.RG&E realizes the importance
of assuring proper operation of these breakers.RG&E has not been lax in its attention to the importance
of testing molded case circuit breakers.This problem was self-identified
by RG&E and was incorporated
into the RCM program.On our own initiative, we developed and implemented
receipt-inspection
testing for all new molded case circuit breakers at Ginna.We have also performed testing on molded, case circuit breakers in an effort to determine their characteristics.
Three years ago, RG&E performed special testing of all of its exist-ing magnetic only, molded case circuit breakers at Ginna Station on a special one-time basis.Successful
operation has indicated no known degradation.
B-1
While the functioniO~
of molded case circuit hkers is important to safety and while there is an NRC requirement
for a test program to assure that safety-related
structures, systems and.components
will perform satisfactorily, there is no specific requirement
to test periodically
every piece of equipment.
As stated.in Appendix B, Criterion XI,"The test program shall include, as appropriate, operational
tests...of structures, systems and components." The term"as appropriate" is applicable
and includes the availability
of appropriate
test methods.Molded case circuit breakers are not designed for in situ testing and would require determination
and retermination
to perform the testing.The vendors of this equipment have also not made recommendations
for periodic testing.Because of generic applicability, periodic testing for molded case circuit breakers has been an industry-wide
issue and no generally accepted test method has been developed at this time.The nuclear industry has responded to the NRC through NUMARC concern-ing molded case circuit breaker testing and RG&E is pursuing this in conjunction
with this effort.CORRECTIVE
STEPS TAKEN: RG&E is continuing
to work toward developing
appropriate
test methods for molded case circuit breakers, as part of the Reliability
Centered Maintenance (RCM)program.The Ginna Nuclear Power Plant is one of the two"pilot plants" in the nation for the EPRI sponsored RCM program.CORRECTIVE
STEPS TO BE TAKEN TO PR1DGQFZ RECURRENCE-
The industry is currently examining the need for, and benefits of, molded case circuit breaker testing.RG&E will continue to work closely with the industry and EPRI to determine appropriate
test methods and.requirements.
DATE WHEN FULL COMPLIANCE
WILL BE ACHIEVED: Although RG&E does not consider this a compliance
matter, RG&E will implement, when available, those testing methods and requirements
endorsed by the industry.Part 2: Undervoltage
Relay Alarm Acceptance
Criteria CORRECTIVE
STEPS TAKEN: The periodic test procedure PT-11"60-Cell Battery Banks'AB'" has been revised to explicitly
define the acceptance
band/criterion
for the dc undervoltage
alarm relays.The dc relays have subsequently
been calibrated
and tested.The relays have been verified to perform within the specified acceptance
criterion.
B-2
~1 0
.CORRECTIVE
STEPS TAKEN TO PREVIXT RE E: The applicability
of this violation has been broadened by RG&E to assure that not.only the test procedures
for dc undervoltage
alarm relays have explicit acceptance
criteria, but also that the test procedures
for the ac undervoltage
relays for the safeguards
buses have explicit acceptance
criteria.The test procedure, PT-11 for the dc undervoltage
alarm relays has been revised and PT-9.1 for the 480V ac safeguards
buses is being revised to provide explicit acceptance
criterion.
DATE WHEN FULL COMPLIANCE
WILL BE ACHIEVED: The test procedure for the dc undervoltage
alarm relays has been revised to provide an explicit acceptance
band/criterion.
This action was completed prior to the receipt of the NRC inspection
report.RG&E is in full compliance.
B-3
0 t'
ENCLOSURE C Identification
of Specific Unresolved
Items Note: The statements
of issues have been directly extracted from the SSFI report.In a few instances the issues have been condensed and paraphrased.
r 0 0 e
.89-81-01 Service r Single Failure Suscept lity Potential loss of cooling water[flow]to both emergency diesel generators
during or following a seismic event.The cooling water for the water jacket heat exchanger and lube oil heat exchanger discharges
through a common non-safety
non-seismic
10-inch discharge pipe.The cooling water discharge pipe would have to fail[or has been postulated
by the NRC SSFI team to fail]so as to prevent[block/pinch
off]the flow of the service water.89-81-02 Resolution
of Safety Concerns The licensee was unable to provide the team with a documented
or verifiable
process available at RG&E that addresses how safety concerns raised outside the normal engineering
process are brought to the attention of the Nuclear Safety and Licensing staff and resolved.89-81-03 RHR Pump NPSH A consultant
independently
evaluated the available NPSH during post-accident recirculation
mode from the containment
sump and, a prelimi-nary result indicates that there may be some modes of operation of the RHR pumps under which adequate NPSH is not available.
Licensee is evaluating
the validity of these modes and the probability
of occurrence.
Licensee is also evaluating
the possibility
that the consultant's
analytical
model was too conservative.
89-81-04 Class 1E Battery Testing Failure to test the batteries with a load profile which truly repre-sented the load demand on the battery is considered
a violation of 10 CFR 50, Appendix B, Criterion III.89-81-05 Electrical
Load Growth Control Program RG&E does not have a mechanism to assure that plant calculations
affected by modifications
are updated to ensure that they are main-tained up-to-date
and accurate.The design process provides guidance to engineers to review the system capacity and other attributes, but the guidance addresses only specific modifications
as they are performed.
There is no formal load tracking system to ensure that system capacity is reviewed for the integrated
effect of several modifications
instead of just one.The licensee stated that an on-line program to capture electrical
load growth and update affected calculations
would be developed.
89-81-06 Molded Case Circuit Breakers and, Undervoltage
Relay Alarms Failure to periodically
test the molded case circuit breakers and not establishing
an acceptance
criteria for the undervoltage
relay alarms are a violation of facility Technical Specifications 6.8.1, which requires testing of safety-related.
components
in accordance
with established
procedures.
.89-81-07A
Calibrate of Control Room Instrume The control room dc voltmeters
are not calibrated
on a periodic basis to ensure reliable system voltage indication
to operators.
89-81-07B Control Room P&IDs Piping and Instrument
Diagram (P&ID)updates and Design Change Requests (DCRs)posted in the control room were reviewed by the team.It was noted that the RHR system P&ID (33013-1247)
did not reflect the current valve position configuration
for the RHR system.Also, the existing DCRs outstanding
against this drawing could not be used.to derive the correct valve positions in that DCRs 1247-4, and 1247-5 had not been approved by RG&E Engineering
and did not reflect the current position of valve 822B.Processing
of DCRs does not always occur in a timely manner such that the control room P&IDs can be immediately
updated.Plant operations
organization
makes permanent changes to system valve positions, there is not an immediate markup or annotation
made on the effected draw-ings.The team noted that permanent changes to valve positions in system operating procedures
are occurring without the prior concurrence
of RG&E engineering.
UFSAR, sections 5.4.5.3.5 and 5.4.5.2, refers to two remotely operat-ed valves which can be utilized to isolate an RHR loop from outside the pump room.The system walkdown and the upgraded P&IDs indicate that there is no longer any method available to isolate an RHR loop remotely (i.e., via reach rods).Although this information
has been removed from the RHR P&ID, there is no identified
punchlist item to delete this information
from the UFSAR.The team noted that uncontrolled
training material (Lesson Texts)have not been updated to reflect system changes accomplished
during the last outage.There is no station requirement
to maintain this training material current.The inspection
team considers that making this type of information
available to control room operators in such an uncontrolled
manner represents
a notable program weakness.The lack of timely operating information
updates for control room use is considered
an unresolved.
item.89-81-08 Equipment Environmental
Qualification
Evaluation
The NRC questioned.
the basis for the assumption
that RHR pump seal failure will occur after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.The NRC requests RG&E to sub-stantiate the method of detecting any leak in the RHR pump room if the pump seal were to fail before the stated 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.C-2
The safety relief valve test procedures
contain general and minimal instructions
for performing
the relief setpoint test.Standard test practices are not always performed.
or documented.
As written, the test procedure requires only one successful
setpoint test.Data from relief valve testing has been recorded inaccurately
and.inconsistent-
ly in some cases.The NRC concluded that RG&E should formalize test procedures
instructions
and data recording requirements.
During the on-going procedure upgrade effort, RG&E should assure that valve test procedures
incorporate
all new (1986)ASME Code Section XI, IWV-3512, and ANSI/ASME OM-1-1981 requirements
for safety relief valves.In particular, more than one successful"pop test" at the designated
lift pressure should be performed and the results comp-letely and accurately
documented.
Valve setpoint and leak testing should also be performed with the allowable specification
listed in the procedure.
Valve test results and data should accurately
reflect'he results of all test activities.
RG&E should also consider the benefits of adding other periodic valve tests such as the as-found relief lift setpoint, valve accumulation, and.valve capacity.89-81-10 Translation
of FSAR Requirements
into Operating Procedures
The Ginna UFSAR contains"operational" information
and data which the inspectors
determined
to be invalid and, without a supporting
design basis., Specifically, Section 5.4.5.3.5 states that in the event of a 50 gpm RHR pump seal leak and loss of both pump room sump pumps, operators have 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to isolate the leak before the RHR pump motors become flooded.The team determined
that a 50 gpm leak into the pump room, with two failed sump motors;cannot be sustained in the RHR pump room for four hours before flooding the pump motors.It was suggested.that the four hour allowance was originally
intended just to indicate a rough system margin for coping with gross leakage in the pump pit.The team was unable to find any consideration
of this in any of the available design documents associated, with the RHR system.It also could not be found in any of the system operating or emergency procedures.
The alarm response procedure for the high sump level alarm requires control room operators to dispatch an auxiliary operator to investigate
possible pump room flooding, however there is no reference to maximum time limit to isolate a leaking RHR train if necessary.
The team reviewed the instrumentation
devices available to control, room operators which would indicate RHR leakage in the pump room.The only known indication
would be from a high level sump alarm.However, the sump alarm instrument
is not qualified for service in a harsh environment.
Operating procedures, emergency procedures, and operator.training material do not reflect the limiting design basis of the system.The apparently
unsupported
4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> flooding limit is considered
an un-resolved item pending verification
of the value by the licensee or correction
of the UFSAR.C-3
O 89-81-11 Engineer'ssurance
The design control measures as implemented/practiced
by the licensee's
engineering
department
were weak, and did not favorably compare to good engineering
assurance practices generally accepted in the industry.There was lack of consistency
in the implementation
of approved engineering
procedures
among the various departments
and engineering
management
did not appear to be cognizant of this incon-sistency.There was a lack of formal design interface control, lack of control over external communication
with design consultants, and a lack of control over design documents/modification
packages during the development
and implementation
phase.C-4
ENCLOSURE.D Preliminary
Categorization
of Issues Note: The categories
contained in Enclosure D were selected topics in 10 CFR 50 Appendix B and other sources.To begin the review of the broader concerns, we have reviewed the SSFI report and the cited issues, and have categorized
them into general topical areas.For example, unresolved
item 89-81-05 involves not having a mechanism to assure that design calculations
are maintained
up-to-date.
We see this specific item as being part of a more general area called design control.Enclosure D is a preliminary
categorization
of the unre-solved items into the general topical areas..It is currently planned to categorize
all the concerns identified
in the inspection
report.
DESIGN CONTROL General Control of Design Inputs Control of Design Process SSFI URI 89-81-05: SSFI URI 89-81-08: Electrical
Load Growth Con-trol Program Equipment Environmental
Qual-ification Evaluation
Control of Design Outputs SSFI URI 89-81-07B:
Control Room P&IDs Control of Design Interfaces
and.Coordination
Control of Design Changes Design Reviews/Engineering
Assurance SSFI URI 89-81-05: Electrical
Load Growth Control Program SSFI URI 89-81-11: Engineering
Assurance Specific Design Concerns SSFI URI 89-81-01: Service Water Single Failure Susceptibility
SSFI URI 89-81-03: RHR Pump NPSH PROCEDURES
SSFI URI 89-81-09: Safety Relief Valve Testing DOCUMENT CONTROL SSFI URI 89-81-07B:
Control Room P&IDs
0
ORGANIZATIONAL
.ACES SSFI URI 89-81-02: Resolution
of Safety Concerns SSFI URI 89-81-07B:
Control Room P&IDs SSFI URI 89-81"10: Translation
of FSAR Require-ments into Operating Proce-dures HANDLING OF SAFETY CONCERNS SSFI URI 89-81-02: Resolution
of Safety Concerns SURVEILLANCE
TESTING MAINTENANCE
Calibration
of Control Room Instruments
SSFI URI 89-81-09: Safety Relief Valve Testing D-2
ENCLOSURE E Resolution
of Specific Issues Note: We have separated.
the schedule information
contained in this enclo-sure into two categories:
resolution
completed, and schedule for resolution.
Listed first are those items for which RG&E has complet-ed resolution.
Those measures taken by RG&E are identified.
Some of the unresolved
items listed cannot be adequately
resolved, without addressing
the broader more programmatic
issues such as design control and engineering
assurance and require more time to resolve than the specific items.The schedules provided for some items may change as RG&E further identifies
the underlying
concerns.An updated schedule will be provided in the 120 day response.
0
~Resolution
Com lete~89-81-04 Class 1E Battery Testing This item was resolved prior to receipt of the SSFI report.see Enclosure A for actions taken for resolution.
Please 89-81-06 Undervoltage
Relay Alarms and Molded Case Circuit Breakers Please see Enclosure B for actions taken for resolution.
89-81-07A Calibration
of Contxol Room Instruments
This item was resolved prior to receipt of the SSFI report.The actions taken to resolve this issue include: 1)Calibration
of all control room dc bus voltmeters
during the recent refueling outage (the voltmeters
were found to be within the specified acceptance
criteria).
2)All dc bus voltmeters
are now calibrated
per Calibration
Proce-, dure CP-514 on an annual basis.3)All emergency diesel generator and various secondary system power meter calibrations
have been added to the CP-500 series procedures, and the meters were calibrated
during the 1990 refueling outage.89-81-10 Translation
of the FSAR Requirements
into Operational
Procedures
This item was resolved promptly.The actions taken to resolve this issue include: 1)Performance
of a reanalysis, during the SSFI inspection, which determined
that operators have two hours to respond.(Design Analysis, 10CFR50.59
Safety Evaluation, NSL-0000-015, Rev.0, dated December 8, 1989, Residual Heat Removal Leakage Provi-sions.)T 2)Update of UFSAR sections 5.4.5.3.5, 5.4.5.2 and 6.3.3.8, submit-ted as part of the UFSAR update on December 16, 1989.3)Revision of Training System Description
RGE-25 during the inspection.
4)Revision of EOPs prior to receipt of the inspection
report.(Procedure
E-1, Loss of Reactor or Secondary Coolant, Step 18 was added and ES-1.3, Transfer to Cold Leg Recirculation, a note before Step 9 was added.)
.Schedule for Resolu.n 89-81-01 Service Water Single Failure Susceptibility
As noted in the inspection
report, the failure of the 10 inch dis-charge line in a manner which would stop service water flow to the diesel generators
is a low probability
event.This event is also beyond the design and licensing basis of the plant.Nevertheless, RG&E plans to further evaluate the potential risk of this scenario during the PRA/IPE effort.Our IPE is currently scheduled to be submitted in the third quarter of 1991.89-81-02 Resolution
of Safety Concerns An interim process for handling safety concerns is under-development
and will be discussed in our 120 day response.89-81-03 RHR Pump NPSH Documentation
of the analysis findings is scheduled to be completed by December 31, 1990.In addition, RG&E plans to consider this matter in the PRA/IPE.89-81-05 Electrical
Load Growth Program RG&E has implemented
an interim process for all modifications
to perform the following actions: Current system loadings for the dc batteries have been estab-lished in Design Analysis, EWR 3341, Sizing of Vital Batteries, and for the diesel generator loads in Design Analysis, EWR 4136, Diesel Generator Loading.2)An Electrical
Engineering
Design Guide, Electrical
Interface Checklist EDG-15D, Rev.0, is being implemented
on all modifica-tions which requires identification
of load changes to the dc batteries and the diesel generator ac loads.3)A process controlled
by Electrical
Design Guide, Design Verifi-cation Model EDG-15B, Rev.0, has been established
within the Electrical
Engineering
Design Verification
Group which updates the loading data for the impacted power supply and determines
the remaining capacity margin for ac and dc loads.We are taking actions to integrate this process into the appropriate
Engineering (QE)procedures.
We anticipate
completion
of these actions by the date of our 120 day response.-
89-81-07B Control Room P&IDs RG&E has considered
the examples identified
by the staff which resulted in the staff's conclusion
that information
updates for control room use are not implemented
in a timely manner.RG&E has resolved several of the examples identified.
These include: E-2
C 4'C 0
.1)RG&E has impleilhted
improved controls inDrawing Change Request (DCR)process.RG&E has assigned Z" Station Engineer with responsibi
lity for tracking and processing
all DCRs.~~2)The UFSAR has been reviewed to assure that the appropriate
information
with regard to the isolation of the RHR pump seal is correct.3)RGGE has revised the lesson text to reflect the revised RHR'pump seal leakage time limitation
of two hours.An interim process for enhancing the update process for control room information
is currently under review and will be discussed, in the 120 day response.89-81-08 Equipment Environmental
Qualification
Evaluation
The passive failure'of a RHR pump seal is assumed to occur at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, consistent
with SRP 15.6.5.The consequences
of this assumed passive failure, concurrent
with the assumed design basis LOCA, was evaluated, by the NRC during the review of SEP Topic XV-19 and found to be acceptable.
Nevertheless, RGGE plans to further evaluate this scenario during the PRA/IPE effort with its attendant requirement
to perform an internal flooding analysis.Our IPE is currently sched-uled to be submitted in the third quarter of 1991.The results of this evaluation
will determine if the upgrade of the sump level switches to a safety-related
status is recommended.
89-81-09 Safety Relief Valve Testing and, Documentation
RGGE has commit/ed to incorporate
ASME Code Section ZI-IWV-3512
(1986)and implement ANSI/ASME OM-1-1987 as part of the IST Program Upgrade.Procedure changes to incorporate
these requirements
were completed.
prior to receipt of the SSFI report.RGfiE will have completed.
all testings under these new requirements
by December 31, 1994.E-3
e