ML17261A821
| ML17261A821 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 01/06/1989 |
| From: | Mecredy R ROCHESTER GAS & ELECTRIC CORP. |
| To: | Russell W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| NUDOCS 8901200046 | |
| Download: ML17261A821 (30) | |
See also: IR 05000244/1988022
Text
A.C CEMRATZD D1SIKBUTJON
DEMONIST'RA,T10N
SYSTEM REGULATORY
INFORMATION
DISTRIBUTION
SYSTEM (RIDS)ACCESSION NBR:8901200046
DOC.DATE: 89/Ol/06 NOTARIZED:
NO DOCKET FACIL:50-244
Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244~~AUTH.,NAME
AUTHOR AFFILIATION
MECREDY,R.C.
Rochester Gas a Electric Corp.RECIP.NAME
RECIPIENT AFFILIATION
RUSSELL,W.T.
Region 1, Ofc of the Director SUBJECT: Responds to violations
noted in Insp Rept 50-244/88-22
RGE t denies violation.Supprorting
info encl.DISTRIBUTION
CODE: IE01D COPIES RECEIVED:LTR
t ENCL t SIZE:/TITLE:.General (50 Dkt)-Insp Rept/Notice
of Violation Response NOTES:License
Exp date in accordance
with 10CFR2,2.109(9/19/72).
D 05000244 S RECIPIENT ID CODE/NAME PDl-3 PD INTERNAL: AEOD DEDRO NRR/DEST DIR NRR/DLPQ/QAB
10 NRR/DREP/EPB
lo NRR/DRIS DIR 9A NUDOCS-ABSTRACT
OGC/HDS2 RGNl FILE 01 COPIES LTTR ENCL 1 1 1 1 1 1 1 1 1 11 1 1 1 1 1 1 1 1 1 RECIPIENT ID CODE/NAME STAHLEiC AEOD/DEIIB
NRR SHANKMANiS
NRR/DLPQ/PEB
11 NRR/DOEA DIR ll NRR/DREP/RPB
10 NRR/PMAS/ILRB12
MANcJ G ILE 02 COPIES LTTR ENCL 1 1 1 1 1 1 1 1 1 1 2 2 1 1 1 1 1 1/h TERNAL: LPDR NSIC 1 1 1 1 NRC PDR REST MARTIN,D 1 1 1 1 R I D NOXE'IO ALL'RIDS" RECZPIENIS:
PIZASE HELP US'IO REDUCE WASTE!CONTACT GHE DOQ3MEPZ CDÃHRL DESK, IZST8 PGR DOCUMEHZS YOU DOMiT NEEDf D S TOTAL NUMBER OF COPIES REQUIRED: LTTR 24 ENCL 24
your$Ta1c ROCHESTER GAS AND ELECTRIC CORPORATION
~eg EAST AVENUE, ROCHESTER, N.Y.14649.0001
January 6, 1989 Ici.cpHoNc
AecA cooc 7ie 546.2700 Mr.William T.Russell Regional Administrator
US Nuclear Regulatory
Commission
Region I 475 Allendale Road King of Prussia, PA 19406 Subject: Inspection
Report 50-244/88-22
Notice of Violation 88-22-02 R.E.Ginna Nuclear Power Plant Docket No.50-244 Dear Mr.Russell: In accordance
with 10 CPR 2.201, RGGE provides Attachment
I to this letter as our response to the Notice of Violation.
RGGE denies the violation as set forth by the NRC and has presented information
supporting
this denial.In addition, Attachment
II addresses the Staff's comments related to this matter as contained in Inspection
Report 88-22 and Attachment
III provides 50.59 safety evaluations
for the tygon tubing and hot water system additions.
All attachments
to this letter provide considerable
information
concerning
these issues.We would be pleased to meet with you and your staff to discuss the issues should you believe this would enhance communications
on this or related issues.Very truly yours, Robert C.-cred General Manager Nuclear Production
GJWK016 Attachments
xc: U.S.Nuclear Regulatory
Commission (original)
Document Control Desk Washington, DC 20555 xc: Ginna Senior Resident Inspector 89pg~pppAb
S~pppp 8qpipb PDR~DOCK pal 6 gP
Violation"10 CFR 50, Appendix B, section III requires, in part, measures shall be established
to assure that appropriate'uality
standards are specified and included in design documents, and deviations
from such standards are controlled.
The Quality Assurance Manual Ginna Station, section 3, step 3.1.3 requires modifications
involving a change to the facility's
described in the Updated Final Safety Analysis Report (UFSAR)have a safety evaluation
in accordance
with 10 CFR 50.59."Contrary to the above, on October 5, 1988 a modification
involving.
a change to the Condensate
Storage Tank, described, in Chapter 10 of the UFSAR as the main source of water for the Auxiliary Feedwater system, was installed without a safety evaluation
in accordance
with 10 CFR 50.59."~Res nse RG&E denies this violation.
RG&E agrees that the Ginna QA manual specifies that a 50.59 evaluation
should be performed for any facility modification
involving a change to the facility as described in the UFSAR.However, we do not agree that the addition of the Tygon tubing to the-Condensate
Storage Tanks constitutes
such a change.What is shown in the UFSAR is a 3/4 inch sampling line which is isolated by closed manual valve 4318A (UFSAR Figure 10.7-5).This valve is locked closed.This configuration
has not been changed by the addition of the Tygon tube.The Tygon tube has been added downstream
of this valve and does not affect the CSTs as explicitly
described in the UFSAR.Even if we consider (and we did)the applicability
of 10 CFR 50.59 to items implicitly
described in the UFSAR, current draft industry guidance (which has been reviewed by the NRC Staff)defines this implicit inclusion or description
as follows: "If the change alters the design, function, or method of performing
the function of the larger structure, system, or component[in this case the CSTs]as described, in the SAR then a safety evaluation
is required." ((NUMARC/NSAC
Draft Guidelines
for 10 CFR 50.59 Safety Evaluations)(December 1988)}.Because the Tygon tube has been installed beyond a manual locked closed valve it was and still is understood
that this modification
does not alter the CSTs design function or method of performance.
In addition, any failure of the Tygon tubing cannot interact with the CSTs or affect any of the surrounding
equipment.
When this modification
was made by RG&E, the appropriate
consideration
was given to the governing requirements
of 10 CFR-50.59.Appropriate
screening criteria were applied, to determine the applicability
of 10 CFR 50.59.Although the'documentation
maintained
for screening this modification
and concluding
that 50.59 did not apply was brief, good engineering
judgment was implemented
and documentation
was provided.RG&E believes that the documentation
supporting
this modification
adequately
addresses the safety issues.Because we believe in the importance
of properly applying the 50.59 requirement, we have continued to institute additional
programmatic
guidance on implementing
10 CFR 50.59.I Additional
Review A more detailed review of the addition of the Tygon tubing has been documented (see Attachment
III).Even under-the scrutiny of a 50.59 safety evaluation (as enclosed), a Unreviewed
Safety Question (USQ)does not result.Pro rammatic I rovements As discussed in Attachment
II, a programmatic
approach to, 50.59 has been and is continuing
to be developed at Ginna.Procedures
have been written, and more comprehensive
procedures
are being developed, to ensure that appropriate
screening criteria forms are filled out in accordance
with the Ginna 50.59 program.These screening forms will provide an adequate basis for applying 50.59 on a case-by-case
basis, and will provide for an adequate documentation
of the basis for the conclusions
of the applicability
screening in those cases where 50.59 does not apply.In addition, RG&E is instituting
training programs on the implementation
of 50.59 to make certain that all personnel involved in performing
such evaluations
understand
the RG&E 50.59 program and the technical considerations
involved in applying the programmatic
guidance.The final form of our program will incorporate
the guidance resulting from present NUMARC-NRC
discussions
on industry-wide
implementation
of 10CFR50.59
programs.Date of Pull C liance RG&E believes it is currently in compliance
with 10 CFR 50.59 and with the Ginna Quality Assurance Manual as it relates to the issues identified.
ATTACHMML'I
I.Introduction
In addition to addressing
the Notice of Violation, we are responding
to some of the Staff's associated
concerns raised within the inspection
report itself.We would like the Staff to understand
the status of our programs, including the critical review of our modification
process and the institution
of our 50.59 process.We believe that it is evident that RG&E is being proactive and that we have a clear understanding
of not only the concerns expressed in this inspection
report, but the evolving concerns of the past few years that relate to these latest issues.It is our intent that the Staff understand
that we have not been idle for 20 months, but have made strides in developing
comprehensive
programs that not only address concerns in a specific manner, but look at the broader picture and can be seen as an overall improvement.
II.50.59 Pro ram Im rovements In the body of Inspection
Report 88-22, the NRC expressed a concern that RG&E's failure to perform Safety Evaluations
has been an NRC identified
concern for more than 20 months and is indicative
of programmatic
weakness in the control of station modifications.
Ginna Station procedures
are clear in the requirement
to develop a Safety Evaluation
in support of modifications.
In addition to reviewing physical changes, RG&E has a detailed screening program and 50.59 guidance for revisions made to procedures.
The safety evaluation
process for both modifications
and procedure changes has been improved through the continuing
development
of detailed guidance.This guidance documents the impacts that each evaluator must consider for a specific type of change.Specific examples for these changes are also provided.RG&E has taken a proactive approach to dealing with the 10 CFR 50.59 process.In many respects, this has been difficult because of-the evolving nature of NRC/industry
guidance in this area.This is evidenced, by the fact that even the most recent NUMARC guidance is still considered
a draft.Even an industry-wide
attempt to develop generally accepted definitions
under 50.59 has been a long involved process, one in which the Staff is still participating.
Despite this, RG&E has been active in this arena and will continue to be so.We are greatly concerned that the Staff perceives that our modification
program.is weak from a programmatic
standpoint
and, as a result, have embarked on a critical review of this process.Part of this critical review is to identify ways to streamline
and clearly proceduralize
the facility change
program.RG&E is planning to develop comprehensive
governing procedures, which control all facility changes (major modifications, minor modifications, temporary modifications).
All'facility changes, including procedure changes and other'programs such as NCRs would be handled with one, 50.59 guidance procedure.
This guidance procedure would contain the latest industry guidance'egarding 10 CFR 50.59, including determination
of applicability.
The objective of this process is to assure that facility modifications
regardless
of type are handled in a consistent
fashion.This includes review of the design criteria, safety analysis, and 50.59 screening and safety evaluations
for appropriate
depth and breadth of content.As the NRC no doubt realizes, such an undertaking
is very intensive, and requires the realignment
of programs and the transfer or addition of personnel to support the changes.As a result, it will take time to complete this task, and RG&E.will discuss with the NRC the schedule for accomplishing
the change in the near future.'I Also, as part of this effort, RG&E will conduct the required retraining
of affected personnel.
This will include discussion
of specific procedural
requirements, identified
interfaces, and the requirements
of the Ginna 50.59 program.Another concern expressed by the Staff was that RG&E is not performing
50.59 evaluations
for all modifications
that involve plant equipment described in the UFSAR.RG&E is committed to the requirements
to perform 50.59 evaluations, but does not base this decision to perform a 50.59 evaluation
simply on whether or not the equipment is described in the UFSAR.It-is our position that if a change affects the facility as described in the UFSAR, either explicitly
or implicitly, a 50.59 evaluation
should be completed.
This is not our last determining
factor, however.RG&E makes great efforts to.conservatively
apply the requirements
of 50.59 without losing the perspective
on the intent of the regulation.
RG&E believes that this regulation
must be applied so that it remains meaningful.
We have developed clear safety evaluation
guidance, and extensive screening criteria to accomplish
this goal.We have done this in an effort to not rely excessively
upon the high level of engineering
expertise of our existing personnel, but to furnish clear programmatic
controls.We understand.
that this program development
has taken time and is still underway, but we believe that the Staff should be aware that we saw the need for such guidance and have taken appropriate
steps.RG&E has neither ignored nor downplayed
the Staff's concerns expressed.
over the past 20 months, but has systematically
set up an overall system to address the root cause of those problems.
III.GDC-34 Concerns Another issue raised as part of the Inspection
Report is that the RG&E review"does not address whether good commercial-grade
engineering
practices meets the requirements
of General Design Criterion (GDC)34".The following discussion
provides supplemental
information
to clarify this terminology
and place it in a context that more accurately
reflects RG&E's past practices regarding the design and quality assurance controls applied to the CSTs.The R.E.Ginna Nuclear Power Plant was designed to the proposed AIF GDC issued for comment on July 10, 1967.It should be noted that there is no comparable
1967 AIF GDC which address the residual heat removal issue identified
in GDC 34.The plant was not originally
designed to meet the General Design Criteria (GDC)of Appendix A to 10 CFR 50, including GDC 34, since these criteria were issued in February of 1971.Specifically, the AFW CSTs were designed to the American Water Works Association
Standard (AWWA)D100, 1965 edition.Since Ginna Provisional
Operating License, the AFW system has been scrutinized
as part of the TMI NUREG-0737
effort and the Ginna Systematic
Evaluation
Program (SEP).During the SEP review of Topic III-1 (Classification
of Structures, Components, and Systems-Seismic and Quality), the Franklin Research Center recognized
that the CSTs as originally
designed might not be capable of meeting current compressive
stress requirements.
Additional
information
regarding the compressive
stress capabilities
of the CSTs was requested in the NRC SER on this topic.The information
supplied by RG&E was accepted by the NRC.The CSTs, due to their location in the Service Building (non-seismic structure), lack of protective
features, and their original design pedigree, have the potential for being rendered inoperable
by the effects of several postulated
hazards (i.e., safe shutdown earthquake, tornadoes, floods, missiles, high energy line break effects on the AFW system).It should be noted that these postulated
hazards are remote events with low probability
of occurrence
during the lifetime of the plant.The Ginna Station design accommodates
these remote occurrences
by incorporating
a Seismic Category I source of water (Service Water system)available to the suction of the AFW pumps and by having available a second"Standby" AFW system.(The SAFW system permits delivery of AFW flow to the Steam Generators
assuming the occurrence
of a high energy line break in the Intermediate
Building.)
In addition, another source of water available is the yard fire hydrant system which can function independent
of all AC power.As a result, the CSTs are not required to remain functional
following these postulated-
hazards.With the use of independent
AFW systems and the availability
of independent
and redundant sources of water, means are available at Ginna station to remove reactor decay heat from
A the secondary side of the Steam Generators
at a rate sufficient
to achieve and maintain a safe shutdown condition following any design basis event.'As stated in 10 CFR 50, Appendix B, Section III, design and quality assurance controls should be"commensurate
with those applied to the original design", including all regulatory
commitments
made since Ginna Provisional
Operating License.Thes'e controls assure that the CSTs and changes made thereto meet quality standards at least as stringent as those originally'pplied
placed.on the CSTs are commensurate
with the controls necessary to assure that the CSTs will properly function for the design basis events that require their operability
while being subjected.
to the effects of these same design basis events.The CSTs function for UFSAR Chapter 15 events, unmitigated
fires, and station blackout.The adverse effects of these events have a limited impact on the operability
of the CSTs, due to the CSTs'ocation
in the Service Building and the assumptions
made for these event scenarios (for instance, the assumption
of a coincident
loss of offsite power, but not the assumption
of a coincident
hazard, such as a safe shutdown earthquake).
Section 2.2 of the R.E.Ginna QA manual recognizes
that the CSTs are safety related, but not Seismic Category I and identifies
the controls that apply to these'anks.The Ginna AFW design was found to be acceptable
as originally
licensed in 1969, as reviewed against NUREG-0737, Items II.E.'1.1 and II.E.1.2, following the TMI accident, and as reviewed against SEP Topics X,"Auxiliary
Feedwater System", and V-10.B,"Residual Heat Removal System Reliability".(Note that the TMI and SEP reviews essentially
reviewed the Ginna AFW systems against the criteria of BTP ASB 10-1 and BTP RSB 5-1.)As a result of NUREG-0737
and the SEP effort, RG&E made numerous commitments
and upgrades to the AFW systems.The QA and design controls applied to the AFW system, including the CSTs, are consistent
with these commitments.
RG&E is aware of the safety importance
of the CSTs and believes that the quality assurance controls applied to the CSTs meet the original design bases, as well as the regulatory
commitments
made since the Provisional
Operating License was issued..When considered
within the overall context of the Ginna Station design, the QA requirement
applied to CSTs are appropriate.
Due to the issuance recently of 10 CFR 50.63, and Regulatory
Guide 1.155, RG&E is performing
an additional
review of the design controls placed on the CSTs in the context of this regulatory
guidance and.will include appropriate
upgrades.
8
IV.Evaluation
of CST Modifications
Another issue discussed in the inspection
report was that a technical evaluation
for the installation
of Tygon tubing and'copper piping had not been provided at the end of the inspection
period.RGGE has performed these 50.59 safety evaluations
for both of these concerns (see also the response to the Notice of Violation regarding the Tygon tubing, which contends that a proper 50.59 screening was performed for the addition of the Tygon tubing prior to its installation).
All other CST modifications
had previously
documented
50.59 safety evaluations.
For both the Tygon tubing and the installation
of the Hot Water system the safety evaluations
conclude that no unreviewed
safety questions have been introduced.
These evaluations
are provided in Attachment
III.
ATTACHKBiT
III Safet Evaluation
for the Hot Water S stem Connection
The Hot Water system connections
to the CSTs are shown on P&XDs 33013-457 and 1234.Suction to the hot water circulation
water pump (MK 102)is taken from the CSTs through manual valves 8271, 8275 (CST B), 8270, 8274 (CST A), and 8276.Hot water is recirculated
back to the CSTs through 8299J, 8282 (CST A), and 8283 (CST B).The'ollowing
sections evaluate the impact to plant safety of the connection
of the Hot Water system to the main AFW CSTs.Postulated
Hazards and Safe Shutdown Ca ahilit The following discussion
applies to the postulated
Hazards listed below:~Adverse weather phenomena including floods, high winds, snow, and tornadoes~Safe Shutdown Earthquake
~High Energy Line Breaks~Externally
or internally
generated missiles The'Hot Water system is located adjacent to the CSTs in the Service Building.-
As a result, the adverse effects of postulated
hazards that can potentially
fail the Hot Water system (and thereby introduce a potential interaction
with the CSTs)also have the potential to fail the CSTs, since in both cases: 1.Neither the CSTs nor the Hot Water system are required to be designed to withstand.the effects of the hazards postulated
for the R.E.Ginna plant.2.Neither the CSTs nor the Hot Water system are protected.
by design features such as physical barriers to preserve~their integrity following postulated
hazards.The Service Building is not a Seismic Category I structure capable of withstanding
adverse weather effects or natural phenomenon.
For postulated
hazards that, potentially
fail the CSTs, an alternate and independent
means of achieving and maintaining
a safe shutdown condition (the safe shutdown function of concern is the removal of reactor decay heat)is available via the Service Water system supplying water to either the main or standby AFW systems.
As a result, no significant
degradation
in the capability
of achieving and maintaining
a safe shutdown condition will result due to the hot water modification
interface with the CSTs;The contingency
actions and alternate means of removing reactor decay heat following a postulated
hazard remain valid for the current CST configuration.
The failure of the Hot Water system following a seismic event which could lead to the draining of the CSTs onto the Service Building floor is bounded.by the present analysis of failure of CSTs per EWR 1023, May 20, 1975.This'looding, scenario is the same as that previously
analyzed since the Hot Water system does not introduce a new source of water.In addition, the Hot Water system does not introduce any high energy line break of concern, or the potential for internally
generated missiles.Fires The main AFW system taking suction from the CSTs is used to remove decay heat following several postulated
unmitigated, fires.Unmitigated
fires can result in a loss of offsite power which would subsequently
result in a loss of the Instrument
Air System (IAS).A review of PAID 33013-457, shows that the"effect" of a fire resulting in a loss of the IAS is minimal on the CSTs as conf igured with Hot Water system connection.
If the Hot Water system is not in use, it can be isolated from the CSTs via manual valves 8275, 8274, and 8299J.If the Hot Water system is in use during a fire, Hot Water pumps (MK102 and 115)would stop on loss of AC power.Although the Hot Water system could become a potential drainage path for CST inventory, the elevation of the hot water users (laundry and.hot showers are at an elevation equal to the top of the CSTs)and the flow resistance
of the Hot Water system piping essentially
make the Hot Water system a closed system to drainage.Although the Hot Water system introduces
combustibles
into the Service Building via the gas supply to heater Mk106, the safe shutdown components
located in the Service Building (CSTs, piping to the AFW pumps)should not be adversely affected.This is in accordance
with the existing analysis which deals with fires in the Service Building.These are mechanical
components
that must maintain their pressure boundary integrity to accomplish
their safe shutdown function.The Appendix R analysis for Ginna assumes that exposure fires do not cause mechanical
components
to lose pressure boundary integrity.
As a result, this modification
does not affect safe shutdown for fires.For fires requiring operation of the main AFW system, the CSTs will be operable, enabling the removal of reactor decay heat to achieve and maintain a safe shutdown condition during and following postulated
fires.
Miti ation of Cha ter 15 Events The following discussion
assesses the safety impact of the connection
of the Hot Water System to the CSTs for the'following
postulated
UFSAR Chapter 15 events, which are the only events for which the AFW system is relied on as a mitigation
f eature:~Main Steam Line Breaks (MSLB)~Main Feed Line Breaks (MFLB)~Loss of Normal Feedwater~Loss of AC to the Station Auxiliaries
~Loss of External Electrical
Loads~Loss of Coolant Accidents (LOCAs)~Steam Generator Tube Rupture (SGTR)To assess the potential degradation
in the capability
of mitigating
the above events, the Hot Water system is examined for its potential adverse interaction
with the CSTs.The function of the CSTs is to maintain an inventory of 22,500 gallons of condensate-grade
water for the removal of reactor decay for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> independent
of any AC power source (TMI Item II.E.1.1).
The CSTs also function as the initial AFW inventory source following the occurrence
of any of the above Chapter 15 events (which result in the subsequent
loss of main feedwater and auto initiation
of AFW).Hence, the adverse"effects" of these Chapter 15 events are examined for the potential to fail the Hot Water system and thereby deplete the CST inventory through an adverse interaction.
There are two"effects" that impact the current CST configuration.
1.MSLBs and MFLBs in the intermediate
building create adverse effects (pipe whip, jet impingement, temperature, pressure, humidity)that have the potential to fail the main AFW system.All three main AFW pumps (2 motor driven pumps, 1 turbine driven pump)and a significant
portion of AFW piping is located in the intermediate
building.The effects of some postulated
MSLBs and, MFLBs can fail the main AFW system.2.Most of the Chapter 15 events assume of offsite power.Loss of normal AC loss of the Instrument
Air System operated valves (AOVs)to fail on loss the coincident
loss power results in a (IAS)causing air of supply air.For the case of MSLBs or MFLBs in the intermediate
building, if the main AFW system fails then the CSTs no longer function as the AFW water source.In this case the standby AFW system is placed into service (10 minutes for operator action is available)
taking suction from the Service Water system.This equipment is located in the Standby Auxiliary Feedwater Pump Building and is a completely
independent
means of removing reactor decay heat.Therefore, the potential effects of MSLBs and MFLBs on the Hot Water system have been bounded by the current Chapter 15 analysis.
As described in the section on"Fires" above, the loss of normal AC power does not create an adverse interaction
between the Hot Water system and the CSTs.The Hot Water system is effectively
a closed system.CST inventory will not be depleted as a result of the assumed coincidence
of a loss of offsite power for the Chapter 15 events.In conclusion, the connection
of the Hot Water system to the CSTs does not result in additional
consequential
failures or new failure modes that create the potential for new"worst single failures",.
or different event scenarios.
Hot Water S stem Safet Evaluation
Conclusion
This section summarizes
the safety evaluation
of the Hot Water system connection
to the CSTs.This summary groups postulated
Fires under the category of Hazards.The connection
of the Hot Water system to the CSTs does not increase the probability
of occurrence
of an accident previously
evaluated in the Ginna Updated FSAR.As discussed previously, the failure of the Hot Water system does not create a plant transient requiring a protective
response from a safety system.The connection
of the Hot Water system to the CSTs does not increase the consequences
of an accident previously
evaluated in the Ginna Updated FSAR.As discussed above, the capability
to achieve and.maintain a safe shutdown condition following the occurrence
of postulated
hazards is not degraded.The Hot Water system does not adversely interact with CSTs for the Chapter 15 event scenarios.
As a result, CST inventory is not degraded, main AFW performance
is not impacted, and the capability
to remove reactor decay heat during and following the postulated
Chapter 15 events is not degraded.Therefore, the integrity of barriers preventing
the release of fission products is not impacted.The connection
of the Hot Water system to the CSTs does not increase the probability
of occurrence
of a malfunction
of equipment important to safety previously
evaluated in the Ginna Updated FSAR.As discussed above, the effects of postulated
Hazards that could fail the Hot Water system would likely also-fail the CSTs since the CSTs were not originally
designed to withstand such effects.As such the effect of failing the hot water system is bounded by the original analysis which assumes failure of the CSTs.The effects of Chapter 15 events have no impact on the capability
of CSTs to maintain their inventory for those events requiring the operation of the main AFW system.There is therefore no change in the failure probability
of the CSTs for Chapter 15 events.
The connection
of the Hot Water system to the CSTs does not increase the consequen"es
of a malfunction
of equipment important to safety.The Hot Water system does not increase the severity of the malfunction
of the CSTs for Hazards or Chapter 15 events.The consequences
of such malfunctions
are therefore unchanged.
The connection
of the Hot Water system to the CSTs does not create the possibility
of an accident of a different type than any previously
evaluated in the Ginna Updated FSAR.The failure of the Hot Water system and its potential for interaction
with the CSTs does not create new plant transients
requiring mitigation.
The connection
of the Hot Water system to the CSTs does not create the possibility
of a malfunction
of equipment important to safety of a different type than any previously
evaluated in the Updated Ginna FSAR.The failure of the Hot Water system is essentially
the same as a failure of the CSTs.The installation
of the Hot Water system does not introduce a new or different failure mode.The connection
of the Hot Water system to the CSTs does not reduce the margin of safety.The AFW system can still function to mitigate Chapter 15 events, as well as to remove decay heat for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> without an A.C.power source.In addition, safe shutdown capability
is not'ffected, and the integrity of fission product barriers are not compromised.
Based on the above conclusions, the connection
of the Hot Water system to the main AFW CSTs does not introduce an unreviewed
safety question as defined by 10 CFR 50.59.Safet Evaluation
for the Addition of on Tuhin Tygon tubing was installed downstream
of locked closed manual valve 4318A to provide a means of local CST level indication
independent
of any A.C.power source.Local CST level indication
via the Tygon tubing would be used to allow local operators to determine when to align and place into operation the Service Water system following:
1.Control Complex Fires (SC-3.30.1)
2.Cable Tunnel Fires (SC-3.30.2)
3.Auxiliary Building Basement/Mezzanine
Fires (SC-3.30.3)
The 22,500 gallon inventory in the CSTs provides for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of reactor decay heat removal.This is considered.
sufficient
time to align and place into operation the Service Water system in the remote event that an AC power source can not be restored for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following the postulated
unmitigated
fires identified
above.Hence, the Tygon.tubing is not essential for safe shutdown following fires.However, it can provide operators with CST level information
to provide a more accurate means of determining
when Service Water should
1 be aligned.It should be noted that the R.E.Ginna Appendix R Alternative
Shutdown Report does not identify CST level indication
as a plant process parameter that must be monitored for supporting
safe shutdown.The installation
of Tygon tubing does not increase the probability
of occurrence
or the consequences
of an accident or malfunction
of equipment important to safety previously
evaluated in the Ginna Updated FSAR.The Tygon tubing is isolated from all plant process systems via locked'closed manual valve 4318A.It therefore, has no effect on previously
analyzed accidents.
The Tygon tubing's a flexible material of low mass that is not'apable
of physically
impacting the systems, structures
or components
in the immediate vicinity.The installation
of Tygon tubing does not create the possibility
of an accident or malfunction
of a different type than any previously
analyzed in the Ginna Updated FSAR.As stated.above, the Tygon tubing is isolated from all plant systems and therefore does not create the potential for process interactions
that can lead to different accidents or malfunctions.
The installation
of Tygon tubing does.not reduce the margin of safety as defined in the basis of any Technical Specification.
The Tygon tubing will provide a local indication
of CST level for fires that result in a loss of all AC.This indication
provides a better means of determining
when to align Service Water in the unlikely event of an unmitigated
fire, prolonged loss of all AC, and depletion of the CSTs.Although'afe
shutdown can be achieved without this local CST level indication, the Tygon tubing is beneficial
for fire recovery efforts.