IR 05000244/1988022
| ML17251A500 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 05/15/1989 |
| From: | Kane W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Mecredy R ROCHESTER GAS & ELECTRIC CORP. |
| References | |
| NUDOCS 8905260363 | |
| Download: ML17251A500 (34) | |
Text
ACCEKZRATED DISHUBUTION DEMONSTRATION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:8905260363 DOC.DATE: 89/05/15 NOTARIZED: NO
DOCKET ¹ FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G
05000244
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AUTH.NAME AUTHOR AFFILIATION ANEPW.F.
Region 1, Ofc of the Director RECIP'.NAME RECIPIENT AFFILIATION MECREDY,g.C.
Rochester Gas a Electric Corp.
-SUBJECT: Withdraws violation re addition. of Tyson tubing from Insp Rept 50-244/88-22.Response dtd 890106 S0217 encl.
DISTRIBUTION CODE:
IEOID COPIES RECEIVED:LTR (
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TITLE: General (50 Dkt)-Insp Rept/Notice of Violation Response'OTES:License Exp date in accordance with 10CFR2,2.109(9/19/72).
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Docket No ~ 50-244
.
HIAY i 5 1989 Rochester Gas and Electric Company ATTN:
Dr. Robert C. Mecredy General Manager Nuclear Production 49 East Avenue Rochester, New York 14649 Gentlemen Subject:
Inspection Report 50-244/88-22; Notice of Violation 88-22-02 This letter refers to your letter, dated January 6,
1989, as amended by your letter, dated February 17, 1989, which denies the violation transmitted in In-spection Report 50-244/88-22 relating to the addition of Tygon tubing to the condensate storage tank.
We have reviewed the new information provided in your submittals and have concluded that the violation is not appropriate and it is hereby withdrawn.
The Tygon tubing does not represent a safety concern as in-stalled and valve 4318A provides an isolable boundary.
However, we understand that the origin of valve 4318A is questionabl'e as to whether it is original equipment or a later modification.
Due to the age of the facility and the level of quality of the design basis documentation, we are unable to clearly establish the need for a
CFR 50.59 evaluation to support its existence.
Your commitment to perform a comprehensive review of the con-densate storage tanks, finalize a systematic approach to performing reviews required by 10 CFR 50.59 and to conduct training on the requirements regarding changes, tests and experiments are appropriate and should alleviate future issues such as this one.
Your cooperation with us in this matter is appreciated.
Sincerely, ofNe41 ~+ +
William F.
Kane, Director Division of Reactor Projects Enclosures:.
1.
RG&E Response dated January 6,
1989 2.
RGKE Letter dated February 17, 1989 gppg~g 363 >
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Harry H. Voigt, Esquire Central Records (4 copies)
Oirector, Power Oivision Public Document Room (POR)
Local Public Oocument Room (LPOR)
Nuclear Safety Information Center (NSIC)
NRC Resident Inspector State of New York, Oepartment, of Law bcc w/encl:
Region I Oocket Room (with concurrences)
Management Assistant, ORMA (w/o encls)
ORP Section Chief C. Stahle, PM, NRR Robert J.
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ENCLOSURE
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ROCHESTER GAS AND ELECTRIC C:.RI-OR4TION
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o"9 EAST AVENUE. ROCHESTER, W.Y.
I~6~9-OOOO'anuary 6,- 1989 Mr. William 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:
Qi In accordance with
CFR 2.201, RG&E provides Attachment I to this letter as our resp6nse to the.Notice of Violation.
RG&E 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 6JWN016 Attachments xc xc U.S. Nuclear Regulatory Commission (original)
Document Control Desk Washington, DC 20555 Ginna Senior Resident Inspector
ATTACHMENTI Violation
"l.0 CFR 50, Appendix B,
section III requires, in part, measures shall be established to assure that appropriate quality 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 as 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."
Response RGGE denies this violation.
RGGE 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
"o 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).'his 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
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.
I
When this modii ation was made by RG&E, the appropriate consideration was given to the governing requirements of
CFR 50. 59.
Appropriate screening criteria were applied to determine the applicability of
CFR 50.59.
Although the documentation ma,intained 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
CFR 50.59.
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 aaaaatic
'ovements As discussed in Attachment II, a
programmatic approach to 50.59 has been and is continuing to be developed at Ginna.
Procedures'ave 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 Ful1 liance RG&E believes it is currently in compliance with
CFR 50.59 and with the Ginna Quality Assurance Manual as it relates to the issues identified.
ATTACHM1?ST II I.
Int1oduction In addition to addressing the Notice of Violation, we are responding to some of the Staff's associated concerns raised.
within the inspection repoxt 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
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 Pr am 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'mpacts that each evaluator must consider for a specific type of change.
Specifi.c examples for these changes are also provided.
RG&E has taken a
proactive approach to dealing with the
CPR 50.59 process.
In many xespects, this has been difficult because of the evolving nature of NRC/industry guidance in'his area.
This is evidenced.
by the fact that even the most recent SBQRC 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 clear ly proceduralize the facility change
$ 1
progran.
RG&E x planning to develop comprenensive governing procedures, whicn control all facility changes (major modifications, minor modifications, temporary modifications).
All'acility 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 regarding
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 const,stent 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.
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 perform'ing 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 syst: em to address the root cause of those problems.
I $
III. GDC-34 Concerns Another.issue raised as part of the Inspection Report is that the RGGE 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 RGGE'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
CFR 50, including GDC 34, since these criteria were 'ssued in February of 1971.
Specifically, the
'AFM 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 gart of the TNI 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 SEE on this topic.
The information supplied by RGGE was accepted by the NRC.
The'STs, 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 gostulated, 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 0he 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
the secondary sade 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.
These controls assure that the CSTs and changes made thereto meet quality standards at least as stringent as those originally applied to the CSTs.
The. QA controls 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 tanks.
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, RGGE 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.
RGGE 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, RGGE 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.
IV.
Evaluation ~f 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.
ATTACHMENTIII Safet Evaluation for the Hot Water S stem Connection The Hot Water system connections to the CSTs are shown on PGIDs 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 following 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 abilit The following discussion applies to the postulated Hazards listed below:
~
Adverse weather phenomena including floods, high winds, snow, and. tornadoes
~
~
Hi h Ener Line Breaks g
gy
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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 Mater 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 flooding 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 Mater 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 P&ID 33013-457, shows that the "effect" of a fire resulting in a loss of the IAS is minimal on the CSTs as configured 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 atioa of Cha ter 15 Events The following discussion assesses the safety impact cf 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'ystem is relied.
on as a
mitigation feature:
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)
'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
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 the coincident loss of offsite power.
Loss of normal AC power results in a loss of the Instrument Air System ( IAS )
causing air operated valves (AOVs) to fail on, loss 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.
Zn this case the standby AFW syst: em 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 NFLBs on the Hot Water system have been hounded by the-current Chapter 15 analysis.
As described in che section on "Fixes" 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.
Zn 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'STs.
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 syst: em 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 syst: em 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
events.
The connection of the Hot Water system to the CSTs does not increase the consequences of a
malfunction of equipment important to safety.
The Hot Water system does. not increase the severity of the malfunction of the CSTs or 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
hours without an A.C.
power source.
In addition, safe shutdown capability is not affected, and. the integrity of fission product barriers are not compromised.'ased 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.
Safe Evaluation for the Addition of on Tubin Tygon tubing was installed downstream oi ~losed manual valve 4318A to provide a means of local CST level indication independent of any A.C.
power source.
Z,ocal 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 opegation the Service Water system in the remote event that an AC power source can not be restored for
hours 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
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 is a
flexible material of low mass that is not capable of physically impacting the systems, structures or components in the immediate vicinity.
The installation of Tygon tubing d'oes 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'
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 safe shutdown can be achieved without this local CST level indication, the Tygon tubing is beneficial for fire recovery efforts.