ML17250A715: Difference between revisions

From kanterella
Jump to navigation Jump to search
(Created page by program invented by StriderTol)
(Created page by program invented by StriderTol)
Line 18: Line 18:


=Text=
=Text=
{{#Wiki_filter:Environmental Qualif ication of Electrical Equipment R.E.Ginna Nuclear Power Plant Docket No.50-244 February 24, 1978 Rev.1, December 1, 1978 Rev.>2, April 25, 1980 Rev.3, October 31, 1980-luanCE THE ATTACHED FILES ARE OFFICIAL RECORDS OF THE DIVISION OF DOCUMENT CONTROL.THEY HAVE BEEN CHARGED TO YOU FOR A LlhhlTED TIME PERIOD AND MUST BE RETURNED TO THE RECORDS FACILITY BRANCH 016.PLEASE DO NOT SEND-DOCUMENTS CHARGED OUT THROUGH THE MAIL.REMOVAL OF ANY" PAGEIS)FROM DOCUMENT FOR REPRODUCTION, MUST BE REFERRED TO FILE PERSONNEL.
{{#Wiki_filter:Environmental Qualif ication of Electrical   Equipment R. E. Ginna   Nuclear Power Plant Docket No. 50-244 February 24, 1978 Rev. 1, December 1, 1978 Rev.> 2, April 25, 1980 Rev. 3, October 31, 1980 luanCE THE ATTACHED FILES ARE OFFICIAL RECORDS OF THE DIVISION OF DOCUMENT CONTROL. THEY HAVE BEEN CHARGED TO YOU FOR A LlhhlTED TIME PERIOD AND MUST BE RETURNED TO THE RECORDS FACILITY BRANCH 016.       PLEASE DO NOT SEND -DOCUMENTS CHARGED OUT THROUGH THE MAIL. REMOVAL OF ANY
Docket@gO~tIO)g So i I Oqc a.S V DEADLINE RETURN DATE gEGlHATORV OOgKET F{K Bo 1.>040'/B'P~H>>@R FY"'"~'A, RBCCIIDB FACILITY BRANCH I P r I ,,i 7g/p1 1 i/ji'N Introduction TABLE OF CONTENTS Pacae Identification of Necessary Safety Related Equipment 3 A.B.C.Events Accompanying a Loss of Coolant Accident 3 Events Accompanying a Main Steam Line Break or 11 a Main Feed Line Break High Energy Line Breaks Outside Containment 16 Identif ication of the Limiting Service Environmental 19 Conditions for Equipment which is Required to Function to Mitigate the Consequences of Events Identified Above A.B.C~D.E.F.G.H.I.Inside Containment Auxiliary Building Intermediate Building Cable Tunnel Control Building Diesel Generator Rooms Turbine Building Auxiliary Building Annex Screen House 19 22 25 27 27 30 30 32 32 Equipment Qual if ication Inf ormation 34 1 I I I LIST OF FIGURES Figure 1 Loss of Coolant Accident fSequence of Events Diagram]Figure 2 Main Steam or Feed Line Break (Sequence of Events Diagram]Figure 3-Plant Layout Figure 4 Pressure Envelope for Ginna (FSAR Figure 1 of Appendix 6E)Figure 5 Temperature Envelope for Ginna (FSAR Figure 2 of Appendix 6E)Figure 6 Radiation Level for Ginna (FSAR Figure 5 of Appendix 6E)
                              "
LIST OF TABLES Table 1 Loss of Coolant Accident[Required Equipment List]Table 2 Main Steam or Feed Line Break[Required Equipment List]Table 3 Equipment Qualif ication Table 4 Environmental Service Conditions
PAGEIS) FROM DOCUMENT FOR REPRODUCTION, MUST BE REFERRED TO FILE PERSONNEL.
Docket@
gO~tIO) g So iI Oqc a. S V DEADLINE RETURN DATE gEGlHATORV OOgKET F{K Bo 1.>040'/B'P~H>>@R             I FY " '" ~'   A,   RBCCIIDB FACILITY BRANCH


Environmental Qualification of Safety-Related Electrical Equipment INTRODUCTION The electrical equipment described in this report is that saf ety-related equipment required to mitigate the ef f ects of high or moderate energy line breaks (HELB)inside or outside containment, and to effect eventual cold shutdown of the reactor.The environmental qualification requirements are described in the"DOR Guidelines", transmitted to RG6E on February 15, 1980.Although the DOR Guidelines address all electrical equipment, the emphasis in this report will be on that equipment exposed to an adverse HELB environment.
P r
This is defined as that equipment located in the containment, Intermediate Building, Turbine Building, and Auxiliary Building basement (radiation only).This revised scope is consistent with the Commission Order of September 19, 1980.Equipment in other"mild" environments will be addressed at a later time.This submittal revises and supersedes our previous reports concerning environmental qualification of electrical equipment, dated February 24, 1978, December 1, 1978, and April 25, 1980.It also consolidates and updates all information submitted on June 10, 1980 and September 24, 1980.Section IV of this report presents an item-by-item response to the Draft Interim Technical Evaluation Report FRC Project C5257, concerning the review of the Ginna electrical equipment P
I
environmental qualif ication, dated August 20, 1980.New references are included with this report.However, references previously submitted are not being resubmitted.
,,i    7g
1n Section IV, it is either shown that each item is adequately qualified to perform its required safety function in its post-accident operating environment, or a commitment for additional testing or replacement is made.In all cases, sufficient justification for continued operation is given.Table 3 summarizes the equipment qualification in the format requested for SEP by the NRC in a September 6, 1978 letter.Table 4 provides the definition of environmental parameters throughout the Ginna plant.This table is comparable to Appendix A of F-C5257, and tabulates the explanatory basis given in Section III of this report.Supplement No.3 to IE Bulletin 79-01B provides the timing for submittal of qualification information for equipment in-stalled to meet the TMI Short Term Lessons Learned.RGSE intends to follow the guidance given in this supplement.
      /p1 1
In a number of cases, it is possible that additional documentation or testing results may become available after November 1, 1980.Since this additional information will be of use in documenting the status of the Ginna environmental qualification, it will be submitted when received.Every effort has been made to ensure that all documentation was obtained for use with this submittal.
i
l II.IDENTIFICATION OF NECESSARY SAFETY RELATED EQUIPMENT This section of the report identifies the necessary safety related equipment for each of the Design Basis Events (DBE)of concern and a brief description of why the equipment is needed.This identification includes all electrical equip-ment required by the Ginna emergency procedures for accomplish-ing the necessary safety functions.
      /
It must be recognized that not all electrical equipment referenced in the procedures is required to function (as opposed to being useful if available), and is therefore not required to be qualified.
ji'N
The emergency operating procedures were not developed by considering safety-related components to the exclusion of all others.While such procedures are written with priority attention given to safety-related equipment, other systems and components are justifiably mentioned.
A realistic evaluation of plant incidents might result in situations and hostile environments significantly less severe than those assumed for the purposes of conducting the environmental qualification program.The absence of full qualification for certain components which fall into this category is not, by itself, a sufficient motive to classify the equipment inoperable.or to remove these components from the procedures.
A.Events Accom an in a Loss Of Coolant Accident Analyses of the course and consequences of loss of coolant accidents have been submitted previously (LOCA 1-4).A discussion of equipment required to function to mitigate the consequences of a loss of coolant accident is presented in the FSAR Chapters 6, 7 and 14.Post-LOCA operator actions are included in the Ginna Emergency Procedures.
These procedures are consistent with the generic Westinghouse guidelines, which have been approved by the NRC.Additional descriptive material is presented in this report to provide summary information as to the sequence of events and the equipment involved at each stage.Figure 1 illustrates the sequence of events following a loss of coolant accident.Table 1 provides a specific equipment list for each numbered block in Figure 1.Also provided in Table 1 is the safety function which is required and the period of time that operability must be ensured.It should be noted that Table 1 includes all redundant equipment, not the minimum safeguards equipment assumed in the safety analysis.In the"required" column it should be noted that equipment listed as"signal initiation" is required to be operable only until its required safety function, the initiation of a safety signal, is performed.
It is important to note that the arbitrary requirement of the DOR Guidelines to qualify equipment to function for at least one hour, even if its only function is completed within seconds, is not well reasoned.In many cases, the environment would not exist unless the equipment safety function had been completed (e.g., flooding to a seven foot level in containment by necessity means that SI was initiated).
RGSE does not agree with this one-hour requirement, and it is therefore not applied as an environmental qualification requirement.
Equipment listed as"long term" is required to provide long term decay heat removal, post-accident monitoring and sampling, or maintaining a safe shutdown condition.
Equipment listed as"short term" is required only for a short period of time (hours).Table 3 provides the environmental qualification require-ments and documentation references for the Ginna Class IE equipment.
1.The first event in the loss of coolant accident following the rupture is the detection of the rupture.Any 2/3 low pressurizer pressure or 2/3 high contain-ment pressure will initiate"safety injection" (SI).la.Instrumentation is available to the operator to distinguish between a LOCA and the other accidents, such as a steam line break or feed line break.It is important to note that the automatic actions and immediate operator actions (first 10 minutes)are identical in the mitigation of these accidents.
2.Upon"safety injection" signal generation, safe-guards sequencing is initiated (see FSAR Table 8.2-4).The diesel generators start and energize the safeguards buses assuming there is a loss of offsite power.With the safeguards buses energized, either by off-site power or the diesels, the three safety injection pumps,


the two residual heat removal pumps," two of the four service water pumps, the two motor driven auxiliary feedwater pumps, and the four containment.
TABLE OF CONTENTS Pacae Introduction Identification of   Necessary Safety Related Equipment    3 A. Events Accompanying a Loss of Coolant Accident      3 B. Events Accompanying a Main Steam Line Break or    11 a Main Feed Line Break C. High Energy Line Breaks Outside Containment      16 Identif ication of the Limiting Service Environmental  19 Conditions for Equipment which is Required to Function to Mitigate the Consequences of Events Identified Above A. Inside Containment                                19 B. Auxiliary Building                                22 C ~  Intermediate Building                            25 D. Cable Tunnel                                      27 E. Control Building                                  27 F. Diesel Generator  Rooms                          30 G. Turbine Building                                  30 H. Auxiliary Building  Annex                        32 I. Screen House                                      32 Equipment Qual ification  Information                  34
fan coolers-will"be loaded sequentially onto the buses.The two containment spray pumps are automatically loaded onto the buses when the 30 psig containment pressure setpoint's reached.3.A break in the reactor coolant system piping actuates the passive accumulator injection system when the reactor coolant system pressure is reduced to 700 ps lg The flow path of the borated water from each accumulator-is through a series of check valves and a normally locked open (with AC control power removed)motor operated valve.The motor operated valves, MOV 841 and NOV 865, are not required to function to mitigate the consequences of the accident[Flood-1].4.The main steam isolation valves 3516 and 3517 close upon receiving a high containment pressure signal and the main and bypass feedwater control valves 4269, 4270, 4271 and 4272 close upon receiving a safety injection signal.The SI signal also causes a trip of the main feedwater pumps (which in turn causes the closing of the feedwater discharge valves).All of this equipment will fail in its safety position on loss of electrical power.  


5."Containment Isolation" and"Containment Ventilation Isolation" (ref erred to collectively as simply,"Containment Isolation")is initiated by the saf ety injection signal.Containment isolation is discussed in detail in Section 5.2 of the FSAR.Most of the containment isolation valves are air operated valves.All air operated containment isolation valves close with safety injection signal with the exception of valves 4561 and 4562 which open full to insure service water supply to the containment recirculation f ans.The f ail saf e position of the valves is the desired safeguard position as described above.Six motor operated valves (313, 813, 814, ATV-1, ATV-2, ATV-3)receive a containment isolation signal.All of these valves are located outside of containment and only valves 313, 813, and 814 are fed from the safeguards buses.During normal operation ATV-1, ATV-2, and ATV-3 are closed with blank flanges installed on their respective penetrations inside containment.
1 I
The use of the process lines associated with these valves occurs only during the containment building integrated leak rate tests.Valve 313, the reactor coolant pumps seal water return line, and valves 813 and 814, reactor coolant support inlet and outlet lines, are closed by the containment isolation signal.
I I


~6.The SI signal trips the reactor and turbine.Other reactor trips are discussed in the FSAR, Section 7.7.The reactor coolant pumps are tripped by manual operator action when low pressurizer pressure (1715.psig)is reached, and SI flow is initiated.
LIST  OF FIGURES Figure 1  Loss of Coolant Accident fSequence  of Events Diagram]
8.Selected valves throughout the plant provide flow paths for the required safeguards equipment with the advent of the SI signal.During normal operation all required valves in the flow paths for high head safety injection'are normally open with the exception of valves 826A and 826C, the dis-charge valves from the boric acid storage tank to the suction of the safety injection pumps.Valves 826A;B, C and D receive the safety injection signal and valves 82 6A and C open providing borated water to the reactor coolant loop cold legs.When the level in the boric acid storage tank decreases to the 10%level, suction for the high head safety in-jection pumps is automatically switched from the boric acid storage tanks to the refueling water storage tank by the automatic opening of, valves 825A and B and closing of valves 826A, B, C and D.During normal operation, all valves in the flow paths for low head safety injection are normally open except for MOV 852A and MOV 852B, the valves in the vessel upper plenum injection lines.These valve's open upon receipt of a safety injection signal and remain open-thereaf ter.The containment spray pumps will automatically start and the discharge valves 860A Bg C and D automatically open, receiving power from the safeguards buses when containment pressure reaches 30 psig.If containment pressure does not reach 30 psig, the operator may manually start the spray pumps after all other safeguards are loaded on the safeguards buses.Automatic NaOH addition via opening of valves HCV 836A, B takes place two minutes after containment spray pump start unless defeated manually.The containment spray pumps are normally aligned to the refueling water storage tank with all suction valves.open.SI system actuation will automatically align the two post accident charcoal f ilters to the containment recirculation system by opening inlet dampers 5871 and 5872, and outlet dampers 5873 and 5874.Loop entry dampers 5875 and 5876 will close.These dampers will fail to their safeguards position upon loss of electric power.9.The control room ventilation is automatically placed in the 100%recirculation mode (with about 25%flow through charcoal filters), when SI is initiated.
Figure 2  Main Steam or Feed  Line Break (Sequence of Events Diagram]
Figure 3 - Plant Layout Figure 4  Pressure Envelope  for Ginna (FSAR Figure 1 of Appendix 6E)
Figure 5  Temperature Envelope for Ginna (FSAR Figure 2 of Appendix 6E)
Figure 6  Radiation Level for Ginna (FSAR Figure 5 of Appendix 6E)


10.Af ter the safety injection pumps are automatically switched from the boric acid storage tanks to the re-fueling water storage tanks, the operator resets safety injection, starts the component cooling water pumps and aligns flow to the RHR heat exchangers, and initiates SW flow to the'CW heat exchangers.
LIST  OF TABLES Table 1 Loss of Coolant Accident [Required Equipment List]
At the 31%RWST alarm, the operator shuts off one CS and one SI pump (if more than one are running).When the refueling water storage tank level is reduced to 10%, the plant operator stops the remaining residual heat removal, containment spray and high head safety injection pumps and establishes f low paths to the reactor vessel f or both high (if required)and low head safety injection from containment sump B.The normal (non-saf ety grade)auxiliary f eedwater supply source is from the condensate storage tanks.If this supply is exhausted the operator opens the motor operated valves 4027 and 4028 and manual operated valves 4344 and 4345 to provide service water to the suction of the auxiliary feedwater pumps.If the AFW system is not functioning properly, the operator can align from the control room the Standby AFW system to the steam generators (using'ervice water suction).11.In the recirculation phase, the operator aligns the RHR pumps to containment sump B by opening valve 850A for pump A and valve 850B for pump B, and closing 10
Table 2 Main Steam or Feed Line Break [Required Equipment List]
Table 3 Equipment Qualif ication Table 4 Environmental Service Conditions


valve 704A, 704B, 856, and 896A or 896B.For low head recirculation, injection is through the vessel nozzles.,For high head recirculation, the RHR pumps discharge to the safety injection pumps through alignment of valve 857A (for RHR pump B)and/or valves 857B and 857C (for RHR pump A).Valves AOV 897, 898 are closed.The high head safety injection pumps then provide water to the cold leg injection points.This alignment also allows CS pump operation, if desired.Long term recirculation to compensate for the possible effects of boron precipitation has been described in Ref[Flood-1]and includes the use of RHR pumped flow to the vessel nozzles and through a high head safety injection pump into either cold leg.Post-accident reactor coolant and containment atmosphere sampling modifications are presently being undertaken, in accordance with the implementation schedule for the TMX Lessons Learned commitments.
Environmental Qualification of Safety-Related      Electrical  Equipment INTRODUCTION The  electrical  equipment described in    this report is that saf ety-related equipment required to mitigate the ef f ects of high or moderate energy line breaks (HELB) inside or outside containment, and to effect eventual cold shutdown of the reactor. The environmental qualification requirements are described in the    "DOR  Guidelines", transmitted to RG6E    on February 15, 1980.     Although the DOR Guidelines address    all electrical  equipment, the emphasis    in this report will be  on that equipment exposed to an adverse HELB environment. This is defined as that equipment located in the containment, Intermediate Building, Turbine Building, and Auxiliary Building basement ( radiation only). This revised scope is consistent with the Commission Order of September 19, 1980.
See[Ref TMI-3].Events Accom an in a Main Steam Line Break or a Main Feed Line Break The analyses of a main steam line break or a main feed line break and the consequences thereof have been discussed in Chapters 6 and'14 of the FSAR and in References
Equipment in other "mild" environments will be addressed at a later time.
[SLB/FLB 2-4].The High Energy Line Break analyses[HELB 1-7]provide additional information regarding steam line breaks outside of containment, as 11
This submittal revises and supersedes      our previous reports concerning environmental    qualification of electrical equipment, dated February 24, 1978,    December 1, 1978, and April 25, 1980. It also  consolidates  and updates  all information submitted on June 10, 1980 and September 24, 1980.        Section IV of this report presents an item-by-item response to the Draft Interim Technical Evaluation Report FRC Project C5257, concerning the review of the Ginna      electrical  equipment


well as feedwater line breaks inside and outside containment.
P environmental qualif ication, dated August 20, 1980.        New references  are included with    this report. However, references previously submitted are not being resubmitted.
Figure 2 illustrates the sequence of events required to mitigate the consequences of a main steam line break.The same initial sequence of events would occur for a feedwater line break.Since the same equipment is re-quired to operate and the same emergency procedure is used following a feedline break as a steam line break, but a steam line break is a more severe accident in 4 terms of RCS cooldown (return to criticality) and mass and energy release to containment, the subsequent discussion will address the main steam line break only.Table 2 lists the required equipment for each numbered block in Figure 2.1.A large main steam line break (greater than approxi-mately one square foot)would first be detected by the low steam line pressure sensors.Low steam line pres-sure sensed by two out of the three steam line pressure transmitters initiates safety injection accompanied by reactor and turbine trip.la.Diagnostic instrumentation is available to the operator to distinguish among accidents, as described in the LOCA discussion.
1n Section IV,  it  is either shown that each item is adequately qualified to perform its required safety function in its post-accident operating environment, or a commitment for additional testing or replacement is made. In all cases, sufficient justification for continued operation is given.
2.Two out of three low pressurizer pressure signals would provide additional protection for a larger steam line break and also provides the initial safety injec-12
Table  3  summarizes  the equipment qualification in the format requested  for SEP  by the NRC in a September  6, 1978  letter.
Table  4 provides the definition of environmental parameters throughout the Ginna plant. This table is comparable to Appendix A of F-C5257, and tabulates the explanatory basis given in Section    III of  this report.
Supplement No. 3 to IE Bulletin  79-01B provides the timing for submittal of qualification information for equipment in-stalled to meet the TMI Short Term Lessons Learned. RGSE intends to follow the guidance given in this supplement.           In a number of cases, it is possible that additional documentation or testing results may become available after November 1, 1980. Since this additional information will be of use in documenting the status of the Ginna environmental qualification, it will be  submitted when received.     Every  effort  has been made  to ensure that    all  documentation  was  obtained  for use with this submittal.


tion signal for smaller breaks.Also, high.containment pressure (6 psig)will initiate safety injection.
l II. IDENTIFICATION OF NECESSARY SAFETY RELATED EQUIPMENT This section of the report    identifies the necessary safety related equipment for   each  of the Design Basis Events (DBE) of concern and a brief description of why the equipment is needed. This identification includes all electrical equip-ment required by the Ginna emergency procedures for accomplish-ing the necessary safety functions. It must be recognized that not all electrical equipment referenced in the procedures is required to function ( as opposed to being useful if available), and is therefore not required to be qualified.
3.The Ginna design includes non-return check valves in each steam line just upstream of the main steam header in the intermediate building.Thus for any break upstream of the check valves, which includes all breaks inside containment, the check valves will preclude blowdown of the intact generator.
The emergency  operating procedures were not developed by considering safety-related components to the exclusion of all others. While such procedures are written with priority attention given to safety-related equipment, other systems and components  are  justifiably mentioned. A realistic evaluation of plant incidents might result in situations and hostile environments significantly less severe than those assumed for the purposes of conducting the environmental qualification program. The absence of full qualification for certain components which fall into this category is not, by itself, a sufficient motive to classify the equipment inoperable .or to remove these components from the procedures.
Reactor trip will result in closing the turbine stop valves.As redundant protection in the event of a steam line break upstream of the check valves, and for all breaks downstream of the check valves, the main steam line isolation valves are closed by several signals.These signals include 2/3 high containment pressure (20 psig);1/2 high steam flow in either steam line plus 2/4 low Tave plus safety injection; and 1/2 high-high steam flow in either steam line plus safety injection.
A. Events Accom an  in  a Loss Of  Coolant Accident Analyses of the course and consequences    of loss of coolant accidents have been submitted previously (LOCA 1-4). A discussion of equipment required to function to mitigate the consequences of a loss of coolant
4.The safety injecti~on signal closes the main and bypass f eedwater control valves, trips the f eedwater pumps and closes their respective discharge valves.5.The safety injection signal initiates containment isolation and containment ventilation isolation as described in the sequence of events in the loss of coolant accident.
6.The safeguards sequence as described in the loss of coolant accident is initiated by the safety injection signal.(For steam breaks outside containment, the spray pumps are not required.)
7.The safety injection signal trips the reactor and turbine.Other reactor trips are discussed in the FSAR, Section 7.8.The reactor coolant pumps'are tripped by manual operator action when low pressurizer pressure (1715 psig)is reached, and SI flow is initiated.
9.All valves associated with the safety injection systems are aligned and automatically function as de-scribed in the loss of coolant accident discussion.
If high containment pressure of 30 psig is reached, the containment spray system operates as described in the LOCA discussion.
10.When the boric acid storage tanks are drained to the 10%level and safety injection pump suction has automatically been aligned to the refueling water storage tank, the operator will reset safety injection and if reactor coolant pressure is above the shut-off head of the RHR pumps, will stop the RHR pumps and place them in the standby mode.A high steam line flow and/or low steam line pressure will indicate to the operator which steam generator has the steam line break.When this has been determined, 14 the operator will terminate AFW flow to the faulted steam generator, and align/maintain flow to the intact steam generator.
The inventory of the reactor coolant will be maintained by the remote manual operation of the high head safety injection pumps in combination with use of the charging p Umps~At least two hours after the start of the accident, supply water for the auxiliary feedwater pumps can be manually transferred from the condensate storage tanks to the service water system, by the method described in the LOCA discussion
[See Ref.SLB/FLB-6]
.If the auxiliary feedwater system is not operating properly, the operator can initiate operation from the control room of the Standby AFW system (using service water suction).11.If conditions and equipment availability permit, the operator can begin a gradual cooldown and depressuri-zation to cold shutdown conditions.
However, the primary safety function is to maintain the RCS in a safe condition at all times, removing decay heat at a rate comparable to the generation rate.Maintenance of this safe shutdown condition is accomplished by a combination of steam dump (to the condenser or atmosphere) with primary and secondary inventory makeup, accomplished by use of the safety injection and/or the charging 15


I pumps, and the auxiliary feedwater system.It is expected that RCS temperature can be lowered to near 212'F by using the steam generators.
accident is presented in the FSAR Chapters 6, 7 and 14.
The safe shutdown conditions can be maintained until a final cooldown and depressurization to ambient conditions can be effected.C.Hi h Ener Line Breaks Outside Containment An analysis has been provided describing the effects of pipe breaks outside containment
Post-LOCA operator actions are included in the Ginna Emergency Procedures. These procedures are consistent with the generic Westinghouse guidelines, which have been approved by the NRC. Additional descriptive material is presented in this report to provide summary information as to the sequence of events and the equipment involved at each stage. Figure 1 illustrates the sequence of events following a loss of coolant accident.
[HELB-1].The report proposed a program of augmented inservice inspection of certain piping welds in order to preclude the necessity to address further full diameter high energy piping breaks.Credible breaks of main steam lines outside containment, that is, those not included in the inspec-tion program, are bounded by a 6 inch main steam line branch connection in the Intermediate Building and a 12 inch main steam line branch connection in the Turbine Building.Credible breaks in the feedwater lines outside containment are bounded by a break in the 20 inch feedwater line in the Turbine Building.The accident environment created by these breaks, and other postulated breaks are provided in References
Table  1 provides a specific equipment list for each numbered block in Figure 1. Also provided in Table 1 is the safety function which is required and the period of time that operability must be ensured. It should      be noted that Table 1 includes all redundant equipment, not the minimum safeguards equipment assumed in the safety analysis. In the "required" column it should be noted that equipment listed as "signal initiation" is required to be operable only until its required safety function, the initiation of a safety signal, is performed.
[HELB 8-11].The program has been accepted by the NRC[Ref.HELB 7,8].Several modifications have been performed at the Ginna Nuclear Plant as a result of high energy line break analyses.Reference[HELB-1]discusses the various modifications, but of particular note is the Standby Auxiliary Feedwater system modification.
It is important to note that the arbitrary requirement of the DOR Guidelines to qualify equipment to function for at least one hour, even if its only function is completed within seconds, is not well reasoned.      In many cases,  the environment would not exist unless the equipment safety function had been completed (e.g.,
A-16
flooding to a seven foot level in containment by necessity means  that SI  was initiated). RGSE does not agree  with


remote-manual controlled standby auxiliary feedwater system, identical to the auxiliary feedwater system in cooling capability, has been installed.
this one-hour requirement,     and it is therefore not applied  as an  environmental qualification requirement.
The pumps are housed in a seismically designed structure (area 6 Figure 3)remote from the auxiliary feedwater and any high energy lines.Any portion of this system required to operate in an emergency is not subjected to an adverse environment.
Equipment  listed  as  "long term" is required to provide long term decay heat removal, post-accident monitoring and sampling, or maintaining a safe shutdown condition.
Ref[HELB-8]includes the NRC's Safety Evaluation Report concerning the RGGE modifications resultant from the review of Ref.[HELB-1].It includes a discussion of the acceptability of the instrumentation relocation and cable re-routing performed to insure that sufficient equipment will be protected from the environmental effects of a HELB outside containment.
Equipment  listed  as  "short term" is required only for  a short period of time (hours).
The failure of steam heating lines in the Auxiliary Building was identified and discussed in Ref.[HELB-1].It has been determined that steam heating lines also traverse other areas in the vicinity of safety related equipment[Ref.HELB-15].Modifications are planned which will isolate the steam heating line to the affected areas in the event of a failure and therefore preclude an adverse environment.
Table  3  provides the environmental qualification require-ments and documentation references for the Ginna Class IE equipment.
The commitment to perform analyses/modifications for those pipe breaks outside containment are given in Reference[HELB-13].
: 1. The  first event  in the loss of coolant accident following the rupture is the detection of the rupture.
Prior to its installation, regular inspections are being performed to reduce the likelihood of a failure creating an adverse environment.
Any 2/3 low pressurizer pressure or 2/3 high contain-ment pressure will initiate "safety injection" (SI).
These inspections, performed-17
la. Instrumentation is available to the operator to distinguish between a LOCA and the other accidents, such as a steam line break or feed line break.      It is important to note that the automatic actions and immediate operator actions (first 10 minutes) are identical in the mitigation of these accidents.
: 2. Upon  "safety injection" signal generation, safe-guards sequencing is initiated (see FSAR Table 8.2-4).
The diesel generators start and energize the safeguards buses assuming    there is a loss of offsite power. With the safeguards buses energized, either by off-site power or the diesels, the three safety injection pumps,


during each plant operating shift, would detect any leakage.Plant procedures (T-35F,"Steam to Auxiliary Building, Screen House, or Diesel Generators and Oil-Room")call for isolation of the affected piping promptly upon detection of the leakage.18 III.IDENTIFICATION OF THE LIMITING SERVICE ENVIRONMENTAL CONDI-TIONS FOR EQUIPMENT WHICH IS REQUIRED TO FUNCTION TO MITIGATE THE CONSEQUENCES OF DESIGN BASIS EVENTS This Section of the report defines the bases for and references to the environmental conditions encountered throughout the plant.A tabular summary is provided in Table 4.A.Inside Containment Post accident containment environmental conditions are discussed in Appendix 6E of the Ginna FSAR.These conditions result from a loss of coolant accident.The temperature and pressure profiles are given in Figures 1 and 2 of Appendix 6E with peak values being 286'F and 60 psig respectively.
the two residual heat removal pumps," two of the four service water pumps, the two motor driven auxiliary feedwater pumps, and the four containment. fan coolers
The radiation profile is presented in Figures 4 and 5 of Appendix 6E and it is seen, for example, that the doses at 30 minutes and one year following a LOCA are 1.7 x 10 and 1.6 x 10 rads, 6 8 respectively.(These figures are repeated as Figures 4,5,and 6 of this report.)Materials compatibility with post-accident chemical environment is discussed in detail in Appendix 6E.100$humidity is assumed.Design parameters
-will "be  loaded sequentially onto the buses. The two containment spray pumps are automatically loaded onto the buses when the 30 psig containment pressure    setpoint's reached.
'for environmental conditions have been conservatively selected for Ginna.As seen in FSAR Figure 14.3.4-2, the calculated peak pressure is less than 53 psig while the design value is 60 psig.The duration of the peak, similarly, bounds the cal-culated values.19 I
: 3. A  break in the reactor coolant system piping actuates the passive accumulator injection system when the reactor coolant system pressure is reduced to 700 ps lg The  flow path of the borated water from each accumulator-is through a series of check valves and a normally locked open (with AC control power removed) motor operated valve. The motor operated valves, MOV 841 and NOV 865, are not required to function to mitigate the consequences of the accident [Flood-1] .
Another example of the conservatism employed is seen in the accident radiation environment used for design purposes.As noted in WCAP 7744, a release of 100%of the noble gases, 50%of the halogens, and 1%, of all remaining fission products is assumed.In addition, no credit is taken for removal of radioactivity from the containment atmosphere by sprays, filters and fission product plateout.Finally, the specific activity in containment was roughly doubled by assuming a contain-ment free volume associated with an ice condenser con-tainment.Thus the radiation environment clearly over-states that which would be present even in a minimum safeguards case.This conservation is apparent from a comparison to the DOR Guidelines, which suggest a post-LOCA integrated dose of 2 x 10 rads gamma.7 Submergence of valves inside containment.
: 4.     The main steam isolation valves 3516 and 3517 close upon receiving a high containment pressure signal and the main and bypass feedwater control valves 4269, 4270, 4271 and 4272 close upon receiving a safety injection signal. The SI signal also causes a trip of the main feedwater pumps (which in turn causes the closing of the feedwater discharge valves). All of this equipment will fail in its safety position on loss of electrical power.
has previously been discussed in Reference[Flood-4]and it has been shown that operation following submergence is not required.Submergence of instrumentation has been discussed in Ref[Flood-5].
: 5.     "Containment Isolation" and "Containment Ventilation Isolation" (ref erred to collectively as simply, "Containment Isolation" ) is initiated by the saf ety injection signal.
Since the instrumentation is not required to function while flooded, no qualification for submergence is specified (see e.g., Section IV.19 of this report).The peak pressure following a MSLB is given in Section 14.2.5 of the FSAR as 52 psig, assuming no credit for containment pressure reducing equipment.
Containment isolation is discussed in detail in Section 5.2 of the FSAR. Most of the containment isolation valves are air operated valves. All air operated containment isolation valves close with safety injection signal with the exception of valves 4561 and 4562 which open full to insure service water supply to the containment recirculation fans. The f ail saf e position of the valves is the desired safeguard position as described above.
Recent analyses 20 for other facilities indicate that the containment vapor temperature following a MSLB in contaiment may briefly exceed those derived for a LOCA.These higher temperatures should not be limiting, however, for qual if ication of equipment required f ol lowing a MSLB, because: 1)the fact that the high temperature transient.
Six motor operated valves (313, 813, 814, ATV-1, ATV-2, ATV-3) receive a containment isolation signal. All of these valves are located outside of containment and only valves 313, 813, and   814 are fed from the safeguards buses.
is very brief and there is superheated steam (with its lower heat transfer capability) as opposed to saturated steam, 2)the equipment is protected from the direct effects of the steam line break by concrete floors and shields, and 3)the sensitive portions of the electrical equipment are not directly exposed to the environment, but are protected by housing, cable jackets, and the like.For these reasons, the humidity and steam environment following a LOCA remains limiting.This is consistent with the NRC's position 4.2 of the"Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors;" Radiation levels in containment following a MSLB are not limiting since fuel failures are not projected to result from a MSLB.Chemical environment and submergence are bounded by the LOCA conditions.
During normal operation ATV-1, ATV-2, and ATV-3 are closed with blank flanges installed on    their respective penetrations inside containment. The use of the process lines associated with these valves occurs only during the containment building integrated leak rate tests.
21
Valve 313, the reactor coolant pumps seal water return line,  and valves 813 and 814,  reactor coolant support inlet  and outlet lines, are closed  by the containment isolation signal.


B.Auxil iar Buil din The auxiliary building has a HVAC system which provides clean, f iltered and tempered air to the operating floor of the auxiliary building, and to the surface of the decontamination and spent fuel storage pits.The system exhausts air from the equipment rooms and open areas of the auxiliary building, and from the decon-tamination and spent fuel storage pits, through a closed exhaust system.The exhaust system includes a 100 percent capacity bank of high efficiency particulate air (HEPA)filters, and redundant 100 percent capacity fans discharging to the a'tmosphere via the plant vent.This arrangement insures the proper direction of air flow for removal of airborne radioactivity from the auxiliary building.Included in the auxiliary building exhaust system is a separate charcoal filter circuit, which exhausts from rooms where fission product activity may accumulate, during normal plant operation, in concentrations exceeding the average levels expected in the rest of the build-ing.Following a loss-of-coolant accident, this circuit is capable of providing exhaust ventilation from the areas containing pumps and related piping and valving which are used to recirculate containment sump liquid.A full flow charcoal filter bank is provided in the circuit, along with two 50 percent capacity exhaust 22 Vg fans.The air operated suction and discharge dampers associated with each fan are interlocked with the fan such that they are fully open when the fan is operating and fully closed when the fan is stopped.These dampers fail to the open position on loss of control signal or control air.The fans discharge to the main auxiliary building exhaust system, containing the HEPA filter bank.To assure a path for the charcoal (and HEPA)filtered exhaust to the plant vent if, the main exhaust fans are not operating, a fail open damper is installed in a bypass circuit around the two main exhaust fans.The residual heat removal, safety injection, containment spray and charging pump motors are provided with addi-tional cooling provisions to maintain ambient temperatures within acceptable limits when'the pumps are operating.
~
The charging pumps and RHR pumps are located in their own rooms, each room being provided with two cooling units consisting of redundant fans, water-cooled heat exchangers, and ductwork for circulating the cooled air.The capacity of each unit is sufficient to maintain acceptable room ambient temperatures with the minimum number of pumps required for system operation in service.The safety injection and containment spray pumps are 0 provided with cooling units providing cool air directly to the motor.There is a separate fan for each of the motors.23
: 6. The SI  signal trips the reactor and turbine.
Other reactor trips are discussed in the FSAR, Section 7.
: 7. The reactor coolant pumps are tripped by manual operator action when low pressurizer pressure (1715
  .psig) is reached, and SI flow is initiated.
: 8. Selected valves throughout the plant provide flow paths  for the required    safeguards    equipment with the advent of the SI signal.
During normal operation     all  required valves in the flow paths for high head safety injection 'are normally open with the exception of valves 826A and 826C, the dis-charge valves from the boric acid storage tank to the suction of the safety injection pumps.
Valves 826A; B,   C and   D  receive the safety injection signal  and valves 82 6A  and C open providing borated water to the reactor coolant loop cold legs.
When  the level in the boric acid storage tank decreases to the   10% level, suction for the high head safety in-jection pumps is automatically switched from the boric acid storage tanks to the refueling water storage tank by the automatic opening of, valves 825A and B and closing of valves  826A, B,     C  and D.
During normal operation,      all  valves in the flow paths for low head safety   injection are normally      open except


In the event of a loss of offsite power, the auxiliary building ventilation system main supply and exhaust f ans would be inoperable.
for  MOV  852A and  MOV  852B, the  valves in the vessel upper plenum    injection lines. These      valve's open upon receipt of   a safety injection signal    and remain open
However, all other fans in the auxiliary building ventilation system are supplied by emergency diesel power including the charcoal filter circuit and the pump cooling circuits for safety related pump motors, as described above.Since the auxiliary building is a very large volume building, it is not'\expected that there would be a post-accident tempera-ture increase except in some local areas near hot piping and large motors.This situation exists only in the basement of the auxiliary building where the safety-related pumps and recirculated sump fluid piping are located.As shown in Reference[HELB-14]the ventila-tion system for these areas is expected to be adequate to maintain the post-accident temperature with normal"ambient" levels.Further detailed evaluation of the environment in these areas is being addressed with the final resolution of the"mild" environment qualification requirements
- thereaf ter.
.The radiation levels in the auxiliary building will increase in the event of a LOCA.Using very conservative post-accident fission product activity levels, the post-accident environment in the auxiliary building was calculated in Appendix A to Reference[TMI-3].It is apparent from Table 5-1 of this reference that the only major radiation field in terms of equipment qualification 24
The containment spray pumps      will automatically start and the discharge valves 860A        Bg C  and  D  automatically open, receiving power from the safeguards          buses when containment pressure reaches      30  psig. If containment pressure does not reach 30 psig, the operator may manually start the spray pumps after all other safeguards are loaded on the safeguards buses. Automatic NaOH addition via opening of valves      HCV  836A,  B  takes place two minutes    after containment spray    pump  start  unless defeated manually.
The containment spray pumps are      normally aligned to the refueling water storage tank with all suction valves.
open.
SI system actuation    will automatically align the two post accident charcoal f ilters to the containment recirculation system by opening inlet dampers 5871 and 5872, and  outlet  dampers  5873 and 5874.      Loop  entry dampers  5875 and 5876    will close. These dampers    will fail  to their safeguards position upon loss of electric power.
: 9. The  control  room  ventilation is automatically placed in the 100% recirculation mode ( with about 25%
flow through charcoal filters), when SI is initiated.
: 10. Af ter the  safety injection pumps are automatically switched from the boric acid storage tanks to the re-fueling water storage tanks, the operator resets safety injection, starts the component cooling water pumps and aligns flow to the RHR heat exchangers, and initiates SW flow to the'CW heat exchangers.         At the 31% RWST alarm, the operator shuts off one CS and one SI pump (if more than one are running). When the refueling water storage tank level is reduced to 10%, the plant operator stops the remaining residual heat removal, containment spray and high head safety injection pumps and establishes f low paths to the reactor vessel for both high ( if required) and low head safety injection from containment    sump   B.
The normal    (non-saf ety grade) auxiliary f eedwater supply source is from the condensate storage tanks. If this supply is exhausted the operator opens the motor operated valves 4027 and 4028 and manual operated valves 4344 and 4345 to provide service water to the suction of the auxiliary feedwater pumps. If the AFW system is not functioning properly, the operator can align from the control      room the Standby  AFW system to the steam generators    ( using'ervice water suction).
: 11. In the recirculation phase, the operator aligns the RHR pumps to containment sump B by opening valve 850A  for  pump A and  valve  850B for pump B, and  closing 10


will be in the vicinity of the recirculated fluid.The required qualification doses are addressed for all the affected equipment in Table 3.The RGEE commitments to-ensure that a HELB in the auxiliary building will not affect the capability of effecting and maintaining a safe shutdown condition is provided in Reference[HELB-13].
valve 704A, 704B, 856, and      896A  or 896B. For low head recirculation, injection is through the vessel nozzles.
Flooding is not a concern in the Auxiliary Building.Even in the event of leakage, two 50 gpm sump pumps are provided in the low point of the*building.
,For high head recirculation, the RHR pumps discharge to the safety injection pumps through alignment of valve 857A (for RHR pump B) and/or valves 857B and 857C (for RHR pump  A). Valves  AOV  897, 898 are closed.      The  high head  safety injection pumps then provide water to the cold leg injection points. This alignment also allows CS pump operation, if desired.
This is described in Section 9.3 of the FSAR, and has been evaluated by the NRC in Reference[HELB-15].
Long term  recirculation to compensate for the possible effects of boron precipitation has been described in Ref [Flood-1] and includes the use of        RHR  pumped  flow to the vessel nozzles and through a high        head safety injection pump into either cold leg.
Intermediate Buildin Implementation of an augmented inservice inspection program for high energy piping outside containment has reduced the probability of pipe breaks in these systems to acceptably low levels[Ref.HELB-7, 8].A six inch main steam line branch connection is the intermediate building DBE.Based on the f ailure capacity of portions of the exterior walls, the limiting pressure is established in Ref.[HELB-1]as being a pressure of 0.80 psig.Assuming saturation conditions, one obtains a limiting I'I temperature of approximately 215'F.A 100%humidity steam-air mixture is assumed.If the pipe crack or branch line break were in a portion of the steam or~~f eed line that could be isolated, the isolation would immediately halt the mass and energy addition to the intermediate building.A pipe crack or branch line 25 which could not be isolated is the limiting DBE for intermediate building environment.
Post-accident reactor coolant and containment atmosphere sampling modifications are presently being undertaken, in accordance with the implementation schedule for the TMX  Lessons  Learned commitments. See  [Ref TMI-3].
Mass and energy release in this case would be limited by the dryout of the steam generators with the duration of the environment dependent on the size of the leak or break.Based on flow through a main steam safety valve (a 6 inch line)of 247 lbs/sec at a steam line pressure of 1100 psia and the inventory available for release from a main steam break of 165,500 lbs (FSAR Section 14.2.5), the mass and energy flow will continue for at least 11 minutes.Smaller leaks may continue substantially longer.Zt is expected that within 30 minutes to an hour, action could be taken to provide added ventilation to the building by opening doors.Within several hours, return to near ambient could be accomplished.
Events Accom an    in  a Main Steam  Line Break or  a Main Feed Line Break The analyses    of  a main steam  line break or  a main  feed line break  and  the consequences thereof have been discussed in Chapters 6 and'14 of the FSAR and in References [SLB/FLB 2-4]. The High Energy Line Break analyses [HELB 1-7] provide additional information regarding steam line breaks outside of containment, as 11
Table 4 provides an estimate of the duration of the environmental transient expected.The exact duration is not critical in terms of affected equipment qualification; therefore, no explicit calculations have been performed.
 
Chemical spray is not a design consider-ation in this building.The effects of submergence need not be considered, as described in References
well  as  feedwater line breaks inside and outside containment.
[HELB-1],[HELB-4], and[FLOOD-11'].
Figure  2  illustrates    the sequence of events required to mitigate the consequences        of a main steam line break.
This latter reference presents the result of an analysis performed to ensure that safety-related equipment would not be flooded in the event of an feed line break in the intermediate building.26 The radiation environment was reviewed in response to the TMI Lessons Learned commitments
The same initial sequence of events would occur for a feedwater line break. Since the same equipment is re-quired to operate and the same emergency procedure is used  following    a  feedline break    as a steam    line break, but  a steam    line break is    a more  severe accident in terms of    RCS 4
[see Ref.TMI-3].It can be seen from Table 5-1 that the radiation environ-ment is not significant in terms of equipment qualification.
cooldown    (return to  criticality)    and mass and energy release      to containment, the subsequent discussion will address the main steam line break only.
Table 2 lists the required equipment for each numbered block in Figure 2.
: 1. A  large main steam line break        ( greater than approxi-mately one square foot) would first be detected by the low steam line pressure sensors.            Low steam line pres-sure sensed by two out of the three steam            line pressure transmitters initiates safety injection accompanied by reactor and turbine trip .
la. Diagnostic instrumentation is available to the operator to distinguish        among  accidents,    as described in the LOCA discussion.
: 2. Two  out of three low pressurizer pressure signals would provide additional protection            for a  larger steam line break    and also provides the      initial  safety injec-12
 
tion signal for smaller breaks.        Also, high. containment pressure  ( 6 psig) will initiate safety injection.
: 3. The Ginna  design includes non-return check valves in  each steam  line just  upstream of the main steam header in the intermediate building.        Thus for  any break upstream of the check valves, which includes all breaks inside containment, the check valves will preclude blowdown of the    intact generator. Reactor trip will result in closing the turbine stop valves. As redundant protection in the event of a steam line break upstream of the check valves, and for all breaks downstream of the check valves, the main steam line isolation valves are closed by several signals. These signals include 2/3 high containment pressure (20 psig); 1/2 high steam flow in either steam line plus 2/4 low Tave plus safety injection; and 1/2 high-high      steam  flow in either steam line plus safety injection.
: 4. The safety injecti~on signal closes the main and bypass f eedwater control valves, trips the f eedwater pumps and closes their respective discharge valves.
: 5. The safety injection signal initiates containment isolation    and containment  ventilation isolation as described in the sequence    of events in the loss of coolant accident.
: 6. The safeguards    sequence  as  described in the loss of coolant accident is initiated by the safety injection signal. ( For steam breaks outside containment, the spray  pumps  are not required.)
: 7. The  safety injection signal trips the reactor      and turbine. Other reactor trips are discussed in the FSAR,  Section 7.
: 8. The  reactor coolant pumps'are tripped by manual operator action when low pressurizer pressure (1715 psig) is reached, and SI flow is initiated.
: 9. All valves associated with the safety injection systems are aligned and automatically        function as  de-scribed in the loss of coolant accident discussion. If high containment pressure of 30 psig is reached, the containment spray system operates as described in the LOCA  discussion.
: 10. When  the boric acid storage tanks are drained to the  10%  level  and safety injection pump suction has automatically    been aligned to the refueling water storage tank, the operator will reset safety injection and if reactor coolant pressure is above the shut-off head of the    RHR  pumps,  will stop  the  RHR pumps and place them in the standby mode.
A  high steam line flow and/or low steam line pressure will indicate to the operator which steam generator has the steam line break. When this has been determined, 14
 
the operator  will terminate    AFW flow to the faulted steam generator,    and  align/maintain flow to the intact steam generator.
The  inventory of the reactor coolant will be maintained by the remote manual operation of the high head safety injection pumps in combination with use of the charging p Umps ~
At least two hours after the      start of the accident, supply water  for the auxiliary feedwater pumps can    be manually transferred from the condensate storage tanks to the service water system, by the method described in the  LOCA  discussion [See Ref . SLB/FLB-6] . If the auxiliary feedwater system is not operating properly, the operator can initiate operation from the control room of the Standby AFW system (using service water suction).
: 11. If conditions  and equipment  availability permit, the operator can begin      a gradual cooldown and depressuri-zation to cold shutdown conditions. However, the primary safety function is to maintain the RCS in a safe condition at all times, removing decay heat at a rate comparable to the generation rate. Maintenance of this safe  shutdown  condition is accomplished by a combination of steam dump ( to the condenser or atmosphere) with primary and secondary inventory makeup, accomplished by use of the safety injection and/or the charging 15
 
I pumps, and the  auxiliary feedwater system. It is expected that  RCS temperature can be lowered to near 212'F by using the steam generators.      The safe shutdown conditions can be maintained until a final cooldown and depressurization to ambient conditions can be effected.
C. Hi h Ener    Line Breaks Outside Containment An  analysis has been provided describing the effects of pipe breaks outside containment [HELB-1]. The report proposed a program of augmented inservice inspection of certain piping welds in order to preclude the necessity to address further full diameter high energy piping breaks. Credible breaks of main steam lines outside containment, that is, those not included in the inspec-tion program, are  bounded by a  6 inch main steam line branch connection in the Intermediate Building and a 12 inch main steam line branch connection in the Turbine Building. Credible breaks in the feedwater lines outside containment are bounded by a break in the      20 inch feedwater line in the Turbine Building. The accident environment created by these breaks, and other postulated breaks are provided in References [HELB 8-11]. The program has been accepted by the NRC [Ref.
HELB 7,8]. Several modifications have been performed at the Ginna Nuclear Plant as a result of high energy line break analyses. Reference [HELB-1] discusses the various modifications, but of particular note is the Standby Auxiliary Feedwater  system  modification. A 16
 
remote-manual  controlled standby auxiliary feedwater system, identical to the auxiliary feedwater system in cooling capability, has been installed. The pumps are housed in a seismically designed structure (area 6 Figure 3) remote from the auxiliary feedwater and any high energy lines. Any portion of this system required to operate in an emergency is not subjected to an adverse environment. Ref [HELB-8] includes the NRC's Safety Evaluation Report concerning the RGGE modifications resultant from the review of Ref. [HELB-1]. It includes a discussion of the acceptability of the instrumentation relocation and cable re-routing performed to insure that sufficient equipment will be protected from the environmental effects of a HELB outside containment.
The  failure of steam heating lines in the Auxiliary Building was identified and discussed in Ref . [HELB-1].
It has  been determined  that  steam heating  lines also traverse other areas in the vicinity of safety related equipment [Ref. HELB-15]. Modifications are planned which will isolate the steam heating line to the affected areas in the event of a failure and therefore preclude an adverse  environment. The commitment  to perform analyses/modifications  for  those pipe breaks outside containment are given in Reference    [HELB-13]. Prior to its installation, regular    inspections are being performed to reduce the likelihood of a failure creating an adverse environment. These inspections, performed 17
 
during each plant operating shift, would detect any leakage. Plant procedures (T-35F, "Steam to Auxiliary Building, Screen House, or Diesel Generators and Oil
- Room" ) call for isolation of the affected piping promptly upon detection of the leakage.
18
 
III. IDENTIFICATION OF    THE  LIMITING SERVICE ENVIRONMENTAL CONDI-TIONS FOR EQUIPMENT WHICH IS REQUIRED TO FUNCTION TO MITIGATE THE CONSEQUENCES    OF DESIGN BASIS EVENTS This Section of the report defines the bases        for and  references to the environmental conditions encountered throughout the plant. A tabular summary is provided in Table 4.
A. Inside Containment Post accident containment environmental conditions are discussed in Appendix 6E of the Ginna FSAR. These conditions result from a loss of coolant accident. The temperature and pressure profiles are given in Figures 1  and  2 of Appendix  6E  with peak values being 286'F and 60  psig respectively.      The radiation profile is presented in Figures 4 and 5 of Appendix 6E and      it is seen, for example, that the doses at 30 minutes and one year following a LOCA are 1.7 x 10 6 and 1.6 x 10 8 rads, respectively.    (These  figures are repeated as Figures 4,5,and 6 of this report.) Materials compatibility with post-accident chemical environment is discussed in detail in Appendix 6E. 100$ humidity is assumed.
Design parameters    'for environmental conditions have been    conservatively selected for Ginna. As seen in FSAR  Figure 14.3.4-2, the calculated peak pressure is less than 53 psig while the design value is 60 psig.
The duration of the peak, similarly, bounds the cal-culated values.
19
 
I Another example of the conservatism employed is seen in the accident radiation environment used for design purposes. As noted in WCAP 7744, a release of 100% of the noble gases, 50% of the halogens, and 1%, of all remaining fission products is assumed.        In addition, no credit is taken for removal of radioactivity from the containment atmosphere by sprays,        filters and  fission product plateout. Finally, the specific activity in containment was roughly doubled by assuming a contain-ment free volume associated with an ice condenser con-tainment. Thus the radiation environment clearly over-states that which would      be present even in  a minimum safeguards  case. This conservation is apparent from      a comparison to the    DOR Guidelines, which suggest      a post-LOCA integrated dose of 2 x 10 7 rads gamma.
Submergence  of valves inside containment. has previously been discussed    in Reference [Flood-4] and    it  has been shown that operation following submergence is not required. Submergence of instrumentation has been discussed in Ref [Flood-5]. Since the instrumentation is not required to function while flooded, no qualification for submergence is specified (see e.g., Section IV.19 of this report) .
The peak  pressure following    a MSLB  is given in Section 14.2.5 of the    FSAR as  52  psig, assuming no credit for containment pressure reducing equipment.        Recent analyses 20
 
for other facilities indicate that    the containment vapor temperature following a    MSLB in contaiment  may briefly exceed  those derived  for a LOCA. These higher temperatures  should not be  limiting, however, for qual ification  of equipment required  fol lowing a MSLB, because:
: 1)    the  fact that the high temperature transient. is very brief and there is superheated steam (with its lower heat transfer capability) as opposed to saturated steam,
: 2)    the equipment  is protected from the direct effects of the steam  line break by concrete floors and shields, and
: 3)    the sensitive portions of the electrical equipment are not directly exposed to the environment, but are protected by housing, cable jackets, and the like.
For these reasons,    the humidity and steam environment following a LOCA remains limiting. This is consistent with the NRC's position 4.2 of the "Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors;" Radiation levels in containment following a MSLB are not limiting since fuel failures are not projected to result from a MSLB. Chemical environment and submergence    are bounded by the  LOCA conditions.
21
 
B. Auxiliar Buildin The auxiliary building has    a HVAC system which  provides clean, f iltered and tempered air to the operating floor of the auxiliary building, and to the surface of the decontamination and spent fuel storage pits. The system exhausts    air  from the equipment rooms and open areas of the auxiliary building, and from the decon-tamination and spent fuel storage pits, through a closed exhaust system. The exhaust system includes a 100 percent capacity bank of high efficiency particulate air (HEPA) filters, and redundant 100 percent capacity fans discharging to the a'tmosphere via the plant vent.
This arrangement insures the proper direction of air flow for removal of airborne radioactivity from the auxiliary building.
Included in the auxiliary building exhaust system is a separate charcoal filter circuit, which exhausts from rooms where  fission product activity    may accumulate, during normal plant operation, in concentrations exceeding the average levels expected in the rest of the build-ing. Following a loss-of-coolant accident, this circuit is capable of providing exhaust ventilation from the areas containing pumps and related piping and valving which are used to    recirculate containment    sump liquid.
A  full flow charcoal filter bank    is provided in the circuit,  along with two 50 percent capacity exhaust 22
 
Vg fans. The  air  operated suction and discharge dampers associated    with each fan are interlocked with the fan such that they are fully open when the fan is operating and fully closed when the fan is stopped.            These dampers fail to the open position on loss of control signal or control air. The fans discharge to the main auxiliary building exhaust system, containing the HEPA filter bank. To  assure  a path  for the charcoal    (and HEPA) filtered    exhaust to the plant vent    if, the  main exhaust fans are not operating,      a  fail open  damper  is installed in  a bypass    circuit  around the two main exhaust fans.
The  residual heat removal, safety injection, containment spray and charging pump motors are provided with addi-tional cooling provisions to maintain ambient temperatures within acceptable limits when'the pumps are operating.
The charging pumps and RHR pumps are located in their own rooms,    each room being provided    with two cooling units consisting of redundant fans, water-cooled heat exchangers,    and ductwork    for circulating the cooled air. The    capacity of each unit is sufficient to maintain acceptable room ambient temperatures        with the minimum number of pumps      required for system operation in service.
The  safety injection and containment spray pumps are 0
provided with cooling units providing cool air directly to the motor. There is a separate fan for each of the motors  .
23
 
In the event of    a  loss of  offsite power, the  auxiliary building ventilation system      main supply and exhaust f ans  would be inoperable. However,  all other fans in the auxiliary building ventilation system are supplied by emergency diesel power including the charcoal filter circuit    and the pump  cooling circuits for safety related pump motors, as described above.      Since the auxiliary building is a very large volume building,      it
                                                '\
is not expected that there would be      a post-accident tempera-ture increase except in some local areas near hot piping and large motors. This situation exists only in the basement of the auxiliary building where the safety-related pumps and recirculated sump fluid piping are located. As shown  in Reference [HELB-14] the ventila-tion  system  for these areas is expected to be adequate to maintain the post-accident temperature with normal "ambient" levels. Further detailed evaluation of the environment in these areas is being addressed with the final resolution of the "mild" environment qualification requirements    .
The  radiation levels in the auxiliary building will increase in the event of a LOCA. Using very conservative post-accident fission product activity levels, the post-accident environment in the auxiliary building was calculated in Appendix A to Reference [TMI-3]. It is apparent from Table 5-1 of this reference that the only major radiation field in terms of equipment qualification 24
 
will be in the vicinity of     the recirculated   fluid. The required qualification doses are addressed         for all the affected equipment in Table 3. The RGEE commitments to
- ensure that a HELB in the auxiliary building will not affect the capability of effecting and maintaining a safe shutdown condition is provided in Reference         [HELB-13].
Flooding is not a     concern in the Auxiliary Building.
Even in the event of leakage, two 50 gpm sump pumps are provided in the low point of the*building. This is described in Section 9.3 of the FSAR, and has been evaluated by the NRC in Reference [HELB-15].
Intermediate Buildin Implementation of an augmented inservice inspection program for high energy piping outside containment has reduced the probability of pipe breaks in these systems to acceptably low levels [Ref . HELB-7, 8] . A six inch main steam line branch connection is the intermediate building DBE. Based on the f ailure capacity of portions of the exterior walls, the limiting pressure is established in Ref . [HELB-1] as being a pressure of 0.80 psig.
Assuming saturation conditions, one obtains a limiting I
                    'I temperature of approximately 215'F.         A 100% humidity steam-air mixture is assumed. If the pipe crack or branch line break were in a portion of the steam or
  ~ ~
f eed line that could be isolated, the isolation would immediately halt the mass and energy addition to the intermediate building.       A pipe crack or branch line 25
 
which could not be isolated   is the limiting DBE for intermediate building environment. Mass and energy release in this case would be limited by the dryout of the steam generators with the duration of the environment dependent on the size of the leak or break. Based on flow through a main steam safety valve (a 6 inch line) of 247 lbs/sec at a steam line pressure of 1100 psia and the inventory available for release from a main steam break of 165,500 lbs ( FSAR Section 14.2.5), the mass and energy flow will continue for at least 11 minutes. Smaller leaks may continue substantially longer. Zt is expected that within 30 minutes to an hour, action could be taken to provide added ventilation to the building by opening doors. Within several hours, return to near ambient could be accomplished.
Table 4 provides an estimate of the duration of the environmental transient expected. The exact duration is not critical in terms of affected equipment qualification; therefore, no explicit calculations have been performed. Chemical spray is not a design consider-ation in this building. The effects of submergence need not be considered, as described in References
[HELB-1], [HELB-4], and [FLOOD-11']. This latter reference presents the result of   an analysis performed to ensure that safety-related equipment would not be flooded in the event of an feed line break in the intermediate building.
26
 
The radiation environment     was reviewed in response   to the TMI Lessons   Learned commitments     [see Ref . TMI-3] .
It can be seen from Table 5-1     that the radiation environ-ment is not significant in terms of equipment qualification.
Cable Tunnel Since the cable tunnel is open to the Intermediate Building, the limiting environmental conditions for the cable tunnel are identical to the Intermediate Building conditions.
Cable Tunnel Since the cable tunnel is open to the Intermediate Building, the limiting environmental conditions for the cable tunnel are identical to the Intermediate Building conditions.
Control Buildin The limiting environment of the Control Building which includes the control room, relay room, and battery rooms, is normal ambient conditions.
Control Buildin The limiting environment of the Control Building which includes the control room, relay room, and battery rooms, is normal ambient conditions. Protection against high energy line breaks and circulating water line breaks which could occur outside the Control Building and affect the Control Building environment       are identified and discussed   in References   [HELB-1, HELB-6, HELB-7, HELB-13, HELB-15, FLOOD-1, and FLOOD-5]       .
Protection against high energy line breaks and circulating water line breaks which could occur outside the Control Building and affect the Control Building environment are identified and discussed in References
The air conditioning   system for   the control room is described in Section 9.9 of the FSAR. The main air handling unit and circulation fans for the control room are powered from   a single Class IE motor control center
[HELB-1, HELB-6, HELB-7, HELB-13, HELB-15, FLOOD-1, and FLOOD-5].The air conditioning system for the control room is described in Section 9.9 of the FSAR.The main air handling unit and circulation fans for the control room are powered from a single Class IE motor control center (MCC-1K), which receives power from a diesel-backed emergency bus (diesel 1A).If there were a failure of this train during the post accident period, it would be possible to crosstie to the 1B diesel.The operator, after assuring that any faults are cleared, would close 27  
( MCC-1K), which   receives power from a diesel-backed emergency bus (diesel 1A). If there were a failure of this train during the post accident period,         it would be possible to crosstie to the 1B diesel. The operator, after assuring that     any faults are cleared,   would close 27
 
the bus tie between buses 14 and 16'to energize the in-operable-Control Room air handling unit from the 1B diesel, while making sure that the operational diesel
- does not become overloaded. This emergency bus cross-ties procedure has previously been included in the Ginna Emergency Procedures  .
The  control room HVAC system has been    out of service several times in the last  11 years  for  maintenance. A satisfactory environment has been maintained by opening the two control room doors and two relay room doors, connecting the two rooms together and with outside      air, to provide natural circulation. Equipment failure has never been experienced during these events because of        a temperature increase due to lack of HVAC.
It is  also possible, of course, to provide for the use of portable air-conditioning units or fans to maintain environmental conditions within proper specifications.
Further evaluation of the long-term effects of the loss of ventilation  will be made at a  later time,  when safety-related equipment not exposed to a "harsh" accident environment is addressed in terms of environmental qualification.
The  relay room is normally cooled by two non-safety-related air conditioning systems, which can be manually aligned to the emergency buses by closing the proper bus-tie breakers.
28
 
Natural circulation with the control room, and the use of portable air-conditioning units and fans, are options available to maintain environmental conditions within the required specifications. Further evaluation con-cerning loss of ventilation will be made at  a later time, together with the control room study.
To  further assure that  a loss of ventilation to the control and relay rooms is not expected to be a concern, RG&E conducted an 8-hour test on September 15, 1980.
It was  demonstrated  that, for a loss of all HVAC, no, significant temperature increase occurred in the control room or relay room. Only the plant computer, located in its own room within the relay room, and not required for accident mitigation or safe shutdown, appeared to be susceptible to overheating.
The  battery rooms have a set of inlet and exhaust fans, as well as an air-conditioning system. Additional fans are to be installed in the near future. These fans will be d.c.-powered 'directly from the batteries.
While this modification is in progress, the present Emergency Procedures provide for manual alignment to the emergency buses by closing of bus-tie breakers. If necessary,  portable fans could be used to provide sufficient air handling capacity to maintain the battery rooms at acceptable ambient conditions.
29
 
F. Diesel Generator  Rooms The emergency  diesel generator  rooms each have  their own HVAC  system, powered from the diesels.      As soon as the diesels are brought    up to speed, stabilized, and their respective circuit breakers closed to their emergency buses, the HVAC systems ( ventilating fans) are energized. Protection against failure of steam heating lines in the rooms is described in Section II.C above. Failure of a steam heating line would affect only one diesel. The other diesel, as well as offsite power, would still be available. This configuration has been reviewed by the    NRC in Reference [HELB-15],  ~
and found  acceptable. Protection agains events outside the rooms is described in References [HELB-1, HELB-6, HELB-7, FLOOD-1, and FLOOD-5].      The limiting environment in the diesel generator rooms    therefore is normal ambient conditions.
G. Turbine Buildin The turbine building does not require an HVAC system per se, but rather utilizes roof vent fans, wall vent vans, windows and unit heaters for control of the en-virons. In the event of loss of power to fans in this building there would be no significant temperature rise, since it is a large volume building with sufficient openings ( windows and access doors) to adequately cir-culate outside air.
30


the bus tie between buses 14 and 16'to energize the in-operable-Control Room air handling unit from the 1B diesel, while making sure that the operational diesel-does not become overloaded.
Analyses have shown  that the limiting pressure are caused by an instantaneous break in the 20 inch feed line in the turbine building. See Reference [HELB-1].
This emergency bus cross-ties procedure has previously been included in the Ginna Emergency Procedures
Peak pressures  are 1.14 psig on the lower two levels of the building and 0. 70 ps ig on the operating floor.
.The control room HVAC system has been out of service several times in the last 11 years for maintenance.
Failure of portions of the exterior wall limit the duration of the pressure pulse to,a f ew seconds.
A satisfactory environment has been maintained by opening the two control room doors and two relay room doors, connecting the two rooms together and with outside air, to provide natural circulation.
Pressure and temperature is limited by the failure capacity of the exterior walls. Assuming saturation conditions, one obtains a limiting temperature of approximately 220'F. A 100% humidity steam-air mixture is assumed. Isolation of the main steam and feed system will isolate the source of energy to the turbine building. Temperature and pressure reduction will be accomplished by opening  exterior doors  and windows and as a  result of leakage through  known openings  to the outside. For conservatism,  it has  been assumed  that the peak temperature condition persists for 30 minutes with return to ambient being accomplished in a total of 3 hours. For conservatism, peak pressures  are assumed to persist for 1 minute with return to ambient being accomplished in a total of 3 hours. (This is tabulated in Table 4). The exact duration of high environmental 31
Equipment failure has never been experienced during these events because of a temperature increase due to lack of HVAC.It is also possible, of course, to provide for the use of portable air-conditioning units or fans to maintain environmental conditions within proper specifications.
Further evaluation of the long-term effects of the loss of ventilation will be made at a later time, when safety-related equipment not exposed to a"harsh" accident environment is addressed in terms of environmental qualification.
The relay room is normally cooled by two non-safety-related air conditioning systems, which can be manually aligned to the emergency buses by closing the proper bus-tie breakers.28


Natural circulation with the control room, and the use of portable air-conditioning units and fans, are options available to maintain environmental conditions within the required specifications.
conditions is not critical in terms of affected equipment qualification; therefore, no explicit calculations have been performed.
Further evaluation con-cerning loss of ventilation will be made at a later time, together with the control room study.To further assure that a loss of ventilation to the control and relay rooms is not expected to be a concern, RG&E conducted an 8-hour test on September 15, 1980.It was demonstrated that, for a loss of all HVAC, no, significant temperature increase occurred in the control room or relay room.Only the plant computer, located in its own room within the relay room, and not required for accident mitigation or safe shutdown, appeared to be susceptible to overheating.
Limiting flood conditions are the result of a circulating water system pipe break and is a water level of 18 inches in the basement [FLOOD-5].
The battery rooms have a set of inlet and exhaust fans, as well as an air-conditioning system.Additional fans are to be installed in the near future.These fans will be d.c.-powered
Auxiliar Buildin Annex This structure, which houses the Standby Auxiliary Feedwater System, is described in References [HELB-1]
'directly from the batteries.
and [HELB-6] . The limiting environment in this structure is normal ambient conditions. The cooling system for this building is redundant and seismically qualified.
While this modification is in progress, the present Emergency Procedures provide for manual alignment to the emergency buses by closing of bus-tie breakers.If necessary, portable fans could be used to provide sufficient air handling capacity to maintain the battery rooms at acceptable ambient conditions.
Flooding is not  a  concern since  all safety-related equipment associated    with the Standby AFW System is elevated so that  a  complete failure of the Condensate Tank would not cause submergence.
29 F.Diesel Generator Rooms G.The emergency diesel generator rooms each have their own HVAC system, powered from the diesels.As soon as the diesels are brought up to speed, stabilized, and their respective circuit breakers closed to their emergency buses, the HVAC systems (ventilating fans)are energized.
Screen House The screen house,    like  the turbine building, does not require an HVAC per se, but utilizes roof vent fans, wall vent fans, windows, and unit heaters for control of the environs. Xn the event of a loss of power to the fans, there would be no significant temperature rise, since it is a large volume building with suf f icient openings to adequately circulate outside air.
Protection against failure of steam heating lines in the rooms is described in Section II.C above.Failure of a steam heating line would affect only one diesel.The other diesel, as well as offsite power, would still be available.
32
This configuration has been reviewed by the NRC in Reference[HELB-15],~and found acceptable.
Protection agains events outside the rooms is described in References
[HELB-1, HELB-6, HELB-7, FLOOD-1, and FLOOD-5].The limiting environment in the diesel generator rooms therefore is normal ambient conditions.
Turbine Buildin The turbine building does not require an HVAC system per se, but rather utilizes roof vent fans, wall vent vans, windows and unit heaters for control of the en-virons.In the event of loss of power to fans in this building there would be no significant temperature rise, since it is a large volume building with sufficient openings (windows and access doors)to adequately cir-culate outside air.30


Analyses have shown that the limiting pressure are caused by an instantaneous break in the 20 inch feed line in the turbine building.See Reference[HELB-1].Peak pressures are 1.14 psig on the lower two levels of the building and 0.70 ps ig on the operating floor.Failure of portions of the exterior wall limit the duration of the pressure pulse to,a f ew seconds.Pressure and temperature is limited by the failure capacity of the exterior walls.Assuming saturation conditions, one obtains a limiting temperature of approximately 220'F.A 100%humidity steam-air mixture is assumed.Isolation of the main steam and feed system will isolate the source of energy to the turbine building.Temperature and pressure reduction will be accomplished by opening exterior doors and windows and as a result of leakage through known openings to the outside.For conservatism, it has been assumed that the peak temperature condition persists for 30 minutes with return to ambient being accomplished in a total of 3 hours.For conservatism, peak pressures are assumed to persist for 1 minute with return to ambient being accomplished in a total of 3 hours.(This is tabulated in Table 4).The exact duration of high environmental 31 conditions is not critical in terms of affected equipment qualification; therefore, no explicit calculations have been performed.
RG&E's commitment  to resolve the HELB environment is provided in Section II. C. Protection against f looding is described in Ref erences  [FLOOD-1] and [FLOOD-5] .
Limiting flood conditions are the result of a circulating water system pipe break and is a water level of 18 inches in the basement[FLOOD-5].
The, limiting environment in the  screenhouse is thus normal ambient conditions.
Auxiliar Buildin Annex This structure, which houses the Standby Auxiliary Feedwater System, is described in References
33
[HELB-1]and[HELB-6].The limiting environment in this structure is normal ambient conditions.
The cooling system for this building is redundant and seismically qualified.
Flooding is not a concern since all safety-related equipment associated with the Standby AFW System is elevated so that a complete failure of the Condensate Tank would not cause submergence.
Screen House The screen house, like the turbine building, does not require an HVAC per se, but utilizes roof vent fans, wall vent fans, windows, and unit heaters for control of the environs.Xn the event of a loss of power to the fans, there would be no significant temperature rise, since it is a large volume building with suf f icient openings to adequately circulate outside air.32


RG&E's commitment to resolve the HELB environment is provided in Section II.C.Protection against f looding is described in Ref erences[FLOOD-1]and[FLOOD-5].The, limiting environment in the screenhouse is thus normal ambient conditions.
IV. EQUIPMENT QUALIFICATION INFORMATION Table 3 summarizes   the qualif ication information of required electrical equipment. This section provides the detailed background information, with emphasis on a response to the August 20, 1980     FRC   Draf t Interim Technical Evaluation Report, Project C5257.         For this reason, the paragraphs are ordered consistent with Section 3 of that report.
33 IV.EQUIPMENT QUALIF ICATION INFORMATION Table 3 summarizes the qualif ication information of required electrical equipment.
: 1. TER Paragraph   3.2.1 Table 3 Item No. 23. Main Steam-line Pressure Transmitter in the Intermediate Building.
This section provides the detailed background information, with emphasis on a response to the August 20, 1980 FRC Draf t Interim Technical Evaluation Report, Project C5257.For this reason, the paragraphs are ordered consistent with Section 3 of that report.1.TER Paragraph 3.2.1-Table 3 Item No.23.Main Steam-line Pressure Transmitter in the Intermediate Building.TER C5257 noted that this instrumentation meets the DOR Guidelines.
TER C5257   noted that   this instrumentation     meets the DOR Guidelines.       In order to provide instru-mentation with all of the proper qualification documentation, there are plans to replace these transmitters by     June 1982. Qualification docu-mentation   will be   made available when received.
In order to provide instru-mentation with all of the proper qualification documentation, there are plans to replace these transmitters by June 1982.Qualification docu-mentation will be made available when received.2.TER Paragraph 3.2.2-Table 3 Item Nos.31, 41.Medium Voltage Switchgear Located Outside Containment (Models DB-50A and DH-350E).TER C5257 found these acceptable, since the breakers are exposed only to a relatively mild (1 psig, 220'F)environment, must function within a short time (generally seconds)and fail-safe on loss of power.No additional information is'onsidered necessary to show proper operational capability under the required accident conditions.
: 2. TER   Paragraph   3.2.2   Table 3 Item Nos. 31, 41.       Medium Voltage Switchgear Located Outside Containment             ( Models DB-50A and DH-350E).
34  
TER C5257   found these acceptable,     since the breakers are exposed only to     a relatively mild   (1 psig, 220'F) environment, must function within           a   short time (generally seconds)       and fail-safe   on   loss of power. No   additional information is'onsidered necessary to show proper operational capability under the required accident conditions.
34


I 3.TER Paragraph 3.2.3-Table 3 Item No.21A.Containment Pressure Transmitters located outside containment.
I
TER C5257 found that these transmitters satisfied the DOR Guidelines.
: 3. TER Paragraph 3. 2. 3 Table 3 Item No. 21A.     Containment Pressure Transmitters located outside containment.
In light of TMI Lessons Learned, f ive of the seven transmitters, which could see a high radiation field following a LOCA, are being replaced with new transmitters (three will have a 10-200 psig span and provide post-accident monitoring).
TER C5257 found that these transmitters       satisfied the DOR Guidelines. In light of TMI Lessons Learned, f ive of the seven transmitters, which could see a high radiation field following a LOCA, are being replaced with new transmitters ( three will have a 10-200 psig span and provide post-accident monitoring). These transmitters will be qualified for the post-LOCA environment and will therefore be qualified for a HELB outside containment environment. All 5 will be   replaced by June 1982.
These transmitters will be qualified for the post-LOCA environment and will therefore be qualified for a HELB outside containment environment.
Qualification documentation will be made available when received. The two transmitters not being replaced are not exposed to     a harsh environment as the result of   a LOCA. For a   high energy line break outside containment, these two transmitters are not required to function.
All 5 will be replaced by June 1982.Qualification documentation will be made available when received.The two transmitters not being replaced are not exposed to a harsh environment as the result of a LOCA.For a high energy line break outside containment, these two transmitters are not required to function.4.TER Paragraph 3.2.4-Table 3 Item No.25 BAST Level Transmitter in the Auxiliary Building.TER C5257 found that these transmitters met the intent of the DOR Guidelines.
: 4. TER Paragraph 3.2.4 Table 3 Item No.       25 BAST Level Transmitter in the Auxiliary Building.
It is important to note that, this instrumentation performs'its safety function following a LOCA or steam line break prior to the time any accident environment is encountered in the Auxiliary Building.For a HELB in the Auxiliary Building, there is no need for the BAST level transmitters to function.No additional information is required for this equip-ment.5.TER Paragraph 3.2.5-Table 3 Item No.18.RWST Level Transmitter in the Auxiliary Building.I TER C5257 notes that this item satisfies the intent of the DOR Guidelines.
TER C5257 found that these transmitters met the intent of the DOR Guidelines. It is important to note that, this instrumentation performs'its safety function following a LOCA or steam line break prior to the time any accident environment is encountered in the Auxiliary Building. For a HELB
For f urther assurance, this transmitter will be replaced by June 1982 with a f ully-qualif ied transmitter.
 
-Qualif ication documentation will be made available when received.6.TER Paragraph 3.2.6-Table 3 Item No.19.RWST Level Switch in Auxiliary Building.TER C5257 notes that this item does not require environmental qualification, since the safety function is performed prior to the onset of an adverse environment.
in the Auxiliary Building, there is no need for the BAST level transmitters to function. No additional information is required for this equip-ment.
This is correct;for added assurance of post-accident monitoring, however, this item is being replaced by June 1982.Qualification documentation will be made available when received.7.TER Paragraph 3.3.1.1-Table 3 Item No.8A.Valve Operators for Valves MOV 841, 865.TER C5257 concludes that, since these valve actuators are locked in the"open" position with power removed with no need to f unction, lack of valid 36-qualification documentation is a moot point.Thus, no qualif ication inf ormation is required f or this item.8.TER Paragraph 3.3.1.2-Table 3 Item Nos.SF, SG.Valve Operator for MOVs 851A, B;878 B, D.TER C5257 concludes that, since these valve actuators)are locked in the"safety" position, with no need to function, environmental qualification is a moot point.Thus, no qualification information is 9.required for this item./TER Paragraph 3.3.1.3-Table 3 Item No.SC.Valve Operators for MOVs 825 A, B.As noted in TER C5257, these valves perform their safety function (open to allow RWST fluid to the suction of the SI pumps)prior to the time an adverse environment would exist in the Auxiliary Building due to sump recirculation.
: 5. TER Paragraph 3.2.5     Table 3   Item No. 18. RWST Level Transmitter in the Auxiliary Building.
No"harsh" environmental qualification is required for these items.10.TER Paragraph 3.3.1.4-Table 3 Item No.SD.Valve Operators for MOVs 4027, 4028, 4007, 4008, 4000A, 4000B.As noted in TER C5257, these valves would not be used in the.event of a HELB in the Intermediate Building.RGGE Emergency Procedures specifically call for actuating the Standby Auxiliary Feedwater 37 i'
I TER C5257   notes that     this item satisfies the intent of the DOR Guidelines. For f urther assurance, this transmitter will be replaced by June 1982 with a f ully-qualified transmitter. Qualif ication
System in the event the AFW system is inoperable.
                                                    -
Since none of the S tandby AFW system components wil l be e osed xp to a HELB, it is concluded that this system will be suff icient to provide the needed saf ety f unction.No"harsh" environmental qualification for the AFW valves ves xs needed.11.TER Para ra h 3 g p.3.1.6-Tables 3 Item No 11 o..Auxiliary Feedwater Pump Motors.As noted in TER C5257 th hese pumps are not required to function in the event of a HELB in the Xnter-mediate Building.The S e tandby AFW System performs the required safety function P d roce ures call for removing the AFW um p ps from the safety-related bus, prior to connecting the standby system.Mechanical interlocks ensure that both sets of pumps cannot be powered from th d'iesels concurrently.
documentation     will be     made available   when received.
No"harsh" environmental qualif ication for the auxiliary f eedwater pumps is required.12.TER Para ra h 3 g p.3.2.1-Table 3 Xtem No.8E.Valve operators for MOVs.850 A, BE 856'57 Ag BJ C 860 Ai Ci Documentation Reference 53 b su mitted to the NRC on September 24 1 980, provides a ref erence to Limitorque Re ort B p 0003.This reference provides assurance that these valves will perform their safet functi'on.Additional information from-38 Limitorque Report B0058 has be'en added to Reference 53, documenting Limitorque's use of generic quali-f ication to qualif y multiple size actuators by one type test.13.TER Paragraph 3.3.2.2-Table 3 Item No.8H Valve Operators for MOVs 852A, B.TER CS257 notes that these valve actuators are not acceptable for long-term service in an accident environment, and are not qualified for submerged operation.
: 6. TER Paragraph 3. 2. 6     Table 3 Item No. 19. RWST Level Switch in Auxiliary Building.
Qualification for short-term post-LOCA operation is shown in Reference 18, however.The f unction of these valves is to open upon receipt of an SI signal, and then to remain open.Quali-f ication for submerged operation is not required.Submergence could occur unless the saf ety f unction of the valves has already occurred.Specif ically, to submerge these valve operators, the entire contents of the primary system, the entire contents of both accumulators, and a portion of the water in the refueling water storage tank must be discharged to the containment.
TER C5257   notes that     this item does not require environmental qualification, since the safety function is performed prior to the onset of an adverse environment. This is correct; for added assurance   of post-accident monitoring, however, this item is being replaced by June 1982.
For this to occur, however, a safety injection signal must have occurred and the valves must have opened.RGSE has incorporated modif ications to these valve operators to prevent undesired operation in the event of submergence.
Qualification documentation will be           made available when received.
The details of these 39 modifications were provided in References
: 7. TER Paragraph   3. 3. 1. 1   Table 3 Item No. 8A. Valve Operators for Valves       MOV 841, 865.
[FLOOD-2, FLOOD-3], transmitted to FRC on May 29, 1980.It is thus considered that these valves are qualified to perform their required safety function.14.TER Paragraph 3.3.2.3-Table 3 Item No.SI.Valve Operators for MOV's 9703A,B;9704A,B;9710A,B in the SAFN System.All of these valve operators are located in the Auxiliary Building Addition, which is a"mild" environment.
TER C5257   concludes     that, since these valve actuators are locked in the "open" position with power removed with no need to function, lack of valid 36
Environmental qualif ication is provided under paragraph 4.3.3 of the"DOR Guide-lines", Areas Normal l Maintained at Room Conditions.
 
The Auxiliary Building Addition is maintained at room conditions by redundant air conditioning systems served by the onsite emergency electrical power system.The room conditions specified in Reference 43 are 60-120'F.The valve specification (Reference 54)states that"the valve actuator shall be designed for a 40 year plant life under ambient conditions of 40F to 120F..." Since there is no change in the environmental conditions between normal and accident conditions,"...no special consideration need be given'to the environ-mental qualification of Class IE equipment in these areas provided the aging requirements discussed in Section 7.0 are satisfied and the areas are maintained at room conditions by redundant air 40 conditioning or ventilation systems served by the onsite emergency electrical power system".Reference 47 describes the program developed at R.E.Ginna for detecting age-related failures.This program was developed to conform to the provisions of Section 7.0 of the"DOR Guidelines" for the"ongoing programs...to review surveillance and maintenance.
qualification documentation is a moot point.
records to assure that equipment which is exhibiting age-related degradation will be identified and replaced as necessary".
Thus, no qualif ication information is required for this item.
15.TER Paragraph 3.3.2.4-Table 3 Item No.13A.Crouse-Hinds Electrical Penetrations
: 8. TER Paragraph 3. 3. 1. 2 Table 3 Item Nos. SF, SG.
.r TER C5257 notes that the Brunswick tests could not be substantiated, since no test description was provided.Reference 45 provides this description.
Valve Operator for   MOVs 851A, B; 878 B, D.
Reference 58 is a letter from Westinghouse stating that the Brunswick data is applicable to qualify the seal, canister, and internal connections.
TER C5257 concludes   that, since these valve actuators
                                                      )
are locked in the "safety" position, with no need to function, environmental qualification is a moot point. Thus, no qualification information is required for this item.
                        /
: 9. TER Paragraph 3. 3. 1. 3 Table 3 Item No. SC. Valve Operators for MOVs   825 A, B.
As noted in TER C5257, these valves perform their safety function (open to allow RWST fluid to the suction of the SI pumps) prior to the time       an adverse environment would     exist in the Auxiliary Building due to sump recirculation. No "harsh" environmental qualification is required for these items.
: 10. TER Paragraph 3.3.1.4     Table 3 Item No. SD. Valve Operators for MOVs   4027, 4028, 4007, 4008, 4000A, 4000B.
As noted in TER C5257,   these valves would not be used in the .event of a HELB in the Intermediate Building. RGGE Emergency Procedures specifically call for actuating the Standby Auxiliary Feedwater 37
 
i' System   in the event the         AFW   system     is inoperable.
Since none of the         S tandby   AFW     system components will  be e xp osed     to   a HELB,     it is     concluded     that this system     will be suff icient to provide the needed   saf ety f unction. No "harsh" environmental qualification for the           AFW   valves ves xs needed.
: 11. TER Para g ra p h 3 .3.1.6     Tables   3   Item   Noo. 11 . Auxiliary Feedwater   Pump     Motors.
As noted in     TER C5257       thhese pumps are not required to function in the event of a HELB in the Xnter-mediate Building. Thee S tandby AFW System performs the required safety function                   P roce d ures call for removing the       AFW p um ps from the         safety-related bus, prior to connecting the               standby system.
Mechanical interlocks ensure that both sets of pumps   cannot be powered from th d'iesels concurrently.
No "harsh" environmental qualif ication for the auxiliary feedwater          pumps   is required.
: 12. TER Para g ra p h 3 .3.2.1     Table   3   Xtem No. 8E.         Valve operators for       MOVs. 850 A, BE       856   '57       Ag   BJ C   860 Ai Ci Documentation Reference 53               su b mitted to the     NRC on September       24   1 980, provides       a ref erence to Limitorque     Re p ort   B 0003. This reference provides assurance     that these valves will perform their safet     functi'on. Additional information                 from 38
 
Limitorque Report B0058 has be'en added to Reference 53, documenting Limitorque's use of generic quali-f ication to qualif y multiple size actuators by one type   test.
: 13. TER Paragraph   3.3.2.2   Table 3 Item No. 8H Valve Operators   for MOVs   852A, B.
TER CS257     notes that these valve actuators are not acceptable     for long-term service in   an accident environment, and are not qualified       for submerged operation. Qualification for short-term post-LOCA operation is shown in Reference 18, however. The function of these valves is to open upon receipt of an SI signal, and then to remain open. Quali-f ication for submerged operation is not required.
Submergence could occur unless the saf ety f unction of the valves has already occurred. Specif ically, to submerge these valve operators, the entire contents of the primary system, the entire contents of both accumulators, and a portion of the water in the refueling water storage tank must be discharged to the containment. For this to occur, however, a safety injection signal must have occurred and the valves must have opened.
RGSE has   incorporated modif ications to these valve operators to prevent undesired operation in the event of submergence.       The details of these 39
 
modifications were provided in References [FLOOD-2, FLOOD-3], transmitted to FRC on May 29, 1980.         It is thus considered that these valves are qualified to perform their required safety function.
: 14. TER Paragraph   3. 3. 2. 3 Table 3 Item No. SI. Valve Operators   for MOV's 9703A,B; 9704A,B; 9710A,B     in the SAFN System.
All of   these valve operators are located in the Auxiliary Building Addition, which is a "mild" environment. Environmental qualif ication is provided under paragraph 4.3.3 of the "DOR Guide-lines", Areas Normal l Maintained at Room Conditions.
The Auxiliary Building Addition is maintained at room conditions by redundant air conditioning systems served by the onsite emergency electrical power system.       The room conditions specified in Reference 43 are 60-120'F. The valve specification (Reference 54) states that "the valve actuator shall   be designed   for a 40 year plant life under ambient conditions of 40F to 120F..."       Since there is no change   in the environmental conditions between normal and accident conditions, "...no special consideration need be given 'to the environ-mental qualification of Class IE equipment in these areas provided the aging requirements discussed in Section 7.0 are satisfied and the areas are maintained at room conditions by redundant       air 40
 
conditioning or ventilation systems served by the onsite emergency electrical power system". Reference 47 describes the program developed at R. E. Ginna for detecting age-related failures. This program was developed to conform to the provisions of Section 7.0 of the "DOR Guidelines" for the "ongoing programs...to review surveillance and maintenance.
records to assure that equipment which is exhibiting age-related degradation     will be identified and replaced as necessary".
: 15. TER Paragraph   3.3.2.4     Table 3   Item No. 13A. Crouse-Hinds Electrical   Penetrations r
                                      .
TER C5257 notes that the Brunswick tests could not be substantiated, since no test description was provided. Reference 45 provides this description.
Reference 58   is a letter   from Westinghouse stating that the Brunswick data is applicable to qualify the seal, canister, and internal connections.
Reference 54 is an evaluation of the capability of the Ginna penetrations to perform their function under elevated and short-circuit electrical loading conditions.
Reference 54 is an evaluation of the capability of the Ginna penetrations to perform their function under elevated and short-circuit electrical loading conditions.
Further, an evaluation (Reference 59)of the functions of the various materials in the penetra-tions disclosed that the organic compounds, which are possibly subject to aging or radiation effects, 41 do not perform any critical insulating or sealing functions.
Further, an evaluation (Reference 59) of the functions of the various materials in the penetra-tions disclosed that the organic compounds, which are possibly subject to aging or radiation effects, 41
These functions are performed by ceramic and metallic components..
 
This evaluation augments the qualification testing performed on these penetrations, confirming that they are N qualified to perform their safety function.16.TER Paragraph 3.3.2.5-Table 3 Item No.13B.Westinghouse Electrical Penetrations
do not perform any critical insulating or sealing functions. These functions are performed by ceramic and metallic components.. This evaluation augments   the qualification testing performed on these penetrations, confirming that they are N
.It is noted in TER C5257 that additional inf ormation concerning the"similar resin", aging characteristics of the insulation on the cable leads, and qual if ied lif e should be provided.Ref erence 61, Research II Report 75-7BS-BIGAL-122, shows that the lower 95%conf idence band on qual if ied lif e at 105'C is greater than 40 years.Also, the author of this report, Mr.J.F.Quirk, has stated that the word"similar" had been used only in the respect that test results of this epoxy were close to the results of other epoxies also being tested.The, epoxy in the Ginna penetrations is identical to that tested.Cable lead insulation aging data is also included in Reference 61.It can be concluded that these penetrations are suitable to perform their required safety functions.
qualified to perform their safety function.
42  
: 16. TER Paragraph   3.3.2.5     Table 3 Item No. 13B.     Westinghouse Electrical   Penetrations   .
It is   noted in     TER C5257 that additional information concerning the "similar resin", aging characteristics of the insulation on the cable leads, and qual ified life  should be provided.
II Ref erence   61, Research Report 75-7BS-BIGAL-122, shows that the lower           95%
conf idence band on qual     ified life  at 105 'C is greater than 40 years. Also, the author of this report, Mr. J. F. Quirk, has stated that the word "similar" had been used only in the respect that test results of this epoxy were close to the results of other epoxies also being tested. The, epoxy in the Ginna penetrations is identical to that tested. Cable lead insulation aging data is also included in Reference 61.
It can be concluded     that these penetrations are suitable to perform their required safety functions.
42
: 17. TER Paragraph  3. 3. 2. 6 - Table  3 Item No. 14. Westinghouse Terminal Blocks Inside Containm'ent.
TER C5257  found that, although      qualification for pressure,  temperature,    and  humidity is acceptable, additional information is needed concerning thermal aging and radiation. Reference 60 is a Proprietary Westinghouse    R&D Report  ( 077-7B7-CBSEL-R3) dated July 13, 1977. It shows that for a criteria of f ailure of 50% of the original flexure strength and impact strength, the 40 year life extrapola-tion is approximately 120'C. This report, is not yet in our possession, but may be audited at the Westinghouse    facility.
Additional information -concerning radiation sus-ceptibility of the terminal blocks is also provided in Reference 60. It is shown that the qualification level is 2 x 10 7 rads. Although not meeting the long-term conservatively calculated radiation dose for Ginna of 1. 6 x 10 8 rads, the DOR Guideline values are met. Based on the protected location
                        '
of these terminal blocks, 2 x 10 7 rads is considered adequate. A detailed evaluation of this post-LOCA radiation dose will be'ade. If the required dose for the long-term monitoring function is greater, replacement or additional protection will be provided.
43
 
As  presently installed, the terminal blocks for pressurizer pressure and level instrumentation would become submerged      after  a LOCA            en  qualified long-term monitoring instrumentation for these functions is installed at Gin    irma, and elevated above the submergence level, the terminal blocks will also  be  el evated. Submergence        and  direct spray impingement    will thus be precluded.              See paragraphs  19 and  20 for a discussion of            the pressurizer pressure      and  level instrumentation.
: 18. TER Paragraph  3.3.2.7  Table 3 Item Nos.              15A, B,  C Kerite Cable Inside Containment.
Reference 51    is the  "Cable Id      t'f'n i z.cation  and Qualification Supplement"          Th'is      ocument can be used to determine the      identity of cable in use throughout the plant. It is shown that all power cable inside containment is Kerite. The most recent and comprehensive qualification testing of Kerite cable was performed in conjunction with the testing of Raychem sleeves (Reference 38). Reference 55 is a lett etter from Kerite verifying that the cable supplied'or the qualification testing in Reference 38 is identical to th a t orig>nally supplied and installed in the Ginna                co t irma containment.
The pre-aging done for the Kerite cable during the Raychem sleeve test establish e d a 93 . 3 year life 44
 
at 140'F  mean  surface temperature. The  Arrhenius data  is conf idential to the    manuf acturer,  but is available at    RG&E  as Reference  63.
RG&E  believes that this recent testing definitively demonstrates the adequacy of the Kerite cable for performing its required safety function.
There are no    safety-related cables inside containment subject to flooding, which are required to perform a safety function during submergence.        Qualification for submergence is thus not required.
: 19. TER  Paragraph  3.3.2.8    Table 3  Item No. 22. Pressurizer Pressure Transmitters.
The  deficiencies noted in TER C5257 included lack of radiation and submergence qualification. RG&E does  not claim credit for the use of this instru-mentation at the time    it would receive excessive radiation exposure, or become submerged. Ginna Emergency Procedures specify that, unless pressurizer pressure, level, and other parameters appear stable and are returning to prescribed levels, safety injection flow is not to be terminated.
Failure to terminate safety injection is not a safety concern. Therefore, lack of qualification for this instrumentation is not considered of immediate safety significance.
45
 
It is  recognized, however, that accurate primary system information would be extremely useful to the operator    for  diagnosing the status of the plant during accident conditions. RG6E, therefore, plans to replace the present instrumentation by June 1982 with f ully-qualified transmitters, located above any possible submergence level.
Qualification documentation will be made available when  received.
: 20. TER Paragraph  3. 3. 2. 9  Table 3  Item No. 24. Pressurizer Level Instrumentation.
The same  information    as  described in 19 above for the pressurizer pressure instrumentation applies to this instrumentation.
: 21. TER Paragraph  3.3.2.10  Table 3 Item No. 30.      Fan Cooler Motors Inside Containment.
TER C5257  concluded that in addition to the information provided in References 18 and 2 0, information needed for complete qualification of the fan cooler motors is a) documentation regarding qualification of motor-lead and lead-to-cable splices, and (b) determination of a qualified life for the motor. Information regarding the splices is given in    Reference    64.
46
 
Aging information      for the insulating material of these motors, as well as the bearing lubricants, is given in    Reference 18, Section 4.      Aging to demonstrate    40  year continuous operation at 120'C was  performed. This is consistent with the data given in Reference 67, and is considered sufficient to qualify the fan cooler motors for continued operation. A program at RG6E to maintain motor bearings and lubricants is given in Reference 65.
This program will ensure that the lubricants used are compatible with the environmental conditions which could occur during a DBE.
Additional information regarding qualification testing of the same type of motors is given in WCAP 7829, "Fan Cooler Motor Unit Test" (Reference
: 70) .
: 22. TER Paragraph    3. 3. 2.11  Table  3  Item No. 34. Raychem Cable Splice Sleeves.
TER C5257    states that    RG&E  should present evidence of  similarity    between the tested and    installed equipment.      This is'documented      in the detailed evaluation and observation of the splice sleeve replacement program, given in IE Inspection Reports 78-20 and 78-21 (Reference 56).
It is  also stated that a determination of qualified life  should be made for the sleeves.        The actual 47
 
test in Reference 38 established      a 12.1 year    life at 60'C ambient. This pre-aging        was  constrained by the concurrent aging of the Kerite cable, which was pre-aged for 93.3 years at 60'C by the same test. Based on  proprietary  Raychem  information (included in Reference 63 and available for audit at RG6E) a 40 year life at 91'C can be expected..
Therefore, these sleeves are considered        fully qualified.
: 23. TER Paragraph  3.3.2.12    Table 3 Xtem No. 20. Steam Flow Transmitters    Enside Containment.
RG&E  has stated    that these transmitters are not required to perform a safety function at a time they could be exposed to a high energy line break environment. Thus, the lack of complete qualification documentation is a moot point for these trans-mitters. For a steam line break inside containment, the steam line non-return check valves will assure that the intact steam generator will not blow down. Steam line isolation would be provided by the high containment pressure signal.
For added assurance    of steam line isolation in the event of  a  steam break'inside containment, these transmitters    will be  replaced by June 1982 with fully-qualified equipment.      Qualification documenta-tion will be    made  available  when  received.
48
: 24. TER Paragraph  3.3.2.13 - Table 3 Item No. 21B. Contain-ment Pressure Transmitters in the Intermediate Building.
As noted in Section IV.3 of this report, five of the seven containment pressure      transmitters, which could be exposed    to high post-LOCA radiation levels, are being replaced with LOCA-qualified units by June 1982, in response to TMI Lessons Learned. Qualif ication documentation will be made available when received.
: 25. TER Paragraph  3.3.2.14    Table 3, Item No. 37, Hydrogen Recombiner  Igniter Exciter TER C5257  requested  that the effects of containment spray and thermal aging be addressed.        This informa-tion has not yet been received. If proper documen-tation is not found concerning these environmental parameters,  RG&E  will commit to    replace the necessary equipment. It is  important to note that the present licensing basis for Ginna does not include the hydrogen recombiner as a means necessary for I
post-LOCA hydrogen control (see the RG&E "Technical Supplement Accompanying    Application for  a Full Term  Operating License," August 1972, Section III.B.7).
: 26. TER Paragraph  3.3.2.15    Table 3, Item No. 38, Hydrogen Recombiner Blower Motor.
49
 
The  only deficiency noted in TER C5257 is that no analysis exists comparing the impact of deviations between the test specimen specific design features, materials, and production procedure and those of the installed equipment.      The only evidence at this time is contained in Section 5.2 of Reference 18, WCAP 7410-L, Vol. II. It is stated that "the 2 hp motor used in the test program is constructed in the same manner as, is the actual 15 hp motor used in the recombiner." Further,      it  has been verified that the Ginna 15 hp    motor has Class  H insulation, the same as the 2    hp motor  tested.
Based on the  available information,  RG6E  believes that there is reasonable assurance that the Ginna recombiner motor will perform its safety function.
Further, as stated in 25 above, the hydrogen recombiner is not required by the present Ginna design basis. Based on the TMI Lessons Learned, however, RGEE will commit to replace the motor if proper environmental qualification documentation is not established.
: 27. TER Paragraph  3.3.3.1  Table  3  Item No. 8B. Valve Operators  for MOVs  826 A,B,C,D; 896 A,B.
The MOVs 826 A,B,C,D    are located at the discharge of the Boric Acid Storage Tanks, and provide suction to the SI pumps in the event of a Safety 50
 
Injection signal. Upon low BAST level, these valves close af ter the 825 A,B valves open. The valves are located in the auxiliary building, and will have completed their function prior to the presence of an adverse environment caused by        sump recirculation fluid.
MOVs  896 A,B are  normally locked-open valves, located at the suction of the SI and CS pumps from the .EST. The valves are closed prior to the time sump recirculation is initiated.        Therefore, these valves  will have  completed  their function orior to the time  an adverse  environment would occur.
In the case of    all six  valves, environmental qualification for    an adverse    environment  is not required.
: 28. TER Paragraph  3.3.3.2  Table    3 Item Nos. 1A, 1B, 1C,
: 5. ASCO  solenoid valves.
The  feedwater control and bypass valves ( items 1A, 1B) fail closed on loss of air. This is supported by Reference 23. In order to further ensure that these valves  will perform their safety function when exposed  to  a HELB  in the Turbine Building, the solenoids  will be  replaced with valves having proper qualification documentation. It is exoected that this  can be accomplished by June 1982.        The fail-safe closure of    the valves ensures    that the 51
 
required safety function can    be performed  until replacement can be effected.
Item 1C, the solenoid control ling LCV112B,      will not experience  an adverse  environment during an accident. Further, an accessible manual bypass valve, valve 358, is used to provide alternative suction for the charging pumps from the RWST.
Since this function would not be required for many hours following an event requiring the maintenance of  a safe shutdown condition, the use of this manual valve  is considered acceptable. Item  1C will thus  be  deleted from Table 3.
Item 5A, the  RHR  discharge valves, are normally open. They need only remain open in the event of an accident. The I/P controller ( rather than a solenoid valve) controlling their position is fail-open. Since no function must be performed by these valves, they have been deleted from Table 3.
Item 5B, the solenoid valves    for  AOVs 897  and 898, are required to close    prior to  sump recirculation.
They  will not  experience an adverse environment prior to the time they    must perform  their safety function. Environmental qualification of these valves will be addressed in a later submittal, concerning electrical equipment located in a "mild" environment.
52
: 29. TER Paragraph    3.3.3.3      Table  3 Item No. 2. Copes-Vulcan Solenoid Valves.
The  valves were purchased from      ASCO  (Series 8300).
Therefore,    all information    from Reference 23 applies to the valves.        Further, since these valves are located in      a  "mild" environment, qualification of these valves will be discussed at a later time.
: 30. TER Paragraph 3.3.3.4 - Table 3 Item Nos. 3A, 3B.
Lawrence Solenoid Valves in Intermediate Building.
Based on the design      principle of these valves, they will perform their safety function by failing in a closed position upon loss of power. However, if power qualification documentation is not established,
        .RGaE will initiate a replacement for these solenoid valves. Qualification documentation will be made available when received. The fail-safe mode of operation ensures no loss of safety function in the interim.
: 31. TER Paragraph    3. 3. 3. 5  Table  3 Item No. 4. Versa Solenoid Valves inside containment.
The  safety function of the solenoid valves controlling the containment      air recirculation dampers is accomplished    through fail-safe operation. This is accomplished immediately with the SI signal following an  accident, before environmental conditions would 53
 
become  very severe.      In ordei to have this safety function accomplished with equipment having the proper qualif ication testing and documentation, replacement of these solenoid valves will be initiated. It is  expected that    this can be accomplished by June 1982.        Qualification docu-mentation  will be    made  available  when received.
: 32. TER  Paragraph  3. 3. 3. 6  Table  3 Item Nos. 6A, 6B.
Versa Solenoid Valves.
The  safety function of these containment purge and depressurization valves immediately following an accident is to close for containment isolation.
This is accomplished by the        fail-close design of these valves.      In order to have this safety function I
accomplished with equipment having the proper qualification testing      and documentation,    replace-ment of these solenoid valves        will be initiated.
It is  expected that      this can be accomplished by June 1982. Qualification documentation will be made  available    when  received.
: 33. TER  Paragraph  3.3.3.7      Table  3 Item No. 7. Control Room  Dampers.
This equipment item is not        electrical,  and there-fore is not addressed in this report.          The solenoid valves operating these dampers are addressed under paragraph TER 3.3.3.24 (Table 3, Item No. 40).
54
: 34. TER Paragraph    3. 3. 3. 8 - Table    3  Item No. 9. Standby'FN Pump  Motors .
Although  this item is not located in          a harsh environment, and therefore does not need to be addressed  at this time,      RGSE    considers the environ-mental  qualification of this item to          be complete and  acceptable.      As stated in Section 4.3.3 of the DOR  Guidelines, "No special consideration need to the environmental qualification of Class be'iven IE equipment in these [non-harsh] areas provided the aging requirements discussed in Section 7.0 are  satisfied    and the areas        are maintained at room conditions by redundant air conditioning or ventila-tion systems served by the onsite emergency electrical power system." This is the case with these motors.
The equipment    specification for these motors (Reference 3) states "Motors shall be rated for operation in an ambient tern erature of 50'C [122'F] ".
(
Tnis is consistent with the ambient operating conditions for the Auxiliary Building Addition of 60-120'F (Ref erence 43)    .      Furthermore, the ongoing
        .program described      in Reference      47  to detect age-related f ailures  includes these motors.            RG&E theref ore considers these motors to have met            all necessary environmental requirements .
P.
55
: 35. TER Paragraph    3.3.3.9 - Table      3  Item Nos. 10A, 10B, 10C, 12A. Motors  for the    Containment Spray Pumps, Component Cooling Water Pumps,          Residual Heat Removal Pumps,  and  Safety Injection      Pumps.
The  first  three of these Ginna motors have Class B insulation made of "Thermalastic Epoxy". The SI pump motor insulation is "PMR" (Premimum Moisture Resistant).      This is shown in Reference 67.
Qualf ication of these systems is given in          WCAP 8754,  ( Ref erence  68 ),  for  the "Thermalas  tie  Epoxy" motors, and the Westinghouse Research Report 71-1C2-RADMC-R1, "The Ef f ect of Radiation on Insulating Materials        Used  in Westinghouse    Medium Motors," December 31, 1970 (Revised April 10, 1971) (Reference 69) for the "PMR" motors. These reports are proprietary, but are available for audit at RGEE and at Westinghouse. Testing does indicate that these motors can withstand an accumulated dose of 10 7 rads during their operating lif e, with an operating      lif e of 20 years. Since these motors are not used at all times (only the CCW pump is used during normal operation,            and even then only one of the two pumps is normal ly in use), the operational capability is at least 40 years. Also, RG&E has a program of insulation inspection once per year (M45.1A, Inspection of Saf eg uard Motor) and replacement        ( if needed)  every five years.
56


17.TER Paragraph 3.3.2.6-Table 3 Item No.14.Westinghouse Terminal Blocks Inside Containm'ent.
r l
TER C5257 found that, although qualification for pressure, temperature, and humidity is acceptable, additional information is needed concerning thermal aging and radiation.
l
Reference 60 is a Proprietary Westinghouse R&D Report (077-7B7-CBSEL-R3) dated July 13, 1977.It shows that for a criteria of f ailure of 50%of the original flexure strength and impact strength, the 40 year life extrapola-tion is approximately 120'C.This report, is not yet in our possession, but may be audited at the Westinghouse facility.Additional information-concerning radiation sus-ceptibility of the terminal blocks is also provided in Reference 60.It is shown that the qualification level is 2 x 10 rads.Although not meeting the 7 long-term conservatively calculated radiation dose f or Ginna of 1.6 x 10 rads, the DOR Guideline 8 values are met.Based on the protected location'7 of these terminal blocks, 2 x 10 rads is considered adequate.A detailed evaluation of this post-LOCA radiation dose will be'ade.If the required dose for the long-term monitoring function is greater, replacement or additional protection will be provided.43 As presently installed, the terminal blocks for pressurizer pressure and level instrumentation would become submerged after a LOCA en qualified long-term monitoring instrumentation for these functions is installed at Gin irma, and elevated above the submergence level, the terminal blocks will also be el evated.Submergence and direct spray impingement will thus be precluded.
See paragraphs 19 and 20 for a discussion of the pressurizer pressure and level instrumentation.
18.TER Paragraph 3.3.2.7-Table 3 Item Nos.15A, B, C Kerite Cable Inside Containment.
Reference 51 is the"Cable Id t'f'n i z.cation and Qualification Supplement" Th'is ocument can be used to determine the identity of cable in use throughout the plant.It is shown that all power cable inside containment is Kerite.The most recent and comprehensive qualification testing of Kerite cable was was performed in conjunction with the testing of Raychem sleeves (Reference 38).Reference 55 is a lett etter from Kerite verifying that the cable supplied'or the qualification testing in Reference 38 is identical to th t a orig>nally supplied and installed in the Ginna co t irma containment.
The pre-aging done for the Kerite cable during the Raychem sleeve test establish d 93 3 e a.year life-44 at 140'F mean surface temperature.
The Arrhenius data is conf idential to the manuf acturer, but is available at RG&E as Reference 63.RG&E believes that this recent testing definitively demonstrates the adequacy of the Kerite cable for performing its required safety function.There are no safety-related cables inside containment subject to flooding, which are required to perform a safety function during submergence.
Qualification for submergence is thus not required.19.TER Paragraph 3.3.2.8-Table 3 Item No.22.Pressurizer Pressure Transmitters.
The deficiencies noted in TER C5257 included lack of radiation and submergence qualification.
RG&E does not claim credit for the use of this instru-mentation at the time it would receive excessive radiation exposure, or become submerged.
Ginna Emergency Procedures specify that, unless pressurizer pressure, level, and other parameters appear stable and are returning to prescribed levels, safety injection flow is not to be terminated.
Failure to terminate safety injection is not a safety concern.Therefore, lack of qualification for this instrumentation is not considered of immediate safety significance.
45 It is recognized, however, that accurate primary system information would be extremely useful to the operator for diagnosing the status of the plant during accident conditions.
RG6E, therefore, plans to replace the present instrumentation by June 1982 with f ully-qualif ied transmitters, located above any possible submergence level.Qualification documentation will be made available when received.20.TER Paragraph 3.3.2.9-Table 3 Item No.24.Pressurizer Level Instrumentation.
The same information as described in 19 above for the pressurizer pressure instrumentation applies to this instrumentation.
21.TER Paragraph 3.3.2.10-Table 3 Item No.30.Fan Cooler Motors Inside Containment.
TER C5257 concluded that in addition to the information provided in References 18 and 2 0, information needed for complete qualification of the fan cooler motors is a)documentation regarding qualification of motor-lead and lead-to-cable splices, and (b)determination of a qualified life for the motor.Information regarding the splices is given in Reference 64.46-Aging information for the insulating material of these motors, as well as the bearing lubricants, is given in Reference 18, Section 4.Aging to demonstrate 40 year continuous operation at 120'C was performed.
This is consistent with the data given in Reference 67, and is considered sufficient to qualify the fan cooler motors for continued operation.
A program at RG6E to maintain motor bearings and lubricants is given in Reference 65.This program will ensure that the lubricants used are compatible with the environmental conditions which could occur during a DBE.Additional information regarding qualification testing of the same type of motors is given in WCAP 7829,"Fan Cooler Motor Unit Test" (Reference 70).22.TER Paragraph 3.3.2.11-Table 3 Item No.34.Raychem Cable Splice Sleeves.TER C5257 states that RG&E should present evidence of similarity between the tested and installed equipment.
This is'documented in the detailed evaluation and observation of the splice sleeve replacement program, given in IE Inspection Reports 78-20 and 78-21 (Reference 56).It is also stated that a determination of qualified life should be made for the sleeves.The actual 47 test in Reference 38 established a 12.1 year life at 60'C ambient.This pre-aging was constrained by the concurrent aging of the Kerite cable, which was pre-aged for 93.3 years at 60'C by the same test.Based on proprietary Raychem information (included in Reference 63 and available for audit at RG6E)a 40 year life at 91'C can be expected..
Therefore, these sleeves are considered fully qualified.
23.TER Paragraph 3.3.2.12-Table 3 Xtem No.20.Steam Flow Transmitters Enside Containment.
RG&E has stated that these transmitters are not required to perform a safety function at a time they could be exposed to a high energy line break environment.
Thus, the lack of complete qualification documentation is a moot point for these trans-mitters.For a steam line break inside containment, the steam line non-return check valves will assure that the intact steam generator will not blow down.Steam line isolation would be provided by the high containment pressure signal.For added assurance of steam line isolation in the event of a steam break'inside containment, these transmitters will be replaced by June 1982 with fully-qualified equipment.
Qualification documenta-tion will be made available when received.48


24.TER Paragraph 3.3.2.13-Table 3 Item No.21B.Contain-ment Pressure Transmitters in the Intermediate Building.As noted in Section IV.3 of this report, five of the seven containment pressure transmitters, which could be exposed to high post-LOCA radiation levels, are being replaced with LOCA-qualified units by June 1982, in response to TMI Lessons Learned.Qualif ication documentation will be made available when received.25.TER Paragraph 3.3.2.14-Table 3, Item No.37, Hydrogen Recombiner Igniter Exciter TER C5257 requested that the effects of containment spray and thermal aging be addressed.
Since the only adverse environm'ent anticipated          for any of these motors      is  a post-LOCA   radiation   dose
This informa-tion has not yet been received.If proper documen-tation is not found concerning these environmental parameters, RG&E will commit to replace the necessary equipment.
( conservatively estimated in Reference [TMI-3] as I
It is important to note that the present licensing basis for Ginna does not include the hydrogen recombiner as a means necessary for I post-LOCA hydrogen control (see the RG&E"Technical Supplement Accompanying Application for a Full Term Operating License," August 1972, Section III.B.7).26.TER Paragraph 3.3.2.15-Table 3, Item No.38, Hydrogen Recombiner Blower Motor.49
: 2. 8 x 10 6 rads) these motors are considered properly qualified both for "life" and radiation.
3 6. TER Paragraph     3.3.3.10    Table 3   Item No. 12B.      Service Water    Pump  Motor.
As  stated in Reference [Flood-15], the effects of jet  impingement and water spray on these motors were evaluated by the      NRC during the review of      SEP Topic  III-5.B,  "Pipe Break Outside Containment".
RGEE  committed to supplement the      NRC  recommenda-tion in Reference      [FLOOD-13.]. Thus, the Service Water  Pump  Motors have been removed from the        HELB environment considerations.         Further review for operation is a "mild" environment        will be  conducted at a later time.
: 37. TER Paragraph     3.3.3.11    Table 3 Item No. 16. Coleman Cable Inside Containment.
Reference 51    is the  "Cable  Identification    and Qualification Supplement".        This reference allows traceability of all cable        used  in the  Ginna  plant, by referencing back to the        original purchase order specifications. It can be        seen that, in addition to the Kerite safeguards cable, the only other cable inside containment used to perform a required 57


The only deficiency noted in TER C5257 is that no analysis exists comparing the impact of deviations between the test specimen specific design features, materials, and production procedure and those of the installed equipment.
post-accident safety f unction is the silicone-rubber insulated cable, which is used for all required safety-related instrumentation and control cable.
The only evidence at this time is contained in Section 5.2 of Reference 18, WCAP 7410-L, Vol.II.It is stated that"the 2 hp motor used in the test program is constructed in the same manner as, is the actual 15 hp motor used in the recombiner." Further, it has been verified that the Ginna 15 hp motor has Class H insulation, the same as the 2 hp motor tested.Based on the available information, RG6E believes that there is reasonable assurance that the Ginna recombiner motor will perform its safety function.Further, as stated in 25 above, the hydrogen recombiner is not required by the present Ginna design basis.Based on the TMI Lessons Learned, however, RGEE will commit to replace the motor if proper environmental qualification documentation is not established.
Reference 46 identifies this as Coleman cable. In addition to the testing stated in Reference 46,       a section of this cable was taken from the Ginna plant, and environmentally qualif ied with the Raychem splice sleeves (documentation of the testing is given in FRC Final Report Supplement, F-C5074  (Supplement),  April 1979, which is included in Reference 51). The cable is specimen number C5074-7 of Table 1 of F-C5074 Supplement.
27.TER Paragraph 3.3.3.1-Table 3 Item No.8B.Valve Operators for MOVs 826 A,B,C,D;896 A,B.The MOVs 826 A,B,C,D are located at the discharge of the Boric Acid Storage Tanks, and provide suction to the SI pumps in the event of a Safety 50
This testing shows that the Coleman silicone-rubber insulated cable will perform     its required safety functions inside containment.
Reference 46 states    that this cable is aged at 200'C for 168 hours. Although no specific Arrhenius plot is available, the application of the "10'C rule" shows an operating life of 40 years at 60'C.
This is considered a reasonable      estimate of the exoected  life of this  cable.
: 38. TER Paragraph   3.3.3.12    Table 3 Items 17A, 17B, 17C.
Coleman, Rome, and General Cables Used Outside Containment.
Reference 51  is the "Cable Identification and Qualification Supplement". From this reference, the type of cable used throughout the Ginna plant 58


Injection signal.Upon low BAST level, these valves close af ter the 825 A,B valves open.The valves are located in the auxiliary building, and will have completed their function prior to the presence of an adverse environment caused by sump recirculation fluid.MOVs 896 A,B are normally locked-open valves, located at the suction of the SI and CS pumps from the.EST.The valves are closed prior to the time sump recirculation is initiated.
can be traced by reference back        to the original purchase order    specification. It is shown that all of the safety-related cable outside containment which is not Kerite cable is PVC-insulated cable.
Therefore, these valves will have completed their function orior to the time an adverse environment would occur.In the case of all six valves, environmental qualification for an adverse environment is not required.28.TER Paragraph 3.3.3.2-Table 3 Item Nos.1A, 1B, 1C, 5.ASCO solenoid valves.The feedwater control and bypass valves (items 1A, 1B)fail closed on loss of air.This is supported by Reference 23.In order to further ensure that these valves will perform their safety function when exposed to a HELB in the Turbine Building, the solenoids will be replaced with valves having proper qualification documentation.
The specif ications included in Reference 51 refer to GAI Specs SP-5324 and SP-5315. Both of these specifications in turn specify the requirements of IPCEA S-61-402 for PVC-Cable.         Information f rom this standard is provided in Reference 10. Additional information for Coleman and Rome cable is provided in Ref erence 4 6.
It is exoected that this can be accomplished by June 1982.The fail-safe closure of the valves ensures that the 51 required safety function can be performed until replacement can be effected.Item 1C, the solenoid control ling LCV112B, wil l not experience an adverse environment during an accident.Further, an accessible manual bypass valve, valve 358, is used to provide alternative suction for the charging pumps from the RWST.Since this function would not be required for many hours following an event requiring the maintenance of a safe shutdown condition, the use of this manual valve is considered acceptable.
The IPCEA  testing of this cable, including insula-tion aging at 121'C (250'F) for 168 hours ( jacket at 212'F), oil immersion, heat shock, and cold shock, shows the     ability  to operate under conditions more severe than those anticipated outside containment.
Item 1C will thus be deleted from Table 3.Item 5A, the RHR discharge valves, are normally open.They need only remain open in the event of an accident.The I/P controller (rather than a solenoid valve)controlling their position is fail-open.
Although no specif ic qualif ication testing was performed, the standard testing of these cable types gives reasonable      assurance  that they are suitable for outside-containment use.
Since no function must be performed by these valves, they have been deleted from Table 3.Item 5B, the solenoid valves for AOVs 897 and 898, are required to close prior to sump recirculation.
: 39. TER Paragraph   3.3.3.13    Table 3 Item No. 27. RTDs Inside Containment.
They will not experience an adverse environment prior to the time they must perform their safety function.Environmental qualification of these valves will be addressed in a later submittal, concerning electrical equipment located in a"mild" environment.
Reference 35    is  a specification sheet    and drawing of the Ginna    RTD  (Rosemount 176JA model).
52 29.TER Paragraph 3.3.3.3-Table 3 Item No.2.Copes-Vulcan Solenoid Valves.The valves were purchased from ASCO (Series 8300).Therefore, all information from Reference 23 applies to the valves.Further, since these valves are located in a"mild" environment, qualification of these valves will be discussed at a later time.30.TER Paragraph 3.3.3.4-Table 3 Item Nos.3A, 3B.Lawrence Solenoid Valves in Intermediate Building.Based on the design principle of these valves, they will perform their safety function by failing in a closed position upon loss of power.However, if power qualification documentation is not established,.RGaE will initiate a replacement for these solenoid valves.Qualification documentation will be made available when received.The fail-safe mode of operation ensures no loss of safety function in the interim.31.TER Paragraph 3.3.3.5-Table 3 Item No.4.Versa Solenoid Valves inside containment.
The  reactor coolant system temperature detectors (RTD) are not required for a loss of coolant 59
The safety function of the solenoid valves controlling the containment air recirculation dampers is accomplished through fail-safe operation.
This is accomplished immediately with the SI signal following an accident, before environmental conditions would 53


become very severe.In ordei to have this safety function accomplished with equipment having the proper qualif ication testing and documentation, replacement of these solenoid valves will be initiated.
accident. In a steam line break accident, low Tave plus high steam flow plus" a safety injection signal will close the main steam line isolation valves. Also, high-high steam flow     will perform this function. As described in Section II.B above, for a break upstream of the non-return check valves, which includes     all breaks inside containment, closure of the main steam isolation valves is not required.
It is expected that this can be accomplished by June 1982.Qualification docu-mentation will be made available when received.32.TER Paragraph 3.3.3.6-Table 3 Item Nos.6A, 6B.Versa Solenoid Valves.The safety function of these containment purge and depressurization valves immediately following an accident is to close for containment isolation.
For breaks downstxeam of the check valves, closure of the main steam isolation valves is desirable, however, in   this case the RTDs are not subjected to an adverse environment. Theref ore, the RTDs do not require environmental qualification to px'ovide their required safety function. However, the RTDs would be useful for post-accident monitoring.
This is accomplished by the fail-close design of these valves.In order to have this safety function I accomplished with equipment having the proper qualification testing and documentation, replace-ment of these solenoid valves will be initiated.
Since the RTDs are not qualified for post-accident use, the pxesent Ginna Emergency Procedux'es specify that, if a 50'F subcooling margin cannot be established or maintained, safety injection flow shall not be terminated. Failure of the RTDs would require that SI flow be maintained. Since the Ginna high head safety injection pumps do not have a high enough shutoff head to open the pressurizer PORVs, continued SI pump operation is not a safety concern.
It is expected that this can be accomplished by June 1982.Qualification documentation will be made available when received.33.TER Paragraph 3.3.3.7-Table 3 Item No.7.Control Room Dampers.This equipment item is not electrical, and there-fore is not addressed in this report.The solenoid valves operating these dampers are addressed under paragraph TER 3.3.3.24 (Table 3, Item No.40).54 34.TER Paragraph 3.3.3.8-Table 3 Item No.9.Standby'FN Pump Motors.Although this item is not located in a harsh environment, and therefore does not need to be addressed at this time, RGSE considers the environ-mental qualification of this item to be complete and acceptable.
However, to avoid the possibility of operator confusion, RG&E will initiate a program to provide 60
As stated in Section 4.3.3 of the DOR Guidelines,"No special consideration need be'iven to the environmental qualification of Class IE equipment in these[non-harsh]
areas provided the aging requirements discussed in Section 7.0 are satisfied and the areas are maintained at room conditions by redundant air conditioning or ventila-tion systems served by the onsite emergency electrical power system." This is the case with these motors.The equipment specification for these motors (Reference 3)states"Motors shall be rated for operation in an ambient tern erature of 50'C[122'F]".(Tnis is consistent with the ambient operating conditions f or the Auxiliary Building Addition of 60-120'F (Ref erence 43).Furthermore, the ongoing.program described in Reference 47 to detect age-related f ailures includes these motors.RG&E theref ore considers these motors to have met all necessary environmental requirements
.P.55 35.TER Paragraph 3.3.3.9-Table 3 Item Nos.10A, 10B, 10C, 12A.Motors for the Containment Spray Pumps, Component Cooling Water Pumps, Residual Heat Removal Pumps, and Safety Injection Pumps.The first three of these Ginna motors have Class B insulation made of"Thermalastic Epoxy".The SI pump motor insulation is"PMR" (Premimum Moisture Resistant).
This is shown in Reference 67.Qualf ication of these systems is given in WCAP 8754, (Ref erence 68), f or the"Thermalas tie Epoxy" motors, and the Westinghouse Research Report 71-1C2-RADMC-R1,"The Ef f ect of Radiation on Insulating Materials Used in Westinghouse Medium Motors," December 31, 1970 (Revised April 10, 1971)(Reference 69)for the"PMR" motors.These reports are proprietary, but are available for audit at RGEE and at Westinghouse.
Testing does indicate that these motors can withstand an accumulated dose of 10 rads during their operating 7 lif e, with an operating lif e of 20 years.Since these motors are not used at all times (only the CCW pump is used during normal operation, and even then only one of the two pumps is normal ly in use), the operational capability is at least 40 years.Also, RG&E has a program of insulation inspection once per year (M45.1A, Inspection of Saf eg uard Motor)and replacement (if needed)every five years.56-r l l Since the only adverse environm'ent anticipated for any of these motors is a post-LOCA radiation dose (conservatively estimated in Reference[TMI-3]as I 6 2.8 x 10 rads)these motors are considered properly qualified both for"life" and radiation.
3 6.TER Paragraph 3.3.3.10-Table 3 Item No.12B.Service Water Pump Motor.As stated in Reference[Flood-15], the effects of jet impingement and water spray on these motors were evaluated by the NRC during the review of SEP Topic III-5.B,"Pipe Break Outside Containment".
RGEE committed to supplement the NRC recommenda-tion in Reference[FLOOD-13.].
Thus, the Service Water Pump Motors have been removed from the HELB environment considerations.
Further review for operation is a"mild" environment will be conducted at a later time.37.TER Paragraph 3.3.3.11-Table 3 Item No.16.Coleman Cable Inside Containment.
Reference 51 is the"Cable Identification and Qualification Supplement".
This reference allows traceability of all cable used in the Ginna plant, by referencing back to the original purchase order specifications.
It can be seen that, in addition to the Kerite safeguards cable, the only other cable inside containment used to perform a required 57 post-accident safety f unction is the silicone-rubber insulated cable, which is used for all required safety-related instrumentation and control cable.Reference 46 identifies this as Coleman cable.In addition to the testing stated in Reference 46, a section of this cable was taken from the Ginna plant, and environmentally qualif ied with the Raychem splice sleeves (documentation of the testing is given in FRC Final Report Supplement, F-C5074 (Supplement), April 1979, which is included in Reference 51).The cable is specimen number C5074-7 of Table 1 of F-C5074 Supplement.
This testing shows that the Coleman silicone-rubber insulated cable will perform its required safety functions inside containment.
Reference 46 states that this cable is aged at 200'C for 168 hours.Although no specific Arrhenius plot is available, the application of the"10'C rule" shows an operating life of 40 years at 60'C.This is considered a reasonable estimate of the exoected life of this cable.38.TER Paragraph 3.3.3.12-Table 3 Items 17A, 17B, 17C.Coleman, Rome, and General Cables Used Outside Containment.
Reference 51 is the"Cable Identification and Qualification Supplement".
From this reference, the type of cable used throughout the Ginna plant 58 can be traced by reference back to the original purchase order specification.
It is shown that all of the safety-related cable outside containment which is not Kerite cable is PVC-insulated cable.The specif ications included in Reference 51 refer to GAI Specs SP-5324 and SP-5315.Both of these specifications in turn specify the requirements of IPCEA S-61-402 for PVC-Cable.
Inf ormation f rom this standard is provided in Reference 10.Additional information for Coleman and Rome cable is provided in Ref erence 4 6.The IPCEA testing of this cable, including insula-tion aging at 121'C (250'F)for 168 hours (jacket at 212'F), oil immersion, heat shock, and cold shock, shows the ability to operate under conditions more severe than those anticipated outside containment.
Although no specif ic qualif ication testing was performed, the standard testing of these cable types gives reasonable assurance that they are suitable for outside-containment use.39.TER Paragraph 3.3.3.13-Table 3 Item No.27.RTDs Inside Containment.
Reference 35 is a specification sheet and drawing of the Ginna RTD (Rosemount 176JA model).The reactor coolant system temperature detectors (RTD)are not required for a loss of coolant-59 accident.In a steam line break accident, low Tave plus high steam flow plus" a safety injection signal will close the main steam line isolation valves.Also, high-high steam flow will perform this function.As described in Section II.B above, for a break upstream of the non-return check valves, which includes all breaks inside containment, closure of the main steam isolation valves is not required.For breaks downstxeam of the check valves, closure of the main steam isolation valves is desirable, however, in this case the RTDs are not subjected to an adverse environment.
Theref ore, the RTDs do not require environmental qualification to px'ovide their required safety function.However, the RTDs would be useful for post-accident monitoring.
Since the RTDs are not qualified for post-accident use, the pxesent Ginna Emergency Procedux'es specify that, if a 50'F subcooling margin cannot be established or maintained, safety injection flow shall not be terminated.
Failure of the RTDs would require that SI flow be maintained.
Since the Ginna high head safety injection pumps do not have a high enough shutoff head to open the pressurizer PORVs, continued SI pump operation is not a safety concern.However, to avoid the possibility of operator confusion, RG&E will initiate a program to provide 60  


qualified RTDs for post-accident monitoring.
qualified   RTDs for post-accident monitoring.
These will be procured and installed by June 1982, I sub ject to equipment availability and procurement/
These will be   procured and installed by June 1982, I
sub ject to equipment     availability and procurement/
delivery schedules.
delivery schedules.
40.TER Paragraph 3.3.3.14-Table 3 Item No.28.Batteries in the Control Building.As noted in TER C5257, the ventilation system is being modified, such that the battery rooms can be considered a"mild" environment.
: 40. TER Paragraph   3.3.3.14 Table   3 Item No. 28. Batteries in the Control Building.
Reference fHELB-13]committed to a resolution of the potential flooding problem.The batteries will thus be further discussed at a later time, together with other equipment located in a"mild" environment.
As noted in   TER C5257, the ventilation system is being modified, such that the battery rooms can be considered   a "mild" environment. Reference fHELB-13]
41.TER Paragraph 3.3.3.15-Table 3 Item No.26.Steam Generator Level Transmitter.
committed to a resolution of the potential flooding problem. The batteries will thus be further discussed at a later time, together with other equipment located in a "mild" environment.
The steam generator level transmitters, although useful for confirming secondary system heat removal capability, are not necessary for performing this function.For an accident inside containment, which could degrade the performance of the SG level transmitters, the main steam pressure transmitters, located outside containment, provide information regarding steam generator status.Auxiliary feedwater flow instrumentation for each steam generator, also located outside containment, provides the primary indication of the steam generator heat 61 removal capability.
: 41. TER Paragraph 3.3.3.15     Table 3 Item No. 26. Steam Generator Level Transmitter.
Based on the latest information provided at the Westinghouse Emergency Operating Instructions seminar, the Ginna Emergency Procedures will be revised to reflect AFW flow indications as being of prime value as the main indication of secondary heat removal capability.
The steam   generator level transmitters, although useful for confirming secondary system heat removal capability, are not necessary for performing this function. For an accident inside containment,     which could degrade the performance of the     SG level transmitters, the main steam pressure transmitters, located outside containment, provide information regarding steam generator status. Auxiliary feedwater flow instrumentation for each steam generator, also located outside containment, provides the primary indication of the steam generator heat 61
Nevertheless, in order to remove the possibility of operator confusion due to misleading instrument indications, the steam generator Level trans-mitters will be replaced by June 1982.Qualifica-tion documentation will be made available when received.42.TER Paragraph 3.3.3.16-Table 3 Item Nos.29A, 29B, 29C.Diesel Generator Electrical Equipment.
 
This equipment is located in a"mild" environment.
removal capability. Based on the latest information provided at the Westinghouse Emergency Operating Instructions seminar, the Ginna Emergency Procedures will be revised to reflect AFW flow indications as being of prime value as the main indication of secondary heat removal capability.
Its qualification will reviewed at a later date.43.TER Paragraph 3.3.3.17-Table 3 Item No.35.Valcor Solenoid Valves for the Pressurizer PORVs.Additional information has been added to Reference 48, consisting of the test results and testing methodology.
Nevertheless,   in order to remove the possibility of operator confusion due to misleading instrument indications, the steam generator Level trans-mitters will be replaced by June 1982. Qualifica-tion documentation will be made available when received.
This was provided to the NRC and FRC on September 24, 1980.The entire test report is also available for audit and review at RGSE.These valves are fully qualified to IEEE-323-1974 to perform their post-accident safety function.62  
: 42. TER Paragraph 3.3.3.16 - Table 3 Item Nos. 29A, 29B, 29C. Diesel Generator Electrical Equipment.
This equipment is located in a "mild" environment.
Its qualification will reviewed at a later date.
: 43. TER Paragraph 3.3.3.17 Table 3 Item No. 35. Valcor Solenoid Valves for the Pressurizer PORVs.
Additional information has been added to Reference 48, consisting of the test results and testing methodology. This was provided to the NRC and FRC on September 24, 1980. The entire test report is also available for audit and review at RGSE.
These valves are   fully qualified to IEEE-323-1974 to perform their post-accident safety function.
62


I 44.TER Paragraph 3.3.3.18-Table 3 item No.36.Sump B Wide Range Level Switch.Ref erence 52, the specif ication sheet f or this item, was provided to the NRC and FRC on September 24, 1980.There is evidence that these level switches can perform their function in a contain-ment post-accident environment.
I
However, not all of the requirements of the DOR Guidelines are met for this instrumentation.
: 44. TER Paragraph   3.3.3.18   Table 3 item No. 36.     Sump B Wide Range Level Switch.
Xt is important to note, however, that these instruments are not used to perf orm any post-accident saf ety-related f unctions, and are not specified for use in the Ginna Emergency Procedures except as confirmatory information.
Ref erence 52, the   specif ication sheet for this item, was provided to the NRC and FRC on September 24, 1980. There is evidence that these level switches can perform their function in a contain-ment post-accident environment.         However, not all of the requirements of the       DOR   Guidelines are met for this instrumentation.       Xt is important to note, however, that these instruments are not used to perf orm any post-accident saf ety-related f unctions, and are not specified for use in the Ginna Emergency Procedures except as confirmatory information.
The saf ety-related f unction of determining the timing of the"sump switchover" procedure is performed by the RWST level instrumentation, located outside containment.
The saf ety-related function of determining the timing of the   "sump switchover" procedure is performed by the RWST level instrumentation, located outside containment.
The TMI Lessons Learned determined that a wide-range sump level indication was to be provided for operator information.
The TMI Lessons     Learned determined     that a wide-range sump   level indication was to be provided for operator information. Fully-qualified equipment will be purchased to meet this requirement. The qualification documentation for this instrumenta-tion will be made available when received.
Fully-qualified equipment will be purchased to meet this requirement.
: 45. TER Paragraph 3.3.3.19 - Table 3 Xtem Nos. 42, 43.
The qualification documentation for this instrumenta-tion will be made available when received.45.TER Paragraph 3.3.3.19-Table 3 Xtem Nos.42, 43.Motors for Cooling Fans for RHR, CS, Sl, and Charging Pumps in Auxiliary Building.63 Reference 69 provides information concerning the life and radiation characteristics of these motors.These motors are capable of operation after a radiation exposure of 1 x 10 rads and 20 years.7 Since these motors are run only intermittently, operational capability for 40 years is shown.Since the only harsh environment experienced by these motors is post-LOCA radiation (estimated at 2.8 x 10 rads), operation under required accident 6 conditions is shown.46.TER Paragraph 3.3.3.20-Table 3 Item Nos.32, 44.IGC Cabinets and Relay Racks in Relay Room.This equipment is located in a mild environment.
Motors for Cooling Fans for RHR, CS, Sl, and Charging Pumps in Auxiliary Building.
Its qualification will be considered at a later time.47.TER Paragraph 3.3.3.21-Table 3 Item No.33A.Control Room HVAC Air Handling Units.This equipment is located in a mild environment.
63
Its qualification will be considered at a later time.48.TER Paragraph 3.3.3.22-Table 3 Item No.33B.Control Room HVAC Fans.This item is not an electrical piece of equipment.
 
It has thus been deleted from Table 3, and from consideration in this report.64 49.TER Paragraph 3.3.3.23-Table 3, Item No.39, Charging Pumo Mo tors.This equipment is located in a mild environment.
Reference   69 provides information concerning the life and radiation characteristics of these motors.
Its qualification will be considered at a later time.50.TER Paragraph 3.3.3.24-Table 3 Item No.40.Control Room HVAC Damper Solenoids.
These motors are capable of operation after a radiation exposure of     1 x 10 7 rads and 20 years.
Since these motors are run only       intermittently, operational capability for 40 years is shown.
Since the only harsh environment experienced by these motors is post-LOCA radiation (estimated at 2.8 x 10 6 rads), operation under required accident conditions is shown.
: 46. TER Paragraph   3.3.3.20   Table 3 Item Nos. 32, 44. IGC Cabinets and Relay Racks in Relay       Room.
This equipment is located in       a mild environment.
Its qualification will be considered at         a later time.
: 47. TER Paragraph   3.3.3.21   Table 3 Item No. 33A. Control Room HVAC Air Handling Units.
This equipment     is located in a mild environment.
Its qualification will be considered at a later time.
: 48. TER Paragraph   3.3.3.22   Table 3 Item No. 33B. Control Room HVAC Fans.
This item is not an     electrical piece of     equipment.
It has   thus been deleted from Table 3, and from consideration in this report.
64
: 49. TER Paragraph   3.3.3.23 Table 3, Item No. 39, Charging Pumo Mo tors .
This equipment is located in   a mild environment.
Its qualification will be considered at     a later time.
: 50. TER Paragraph   3.3.3.24 Table 3 Item No. 40. Control Room HVAC   Damper Solenoids.
This equipment is located in a mild environment.
This equipment is located in a mild environment.
Its qualification will.be considered at a later time.65 LOSS OF COOLANT ACCIDENT 1.2/3 HIGH CONTA I NMENT PRESSURE HI HI j 2/3 LOW PRESSURIZER PRESSURE FIGURE 1 SAFETY INJECTION la ACCIDENT DIAGNOSTICS 4.HAIN STEAM LINE ISOLATION 3.ACCUtlULATOR DUtlP 2.SAFETY INJECTION SEQUENCE (AUTO)4.FEEDl<ATER LINE ISOLATION 5.CONTA I Nf 1ENT ISOLATION 6.REACTOR TRIP VALVES 7.REACTOR COOLANT PUf'lp TRIP 9: CONTROL ROOM VENTILATION 10.MANUAL ACTIONS RECIRC-ULATION TABLE 1 LOSS OF COOLANT ACCIDENT BLOCK NO./EQUIPMENT SAFETY FUNCTION REQUIRED OPERATION TIME 1.High Containment Pressure Low Pressurizer Pressure PT 945, 946, 947 PT 948)949, 950 Provide signals for Contain-ment Spray, Safety Injection, Containment Isolation, and Main Steam and Feedwater Line Isolation Signal Initiation PT 429)430, 433.)449 Accident Diagnostics Provide Reactor trip and Safety Injection signals Short term Signal Initiation Splice Sleeves, Terminal Blocks, Electrical Pene-trations, Electrical Cable Accident Diagnostics Control and Power Signal Transmission Short term Long term la.Steam Line Pressure PT 468)469)482 PT 478, 479)483 Accident Diagnostics Short term'ontainment Radiation[Being provided per TMI STLL]Accident Diagnostics Short term Containment'sump level IT 942, LT 943 Accident Diagnostics Short term 2.Safety Injection Sequence (Auto)Batteries lA, 1B Diesel Generator and Auxiliaries D.C.Power Power supply to safeguards busses during loss of out-side AC Power Long Term Long term 480 Volt Safeguards busses 14, 16, 17, 18 Provide.the distribution of power to safeguards equipment Long term lA, 1B, 1C Safety Injec-tion Pumps High head injection of bo-rated water to Reactor Coolant System Long term lA, 1B Containment Spray Pumps (only on hi-hi Cont.pressure)Containment Pressure, Tem-perature, and Iodine control Long term TABLE 1 ,f BL CK'NO./EQUIPMENT LOSS OF COOPT ACCIDENT SAFETY FUNCTION RE(}UIRED.
Its qualification will .be considered at a later time.
OPERATION TIME 1.<, 1B Residual Heat Re-.moval Pumps/1A;1B, 1C, 1D Service Mater Pumps Low head injection of borated water to Reactor Vessel Cooling water to RHR and CCN Heat Exchangers Long term Long term 1A, 1B, 1C," lD Contain-ment Recirc.Units Containment Pressure, Tem-perature, and Iodine control Long term Cooling Units for pump motors (SI, RHR, CS, and Charging)Haintain motors within proper ambient temperature limits Long Term 1A, 1B Hotor Driven Aux.Feedwater Pumps Cooling water to Steam Gen-erators Long term 480 Volt Safeguards MCC's Provide the distribution of power to safeguards equipment Long term 3~Accumulator Dump HOV 841 (N.O.)-'OV 865 (N.O.)Provide path to Reactor Vessel from Accumulators for injection of borated water Not required to function 4.Main Steam Line Isolation Feedwater Line Isolation AOV.3516 AOV 3517 AOV 4269 AOV 4270 AOV 4271 AOV 4272 Isolate 1A, 1B Steam Generators Isolate Hain Feedwater System 5 Seconds after signal 5 Seconds a f ter signal 5.Containment Isolation See Text, Section II.A.5 6.Reactor Trip Reactor trip breakers 0 Provide means to trip the reactor Required for Reactor Trip Reactor protection and in-strumentation cabinets Provide the instrumentation and protection circuits for the con-trol and tripping of the Reactor Required for Reactor Trip 7.RCP Trip RCP Trip Breakers Provide means to trip RCP's Short term N.O.=Normally Open I
65
CK NO./EQUIPHENT LOSS OF COOLANT ACCIDENT SAFETY FUNCTION REQUIRED OPERATION TIHE alves HOV 825 A)B HOV,826 A)B)C D (Baa N.O.)AOV 836 A)B Provide path to SI Pumps for bor-ated water to high head safety injection Provide controlled addition of NaOH to Containment Spray for Iodine control 10/BAST Level or-1/2 hour Short term HOV 852 A)B HOV 860 A)B,C)D BAST Level IT 102)106, 171)172 HOV 878 B)D (N.O.)Provide path to Reactor Vessel of borated water for low head safety injection Provide path to Containment Spray headers for CS Pumps Indicate BAST Level for automatic transfer of SI Pump suction from BAST to RMST Provide path to cold legs of RCS from high head safety injection SI'initiation I,ong term 10%BAST Ievel or-1/2 hour not required to function HOV 4007, 4008 1A, 1B Steam Generators Provide path for Aux.Feedwater to Short term AOV 5871, 5872, 5873 AOV 5874, 5875)5876 9.Control Room Ventilation Dampers and AiiU 10: Hanual Provide path for cleaning of cont.atmosphere by fan coolers Provide cleaning of Control Room atmosphere signal initiation Short term Safety Injection Reset Button 1A, 1B Component Cooling Mater Pumps 1A, 1B Containment Spray Pumps (if Cont.Pressure (30 psig)Reset Safety Injection signal after, automatic S.I.Sequencing is complete Cooling water for safeguards equipment Containment Pressure, Temperature and Iodine control less than 24 hours Long term Long term RWST Level LT 920, LIC 921 Indicate RMST Level for operator less than 24 hours transfer from S.I.phase to Recirculation phase
 
'N I TABLE 1 f BLOCK NO./EQUIPHENT LOSS OF COOLS'CCIDENT SAFETY FUNCTION REQUIRED OPERATION TIHE HOV 4027, 4028 HOV 4000A, 4000B HOV 4734)4735)4615, 4616 HOV 738 A)B Standby AFW Pumps Provide Service Mater to Hotor Driven Aux.Feedwater Pumps suction Provide AFW Cross-Connect Direct SW Flow to CCW HX's Direct CCW Flow to RHR HX's AFW Flow to SG's if normal AFM System inoperable within-2 hours Short term less than 24 hours less than 24 hours Long term HOV 9629 A,B Provide SW to suction of standby Long term AFM Pumps HOV 9710 A,B;9703 A,B;9704 A)B Steam Generator Level LT 460, 461, 462, 463 LT 470)471, 472)473 Sampling (being provided per THI)e Hydrogen Recombiners Pressurizer PORVs.11.Recirculation HOV 850 A,B outside cont.HOV 851 A,B (N.O.)inside cont.Standby AFM Discharge Valves to provide flow to SG's Honitoring Sample containment atmosphere and reactor coolant Haintain hydrogen control RC Pressure Control Provide path to RHR suction from B sump for low head safety injec-tion Long term Long term I,ong term Long term Long term Long term HOV 856 (N.O.)HOV 896 A,B (N.O.)HOV 857 A,B,C AOV 897)898 RWST isolation valve to RHR pumps suction, must close after RMST is drained RMST isolation valve, must close after RWST is drained Provide path to suction of SI and CS Pumps from RER pumps discharge Isolate high head recirculation flow to RWST during sump recir-culation required to func-tion to switch to recirc phase required to func-tion to switch to recirc phase required to func-tion to switch to recirc phase Short term HOV 704 A)B recirculation Close during switch to sump less than 24 hours  
LOSS OF COOLANT ACCIDENT 1 .       2/3                  2/3 HIGH               LOW CONTA I NMENT           PRESSURIZER PRESSURE            PRESSURE FIGURE 1 HI HI j
la SAFETY             ACCIDENT INJECTION           DIAGNOSTICS
: 4.         3.               2.                     4.                 5.              6.
HAIN    ACCUtlULATOR          SAFETY            FEEDl<ATER         CONTA I Nf 1ENT   REACTOR STEAM LINE    DUtlP          INJECTION                  LINE            ISOLATION         TRIP ISOLATION                    SEQUENCE              ISOLATION (AUTO) 7.
REACTOR VALVES                COOLANT PUf'lp TRIP 9:
CONTROL ROOM VENTILATION 10.
MANUAL ACTIONS RECIRC-ULATION
 
TABLE 1                             LOSS OF COOLANT ACCIDENT REQUIRED BLOCK NO./EQUIPMENT                     SAFETY FUNCTION                 OPERATION TIME
: 1. High Containment Pressure Low Pressurizer Pressure PT 945, 946, 947               Provide signals    for Contain-  Signal  Initiation PT 948) 949, 950               ment Spray, Safety     Injection, Containment   Isolation, and Main Steam and Feedwater Line Isolation Accident Diagnostics              Short term PT 429) 430, 433.) 449         Provide Reactor trip and         Signal  Initiation Safety Injection signals Accident Diagnostics              Short term Splice Sleeves, Terminal       Control and Power Signal          Long term Blocks, Electrical Pene-       Transmission trations, Electrical     Cable la. Steam Line Pressure           Accident Diagnostics              Short term PT 468 ) 469 ) 482 PT 478, 479) 483 Radiation      Accident Diagnostics             Short term 'ontainment
[Being provided per TMI STLL]
Containment'sump    level      Accident Diagnostics             Short term IT 942,   LT 943
: 2. Safety Injection Sequence   (Auto)
Batteries                     D. C. Power                      Long Term lA, 1B Diesel Generator     Power supply   to safeguards     Long term and  Auxiliaries              busses   during loss of out-side AC Power 480 Volt Safeguards           Provide. the  distribution of    Long term busses   14, 16, 17, 18       power to safeguards     equipment lA, 1B, 1C   Safety Injec-   High head   injection of bo-     Long term tion  Pumps                  rated water to Reactor Coolant System lA, 1B Containment Spray     Containment Pressure,    Tem-    Long term Pumps  (only on hi-hi Cont. perature,   and Iodine control pressure)
 
TABLE 1                                   LOSS OF  COOPT ACCIDENT
          ,f RE(}UIRED.
BL CK 'NO./EQUIPMENT                           SAFETY FUNCTION                     OPERATION TIME 1.<, 1B Residual Heat Re-             Low head injection of borated        Long term
      .moval Pumps                         water to Reactor Vessel
          /
1A; 1B, 1C, 1D Service               Cooling water to    RHR  and CCN      Long term Mater   Pumps                         Heat Exchangers 1A, 1B, 1C," lD Contain-             Containment Pressure,     Tem-       Long term ment Recirc. Units                    perature,   and Iodine   control Cooling Units for           pump     Haintain motors within proper        Long Term motors (SI, RHR, CS,                 ambient temperature    limits and Charging) 1A, 1B Hotor           Driven         Cooling water to Steam Gen-           Long term Aux. Feedwater Pumps                  erators 480 Volt Safeguards                   Provide the distribution of           Long term MCC's                                power to safeguards   equipment 3 ~ Accumulator           Dump HOV 841   (N.O.)-'OV Provide path to Reactor Vessel        Not required 865 (N.O.)                     from Accumulators for injection       to function of borated water
: 4. Main Steam Line Isolation Feedwater Line Isolation AOV .3516                             Isolate  1A, 1B Steam Generators    5 Seconds  after AOV 3517                                                                    signal AOV   4269 4270 Isolate Hain Feedwater     System     5 Seconds   a fter AOV                                                                          signal AOV  4271 AOV  4272
: 5. Containment           Isolation       See Text, Section II.A.5
: 6. Reactor Trip 0
Reactor   trip         breakers       Provide means to   trip the reactor Required for Reactor Trip Reactor protection and             in- Provide the instrumentation and       Required for strumentation cabinets                protection circuits for the con-     Reactor Trip trol and tripping of the Reactor
: 7. RCP   Trip RCP   Trip Breakers                   Provide means to   trip   RCP's       Short term N.O. = Normally Open
 
I LOSS OF COOLANT ACCIDENT REQUIRED CK NO./EQUIPHENT                   SAFETY FUNCTION                     OPERATION TIHE alves HOV 825   A)B                 Provide path to SI Pumps for bor-    10/  BAST  Level HOV,826 A ) B ) C D           ated water to high head safety      or-1/2 hour (Baa N.O.)                   injection AOV 836   A)B                 Provide controlled addition of      Short term NaOH  to Containment Spray for Iodine control HOV 852   A)B               Provide path to Reactor Vessel      SI'initiation of borated water for low head safety injection HOV 860   A)B,C)D             Provide path to Containment Spray    I,ong term headers  for  CS Pumps BAST   Level                   Indicate BAST Level for automatic    10% BAST Ievel IT  102) 106, 171) 172       transfer of SI Pump suction from    or-1/2 hour BAST  to RMST HOV 878 B)D                   Provide path to cold legs of RCS    not required (N.O.)                         from high head safety injection     to function HOV  4007, 4008              Provide path for Aux. Feedwater to   Short term 1A, 1B Steam Generators AOV  5871, 5872, 5873        Provide path for cleaning of cont. signal  initiation AOV  5874, 5875) 5876        atmosphere by fan coolers
: 9. Control    Room  Ventilation  Provide cleaning of Control    Room  Short term Dampers and AiiU              atmosphere 10: Hanual Safety Injection Reset        Reset Safety  Injection signal      less than 24 hours Button                        after, automatic S.I. Sequencing is complete 1A, 1B Component    Cooling  Cooling water  for safeguards      Long term Mater  Pumps                  equipment 1A, 1B Containment Spray      Containment Pressure,   Temperature Long term Pumps    (if Cont. Pressure  and Iodine control (30 psig)
RWST  Level                  Indicate RMST Level for operator    less than  24 hours LT 920, LIC 921                transfer from S.I. phase to Recirculation phase
 
'N I
 
TABLE 1                            LOSS OF  COOLS'CCIDENT f
REQUIRED BLOCK NO./EQUIPHENT                    SAFETY FUNCTION                    OPERATION TIHE HOV  4027, 4028              Provide Service Mater to Hotor      within-2 hours Driven Aux. Feedwater Pumps suction HOV  4000A, 4000B            Provide  AFW  Cross-Connect        Short term HOV 4734) 4735)   4615, 4616 Direct  SW Flow  to CCW HX's        less than 24 hours HOV 738 A)B                  Direct  CCW Flow  to RHR HX's      less than 24 hours Standby  AFW Pumps          AFW  Flow to SG's System inoperable if normal AFM  Long term HOV 9629    A,B              Provide  SW  to suction of standby  Long term AFM Pumps HOV   9710 A,B; 9703 A,B;    Standby  AFM  Discharge Valves to  Long term 9704 A)B                     provide flow to SG's Steam Generator Level        Honitoring                          Long term LT 460, 461, 462, 463 LT 470) 471, 472) 473 Sampling (being provided     Sample containment atmosphere      I,ong term per THI)                     and reactor coolant e
Hydrogen Recombiners         Haintain hydrogen control          Long term Pressurizer   PORVs         RC Pressure  Control              Long term
. 11. Recirculation HOV 850   A,B outside cont. Provide path to RHR suction from    Long term HOV 851   A,B (N.O.) inside   B sump for low head safety injec-cont.                         tion HOV 856    (N.O.)            RWST  isolation valve to RHR pumps  required to func-suction, must close after RMST is  tion to switch to drained                            recirc phase HOV 896  A,B (N.O.)         RMST  isolation valve,  must close required to func-after  RWST is drained              tion to switch to recirc  phase HOV 857  A,B,C               Provide path to suction of SI and  required to func-CS Pumps from RER pumps discharge   tion to switch to recirc  phase AOV  897) 898                Isolate high head recirculation     Short term flow to RWST during sump recir-culation HOV 704    A)B                Close during switch to sump        less than 24 hours recirculation
 
MAIN STEAM OR FEED LINE    B FIGURE 2
: 3.        2/3    1.          2/3    2.        2/3 HIGH                                    LOM CONTAINMENT        STEAN LINE          PRESSURIZER PRESSURE          PRESSURE          PRESSURE HI HI
: 3.        2/3      3.        2/4                                                      3.        2/4 I
STEAN LINE            LOW                        SAFETY            ACCIDENT          OVERPOWER FLOIA              T ave                      INJECTION        OIAGIIOSTICS            hT I
HI  1
(
I.
: 4.                  6.                                      5.
MAIN              SAFETY          FEEDllATER STEAN LINE            INJECT ION            LINE          CONTAINMENT          REACTOR ISOLATION          SEQUENCE          ISOLATION            ISOLATION            TRIP (AUTO)
: 9.                  8.
REACTOR VALVES            COOLANT PUMP TRIP 10.
MANUAL ACTIONS 11.
CONTINUED SAFE SHUTDOWN
 
TABLE 2                                MAIN STEAM LINE BREAK SAFETY FUNCTION/BREAK LOCATION REQUIRED BLOCK NO./EQUIPMENT                    SAFETY FUNCTION                  OPERATION TIME INSIDE CV          OUTSIDE CV
: 1. Steam Line Pressure            Provide signal for        same    signal  initiation PT 468, 469, 482              SI on low steam line PT 478) 479) 483              pressure la. Steam Line Pressure            Accident Diagnostics      same    short term (see  1 above)
Containment Radiation          Accident Diagnostics      NA      short term Containment    Sump Level      Accident Diagnostics      NA      short term High Containment Pressure      Accident Diagnostics      NA      short term (see 3 below)
: 2. Low  Pressurizer Pressure PT 429,  430, 431) 449        Provide Reactor trip      same    signal  initiation and Safety Injection signals Electrical Penetrations,      Provide control and      same    long term Cable, Sleeves, and            Power Signal Terminal Blocks                Transmission High Containment Pressure PT 945) 946, 947              Provide signals for      NA      signal  initiation PT 948) 949~ 950              Containment Spray, Safety Injection, Containment  Isola-tion,  and Main- Steam Line Isolation Steam Line Flow FT 464, 465                    Provide signals for      same    signal  initiation FT 474, 475                  Reactor trip and Main Steam Line Iso-lation Reactor Coolant Temperature Loop A Hot Ieg                Provide Iow Tave 6        same    signal initiation TE 401A, 402A)                6 signals  for'Reactor 405A, 406A,                trip,  Safety Injec-409A                      tion  and Main Steam Line Isolation
 
TABLE 2                                    MAIN STEAM LINE BREAK                      -  2-SAFETY FUNCTION/BREAK LOCATION REQUIRED BLOCK NO./EQUIPMENT                        SAFETY FUNCTION                  OPERATION TIME INSIDE  CV          OUTSIDE CV Loop A Cold Leg TE 401B> 404A, 407A>
408A, 410A Loop  B  Hot Leg TE 403B> 404B, 407B, 408B, 410B Loop  B  Cold I,eg TE 403B> 404B> 407B, 408B> 410B Main Steam    Isolation AOV 3516                          Isolate  1A, B Steam      same    5 seconds  after signal AOV 3517                          Generators Feedwater Line      Isolation AOV  4269                          Isolate  Main Feed-      same    5  seconds  after signal AOV  4270                          water system AOV  4271 AOV  4272 Containment    Isolation          See  Text, Section        same II.B.5 Safety Injection Sequence    (Auto)
Batteries                          D.C. Power                same    Long term 1A, 1B    Diesel                  Power supply to safe-      same    Long term Generators and                    guards busses during auxiliaries                        loss of.,outside  AC Power 480  Volt Safeguards              Provide  distribution    same    Long term busses    14, 16, 17, 18          of power to safe-guards equipment 1A, 1B,    1C  Safety In-          High head. injection      same    Long term jection    pumps                  of borated water to Reactor Coolant System lA,  B  Containment Spray          Containment Pressure      N/A    I,ong term Pumps    (only on  hi-hi cont. and Temperature Pressure)                          control 1A, 1B, 1C,    1D  Service        Cooling Water to          same    Long term Water Pumps                        CCW  Heat Exchanger
 
HAIN STEAM LINE BREAK SAFETY FUNCTION/BREAK IOCATION REQUIRED BLOCK NO./EQUIPMENT                      SAFETY FUNCTION                                  OPERATION TIME INSIDE CV                        OUTSIDE CV 1A, 1B, 1C, 1D Containment      Containment Pressure                    N/A      Long term Recirc Units                    and Temperature con-trol 1A, 1B Motor Driven Aux.        Cooling w'ater supply                  same      Long term Feedwater Pumps                  to Steam Generators Cooling Units for SI,    CS,    Maintain motors                        same      Long term RHR, and  Charging  Pump        within proper ambient temperature  limits 480  Volt Safeguards            Provide the  distribu-                same      Long term HCCs                            tion of  power to safeguards equipment
: 7. Reactor Trip Reactor  trip  breakers        Provide means to                        same      Required  for trip the reactor                                'eactor    Trip Reactor Protection and          Provide the instru-                    same      Required for Instrumentation                  mentation and pro-                                Reactor Trip Cabinets                        tion circuits for the control and tripping of the reactor
: 8. Reactor Coolant Pump Trip        Provide  means  to              trip    NA        Short term RCP Trip Breakers                RCPs
: 9. Valves HOV 825A> B                      Provide path to SI                      same      10/ BAST  Level HOV 826A, B) C, D                Pumps for borated.                                o~l/2  hour (Baa N.O.)                      water to high head safety injection AOV  836A,  B .                  Provide needed NaOH  to              CS if              Short term HOV  860A, B, C)  D            Provide path to Con-                    N/A      Long term tainment, Spray headers for CS Pumps'rovide HOV  878, B,  D                          path to                        same      not required to (N.O.)                          cold legs of,RCS                                  function from high head safety injection
 
TABLE 2                                  MAIN STEAM LINE BREAK SAFETY FUNCTION/BREAK LOCATION REQUIRED BIOCK NO./EQUIPMENT                      SA'FETY FUNCTION                    OPERATION TIME INSIDE CV              OUTSIDE CV HOV      896)A)B)(NO)            Provide path from            same    short-term (to close RWST  of borated                    if  need sump water for SI and                    recirculaton)
CS  pumps  suction MOV      4007) 4008              Provide path for Aux.        same    Short term Feedwater to Steam Generators AOV      5871) 5872) 5873        Provide path for            N/A    signal  initiation AOV      5874, 5875) 5876        cleaning by fan coolers, cooling of cont. Atmosphere BAST      Level            1 Indicate  BAST  Level      same    10/ BAST I,evel LT 102) 106)        171)'72      for automatic trans-                or~1/2 hour fer of SI Pump suction from    BAST to RWST MOV      852A,  B                Provide path for low        same    Signal  Initiation head SI to Reactor Vessel
: 10. Manual'G Level Instrumentation    Determine affected    SG    same    Short term LT 470, 471, 472, 473 LT 460, 461, 462) 463 Safety Injection Reset            Reset SI signal after        same    less than 24 hours Button                            Automatic SI sequenc-ing is complete 1A, 1B Component Cooling          Cooling Water for            same    Long term Water      Pumps                safeguards equipment 1A, 1B Containment                Containment Pressure        N/A    Long term Spray Pump (If cont.              and Temperature con-Pressure < 30 psig)              trol MOV      402?, 4028              Provide Service Water        same    within ~2 hours to Motor Driven Aux.
Feedwater Pumps Suction Charging pumps                    Inventory control to        same    Long term RCS
 
TABLE 2                            HAIN STEAH LINE BREAK SAFETY FUNCTION/BREAK LOCATION REQUIRED BLOCK NO./EQUIPHENT                  SAFETY FUNCTION                    OPERATION TIHE INSIDE  CV            OUTSIDE CV Standby  AFW pumps          Provide flow to              same    I,ong term SGs  if  AFW  system  in-operable HOV  9629A, B              Provide  SW  to suction    same    Long term of Standby  AFW Pumps MOV  9710A, B; 9703A, B;    Standby AFW discharge        same    Long term 9704A, B                    valves to provide AFW flow to SGs HOV  4000A, B              AFW  Cross-Connect          same    Short term Valves
: 11. Continued Safe Shutdown Sampling (per THI)          Sample Containment          same    Long term Atmosphere and Reactor Coolant Pressurizer  PORVs          RC  Pressure  Control      same    Long term
 
Accident References LOCA  analysis    [LOCA]
FSAR
: 2.    "ECCS Analysis  for the R. E. Ginna Reactor  with ENC WREM-2 PWR Evaluation Model"    dated December 1977 sub-mitted with Application for    Amendement to Operating License, on January 6, 1978.
: 3. ECCS Analysis submitted by letter dated April 7, 1977 from L. D. White, Jr., RG&E to A. Schwencer, Chief, Operating Reactors Branch    Il,  USNRC.
4,    ECCS Analysis    for the R. E. Ginna Reactor with ENC WREM-2 PWR Evaluation Model.      Exxon Nuclear Co.
Report XN-NF-77-58.
: 5. Ginna Emergency Procedures E1.1 and E1.2, submitted by letter dated February 26, 1980 from L. D. White, Jr.
RG&E, to D. L. Ziemann, USNRC.
Steam  Line Break and Feedwater Line Break [SLB/FLB]
: 2. Steam line break analyses submitted with Application for Amendment to Operating License on September 22, 1975.
: 3. Plant 'Transient. Analysis for the R. E. Ginna Unit 1 Nuclear Power Plant, Exxon Report XN-NF-77-40 (11/77 and updated 12/15/78 and March, 1980.
Letter dated May 24, 1977 from K. W. Amish, RG&E to J. F.- O'eary, NRC.
: 5. Ginna Emergency Procedures E1.1 and E1.3, submitted by letter dated February 26, 1980 from L. D. White, Jr.,
RG&E to D. L. Ziemann, USNRC.
6'. Letter from L. D. White, Jr.,  RG&E,  to D. L. Ziemann, NRC,  March 28, 1980.
High Energy Line Break [HELB]
            "Effects of Postulated'Pipe Breaks Outside the Con-tainment Building", GAI Report No. 1815, submitted by letter  dated November 1, 1973 from K. W. Amish, RG&E, to  A, Giambuso, Deputy Director for Reactor Projects, USNRC.
 
Letter dated  May 24, 1974 from K. W. Amish, RG&E,      to J. F. O'eary, Director, Directorate of Licensing, USNRC.
Letter dated September 4, 1974 for R. R. Koprowski, RG&E  to Edson Case, Acting Director, Directorate of Licensing,  USNRC.
Letter dated November 1, 1974 from K. W. Amish, RG&E, to Edson Case, Acting Director, Directorate of Li-censing,  USNRC.
Letter dated May 20, 1977 from L. D. White, Jr., RG&E, to A. Schwencer, Chief Operating Reactors Branch 51, USNRC.
Letter dated February    6, '1978 from L. D. White,    Jr.,
RG&E, to A. Schwencer,  Chief, Operating Reactors Branch Ol, USNRC.
Amendment No. 7    to Provisional Operating License DPR-18, transmitted, by letter dated May 14, 1975 from Robert A.
Purple, Chief, Operating Reactors Branch-51, USNRC, to L. D. White, Jr , RG&E.
Amendment No. 29 to Provisional Operating License DPR-18, transmitted by letter dated August 24, 1979 from Dennis L.
Ziemann, Chief, ORB 52, to L. D. White, Jr., RG&E.
Letter, L. D. White, Jr., RG&E, to D. L. Ziemann, May 17, 1979.
Letter, L. D. White,  Jr.,  RG&E,  to  D. L. Ziemann,  USNRC, June 27, 1979.
Letter, L. D. White,  Jr.,  RG&E,  to  D. L. Ziemann,  USNRC July  6, 1979.
Letter,  R. E. Anderson,    Gilbert/Commonwealth to James J.
Shea,  USNRC,  June 11, 1979.
Letter,  L. D. White,  Jr., RG&E, to D. M. Crutchfield, NRC, SEP  Topic  III-5.B, "Pipe Break Outside Containment,"
August 7, 1980.
Letter, J. Wenclawiak    and T. Snyder, Catalytic, to G. Wrobel, RG&E, "Equipment Environmental Qualification,"
October 27, 1980.
Letter from  D. M. Crutchfield,    NRC,  to L. D. White, Jr.
RG&E, SEP  Topic  III-S.B, "Pipe    Break Outside Containment,"
June 24, 1980.
 
Effects of Flooding [Flood]
Letter dated May 13, 1975 from L. D. White, Jr., RG&E, to Benard C. Rusche, Director, Office of Nuclear- Reactor Regulation,  USNRC.
: 2. Letter dated May 20, 1975 from L.- D'. White, Jr., RG&E, to Robert A. Purple, Chief, Operating Reactors Branch 51, Division of Reactor Licensing.
3.,    Letter dated May 30, 1975 from L. D. White, Jr., RG&E, to Robert A. Purple.
t Letter dated June 16, 1975 from L. D. White', Jr., RG&E, to Robert A. Purple.
: 5. Letter dated July 3, 1975 from Robert A. Purple to L. D. White,  Jr.,  RG&E.
: 6. Letter dated August. 8, 1972 from Donald J. Skovholt, Assistant Director for Operating Reactors, USAEC, to Edward J. Nelson, RG&E.
: 7. Letter dated October 3, 1972 from K. W. Amish, RG&E, to Donald J. Skovholt, Assistant Director for Operating Reactors,  USAEC.
: 8. Letter dated May 31, 1973 from K. W. Amish, RG&E, to Donald J. Skovholt, Assistant, Director for Operating Reactors,  USAEC.
: 9. Application for Amendment to Operating License, sub-mitted March 10, 1975.
: 10. Amendment, No. 14 to Provisional Operating License DPR-18, transmitted by letter dated June 1, 1977 from A. Schwencer,  Chief, Operating Reactors Branch 51, USNRC.
Letter, L. D. White, Jr. RG&E, to Dennis L. Ziemann, USNRC,  High Energy Line Breaks Outside Containment, June 27, 1979.
TMI Lessons  Learned [TMI]
RG&E  letter of  October 17, 1979, L. D. White, Jr.,
RG&E,  to D. L. Ziemann, USNRC, "TMI  Short Term Lessons Learned Requirements."
: 2. RG&E  letter of  November 19, 1979, L. D. White, Jr. to D. L. Ziemann,  USNRC,  "TMI Short Term Lessons Learned."
: 3. RG&E  letter of  December 28, 1979, L. D. White, Jr. to D.,L. Ziemann,  USNRC, "TMI Short Term Lessons Learned."
 
I I
  '\
,(
l


MAIN STEAM OR FEED LINE B FIGURE 2 3.2/3 HIGH CONTAINMENT PRESSURE 1.2/3 STEAN LINE PRESSURE 2.2/3 LOM PRESSURIZER PRESSURE HI HI 3.2/3 STEAN LINE FLOIA 3.LOW T ave 2/4 SAFETY INJECTION I ACCIDENT OIAGIIOSTICS I 3.2/4 OVERPOWER hT HI 1 (I.4.MAIN STEAN LINE ISOLATION 6.SAFETY INJECT ION SEQUENCE (AUTO)FEEDllATER LINE ISOLATION 5.CONTAINMENT ISOLATION REACTOR TRIP 9.VALVES 8.REACTOR COOLANT PUMP TRIP 10.MANUAL ACTIONS 11.CONTINUED SAFE SHUTDOWN TABLE 2 MAIN STEAM LINE BREAK BLOCK NO./EQUIPMENT SAFETY FUNCTION/BREAK LOCATION SAFETY FUNCTION REQUIRED OPERATION TIME INSIDE CV OUTSIDE CV 1.Steam Line Pressure PT 468, 469, 482 PT 478)479)483 la.Steam Line Pressure (see 1 above)Provide signal for SI on low steam line pressure Accident Diagnostics same same signal initiation short term Containment Radiation Containment Sump Level High Containment Pressure (see 3 below)Accident Diagnostics Accident Diagnostics Accident Diagnostics NA NA NA short term short term short term 2.Low Pressurizer Pressure PT 429, 430, 431)449 Electrical Penetrations, Cable, Sleeves, and Terminal Blocks Provide Reactor trip and Safety Injection signals Provide control and Power Signal Transmission same same signal initiation long term High Containment Pressure PT 945)946, 947 PT 948)949~950 Provide signals for Containment Spray, Safety Injection, Containment Isola-tion, and Main-Steam Line Isolation NA signal initiation Steam Line Flow FT 464, 465 FT 474, 475Provide signals for Reactor trip and Main Steam Line Iso-lation same signal initiation Reactor Coolant Temperature Loop A Hot Ieg TE 401A, 402A)405A, 406A, 409A Provide Iow Tave 6 6 signals for'Reactor trip, Safety Injec-tion and Main Steam Line Isolation same signal initiation TABLE 2 MAIN STEAM LINE BREAK-2-BLOCK NO./EQUIPMENT SAFETY FUNCTION/BREAK LOCATION SAFETY FUNCTION REQUIRED OPERATION TIME INSIDE CV OUTSIDE CV Loop A Cold Leg TE 401B>404A, 407A>408A, 410A Loop B Hot Leg TE 403B>404B, 407B, 408B, 410B Loop B Cold I,eg TE 403B>404B>407B, 408B>410B Main Steam Isolation AOV 3516 AOV 3517 Isolate 1A, B Steam Generators same 5 seconds after signal Feedwater Line Isolation AOV 4269 AOV 4270 AOV 4271 AOV 4272 Isolate Main Feed-water system same 5 seconds after signal Containment Isolation See Text, Section II.B.5 same Safety Injection Sequence (Auto)Batteries 1A, 1B Diesel Generators and auxiliaries D.C.Power Power supply to safe-guards busses during loss of.,outside AC Power same same Long term Long term 480 Volt Safeguards busses 14, 16, 17, 18 1A, 1B, 1C Safety In-jection pumps lA, B Containment Spray Pumps (only on hi-hi cont.Pressure)1A, 1B, 1C, 1D Service Water Pumps Provide distribution of power to safe-guards equipment High head.injection of borated water to Reactor Coolant System Containment Pressure and Temperature control Cooling Water to CCW Heat Exchanger same same N/A same Long term Long term I,ong term Long term HAIN STEAM LINE BREAK BLOCK NO./EQUIPMENT SAFETY FUNCTION/BREAK IOCATION SAFETY FUNCTION REQUIRED OPERATION TIME INSIDE CV OUTSIDE CV 1A, 1B, 1C, 1D Containment Recirc Units Containment Pressure N/A and Temperature con-trol Long term 1A, 1B Motor Driven Aux.Feedwater Pumps Cooling w'ater supply same to Steam Generators Long term Cooling Units for SI, CS, RHR, and Charging Pump Maintain motors within proper ambient temperature limits same Long term 480 Volt Safeguards HCCs 7.Reactor Trip Provide the distribu-same tion of power to safeguards equipment Long term Reactor trip breakers Reactor Protection and Instrumentation Cabinets Provide means to trip the reactor Provide the instru-mentation and pro-tion circuits for the control and tripping of the reactor same same Required for'eactor Trip Required for Reactor Trip 8.Reactor Coolant Pump Trip RCP Trip Breakers Provide means to trip NA RCPs Short term 9.Valves HOV 825A>B HOV 826A, B)C, D (Baa N.O.)AOV 836A, B.Provide path to SI Pumps for borated.water to high head safety injection Provide NaOH to CS if needed same 10/BAST Level o~l/2 hour Short term HOV 860A, B, C)D HOV 878, B, D (N.O.)Provide path to Con-tainment, Spray headers for CS Pumps'rovide path to cold legs of,RCS from high head safety injection N/A same Long term not required to function TABLE 2 MAIN STEAM LINE BREAK BIOCK NO./EQUIPMENT SAFETY FUNCTION/BREAK LOCATION SA'FETY FUNCTION REQUIRED OPERATION TIME INSIDE CV OUTSIDE CV HOV 896)A)B)(NO)
Table    3                                                                                                                 Page 1 Reactor:      GINNA                                                  SYSTElTIC  EVALUATION PROGRAM Tame              ENVIRONMENT                  Qua . Document Equipment Type              Location    Needed  Parameter      Require    Qua          Method    Reference            Comments Solenoid Valve        Area 57  SI Signal  Temp      ('F)  See        Amb.     Experience    23              DBE  Main SLB ASCO/                                     ,Pr (psia)       Comments  Atm.     Experience                    in Turbine Bldg.
MOV 4007)4008 Provide path from RWST of borated water for SI and CS pumps suction Provide path for Aux.Feedwater to Steam Generators same same short-term (to close if need sump recirculaton)
V-4269, V-4270                            RH (%)                     Amb.     Experience                    Fail-Safe (closed)
Short term AOV 5871)5872)5873 AOV 5874, 5875)5876 BAST Level 1 LT 102)106)171)'72 Provide path for cleaning by fan coolers, cooling of cont.Atmosphere Indicate BAST Level for automatic trans-fer of SI Pump suction from BAST to RWST N/A same signal initiation 10/BAST I,evel or~1/2 hour MOV 852A, B Provide path for low head SI to Reactor Vessel same Signal Initiation 10.Manual'G Level Instrumentation LT 470, 471, 472, 473 LT 460, 461, 462)463 Safety Injection Reset Button Determine affected SG same Reset SI signal after same Automatic SI sequenc-ing is complete Short term less than 24 hours 1A, 1B Component Cooling Water Pumps Cooling Water for safeguards equipment same Long term 1A, 1B Containment Spray Pump (If cont.Pressure<30 psig)Containment Pressure N/A and Temperature con-trol Long term MOV 402?, 4028 Provide Service Water to Motor Driven Aux.Feedwater Pumps Suction same within~2 hours Charging pumps Inventory control to RCS same Long term TABLE 2 BLOCK NO./EQUIPHENT HAIN STEAH LINE BREAK SAFETY FUNCTION/BREAK LOCATION SAFETY FUNCTION REQUIRED OPERATION TIHE INSIDE CV OUTSIDE CV Standby AFW pumps HOV 9629A, B MOV 9710A, B;9703A, B;9704A, B HOV 4000A, B Provide flow to SGs if AFW system in-operable Provide SW to suction of Standby AFW Pumps Standby AFW discharge valves to provide AFW flow to SGs AFW Cross-Connect Valves same same same same I,ong term Long term Long term Short term 11.Continued Safe Shutdown Sampling (per THI)Pressurizer PORVs Sample Containment Atmosphere and Reactor Coolant RC Pressure Control same same Long term Long term
LB 8300 B 61 U                            Chem (FW Control Valves)                       Rad.
V-4271, V-4272                            Sub.
LB 8300 B 64 RU (FW Bypass Valves)
                                                      'emp
: 2. Solenoid Valve        Area 52  Minutes              ('F) See        Amb.       Experience  23              These valves were
    'Copes-Vulcan                              Pr (psia)       Comments  Atm.       Experience                    purchased from ASCO.
AOV 836 A,B                               RH  (%)                   Amb.       Experience                    8200 series. They
    .(NaOH   to   CS)                           Chem.                                                              are fail safe Rad:                                                                (open).
Sub.
: 3. Solenoid Valve      Area I3  Seconds    Temp      ('F)   See      250        Vendor Data  25-,             En'closed  in NEMA-2 Lawrence/                                 Pr (psia)        Comments  Atm.                        .'xperience drip-proof enclosure 110114W  -  Supply                      RH (%)                     Amb.      Experience                    which is subjected 125434W  -  Vent                        Chem.                                                              to salt water spray V-3516, V-3517                            Rad.                                                                qualification test.
(Main Steam Isola-                       Sub.                                                                Fail safe (closed) tion)
: 4. Solenoid Valve      Area 51  Seconds    Temp      ('F)   See      200        Vendor Data  26              Fail safe. Per-Versa/VSG                                Pr (psia) Comments        Atm.       Experience                    forms safety V-5871, V-5872,                           RH (%)                     Amb.       Experience                    function within
    ~ V-5873, V-5874,                           Chem.                      Yes                                      seconds of start of
    'V-5875, V-5876                              Rad.                      No                                      DBE. Not required (Containment.Recir-                      Sub.                                                                to operate when culation    System                                                                                            accident conditions Dampers)                                                                                                       are reached.


Accident References LOCA analysis[LOCA]FSAR 2.3.4, 5."ECCS Analysis for the R.E.Ginna Reactor with ENC WREM-2 PWR Evaluation Model" dated December 1977 sub-mitted with Application for Amendement to Operating License, on January 6, 1978.ECCS Analysis submitted by letter dated April 7, 1977 from L.D.White, Jr., RG&E to A.Schwencer, Chief, Operating Reactors Branch Il, USNRC.ECCS Analysis for the R.E.Ginna Reactor with ENC WREM-2 PWR Evaluation Model.Exxon Nuclear Co.Report XN-NF-77-58.
l r
Ginna Emergency Procedures E1.1 and E1.2, submitted by letter dated February 26, 1980 from L.D.White, Jr.RG&E, to D.L.Ziemann, USNRC.Steam Line Break and Feedwater Line Break[SLB/FLB]2.3.5.6'.Steam line break analyses submitted with Application for Amendment to Operating License on September 22, 1975.Plant'Transient.
Analysis for the R.E.Ginna Unit 1 Nuclear Power Plant, Exxon Report XN-NF-77-40 (11/77 and updated 12/15/78 and March, 1980.Letter dated May 24, 1977 from K.W.Amish, RG&E to J.F.-O'eary, NRC.Ginna Emergency Procedures E1.1 and E1.3, submitted by letter dated February 26, 1980 from L.D.White, Jr., RG&E to D.L.Ziemann, USNRC.Letter from L.D.White, Jr., RG&E, to D.L.Ziemann, NRC, March 28, 1980.High Energy Line Break[HELB]"Effects of Postulated'Pipe Breaks Outside the Con-tainment Building", GAI Report No.1815, submitted by letter dated November 1, 1973 from K.W.Amish, RG&E, to A, Giambuso, Deputy Director for Reactor Projects, USNRC.
Letter dated May 24, 1974 from K.W.Amish, RG&E, to J.F.O'eary, Director, Directorate of Licensing, USNRC.Letter dated September 4, 1974 for R.R.Koprowski, RG&E to Edson Case, Acting Director, Directorate of Licensing, USNRC.Letter dated November 1, 1974 from K.W.Amish, RG&E, to Edson Case, Acting Director, Directorate of Li-censing, USNRC.Letter dated May 20, 1977 from L.D.White, Jr., RG&E, to A.Schwencer, Chief Operating Reactors Branch 51, USNRC.Letter dated February 6,'1978 from L.D.White, Jr., RG&E, to A.Schwencer, Chief, Operating Reactors Branch Ol, USNRC.Amendment No.7 to Provisional Operating License DPR-18, transmitted, by letter dated May 14, 1975 from Robert A.Purple, Chief, Operating Reactors Branch-51, USNRC, to L.D.White, Jr , RG&E.Amendment No.29 to Provisional Operating License DPR-18, transmitted by letter dated August 24, 1979 from Dennis L.Ziemann, Chief, ORB 52, to L.D.White, Jr., RG&E.Letter, L.D.White, Jr., RG&E, to D.L.Ziemann, May 17, 1979.Letter, L.D.White, Jr., RG&E, to D.L.Ziemann, USNRC, June 27, 1979.Letter, L.D.White, Jr., RG&E, to D.L.Ziemann, USNRC July 6, 1979.Letter, R.E.Anderson, Gilbert/Commonwealth to James J.Shea, USNRC, June 11, 1979.Letter, L.D.White, Jr., RG&E, to D.M.Crutchfield, NRC, SEP Topic III-5.B,"Pipe Break Outside Containment," August 7, 1980.Letter, J.Wenclawiak and T.Snyder, Catalytic, to G.Wrobel, RG&E,"Equipment Environmental Qualification," October 27, 1980.Letter from D.M.Crutchfield, NRC, to L.D.White, Jr.RG&E, SEP Topic III-S.B,"Pipe Break Outside Containment," June 24, 1980.
Effects of Flooding[Flood]Letter dated May 13, 1975 from L.D.White, Jr., RG&E, to Benard C.Rusche, Director, Office of Nuclear-Reactor Regulation, USNRC.2.3., 5.6.7.8.9.10.Letter dated May 20, 1975 from L.-D'.White, Jr., RG&E, to Robert A.Purple, Chief, Operating Reactors Branch 51, Division of Reactor Licensing.
Letter dated May 30, 1975 from L.D.White, Jr., RG&E, to Robert A.Purple.t Letter dated June 16, 1975 from L.D.White', Jr., RG&E, to Robert A.Purple.Letter dated July 3, 1975 from Robert A.Purple to L.D.White, Jr., RG&E.Letter dated August.8, 1972 from Donald J.Skovholt, Assistant Director for Operating Reactors, USAEC, to Edward J.Nelson, RG&E.Letter dated October 3, 1972 from K.W.Amish, RG&E, to Donald J.Skovholt, Assistant Director for Operating Reactors, USAEC.Letter dated May 31, 1973 from K.W.Amish, RG&E, to Donald J.Skovholt, Assistant, Director for Operating Reactors, USAEC.Application for Amendment to Operating License, sub-mitted March 10, 1975.Amendment, No.14 to Provisional Operating License DPR-18, transmitted by letter dated June 1, 1977 from A.Schwencer, Chief, Operating Reactors Branch 51, USNRC.Letter, L.D.White, Jr.RG&E, to Dennis L.Ziemann, USNRC, High Energy Line Breaks Outside Containment, June 27, 1979.TMI Lessons Learned[TMI]RG&E letter of October 17, 1979, L.D.White, Jr., RG&E, to D.L.Ziemann, USNRC,"TMI Short Term Lessons Learned Requirements." 2.3.RG&E letter of November 19, 1979, L.D.White, Jr.to D.L.Ziemann, USNRC,"TMI Short Term Lessons Learned." RG&E letter of December 28, 1979, L.D.White, Jr.to D.,L.Ziemann, USNRC,"TMI Short Term Lessons Learned."
I I'\,(l Table 3 Page 1 Reactor: GINNA SYSTElTIC EVALUATION PROGRAM Equipment Type Location Tame Needed ENVIRONMENT Parameter Require Qua Qua.Document Method Reference Comments Solenoid Valve ASCO/V-4269, V-4270 LB 8300 B 61 U (FW Control Valves)V-4271, V-4272 LB 8300 B 64 RU (FW Bypass Valves)2.Solenoid Valve'Copes-Vulcan AOV 836 A,B.(NaOH to CS)3.Solenoid Valve Lawrence/110114W-Supply 125434W-Vent V-3516, V-3517 (Main Steam Isola-tion)4.Solenoid Valve Versa/VSG V-5871, V-5872,~V-5873, V-5874,'V-5875, V-5876 (Containment.Recir-culation System Dampers)Area 57 SI Signal Area 52 Minutes Area I3 Seconds Area 51 Seconds Temp ('F),Pr (psia)RH (%)Chem Rad.Sub.'emp ('F)Pr (psia)RH (%)Chem.Rad: Sub.Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.See Comments See Comments See Comments See Comments Amb.Atm.Amb.Amb.Atm.Amb.250 Atm.Amb.200 Atm.Amb.Yes No Experience 23 Experience Experience Experience 23 Experience Experience Vendor Data 25-,.'xperience Experience Vendor Data 26 Experience Experience DBE-Main SLB in Turbine Bldg.Fail-Safe (closed)These valves were purchased from ASCO.8200 series.They are fail safe (open).En'closed in NEMA-2 drip-proof enclosure which is subjected to salt water spray qualification test.Fail safe (closed)Fail safe.Per-forms safety function within seconds of start of DBE.Not required to operate when accident conditions are reached.
l r Table 3~(]]Page 2 Reactor: GINNA Equipment Type Location Tame Needed SYSTEMATIC EVALUATION PROGRAM ENV I RONMENT Qua.Document Parameter Require Qua.Method Reference Comments 5.Solenoid Valve ASCO AOV-897, AOV-898 (SI Recirculation)
Area 42 Short-Term (Before Sump Recirculation)
Temp ('F)Pr (psia)RH (%).Chem.Rad.Sub.See Amb.Comments Atm.Amb.Experience
, 23 Experience Experience"Mild" Envt.to be addressed later 6.Solenoid Valve Versa/Area 51 VSG-3731 Area 53 (Cont.Purge Valves)VSG-3421 (Cont.Depressuriza-tlon)Seconds Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.See 200 Comments Atm.Amb.Vendor data 26 Experience Experience Fail-close to perform con-tainment isola-tion function 7.Control Room Dampers D-81+D-87 8a.Limitorque SMB-2 Reliance Motor MOV 841, 865 (Accumulator Discharge)
Area 41 Not required to operate Temp (oF Pr (psia)RH (%)Chem.Rad.Sub.See 320 Comments 105 100 Yes 2 x 10 No Test Test Test Test Test 18,19 18,19 18, 19 18, 19 18, 19 37 Not Electrical.
Deleted from Report Valves are locked-open with power removed.No need to function.t j 8b.Limitorque SMB-OO, Peerless MOV 826 A,B,C,D (BAST to SI Pumps)MOV 896 A,B (RWST to SI Pumps)Area 52 Short-Term (Before Sump recirculation)
Temp ('F)Amb.Pr (psia)Atm.RH (%)Amb.Chem.No Rad.No Sub.No Amb.Atm.Amb.Experience 13 Experience Experience Not exposed to DBE environment


Table 3 Page 3 Reactor: GINNA Equipment Type SYSTEMATIC EVALUATION PROGRAM Tame ENVIRONMENT Qua.Document Location Needed Parameter Require Qua.Method Reference Comments 8c.Iimitorque SMB-00'Reliance Motor MOV 825 A,B{RWST to SI Pumps)Area 52 Short-Term (Before Sump Recirculation)
                                                                                                              ~(
Temp ('F)Pr (psia)RH (%)Chem.*Rad.Sub.Amb.Atm.Amb.No No No Amb.Atm.Amb.Experience
                                                                                                                                  ]
'3 Experience Exp'erience No exposed to DBE environment Sd.8e.Limitorque SMB-00 Reliance Motor MOV 4007, 4008 (AFW Discharge)
                                                                                                                                  ]
MOV 4027, 4028 (AFW Suction)4000 A,B (AFW Cross-Connect)
Table 3                                                                                                             Page 2 Reactor:   GINNA                                                     SYSTEMATIC EVALUATION PROGRAM Tame             ENV I RONMENT                  Qua .     Document Equipment Type            Location     Needed     Parameter   Require     Qua .       Method       Reference     Comments
Limitorque SMB-00 Reliance V-850 A,B (Sump Valves)MOV 856 (RWST to RHR)V-857 A,B,C (RHR to SI)V-860 A,B,C,D (CS Valves)Area 43 Area 02 Long Short-Term.
: 5. Solenoid Valve        Area 42  Short-Term     Temp ('F)   See        Amb.      Experience  , 23        "Mild" Envt. to ASCO                            (Before Sump  Pr (psia)     Comments  Atm.     Experience               be addressed  later AOV-897, AOV-898                Recirculation) RH (%) .                 Amb.     Experience (SI Recirculation)                           Chem.
Only for DBEs not in area N.See Comment.Temp (4F)Pr (psia)RH (%)Chem.Rad.Sub.Temp ('F)Pr{psia)RH (%)Chem.Rad.Sub.See Comment Amb.Atm.Amb.No 3 x 10 No Amb.Atm.Amb.320 105 100 Yes 2 x 10 Experience Experience Experience Test Test Test Test Test 18,19,53 18ilgi53 18,19,53 18,19,53 18,19,53 Not required to operate in harsh DBE envt.Alter-native SAFW system available.
Rad.
Not exposed to DBE environment except post-LOCA sump water recir-culation 8f.Limitorque SMB-00 MOV-851 A,B Area 51 Not required to operate emp (oF)Pr (psia)RH (%)Chem.Rad.Sub.See Amb.Comment Atm.Amb.No No No Experience 13 Experience Experience Not required to function for DBE.Valves are in locked-open posi-tion as required for SI.
Sub.
Table 3 Page 4 Reactor: GINNA Equipment Type Tame ,Location-Needed SYSTEMATIC EVALUATION PROGRAM ENVIRONMENT Qua.Document Parameter Require Qua.Method Reference Comments g.Limitorque"SMB-00 Peerless Motor MOV 878 B,D (SI to cold legs)8h.Limitorque SMB-1 Reliance Motor MOV 852 A,B (core deluge)Area 51 Not required to operate Area 01 SI Signal Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.Amb.Atm.Amb.286 75 100 Yes 1.6 x 10 No Amb.Atm.Amb.320 105 100 Yes 2 x 10 No Experience
: 6. Solenoid Valve                                Temp ('F)   See        200      Vendor data    26        Fail-close Versa/                Area 51  Seconds        Pr (psia)     Comments  Atm.     Experience               to perform con-VSG-3731              Area 53                RH (%)                   Amb.     Experience              tainment isola-(Cont. Purge Valves)                          Chem.                                                       tion function VSG-3421                                      Rad.
-Experience Experience Test Test Test Test Test 13 18,19 18,19 18,19 18,19 18,19 37 Not required to function for DBE.Valves are locked in open position, as needed for SI.Valve completes safety function (to open)early into accident 8i.Limitorque SMB-00 Reliance Motor MOV 9703 A,B;9704 A, B;9710 A, B (Standby AFW System)9.Motor, Pump General Electric (Standby AFW)Area 46 Long Term Area 86 Long Term Temp (4F)120 Pr (psia)Atm.RH (%)Amb.Chem.No Rad.No Sub.No Temp ('F)120 Pr (psia)Atm.RH (%)Amb.Chem.No Rad.No Sub.No 120 Atm.Amb.122 Atm.Amb.Vendor Data Experience Experience Vendor Data Experience Experience 43,47,54 2,3,43,47 Standby AFW System located in con-trolled envt.Standby AFW pumps located in aux.bldg.annex which has controlled envt.1Q.Motor, Pump Westinghouse 444 TS TBDP 445 TS TBDP (Containment Spray, RHR, Component Cooling)Area 52 Long Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.Amb.Atm.Amb.No 3 x 10 No 104 F Atm.Amb.1 x 10 Spec Experience Experience Test 15,16,67 Only DBE environ-ment is post-accident radiation 69 Table 3 Page 5 Reactor: GINNA SYSTEMATIC EVALUATION PROGRAM Equipment Type Location Tame Needed ENVIRONMENT Parameter Require Qua.Qua Method Document Reference Comments ll.Motor, Pump Westinghouse 505 US ABDP (Auxiliary Feed-water)Area ()3 Long Temp ('F)See Pr (psia)Comment RH (%)Chem.Rad.Sub.1040F Atm.Amb.2 x 10 Spec Experience Experience Test 8,16,67 68 Have installed totally redundant system not exposed to DBE (standby AFW)12a.Motor, Pump Westinghouse 509 US AFDP (Safety Injection) 12b.Motor, Pump 509 UPH ABDP (Service Water)Area C3 Long Area N5 Long Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.Amb.Atm.Amb.No 3xlo No Amb.Atm.Amb.No No No 104oF Atm.Amb.2 x 10 See Comment Spec Experience Experience Test Experience Experience Experience 15,16,67 68 67 Only DBE environ-ment is post-accident radiation This item is in a"mild" environ-ment.It will be addressed later.13a.Penetrations, Electrical Crouse-Hinds Area 41 Long Temp ('F)286 F Pr (psia)75 RH (%)100%Chem.Yes Rad.1.6xl0 Sub.No 340oF 105 100%Yes 1.17x10 Test Test Test Test Test 1,45,54,58 1,4S,S4,S8 1,45,54,58 58 45,64 Radiation level at location of pene-trltions<1.6 x 10 rads.Qualifi-fication test is greater than DOR guidelines value of 2 x 10 rads.13b.Penetrations, Electrical Westinghouse Area Nl Long Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.286oF 75 100%es 8 1.6x10 No 340oF 75 100%s 8 2.1x10 Test Test Test Test 29,30,59 29,30,59 29,30,59 29,30,59
(Cont. Depressuriza-                         Sub.
tlon)
: 7. Control  Room Dampers                                                                                    Not Electrical.
D-81 + D-87                                                                                                Deleted from Report 8a. Limitorque             Area 41  Not required   Temp (oF      See        320       Test         18,19     Valves are locked-SMB-2                          to operate    Pr (psia)     Comments  105        Test          18,19    open with power Reliance Motor                                RH (%)                   100        Test          18, 19    removed. No need MOV  841, 865                                Chem.                   Yes        Test          18, 19    to function        .t (Accumulator                                  Rad.                     2 x 10    Test          18, 19 Discharge)                                   Sub.                     No                       37 j
8b. Limitorque            Area 52  Short-Term SMB-OO, Peerless              (Before Sump  Temp ('F)   Amb.       Amb.       Experience   13        Not exposed to MOV  826 A,B,C,D              recirculation) Pr (psia)     Atm.       Atm.       Experience              DBE environment (BAST to SI Pumps)                           RH (%)       Amb.       Amb.       Experience MOV 896 A,B                                  Chem.         No (RWST to SI Pumps)                           Rad.         No Sub.         No


Table 3 Page 6 Reactor: GINNA Equipment Type Location Tame Needed SYSTEMATIC EVALUATION PROGRAM ENVIRONMENT Qua.Document Parameter Require Qua.Method Reference Comments 14.Terminal Block Westinghouse 542247 Area 51 Long Temp ('F)Pr (psia)RH (%)chem.Rad.Sub.286oF 75 100%o es 8 1.6x10 No 3400F 121 100%Yes 7 2x10 Test Test Test Test Test 50 50 50 50 60 Location of blocks7is such that 2 x 10 rads, a value equal to the DOR guidelines value, should be acceptable.
Table 3                                                                                                           Page 3 Reactor:   GINNA                                                   SYSTEMATIC EVALUATION PROGRAM Tame              ENVIRONMENT                 Qua . Document Equipment Type            Location    Needed    Parameter   Require   Qua .       Method     Reference     Comments 8c. Iimitorque SMB-00 Area 52  Short-Term (Before Sump Temp ('F)
Also, terminal blocks will be elevated.15a.Cable Kerite HT Area Il Long Pr (psia)RH (%)Chem.Rad.Sub.75 100%es 8 1.6xlO No Temp (oF)286 F 340oF 118 100%Yes 8 2xlO Test Test Test Test Test 11,38,51, 55,63 15b.Cable Kerite HT All Long Pr (psia)RH (%)Chem.Rad.Sub.15.8 100 No No No Temp (oF)220oF 340oF 118 100 Yes 8 2x10 Test Test Test Test Test 11,38,51, 55,63 16.Cable Coleman Cable Area Nl Long Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.286 75 100 Yes 1.6xlo No 340 118 100 es 8 2xlO Test Test Test Test Test 46, 51 46,51 46,51 46,51 46,51 Table 3 Page 7 Reactor: GINNA SYSTEMATIC EVALUATION PROGRAM Equipment Type Location Tame Needed ENVIRONMENT Parameter Require Qua.Qua.Method Document Reference Comments 17.Cable Coleman Cable Rome Cable General Cable/18.Transmitter, Level Foxboro (RWST Level)All Long Area N2 Short Term (Before Sump Recirculation)
Pr (psia)
Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.220 15.8 100 No No No Amb.Atm.Amb.No No No 250 Atm.Amb.Amb.Atm.Amb.Test Experience Experience Experience Experience.
Amb.
Experience 5,10,46 In lieu of 100/RH, an owl zmmersxon test performed per IPCEA S-61-402 Not exposed to DBE when required to to function 19.Transmitter, Level Area 42 Short Term Barton 289 (Before Sump (RWST Level)Recirculation) 20.Transmitter, Flow Area 51 Seconds Barton 332 (Steam Flow)Temp (oF)Pr (psia)RH (%)Chem.Rad.Sub.Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.Amb.Atm.Amb.No No No 286 75 100 Yes 1.6x10 No 200 Atm.Amb.See Comments Vendor Data Experience Experience See Comments 34 31 Not exposed to DBE envt.when required to function.Not exposed to to DBE when required to function.21.Transmitter, Pres.Areas 2,3 Long , Barton 332 (Cont.Pressure)Temp (oF)Pr (psia)RH (%)Chem.Rad.Sub.Amb.Atm.Amb.No No No See Comments See Comments 31 Not exposed to DBE when required to function.  
Atm.
Amb.
Atm.
Experience Experience
                                                                                                '3          No exposed DBE to environment
    'Reliance Motor                Recirculation) RH (%)       Amb.      Amb.      Exp'erience
* MOV  825 A,B                                Chem.       No
{RWST  to SI Pumps)                          Rad.         No Sub.         No Sd. Limitorque            Area 43  Short-Term. Temp (4F)   See        Amb.     Experience              Not required to SMB-00                        Only for DBEs  Pr (psia)   Comment    Atm.      Experience              operate in harsh Reliance Motor                not in area N. RH (%)                 Amb.      Experience              DBE  envt. Alter-MOV  4007, 4008              See Comment. Chem.                                                     native SAFW (AFW  Discharge)                            Rad.                                                     system available.
MOV  4027, 4028                              Sub.
(AFW  Suction) 4000 A,B (AFW  Cross-Connect) 8e. Limitorque            Area 02  Long           Temp ('F)   Amb.      320      Test          18,19,53  Not exposed to SMB-00                                      Pr {psia)   Atm.      105      Test          18ilgi53  DBE environment Reliance                                    RH (%)       Amb.       100       Test         18,19,53  except post-LOCA V-850 A,B (Sump                              Chem.       No        Yes      Test          18,19,53  sump water recir-Valves)                                                                                                culation MOV  856 (RWST to                          Rad.         3  x 10    2 x 10    Test          18,19,53 RHR)
V-857 A,B,C (RHR                            Sub.         No to SI)
V-860 A,B,C,D (CS Valves) 8f. Limitorque            Area 51 Not required    emp (oF)     See      Amb.       Experience  13        Not required to SMB-00                        to operate    Pr (psia)     Comment  Atm.       Experience             function for    DBE.
MOV-851 A,B                                  RH (%)                 Amb.      Experience            Valves are in Chem.                   No                                locked-open posi-Rad.                   No                                tion as required Sub.                   No                               for SI.


Table 3 pPage 8 Reactor: GINNA SYSTEMATIC EVALUATION PROGRAM Equipment Type Location Tame Needed ENVI RONMENT Qua.Document Parameter Require Qua.Method Reference Comments 22.Transmitter, Pressure Foxboro 611 GM-DSI~(PRZR Pressure)23.Transmitter, Pressure Foxboro 611 GM-DSI (Steam Pressure)24.Transmitter, Level Foxboro 613 M-MDL Modified (Przr Level)Area 41 Short Area 43 Short Area 51 Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.Temp ('F)Pr (psia)RH(%)Chem.Rad.Sub.Temp ('F)Pr (psia)RH (%)Chem.Sub.286 75 100 Yes l.7xl0 No See Comments See Comments 286 75 100 Yes<3x10 See Comments See Comments Test Test Test Test Evaluation See Comments See Comments 18,19,33 18,19,33 18,19,33 18,19,33 18,19 18,19 18,19 18,19 18,19 18,19 Adequate for short-term function.Will be replaced and elevated to perform post-accident monitoring function Not exposed to DBE when required to function Not required for a short-term safety function.Will be replaced for long-term monitoring 25.Transmitter, Level Area 52 Sort Foxboro 613 DM-MSI (BAST Level)26.Transmitter, Level Area 51 Foxboro 613 (SG Level)Temp (4F)Pr (psia)RH (%)Chem.Rad.Sub.Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.Amb.Atm.Amb.No No No See Comments Amb.Atm.Amb.See Comments Experience Experience Experience See Comments Not exposed to DBE Alternative instrumentation available to per-form safety function.Will be replaced for long-term monitoring.
Table   3                                                                                                         Page 4 Reactor:   GINNA                                                 SYSTEMATIC EVALUATION PROGRAM Tame            ENVIRONMENT                  Qua . Document Equipment Type           ,Location   - Needed   Parameter   Require     Qua .       Method     Reference   Comments
II Table 3 Page 9 Reactor: GINNA Equipment Type Location Tame Needed SYSTEMATIC EVALUATION PROGRAM ENVIRONMENT qua.Document Parameter Require Qua.Method Reference Comments 27.Temp Element Rosemount/176JA (,RTDs)28.Battery Gould/FTA-19 Area&#xb9;1 Area&#xb9;8 Long Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.See Comments Amb.Atm.Amb.No No No 200 Atm.Amb.200 R/hr 110 Atm.Amb.Spec 35 Experience Experience Spec 35 Vendor Data 9,32 Experience Experience Not required to function for short-term DBE.Will be replaced for long-term monitoring Not exposed to DBE 29a.Diesel Generator Area&#xb9;4 Long ALCO Diesel 251F b.Westinghouse 1900 KW Generator c.Westinghouse fuel oil transfer pump-1 HP-model TEFC Class PMF Insulation Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.Amb.Atm.Amb.No No No Amb.Atm.Amb.Experience 7 Experience Experience Not exposed to DBE 30.Motor, Containment Area&#xb9;1 Long Fan Coolers Westinghouse 588.5-CSP Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.286 75 100 Yes 1.6x10 No 320 95 100 Yes 8 2xlo Test Test Test Test Test 18,19,20, 64,65, 67,70 31.Circuit Breaker Westinghouse DB-50A 1600A Area&#xb9;3 Seconds Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.See Comments Amb.Atm.Amb.Experience Experience Experience Equipment will fail-safe on loss of power Table 3 Page 10 Reactor: GINNA SYSTEMATIC EVALUATION PROGRAM Equipment Type 32.IRC Cabinets Foxboro Location Tame Needed Area 08 Long Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.Amb.Atm.Amb.No No No Amb.Atm.Amb.ENV I RONMENT Parameter Require Qua.qua.Method Experience Experience Experience Document Reference Comments Not exposed to DBE 33.HVAC Westinghouse 2162{Control Room AHU)Area 58 Long Temp ('F)Pr (psia)(%)Chem.Rad.Sub.Amb.Atm.Amb.No No No 122 Atm.Amb.Spec 4,6 Experience Experience Not exposed to DBE 34.Splice Sleeves Area 51 Long Temp (4F)286 340 Test 36,38,51 56,62 Raychem WCSF-N 35.Solenoids/
: g. Limitorque            Area 51 Not required Temp ('F) Amb.        Amb.      Experience  13        Not required to "SMB-00                        to operate  Pr (psia)   Atm.        Atm.    - Experience            function for DBE.
Valcor V57300 (Pressurizer PORVs),'36.Level Switches GEM Corp.Model:Special-Similar to LS-1900 (Containment Sump"B" Level)Area Ol Long Area 41 Pr (psia)RH{%)Chem.Rad.Sub.Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.75 100 Yes 1.6x10 No 286 75 100 Yes 1.6x10 No See Comments 118 100 es 8 2x10 346 128 100 Yes 8 2x10 See Comments Test Test Test Test Test Test Test Test 52 Not required to perform safety function.How-will be replaced for TMI-STLL
Peerless Motor                              RH (%)     Amb.       Amb.       Experience            Valves are locked MOV  878 B,D                                Chem.                                                     in open position, (SI to cold legs)                           Rad.                                                     as needed  for SI.
Sub.
8h. Limitorque            Area 01  SI Signal    Temp ('F) 286        320        Test        18,19    Valve completes SMB-1                                      Pr (psia)     75        105        Test        18,19    safety function Reliance Motor                              RH (%)     100         100       Test         18,19     (to open) early MOV  852 A,B                                Chem.      Yes        Yes        Test        18,19     into accident (core deluge)                              Rad.        1.6 x 10    2 x 10    Test        18,19 Sub.       No          No                      37 8i. Limitorque            Area 46  Long Term    Temp (4F) 120        120        Vendor Data  43,47,54  Standby  AFW  System SMB-00                                      Pr (psia)   Atm.       Atm.       Experience            located in con-Reliance Motor                              RH (%)     Amb.       Amb.       Experience             trolled envt.
MOV 9703 A,B;                               Chem.       No 9704 A, B; 9710 A, B                        Rad.       No (Standby AFW System)                       Sub.       No
: 9. Motor, Pump          Area 86  Long Term    Temp ('F)   120        122        Vendor Data  2,3,43,47 Standby  AFW pumps General Electric                            Pr (psia)   Atm.       Atm.       Experience             located in aux.
(Standby AFW)                               RH (%)     Amb.       Amb.       Experience            bldg. annex which Chem.       No                                           has controlled Rad.        No                                           envt.
Sub.       No 1Q. Motor,  Pump        Area 52  Long         Temp ('F)   Amb.       104 F      Spec         15,16,67  Only  DBE environ-Westinghouse                                Pr (psia)   Atm.       Atm.       Experience            ment  is post-444 TS TBDP                                RH (%)       Amb.       Amb.       Experience            accident radiation 445 TS TBDP                                Chem.       No (Containment Spray,                        Rad.         3 x 10    1  x 10    Test        69 RHR, Component                              Sub.         No Cooling)


c, Table 3 Page ll Reactor: GINNA SYSTEMATIC EVALUATION PROGRAM Eguipment Type Location T1me Needed ENVIRONMENT Parameter Requ1re Qua.Qua Method Document Reference Comments 37.H2 Recombiner Area 41 Igniter Exciter Unit GLA Part No.43737, Rev.A, Serial 001 Long Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.286 75 100 Yes 1.6xlo No 315 105 100 Yes 1.73x10 Test Test Test Test Test 18,19,49 18,19,49 18,19,49 18,19,49 18,19,49 38.39.40.41.I H2 Recombiner Blower Motor (2/15 Scale)W 2 HP, Class H Ins., Model TBFC SO 68C24196 Pump Motor U.S.Electrical Motors Model VEU, 100 HP Frame 84-445 U Insulation Class B (Charging Pump)Solenoids/
Table 3 Page 5 Reactor:   GINNA                                             SYSTEMATIC EVALUATION PROGRAM Tame          ENVIRONMENT                  Qua      Document Equipment Type         Location     Needed Parameter   Require    Qua .       Method     Reference     Comments ll. Motor,  Pump Westinghouse Area ()3 Long       Temp ('F) See        1040F    Spec          8,16,67    Have  installed 505 US ABDP Pr (psia)   Comment    Atm.      Experience              totally  redundant (Auxiliary Feed-                        RH (%)                 Amb.      Experience              system not exposed water)
Johnson Controls Model D251 (Control Room Air Handling Unit Dampers)Medium Voltage Switchgear Westinghouse DH-350E 1200 A Breakers (RCP Trip Breakers)Area 51 Long Area N2 Long Area 58 Short Area 07 Short Temp (OF Pr (psia)RH (%)Chem.Rad.Sub.Temp (OF)Pr (psia)RH (%)Chem.Rad.Sub.Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.286 75 100 Yes 1.6xl0 No Amb.Atm.Amb.No No No Amb.Atm.Amb.No No No Amb.Atm.Amb.No No No 286 75 100 Yes 2.0x10 No Amb.Atm.Amb.Amb.Atm.Amb.Amb.Atm.Amb.Te'st Test Test Test Test Experience Experience Experience Experience Experience Experience Experience Experience Experience 18,19,49 18,19,49 18,19,49 18,19,49 18, 19,49 Not exposed to DBE environment Not exposed to DBE environment Breakers need only open for LOCA inside containment to stop RC pumps.Not exposed to DBE when needed to function.cc
Chem.                                                     to DBE  (standby Rad.                   2  x 10  Test          68        AFW)
Sub.
12a. Motor, Pump        Area C3  Long              ('F)             104oF Westinghouse Temp        Amb.                 Spec          15,16,67  Only  DBE  environ-Pr (psia)   Atm.       Atm.     Experience              ment  is post-509 US AFDP (Safety Injection)
RH (%)     Amb.      Amb.      Experience              accident radiation Chem.       No Rad.
Sub.
3xlo No 2  x 10  Test          68 12b. Motor, Pump        Area    Long              ('F) 509 UPH ABDP N5              Temp        Amb.      See      Experience    67        This item is in    a Pr (psia)   Atm.      Comment  Experience              "mild" environ-(Service Water)                        RH (%)     Amb.                Experience              ment. It will be Chem.
Rad.
No No addressed  later.
Sub.       No 13a. Penetrations,      Area 41  Long              ('F)
Electrical                              Temp        286 F      340oF    Test          1,45,54,58 Radiation level at Pr (psia)   75        105      Test          1,4S,S4,S8 location of pene-Crouse-Hinds                            RH (%)     100%      100%      Test          1,45,54,58 trltions < 1.6 x Chem.       Yes      Yes        Test          58        10 rads. Qualifi-Rad.       1.6xl0     1.17x10  Test          45,64      fication test is Sub.       No                                           greater than DOR guidelines value of 2 x 10 rads.
13b. Penetrations,      Area Nl  Long        Temp  ('F)  286oF    340oF      Test Electrical                                                                              29,30,59 Pr (psia)  75        75        Test          29,30,59 Westinghouse                            RH (%)      100%      100%      Test          29,30,59 Chem.        es          s      Test          29,30,59 Rad.        1.6x10 8  2.1x10 8 Sub.       No


Table 3 Page 12 Reactor: GINNA SYSTEMATIC EVALUATION PROGRAM Equipment Type Location Tame Needed ENVIRONMENT Qua-Document, Parameter Require Qua.Method Reference Comments 42.RHR Pump Cooling System Fan Motors Westinghouse Model SBDP Class B Insulation-2HP Area 02 Long Temp ('F)Pr (psia)RH (%)Chem.Rad.Sub.Amb.Atm.3xlO No No Amb.Atm.Amb.7 lx10 Experience Experience Experience Test 69 Only exposed to DBE radiation environment 43.Cont Spray/SI Pump and Charging Pump Cooling Systems Fan Motors Westinghouse Model SBDP Class B Insulation-3HP 44.Main Control Board Reactor Trip Racks Relay Logic and Test Racks Miscellaneous Racks Auxiliary Relay Racks Safeguard Racks Reactor Coolant System Racks CVCS Racks Feedwater Control Racks SI Sequence Racks Area 52 Long Area N2 Long Temp ('F)Amb.Pr (psia)Atm.RH (%)Amb.6 Chem.3x10 Rad.No Sub.No See Comments Amb.Atm.Amb.>1x10 Experience Experience Experience Test 69 Only exposed to DBE radiation environment"Mild" Environment.
Table 3                                                                                                 Page 6 Reactor:   GINNA                                         SYSTEMATIC EVALUATION PROGRAM Tame          ENVIRONMENT                  Qua . Document Equipment Type     Location     Needed Parameter   Require     Qua .       Method     Reference   Comments
be addressed at a later time C I Table 4 Environmental Service Conditions Inside Containment Normal 0 eration Temperature:
: 14. Terminal Block Area 51  Long       Temp ('F) 286oF      3400F    Test          50        Location of Westinghouse                        Pr (psia)   75          121      Test          50        blocks7is such that 542247                              RH (%)     100%o      100%      Test          50        2 x 10 rads, a chem.       es        Yes 7    Test          50        value equal to the Rad.       1.6x10 8    2x10      Test          60        DOR guidelines Sub.       No                                           value, should be acceptable. Also, terminal blocks will be elevated.
Pressure: Humidity: Radiation:
15a. Cable        Area Il  Long       Temp (oF)  286 F       340oF    Test          11,38,51, 55,63 Kerite                              Pr (psia)   75          118      Test HT                                  RH (%)     100%        100%      Test Chem.       es    8 Yes 8    Test Rad.       1.6xlO      2xlO      Test Sub.       No 15b. Cable        All      Long        Temp  (oF)  220oF      340oF    Test          11,38,51, 55,63 Kerite                              Pr (psia)  15.8        118      Test HT                                  RH (%)     100        100      Test Chem.      No          Yes 8    Test Rad.       No          2x10      Test Sub.       No
60-120 F 0 psig 50%(nominal)1 Rad/hr general.Can be higher or lower near specific components.
: 16. Cable          Area Nl  Long        Temp  ('F) 286        340      Test          46, 51 Coleman Cable                      Pr (psia)   75          118      Test          46,51 RH (%)     100        100      Test          46,51 Chem.      Yes          es      Test          46,51 Rad.       1.6xlo      2xlO 8   Test          46,51 Sub.       No
Temperature:
Pressure: Humidity: Radiation:
Chem.Spray: Flooding: Auxiliar Buildin Normal 0 eration Figur'e 5 (286'F max)Figure 4 (60 psig design)100%Figure 6 (1.6 x 10 total)Solution of boric acid (2000 to 3000 ppm boron)plus NaOH in water.Solution pH between 8 and 10.7 ft (approx)Temperature:
Pressure: Humidity: Radiation:
50-104 F 0 psig, 60%(nominal)10 mr/hr general, with areas near RHR piping<100 mr/hr during shutdown operation Accident Conditions includin sum recirculation Temperature:
Pressure: Humidity: Radiation:
Spray: Flooding: 50-104'F (122'F near motors)0 psig 60%(nominal)Operating Floor (271'lev.):
Near Bus 14 and NCC 1C 6 1L: 100 rad Other Areas: less than 50 rad Intermediate Floor (253'lev.):
Near Bus 16 and MCC 1D 8 1N: 900 rad Other Areas: less than 500 rad Basement Floor (236'lev.):
Near CS, RHR, an(SI Pumps: 2.8 x 10 pads Other areas:<10 rads N/A N/A


C.Intermediate Buildin Normal 0 eratzon Temperature:
Table  3 Page 7 Reactor:    GINNA                                                SYSTEMATIC EVALUATION PROGRAM Tame            ENVIRONMENT                  Qua . Document Equipment Type        Location      Needed    Parameter  Require    Qua .       Method      Reference    Comments
Pressure: Humidity: Radiation:
: 17. Cable              All      Long          Temp  ('F)  220        250      Test          5,10,46  In lieu of 100/  RH, Coleman Cable                              Pr (psia)   15.8      Atm.      Experience              an owl zmmersxon Rome  Cable                                RH (%)     100        Amb.      Experience              test performed per General Cable                              Chem.      No                                          IPCEA S-61-402
50-104'F 0 psig 60%(nominal)1 mr/hr (higher near reactor coolant sampling lines)Accident Condition Based u on HELB or MELB Temperature:
        /                                      Rad.        No Sub.        No
Pressure: Humidity: Radiation:
: 18. Transmitter, Level Area N2  Short Term    Temp  ('F)  Amb.      Amb.      Experience              Not exposed to   DBE Foxboro                    (Before Sump  Pr (psia)  Atm.      Atm.      Experience.             when  required to (RWST  Level)              Recirculation) RH (%)      Amb.      Amb.      Experience              to function Chem.      No Rad.        No Sub.        No
Spray: Flooding: 215'F for 30 minutes;then reducing to 104 within 3 hrs 0.8 psig for 30 minutes;then reducing to O,psig within 3 hrs 100%indefinitely N/A N/A 0 Based u on LOCA conditions Temperature:
: 19. Transmitter, Level Area 42  Short Term    Temp ( oF)  Amb.      200      Vendor Data  34        Not exposed to Barton 289                  (Before Sump  Pr (psia)  Atm.      Atm.      Experience              DBE  envt. when (RWST Level)                Recirculation) RH (%)      Amb.      Amb.      Experience              required to Chem.      No                                          function.
Pressure: Humidity: Radiation:
Rad.        No Sub.       No
Spray: Flooding: D.Cable Tunnel 115'F indefinitely*
: 20. Transmitter, Flow  Area 51  Seconds        Temp  ('F)  286        See      See          31        Not exposed to Barton 332                                Pr (psia)  75        Comments  Comments                to DBE when (Steam Flow)                              RH (%)      100                                         required to Chem.      Yes                                          function.
near large motors and FW and SL piping.104'F in open areas 0 psig 100%Negligible N/A 0 E.Same as Intermediate Control Buildin Control Room Normal 0 eration Building Temperature:
Rad.        1.6x10 Sub.        No
Pressure: Humidity: Radiation:
: 21. Transmitter, Pres. Areas 2,3 Long          Temp (oF)   Amb.      See      See          31        Not exposed to
Accident Conditions Temperature:
  , Barton 332                                Pr (psia)   Atm.       Comments  Comments                DBE when  required (Cont. Pressure)                           RH (%)     Amb.                                        to function.
Pressure: Humidity: Radiation:
Chem.      No Rad.        No Sub.        No
Spray: Flooding: 50-104'F (usually 70-78'F)0 psig 60%(nominal)Negligible 104oF 0.psig 60%(nominal)Negligible N/A N/A*Estimated (no explicit calculations performed)  


~1 Normal 0 eration Temperature:
Table  3                                                                                                    pPage  8 Reactor:     GINNA                                              SYSTEMATIC EVALUATION PROGRAM Tame          ENVIRONMENT                  Qua . Document Equipment Type            Location      Needed Parameter  Require    Qua .      Method      Reference    Comments
Pressure: Humidity: Radiation:
: 22. Transmitter,          Area 41  Short      Temp  ('F)  286        286      Test          18,19,33  Adequate  for short-Pressure                                  Pr (psia)   75        75        Test          18,19,33  term function. Will Foxboro                                  RH (%)     100        100      Test          18,19,33  be replaced and 611 GM-DSI                                Chem.      Yes        Yes <    Test          18,19,33  elevated to perform
J Accident Conditions Temperature:
    ~
Pressure: Humidity.Radiation:
(PRZR  Pressure)                          Rad.        l. 7xl0    3x10      Evaluation    18,19    post-accident Sub.        No                                          monitoring function
Spray: Flooding: 50-104 F 0 psig 60%(nominal)Negligible 104 F 0 psig 60%(nominal)Negligible N/A N/A Normal 0 eration Temperature:
: 23. Transmitter,          Area 43  Short      Temp  ('F)  See        See      See          18,19    Not exposed to Pressure                                  Pr (psia)  Comments  Comments  Comments      18,19    DBE when required Foxboro                                  RH(%)                                          18,19    to function 611 GM-DSI                                Chem.                                          18,19 (Steam Pressure)                         Rad.                                          18,19 Sub.
Pressure: Humidity: Radiation:
: 24. Transmitter,          Area 51              Temp  ('F)  See        See      See                    Not required  for Level                                    Pr (psia)  Comments  Comments  Comments                a  short-term Foxboro                                  RH (%)                                                  safety function.
Accident Conditions Temperature:
613 M-MDL  Modified                      Chem.                                                    Will be replaced (Przr Level)                             Sub.                                                    for long-term monitoring
Pressure: Humidity: Radiation:
: 25. Transmitter, Level    Area 52  Sort        Temp  (4F)  Amb.      Amb.      Experience              Not exposed to Foxboro                                  Pr (psia)  Atm.      Atm.      Experience              DBE 613 DM-MSI                                RH (%)      Amb.      Amb.      Experience (BAST Level)                             Chem.      No Rad.        No Sub.        No
Spray Flooding: 50-104 F 0 psig 60%(nominal)Negligible
: 26. Transmitter, Level    Area 51              Temp  ('F)  See        See      See                    Alternative Foxboro 613                              Pr (psia)  Comments  Comments  Comments                instrumentation (SG  Level)                              RH (%)                                                   available to per-Chem.                                                   form safety Rad.                                                    function. Will be Sub.                                                     replaced for long-term monitoring.
<104'F 0 psig 60%(nominal)Negligible N/A N/A Necbanical E i ment Room Normal 0 eratzon Temperature:
Pressure: Humidity: Radiation:
Accident, Conditions Temperature:
Pressure: Humidity: Radiation:
Spray: Flooding: 50-104 F 0 psig 60%(nominal)Negligible
<104'F 0 psig 60%(nominal).Negligible None 3 ft.(estimated for a service water line leak)


F.Diesel Generator Rooms Normal 0 eratxon Temperature:
II Table  3                                                                                                        Page 9 Reactor:      GINNA                                              SYSTEMATIC EVALUATION PROGRAM Tame          ENVIRONMENT                  qua . Document Equipment Type              Location    Needed Parameter    Require    Qua .        Method    Reference    Comments
Pressure: Humidity: Radiation:
: 27. Temp Element            Area &#xb9;1            Temp  ('F)  See        200      Spec          35        Not required to Rosemount 176JA
Accident Conditions 60-104 F 0 psig 60%(nominal)Negligible Temperature:
                  /                            Pr (psia)
Pressure: Humidity: Radiation:
RH (%)
Spray: Flooding: G.Turbine Buildin Normal 0 eration 104 F 0 psig 90%(estimated)
Comments  Atm.
Negligible N/A 0 ft**Temperature:
Amb.
Pressure: Humidity: Radiation:
Experience Experience function for short-term DBE. Will be
Accident Conditions 50-104 F 0 psig 60%(nominal)Negligible Temperature:
(,RTDs )                                  Chem.                                                    replaced for long-Rad.                    200  R/hr  Spec        35        term monitoring Sub.
Pressure: Humidity: Radiation:
: 28. Battery                Area &#xb9;8  Long      Temp  ('F)  Amb.      110      Vendor Data  9,32      Not exposed Gould/FTA-19                              Pr (psia)    Atm.      Atm.      Experience              to  DBE RH (%)      Amb.      Amb.      Experience Chem.        No Rad.        No Sub.        No 29a. Diesel Generator       Area &#xb9;4  Long      Temp  ('F)  Amb.      Amb.      Experience  7        Not exposed to ALCO  Diesel                              Pr (psia)    Atm.      Atm.      Experience            DBE 251F                                      RH (%)      Amb.      Amb.      Experience
Spray: Flooding: H.Auxiliar Buildin Annex Normal 0 eratzon 220'F'or 30 minutes, reduce to 100'F within 3 hrs.1.14 psig on mezzanine and basement levels, 0.7 psig on operating floor 100%Negligible N/A 18'~in basement (Circ.Water Break)Temperature:
: b. Westinghouse 1900    KW                    Chem.        No Generator                                  Rad.        No
Pressure: Humidity: Radiation:
: c. Westinghouse fuel  oil                  Sub.          No transfer pump -  1  HP-model TEFC Class  PMF Insulation
Accident.'Conditions 60-120 F 0 psig 60%(nominal)Negligible Temperature:
: 30. Motor, Containment      Area &#xb9;1  Long      Temp  ('F)  286      320        Test        18,19,20, Fan Coolers                                Pr (psia)    75        95        Test        64,65, Westinghouse                              RH (%)        100      100        Test        67,70 588.5-CSP                                  Chem.        Yes      Yes 8      Test Rad.          1.6x10    2xlo      Test Sub.        No
Pressure: Humidity: Radiation:
: 31. Circuit Breaker        Area &#xb9;3  Seconds    Temp  ('F)  See      Amb.      Experience              Equipment  will Westinghouse                              Pr (psia)    Comments  Atm.      Experience              fail-safe on DB-50A 1600A                                RH (%)                Amb.      Experience              loss of power Chem.
Spray: Flooding: 60-120 F.0 psig 60%(normal)Negligible N/A 2 ft.**Service water line crack would affect only one room (see FEOOD-15)
Rad.
Screenhouse Normal 0 eration Temperature:
Sub.
Pressure: Humidity: Radiation:
 
Accident Conditions:
Table  3                                                                                                      Page 10 Reactor:  GINNA                                                SYSTEMATIC EVALUATION PROGRAM Tame            I ENV RONMENT                  qua . Document Equipment Type          Location      Needed Parameter    Require    Qua .      Method      Reference    Comments
50-104 F 0 psig 60%(nominal)Negligible Temperature:
: 32. IRC Cabinets        Area 08  Long        Temp  ('F)  Amb.      Amb.      Experience              Not exposed Foxboro                                  Pr (psia)    Atm.      Atm.      Experience              to DBE RH (%)      Amb.      Amb.      Experience Chem.        No Rad.        No Sub.        No
Pressure: Humidity: Radiation:
: 33. HVAC                Area 58  Long        Temp  ('F)  Amb.      122      Spec          4,6      Not exposed to Westinghouse                              Pr (psia)    Atm.      Atm.      Experience              DBE 2162                                          (%)      Amb.      Amb.      Experience
Spray: Flooding:<104 F 0 psig 60%(nominal)Negligible N/A 18" (Circ.Water Break)
{Control  Room AHU)                      Chem.        No Rad.        No Sub.        No
Deeda Basf.s Accident Temperature
: 34. Splice Sleeves      Area 51  Long        Temp  (4F)  286        340      Test          36,38,51 56,62 Raychem                                  Pr (psia)    75        118      Test WCSF-N                                    RH {%)      100        100      Test Chem.        Yes        es      Test Rad.        1.6x10    2x10 8 Sub.        No
-.Time Curve$000I I I I 5$0--150~Containment Temperature o~~Sump Temperature
: 35. Solenoids/          Area Ol  Long        Temp  ('F)  286        346      Test Valcor V57300                            Pr (psia)    75        128      Test (Pressurizer PORVs)                      RH (%)      100        100      Test Chem.        Yes        Yes 8    Test Rad.        1.6x10    2x10      Test Sub.        No
~..'l o Ii.l o I~~~~I I~I ,.~)I~~o~'I I I Heat Exchanger Outlet I.Temperature
,'36. Level Switches      Area 41              Temp  ('F)  See        See                    52        Not required to GEM Corp.                                Pr (psia)    Comments  Comments                          perform safety Model:Special-                            RH (%)                                                    function. How-Similar to LS-1900                        Chem.                                                    will be  replaced (Containment "B" Level)
~~H I o I I,'.-~I I~~I~)I~~o I I~I*I I'I~~~~l.I.~~.'II'~'I t.l~.I'~~~~i r~~r I~~*\I~~~~t oo l I~I~I I.'..--i;:>>~I-:~
Sump                      Rad.                                                      for  TMI-STLL Sub.
f--1O O I'"'I'-'I"-I-l-j..I~I~.,I l la:..-..j.L I.~~.I~~~~~~~~I i).,I.o.Ju f I.j.l P~-l-i I'I--lj,j~---., I.o I~I I-s~..~~~(~~r I t g~~l f...I q,).I.I.-~~o Conservative Representation I~~i~oE Containment Prcssure Q I 5 IO:,.'0~:.~~':-.::.'~:~.':.I 10 I~~~~I~I 10 I~:;i.16~6+18-.': '.~I I~~<-~',-~t~4-t-I.I I)Jay-t~'03 10 Tim After 9 sl.gn 2asis Accident ('econds)10 Po g g Q g Q
 
c, Table  3 Page  ll Reactor:    GINNA                                              SYSTEMATIC EVALUATION PROGRAM T1me          ENVIRONMENT                  Qua      Document Eguipment Type            Location      Needed Parameter  Requ1re    Qua .      Method      Reference    Comments
: 37. H2  Recombiner      Area 41  Long        Temp  ('F)  286        315      Test          18,19,49 Igniter Exciter Unit                      Pr (psia)  75        105      Test          18,19,49 GLA  Part No. 43737,                      RH (%)      100        100      Test          18,19,49 Rev. A,                                    Chem.      Yes        Yes      Test          18,19,49 Serial 001                                Rad.        1.6xlo    1.73x10  Test          18,19,49 Sub.        No
: 38. H2  Recombiner      Area 51  Long        Temp (OF    286        286      Te'st        18,19,49 Blower Motor (2/15                        Pr (psia)  75        75        Test          18,19,49 Scale) W 2 HP,                            RH (%)      100        100      Test          18,19,49 Class H Ins., Model                        Chem.      Yes        Yes      Test          18,19,49 TBFC                                      Rad.        1.6xl0    2.0x10    Test SO  68C24196                                                                            18, 19,49 Sub.        No        No
: 39. Pump Motor            Area N2  Long        Temp (OF)  Amb.      Amb.      Experience              Not exposed to U.S. Electrical                            Pr (psia)  Atm.      Atm.      Experience              DBE environment Motors                                    RH (%)      Amb.      Amb.      Experience Model VEU, 100 HP                          Chem.      No Frame 84-445 U                            Rad.        No Insulation Class B                        Sub.        No (Charging Pump)
: 40. Solenoids/            Area 58  Short      Temp  ('F)  Amb.      Amb.      Experience              Not exposed to Johnson Controls                          Pr (psia)  Atm.      Atm.      Experience              DBE environment Model D251                                RH (%)      Amb.      Amb.      Experience (Control    Room Air                      Chem.      No Handling                                  Rad.        No Unit  Dampers)                            Sub.        No
: 41. Medium Voltage        Area 07  Short      Temp  ('F)  Amb.      Amb.      Experience              Breakers need I    Switchgear                                Pr (psia)  Atm.      Atm.      Experience              only open for Westinghouse                              RH (%)      Amb.      Amb.      Experience              LOCA inside DH  -  350E                                Chem.      No                                          containment to cc 1200 A Breakers                            Rad.        No                                          stop RC pumps.
(RCP  Trip Breakers)                      Sub.        No                                          Not exposed to DBE when  needed to function.
 
Table 3                                                                                                      Page 12 Reactor:  GINNA                                              SYSTEMATIC EVALUATION PROGRAM Tame          ENVIRONMENT                  Qua  -    Document, Equipment Type          Location    Needed Parameter  Require    Qua .      Method      Reference    Comments
: 42. RHR Pump  Cooling    Area 02  Long      Temp  ('F)  Amb.      Amb.      Experience              Only exposed to System Fan Motors                        Pr (psia)  Atm.      Atm.      Experience              DBE  radiation Westinghouse Model                      RH (%)                Amb.
7 Experience              environment SBDP                                    Chem.      3xlO      lx10      Test          69 Class  B Insulation-                    Rad.        No 2HP                                      Sub.        No
: 43. Cont Spray/SI Pump  Area 52  Long      Temp  ('F)  Amb.      Amb.      Experience              Only exposed and Charging Pump                        Pr (psia)  Atm.      Atm.      Experience              to  DBE radiation Cooling Systems                          RH (%)      Amb.6      Amb.      Experience              environment Fan Motors                              Chem.      3x10      1x10 >    Test          69 Westinghouse Model                      Rad.        No SBDP                                    Sub.        No Class  B Insulation-3HP
: 44. Main Control Board  Area N2  Long      See                                                      "Mild" Environment.
Reactor Trip Racks                      Comments                                                be addressed at Relay Logic and                                                                                  a  later time Test Racks Miscellaneous Racks Auxiliary Relay Racks Safeguard Racks Reactor Coolant System Racks CVCS  Racks Feedwater Control Racks SI Sequence  Racks
 
C I
 
Table 4 Environmental Service Conditions Inside Containment Normal 0 eration Temperature:        60-120 F Pressure:            0 psig Humidity:            50% (nominal)
Radiation:              1 Rad/hr general.      Can be higher or lower near specific components.
Temperature:         Figur'e  5  (286'F max)
Pressure:           Figure  4 (60  psig design)
Humidity:           100%
Radiation:           Figure 6 (1.6 x 10 total)
Chem. Spray:        Solution of boric acid (2000 to 3000 ppm boron) plus NaOH in water.
Solution  pH between 8 and 10.
Flooding:            7  ft  (approx)
Auxiliar Buildin Normal 0 eration Temperature:        50-104  F Pressure:            0  psig, Humidity:            60%  (nominal)
Radiation:              10  mr/hr general, with areas near      RHR piping < 100 mr/hr during shutdown operation Accident Conditions   includin    sum  recirculation Temperature:        50-104'F (122'F near motors)
Pressure:            0 psig Humidity:            60% (nominal)
Radiation:          Operating Floor      (271'lev.):
Near Bus 14 and  NCC 1C  6 1L:
100  rad Other Areas: less than 50 rad Intermediate Floor (253'lev.):
Near Bus 16 and MCC 1D 8  1N:  900  rad Other Areas:    less than  500 rad Basement Floor (236'lev.):
Near CS, RHR, an( SI Pumps:    2.8 x 10 pads Other areas:    < 10  rads Spray:              N/A Flooding:            N/A
 
C. Intermediate Buildin Normal 0 eratzon Temperature:          50-104'F Pressure:              0 psig Humidity:              60% (nominal)
Radiation:                1 mr/hr (higher  near reactor coolant sampling lines)
Accident Condition Based u on HELB or    MELB Temperature:           215'F  for  30 minutes; then reducing to 104 within 3 hrs Pressure:             0.8 psig for 30 minutes; then reducing to O,psig within 3 hrs Humidity:             100%  indefinitely Radiation:             N/A Spray:                N/A Flooding:              0 Based u on LOCA    conditions Temperature:          115'F  indefinitely* near large motors and  FW  and SL piping. 104'F in open areas Pressure:              0  psig Humidity:              100%
Radiation:            Negligible Spray:                 N/A Flooding:             0 D. Cable Tunnel Same  as  Intermediate Building E. Control Buildin Control Room Normal 0   eration Temperature:          50-104'F (usually 70-78'F)
Pressure:              0 psig Humidity:              60% (nominal)
Radiation:            Negligible Accident Conditions Temperature:              104oF Pressure:              0. psig Humidity:              60% (nominal)
Radiation:            Negligible Spray:                N/A Flooding:              N/A
*Estimated (no  explicit calculations    performed)
 
~1 Normal 0    eration Temperature:            50-104 F Pressure:              0 psig Humidity:              60% (nominal)
Radiation:              Negligible J
Accident Conditions Temperature:               104 F Pressure:              0 psig Humidity.              60% (nominal)
Radiation:              Negligible Spray:                  N/A Flooding:              N/A Normal 0    eration Temperature:            50-104  F Pressure:               0 psig Humidity:               60% (nominal)
Radiation:             Negligible Accident Conditions Temperature:            <  104'F Pressure:              0 psig Humidity:              60% (nominal)
Radiation:              Negligible Spray                  N/A Flooding:              N/A Necbanical  E    i ment Room Normal 0 eratzon Temperature:            50-104 F Pressure:              0 psig Humidity:              60% (nominal)
Radiation:              Negligible Accident, Conditions Temperature:           <  104'F Pressure:              0 psig Humidity:              60% (nominal)
Radiation:            .Negligible Spray:                  None Flooding:              3  ft.  (estimated for a service water line leak)
 
F. Diesel Generator Rooms Normal 0 eratxon Temperature:            60-104 F Pressure:               0 psig Humidity:               60% (nominal)
Radiation:             Negligible Accident Conditions Temperature:                104 F Pressure:              0 psig Humidity:              90% (estimated)
Radiation:              Negligible Spray:                 N/A Flooding:               0 ft **
G. Turbine Buildin Normal 0   eration Temperature:            50-104    F Pressure:              0 psig Humidity:              60% (nominal)
Radiation:              Negligible Accident Conditions Temperature:            220'F'or   30 minutes, reduce to 100'F within 3 hrs.
Pressure:              1.14 psig on mezzanine and basement levels, 0.7 psig on operating floor Humidity:              100%
Radiation:              Negligible Spray:                  N/A Flooding:              18'~ in basement (Circ. Water Break)
H. Auxiliar Buildin    Annex Normal 0 eratzon Temperature:           60-120 F Pressure:              0 psig Humidity:              60% (nominal)
Radiation:             Negligible Accident.'Conditions Temperature:           60-120 F Pressure:            . 0 psig Humidity:              60% (normal)
Radiation:              Negligible Spray:                  N/A Flooding:                  2 ft.
**Service water line crack would affect only     one room (see FEOOD-15)
 
Screenhouse Normal 0 eration Temperature:         50-104 F Pressure:            0 psig Humidity:            60% (nominal)
Radiation:           Negligible Accident Conditions:
Temperature:         < 104 F Pressure:            0 psig Humidity:            60% (nominal)
Radiation:          Negligible Spray:              N/A Flooding:            18" (Circ. Water Break)
 
Deeda Basf.s Accident Temperature                                         -. Time Curve I       ~
                                                                                              ~           ~   ~
                                                                                                                  ~ .. ~ ~     ~
(
                                                                                                                                    ~
                                                                                                                                        ~     r      I t  g
                                                                                                                                                                  ~              ~l f
                                                                                                                                                                                                        ...I q,) .I. I.-          ~
                                                                                                                                                                                                                            ~   o t
$ 00 oo
                                                      ~
Containment              ~
                                                                                          ~
Temperature                                                  Conservative Representation                                                I~        ~
i  ~
o                                                             oE Containment Prcssure Q
                                                                                                              ~
I
                                                                              ~   ~ ~
I   5 0I                                                                                I
                                                                                        ~ I       , .
I                                                                                                                         IO
                                                      'l
                                                                                                          ~
                                                                                                              )
I           Sump  Temperature    ~
                                            .        .
I I
o I ~
o
                                                                                                            ~
                                                                                                            ~
                                                                                                          'I                            :,      .
                                                                                                                                                      '
0~:.               ~           ~   ':-.::.'~: ':.       ~
                                                                                                                                                                                                                            .
                                                                                                                                                                      .'.. -i;:>>~I-:~ f--
I 5$ 0--                                               Ii.l I                          l        I          ~
I        ~    I I 1O O o
                                                                                            ~  ~    H Heat Exchanger Outlet              I
                                                      .
150 Temperature I*                                        I      ~                                          I  '" '                   I
                                                                                                                                                                                            '-' I" -I-l-j..
I I
                                                                          'I
                                                            ~  ~
: l. I.
                                      ~
                                          ~
                                              ~
                                                  ~
                                                      .
                                                          'II'                    \                        ~
* I                                                     ~   I ~     .,I l
                                                            ~ '
t.l  ~  .        I                                                                                  la:           ....j.L I o                                  ~
I'                                                                              I. ~ ~ . I~       ~
I
                                                                                                                                                                        ~
i) .,I
                                                                                                                                                                                ~   ~   ~
                                                                                                                                                                                          .
                                                                                                                                                                                            ~ ~
lj,j I                                                                                                                                                      o I~
I I,'.-  ~      ~ ~ o I
                                                                                                                                                                                    .Ju                                             I I    I  ~
                                                ~ ~
i
                                                    ~
r l-i I    'I-f       I   .j.l     P~-                     ~  -- -., I.o I- s
                                              ~  ~
I ~
                ~ I~ )
r I ~
          ~ ~        ~
I    ~  I ~:;i       .      16 ~ 6+18-.':          '.    ~
I   I ~    ~    <-       ~ ',  -t ~    ~   4- t-I .I      I    ) Jay              -t~
I        I
                                                                                                                                                                                                '03 10          10                                                                                        10                                                                                                                    10 Tim          After 9          sl.gn 2asis Accident ('econds)
Po  g gQg Q
 
Post-Accident ConMinnent Y~terials Design Conditions
                  ~      ~ ~, ~ ~  ~,
                                    ~      ~  ~
r.
                                                  ~,
                                                ~ ~
                                                        ~
                                                          ~ ~
                                                            '        ',;
I ~....'
                                                                                                        ',      ~
                                                                                                                    ~
I  I
                                                                                                                          ~
                                                                                                                              ~,
                                                                                                                                '
                                                                                                                                        ~
                                                                                                                                          ~
                                                                                                                                            !
                                                                                                                                            '
                                                                                                                                                ~
                                                                                                                                                    ~ ';;,
I   ~ ~
                                                                                                                                                              ~ '
                                                                                                                                                                      ~ .
I
                                                                                                                                                                          ~
                                                                                                                                                                              ~
                                                                                                                                                                                ~
                                                                                                                                                                                'gl'
                                                                                                                                                                                      .
                                                                                                                                                                                        '
3
                                                                                                                                                                                              ~,!
                                                                                                                                                                                                  ~, ''.
                                                                                                                                                                                                  ~,
j
                                                                                                                                                                                                                !  ~
                                                                                                                                                                                                                  ...,~    ~
                                                                                                                                                                                                                            ~            ~ ~
                                                                                                                                    ~,
                                  ~   ~
I%PE' P
                                                                        ~
                                                                            ~   '
                                                                                    ~
I   ~  0      ~  y    w,t               I                  I    ~
I        ~  ~  I    ~        ~~    L  ~
                                                                    ~ ~
                                                                                                                                                                                                                                ~  I    ~
250    .
                                                    ~ I                      ~  '                                                                              ~    ~
2OO I
g       C 0      !                                                                                                    I    ~             I O                                                                                                                                  I~~
I 1.~'.-    .
                      ~ I~                                      +~
I I
                                                                          ~
                                                                                "      ''      ',   !    .'              ~    '          '' '    ''              '              '                '  '
1M' j
                ~  ~    ~
              !'                                                                                                                I                                    a                                      4                    \ ~ 'II
                                    ~
                                        ~
                                                    ~      ~  ~  ~                g        g
                                                                                                      ~
4
                                                                                                          ~    ~
                                                                                                                        ~  I 4      ~      I  ~      l
                                                                                                                                                              ~                    ~
                                                                                                                                                                                          ~,
I
                                                                                                                                                                                                    ~        '  ~  ~ ~     4
                                                                                                                                                                                                                      ~ I Figurc            5
 
Contninment Atmosphere Intcgrnted                                                                          Cnmmn  Dose      Level 109,                                      5 4        S      6799)                                            4        5        67091                          2            5    4      5      6769)                                  4    5 I)            799)                2              5      4      5      67991 6
                        ~  ~
lt
                                  ~     ~ I'i ti,i              ~ <<                                                  +                            i  ~                                r  ifr                                                                                                      Ia
                  ~ it)            I   t!l                              ~    'Ial  1~  If Ii rti  ~
Zjj                          )li  C                          L    ia<<                                                          aa'                          ~ t  it!    Ia                              L 7
6
                                                                                                                                                                                            'Ia    air                                                                                                                  ~,,',i S
                                                                                                                          '.I '      I    I  ~ I  ~ ~  ''f  I~i.i I
                                                                                                                                                                      ~
il IIJ tra                          aig ita        3
                                                                                                                                                                                              ~ I                                                                                    Bf    ~
J 2
I                                                                                          )ra    -'p.
                                                                                                                          ~  li  ~
a ~
cl gl 10                                                                                                                                                                                                                                          M  ~
          )<<
6
                                ~ ~  I                  ~    I                          a ~
I'.                                               ~  I ~ ~  I
                                                                                                                                                                                                    'I
                                                                                                                                                                                                                      ~ ~
a
                                                                                                                                                                                                                                                                                      ~ I                            ~ ~      ', "I' j<<:               I L'j    ~ ~                        I<<                                                                                                                                    I~I                                                  I~                                          ~ I                  ilia I,
                                                                                                              ~
gj IZ}      7                                                                                                                                          '                                                                                                                                      I at! iiia II                                                            ~                                           La I
6 ll I ar                                                                                                        .Z  ':
F
                )4    j.'i'l          ~~a'          ~
ta
                                                        ~ I~
I            JII+II  ~
Ii' Iiii  ~
ia                                                      I ~I rI'I                                                        I                          :~
a I
oIJ                                                  I I II                                                                                                                                                                                                    I
                                                                                                    !;I
                                                                                                                                                      'I
                                                                                                                                                                    ~~
Ia,!
                                                                                                                                                                                                                                  ~ Il
                                                                                                                                                                                                                                                                                                            't                t" iF) 2
                                                                                                      ''ltt                                                                                                                                                                                                                  Ix I                                                                          Jl'I Ill;  ~ I IL 10 wa
                                                                                          ~ I
                                                                                                .~
3        II!
9 6
                                    ~~                                                              t'                    ~  I ta    tr          t'  ',  ~   I                      I~ I                                          a ~
I}  ~ !                                                                      ~ il      .;1                            t f~I   ~ !at          )                                                                                                                        ~ jj:    ~  I 7                                                                                                          '      I                              '                                                                                              ~
I 6                                                                                                                                                                                                                                                                                                                      L IR'} J~li:                                                                                                                                                                                                                       I                      'I 5                                                                        . )p,                                                                                                          ; I'9
                                                                                                                                                                                                                                                                'Pi, I'l I
I                    ~
WI.~J I                  Jli                            a I
                                                                                                                                    <<I,                                      I  ~ ~
                                                                                                                                                                                                    ~
                ;I I    '                              Il'i':.                                                                                                         }~ii    Ir I:.y iI I                                                                                          11                          I rl            tl                                            I) I                                            . II':
j3'i
                      ~                                                      ~
I 1 la!                                                                                                                                            it il I!I:                                                I        I:
1 L
i)i'}                ""') IIII                                              II
                                                                                                                                        <<                                                                                          I~
Ii}!J:":
La
                                                                                                                                                                                                                                                                                              }})J
                                                                                                                                                                                                                                                                                                              ~
                                                                                                                                                                                                                                                                                                                  -'}LI II'I                                                      il}                                                                                              aMI                    illiI I ~                 ~:                          tj
                                                                                                                                                                                                                                                            )lI:,:                                                          ~
:',I:
                -"  xg WW                                                                5))                  a il!
                                                                                                                                                                                                                                          ~a
                                                                                                            'I ~ I          '                                                                ~ I                          1+4'.I Il"
                                                                                        }
                                                                                            ~    ~
I IL "
ii I
                                                                                                                                          'al                          l dny            I ) I;.                       1 week I  II 1 month                                                                  1 yent 10                                                                                                                                                                                  I                                               I I t 2                                                      10-'-
10                                                          10                                                                        10                                                            J.Lf                                                                                                                    10 Time After. Activity Relensc (hours)
Figur)
 
I
,I'
    >i
    <I'r
  )~f


Post-Accident ConMinnent Y~terials Design Conditions I~....'~~,~!~I~~~.~~.3~,''.!~~,~~,~~~,~~~~~~~~~~~~I I~'I~'gl'~'',;',~'~'~';;,'~,!~, j...,~~~~~~~r.I%PE'P~'I~0~y w,t I~, I~I~~I~~~L~~~~~~I~250.~~I~I~'2OO I g C 0!O I~I I~~I~a~,'!'~~~~~~~~~g g 4~I 4~I~l I 1.~&#x17d;.-.I~~I I+~~"''',!.'~''''''''''1M'j~~~~~4~~~4\~'II~I Figurc 5 109, 5 4 S 6799)Contninment Atmosphere Intcgrnted Cnmmn Dose Level 4 5 67091 2 5 4 5 6769)4 5 I)799)2 5 4 5 67991 6 7 6 S~~~~it)lt I~I'i ti,i t!l~<<~'Ial 1~I f Ii~rti+Zjj i~)li C I~i.i I il IIJ r L ifr ia<<'Ia air aa'Ia~t it!Ia~,,',i L'.I'I I~I~~''f~tra aig ita 3~I Bf~J 2~li~I a~)ra-'p.gl cl 10)<<j<<: 6 IZ}7 6 F IJ o 2 I 10 9 6 7 6 5 10-~~I I L'j~~wa~~I}~!I rl 1;I~'I tl L-" xg WW)4 j.'i'l~~a'~I I<<~~I~ta I I I II~Il'i':.i)i'}a~~I JII+I Ii'Jl'I~I.~IR'}J~li:.)p, WI.~J I j3'i~I)I I""')IIII I~~: 5))}~~I~ia Iiii!;I''ltt Ill;~I IL t'~il Jli 1 la!a'I~I I'.~IL"~I.;1'I tj il!'ii I a I<<I,.II': I I<<II'I'al~I~~I'I I I I I'~r I t f~I~!at''I~~3 II!ar ta)tr I~~}~ii Ir I:.y l dny I I~I'I iI I il}:',I:~I)I;.I;I'9~I t'',~I~~I,~I Ia,!~Il I~I I~week I II~a 1+4'.I Il" 1 M~a I~La a I:~I~'Pi, it il Ii}!I!I: J:":)lI:,: 1 month I I a~~I 11 La~I I at!I I ll iiia I'l}})J aMI~jj: I I.Z': 't~I~~ilia~I I I:-'}LI ill i I~','gj"I t" iF)Ix L'I 1 yent 10 10 10 t 2 J.Lf 10-'-10 Time After.Activity Relensc (hours)Figur)
I ,I'>i<I'r)~f GINNA STATION (DOCUMENTATION REFERENCE) l.2~3~4, 5.6.7~8.9~10.11.12.13.14.15.16.17.18.19.20'1-22'3'4'5.26'7.28'9'0'1'2'3.34'5'6'7'8'9'0'3.d s tions na 1974 f rom L.D.White on Report F-C5074, Splice Sleeves Crouse-Hinds Penetration Test Report Gilbert Spec.520-Standby AFN Pumps Gilbert Spec.711-Standby AFW Pump Motors Gilbert Spec.5201-Large Motors Deleted.Included in Reference 51 Gilbert Spec.5342-HVAC Throughout Ginna Gilbert Spec.RO-2239-Diesel Generators Gilbert Spec.RO-2267-Auxiliary Feedwater Pumps Gilbert Spec.RO-2400-Batteries IPCEA Std.S-61-402, Sect.3.8 and 4.3.1 Kerite Memo 7/22/68 NEMA Std.SG-3, Low Voltage Circuit Breakers Nestinghouse Spec.676258-Motor Operated Valves Westinghouse Spec.676270-Control Valves Westinghouse Spec.676370-Auxiliary Pumps Westinghouse Spec.676427-Auxiliary Pump Motors NCAP 7343 June, 1969 NCAP 7410-L, Vol.I&II WCAP 7744, Vol.I 8 II NCAP 9003, January, 1969 Deleted.Included in Reference 45 Deleted Report NS-CE-775, Pail-Safe Operation of ASCO Solen.Copes-Vulcan Solenoid Valves Vendor Data on Laurence Solenoid Vendor Data on Versa Solenoid WCAP 7153 Deleted.Included in Reference 45 Gilbert Spec.504-Westinghouse Electrical Penetra Technical.Proposal for Electric Penetration for Gin Containment Structure by Nesti'nghouse
-September 4 NCAP 7354-L Vendor Data on Gould Batteries Westinghouse Spec.Sheet for Foxboro Transmitters Vendor Data on Barton 209 Transmitter Rosemont RTD Spec.Vendor Data on Raychem Splice Sleeves June 16, 1975 Letter to R-.A.Purple Containment Flooding April 4, 1979 FRC Final and Cable Deleted Deleted (I J I R)
GINNA STATION (DOCUMENTATION REFERENCE)
GINNA STATION (DOCUMENTATION REFERENCE)
-CONT'D 41'2'3.44~45'6'7.48'9;50'1.52.53.54.55.56.57.58.59.60'1'2'3'4.65'6~67'8.69'0'eleted Deleted Design Criteria-Standby Aux.Feedwater System-October 24, 1974 Limit Switches Design Approval Test on Material Used in Westinghouse Penetrations for the Brunswick Station of Carolina Power and Light Company-August ll, 1972 Test Data for Coleman and Rome Cable Aging Failure Detect.ion Program Valcor Solenoid Valve: Vendor Data and Test Report Extracts WCAP-9001 Westinghouse Terminal Blocks Cable Identificat.ion and Qualification Supplement, Including F-C5074 (Supplement)
: l.      Crouse-Hinds Penetration Test Report 2 ~    Gilbert Spec. 520  Standby AFN Pumps 3 ~    Gilbert Spec. 711  Standby AFW Pump Motors 4,      Gilbert Spec. 5201  Large Motors
Concerning Silicone-Rubber-Insulated Cable Qualificat.ion Wide-Range Sump Level Switch Specification Limitorque Valve Operator Data, Including Limitorque Report B0003 and Section 4.1.4 of B0058.Containment, Electrical Penetrations Kerite Letter, June 26, 1980 IE Inspections 78-20 and 78-21-Reports Concerning Installation of Splice Sleeves Control Valve Specification SP-513-044666-000, September 27., 1974, Concerning.Standby ApW Valves Westinghouse 10/10/80 Letter Concerning Crouse-Hinds Electrical Penetrations Evaluation of Organic Materials on Crouse-Hinds Electrical Penetrations Westinghouse Terminal Block Information on Aging and Radiation Aging Evaluation of Westinghouse Electrical Penetrat.ions Raychem Splice Sleeve Aging Information Kerite Cable Aging Information Containment Fan Cooler Motor Splices Safety-Rel'ated Motor Bearings.-Maintenance and Lubrication Safety-Related Motor Characteristics (Insulation)
: 5.      Deleted. Included in Reference 51
WCAP-8754 Westinghouse Research Report 71-1C2-RADMC-Rl, December 31, 1970 (Revised April 10,'1971), Concerning"The Effect, of Radiation on Insulating Materials Used in Westinghouse Medium Motors" WCAP-7829,"Fan Cooler Motor Unit Test" I J J J;P~f}}
: 6.      Gilbert Spec. 5342  HVAC Throughout Ginna 7 ~    Gilbert Spec. RO-2239  Diesel Generators
: 8.      Gilbert Spec. RO-2267  Auxiliary Feedwater    Pumps 9 ~    Gilbert Spec. RO-2400  Batteries
: 10. IPCEA Std. S-61-402, Sect. 3.8 and 4.3.1
: 11. Kerite Memo 7/22/68
: 12. NEMA Std. SG-3, Low Voltage Circuit Breakers
: 13. Nestinghouse Spec. 676258 - Motor Operated Valves
    '3.d
: 14. Westinghouse Spec. 676270  Control Valves
: 15. Westinghouse Spec. 676370  Auxiliary Pumps
: 16. Westinghouse Spec. 676427  Auxiliary Pump Motors
: 17.      NCAP 7343 June, 1969
: 18.      NCAP 7410-L, Vol. I & II
: 19.       WCAP  7744, Vol. I 8  II 20        NCAP 9003, January, 1969
  '1-Deleted. Included in Reference    45 22        Deleted
  '3 Report NS-CE-775, Pail-Safe Operation of    ASCO  Solen    s
    '4
        .Copes-Vulcan Solenoid Valves
  '5.
Vendor Data on Laurence Solenoid 26        Vendor Data on Versa Solenoid
    '7.
WCAP  7153 28        Deleted. Included in Reference  45
  '9 Gilbert  Spec. 504  Westinghouse  Electrical  Penetra tions
    '0 Technical .Proposal for Electric Penetration for Gin na
  '1     Containment Structure by Nesti'nghouse  September      4  1974 NCAP 7354-L
    '2 Vendor Data on Gould Batteries
    '3.
Westinghouse Spec. Sheet for Foxboro Transmitters 34        Vendor Data on Barton 209 Transmitter
    '5 Rosemont RTD Spec.
      '6 Vendor Data on Raychem Splice Sleeves
    '7 June 16, 1975 Letter to R-.A. Purple        from  L. D. White on
    '8   Containment Flooding April 4, 1979 FRC Final      Report  F-C5074,    Splice  Sleeves
    '9  and Cable Deleted
    '0 Deleted
 
(
I I
R J
    )
 
GINNA STATION (DOCUMENTATION REFERENCE)          CONT'D 41
    '2 Deleted
  '3.
Design Criteria Standby Aux. Feedwater System October 24, 1974 44  ~  Limit Switches 45      Design Approval Test on Material Used in Westinghouse
  'eleted Penetrations for the Brunswick Station of Carolina Power
    '6 and Light Company August     ll, Test Data for Coleman and Rome Cable 1972
    '7.
Aging Failure Detect.ion Program 48      Valcor Solenoid Valve: Vendor Data and Test Report Extracts
    '9; WCAP-9001 50      Westinghouse Terminal Blocks
    '1.
Cable Identificat.ion and Qualification Supplement, Including F-C5074   (Supplement)   Concerning Silicone-Rubber-Insulated Cable Qualificat.ion
: 52.      Wide-Range Sump Level Switch Specification
: 53.      Limitorque Valve Operator Data, Including Limitorque Report B0003 and Section 4.1.4 of B0058.
: 54.      Containment, Electrical Penetrations
: 55.      Kerite Letter, June 26, 1980
: 56.      IE Inspections 78-20 and 78-21 Reports Concerning Installation of Splice   Sleeves
: 57.      Control   Valve Specification SP-513-044666-000,         September 27., 1974, Concerning .Standby ApW Valves
: 58.      Westinghouse 10/10/80 Letter Concerning Crouse-Hinds     Electrical Penetrations
: 59.      Evaluation of Organic Materials on Crouse-Hinds Electrical Penetrations 60      Westinghouse Terminal Block Information on Aging and Radiation
  '1 Aging Evaluation of Westinghouse Electrical Penetrat.ions Raychem Splice Sleeve Aging Information
    '2
  '3 Kerite Cable Aging Information
  '4.
Containment Fan Cooler Motor Splices 65 '6 Safety-Rel'ated Motor Bearings Maintenance and Lubrication
                                            .
  ~
67      Safety-Related Motor Characteristics (Insulation)
  '8.
WCAP-8754 69      Westinghouse     Research   Report     71-1C2-RADMC-Rl,   December 31, 1970   (Revised April 10,   '1971), Concerning "The Effect,
  '0    of Radiation on Insulating Materials Used in Westinghouse Medium Motors" WCAP-7829, "Fan   Cooler Motor Unit Test"
 
I J
J J;
P
~
f}}

Revision as of 20:51, 29 October 2019

Environ Qualification of Electrical Equipment, Revision 3
ML17250A715
Person / Time
Site: Ginna Constellation icon.png
Issue date: 10/31/1980
From:
ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17250A714 List:
References
TASK-03-12, TASK-3-12, TASK-RR NUDOCS 8011040240
Download: ML17250A715 (159)
v • d • e


Text

Environmental Qualif ication of Electrical Equipment R. E. Ginna Nuclear Power Plant Docket No. 50-244 February 24, 1978 Rev. 1, December 1, 1978 Rev.> 2, April 25, 1980 Rev. 3, October 31, 1980 luanCE THE ATTACHED FILES ARE OFFICIAL RECORDS OF THE DIVISION OF DOCUMENT CONTROL. THEY HAVE BEEN CHARGED TO YOU FOR A LlhhlTED TIME PERIOD AND MUST BE RETURNED TO THE RECORDS FACILITY BRANCH 016. PLEASE DO NOT SEND -DOCUMENTS CHARGED OUT THROUGH THE MAIL. REMOVAL OF ANY

"

PAGEIS) FROM DOCUMENT FOR REPRODUCTION, MUST BE REFERRED TO FILE PERSONNEL.

Docket@

gO~tIO) g So iI Oqc a. S V DEADLINE RETURN DATE gEGlHATORV OOgKET F{K Bo 1.>040'/B'P~H>>@R I FY " '" ~' A, RBCCIIDB FACILITY BRANCH

P r

I

,,i 7g

/p1 1

i

/

ji'N

TABLE OF CONTENTS Pacae Introduction Identification of Necessary Safety Related Equipment 3 A. Events Accompanying a Loss of Coolant Accident 3 B. Events Accompanying a Main Steam Line Break or 11 a Main Feed Line Break C. High Energy Line Breaks Outside Containment 16 Identif ication of the Limiting Service Environmental 19 Conditions for Equipment which is Required to Function to Mitigate the Consequences of Events Identified Above A. Inside Containment 19 B. Auxiliary Building 22 C ~ Intermediate Building 25 D. Cable Tunnel 27 E. Control Building 27 F. Diesel Generator Rooms 30 G. Turbine Building 30 H. Auxiliary Building Annex 32 I. Screen House 32 Equipment Qual ification Information 34

1 I

I I

LIST OF FIGURES Figure 1 Loss of Coolant Accident fSequence of Events Diagram]

Figure 2 Main Steam or Feed Line Break (Sequence of Events Diagram]

Figure 3 - Plant Layout Figure 4 Pressure Envelope for Ginna (FSAR Figure 1 of Appendix 6E)

Figure 5 Temperature Envelope for Ginna (FSAR Figure 2 of Appendix 6E)

Figure 6 Radiation Level for Ginna (FSAR Figure 5 of Appendix 6E)

LIST OF TABLES Table 1 Loss of Coolant Accident [Required Equipment List]

Table 2 Main Steam or Feed Line Break [Required Equipment List]

Table 3 Equipment Qualif ication Table 4 Environmental Service Conditions

Environmental Qualification of Safety-Related Electrical Equipment INTRODUCTION The electrical equipment described in this report is that saf ety-related equipment required to mitigate the ef f ects of high or moderate energy line breaks (HELB) inside or outside containment, and to effect eventual cold shutdown of the reactor. The environmental qualification requirements are described in the "DOR Guidelines", transmitted to RG6E on February 15, 1980. Although the DOR Guidelines address all electrical equipment, the emphasis in this report will be on that equipment exposed to an adverse HELB environment. This is defined as that equipment located in the containment, Intermediate Building, Turbine Building, and Auxiliary Building basement ( radiation only). This revised scope is consistent with the Commission Order of September 19, 1980.

Equipment in other "mild" environments will be addressed at a later time.

This submittal revises and supersedes our previous reports concerning environmental qualification of electrical equipment, dated February 24, 1978, December 1, 1978, and April 25, 1980. It also consolidates and updates all information submitted on June 10, 1980 and September 24, 1980. Section IV of this report presents an item-by-item response to the Draft Interim Technical Evaluation Report FRC Project C5257, concerning the review of the Ginna electrical equipment

P environmental qualif ication, dated August 20, 1980. New references are included with this report. However, references previously submitted are not being resubmitted.

1n Section IV, it is either shown that each item is adequately qualified to perform its required safety function in its post-accident operating environment, or a commitment for additional testing or replacement is made. In all cases, sufficient justification for continued operation is given.

Table 3 summarizes the equipment qualification in the format requested for SEP by the NRC in a September 6, 1978 letter.

Table 4 provides the definition of environmental parameters throughout the Ginna plant. This table is comparable to Appendix A of F-C5257, and tabulates the explanatory basis given in Section III of this report.

Supplement No. 3 to IE Bulletin 79-01B provides the timing for submittal of qualification information for equipment in-stalled to meet the TMI Short Term Lessons Learned. RGSE intends to follow the guidance given in this supplement. In a number of cases, it is possible that additional documentation or testing results may become available after November 1, 1980. Since this additional information will be of use in documenting the status of the Ginna environmental qualification, it will be submitted when received. Every effort has been made to ensure that all documentation was obtained for use with this submittal.

l II. IDENTIFICATION OF NECESSARY SAFETY RELATED EQUIPMENT This section of the report identifies the necessary safety related equipment for each of the Design Basis Events (DBE) of concern and a brief description of why the equipment is needed. This identification includes all electrical equip-ment required by the Ginna emergency procedures for accomplish-ing the necessary safety functions. It must be recognized that not all electrical equipment referenced in the procedures is required to function ( as opposed to being useful if available), and is therefore not required to be qualified.

The emergency operating procedures were not developed by considering safety-related components to the exclusion of all others. While such procedures are written with priority attention given to safety-related equipment, other systems and components are justifiably mentioned. A realistic evaluation of plant incidents might result in situations and hostile environments significantly less severe than those assumed for the purposes of conducting the environmental qualification program. The absence of full qualification for certain components which fall into this category is not, by itself, a sufficient motive to classify the equipment inoperable .or to remove these components from the procedures.

A. Events Accom an in a Loss Of Coolant Accident Analyses of the course and consequences of loss of coolant accidents have been submitted previously (LOCA 1-4). A discussion of equipment required to function to mitigate the consequences of a loss of coolant

accident is presented in the FSAR Chapters 6, 7 and 14.

Post-LOCA operator actions are included in the Ginna Emergency Procedures. These procedures are consistent with the generic Westinghouse guidelines, which have been approved by the NRC. Additional descriptive material is presented in this report to provide summary information as to the sequence of events and the equipment involved at each stage. Figure 1 illustrates the sequence of events following a loss of coolant accident.

Table 1 provides a specific equipment list for each numbered block in Figure 1. Also provided in Table 1 is the safety function which is required and the period of time that operability must be ensured. It should be noted that Table 1 includes all redundant equipment, not the minimum safeguards equipment assumed in the safety analysis. In the "required" column it should be noted that equipment listed as "signal initiation" is required to be operable only until its required safety function, the initiation of a safety signal, is performed.

It is important to note that the arbitrary requirement of the DOR Guidelines to qualify equipment to function for at least one hour, even if its only function is completed within seconds, is not well reasoned. In many cases, the environment would not exist unless the equipment safety function had been completed (e.g.,

flooding to a seven foot level in containment by necessity means that SI was initiated). RGSE does not agree with

this one-hour requirement, and it is therefore not applied as an environmental qualification requirement.

Equipment listed as "long term" is required to provide long term decay heat removal, post-accident monitoring and sampling, or maintaining a safe shutdown condition.

Equipment listed as "short term" is required only for a short period of time (hours).

Table 3 provides the environmental qualification require-ments and documentation references for the Ginna Class IE equipment.

1. The first event in the loss of coolant accident following the rupture is the detection of the rupture.

Any 2/3 low pressurizer pressure or 2/3 high contain-ment pressure will initiate "safety injection" (SI).

la. Instrumentation is available to the operator to distinguish between a LOCA and the other accidents, such as a steam line break or feed line break. It is important to note that the automatic actions and immediate operator actions (first 10 minutes) are identical in the mitigation of these accidents.

2. Upon "safety injection" signal generation, safe-guards sequencing is initiated (see FSAR Table 8.2-4).

The diesel generators start and energize the safeguards buses assuming there is a loss of offsite power. With the safeguards buses energized, either by off-site power or the diesels, the three safety injection pumps,

the two residual heat removal pumps," two of the four service water pumps, the two motor driven auxiliary feedwater pumps, and the four containment. fan coolers

-will "be loaded sequentially onto the buses. The two containment spray pumps are automatically loaded onto the buses when the 30 psig containment pressure setpoint's reached.

3. A break in the reactor coolant system piping actuates the passive accumulator injection system when the reactor coolant system pressure is reduced to 700 ps lg The flow path of the borated water from each accumulator-is through a series of check valves and a normally locked open (with AC control power removed) motor operated valve. The motor operated valves, MOV 841 and NOV 865, are not required to function to mitigate the consequences of the accident [Flood-1] .
4. The main steam isolation valves 3516 and 3517 close upon receiving a high containment pressure signal and the main and bypass feedwater control valves 4269, 4270, 4271 and 4272 close upon receiving a safety injection signal. The SI signal also causes a trip of the main feedwater pumps (which in turn causes the closing of the feedwater discharge valves). All of this equipment will fail in its safety position on loss of electrical power.
5. "Containment Isolation" and "Containment Ventilation Isolation" (ref erred to collectively as simply, "Containment Isolation" ) is initiated by the saf ety injection signal.

Containment isolation is discussed in detail in Section 5.2 of the FSAR. Most of the containment isolation valves are air operated valves. All air operated containment isolation valves close with safety injection signal with the exception of valves 4561 and 4562 which open full to insure service water supply to the containment recirculation fans. The f ail saf e position of the valves is the desired safeguard position as described above.

Six motor operated valves (313, 813, 814, ATV-1, ATV-2, ATV-3) receive a containment isolation signal. All of these valves are located outside of containment and only valves 313, 813, and 814 are fed from the safeguards buses.

During normal operation ATV-1, ATV-2, and ATV-3 are closed with blank flanges installed on their respective penetrations inside containment. The use of the process lines associated with these valves occurs only during the containment building integrated leak rate tests.

Valve 313, the reactor coolant pumps seal water return line, and valves 813 and 814, reactor coolant support inlet and outlet lines, are closed by the containment isolation signal.

~

6. The SI signal trips the reactor and turbine.

Other reactor trips are discussed in the FSAR, Section 7.

7. The reactor coolant pumps are tripped by manual operator action when low pressurizer pressure (1715

.psig) is reached, and SI flow is initiated.

8. Selected valves throughout the plant provide flow paths for the required safeguards equipment with the advent of the SI signal.

During normal operation all required valves in the flow paths for high head safety injection 'are normally open with the exception of valves 826A and 826C, the dis-charge valves from the boric acid storage tank to the suction of the safety injection pumps.

Valves 826A; B, C and D receive the safety injection signal and valves 82 6A and C open providing borated water to the reactor coolant loop cold legs.

When the level in the boric acid storage tank decreases to the 10% level, suction for the high head safety in-jection pumps is automatically switched from the boric acid storage tanks to the refueling water storage tank by the automatic opening of, valves 825A and B and closing of valves 826A, B, C and D.

During normal operation, all valves in the flow paths for low head safety injection are normally open except

for MOV 852A and MOV 852B, the valves in the vessel upper plenum injection lines. These valve's open upon receipt of a safety injection signal and remain open

- thereaf ter.

The containment spray pumps will automatically start and the discharge valves 860A Bg C and D automatically open, receiving power from the safeguards buses when containment pressure reaches 30 psig. If containment pressure does not reach 30 psig, the operator may manually start the spray pumps after all other safeguards are loaded on the safeguards buses. Automatic NaOH addition via opening of valves HCV 836A, B takes place two minutes after containment spray pump start unless defeated manually.

The containment spray pumps are normally aligned to the refueling water storage tank with all suction valves.

open.

SI system actuation will automatically align the two post accident charcoal f ilters to the containment recirculation system by opening inlet dampers 5871 and 5872, and outlet dampers 5873 and 5874. Loop entry dampers 5875 and 5876 will close. These dampers will fail to their safeguards position upon loss of electric power.

9. The control room ventilation is automatically placed in the 100% recirculation mode ( with about 25%

flow through charcoal filters), when SI is initiated.

10. Af ter the safety injection pumps are automatically switched from the boric acid storage tanks to the re-fueling water storage tanks, the operator resets safety injection, starts the component cooling water pumps and aligns flow to the RHR heat exchangers, and initiates SW flow to the'CW heat exchangers. At the 31% RWST alarm, the operator shuts off one CS and one SI pump (if more than one are running). When the refueling water storage tank level is reduced to 10%, the plant operator stops the remaining residual heat removal, containment spray and high head safety injection pumps and establishes f low paths to the reactor vessel for both high ( if required) and low head safety injection from containment sump B.

The normal (non-saf ety grade) auxiliary f eedwater supply source is from the condensate storage tanks. If this supply is exhausted the operator opens the motor operated valves 4027 and 4028 and manual operated valves 4344 and 4345 to provide service water to the suction of the auxiliary feedwater pumps. If the AFW system is not functioning properly, the operator can align from the control room the Standby AFW system to the steam generators ( using'ervice water suction).

11. In the recirculation phase, the operator aligns the RHR pumps to containment sump B by opening valve 850A for pump A and valve 850B for pump B, and closing 10

valve 704A, 704B, 856, and 896A or 896B. For low head recirculation, injection is through the vessel nozzles.

,For high head recirculation, the RHR pumps discharge to the safety injection pumps through alignment of valve 857A (for RHR pump B) and/or valves 857B and 857C (for RHR pump A). Valves AOV 897, 898 are closed. The high head safety injection pumps then provide water to the cold leg injection points. This alignment also allows CS pump operation, if desired.

Long term recirculation to compensate for the possible effects of boron precipitation has been described in Ref [Flood-1] and includes the use of RHR pumped flow to the vessel nozzles and through a high head safety injection pump into either cold leg.

Post-accident reactor coolant and containment atmosphere sampling modifications are presently being undertaken, in accordance with the implementation schedule for the TMX Lessons Learned commitments. See [Ref TMI-3].

Events Accom an in a Main Steam Line Break or a Main Feed Line Break The analyses of a main steam line break or a main feed line break and the consequences thereof have been discussed in Chapters 6 and'14 of the FSAR and in References [SLB/FLB 2-4]. The High Energy Line Break analyses [HELB 1-7] provide additional information regarding steam line breaks outside of containment, as 11

well as feedwater line breaks inside and outside containment.

Figure 2 illustrates the sequence of events required to mitigate the consequences of a main steam line break.

The same initial sequence of events would occur for a feedwater line break. Since the same equipment is re-quired to operate and the same emergency procedure is used following a feedline break as a steam line break, but a steam line break is a more severe accident in terms of RCS 4

cooldown (return to criticality) and mass and energy release to containment, the subsequent discussion will address the main steam line break only.

Table 2 lists the required equipment for each numbered block in Figure 2.

1. A large main steam line break ( greater than approxi-mately one square foot) would first be detected by the low steam line pressure sensors. Low steam line pres-sure sensed by two out of the three steam line pressure transmitters initiates safety injection accompanied by reactor and turbine trip .

la. Diagnostic instrumentation is available to the operator to distinguish among accidents, as described in the LOCA discussion.

2. Two out of three low pressurizer pressure signals would provide additional protection for a larger steam line break and also provides the initial safety injec-12

tion signal for smaller breaks. Also, high. containment pressure ( 6 psig) will initiate safety injection.

3. The Ginna design includes non-return check valves in each steam line just upstream of the main steam header in the intermediate building. Thus for any break upstream of the check valves, which includes all breaks inside containment, the check valves will preclude blowdown of the intact generator. Reactor trip will result in closing the turbine stop valves. As redundant protection in the event of a steam line break upstream of the check valves, and for all breaks downstream of the check valves, the main steam line isolation valves are closed by several signals. These signals include 2/3 high containment pressure (20 psig); 1/2 high steam flow in either steam line plus 2/4 low Tave plus safety injection; and 1/2 high-high steam flow in either steam line plus safety injection.
4. The safety injecti~on signal closes the main and bypass f eedwater control valves, trips the f eedwater pumps and closes their respective discharge valves.
5. The safety injection signal initiates containment isolation and containment ventilation isolation as described in the sequence of events in the loss of coolant accident.
6. The safeguards sequence as described in the loss of coolant accident is initiated by the safety injection signal. ( For steam breaks outside containment, the spray pumps are not required.)
7. The safety injection signal trips the reactor and turbine. Other reactor trips are discussed in the FSAR, Section 7.
8. The reactor coolant pumps'are tripped by manual operator action when low pressurizer pressure (1715 psig) is reached, and SI flow is initiated.
9. All valves associated with the safety injection systems are aligned and automatically function as de-scribed in the loss of coolant accident discussion. If high containment pressure of 30 psig is reached, the containment spray system operates as described in the LOCA discussion.
10. When the boric acid storage tanks are drained to the 10% level and safety injection pump suction has automatically been aligned to the refueling water storage tank, the operator will reset safety injection and if reactor coolant pressure is above the shut-off head of the RHR pumps, will stop the RHR pumps and place them in the standby mode.

A high steam line flow and/or low steam line pressure will indicate to the operator which steam generator has the steam line break. When this has been determined, 14

the operator will terminate AFW flow to the faulted steam generator, and align/maintain flow to the intact steam generator.

The inventory of the reactor coolant will be maintained by the remote manual operation of the high head safety injection pumps in combination with use of the charging p Umps ~

At least two hours after the start of the accident, supply water for the auxiliary feedwater pumps can be manually transferred from the condensate storage tanks to the service water system, by the method described in the LOCA discussion [See Ref . SLB/FLB-6] . If the auxiliary feedwater system is not operating properly, the operator can initiate operation from the control room of the Standby AFW system (using service water suction).

11. If conditions and equipment availability permit, the operator can begin a gradual cooldown and depressuri-zation to cold shutdown conditions. However, the primary safety function is to maintain the RCS in a safe condition at all times, removing decay heat at a rate comparable to the generation rate. Maintenance of this safe shutdown condition is accomplished by a combination of steam dump ( to the condenser or atmosphere) with primary and secondary inventory makeup, accomplished by use of the safety injection and/or the charging 15

I pumps, and the auxiliary feedwater system. It is expected that RCS temperature can be lowered to near 212'F by using the steam generators. The safe shutdown conditions can be maintained until a final cooldown and depressurization to ambient conditions can be effected.

C. Hi h Ener Line Breaks Outside Containment An analysis has been provided describing the effects of pipe breaks outside containment [HELB-1]. The report proposed a program of augmented inservice inspection of certain piping welds in order to preclude the necessity to address further full diameter high energy piping breaks. Credible breaks of main steam lines outside containment, that is, those not included in the inspec-tion program, are bounded by a 6 inch main steam line branch connection in the Intermediate Building and a 12 inch main steam line branch connection in the Turbine Building. Credible breaks in the feedwater lines outside containment are bounded by a break in the 20 inch feedwater line in the Turbine Building. The accident environment created by these breaks, and other postulated breaks are provided in References [HELB 8-11]. The program has been accepted by the NRC [Ref.

HELB 7,8]. Several modifications have been performed at the Ginna Nuclear Plant as a result of high energy line break analyses. Reference [HELB-1] discusses the various modifications, but of particular note is the Standby Auxiliary Feedwater system modification. A 16

remote-manual controlled standby auxiliary feedwater system, identical to the auxiliary feedwater system in cooling capability, has been installed. The pumps are housed in a seismically designed structure (area 6 Figure 3) remote from the auxiliary feedwater and any high energy lines. Any portion of this system required to operate in an emergency is not subjected to an adverse environment. Ref [HELB-8] includes the NRC's Safety Evaluation Report concerning the RGGE modifications resultant from the review of Ref. [HELB-1]. It includes a discussion of the acceptability of the instrumentation relocation and cable re-routing performed to insure that sufficient equipment will be protected from the environmental effects of a HELB outside containment.

The failure of steam heating lines in the Auxiliary Building was identified and discussed in Ref . [HELB-1].

It has been determined that steam heating lines also traverse other areas in the vicinity of safety related equipment [Ref. HELB-15]. Modifications are planned which will isolate the steam heating line to the affected areas in the event of a failure and therefore preclude an adverse environment. The commitment to perform analyses/modifications for those pipe breaks outside containment are given in Reference [HELB-13]. Prior to its installation, regular inspections are being performed to reduce the likelihood of a failure creating an adverse environment. These inspections, performed 17

during each plant operating shift, would detect any leakage. Plant procedures (T-35F, "Steam to Auxiliary Building, Screen House, or Diesel Generators and Oil

- Room" ) call for isolation of the affected piping promptly upon detection of the leakage.

18

III. IDENTIFICATION OF THE LIMITING SERVICE ENVIRONMENTAL CONDI-TIONS FOR EQUIPMENT WHICH IS REQUIRED TO FUNCTION TO MITIGATE THE CONSEQUENCES OF DESIGN BASIS EVENTS This Section of the report defines the bases for and references to the environmental conditions encountered throughout the plant. A tabular summary is provided in Table 4.

A. Inside Containment Post accident containment environmental conditions are discussed in Appendix 6E of the Ginna FSAR. These conditions result from a loss of coolant accident. The temperature and pressure profiles are given in Figures 1 and 2 of Appendix 6E with peak values being 286'F and 60 psig respectively. The radiation profile is presented in Figures 4 and 5 of Appendix 6E and it is seen, for example, that the doses at 30 minutes and one year following a LOCA are 1.7 x 10 6 and 1.6 x 10 8 rads, respectively. (These figures are repeated as Figures 4,5,and 6 of this report.) Materials compatibility with post-accident chemical environment is discussed in detail in Appendix 6E. 100$ humidity is assumed.

Design parameters 'for environmental conditions have been conservatively selected for Ginna. As seen in FSAR Figure 14.3.4-2, the calculated peak pressure is less than 53 psig while the design value is 60 psig.

The duration of the peak, similarly, bounds the cal-culated values.

19

I Another example of the conservatism employed is seen in the accident radiation environment used for design purposes. As noted in WCAP 7744, a release of 100% of the noble gases, 50% of the halogens, and 1%, of all remaining fission products is assumed. In addition, no credit is taken for removal of radioactivity from the containment atmosphere by sprays, filters and fission product plateout. Finally, the specific activity in containment was roughly doubled by assuming a contain-ment free volume associated with an ice condenser con-tainment. Thus the radiation environment clearly over-states that which would be present even in a minimum safeguards case. This conservation is apparent from a comparison to the DOR Guidelines, which suggest a post-LOCA integrated dose of 2 x 10 7 rads gamma.

Submergence of valves inside containment. has previously been discussed in Reference [Flood-4] and it has been shown that operation following submergence is not required. Submergence of instrumentation has been discussed in Ref [Flood-5]. Since the instrumentation is not required to function while flooded, no qualification for submergence is specified (see e.g.,Section IV.19 of this report) .

The peak pressure following a MSLB is given in Section 14.2.5 of the FSAR as 52 psig, assuming no credit for containment pressure reducing equipment. Recent analyses 20

for other facilities indicate that the containment vapor temperature following a MSLB in contaiment may briefly exceed those derived for a LOCA. These higher temperatures should not be limiting, however, for qual ification of equipment required fol lowing a MSLB, because:

1) the fact that the high temperature transient. is very brief and there is superheated steam (with its lower heat transfer capability) as opposed to saturated steam,
2) the equipment is protected from the direct effects of the steam line break by concrete floors and shields, and
3) the sensitive portions of the electrical equipment are not directly exposed to the environment, but are protected by housing, cable jackets, and the like.

For these reasons, the humidity and steam environment following a LOCA remains limiting. This is consistent with the NRC's position 4.2 of the "Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors;" Radiation levels in containment following a MSLB are not limiting since fuel failures are not projected to result from a MSLB. Chemical environment and submergence are bounded by the LOCA conditions.

21

B. Auxiliar Buildin The auxiliary building has a HVAC system which provides clean, f iltered and tempered air to the operating floor of the auxiliary building, and to the surface of the decontamination and spent fuel storage pits. The system exhausts air from the equipment rooms and open areas of the auxiliary building, and from the decon-tamination and spent fuel storage pits, through a closed exhaust system. The exhaust system includes a 100 percent capacity bank of high efficiency particulate air (HEPA) filters, and redundant 100 percent capacity fans discharging to the a'tmosphere via the plant vent.

This arrangement insures the proper direction of air flow for removal of airborne radioactivity from the auxiliary building.

Included in the auxiliary building exhaust system is a separate charcoal filter circuit, which exhausts from rooms where fission product activity may accumulate, during normal plant operation, in concentrations exceeding the average levels expected in the rest of the build-ing. Following a loss-of-coolant accident, this circuit is capable of providing exhaust ventilation from the areas containing pumps and related piping and valving which are used to recirculate containment sump liquid.

A full flow charcoal filter bank is provided in the circuit, along with two 50 percent capacity exhaust 22

Vg fans. The air operated suction and discharge dampers associated with each fan are interlocked with the fan such that they are fully open when the fan is operating and fully closed when the fan is stopped. These dampers fail to the open position on loss of control signal or control air. The fans discharge to the main auxiliary building exhaust system, containing the HEPA filter bank. To assure a path for the charcoal (and HEPA) filtered exhaust to the plant vent if, the main exhaust fans are not operating, a fail open damper is installed in a bypass circuit around the two main exhaust fans.

The residual heat removal, safety injection, containment spray and charging pump motors are provided with addi-tional cooling provisions to maintain ambient temperatures within acceptable limits when'the pumps are operating.

The charging pumps and RHR pumps are located in their own rooms, each room being provided with two cooling units consisting of redundant fans, water-cooled heat exchangers, and ductwork for circulating the cooled air. The capacity of each unit is sufficient to maintain acceptable room ambient temperatures with the minimum number of pumps required for system operation in service.

The safety injection and containment spray pumps are 0

provided with cooling units providing cool air directly to the motor. There is a separate fan for each of the motors .

23

In the event of a loss of offsite power, the auxiliary building ventilation system main supply and exhaust f ans would be inoperable. However, all other fans in the auxiliary building ventilation system are supplied by emergency diesel power including the charcoal filter circuit and the pump cooling circuits for safety related pump motors, as described above. Since the auxiliary building is a very large volume building, it

'\

is not expected that there would be a post-accident tempera-ture increase except in some local areas near hot piping and large motors. This situation exists only in the basement of the auxiliary building where the safety-related pumps and recirculated sump fluid piping are located. As shown in Reference [HELB-14] the ventila-tion system for these areas is expected to be adequate to maintain the post-accident temperature with normal "ambient" levels. Further detailed evaluation of the environment in these areas is being addressed with the final resolution of the "mild" environment qualification requirements .

The radiation levels in the auxiliary building will increase in the event of a LOCA. Using very conservative post-accident fission product activity levels, the post-accident environment in the auxiliary building was calculated in Appendix A to Reference [TMI-3]. It is apparent from Table 5-1 of this reference that the only major radiation field in terms of equipment qualification 24

will be in the vicinity of the recirculated fluid. The required qualification doses are addressed for all the affected equipment in Table 3. The RGEE commitments to

- ensure that a HELB in the auxiliary building will not affect the capability of effecting and maintaining a safe shutdown condition is provided in Reference [HELB-13].

Flooding is not a concern in the Auxiliary Building.

Even in the event of leakage, two 50 gpm sump pumps are provided in the low point of the*building. This is described in Section 9.3 of the FSAR, and has been evaluated by the NRC in Reference [HELB-15].

Intermediate Buildin Implementation of an augmented inservice inspection program for high energy piping outside containment has reduced the probability of pipe breaks in these systems to acceptably low levels [Ref . HELB-7, 8] . A six inch main steam line branch connection is the intermediate building DBE. Based on the f ailure capacity of portions of the exterior walls, the limiting pressure is established in Ref . [HELB-1] as being a pressure of 0.80 psig.

Assuming saturation conditions, one obtains a limiting I

'I temperature of approximately 215'F. A 100% humidity steam-air mixture is assumed. If the pipe crack or branch line break were in a portion of the steam or

~ ~

f eed line that could be isolated, the isolation would immediately halt the mass and energy addition to the intermediate building. A pipe crack or branch line 25

which could not be isolated is the limiting DBE for intermediate building environment. Mass and energy release in this case would be limited by the dryout of the steam generators with the duration of the environment dependent on the size of the leak or break. Based on flow through a main steam safety valve (a 6 inch line) of 247 lbs/sec at a steam line pressure of 1100 psia and the inventory available for release from a main steam break of 165,500 lbs ( FSAR Section 14.2.5), the mass and energy flow will continue for at least 11 minutes. Smaller leaks may continue substantially longer. Zt is expected that within 30 minutes to an hour, action could be taken to provide added ventilation to the building by opening doors. Within several hours, return to near ambient could be accomplished.

Table 4 provides an estimate of the duration of the environmental transient expected. The exact duration is not critical in terms of affected equipment qualification; therefore, no explicit calculations have been performed. Chemical spray is not a design consider-ation in this building. The effects of submergence need not be considered, as described in References

[HELB-1], [HELB-4], and [FLOOD-11']. This latter reference presents the result of an analysis performed to ensure that safety-related equipment would not be flooded in the event of an feed line break in the intermediate building.

26

The radiation environment was reviewed in response to the TMI Lessons Learned commitments [see Ref . TMI-3] .

It can be seen from Table 5-1 that the radiation environ-ment is not significant in terms of equipment qualification.

Cable Tunnel Since the cable tunnel is open to the Intermediate Building, the limiting environmental conditions for the cable tunnel are identical to the Intermediate Building conditions.

Control Buildin The limiting environment of the Control Building which includes the control room, relay room, and battery rooms, is normal ambient conditions. Protection against high energy line breaks and circulating water line breaks which could occur outside the Control Building and affect the Control Building environment are identified and discussed in References [HELB-1, HELB-6, HELB-7, HELB-13, HELB-15, FLOOD-1, and FLOOD-5] .

The air conditioning system for the control room is described in Section 9.9 of the FSAR. The main air handling unit and circulation fans for the control room are powered from a single Class IE motor control center

( MCC-1K), which receives power from a diesel-backed emergency bus (diesel 1A). If there were a failure of this train during the post accident period, it would be possible to crosstie to the 1B diesel. The operator, after assuring that any faults are cleared, would close 27

the bus tie between buses 14 and 16'to energize the in-operable-Control Room air handling unit from the 1B diesel, while making sure that the operational diesel

- does not become overloaded. This emergency bus cross-ties procedure has previously been included in the Ginna Emergency Procedures .

The control room HVAC system has been out of service several times in the last 11 years for maintenance. A satisfactory environment has been maintained by opening the two control room doors and two relay room doors, connecting the two rooms together and with outside air, to provide natural circulation. Equipment failure has never been experienced during these events because of a temperature increase due to lack of HVAC.

It is also possible, of course, to provide for the use of portable air-conditioning units or fans to maintain environmental conditions within proper specifications.

Further evaluation of the long-term effects of the loss of ventilation will be made at a later time, when safety-related equipment not exposed to a "harsh" accident environment is addressed in terms of environmental qualification.

The relay room is normally cooled by two non-safety-related air conditioning systems, which can be manually aligned to the emergency buses by closing the proper bus-tie breakers.

28

Natural circulation with the control room, and the use of portable air-conditioning units and fans, are options available to maintain environmental conditions within the required specifications. Further evaluation con-cerning loss of ventilation will be made at a later time, together with the control room study.

To further assure that a loss of ventilation to the control and relay rooms is not expected to be a concern, RG&E conducted an 8-hour test on September 15, 1980.

It was demonstrated that, for a loss of all HVAC, no, significant temperature increase occurred in the control room or relay room. Only the plant computer, located in its own room within the relay room, and not required for accident mitigation or safe shutdown, appeared to be susceptible to overheating.

The battery rooms have a set of inlet and exhaust fans, as well as an air-conditioning system. Additional fans are to be installed in the near future. These fans will be d.c.-powered 'directly from the batteries.

While this modification is in progress, the present Emergency Procedures provide for manual alignment to the emergency buses by closing of bus-tie breakers. If necessary, portable fans could be used to provide sufficient air handling capacity to maintain the battery rooms at acceptable ambient conditions.

29

F. Diesel Generator Rooms The emergency diesel generator rooms each have their own HVAC system, powered from the diesels. As soon as the diesels are brought up to speed, stabilized, and their respective circuit breakers closed to their emergency buses, the HVAC systems ( ventilating fans) are energized. Protection against failure of steam heating lines in the rooms is described in Section II.C above. Failure of a steam heating line would affect only one diesel. The other diesel, as well as offsite power, would still be available. This configuration has been reviewed by the NRC in Reference [HELB-15], ~

and found acceptable. Protection agains events outside the rooms is described in References [HELB-1, HELB-6, HELB-7, FLOOD-1, and FLOOD-5]. The limiting environment in the diesel generator rooms therefore is normal ambient conditions.

G. Turbine Buildin The turbine building does not require an HVAC system per se, but rather utilizes roof vent fans, wall vent vans, windows and unit heaters for control of the en-virons. In the event of loss of power to fans in this building there would be no significant temperature rise, since it is a large volume building with sufficient openings ( windows and access doors) to adequately cir-culate outside air.

30

Analyses have shown that the limiting pressure are caused by an instantaneous break in the 20 inch feed line in the turbine building. See Reference [HELB-1].

Peak pressures are 1.14 psig on the lower two levels of the building and 0. 70 ps ig on the operating floor.

Failure of portions of the exterior wall limit the duration of the pressure pulse to,a f ew seconds.

Pressure and temperature is limited by the failure capacity of the exterior walls. Assuming saturation conditions, one obtains a limiting temperature of approximately 220'F. A 100% humidity steam-air mixture is assumed. Isolation of the main steam and feed system will isolate the source of energy to the turbine building. Temperature and pressure reduction will be accomplished by opening exterior doors and windows and as a result of leakage through known openings to the outside. For conservatism, it has been assumed that the peak temperature condition persists for 30 minutes with return to ambient being accomplished in a total of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. For conservatism, peak pressures are assumed to persist for 1 minute with return to ambient being accomplished in a total of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. (This is tabulated in Table 4). The exact duration of high environmental 31

conditions is not critical in terms of affected equipment qualification; therefore, no explicit calculations have been performed.

Limiting flood conditions are the result of a circulating water system pipe break and is a water level of 18 inches in the basement [FLOOD-5].

Auxiliar Buildin Annex This structure, which houses the Standby Auxiliary Feedwater System, is described in References [HELB-1]

and [HELB-6] . The limiting environment in this structure is normal ambient conditions. The cooling system for this building is redundant and seismically qualified.

Flooding is not a concern since all safety-related equipment associated with the Standby AFW System is elevated so that a complete failure of the Condensate Tank would not cause submergence.

Screen House The screen house, like the turbine building, does not require an HVAC per se, but utilizes roof vent fans, wall vent fans, windows, and unit heaters for control of the environs. Xn the event of a loss of power to the fans, there would be no significant temperature rise, since it is a large volume building with suf f icient openings to adequately circulate outside air.

32

RG&E's commitment to resolve the HELB environment is provided in Section II. C. Protection against f looding is described in Ref erences [FLOOD-1] and [FLOOD-5] .

The, limiting environment in the screenhouse is thus normal ambient conditions.

33

IV. EQUIPMENT QUALIFICATION INFORMATION Table 3 summarizes the qualif ication information of required electrical equipment. This section provides the detailed background information, with emphasis on a response to the August 20, 1980 FRC Draf t Interim Technical Evaluation Report, Project C5257. For this reason, the paragraphs are ordered consistent with Section 3 of that report.

1. TER Paragraph 3.2.1 Table 3 Item No. 23. Main Steam-line Pressure Transmitter in the Intermediate Building.

TER C5257 noted that this instrumentation meets the DOR Guidelines. In order to provide instru-mentation with all of the proper qualification documentation, there are plans to replace these transmitters by June 1982. Qualification docu-mentation will be made available when received.

2. TER Paragraph 3.2.2 Table 3 Item Nos. 31, 41. Medium Voltage Switchgear Located Outside Containment ( Models DB-50A and DH-350E).

TER C5257 found these acceptable, since the breakers are exposed only to a relatively mild (1 psig, 220'F) environment, must function within a short time (generally seconds) and fail-safe on loss of power. No additional information is'onsidered necessary to show proper operational capability under the required accident conditions.

34

I

3. TER Paragraph 3. 2. 3 Table 3 Item No. 21A. Containment Pressure Transmitters located outside containment.

TER C5257 found that these transmitters satisfied the DOR Guidelines. In light of TMI Lessons Learned, f ive of the seven transmitters, which could see a high radiation field following a LOCA, are being replaced with new transmitters ( three will have a 10-200 psig span and provide post-accident monitoring). These transmitters will be qualified for the post-LOCA environment and will therefore be qualified for a HELB outside containment environment. All 5 will be replaced by June 1982.

Qualification documentation will be made available when received. The two transmitters not being replaced are not exposed to a harsh environment as the result of a LOCA. For a high energy line break outside containment, these two transmitters are not required to function.

4. TER Paragraph 3.2.4 Table 3 Item No. 25 BAST Level Transmitter in the Auxiliary Building.

TER C5257 found that these transmitters met the intent of the DOR Guidelines. It is important to note that, this instrumentation performs'its safety function following a LOCA or steam line break prior to the time any accident environment is encountered in the Auxiliary Building. For a HELB

in the Auxiliary Building, there is no need for the BAST level transmitters to function. No additional information is required for this equip-ment.

5. TER Paragraph 3.2.5 Table 3 Item No. 18. RWST Level Transmitter in the Auxiliary Building.

I TER C5257 notes that this item satisfies the intent of the DOR Guidelines. For f urther assurance, this transmitter will be replaced by June 1982 with a f ully-qualified transmitter. Qualif ication

-

documentation will be made available when received.

6. TER Paragraph 3. 2. 6 Table 3 Item No. 19. RWST Level Switch in Auxiliary Building.

TER C5257 notes that this item does not require environmental qualification, since the safety function is performed prior to the onset of an adverse environment. This is correct; for added assurance of post-accident monitoring, however, this item is being replaced by June 1982.

Qualification documentation will be made available when received.

7. TER Paragraph 3. 3. 1. 1 Table 3 Item No. 8A. Valve Operators for Valves MOV 841, 865.

TER C5257 concludes that, since these valve actuators are locked in the "open" position with power removed with no need to function, lack of valid 36

qualification documentation is a moot point.

Thus, no qualif ication information is required for this item.

8. TER Paragraph 3. 3. 1. 2 Table 3 Item Nos. SF, SG.

Valve Operator for MOVs 851A, B; 878 B, D.

TER C5257 concludes that, since these valve actuators

)

are locked in the "safety" position, with no need to function, environmental qualification is a moot point. Thus, no qualification information is required for this item.

/

9. TER Paragraph 3. 3. 1. 3 Table 3 Item No. SC. Valve Operators for MOVs 825 A, B.

As noted in TER C5257, these valves perform their safety function (open to allow RWST fluid to the suction of the SI pumps) prior to the time an adverse environment would exist in the Auxiliary Building due to sump recirculation. No "harsh" environmental qualification is required for these items.

10. TER Paragraph 3.3.1.4 Table 3 Item No. SD. Valve Operators for MOVs 4027, 4028, 4007, 4008, 4000A, 4000B.

As noted in TER C5257, these valves would not be used in the .event of a HELB in the Intermediate Building. RGGE Emergency Procedures specifically call for actuating the Standby Auxiliary Feedwater 37

i' System in the event the AFW system is inoperable.

Since none of the S tandby AFW system components will be e xp osed to a HELB, it is concluded that this system will be suff icient to provide the needed saf ety f unction. No "harsh" environmental qualification for the AFW valves ves xs needed.

11. TER Para g ra p h 3 .3.1.6 Tables 3 Item Noo. 11 . Auxiliary Feedwater Pump Motors.

As noted in TER C5257 thhese pumps are not required to function in the event of a HELB in the Xnter-mediate Building. Thee S tandby AFW System performs the required safety function P roce d ures call for removing the AFW p um ps from the safety-related bus, prior to connecting the standby system.

Mechanical interlocks ensure that both sets of pumps cannot be powered from th d'iesels concurrently.

No "harsh" environmental qualif ication for the auxiliary feedwater pumps is required.

12. TER Para g ra p h 3 .3.2.1 Table 3 Xtem No. 8E. Valve operators for MOVs. 850 A, BE 856 '57 Ag BJ C 860 Ai Ci Documentation Reference 53 su b mitted to the NRC on September 24 1 980, provides a ref erence to Limitorque Re p ort B 0003. This reference provides assurance that these valves will perform their safet functi'on. Additional information from 38

Limitorque Report B0058 has be'en added to Reference 53, documenting Limitorque's use of generic quali-f ication to qualif y multiple size actuators by one type test.

13. TER Paragraph 3.3.2.2 Table 3 Item No. 8H Valve Operators for MOVs 852A, B.

TER CS257 notes that these valve actuators are not acceptable for long-term service in an accident environment, and are not qualified for submerged operation. Qualification for short-term post-LOCA operation is shown in Reference 18, however. The function of these valves is to open upon receipt of an SI signal, and then to remain open. Quali-f ication for submerged operation is not required.

Submergence could occur unless the saf ety f unction of the valves has already occurred. Specif ically, to submerge these valve operators, the entire contents of the primary system, the entire contents of both accumulators, and a portion of the water in the refueling water storage tank must be discharged to the containment. For this to occur, however, a safety injection signal must have occurred and the valves must have opened.

RGSE has incorporated modif ications to these valve operators to prevent undesired operation in the event of submergence. The details of these 39

modifications were provided in References [FLOOD-2, FLOOD-3], transmitted to FRC on May 29, 1980. It is thus considered that these valves are qualified to perform their required safety function.

14. TER Paragraph 3. 3. 2. 3 Table 3 Item No. SI. Valve Operators for MOV's 9703A,B; 9704A,B; 9710A,B in the SAFN System.

All of these valve operators are located in the Auxiliary Building Addition, which is a "mild" environment. Environmental qualif ication is provided under paragraph 4.3.3 of the "DOR Guide-lines", Areas Normal l Maintained at Room Conditions.

The Auxiliary Building Addition is maintained at room conditions by redundant air conditioning systems served by the onsite emergency electrical power system. The room conditions specified in Reference 43 are 60-120'F. The valve specification (Reference 54) states that "the valve actuator shall be designed for a 40 year plant life under ambient conditions of 40F to 120F..." Since there is no change in the environmental conditions between normal and accident conditions, "...no special consideration need be given 'to the environ-mental qualification of Class IE equipment in these areas provided the aging requirements discussed in Section 7.0 are satisfied and the areas are maintained at room conditions by redundant air 40

conditioning or ventilation systems served by the onsite emergency electrical power system". Reference 47 describes the program developed at R. E. Ginna for detecting age-related failures. This program was developed to conform to the provisions of Section 7.0 of the "DOR Guidelines" for the "ongoing programs...to review surveillance and maintenance.

records to assure that equipment which is exhibiting age-related degradation will be identified and replaced as necessary".

15. TER Paragraph 3.3.2.4 Table 3 Item No. 13A. Crouse-Hinds Electrical Penetrations r

.

TER C5257 notes that the Brunswick tests could not be substantiated, since no test description was provided. Reference 45 provides this description.

Reference 58 is a letter from Westinghouse stating that the Brunswick data is applicable to qualify the seal, canister, and internal connections.

Reference 54 is an evaluation of the capability of the Ginna penetrations to perform their function under elevated and short-circuit electrical loading conditions.

Further, an evaluation (Reference 59) of the functions of the various materials in the penetra-tions disclosed that the organic compounds, which are possibly subject to aging or radiation effects, 41

do not perform any critical insulating or sealing functions. These functions are performed by ceramic and metallic components.. This evaluation augments the qualification testing performed on these penetrations, confirming that they are N

qualified to perform their safety function.

16. TER Paragraph 3.3.2.5 Table 3 Item No. 13B. Westinghouse Electrical Penetrations .

It is noted in TER C5257 that additional information concerning the "similar resin", aging characteristics of the insulation on the cable leads, and qual ified life should be provided.

II Ref erence 61, Research Report 75-7BS-BIGAL-122, shows that the lower 95%

conf idence band on qual ified life at 105 'C is greater than 40 years. Also, the author of this report, Mr. J. F. Quirk, has stated that the word "similar" had been used only in the respect that test results of this epoxy were close to the results of other epoxies also being tested. The, epoxy in the Ginna penetrations is identical to that tested. Cable lead insulation aging data is also included in Reference 61.

It can be concluded that these penetrations are suitable to perform their required safety functions.

42

17. TER Paragraph 3. 3. 2. 6 - Table 3 Item No. 14. Westinghouse Terminal Blocks Inside Containm'ent.

TER C5257 found that, although qualification for pressure, temperature, and humidity is acceptable, additional information is needed concerning thermal aging and radiation. Reference 60 is a Proprietary Westinghouse R&D Report ( 077-7B7-CBSEL-R3) dated July 13, 1977. It shows that for a criteria of f ailure of 50% of the original flexure strength and impact strength, the 40 year life extrapola-tion is approximately 120'C. This report, is not yet in our possession, but may be audited at the Westinghouse facility.

Additional information -concerning radiation sus-ceptibility of the terminal blocks is also provided in Reference 60. It is shown that the qualification level is 2 x 10 7 rads. Although not meeting the long-term conservatively calculated radiation dose for Ginna of 1. 6 x 10 8 rads, the DOR Guideline values are met. Based on the protected location

'

of these terminal blocks, 2 x 10 7 rads is considered adequate. A detailed evaluation of this post-LOCA radiation dose will be'ade. If the required dose for the long-term monitoring function is greater, replacement or additional protection will be provided.

43

As presently installed, the terminal blocks for pressurizer pressure and level instrumentation would become submerged after a LOCA en qualified long-term monitoring instrumentation for these functions is installed at Gin irma, and elevated above the submergence level, the terminal blocks will also be el evated. Submergence and direct spray impingement will thus be precluded. See paragraphs 19 and 20 for a discussion of the pressurizer pressure and level instrumentation.

18. TER Paragraph 3.3.2.7 Table 3 Item Nos. 15A, B, C Kerite Cable Inside Containment.

Reference 51 is the "Cable Id t'f'n i z.cation and Qualification Supplement" Th'is ocument can be used to determine the identity of cable in use throughout the plant. It is shown that all power cable inside containment is Kerite. The most recent and comprehensive qualification testing of Kerite cable was performed in conjunction with the testing of Raychem sleeves (Reference 38). Reference 55 is a lett etter from Kerite verifying that the cable supplied'or the qualification testing in Reference 38 is identical to th a t orig>nally supplied and installed in the Ginna co t irma containment.

The pre-aging done for the Kerite cable during the Raychem sleeve test establish e d a 93 . 3 year life 44

at 140'F mean surface temperature. The Arrhenius data is conf idential to the manuf acturer, but is available at RG&E as Reference 63.

RG&E believes that this recent testing definitively demonstrates the adequacy of the Kerite cable for performing its required safety function.

There are no safety-related cables inside containment subject to flooding, which are required to perform a safety function during submergence. Qualification for submergence is thus not required.

19. TER Paragraph 3.3.2.8 Table 3 Item No. 22. Pressurizer Pressure Transmitters.

The deficiencies noted in TER C5257 included lack of radiation and submergence qualification. RG&E does not claim credit for the use of this instru-mentation at the time it would receive excessive radiation exposure, or become submerged. Ginna Emergency Procedures specify that, unless pressurizer pressure, level, and other parameters appear stable and are returning to prescribed levels, safety injection flow is not to be terminated.

Failure to terminate safety injection is not a safety concern. Therefore, lack of qualification for this instrumentation is not considered of immediate safety significance.

45

It is recognized, however, that accurate primary system information would be extremely useful to the operator for diagnosing the status of the plant during accident conditions. RG6E, therefore, plans to replace the present instrumentation by June 1982 with f ully-qualified transmitters, located above any possible submergence level.

Qualification documentation will be made available when received.

20. TER Paragraph 3. 3. 2. 9 Table 3 Item No. 24. Pressurizer Level Instrumentation.

The same information as described in 19 above for the pressurizer pressure instrumentation applies to this instrumentation.

21. TER Paragraph 3.3.2.10 Table 3 Item No. 30. Fan Cooler Motors Inside Containment.

TER C5257 concluded that in addition to the information provided in References 18 and 2 0, information needed for complete qualification of the fan cooler motors is a) documentation regarding qualification of motor-lead and lead-to-cable splices, and (b) determination of a qualified life for the motor. Information regarding the splices is given in Reference 64.

46

Aging information for the insulating material of these motors, as well as the bearing lubricants, is given in Reference 18, Section 4. Aging to demonstrate 40 year continuous operation at 120'C was performed. This is consistent with the data given in Reference 67, and is considered sufficient to qualify the fan cooler motors for continued operation. A program at RG6E to maintain motor bearings and lubricants is given in Reference 65.

This program will ensure that the lubricants used are compatible with the environmental conditions which could occur during a DBE.

Additional information regarding qualification testing of the same type of motors is given in WCAP 7829, "Fan Cooler Motor Unit Test" (Reference

70) .
22. TER Paragraph 3. 3. 2.11 Table 3 Item No. 34. Raychem Cable Splice Sleeves.

TER C5257 states that RG&E should present evidence of similarity between the tested and installed equipment. This is'documented in the detailed evaluation and observation of the splice sleeve replacement program, given in IE Inspection Reports 78-20 and 78-21 (Reference 56).

It is also stated that a determination of qualified life should be made for the sleeves. The actual 47

test in Reference 38 established a 12.1 year life at 60'C ambient. This pre-aging was constrained by the concurrent aging of the Kerite cable, which was pre-aged for 93.3 years at 60'C by the same test. Based on proprietary Raychem information (included in Reference 63 and available for audit at RG6E) a 40 year life at 91'C can be expected..

Therefore, these sleeves are considered fully qualified.

23. TER Paragraph 3.3.2.12 Table 3 Xtem No. 20. Steam Flow Transmitters Enside Containment.

RG&E has stated that these transmitters are not required to perform a safety function at a time they could be exposed to a high energy line break environment. Thus, the lack of complete qualification documentation is a moot point for these trans-mitters. For a steam line break inside containment, the steam line non-return check valves will assure that the intact steam generator will not blow down. Steam line isolation would be provided by the high containment pressure signal.

For added assurance of steam line isolation in the event of a steam break'inside containment, these transmitters will be replaced by June 1982 with fully-qualified equipment. Qualification documenta-tion will be made available when received.

48

24. TER Paragraph 3.3.2.13 - Table 3 Item No. 21B. Contain-ment Pressure Transmitters in the Intermediate Building.

As noted in Section IV.3 of this report, five of the seven containment pressure transmitters, which could be exposed to high post-LOCA radiation levels, are being replaced with LOCA-qualified units by June 1982, in response to TMI Lessons Learned. Qualif ication documentation will be made available when received.

25. TER Paragraph 3.3.2.14 Table 3, Item No. 37, Hydrogen Recombiner Igniter Exciter TER C5257 requested that the effects of containment spray and thermal aging be addressed. This informa-tion has not yet been received. If proper documen-tation is not found concerning these environmental parameters, RG&E will commit to replace the necessary equipment. It is important to note that the present licensing basis for Ginna does not include the hydrogen recombiner as a means necessary for I

post-LOCA hydrogen control (see the RG&E "Technical Supplement Accompanying Application for a Full Term Operating License," August 1972,Section III.B.7).

26. TER Paragraph 3.3.2.15 Table 3, Item No. 38, Hydrogen Recombiner Blower Motor.

49

The only deficiency noted in TER C5257 is that no analysis exists comparing the impact of deviations between the test specimen specific design features, materials, and production procedure and those of the installed equipment. The only evidence at this time is contained in Section 5.2 of Reference 18, WCAP 7410-L, Vol. II. It is stated that "the 2 hp motor used in the test program is constructed in the same manner as, is the actual 15 hp motor used in the recombiner." Further, it has been verified that the Ginna 15 hp motor has Class H insulation, the same as the 2 hp motor tested.

Based on the available information, RG6E believes that there is reasonable assurance that the Ginna recombiner motor will perform its safety function.

Further, as stated in 25 above, the hydrogen recombiner is not required by the present Ginna design basis. Based on the TMI Lessons Learned, however, RGEE will commit to replace the motor if proper environmental qualification documentation is not established.

27. TER Paragraph 3.3.3.1 Table 3 Item No. 8B. Valve Operators for MOVs 826 A,B,C,D; 896 A,B.

The MOVs 826 A,B,C,D are located at the discharge of the Boric Acid Storage Tanks, and provide suction to the SI pumps in the event of a Safety 50

Injection signal. Upon low BAST level, these valves close af ter the 825 A,B valves open. The valves are located in the auxiliary building, and will have completed their function prior to the presence of an adverse environment caused by sump recirculation fluid.

MOVs 896 A,B are normally locked-open valves, located at the suction of the SI and CS pumps from the .EST. The valves are closed prior to the time sump recirculation is initiated. Therefore, these valves will have completed their function orior to the time an adverse environment would occur.

In the case of all six valves, environmental qualification for an adverse environment is not required.

28. TER Paragraph 3.3.3.2 Table 3 Item Nos. 1A, 1B, 1C,
5. ASCO solenoid valves.

The feedwater control and bypass valves ( items 1A, 1B) fail closed on loss of air. This is supported by Reference 23. In order to further ensure that these valves will perform their safety function when exposed to a HELB in the Turbine Building, the solenoids will be replaced with valves having proper qualification documentation. It is exoected that this can be accomplished by June 1982. The fail-safe closure of the valves ensures that the 51

required safety function can be performed until replacement can be effected.

Item 1C, the solenoid control ling LCV112B, will not experience an adverse environment during an accident. Further, an accessible manual bypass valve, valve 358, is used to provide alternative suction for the charging pumps from the RWST.

Since this function would not be required for many hours following an event requiring the maintenance of a safe shutdown condition, the use of this manual valve is considered acceptable. Item 1C will thus be deleted from Table 3.

Item 5A, the RHR discharge valves, are normally open. They need only remain open in the event of an accident. The I/P controller ( rather than a solenoid valve) controlling their position is fail-open. Since no function must be performed by these valves, they have been deleted from Table 3.

Item 5B, the solenoid valves for AOVs 897 and 898, are required to close prior to sump recirculation.

They will not experience an adverse environment prior to the time they must perform their safety function. Environmental qualification of these valves will be addressed in a later submittal, concerning electrical equipment located in a "mild" environment.

52

29. TER Paragraph 3.3.3.3 Table 3 Item No. 2. Copes-Vulcan Solenoid Valves.

The valves were purchased from ASCO (Series 8300).

Therefore, all information from Reference 23 applies to the valves. Further, since these valves are located in a "mild" environment, qualification of these valves will be discussed at a later time.

30. TER Paragraph 3.3.3.4 - Table 3 Item Nos. 3A, 3B.

Lawrence Solenoid Valves in Intermediate Building.

Based on the design principle of these valves, they will perform their safety function by failing in a closed position upon loss of power. However, if power qualification documentation is not established,

.RGaE will initiate a replacement for these solenoid valves. Qualification documentation will be made available when received. The fail-safe mode of operation ensures no loss of safety function in the interim.

31. TER Paragraph 3. 3. 3. 5 Table 3 Item No. 4. Versa Solenoid Valves inside containment.

The safety function of the solenoid valves controlling the containment air recirculation dampers is accomplished through fail-safe operation. This is accomplished immediately with the SI signal following an accident, before environmental conditions would 53

become very severe. In ordei to have this safety function accomplished with equipment having the proper qualif ication testing and documentation, replacement of these solenoid valves will be initiated. It is expected that this can be accomplished by June 1982. Qualification docu-mentation will be made available when received.

32. TER Paragraph 3. 3. 3. 6 Table 3 Item Nos. 6A, 6B.

Versa Solenoid Valves.

The safety function of these containment purge and depressurization valves immediately following an accident is to close for containment isolation.

This is accomplished by the fail-close design of these valves. In order to have this safety function I

accomplished with equipment having the proper qualification testing and documentation, replace-ment of these solenoid valves will be initiated.

It is expected that this can be accomplished by June 1982. Qualification documentation will be made available when received.

33. TER Paragraph 3.3.3.7 Table 3 Item No. 7. Control Room Dampers.

This equipment item is not electrical, and there-fore is not addressed in this report. The solenoid valves operating these dampers are addressed under paragraph TER 3.3.3.24 (Table 3, Item No. 40).

54

34. TER Paragraph 3. 3. 3. 8 - Table 3 Item No. 9. Standby'FN Pump Motors .

Although this item is not located in a harsh environment, and therefore does not need to be addressed at this time, RGSE considers the environ-mental qualification of this item to be complete and acceptable. As stated in Section 4.3.3 of the DOR Guidelines, "No special consideration need to the environmental qualification of Class be'iven IE equipment in these [non-harsh] areas provided the aging requirements discussed in Section 7.0 are satisfied and the areas are maintained at room conditions by redundant air conditioning or ventila-tion systems served by the onsite emergency electrical power system." This is the case with these motors.

The equipment specification for these motors (Reference 3) states "Motors shall be rated for operation in an ambient tern erature of 50'C [122'F] ".

(

Tnis is consistent with the ambient operating conditions for the Auxiliary Building Addition of 60-120'F (Ref erence 43) . Furthermore, the ongoing

.program described in Reference 47 to detect age-related f ailures includes these motors. RG&E theref ore considers these motors to have met all necessary environmental requirements .

P.

55

35. TER Paragraph 3.3.3.9 - Table 3 Item Nos. 10A, 10B, 10C, 12A. Motors for the Containment Spray Pumps, Component Cooling Water Pumps, Residual Heat Removal Pumps, and Safety Injection Pumps.

The first three of these Ginna motors have Class B insulation made of "Thermalastic Epoxy". The SI pump motor insulation is "PMR" (Premimum Moisture Resistant). This is shown in Reference 67.

Qualf ication of these systems is given in WCAP 8754, ( Ref erence 68 ), for the "Thermalas tie Epoxy" motors, and the Westinghouse Research Report 71-1C2-RADMC-R1, "The Ef f ect of Radiation on Insulating Materials Used in Westinghouse Medium Motors," December 31, 1970 (Revised April 10, 1971) (Reference 69) for the "PMR" motors. These reports are proprietary, but are available for audit at RGEE and at Westinghouse. Testing does indicate that these motors can withstand an accumulated dose of 10 7 rads during their operating lif e, with an operating lif e of 20 years. Since these motors are not used at all times (only the CCW pump is used during normal operation, and even then only one of the two pumps is normal ly in use), the operational capability is at least 40 years. Also, RG&E has a program of insulation inspection once per year (M45.1A, Inspection of Saf eg uard Motor) and replacement ( if needed) every five years.

56

r l

l

Since the only adverse environm'ent anticipated for any of these motors is a post-LOCA radiation dose

( conservatively estimated in Reference [TMI-3] as I

2. 8 x 10 6 rads) these motors are considered properly qualified both for "life" and radiation.

3 6. TER Paragraph 3.3.3.10 Table 3 Item No. 12B. Service Water Pump Motor.

As stated in Reference [Flood-15], the effects of jet impingement and water spray on these motors were evaluated by the NRC during the review of SEP Topic III-5.B, "Pipe Break Outside Containment".

RGEE committed to supplement the NRC recommenda-tion in Reference [FLOOD-13.]. Thus, the Service Water Pump Motors have been removed from the HELB environment considerations. Further review for operation is a "mild" environment will be conducted at a later time.

37. TER Paragraph 3.3.3.11 Table 3 Item No. 16. Coleman Cable Inside Containment.

Reference 51 is the "Cable Identification and Qualification Supplement". This reference allows traceability of all cable used in the Ginna plant, by referencing back to the original purchase order specifications. It can be seen that, in addition to the Kerite safeguards cable, the only other cable inside containment used to perform a required 57

post-accident safety f unction is the silicone-rubber insulated cable, which is used for all required safety-related instrumentation and control cable.

Reference 46 identifies this as Coleman cable. In addition to the testing stated in Reference 46, a section of this cable was taken from the Ginna plant, and environmentally qualif ied with the Raychem splice sleeves (documentation of the testing is given in FRC Final Report Supplement, F-C5074 (Supplement), April 1979, which is included in Reference 51). The cable is specimen number C5074-7 of Table 1 of F-C5074 Supplement.

This testing shows that the Coleman silicone-rubber insulated cable will perform its required safety functions inside containment.

Reference 46 states that this cable is aged at 200'C for 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />. Although no specific Arrhenius plot is available, the application of the "10'C rule" shows an operating life of 40 years at 60'C.

This is considered a reasonable estimate of the exoected life of this cable.

38. TER Paragraph 3.3.3.12 Table 3 Items 17A, 17B, 17C.

Coleman, Rome, and General Cables Used Outside Containment.

Reference 51 is the "Cable Identification and Qualification Supplement". From this reference, the type of cable used throughout the Ginna plant 58

can be traced by reference back to the original purchase order specification. It is shown that all of the safety-related cable outside containment which is not Kerite cable is PVC-insulated cable.

The specif ications included in Reference 51 refer to GAI Specs SP-5324 and SP-5315. Both of these specifications in turn specify the requirements of IPCEA S-61-402 for PVC-Cable. Information f rom this standard is provided in Reference 10. Additional information for Coleman and Rome cable is provided in Ref erence 4 6.

The IPCEA testing of this cable, including insula-tion aging at 121'C (250'F) for 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> ( jacket at 212'F), oil immersion, heat shock, and cold shock, shows the ability to operate under conditions more severe than those anticipated outside containment.

Although no specif ic qualif ication testing was performed, the standard testing of these cable types gives reasonable assurance that they are suitable for outside-containment use.

39. TER Paragraph 3.3.3.13 Table 3 Item No. 27. RTDs Inside Containment.

Reference 35 is a specification sheet and drawing of the Ginna RTD (Rosemount 176JA model).

The reactor coolant system temperature detectors (RTD) are not required for a loss of coolant 59

accident. In a steam line break accident, low Tave plus high steam flow plus" a safety injection signal will close the main steam line isolation valves. Also, high-high steam flow will perform this function. As described in Section II.B above, for a break upstream of the non-return check valves, which includes all breaks inside containment, closure of the main steam isolation valves is not required.

For breaks downstxeam of the check valves, closure of the main steam isolation valves is desirable, however, in this case the RTDs are not subjected to an adverse environment. Theref ore, the RTDs do not require environmental qualification to px'ovide their required safety function. However, the RTDs would be useful for post-accident monitoring.

Since the RTDs are not qualified for post-accident use, the pxesent Ginna Emergency Procedux'es specify that, if a 50'F subcooling margin cannot be established or maintained, safety injection flow shall not be terminated. Failure of the RTDs would require that SI flow be maintained. Since the Ginna high head safety injection pumps do not have a high enough shutoff head to open the pressurizer PORVs, continued SI pump operation is not a safety concern.

However, to avoid the possibility of operator confusion, RG&E will initiate a program to provide 60

qualified RTDs for post-accident monitoring.

These will be procured and installed by June 1982, I

sub ject to equipment availability and procurement/

delivery schedules.

40. TER Paragraph 3.3.3.14 Table 3 Item No. 28. Batteries in the Control Building.

As noted in TER C5257, the ventilation system is being modified, such that the battery rooms can be considered a "mild" environment. Reference fHELB-13]

committed to a resolution of the potential flooding problem. The batteries will thus be further discussed at a later time, together with other equipment located in a "mild" environment.

41. TER Paragraph 3.3.3.15 Table 3 Item No. 26. Steam Generator Level Transmitter.

The steam generator level transmitters, although useful for confirming secondary system heat removal capability, are not necessary for performing this function. For an accident inside containment, which could degrade the performance of the SG level transmitters, the main steam pressure transmitters, located outside containment, provide information regarding steam generator status. Auxiliary feedwater flow instrumentation for each steam generator, also located outside containment, provides the primary indication of the steam generator heat 61

removal capability. Based on the latest information provided at the Westinghouse Emergency Operating Instructions seminar, the Ginna Emergency Procedures will be revised to reflect AFW flow indications as being of prime value as the main indication of secondary heat removal capability.

Nevertheless, in order to remove the possibility of operator confusion due to misleading instrument indications, the steam generator Level trans-mitters will be replaced by June 1982. Qualifica-tion documentation will be made available when received.

42. TER Paragraph 3.3.3.16 - Table 3 Item Nos. 29A, 29B, 29C. Diesel Generator Electrical Equipment.

This equipment is located in a "mild" environment.

Its qualification will reviewed at a later date.

43. TER Paragraph 3.3.3.17 Table 3 Item No. 35. Valcor Solenoid Valves for the Pressurizer PORVs.

Additional information has been added to Reference 48, consisting of the test results and testing methodology. This was provided to the NRC and FRC on September 24, 1980. The entire test report is also available for audit and review at RGSE.

These valves are fully qualified to IEEE-323-1974 to perform their post-accident safety function.

62

I

44. TER Paragraph 3.3.3.18 Table 3 item No. 36. Sump B Wide Range Level Switch.

Ref erence 52, the specif ication sheet for this item, was provided to the NRC and FRC on September 24, 1980. There is evidence that these level switches can perform their function in a contain-ment post-accident environment. However, not all of the requirements of the DOR Guidelines are met for this instrumentation. Xt is important to note, however, that these instruments are not used to perf orm any post-accident saf ety-related f unctions, and are not specified for use in the Ginna Emergency Procedures except as confirmatory information.

The saf ety-related function of determining the timing of the "sump switchover" procedure is performed by the RWST level instrumentation, located outside containment.

The TMI Lessons Learned determined that a wide-range sump level indication was to be provided for operator information. Fully-qualified equipment will be purchased to meet this requirement. The qualification documentation for this instrumenta-tion will be made available when received.

45. TER Paragraph 3.3.3.19 - Table 3 Xtem Nos. 42, 43.

Motors for Cooling Fans for RHR, CS, Sl, and Charging Pumps in Auxiliary Building.

63

Reference 69 provides information concerning the life and radiation characteristics of these motors.

These motors are capable of operation after a radiation exposure of 1 x 10 7 rads and 20 years.

Since these motors are run only intermittently, operational capability for 40 years is shown.

Since the only harsh environment experienced by these motors is post-LOCA radiation (estimated at 2.8 x 10 6 rads), operation under required accident conditions is shown.

46. TER Paragraph 3.3.3.20 Table 3 Item Nos. 32, 44. IGC Cabinets and Relay Racks in Relay Room.

This equipment is located in a mild environment.

Its qualification will be considered at a later time.

47. TER Paragraph 3.3.3.21 Table 3 Item No. 33A. Control Room HVAC Air Handling Units.

This equipment is located in a mild environment.

Its qualification will be considered at a later time.

48. TER Paragraph 3.3.3.22 Table 3 Item No. 33B. Control Room HVAC Fans.

This item is not an electrical piece of equipment.

It has thus been deleted from Table 3, and from consideration in this report.

64

49. TER Paragraph 3.3.3.23 Table 3, Item No. 39, Charging Pumo Mo tors .

This equipment is located in a mild environment.

Its qualification will be considered at a later time.

50. TER Paragraph 3.3.3.24 Table 3 Item No. 40. Control Room HVAC Damper Solenoids.

This equipment is located in a mild environment.

Its qualification will .be considered at a later time.

65

LOSS OF COOLANT ACCIDENT 1 . 2/3 2/3 HIGH LOW CONTA I NMENT PRESSURIZER PRESSURE PRESSURE FIGURE 1 HI HI j

la SAFETY ACCIDENT INJECTION DIAGNOSTICS

4. 3. 2. 4. 5. 6.

HAIN ACCUtlULATOR SAFETY FEEDl<ATER CONTA I Nf 1ENT REACTOR STEAM LINE DUtlP INJECTION LINE ISOLATION TRIP ISOLATION SEQUENCE ISOLATION (AUTO) 7.

REACTOR VALVES COOLANT PUf'lp TRIP 9:

CONTROL ROOM VENTILATION 10.

MANUAL ACTIONS RECIRC-ULATION

TABLE 1 LOSS OF COOLANT ACCIDENT REQUIRED BLOCK NO./EQUIPMENT SAFETY FUNCTION OPERATION TIME

1. High Containment Pressure Low Pressurizer Pressure PT 945, 946, 947 Provide signals for Contain- Signal Initiation PT 948) 949, 950 ment Spray, Safety Injection, Containment Isolation, and Main Steam and Feedwater Line Isolation Accident Diagnostics Short term PT 429) 430, 433.) 449 Provide Reactor trip and Signal Initiation Safety Injection signals Accident Diagnostics Short term Splice Sleeves, Terminal Control and Power Signal Long term Blocks, Electrical Pene- Transmission trations, Electrical Cable la. Steam Line Pressure Accident Diagnostics Short term PT 468 ) 469 ) 482 PT 478, 479) 483 Radiation Accident Diagnostics Short term 'ontainment

[Being provided per TMI STLL]

Containment'sump level Accident Diagnostics Short term IT 942, LT 943

2. Safety Injection Sequence (Auto)

Batteries D. C. Power Long Term lA, 1B Diesel Generator Power supply to safeguards Long term and Auxiliaries busses during loss of out-side AC Power 480 Volt Safeguards Provide. the distribution of Long term busses 14, 16, 17, 18 power to safeguards equipment lA, 1B, 1C Safety Injec- High head injection of bo- Long term tion Pumps rated water to Reactor Coolant System lA, 1B Containment Spray Containment Pressure, Tem- Long term Pumps (only on hi-hi Cont. perature, and Iodine control pressure)

TABLE 1 LOSS OF COOPT ACCIDENT

,f RE(}UIRED.

BL CK 'NO./EQUIPMENT SAFETY FUNCTION OPERATION TIME 1.<, 1B Residual Heat Re- Low head injection of borated Long term

.moval Pumps water to Reactor Vessel

/

1A; 1B, 1C, 1D Service Cooling water to RHR and CCN Long term Mater Pumps Heat Exchangers 1A, 1B, 1C," lD Contain- Containment Pressure, Tem- Long term ment Recirc. Units perature, and Iodine control Cooling Units for pump Haintain motors within proper Long Term motors (SI, RHR, CS, ambient temperature limits and Charging) 1A, 1B Hotor Driven Cooling water to Steam Gen- Long term Aux. Feedwater Pumps erators 480 Volt Safeguards Provide the distribution of Long term MCC's power to safeguards equipment 3 ~ Accumulator Dump HOV 841 (N.O.)-'OV Provide path to Reactor Vessel Not required 865 (N.O.) from Accumulators for injection to function of borated water

4. Main Steam Line Isolation Feedwater Line Isolation AOV .3516 Isolate 1A, 1B Steam Generators 5 Seconds after AOV 3517 signal AOV 4269 4270 Isolate Hain Feedwater System 5 Seconds a fter AOV signal AOV 4271 AOV 4272
5. Containment Isolation See Text,Section II.A.5
6. Reactor Trip 0

Reactor trip breakers Provide means to trip the reactor Required for Reactor Trip Reactor protection and in- Provide the instrumentation and Required for strumentation cabinets protection circuits for the con- Reactor Trip trol and tripping of the Reactor

7. RCP Trip RCP Trip Breakers Provide means to trip RCP's Short term N.O. = Normally Open

I LOSS OF COOLANT ACCIDENT REQUIRED CK NO./EQUIPHENT SAFETY FUNCTION OPERATION TIHE alves HOV 825 A)B Provide path to SI Pumps for bor- 10/ BAST Level HOV,826 A ) B ) C D ated water to high head safety or-1/2 hour (Baa N.O.) injection AOV 836 A)B Provide controlled addition of Short term NaOH to Containment Spray for Iodine control HOV 852 A)B Provide path to Reactor Vessel SI'initiation of borated water for low head safety injection HOV 860 A)B,C)D Provide path to Containment Spray I,ong term headers for CS Pumps BAST Level Indicate BAST Level for automatic 10% BAST Ievel IT 102) 106, 171) 172 transfer of SI Pump suction from or-1/2 hour BAST to RMST HOV 878 B)D Provide path to cold legs of RCS not required (N.O.) from high head safety injection to function HOV 4007, 4008 Provide path for Aux. Feedwater to Short term 1A, 1B Steam Generators AOV 5871, 5872, 5873 Provide path for cleaning of cont. signal initiation AOV 5874, 5875) 5876 atmosphere by fan coolers

9. Control Room Ventilation Provide cleaning of Control Room Short term Dampers and AiiU atmosphere 10: Hanual Safety Injection Reset Reset Safety Injection signal less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Button after, automatic S.I. Sequencing is complete 1A, 1B Component Cooling Cooling water for safeguards Long term Mater Pumps equipment 1A, 1B Containment Spray Containment Pressure, Temperature Long term Pumps (if Cont. Pressure and Iodine control (30 psig)

RWST Level Indicate RMST Level for operator less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> LT 920, LIC 921 transfer from S.I. phase to Recirculation phase

'N I

TABLE 1 LOSS OF COOLS'CCIDENT f

REQUIRED BLOCK NO./EQUIPHENT SAFETY FUNCTION OPERATION TIHE HOV 4027, 4028 Provide Service Mater to Hotor within-2 hours Driven Aux. Feedwater Pumps suction HOV 4000A, 4000B Provide AFW Cross-Connect Short term HOV 4734) 4735) 4615, 4616 Direct SW Flow to CCW HX's less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> HOV 738 A)B Direct CCW Flow to RHR HX's less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Standby AFW Pumps AFW Flow to SG's System inoperable if normal AFM Long term HOV 9629 A,B Provide SW to suction of standby Long term AFM Pumps HOV 9710 A,B; 9703 A,B; Standby AFM Discharge Valves to Long term 9704 A)B provide flow to SG's Steam Generator Level Honitoring Long term LT 460, 461, 462, 463 LT 470) 471, 472) 473 Sampling (being provided Sample containment atmosphere I,ong term per THI) and reactor coolant e

Hydrogen Recombiners Haintain hydrogen control Long term Pressurizer PORVs RC Pressure Control Long term

. 11. Recirculation HOV 850 A,B outside cont. Provide path to RHR suction from Long term HOV 851 A,B (N.O.) inside B sump for low head safety injec-cont. tion HOV 856 (N.O.) RWST isolation valve to RHR pumps required to func-suction, must close after RMST is tion to switch to drained recirc phase HOV 896 A,B (N.O.) RMST isolation valve, must close required to func-after RWST is drained tion to switch to recirc phase HOV 857 A,B,C Provide path to suction of SI and required to func-CS Pumps from RER pumps discharge tion to switch to recirc phase AOV 897) 898 Isolate high head recirculation Short term flow to RWST during sump recir-culation HOV 704 A)B Close during switch to sump less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> recirculation

MAIN STEAM OR FEED LINE B FIGURE 2

3. 2/3 1. 2/3 2. 2/3 HIGH LOM CONTAINMENT STEAN LINE PRESSURIZER PRESSURE PRESSURE PRESSURE HI HI
3. 2/3 3. 2/4 3. 2/4 I

STEAN LINE LOW SAFETY ACCIDENT OVERPOWER FLOIA T ave INJECTION OIAGIIOSTICS hT I

HI 1

(

I.

4. 6. 5.

MAIN SAFETY FEEDllATER STEAN LINE INJECT ION LINE CONTAINMENT REACTOR ISOLATION SEQUENCE ISOLATION ISOLATION TRIP (AUTO)

9. 8.

REACTOR VALVES COOLANT PUMP TRIP 10.

MANUAL ACTIONS 11.

CONTINUED SAFE SHUTDOWN

TABLE 2 MAIN STEAM LINE BREAK SAFETY FUNCTION/BREAK LOCATION REQUIRED BLOCK NO./EQUIPMENT SAFETY FUNCTION OPERATION TIME INSIDE CV OUTSIDE CV

1. Steam Line Pressure Provide signal for same signal initiation PT 468, 469, 482 SI on low steam line PT 478) 479) 483 pressure la. Steam Line Pressure Accident Diagnostics same short term (see 1 above)

Containment Radiation Accident Diagnostics NA short term Containment Sump Level Accident Diagnostics NA short term High Containment Pressure Accident Diagnostics NA short term (see 3 below)

2. Low Pressurizer Pressure PT 429, 430, 431) 449 Provide Reactor trip same signal initiation and Safety Injection signals Electrical Penetrations, Provide control and same long term Cable, Sleeves, and Power Signal Terminal Blocks Transmission High Containment Pressure PT 945) 946, 947 Provide signals for NA signal initiation PT 948) 949~ 950 Containment Spray, Safety Injection, Containment Isola-tion, and Main- Steam Line Isolation Steam Line Flow FT 464, 465 Provide signals for same signal initiation FT 474, 475 Reactor trip and Main Steam Line Iso-lation Reactor Coolant Temperature Loop A Hot Ieg Provide Iow Tave 6 same signal initiation TE 401A, 402A) 6 signals for'Reactor 405A, 406A, trip, Safety Injec-409A tion and Main Steam Line Isolation

TABLE 2 MAIN STEAM LINE BREAK - 2-SAFETY FUNCTION/BREAK LOCATION REQUIRED BLOCK NO./EQUIPMENT SAFETY FUNCTION OPERATION TIME INSIDE CV OUTSIDE CV Loop A Cold Leg TE 401B> 404A, 407A>

408A, 410A Loop B Hot Leg TE 403B> 404B, 407B, 408B, 410B Loop B Cold I,eg TE 403B> 404B> 407B, 408B> 410B Main Steam Isolation AOV 3516 Isolate 1A, B Steam same 5 seconds after signal AOV 3517 Generators Feedwater Line Isolation AOV 4269 Isolate Main Feed- same 5 seconds after signal AOV 4270 water system AOV 4271 AOV 4272 Containment Isolation See Text, Section same II.B.5 Safety Injection Sequence (Auto)

Batteries D.C. Power same Long term 1A, 1B Diesel Power supply to safe- same Long term Generators and guards busses during auxiliaries loss of.,outside AC Power 480 Volt Safeguards Provide distribution same Long term busses 14, 16, 17, 18 of power to safe-guards equipment 1A, 1B, 1C Safety In- High head. injection same Long term jection pumps of borated water to Reactor Coolant System lA, B Containment Spray Containment Pressure N/A I,ong term Pumps (only on hi-hi cont. and Temperature Pressure) control 1A, 1B, 1C, 1D Service Cooling Water to same Long term Water Pumps CCW Heat Exchanger

HAIN STEAM LINE BREAK SAFETY FUNCTION/BREAK IOCATION REQUIRED BLOCK NO./EQUIPMENT SAFETY FUNCTION OPERATION TIME INSIDE CV OUTSIDE CV 1A, 1B, 1C, 1D Containment Containment Pressure N/A Long term Recirc Units and Temperature con-trol 1A, 1B Motor Driven Aux. Cooling w'ater supply same Long term Feedwater Pumps to Steam Generators Cooling Units for SI, CS, Maintain motors same Long term RHR, and Charging Pump within proper ambient temperature limits 480 Volt Safeguards Provide the distribu- same Long term HCCs tion of power to safeguards equipment

7. Reactor Trip Reactor trip breakers Provide means to same Required for trip the reactor 'eactor Trip Reactor Protection and Provide the instru- same Required for Instrumentation mentation and pro- Reactor Trip Cabinets tion circuits for the control and tripping of the reactor
8. Reactor Coolant Pump Trip Provide means to trip NA Short term RCP Trip Breakers RCPs
9. Valves HOV 825A> B Provide path to SI same 10/ BAST Level HOV 826A, B) C, D Pumps for borated. o~l/2 hour (Baa N.O.) water to high head safety injection AOV 836A, B . Provide needed NaOH to CS if Short term HOV 860A, B, C) D Provide path to Con- N/A Long term tainment, Spray headers for CS Pumps'rovide HOV 878, B, D path to same not required to (N.O.) cold legs of,RCS function from high head safety injection

TABLE 2 MAIN STEAM LINE BREAK SAFETY FUNCTION/BREAK LOCATION REQUIRED BIOCK NO./EQUIPMENT SA'FETY FUNCTION OPERATION TIME INSIDE CV OUTSIDE CV HOV 896)A)B)(NO) Provide path from same short-term (to close RWST of borated if need sump water for SI and recirculaton)

CS pumps suction MOV 4007) 4008 Provide path for Aux. same Short term Feedwater to Steam Generators AOV 5871) 5872) 5873 Provide path for N/A signal initiation AOV 5874, 5875) 5876 cleaning by fan coolers, cooling of cont. Atmosphere BAST Level 1 Indicate BAST Level same 10/ BAST I,evel LT 102) 106) 171)'72 for automatic trans- or~1/2 hour fer of SI Pump suction from BAST to RWST MOV 852A, B Provide path for low same Signal Initiation head SI to Reactor Vessel

10. Manual'G Level Instrumentation Determine affected SG same Short term LT 470, 471, 472, 473 LT 460, 461, 462) 463 Safety Injection Reset Reset SI signal after same less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Button Automatic SI sequenc-ing is complete 1A, 1B Component Cooling Cooling Water for same Long term Water Pumps safeguards equipment 1A, 1B Containment Containment Pressure N/A Long term Spray Pump (If cont. and Temperature con-Pressure < 30 psig) trol MOV 402?, 4028 Provide Service Water same within ~2 hours to Motor Driven Aux.

Feedwater Pumps Suction Charging pumps Inventory control to same Long term RCS

TABLE 2 HAIN STEAH LINE BREAK SAFETY FUNCTION/BREAK LOCATION REQUIRED BLOCK NO./EQUIPHENT SAFETY FUNCTION OPERATION TIHE INSIDE CV OUTSIDE CV Standby AFW pumps Provide flow to same I,ong term SGs if AFW system in-operable HOV 9629A, B Provide SW to suction same Long term of Standby AFW Pumps MOV 9710A, B; 9703A, B; Standby AFW discharge same Long term 9704A, B valves to provide AFW flow to SGs HOV 4000A, B AFW Cross-Connect same Short term Valves

11. Continued Safe Shutdown Sampling (per THI) Sample Containment same Long term Atmosphere and Reactor Coolant Pressurizer PORVs RC Pressure Control same Long term

Accident References LOCA analysis [LOCA]

FSAR

2. "ECCS Analysis for the R. E. Ginna Reactor with ENC WREM-2 PWR Evaluation Model" dated December 1977 sub-mitted with Application for Amendement to Operating License, on January 6, 1978.
3. ECCS Analysis submitted by letter dated April 7, 1977 from L. D. White, Jr., RG&E to A. Schwencer, Chief, Operating Reactors Branch Il, USNRC.

4, ECCS Analysis for the R. E. Ginna Reactor with ENC WREM-2 PWR Evaluation Model. Exxon Nuclear Co.

Report XN-NF-77-58.

5. Ginna Emergency Procedures E1.1 and E1.2, submitted by letter dated February 26, 1980 from L. D. White, Jr.

RG&E, to D. L. Ziemann, USNRC.

Steam Line Break and Feedwater Line Break [SLB/FLB]

2. Steam line break analyses submitted with Application for Amendment to Operating License on September 22, 1975.
3. Plant 'Transient. Analysis for the R. E. Ginna Unit 1 Nuclear Power Plant, Exxon Report XN-NF-77-40 (11/77 and updated 12/15/78 and March, 1980.

Letter dated May 24, 1977 from K. W. Amish, RG&E to J. F.- O'eary, NRC.

5. Ginna Emergency Procedures E1.1 and E1.3, submitted by letter dated February 26, 1980 from L. D. White, Jr.,

RG&E to D. L. Ziemann, USNRC.

6'. Letter from L. D. White, Jr., RG&E, to D. L. Ziemann, NRC, March 28, 1980.

High Energy Line Break [HELB]

"Effects of Postulated'Pipe Breaks Outside the Con-tainment Building", GAI Report No. 1815, submitted by letter dated November 1, 1973 from K. W. Amish, RG&E, to A, Giambuso, Deputy Director for Reactor Projects, USNRC.

Letter dated May 24, 1974 from K. W. Amish, RG&E, to J. F. O'eary, Director, Directorate of Licensing, USNRC.

Letter dated September 4, 1974 for R. R. Koprowski, RG&E to Edson Case, Acting Director, Directorate of Licensing, USNRC.

Letter dated November 1, 1974 from K. W. Amish, RG&E, to Edson Case, Acting Director, Directorate of Li-censing, USNRC.

Letter dated May 20, 1977 from L. D. White, Jr., RG&E, to A. Schwencer, Chief Operating Reactors Branch 51, USNRC.

Letter dated February 6, '1978 from L. D. White, Jr.,

RG&E, to A. Schwencer, Chief, Operating Reactors Branch Ol, USNRC.

Amendment No. 7 to Provisional Operating License DPR-18, transmitted, by letter dated May 14, 1975 from Robert A.

Purple, Chief, Operating Reactors Branch-51, USNRC, to L. D. White, Jr , RG&E.

Amendment No. 29 to Provisional Operating License DPR-18, transmitted by letter dated August 24, 1979 from Dennis L.

Ziemann, Chief, ORB 52, to L. D. White, Jr., RG&E.

Letter, L. D. White, Jr., RG&E, to D. L. Ziemann, May 17, 1979.

Letter, L. D. White, Jr., RG&E, to D. L. Ziemann, USNRC, June 27, 1979.

Letter, L. D. White, Jr., RG&E, to D. L. Ziemann, USNRC July 6, 1979.

Letter, R. E. Anderson, Gilbert/Commonwealth to James J.

Shea, USNRC, June 11, 1979.

Letter, L. D. White, Jr., RG&E, to D. M. Crutchfield, NRC, SEP Topic III-5.B, "Pipe Break Outside Containment,"

August 7, 1980.

Letter, J. Wenclawiak and T. Snyder, Catalytic, to G. Wrobel, RG&E, "Equipment Environmental Qualification,"

October 27, 1980.

Letter from D. M. Crutchfield, NRC, to L. D. White, Jr.

RG&E, SEP Topic III-S.B, "Pipe Break Outside Containment,"

June 24, 1980.

Effects of Flooding [Flood]

Letter dated May 13, 1975 from L. D. White, Jr., RG&E, to Benard C. Rusche, Director, Office of Nuclear- Reactor Regulation, USNRC.

2. Letter dated May 20, 1975 from L.- D'. White, Jr., RG&E, to Robert A. Purple, Chief, Operating Reactors Branch 51, Division of Reactor Licensing.

3., Letter dated May 30, 1975 from L. D. White, Jr., RG&E, to Robert A. Purple.

t Letter dated June 16, 1975 from L. D. White', Jr., RG&E, to Robert A. Purple.

5. Letter dated July 3, 1975 from Robert A. Purple to L. D. White, Jr., RG&E.
6. Letter dated August. 8, 1972 from Donald J. Skovholt, Assistant Director for Operating Reactors, USAEC, to Edward J. Nelson, RG&E.
7. Letter dated October 3, 1972 from K. W. Amish, RG&E, to Donald J. Skovholt, Assistant Director for Operating Reactors, USAEC.
8. Letter dated May 31, 1973 from K. W. Amish, RG&E, to Donald J. Skovholt, Assistant, Director for Operating Reactors, USAEC.
9. Application for Amendment to Operating License, sub-mitted March 10, 1975.
10. Amendment, No. 14 to Provisional Operating License DPR-18, transmitted by letter dated June 1, 1977 from A. Schwencer, Chief, Operating Reactors Branch 51, USNRC.

Letter, L. D. White, Jr. RG&E, to Dennis L. Ziemann, USNRC, High Energy Line Breaks Outside Containment, June 27, 1979.

TMI Lessons Learned [TMI]

RG&E letter of October 17, 1979, L. D. White, Jr.,

RG&E, to D. L. Ziemann, USNRC, "TMI Short Term Lessons Learned Requirements."

2. RG&E letter of November 19, 1979, L. D. White, Jr. to D. L. Ziemann, USNRC, "TMI Short Term Lessons Learned."
3. RG&E letter of December 28, 1979, L. D. White, Jr. to D.,L. Ziemann, USNRC, "TMI Short Term Lessons Learned."

I I

'\

,(

l

Table 3 Page 1 Reactor: GINNA SYSTElTIC EVALUATION PROGRAM Tame ENVIRONMENT Qua . Document Equipment Type Location Needed Parameter Require Qua Method Reference Comments Solenoid Valve Area 57 SI Signal Temp ('F) See Amb. Experience 23 DBE Main SLB ASCO/ ,Pr (psia) Comments Atm. Experience in Turbine Bldg.

V-4269, V-4270 RH (%) Amb. Experience Fail-Safe (closed)

LB 8300 B 61 U Chem (FW Control Valves) Rad.

V-4271, V-4272 Sub.

LB 8300 B 64 RU (FW Bypass Valves)

'emp

2. Solenoid Valve Area 52 Minutes ('F) See Amb. Experience 23 These valves were

'Copes-Vulcan Pr (psia) Comments Atm. Experience purchased from ASCO.

AOV 836 A,B RH (%) Amb. Experience 8200 series. They

.(NaOH to CS) Chem. are fail safe Rad: (open).

Sub.

3. Solenoid Valve Area I3 Seconds Temp ('F) See 250 Vendor Data 25-, En'closed in NEMA-2 Lawrence/ Pr (psia) Comments Atm. .'xperience drip-proof enclosure 110114W - Supply RH (%) Amb. Experience which is subjected 125434W - Vent Chem. to salt water spray V-3516, V-3517 Rad. qualification test.

(Main Steam Isola- Sub. Fail safe (closed) tion)

4. Solenoid Valve Area 51 Seconds Temp ('F) See 200 Vendor Data 26 Fail safe. Per-Versa/VSG Pr (psia) Comments Atm. Experience forms safety V-5871, V-5872, RH (%) Amb. Experience function within

~ V-5873, V-5874, Chem. Yes seconds of start of

'V-5875, V-5876 Rad. No DBE. Not required (Containment.Recir- Sub. to operate when culation System accident conditions Dampers) are reached.

l r

~(

]

]

Table 3 Page 2 Reactor: GINNA SYSTEMATIC EVALUATION PROGRAM Tame ENV I RONMENT Qua . Document Equipment Type Location Needed Parameter Require Qua . Method Reference Comments

5. Solenoid Valve Area 42 Short-Term Temp ('F) See Amb. Experience , 23 "Mild" Envt. to ASCO (Before Sump Pr (psia) Comments Atm. Experience be addressed later AOV-897, AOV-898 Recirculation) RH (%) . Amb. Experience (SI Recirculation) Chem.

Rad.

Sub.

6. Solenoid Valve Temp ('F) See 200 Vendor data 26 Fail-close Versa/ Area 51 Seconds Pr (psia) Comments Atm. Experience to perform con-VSG-3731 Area 53 RH (%) Amb. Experience tainment isola-(Cont. Purge Valves) Chem. tion function VSG-3421 Rad.

(Cont. Depressuriza- Sub.

tlon)

7. Control Room Dampers Not Electrical.

D-81 + D-87 Deleted from Report 8a. Limitorque Area 41 Not required Temp (oF See 320 Test 18,19 Valves are locked-SMB-2 to operate Pr (psia) Comments 105 Test 18,19 open with power Reliance Motor RH (%) 100 Test 18, 19 removed. No need MOV 841, 865 Chem. Yes Test 18, 19 to function .t (Accumulator Rad. 2 x 10 Test 18, 19 Discharge) Sub. No 37 j

8b. Limitorque Area 52 Short-Term SMB-OO, Peerless (Before Sump Temp ('F) Amb. Amb. Experience 13 Not exposed to MOV 826 A,B,C,D recirculation) Pr (psia) Atm. Atm. Experience DBE environment (BAST to SI Pumps) RH (%) Amb. Amb. Experience MOV 896 A,B Chem. No (RWST to SI Pumps) Rad. No Sub. No

Table 3 Page 3 Reactor: GINNA SYSTEMATIC EVALUATION PROGRAM Tame ENVIRONMENT Qua . Document Equipment Type Location Needed Parameter Require Qua . Method Reference Comments 8c. Iimitorque SMB-00 Area 52 Short-Term (Before Sump Temp ('F)

Pr (psia)

Amb.

Atm.

Amb.

Atm.

Experience Experience

'3 No exposed DBE to environment

'Reliance Motor Recirculation) RH (%) Amb. Amb. Exp'erience

  • MOV 825 A,B Chem. No

{RWST to SI Pumps) Rad. No Sub. No Sd. Limitorque Area 43 Short-Term. Temp (4F) See Amb. Experience Not required to SMB-00 Only for DBEs Pr (psia) Comment Atm. Experience operate in harsh Reliance Motor not in area N. RH (%) Amb. Experience DBE envt. Alter-MOV 4007, 4008 See Comment. Chem. native SAFW (AFW Discharge) Rad. system available.

MOV 4027, 4028 Sub.

(AFW Suction) 4000 A,B (AFW Cross-Connect) 8e. Limitorque Area 02 Long Temp ('F) Amb. 320 Test 18,19,53 Not exposed to SMB-00 Pr {psia) Atm. 105 Test 18ilgi53 DBE environment Reliance RH (%) Amb. 100 Test 18,19,53 except post-LOCA V-850 A,B (Sump Chem. No Yes Test 18,19,53 sump water recir-Valves) culation MOV 856 (RWST to Rad. 3 x 10 2 x 10 Test 18,19,53 RHR)

V-857 A,B,C (RHR Sub. No to SI)

V-860 A,B,C,D (CS Valves) 8f. Limitorque Area 51 Not required emp (oF) See Amb. Experience 13 Not required to SMB-00 to operate Pr (psia) Comment Atm. Experience function for DBE.

MOV-851 A,B RH (%) Amb. Experience Valves are in Chem. No locked-open posi-Rad. No tion as required Sub. No for SI.

Table 3 Page 4 Reactor: GINNA SYSTEMATIC EVALUATION PROGRAM Tame ENVIRONMENT Qua . Document Equipment Type ,Location - Needed Parameter Require Qua . Method Reference Comments

g. Limitorque Area 51 Not required Temp ('F) Amb. Amb. Experience 13 Not required to "SMB-00 to operate Pr (psia) Atm. Atm. - Experience function for DBE.

Peerless Motor RH (%) Amb. Amb. Experience Valves are locked MOV 878 B,D Chem. in open position, (SI to cold legs) Rad. as needed for SI.

Sub.

8h. Limitorque Area 01 SI Signal Temp ('F) 286 320 Test 18,19 Valve completes SMB-1 Pr (psia) 75 105 Test 18,19 safety function Reliance Motor RH (%) 100 100 Test 18,19 (to open) early MOV 852 A,B Chem. Yes Yes Test 18,19 into accident (core deluge) Rad. 1.6 x 10 2 x 10 Test 18,19 Sub. No No 37 8i. Limitorque Area 46 Long Term Temp (4F) 120 120 Vendor Data 43,47,54 Standby AFW System SMB-00 Pr (psia) Atm. Atm. Experience located in con-Reliance Motor RH (%) Amb. Amb. Experience trolled envt.

MOV 9703 A,B; Chem. No 9704 A, B; 9710 A, B Rad. No (Standby AFW System) Sub. No

9. Motor, Pump Area 86 Long Term Temp ('F) 120 122 Vendor Data 2,3,43,47 Standby AFW pumps General Electric Pr (psia) Atm. Atm. Experience located in aux.

(Standby AFW) RH (%) Amb. Amb. Experience bldg. annex which Chem. No has controlled Rad. No envt.

Sub. No 1Q. Motor, Pump Area 52 Long Temp ('F) Amb. 104 F Spec 15,16,67 Only DBE environ-Westinghouse Pr (psia) Atm. Atm. Experience ment is post-444 TS TBDP RH (%) Amb. Amb. Experience accident radiation 445 TS TBDP Chem. No (Containment Spray, Rad. 3 x 10 1 x 10 Test 69 RHR, Component Sub. No Cooling)

Table 3 Page 5 Reactor: GINNA SYSTEMATIC EVALUATION PROGRAM Tame ENVIRONMENT Qua Document Equipment Type Location Needed Parameter Require Qua . Method Reference Comments ll. Motor, Pump Westinghouse Area ()3 Long Temp ('F) See 1040F Spec 8,16,67 Have installed 505 US ABDP Pr (psia) Comment Atm. Experience totally redundant (Auxiliary Feed- RH (%) Amb. Experience system not exposed water)

Chem. to DBE (standby Rad. 2 x 10 Test 68 AFW)

Sub.

12a. Motor, Pump Area C3 Long ('F) 104oF Westinghouse Temp Amb. Spec 15,16,67 Only DBE environ-Pr (psia) Atm. Atm. Experience ment is post-509 US AFDP (Safety Injection)

RH (%) Amb. Amb. Experience accident radiation Chem. No Rad.

Sub.

3xlo No 2 x 10 Test 68 12b. Motor, Pump Area Long ('F) 509 UPH ABDP N5 Temp Amb. See Experience 67 This item is in a Pr (psia) Atm. Comment Experience "mild" environ-(Service Water) RH (%) Amb. Experience ment. It will be Chem.

Rad.

No No addressed later.

Sub. No 13a. Penetrations, Area 41 Long ('F)

Electrical Temp 286 F 340oF Test 1,45,54,58 Radiation level at Pr (psia) 75 105 Test 1,4S,S4,S8 location of pene-Crouse-Hinds RH (%) 100% 100% Test 1,45,54,58 trltions < 1.6 x Chem. Yes Yes Test 58 10 rads. Qualifi-Rad. 1.6xl0 1.17x10 Test 45,64 fication test is Sub. No greater than DOR guidelines value of 2 x 10 rads.

13b. Penetrations, Area Nl Long Temp ('F) 286oF 340oF Test Electrical 29,30,59 Pr (psia) 75 75 Test 29,30,59 Westinghouse RH (%) 100% 100% Test 29,30,59 Chem. es s Test 29,30,59 Rad. 1.6x10 8 2.1x10 8 Sub. No

Table 3 Page 6 Reactor: GINNA SYSTEMATIC EVALUATION PROGRAM Tame ENVIRONMENT Qua . Document Equipment Type Location Needed Parameter Require Qua . Method Reference Comments

14. Terminal Block Area 51 Long Temp ('F) 286oF 3400F Test 50 Location of Westinghouse Pr (psia) 75 121 Test 50 blocks7is such that 542247 RH (%) 100%o 100% Test 50 2 x 10 rads, a chem. es Yes 7 Test 50 value equal to the Rad. 1.6x10 8 2x10 Test 60 DOR guidelines Sub. No value, should be acceptable. Also, terminal blocks will be elevated.

15a. Cable Area Il Long Temp (oF) 286 F 340oF Test 11,38,51, 55,63 Kerite Pr (psia) 75 118 Test HT RH (%) 100% 100% Test Chem. es 8 Yes 8 Test Rad. 1.6xlO 2xlO Test Sub. No 15b. Cable All Long Temp (oF) 220oF 340oF Test 11,38,51, 55,63 Kerite Pr (psia) 15.8 118 Test HT RH (%) 100 100 Test Chem. No Yes 8 Test Rad. No 2x10 Test Sub. No

16. Cable Area Nl Long Temp ('F) 286 340 Test 46, 51 Coleman Cable Pr (psia) 75 118 Test 46,51 RH (%) 100 100 Test 46,51 Chem. Yes es Test 46,51 Rad. 1.6xlo 2xlO 8 Test 46,51 Sub. No

Table 3 Page 7 Reactor: GINNA SYSTEMATIC EVALUATION PROGRAM Tame ENVIRONMENT Qua . Document Equipment Type Location Needed Parameter Require Qua . Method Reference Comments

17. Cable All Long Temp ('F) 220 250 Test 5,10,46 In lieu of 100/ RH, Coleman Cable Pr (psia) 15.8 Atm. Experience an owl zmmersxon Rome Cable RH (%) 100 Amb. Experience test performed per General Cable Chem. No IPCEA S-61-402

/ Rad. No Sub. No

18. Transmitter, Level Area N2 Short Term Temp ('F) Amb. Amb. Experience Not exposed to DBE Foxboro (Before Sump Pr (psia) Atm. Atm. Experience. when required to (RWST Level) Recirculation) RH (%) Amb. Amb. Experience to function Chem. No Rad. No Sub. No
19. Transmitter, Level Area 42 Short Term Temp ( oF) Amb. 200 Vendor Data 34 Not exposed to Barton 289 (Before Sump Pr (psia) Atm. Atm. Experience DBE envt. when (RWST Level) Recirculation) RH (%) Amb. Amb. Experience required to Chem. No function.

Rad. No Sub. No

20. Transmitter, Flow Area 51 Seconds Temp ('F) 286 See See 31 Not exposed to Barton 332 Pr (psia) 75 Comments Comments to DBE when (Steam Flow) RH (%) 100 required to Chem. Yes function.

Rad. 1.6x10 Sub. No

21. Transmitter, Pres. Areas 2,3 Long Temp (oF) Amb. See See 31 Not exposed to

, Barton 332 Pr (psia) Atm. Comments Comments DBE when required (Cont. Pressure) RH (%) Amb. to function.

Chem. No Rad. No Sub. No

Table 3 pPage 8 Reactor: GINNA SYSTEMATIC EVALUATION PROGRAM Tame ENVIRONMENT Qua . Document Equipment Type Location Needed Parameter Require Qua . Method Reference Comments

22. Transmitter, Area 41 Short Temp ('F) 286 286 Test 18,19,33 Adequate for short-Pressure Pr (psia) 75 75 Test 18,19,33 term function. Will Foxboro RH (%) 100 100 Test 18,19,33 be replaced and 611 GM-DSI Chem. Yes Yes < Test 18,19,33 elevated to perform

~

(PRZR Pressure) Rad. l. 7xl0 3x10 Evaluation 18,19 post-accident Sub. No monitoring function

23. Transmitter, Area 43 Short Temp ('F) See See See 18,19 Not exposed to Pressure Pr (psia) Comments Comments Comments 18,19 DBE when required Foxboro RH(%) 18,19 to function 611 GM-DSI Chem. 18,19 (Steam Pressure) Rad. 18,19 Sub.
24. Transmitter, Area 51 Temp ('F) See See See Not required for Level Pr (psia) Comments Comments Comments a short-term Foxboro RH (%) safety function.

613 M-MDL Modified Chem. Will be replaced (Przr Level) Sub. for long-term monitoring

25. Transmitter, Level Area 52 Sort Temp (4F) Amb. Amb. Experience Not exposed to Foxboro Pr (psia) Atm. Atm. Experience DBE 613 DM-MSI RH (%) Amb. Amb. Experience (BAST Level) Chem. No Rad. No Sub. No
26. Transmitter, Level Area 51 Temp ('F) See See See Alternative Foxboro 613 Pr (psia) Comments Comments Comments instrumentation (SG Level) RH (%) available to per-Chem. form safety Rad. function. Will be Sub. replaced for long-term monitoring.

II Table 3 Page 9 Reactor: GINNA SYSTEMATIC EVALUATION PROGRAM Tame ENVIRONMENT qua . Document Equipment Type Location Needed Parameter Require Qua . Method Reference Comments

27. Temp Element Area ¹1 Temp ('F) See 200 Spec 35 Not required to Rosemount 176JA

/ Pr (psia)

RH (%)

Comments Atm.

Amb.

Experience Experience function for short-term DBE. Will be

(,RTDs ) Chem. replaced for long-Rad. 200 R/hr Spec 35 term monitoring Sub.

28. Battery Area ¹8 Long Temp ('F) Amb. 110 Vendor Data 9,32 Not exposed Gould/FTA-19 Pr (psia) Atm. Atm. Experience to DBE RH (%) Amb. Amb. Experience Chem. No Rad. No Sub. No 29a. Diesel Generator Area ¹4 Long Temp ('F) Amb. Amb. Experience 7 Not exposed to ALCO Diesel Pr (psia) Atm. Atm. Experience DBE 251F RH (%) Amb. Amb. Experience
b. Westinghouse 1900 KW Chem. No Generator Rad. No
c. Westinghouse fuel oil Sub. No transfer pump - 1 HP-model TEFC Class PMF Insulation
30. Motor, Containment Area ¹1 Long Temp ('F) 286 320 Test 18,19,20, Fan Coolers Pr (psia) 75 95 Test 64,65, Westinghouse RH (%) 100 100 Test 67,70 588.5-CSP Chem. Yes Yes 8 Test Rad. 1.6x10 2xlo Test Sub. No
31. Circuit Breaker Area ¹3 Seconds Temp ('F) See Amb. Experience Equipment will Westinghouse Pr (psia) Comments Atm. Experience fail-safe on DB-50A 1600A RH (%) Amb. Experience loss of power Chem.

Rad.

Sub.

Table 3 Page 10 Reactor: GINNA SYSTEMATIC EVALUATION PROGRAM Tame I ENV RONMENT qua . Document Equipment Type Location Needed Parameter Require Qua . Method Reference Comments

32. IRC Cabinets Area 08 Long Temp ('F) Amb. Amb. Experience Not exposed Foxboro Pr (psia) Atm. Atm. Experience to DBE RH (%) Amb. Amb. Experience Chem. No Rad. No Sub. No
33. HVAC Area 58 Long Temp ('F) Amb. 122 Spec 4,6 Not exposed to Westinghouse Pr (psia) Atm. Atm. Experience DBE 2162 (%) Amb. Amb. Experience

{Control Room AHU) Chem. No Rad. No Sub. No

34. Splice Sleeves Area 51 Long Temp (4F) 286 340 Test 36,38,51 56,62 Raychem Pr (psia) 75 118 Test WCSF-N RH {%) 100 100 Test Chem. Yes es Test Rad. 1.6x10 2x10 8 Sub. No
35. Solenoids/ Area Ol Long Temp ('F) 286 346 Test Valcor V57300 Pr (psia) 75 128 Test (Pressurizer PORVs) RH (%) 100 100 Test Chem. Yes Yes 8 Test Rad. 1.6x10 2x10 Test Sub. No

,'36. Level Switches Area 41 Temp ('F) See See 52 Not required to GEM Corp. Pr (psia) Comments Comments perform safety Model:Special- RH (%) function. How-Similar to LS-1900 Chem. will be replaced (Containment "B" Level)

Sump Rad. for TMI-STLL Sub.

c, Table 3 Page ll Reactor: GINNA SYSTEMATIC EVALUATION PROGRAM T1me ENVIRONMENT Qua Document Eguipment Type Location Needed Parameter Requ1re Qua . Method Reference Comments

37. H2 Recombiner Area 41 Long Temp ('F) 286 315 Test 18,19,49 Igniter Exciter Unit Pr (psia) 75 105 Test 18,19,49 GLA Part No. 43737, RH (%) 100 100 Test 18,19,49 Rev. A, Chem. Yes Yes Test 18,19,49 Serial 001 Rad. 1.6xlo 1.73x10 Test 18,19,49 Sub. No
38. H2 Recombiner Area 51 Long Temp (OF 286 286 Te'st 18,19,49 Blower Motor (2/15 Pr (psia) 75 75 Test 18,19,49 Scale) W 2 HP, RH (%) 100 100 Test 18,19,49 Class H Ins., Model Chem. Yes Yes Test 18,19,49 TBFC Rad. 1.6xl0 2.0x10 Test SO 68C24196 18, 19,49 Sub. No No
39. Pump Motor Area N2 Long Temp (OF) Amb. Amb. Experience Not exposed to U.S. Electrical Pr (psia) Atm. Atm. Experience DBE environment Motors RH (%) Amb. Amb. Experience Model VEU, 100 HP Chem. No Frame 84-445 U Rad. No Insulation Class B Sub. No (Charging Pump)
40. Solenoids/ Area 58 Short Temp ('F) Amb. Amb. Experience Not exposed to Johnson Controls Pr (psia) Atm. Atm. Experience DBE environment Model D251 RH (%) Amb. Amb. Experience (Control Room Air Chem. No Handling Rad. No Unit Dampers) Sub. No
41. Medium Voltage Area 07 Short Temp ('F) Amb. Amb. Experience Breakers need I Switchgear Pr (psia) Atm. Atm. Experience only open for Westinghouse RH (%) Amb. Amb. Experience LOCA inside DH - 350E Chem. No containment to cc 1200 A Breakers Rad. No stop RC pumps.

(RCP Trip Breakers) Sub. No Not exposed to DBE when needed to function.

Table 3 Page 12 Reactor: GINNA SYSTEMATIC EVALUATION PROGRAM Tame ENVIRONMENT Qua - Document, Equipment Type Location Needed Parameter Require Qua . Method Reference Comments

42. RHR Pump Cooling Area 02 Long Temp ('F) Amb. Amb. Experience Only exposed to System Fan Motors Pr (psia) Atm. Atm. Experience DBE radiation Westinghouse Model RH (%) Amb.

7 Experience environment SBDP Chem. 3xlO lx10 Test 69 Class B Insulation- Rad. No 2HP Sub. No

43. Cont Spray/SI Pump Area 52 Long Temp ('F) Amb. Amb. Experience Only exposed and Charging Pump Pr (psia) Atm. Atm. Experience to DBE radiation Cooling Systems RH (%) Amb.6 Amb. Experience environment Fan Motors Chem. 3x10 1x10 > Test 69 Westinghouse Model Rad. No SBDP Sub. No Class B Insulation-3HP
44. Main Control Board Area N2 Long See "Mild" Environment.

Reactor Trip Racks Comments be addressed at Relay Logic and a later time Test Racks Miscellaneous Racks Auxiliary Relay Racks Safeguard Racks Reactor Coolant System Racks CVCS Racks Feedwater Control Racks SI Sequence Racks

C I

Table 4 Environmental Service Conditions Inside Containment Normal 0 eration Temperature: 60-120 F Pressure: 0 psig Humidity: 50% (nominal)

Radiation: 1 Rad/hr general. Can be higher or lower near specific components.

Temperature: Figur'e 5 (286'F max)

Pressure: Figure 4 (60 psig design)

Humidity: 100%

Radiation: Figure 6 (1.6 x 10 total)

Chem. Spray: Solution of boric acid (2000 to 3000 ppm boron) plus NaOH in water.

Solution pH between 8 and 10.

Flooding: 7 ft (approx)

Auxiliar Buildin Normal 0 eration Temperature: 50-104 F Pressure: 0 psig, Humidity: 60% (nominal)

Radiation: 10 mr/hr general, with areas near RHR piping < 100 mr/hr during shutdown operation Accident Conditions includin sum recirculation Temperature: 50-104'F (122'F near motors)

Pressure: 0 psig Humidity: 60% (nominal)

Radiation: Operating Floor (271'lev.):

Near Bus 14 and NCC 1C 6 1L:

100 rad Other Areas: less than 50 rad Intermediate Floor (253'lev.):

Near Bus 16 and MCC 1D 8 1N: 900 rad Other Areas: less than 500 rad Basement Floor (236'lev.):

Near CS, RHR, an( SI Pumps: 2.8 x 10 pads Other areas: < 10 rads Spray: N/A Flooding: N/A

C. Intermediate Buildin Normal 0 eratzon Temperature: 50-104'F Pressure: 0 psig Humidity: 60% (nominal)

Radiation: 1 mr/hr (higher near reactor coolant sampling lines)

Accident Condition Based u on HELB or MELB Temperature: 215'F for 30 minutes; then reducing to 104 within 3 hrs Pressure: 0.8 psig for 30 minutes; then reducing to O,psig within 3 hrs Humidity: 100% indefinitely Radiation: N/A Spray: N/A Flooding: 0 Based u on LOCA conditions Temperature: 115'F indefinitely* near large motors and FW and SL piping. 104'F in open areas Pressure: 0 psig Humidity: 100%

Radiation: Negligible Spray: N/A Flooding: 0 D. Cable Tunnel Same as Intermediate Building E. Control Buildin Control Room Normal 0 eration Temperature: 50-104'F (usually 70-78'F)

Pressure: 0 psig Humidity: 60% (nominal)

Radiation: Negligible Accident Conditions Temperature: 104oF Pressure: 0. psig Humidity: 60% (nominal)

Radiation: Negligible Spray: N/A Flooding: N/A

  • Estimated (no explicit calculations performed)

~1 Normal 0 eration Temperature: 50-104 F Pressure: 0 psig Humidity: 60% (nominal)

Radiation: Negligible J

Accident Conditions Temperature: 104 F Pressure: 0 psig Humidity. 60% (nominal)

Radiation: Negligible Spray: N/A Flooding: N/A Normal 0 eration Temperature: 50-104 F Pressure: 0 psig Humidity: 60% (nominal)

Radiation: Negligible Accident Conditions Temperature: < 104'F Pressure: 0 psig Humidity: 60% (nominal)

Radiation: Negligible Spray N/A Flooding: N/A Necbanical E i ment Room Normal 0 eratzon Temperature: 50-104 F Pressure: 0 psig Humidity: 60% (nominal)

Radiation: Negligible Accident, Conditions Temperature: < 104'F Pressure: 0 psig Humidity: 60% (nominal)

Radiation: .Negligible Spray: None Flooding: 3 ft. (estimated for a service water line leak)

F. Diesel Generator Rooms Normal 0 eratxon Temperature: 60-104 F Pressure: 0 psig Humidity: 60% (nominal)

Radiation: Negligible Accident Conditions Temperature: 104 F Pressure: 0 psig Humidity: 90% (estimated)

Radiation: Negligible Spray: N/A Flooding: 0 ft **

G. Turbine Buildin Normal 0 eration Temperature: 50-104 F Pressure: 0 psig Humidity: 60% (nominal)

Radiation: Negligible Accident Conditions Temperature: 220'F'or 30 minutes, reduce to 100'F within 3 hrs.

Pressure: 1.14 psig on mezzanine and basement levels, 0.7 psig on operating floor Humidity: 100%

Radiation: Negligible Spray: N/A Flooding: 18'~ in basement (Circ. Water Break)

H. Auxiliar Buildin Annex Normal 0 eratzon Temperature: 60-120 F Pressure: 0 psig Humidity: 60% (nominal)

Radiation: Negligible Accident.'Conditions Temperature: 60-120 F Pressure: . 0 psig Humidity: 60% (normal)

Radiation: Negligible Spray: N/A Flooding: 2 ft.

    • Service water line crack would affect only one room (see FEOOD-15)

Screenhouse Normal 0 eration Temperature: 50-104 F Pressure: 0 psig Humidity: 60% (nominal)

Radiation: Negligible Accident Conditions:

Temperature: < 104 F Pressure: 0 psig Humidity: 60% (nominal)

Radiation: Negligible Spray: N/A Flooding: 18" (Circ. Water Break)

Deeda Basf.s Accident Temperature -. Time Curve I ~

~ ~ ~

~ .. ~ ~ ~

(

~

~ r I t g

~ ~l f

...I q,) .I. I.- ~

~ o t

$ 00 oo

~

Containment ~

~

Temperature Conservative Representation I~ ~

i ~

o oE Containment Prcssure Q

~

I

~ ~ ~

I 5 0I I

~ I , .

I IO

'l

~

)

I Sump Temperature ~

. .

I I

o I ~

o

~

~

'I  :, .

'

0~:. ~ ~ ':-.::.'~: ':. ~

.

.'.. -i;:>>~I-:~ f--

I 5$ 0-- Ii.l I l I ~

I ~ I I 1O O o

~ ~ H Heat Exchanger Outlet I

.

150 Temperature I* I ~ I '" ' I

'-' I" -I-l-j..

I I

'I

~ ~

l. I.

~

~

~

~

.

'II' \ ~

  • I ~ I ~ .,I l

~ '

t.l ~ . I la: ....j.L I o ~

I' I. ~ ~ . I~ ~

I

~

i) .,I

~ ~ ~

.

~ ~

lj,j I o I~

I I,'.- ~ ~ ~ o I

.Ju I I I ~

~ ~

i

~

r l-i I 'I-f I .j.l P~- ~ -- -., I.o I- s

~ ~

I ~

~ I~ )

r I ~

~ ~ ~

I ~ I ~:;i . 16 ~ 6+18-.': '. ~

I I ~ ~ <- ~ ', -t ~ ~ 4- t-I .I I ) Jay -t~

I I

'03 10 10 10 10 Tim After 9 sl.gn 2asis Accident ('econds)

Po g gQg Q

Post-Accident ConMinnent Y~terials Design Conditions

~ ~ ~, ~ ~ ~,

~ ~ ~

r.

~,

~ ~

~

~ ~

' ',;

I ~....'

', ~

~

I I

~

~,

'

~

~

!

'

~

~ ';;,

I ~ ~

~ '

~ .

I

~

~

~

'gl'

.

'

3

~,!

~, .

~,

j

! ~

...,~ ~

~ ~ ~

~,

~ ~

I%PE' P

~

~ '

~

I ~ 0 ~ y w,t I I ~

I ~ ~ I ~ ~~ L ~

~ ~

~ I ~

250 .

~ I ~ ' ~ ~

2OO I

g C 0  ! I ~ I O I~~

I 1.~'.- .

~ I~ +~

I I

~

" ',  ! .' ~ ' ' ' ' ' '

1M' j

~ ~ ~

!' I a 4 \ ~ 'II

~

~

~ ~ ~ ~ g g

~

4

~ ~

~ I 4 ~ I ~ l

~ ~

~,

I

~ ' ~ ~ ~ 4

~ I Figurc 5

Contninment Atmosphere Intcgrnted Cnmmn Dose Level 109, 5 4 S 6799) 4 5 67091 2 5 4 5 6769) 4 5 I) 799) 2 5 4 5 67991 6

~ ~

lt

~ ~ I'i ti,i ~ << + i ~ r ifr Ia

~ it) I t!l ~ 'Ial 1~ If Ii rti ~

Zjj )li C L ia<< aa' ~ t it! Ia L 7

6

'Ia air ~,,',i S

'.I ' I I ~ I ~ ~ f I~i.i I

~

il IIJ tra aig ita 3

~ I Bf ~

J 2

I )ra -'p.

~ li ~

a ~

cl gl 10 M ~

)<<

6

~ ~ I ~ I a ~

I'. ~ I ~ ~ I

'I

~ ~

a

~ I ~ ~ ', "I' j<<: I L'j ~ ~ I<< I~I I~ ~ I ilia I,

~

gj IZ} 7 ' I at! iiia II ~ La I

6 ll I ar .Z ':

F

)4 j.'i'l ~~a' ~

ta

~ I~

I JII+II ~

Ii' Iiii ~

ia I ~I rI'I I  :~

a I

oIJ I I II I

!;I

'I

~~

Ia,!

~ Il

't t" iF) 2

ltt Ix I Jl'I Ill; ~ I IL 10 wa

~ I

.~

3 II!

9 6

~~ t' ~ I ta tr t' ', ~ I I~ I a ~

I} ~ ! ~ il .;1 t f~I ~ !at ) ~ jj: ~ I 7 ' I ' ~

I 6 L IR'} J~li: I 'I 5 . )p,  ; I'9

'Pi, I'l I

I ~

WI.~J I Jli a I

<<I, I ~ ~

~

I I ' Il'i'
. }~ii Ir I:.y iI I 11 I rl tl I) I . II':

j3'i

~ ~

I 1 la! it il I!I: I I:

1 L

i)i'} ""') IIII II

<< I~

Ii}!J:":

La)J 

                                                                                                                                                                                                                                                                                                             ~
                                                                                                                                                                                                                                                                                                                  -'}LI II'I                                                      il}                                                                                              aMI                    illiI I ~                 ~:                           tj
                                                                                                                                                                                                                                                           )lI:,:                                                          ~
',I:
               -"   xg WW                                                                5))                  a il!
                                                                                                                                                                                                                                          ~a
                                                                                                           'I ~ I          '                                                                 ~ I                           1+4'.I Il"
                                                                                       }
                                                                                           ~     ~

I IL " ii I 

                                                                                                                                          'al                           l dny             I ) I;.                       1 week I  II 1 month                                                                  1 yent 10                                                                                                                                                                                   I                                               I I t 2                                                      10-'-

10 10 10 J.Lf 10 Time After. Activity Relensc (hours) Figur)

I ,I' 

   >i
   <I'r
 )~f

GINNA STATION (DOCUMENTATION REFERENCE)

l. Crouse-Hinds Penetration Test Report 2 ~ Gilbert Spec. 520 Standby AFN Pumps 3 ~ Gilbert Spec. 711 Standby AFW Pump Motors 4, Gilbert Spec. 5201 Large Motors
5. Deleted. Included in Reference 51
6. Gilbert Spec. 5342 HVAC Throughout Ginna 7 ~ Gilbert Spec. RO-2239 Diesel Generators
8. Gilbert Spec. RO-2267 Auxiliary Feedwater Pumps 9 ~ Gilbert Spec. RO-2400 Batteries
10. IPCEA Std. S-61-402, Sect. 3.8 and 4.3.1
11. Kerite Memo 7/22/68
12. NEMA Std. SG-3, Low Voltage Circuit Breakers
13. Nestinghouse Spec. 676258 - Motor Operated Valves
    '3.d
14. Westinghouse Spec. 676270 Control Valves
15. Westinghouse Spec. 676370 Auxiliary Pumps
16. Westinghouse Spec. 676427 Auxiliary Pump Motors
17. NCAP 7343 June, 1969
18. NCAP 7410-L, Vol. I & II
19. WCAP 7744, Vol. I 8 II 20 NCAP 9003, January, 1969
  '1-Deleted. Included in Reference     45 22        Deleted
  '3 Report NS-CE-775, Pail-Safe Operation of     ASCO  Solen    s
   '4
        .Copes-Vulcan Solenoid Valves
  '5.

Vendor Data on Laurence Solenoid 26 Vendor Data on Versa Solenoid 

   '7.

WCAP 7153 28 Deleted. Included in Reference 45 

  '9 Gilbert  Spec. 504  Westinghouse  Electrical  Penetra tions
   '0 Technical .Proposal for Electric Penetration for Gin na
  '1     Containment Structure by Nesti'nghouse  September      4  1974 NCAP 7354-L
    '2 Vendor Data on Gould Batteries
    '3.

Westinghouse Spec. Sheet for Foxboro Transmitters 34 Vendor Data on Barton 209 Transmitter 

    '5 Rosemont RTD Spec.
     '6 Vendor Data on Raychem Splice Sleeves
    '7 June 16, 1975 Letter to R-.A. Purple        from  L. D. White on
    '8   Containment Flooding April 4, 1979 FRC Final       Report  F-C5074,    Splice   Sleeves
    '9   and Cable Deleted
   '0 Deleted

( I I R J 

   )

GINNA STATION (DOCUMENTATION REFERENCE) CONT'D 41 

   '2 Deleted
  '3.

Design Criteria Standby Aux. Feedwater System October 24, 1974 44 ~ Limit Switches 45 Design Approval Test on Material Used in Westinghouse 

  'eleted Penetrations for the Brunswick Station of Carolina Power
    '6 and Light Company  August     ll, Test Data for Coleman and Rome Cable 1972
   '7.

Aging Failure Detect.ion Program 48 Valcor Solenoid Valve: Vendor Data and Test Report Extracts 

    '9; WCAP-9001 50       Westinghouse Terminal Blocks
    '1.

Cable Identificat.ion and Qualification Supplement, Including F-C5074 (Supplement) Concerning Silicone-Rubber-Insulated Cable Qualificat.ion

52. Wide-Range Sump Level Switch Specification
53. Limitorque Valve Operator Data, Including Limitorque Report B0003 and Section 4.1.4 of B0058.
54. Containment, Electrical Penetrations
55. Kerite Letter, June 26, 1980
56. IE Inspections 78-20 and 78-21 Reports Concerning Installation of Splice Sleeves
57. Control Valve Specification SP-513-044666-000, September 27., 1974, Concerning .Standby ApW Valves
58. Westinghouse 10/10/80 Letter Concerning Crouse-Hinds Electrical Penetrations
59. Evaluation of Organic Materials on Crouse-Hinds Electrical Penetrations 60 Westinghouse Terminal Block Information on Aging and Radiation
  '1 Aging Evaluation of Westinghouse Electrical Penetrat.ions Raychem Splice Sleeve Aging Information
   '2
  '3 Kerite Cable Aging Information
  '4.

Containment Fan Cooler Motor Splices 65 '6 Safety-Rel'ated Motor Bearings Maintenance and Lubrication 

                                           .
  ~

67 Safety-Related Motor Characteristics (Insulation) 

  '8.

WCAP-8754 69 Westinghouse Research Report 71-1C2-RADMC-Rl, December 31, 1970 (Revised April 10, '1971), Concerning "The Effect, 

  '0    of Radiation on Insulating Materials Used in Westinghouse Medium Motors" WCAP-7829, "Fan    Cooler Motor Unit Test"

I J J J; P ~ f}}

Retrieved from "https://kanterella.com/w/index.php?title=ML17250A715&oldid=2000676"