ML17250A715

Environ Qualification of Electrical Equipment, Revision 3

ML17250A715
Person / Time
Site:	Ginna
Issue date:	10/31/1980
From:	ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17250A714	List: ML17250A713 ML17250A715
References
TASK-03-12, TASK-3-12, TASK-RR NUDOCS 8011040240
Download: ML17250A715 (159)

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.

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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

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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.

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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.

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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.

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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.

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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 .

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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.

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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 ~

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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"

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