ML19309F380

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Environ Qualification of Electrical Equipment, Revision 2
ML19309F380
Person / Time
Site: Ginna 
Issue date: 04/25/1980
From:
ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17249A855 List:
References
TASK-03-12, TASK-3-12, TASK-RR NUDOCS 8004290381
Download: ML19309F380 (75)
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{{#Wiki_filter:i.. 80042903 { 8 ( '8 %.) Environmental Qualification 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

TABLE OF CONTENTS Page I. Introduction 1 II. Identification of Necessary Safety Related Equipment 2 A. Events Accompanying a Loss of Coolant Accident 2 B. Events Accompanying a Main Steam Line Break or a Main Feed Line Break 9 C. High Energy Line Breaks Outside Containment 14 D. Flooding Outside Containment 16 III. Identification of the Limiting Service Environmental Conditions for Equipment which is Required to Function to Mitigate the Consequences of Events Identified Above 17 A. Inside Containment 17 B. Auxiliary Building 19 C. Intermediate Building 19 D. Cable Tunnel 20 E. Control Building 20 F. Diesel Generator Rooms 21 G. Turbine Building 21 H. Auxiliary Building Annex 21 I. Screen House 21 J. Loss of Air Conditioning 22 IV. Equipment Qualification Information 23 A. Auxiliary Feedwater Pumps 23 B. Valves 878 A,B,C,D 23 C. Main Steam Isolation Valves 23 D. Main Feedwater and Bypass Isolation Valves 24 E. Containment Fan Cooler Dampers 24 F. Barton 332 Transmitter 25 G. Foxboro 611 Transmitter 26 H. Foxboro 613 Transmitter 27 I. Instrumentation Terminal Blocks 28 J. Cable 29 K. Reactor Coolant System Temperature Detectors 30 L. Safety Related Cable Splices Subject to LOCA and MSLB Effects 31 M. Aging of Equipment Prior to Qualification Testing 32 V. Conclusions 35

LIST OF FIGURES Figure 1 Loss of Coolant Accident [ Sequence of Events Diagram] Figure 2 Main Steam or Feed Line Break [ Sequence of Events Diagram] Figure 3 Plant Layout

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 Qualification i r l

Environmental Qualification of Safety-Related Electrical Equipment I. INTRODUCTION The equipment and systems identified in this report are classified in accordance with IEEE 308-1974 as Class IE. The purpose of this report is to establish the level of environmental qualification of the Class IE equipment required during and after a loss of coolant accident (LOCA), main steam line break (MSLB), main feedwater line break (MFLB) or high energy line break outside containment in a format which pro-vides for convenient comparison with the limiting design en-vironment in which it is required to perform a safety function. Table 3 summarizes this information. o 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 avail-able), and is therefore not required to be qualified. A. Events Accomoanvinc 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 Emer-gency Procedures. 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 is performed. Equipment listed as "long term" is required to provide long term decay heat removal, post-accident monitoring and sampling, or achieving and maintaining a safe shutdown condition. Equipment listed as "short term" is required only for a short period of time (hours-days). Table 3 provides environmental qualification of all the Ginna Class IE equipment, in the format requested for SEP by the NRC's September 6, 1978 letter. 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". 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 busses assuming there is a loss of offsite power. With the safeguards busses 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 auxiliary feedwater pumps, and the four containment fan coolers will be loaded sequentially onto the busses. The two containment spray pumps are automatically loaded onto the busses when the appropriate containment pressure setpoint is 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 psig. 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 MOV 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 in-jection signal. The SI signal also causes a trip of the main feedwater pumps (which in-turn causes the closing of the feedwater discharge valves). i w

5. " Containment Isolation" and " Containment Ventilation Isolation" (referred to as simply " Containment Iso-lation") is initiated by the safety 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 con-tainment 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 con-tainment recirculation fans. The fail safe 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 safe-guards busses. 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 i the containment building integrated leak rate tests. Valve 313, the reactor coolant pumps seal water return i line, and valves 813 and 814, reactor coolant support l inlet and outlet lines, are closed by the containment isolation signal. -

6. The S.I. 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 S.I. 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 discharge 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 826A & 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 clos-ing 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 injection lines. These valves open upon receipt of a safety injection signal and remain open thereafter. The containment spray pumps will automatically start and the discharge valves 860A, B, C and D automatically open, receiving power from the safeguards busses 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 safe-guards are loaded on the safeguards busses. 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. SIS actuation will automatically align the two post accident charcoal filters to the containment recircula-tion system by opening inlet valves 5871 and 5872, and outlet valves 5873 and 5874. Loop entry dampers 5875 and 5876 will close. 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. After the safety injection pumps are automatically switched from the boric acid storage tanks to the re-fueling water storage tanks, the operator resets safety l 1 l

injection, starts the component cooling water pumps and aligns flow to the RHR heat exchangers, and initiates SW flow to the CCW 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 flow paths to the reactor vessel for both high and low head safety injection from containment sump B. The normal (non-safety grade) auxiliary feedwater supply source is from the condensate storage tanks. If this supply is exhausted the operator opens 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 the Stand-by AFW system to the SG's (using Service Water suction) from the control room, 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 valve 704A, B, 856, and 896A or B. For low head recirculation, injection is through the vessel nozzles. For high head recirculation, the RHR pumps discharge to the safety in-jection pumps through alignment of valve 857A (for RHR pump B) and/or valves 857B and 857C (for RER pump A). Valves AOV 897, 898 are closed. The high head safety injection pumps then provide water to the cold leg in-jection 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 Deing undertaken, in accordance with the implementation schedule for the TMI Lessons Learned commitments. See [Ref TMI-1]. B. Events Accomoanying 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 l-7] provides additional analysis for steam line breaks outside of containment, as 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 wor.1d occur for a feedwater line break. Since the same equipment is re-quired to operate, and the same emar,ency procedure is _9-

used to mitigate, a feedline break as a steam line break, but a steam line break is a more severe accident in terms of RCS cooldown (return to criticality) and mass and energy release to containment, the subsequent discussion will include 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-tion signal for smaller breaks. 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 redun-dant protection in the event of a steam line break upstream of the check valves, and for all breaks down-stream of the check valves, the main steam line isola-tion 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. Additionally, high containment pressure (30 psig) will initiate Safety Injection. 4. The safety injection signal closes the main and bypass feedwater control valves, trips the feedwater 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 Isolation and Containment Ventilation Isolation valves perform 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 Tafety 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. For a main steam line break inside containment the operator may start the containment spray pumps manually if containment pressure is below 30 psig. 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, the operator will terminate AFW flow to the faulted 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 pumps. At least two hours after the start of the accident, supply water for the auxiliary feedwater pumps will 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 of the Standby AFW system (using Service Water suction) from the control room 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 atmo-sphere) with primary and secondary inventory makeup, accomplished by use of the safety injection and/or the charging 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. High Energy 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. 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 modi-fication. A "reuote-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 l an adverse environment. Ref. [HELB-8] includes the

i i

NRC's Safety Evaluation Report concerning the RG&E modifications resultant from the review of Ref. [HELB-1]. It includes a discussion of the acceptability of the instrumentation relocation and cable re-routing per-formed to insure that sufficient equipment will be protected from the environmental effects of a HELB outside containment. This portion of [HELB-8] is attached to this report. 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 the diesel generator rooms and the screen house in the vicinity of safety related equipment. Modifications are planned which will isolate the steam heating line to the affected area in the event of a l failure and therefore preclude an adverse environment. Prior to its installation, regular inspections are being performed to reduce the likelihood of a failure creating an adverse environment. These inspections, performed during each plant operating shift, would detect any leakage. Plant procedures call for isolation of the affected piping promptly upon detection of the leakage. In addition, confidence of the low likelihood of an adverse environment in the vicinity of safety related equipment is provided by the fact that, in almost 11 years of plant operation, no adverse environment has been created by any mechanism, including failure of the steam heating line. 1 )

_. _ _.. ~. l i ~ I i [ i i D. Flooding Outside Containment l The potential for and protection for submergence of t equipment due to postulated failures in the circulating water system is discussed in References [ Flood 6-10]. 1 i b t t I l l \\ i I J h 4 ) I i 4 4 3 4 i l i e n i j - i ll- ~._

III. IDENTIFICATION OF THE LIMITING SERVICE ENVIRONMENTAL CONDI-TIONS FOR EQUIPMENT WHICH IS REQUIRED TO FUNCTION TO MITIGATE THE CONSEQUENCES OF EVENTS IDENTIFIED ABOVE 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 Figure 1 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 ex-ample, that the doses at 30 minutes and one year follow-6 8 ing a LOCA are 1.7 x 10 R and 1.6 x 10 R, respectively. 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 calculated values. Another example of the conservitism employed is seen in the accident radiation environment used for design purposes. As noted in Raf (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 renoval 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 containment. Thus the radiation environment clearly overstates that which would be present even in a minimum safeguards case. Submergence of valves has previously been discussed in Reference [ Flood-4] and it has been shown that opera-tion following submergence is not required. Submergence of instrumentation has been discussed in Ref [ Flood-5] and in further discussed in Sections IV.G and IV.H of this report. It is shown that operation following sub-mergences is not required. Therefore, no qualification for submerged service is required. 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 for other facilities indicate that the containment vapor temperature following a MSLB in containment may briefly exceed those derived for a LOCA. These higher temperatures should not be limiting, however, for qualification of equipment required following a MSLB, because of the nature of the transient, that is, the fact that the high temperature transient is very brief 1 and there is superheated steam as opposed to saturated i

steam; the location of equipment relative to the steam lines; and equipment configuration, that is, thermal lag. For these reasons, the humidity and steam environ-ment following a LOCA remains limiting. This is con-sistent 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. l\\. Auxiliary Building Post accident environmental conditions in the Auxiliary Building are limited to normal ambient levels. (see also Section J below for a discussion of loss of air-conditioning"). In this regard, measures designed to prevent flooding of the residual heat removal pumps are addressed in Section 9.3.3 of the Ginna FSAR. Radiation environment is presently under review as a result of the TMI Lessons Learned Commitments [see Ref. TMI-1]. C. Intermediate Building Implementation of an augmented inservice inspection program for high energy piping outside containment has reduced the probability of pipe breaks in these systems to acceptable low levels [Ref. HELB-7, 8]. A six inch main steam line branch connection is the Intermediate Building DBE. The limiting pressure is established in Ref. HELB-1 as being a pressure of 0.80 psig. Assuming saturation conditions, one obtains a limiting tempera-ture of approximately 215'F. A 100% humidity steam-air mixture is assumed. Radiation, and chemical spray, the same as for the Auxiliary Building, do not require qualification. The effects of submergence need not be considered, as described in References [HELB-1] and [HELB-4]. Radiation environment is presently under review as a result of the 9HI Lessons Learned Commitments [see Ref. TMI-1]. D. Cable Tunnel Since the cable tunnal is open to the Intermediate Building, the limiting environmental conditions for the cable tunnel ure identical to the Intermediate Building conditions. E. Control Building The limiting environment of the Control Building is normal ambient conditions. (See also Section J below for a discussion of " loss of air-conditioning") Protec-tion against events which could occur outside the control Building and affect the Control Building environ-ment (see Ref. HELB-1) are identified and discussed in i References HELB-1, HELB-6, HELB-7, FLOOD-1, and FLOOD-5. l F. Diesel Generator Rooms The limiting environment in the diesel generator rooms is normal ambient conditions. Protection against failure of steam heating lines in the rooms is des-cribed in Section II C above. Protection against events outside the rooms is described in References 4 HELB-1, HELB-6, HELB-7, FLOOD-1 and Flood-5. G. Turbine Building A pressure of 1.14 psig on the mezzanine and basement floors ar.a 0.7 psig on the operating floor, with saturated steam conditions, is the limiting environment of the Turbine Building. H. Auxiliary Building 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. I. Screen House The limiting environment-in the Screen House is normal ambient conditions. frotection against flooding is described in-References FLOOD-1 and FLOOD-5. (Also see Section J belc' for a discusssion of " loss of air-con-ditioning") i

J. Loss of Air-Conditioning The Ginna plant does not include a redundant safety re-lated air-conditioning system for any area except the containment and the newly-installed Auxiliary Building annex. This was discusse'd during the NRC's SEP Site Hazards site vsit of September, 1978. In the event of a loss of air-conditioning, most areas of the plant would not be subject to significant temperature buildup, since these areas (auxiliary building, intermediate building, and screenhouse), are large volume buildings. The diesel generator rooms can be brought to essentially ambient temperature by opening the access Soors at the north end of each room. This provides an un.timited supply of outside air. These spaces do not require continuous manning during an emergency and therefore the potential for radiological exposure of personnel in this mode of operation is very small. The control room air handling unit is powered from a single Class lE motor control center (MCC 1K). If there is a failure of this train (Mcc 1C which.is fed by the 1A diesel) during the post accident period, it is pos-sible to crosstie to the 1B diesel. The operator, after assuring that any faults are cleared, closes the bus tie between busses 14 and 16 to energize the inoperative Control Room air handling unit from the 1B diesel, while making sure that the operational diesel does not become overloaded..

IV. EQUIPMENT QUALIFICATION INFORMATivli Table 3 summarizes the qualification of electrical equipment. Information contained in this section augments the qualifi-cation information of Table 3. A. Auxiliarv Feedwater Pumps The auxiliary feedwater pumps are located in the base-ment of the intermediate building (area 2). The environ-ment qualifi. tion of these pumps is standard industrial level. The consequences of an unlikely failure of one or both of these pumps is acceptable as this system has a redundant system (the Standby AFW System) which is not exposed to an adverse environment (Section III C). B. Valves 878A, B, C and D Valves 878 A, B, C and D are located in the containment basement and are Limitorque valves with Peerless motors. Juring normal operation these valves are positioned in their safeguards position with AC power removed so as ta preclude failure in the event of an accident. Therafore, exposure to an adverse environment is of no consequence to valve performance. C. Main Steam Isolation Valves The main steam isolation valves are located in the intermediate building (area 3). As discussed in Section II.B.3 above, the steam line non-return check valves provide protection redundant to ~ the main steam isolation valves for breaks upstream of the check valves. Due to equipment configuration, the check valves provide protection for all branch line breaks in the Intermediate Building and for all postulated crack breaks in the Intermediate Building except for - cracks in the few feet of pipe between the check valves and the Intermediate Building and Turbine Building wall. Based on vendor data, it is expected that the valves will perform their function up to a temperature of 250*F, since the solenoid valves are enclosed in a NEMA-2 drip-proof enclosure. D. Main Feedwater and Feedwater Bypass Isolation Valves The main feedwater and feedwater bypass isolation valves are located in the turbine building at the intermediate floor level. These valves close upon receiving the safety injection signal for a main steam line break or loss of coolant accident and fail closed with a loss of air or power. E. Containment Fan Cooler Dampers The containment fan cooler dampers are located in the basement of the containment building and will be sub-jected to the environment of the main steam line break (inside containment) and the loss of coolant accidents. These valves do fail in the " fail-safe" accident position [FSAR - Section 6.3]. i i 1 '

E. Barton 332 Transmitter The Barton 332 transmitters are used to detect steam flow for each steam generator and generate a portion of the signal for steam line isolation. These transmitters are located on the operating floor of the containment and would be exposed to the adverse environment produced by a main steam line break (inside containment). The Barton 332 transmitters are housed in a Nema 4 enclosure for protection against water. For temperature qualification an ovendry bake test at 320*F for 66 hours has been successfully performed. These transmitters do not need to function for any steam line break where they would be exposed to an elevated environment. For breaks inside containment, the non-return check valves which are provided in each steam line will assure that the intact steam generator is properly isolated. A signal to the main steam isolation valves is also provided by high containment pressure. For a break outside containment, the nonreturn check valves will provide isolation if the break is upstream of the header. The steam flow transmitters, which will be unaffected by the break, will provide for main steam isolation valve closure. G. Foxboro 611 Transmitters Pressurizer pressure measurements are determined by the Foxboro 611 GM-DSI transmitter. As discussed pre-viously, two out of three low pressurizer pressure signals initiate safety injection for the DBE's of concern. These Foxboro 611 transmitters have successfully under-gone pressure, temperature, and humidity tests as described in WCAP 7410-L and WCAP 7354-L for the LOCA DBE. Radiation exposure testing and knowledge of material properties indicates that the transmitter may 4 experience failure at exposures of greater than 3 x 10 rads. LOCA analyses, References [LOCA 2-4] demonstrate that fuel failures in a LOCA do not occur until well after the pressurizer pressure has decreased below the SI initiation set point. No fuel failures are predicted in a MSLB. Thus, it can be concluded that the safety injection initiation function of the Foxboro 611 trans-mitters will not be affected by irradiation. For post-accident monitoring, these transmitters would be used for operator indication in the event of a small LOCA or a secondary side break. If unavailable due to radiation or flooding failure, however, backup instrumer.- tation is available (SI flow). Furthermore, accident mitigation emergency procedures dictate that, with no indication of stable or increasing pressurizer pressure.

available, SI flow would be maintained to the reactor to ensure continued core cooling. This Foxboro 611 transmitter is also used to measure main steam pressure. These transmitters are located in the intermediate building, and are thus not subject to a post-accident containment environment. For breaks in the intermediate building, other instrumentation located inside contain-ment, such as steam generator level, could be used to perform the required accident functions. H. Foxboro 613 Transmitters The Foxboro 613M-MDL modified transmitters are used to measure pressurizer level. These transmitters have been tested for LOCA environmental qualification condi-tions, as described in WCAP 7410-L and WCAP 7354-L. Tests by Foxboro at 318 F and 90 psig demonstrate that the transmitter will perform its intended safety function. One test by Westinghouse indicates the transmitter had a 7.8% high error after 10 seconds. A second test showed the instrument to perform satis-factorily with errors throughout the test at less than +4% and -5%. Radiation capabilities of the 613 tratsmitters are ex-pected to be similar to those at the Foxboro 611 trans-mitters; however, as with the Foxboro 611, qualification is not required. Pressurizer level is no longer used to generate a SI signal (This change is described in Amendment No. 27 to the Ginna POL, June 15, 1979). Its primary purpose is to provide post-accident information to the operator. However, in the event of a LOCA or loss of secondary coolant, when an adverse radiation or flooding environment could damage this instrumentation, emergency procedures provide for the continued addition of safety injection flow if there is no indication of returning pressurizer level. This ensures continued adequate core cooling. The Foxboro 613 EM-HSI transmitter is used to measure steam generator level. Although this transmitter is not qualified to withstand the post-accident containment environment, instrumentation to determine heat removal via the steam generators is available. The main steam pressure transmitters, located outside containment, i gives indication of the steam generator status. Auxiliary feedwater flow instrumentation, also outside the contain-ment, provides indication of sufficient flow to each steam generator. I. Instrumentation Terminal Blocks Within the containment building, the only safety related instrumentation circuits of concern that utilize ter-minal blocks are those associated with the pressurizer level and pressure transmitters. The transmitters and associated terminal blocks are enc]^ sed in instrument cabinets located at the basement level of the contain-ment., 9

The terminal blocks are Westinghouse 542247 and have the required documentation in support of environmental qualification for use in a DBE environment (see letter from L. D. White, Jr. to B. Grier, USNRC, dated February 10, 1978). Inside containment, all other safety related instruments of concern have terminations within the transmitter enclosure and therefore, will not be exposed to a DBE environment. J. Cable The safety related power, control and instrumentation cable in the containment have been successfully tested for the DBE's. The qualification tests and results are described in WCAP 7410-L, Volume I. All safety related power cable outside of containment is Kerite HT-FR, the same power cable qualified for, and used, in containment. Control cable used outside containment was supplied by General Cable Company, rated 600 volt, PVC insulated, glass braid covering each conductor, 3 mil nylon jacket with overall black PVC jacket [HELB-1, Section 5]. The PVC jacket may soften around 212 F, however the glass braid over the insulation will retain the con-ductor insulation and prevent separation from the con-ductor. The highest credible temperature that this type cable could be exposed to would be a 220 F temperature asso-ciated with a DBE in the turbine building. Thus, a failure of this type control cable is unlikely. Instrumentation cable outside containment for all low level analog signals was supplied by Rome Cable Company, with "Synthenol" insulation covered with glass braid with an overall synthenol jacket. Test reports verify that this cable was heat aged at a temperature of 248*F for 168 hours and the jacket heat aged at 212 F for 120 hours. This cable is not required for safe shutdown of the plant during a high energy line break in the Inter-mediate Building. The safety related instrumentation cable used inside containment was supplied by Coleman Cable and Wire Company. The insulation is silicone rubber with glass jacketing. Silicone rubber insulated cable was qualified for LOCA conditions as described in WCAP 7410-L. K. Reactor Coolant System Temperature Detectors The reactor coolant system temperature detectors (RTD) are not required for a loss of coolant 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 1 Section IV.C 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 downstream 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. Therefore, the RTDs do not require environ-mental qualification to provide their required safety function. The RTDs are alco used as inputs to the subcooling metet. Additional backup to this function is available by use of the in-core thermocouples. The subcooling meter instrumentation design and qualification parameters are presently under review for installation within the TMI Short Term Lessons Learned implementation schedule [Ref. TMI-1]. L. Safety Related Cable Splices Subject to LOCA and MSLB Effects Cable for the safety related pressurizer instrumentation, the core deluge valves, MOV 852 A and B, and the 480V power cable for the fan coolers utilizes Raychem Thermofit, WCSF-N, heat shrink sleeves on the splices subject to LOCA and MSLB effects. These sleeves have been qualified in tests which exceed the worst case Ginna accident environments. Refer to Franklin Institute Research Laboratories Test Report F-C4033-3. This report has been submitted to the NRC by Raychem Corporation. In order to provide further assurance that the splice sleeves used for fan cooler 480V power cables are capable of operating in the accident environment, LOCA/ MSLB tests were parformed on a mock-up of the actual splices in the Ginna containment. This test also included in line splice samples to further verify compatibility of the Raychem splice sleeve materials with the existing cable materials at Ginna. These tests are documented in Franklin Research Center Final Report F-C5074 dated April 1979. M. Aging of Equipment Prior to Qualification Testing Electrical equipment in general consists of components and materials with widely diverse physical properties. When establishing the design lifetime for such equipment, a program for accelerated aging is appropriate only when there is sufficient empirical data or a well understood aging mechanism upon which to base a quantitative estimate, such bases exist for thermally accelerated aging of mrtor and cable insulation. When electronic components are involved, such as in transmitters, no such bases exist and accelerated aging is not appropriate. This type of equipment normally exhibits a failure rate of the " bath-l tub curve" type. That is a period of relatively high failure rate early in life, " infant mortality", followed by a long period of low, constant, random failure rate, which finally terminates at what might be called the "end \\ I

of design life", characterized by a very rapidly incree~- ing failure rate. The objective of aging is to put samples in a condition equivalent to the end-of-life condition. When the equipment or component is of the first type described above, this objective is met by a period of thermal aging, which, based on the data available, can be shown to be equivalent to the design lifetime under normal operation conditions. Qualification tests for motor and cable insulation used in the Ginna Nuclear Plant were conducted using this approach as described in references. For equipment of the second type there is a large amount of evidence to indicate that after an initial i " burn in", sufficient to eliminate components which will be subject to " infant mortality", the failure mechanisms are truly random. That is, the probability 4 of failure, per unit time, is low and constant, and most importantly there is no particular predominant failure mechanism. Under these conditions there is no coupling between the design basis event for which the equipment is to be qualified and the age of the equipment. For such equipment, the "end-of-life condition" is identical with the condition immediately after " burn in". At the Ginna Nuclear Plant the period of operation of such equipment has been long enough to assure that " burn in" has been accomplished. Periodic testing and maintenance assures that any equipment degradation is detected and thoroughly investigated for corrective action. All failures of safety related equipment at Ginna Station are reported, documented, and reviewed for appropriate action (LER, Engineering review, etc.). Component failures are documented and logged under the NPRD data collection program in accordance with its l standard format under the direction of the Ginna Technical Assistant, Operation Assessment Engineer. Since the safety related systems and components generally exhibit very low failure rates, the most sensitve measure of degradation is direct observation by test and maintenance personnel, rather than quantitative analysis of failure rates. Accordingly, the plant procedures which describe the job functions of th Electrical and I & C foremen, and the Tests and Results Department personnel, specifically call attention to I the necessity to evaluate any unusual changes in equipment performance or failure rate. l 1 t i V. CONCLUSIONS 1 It has been determined that the facility will adequately respond to the design basis events as presented in this report and is acceptable for continued operation. I 1 i 'i l I h 4 i d 4 'J l [ 2 f' 4 i L 1 4 I 35 - i f'-- i.

LOSS OF COOLANT ACCIDEF4T 1. 2/3 1. 2/3 HIGH LOW CONTAINMENT PRESSURIZER PRESSURE PRESSURE FIGURE 1 HI HI l l l 'n I s-i l 1 1 SAFETY --i ACCIDENT i l INJECTION DIAGNOSTICS l l u____l U l f h P p i 0 4. 3. 2. 4. 5. 6. fiAIN ACCUMULATOR SAFETY FEEDWATER CONTAINMENT REACTOR STEAM LIflE DUMP INJECTION LINE ISOLATION TRIP ISOLATION SEQUEftCE ISOLATION (AUT0) iL____ v y c. 7. REACTOR VALVES COOLANT PUMP J TRIP Y 9. CONTROL ROOM VENTILATION U 10. MANUAL ACTIONS 11. RECIRC-ULATION e

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 Line Isolation Accident Diagnostics Short Term PT 429, 430, 431, 449 Provide Reactor trip and Short Term Safety Injection signals Signal Initiation Accident Diagonstics la. Steam Line Pressure Accident Diagnostics Short term PT 468, 469, 482 PT 478, 479, 483 Containment Radiation Accident Diagnostics Short term [Being provided per TMI STLL) Containment sump level Accident Diagnostics Short term LT 942, LT 943 2. Safety Injection Sequence (Auto) 1A, IB Diesel Cenerator Power supply to safeguards Long term and Auxiliaries bussess 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 IA, IB,1C Safety Injec-High head injection of bo-Long term tion Pumps rated water to Reactor Coolant System 1A, 1B Containment Spray Containment Pressure, Tem-Long term Pumps (only on hi-hi Cont. perature, and Iodine control pressure) 1A, 1B Residual Heat Re-Low head injection of borated Long term moval Pumps water to Reactor Vessel IA, IB, IC, ID Service Cooling water to RHR and CCW Long term kater Pumps Heat Exchangers IA, 1B, IC, ID Contain-Containment Pressure, Tem-Long term ment Recire. Units perature, and Iodine control 1

-~ .. ~.- TABLE 1 LOSS OF COOLANT.iCCIDENT REQUIRED BLOCK NO./ EQUIPMENT SAFETY FUNCTION OPERATION TIME 1A, IB Motor Driven Cooling water to Steam Gen-Long term Auz. Feedwater Pumps erators 480 Volt Safeguards Provide the distribution of Long term j MCC's IC, ID power to safeguards equipment 3. Accumulator Dump q MOV 841 (N.O.)* Provide path to Reactor Vessel Not required i MOV 865 (N.0,.) from Accumulators for injection to function of borated water 4. Main Steam Line Isolation Feedwater Line Isolation i A0V 3516 Isolate 1A, IB Steam Generators 5 Seconds after A0V 3517 signal A0V 4269 Isolate Main Feedwater System 5 Seconds after ] A0V 4270 signal A0V 4271 A0V 4272 5. Containment Isolation See Text, Section II.A.5 ] 6. Reactor Trip Reactor trip breakers Provide means to trip the reactor Required for Reactor Trip l Reactor protection and in-Provide the instrumentation and Required for i strumentation cabinets protectico circuits for the con-Reactor Trip l trol an4 tripping of the Reactor { 7. RCP Trip RCP Trip Breakers Provide means to trip RCP's Short term ) 8. Valves MOV 825 A,B Provide path to SI Pumps for bor-10% BAST Level M0V 826 A,B,C,D ated water to high head safety or-1/2 hour (B&D N.O.) injection A0V 836 A,B Provide controlled addition of Short term NaOH to Containment Spray for Iodine control MOV 852 A,B Provide path to Reactor Vessel SI initiation of bcrated water for low head safety injection l MOV 860 A,B,C,D Provide path to Containment Spray Long term headers for CS Pumps

  • N.O. = Normally Open

TABLE 1 LOSS OF COOLANT ACCIDENT

REQUIRED BLOCK NO./ EQUIPMENT SAFETY FUNCTION OPERATION TIME BAST Level Indicate BAST Level for automatic 10% BAST Level LT 102, 106, 171, 172 transfer of SI Pump suction from or-1/2 hour BAST to RWST MOV 878 B,D Provide path to cold legs of RCS not required (N.O.)

from high head safety injection to function j MOV 4007, 4008-Provide path for Aux. Feedwater to Short term 1A, 1B Steam Generators A0V 5871, 5872, 5873 Provide path for cleaning of cont. signal initiation A0V 5874, 5875, 5876 atmosphere by fan coolers 9. 2ontrol Room Ventilation Provide cleaning of Control Room Short term I Dampers atmosphere Fans 10. Manual Safety Injection Reset Reset Safety Injection signal less than 24 hours Button after automatic S.I. Sequencing is complete IA, 1B Component Cooling Cooling water for safeguards Long term Water Pumps equipment IA, IB Containment Spray Containment Pressure, Temperature Long term Pumps (if Cont. Pressure and Iodine control <30 psig) 4 RWST Level Indicate RWST Level for operator less than 24 hours i LT 920, LIC 921 transfer from S.I. phase to l Recirculation phase l MOV 4027, 4028 Provide Service Water to Motor within~2 hours l Driven Aux. Feedwater Pumps succion MOV 4734, 4735, 4615, 4616 Direct SW Flow to CCW EX's less than 24 hours MOV 738 A,B Direct CCW Flow to RHR HX's less than 24 hours Standby AFW Pumps AFW Flow to SG's if normal AFW Long term i System inoperable MOV 9629 A,B Provide SW to suction of standby Long term AFW Pumps MOV 9710 A,B; 9703 A,B; Standby AFW Discharge Valves to Long term 9704 A,B provide flow to SG's v

TABLE 1 LOSS OF COOLANT ACCIDENT REQUIRED BLOCK NO./ EQUIPMENT SAFETY FUNCTION OPERATION TIME 11. Recirculation MOV 850 A,B outside cont. Provide path to RHR suction from long term MOV 851 A,B (N.O.) inside B sump for low head safety injec-cont. tion MOV 856 (N.O.) RWST isolation valve to RHR pumps required to func-suction, must close af ter RWST is tion to switch to drained recirc phase MOV 896 A,B 9 (N.0.) RWST isolation valve, must close required to func-after RWST'is drained tion to switch to recirc phase MOV 857 A,B,C Provide path to suction of SI and required to func-CS Pumps from RHR pumps discharge tion to switch to 4 recire phase A0V 897, 898 Isolate high haad recirculation Short term flow to RWST during sump recir-culation MOV 704 A,B Close during switch to sump less than 24 hours recirculation 1 I l 1 ) 1 t f

MAIN STEAM OR FEED LINE BREAK FlGURE 2 h. 2/3 1. 2/3 2. 2/3 HIGH LOW CONTAINMENT STEAM LINE PRESSURIZER PRESSURE PRESSURE PRESSURE i HI HI l l 4 ka[ l 3. 2/4 3. 2/3 3. 2/4! !I I l OVERPOWER y ACCIDENT STEAM LIllE LOW j SAFETY , DIAGNOSTICS l AT FLOW T ave INJECTION l l l L___ l ' "' O l L"'.9c ]F" O f i_ J, i r-J o r n 4. s. 4. s. 7. MAlti SAFETY FEEDWATER STEAM LINE INJECTION LINE CONTAINMENT REACTOR ISOLATION SEQUENCE ISOLATION ISOLATION TRIP (AUT0) p _ _ _ _ _ _Y f 9. 8. REACTOR VALVES COOLANT PUMP TRIP l T 10. MANUAL ACTIONS U 11. CONTINUED SAFE SHUTDOWN

TABLE 2 MAIN STEAM LINE BREAK SAFETY FUNCTION / BREAK LOCATION REQUIRED BLOCK NO./ EQUIPMENT OPERATION TIME INSIDE CV OUTSIDE CV i 1. Steam Line Pressure Provide signal for same signal initiatic PT 468, 469, 482 SI on low steam line PT 478, 479, 483 pressure la. Steam Line Pressure Accident Diagnostics same short term i (see 1 above) Containment Radiation Accident Diagnostics NA short term I 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 3. 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 React.or Coolant Temperature Loop A Hot Leg Provide Low Tave & same signal initiation TE 401A, 402A, aT signals for Reactor 405A, 406A, trip, Safety Injec-409A tion and Main Steam Line Isolation Loop A Cold Leg TE 401B, 402B, 405B, 406B, 409B Loop B Hot Leg TE 403A, 404A, 407A, 408A, 410A Loop B Cold Leg TE 403B, 404B, 407B, { 408B, 410B j 1

TABI 2 MAIN STEAM LINE BREAK ~ l SAFETY FUNCTION / BREAK LOCATION REQUIRED l BLOCK NO./ EQUIPMENT OPERATION TIME INSIDE CV OUTSIDE CV 4. Main Steam Line Isolation A0V 3516 Isolate 1A, B Steam same 5 seconds after signal A0V 3517 Generators Feedwater Line Isolation A0V 4269 Isolate Main Feed-same 5 seconds after signal A0V 4270 water system A0V 4271 A0V 4272 i S. Containment Isolation See Text, Section same II.B.5 6. Safety Injection Sequence (Auto) 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 lA, IB, IC Safety In-High head injection same Long term Jection Pumps of borated water to Reactor Coolant System 4 1A, B Containment Spray Containment Pressure N/A Long term Pumps (only on hi-hi cont. and Temperature j Pressure) control 1A, IB, IC, ID Service Cooling Water to same Long term Water Pumps CCW Heat Exchanger IA, IB, IC, ID Containment Containment Pressure N/A Long term Recire Units .and Temperature con-trol 1A, IB Motor Driven Aux. Cooling water supply same Long term Feedwater Pumps to Steam Generators

TABLE 2 MAIN STEAM LINE BREAK SAFETY FUNCTION / BREAK LOCATION REQUIRED BLOCK NO./ EQUIPMENT OPERATION TIME INSIDE CV OUTSIDE CV 480 Volt Safeguards Provide the distribu-same Long term MCC's IC, ID tion of power to safeguards equipment 7. Reactor Trip Reactor trip breakers Provide means to same Required for trip the reactor Reactor Trip Reactor Protection and Provide the instru-same Required for Instrumentation mentation and pro-Reactor Trip Cabinets tection circuits for the control and tripping of the reactor 8. Reactor Coolant Pump Trip Provide means to trip same Short term RCP Trip Breakers RCP's 9. Valves MOV 825A, B Provide path to SI same 10% BAST Level MOV 826A, B, C, D Pumps for borated or-1/2 hour (B&D N.O.) water to high head safety injection A0V 836A, B Provide NaOH to CS if Short term needed MOV 860A, B, C, D Provide path to Con-N/A Long term tainment Spray i headers for CS Pumps MOV 878, B, D Provide path to same not required to (N.O.) cold legs of RCS function from high head i safety injection MOV 896, A, B (N.0.) Provide path from same short-term (to close RWST of borated if need sump water for SI and recirculation) CS pumps suction MOV 4007, 4008 Provide path for Aux. same Short term Feedwater to Steam Generators 1 A0V 5871, 5872, 5873 Provide path for N/A gnal initiation -A0V 5874, 5875, 5876 cleaning by fan l coolers, cooling of i cont. Atmosphere l \\ l

TABLE 2 MAIN STEAM LINE BREAK SAFETY FUNCTION / BREAK LOCATI0h REQUIRED BLOCK NO./ EQUIPMENT OPERATION TIME INSIDE CV OUTSIDE CV BAST Level Indicate BAST Level same 10% BAST Level LT 102, 106, 171, 172 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 had SI to Reactor Vessel 10. Manual SG Level Instrumentation Determine affected SG same short term i LT 470, 471, 472, 473 LT 460, 461, 462, 463 Safety Injection Reset Reset SI signal after same less than 24 hours Button Automatic SI sequenc-ing is complete IA, 1B Component Cooling Cooling Water for same Long term Water Pumps safeguards equipment 1A, IB Containment Containment Pressure N/A Long term Spray Pump (If cont. and Temperature con-Pressure < 30 psig) trol MOV 4027, 4028 Provide Service Water same within ~2 hours to Motor Driven Aux. Feedwater Pumps Suction j Charging pumps Inventory control to same Long term RCS Standby AFW pumps Provide AFW flow to same Long term SG's if AFW system in-operable MOV 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 SG's MOV 4000A, B AFW Cross-Connect same-Short. term Valves i i I

Accident References LOCA analysis [LOCA] 1. 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 #1, 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 El.1 and El.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] 1. FSAR 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. 4. Letter dated May 24, 1977 from K. W. Amish, RG&E to J. F. O' Leary, NRC. 5. Ginna Emergency Procedures El.1 P.nd El.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] 1. " 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.

2. Letter dated May 24, 1974 from K. W. Amish, RG&E, to J. F. O' Leary, Director, Directorate of Licensing, USNRC. 3. Letter dated September 4, 1974 for R. R. Koprowski, RG&E to Edson Case, Acting Director, Directorate of Licensing, USNRC. 4. Letter dated November 1, 1974 from K. W. Amish, RG&E, to Edsca Case, Acting Director, Directorate of Li-cen.cing, USNRC. 5. Letter dated May 20, 1977 from L. D. White, Jr., RG&E, to A. Schwencer, Chief Operating Reactors Branch #1, USNRC. 6. Letter dated February 6, 1978 from L. D. White, Jr., RG&E, to A. Schwencer, Chief, Operating Reactors Branch

  1. 1, USNRC.

7. Amendment No. 7 to Provisional Operating License DPR-18, transmitted, by letter dated May 14, 1975 from Robert A. Purple, Chief, Operating Reactors Branch #1, USNRC, to i L. D. White, Jr., RG&E. 8. Amendment No. 29 to Provisional Operating License DPR-18, transmitted by letter dated August 24, 1979 from Dennis L. Ziemann, Chief, ORB #2, to L. D. White, Jr., RG&E. 9.

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

10.

Letter, L. D. White, Jr., RG&E, to D.

L. Ziemann, USNRC, June 27, 1979. 11.

Letter, L. D. White, Jr.,

RG&E, to D. L. Ziemann, USNRC July 6, 1979. Effects of Flooding [ Flood] 1. 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 #1, Division of Reactor Licensing. 3. Letter dated May 30, 1975 from L. D. White, Jr., RG&E, to Robert A. Purple. 4. 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 #1, USNRC. TMI Lessons Learned [TMI] 1. RG&E letters of October 17, November 19, and December 28,

1979, L. D. White, Tr., RG&E, to D. L. Ziemann, USNRC, "TMI Short Term Lessons Learned Requirements."

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