LD-91-014, Forwards Response to NRC Request for Addl Info Re Design Certification,CESSAR-DC,per
| ML20070R325 | |
| Person / Time | |
|---|---|
| Site: | 05000470 |
| Issue date: | 03/26/1991 |
| From: | Erin Kennedy ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| PROJECT-675A LD-91-014, LD-91-14, NUDOCS 9104010154 | |
| Download: ML20070R325 (98) | |
Text
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a LJ ABB ASEA BROWN BOVE Al March 26, 1991 LD-91-014 Project-No. 675 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
Subject:
Response to NRC Requests for Additional Information
Reference:
NRC Letter, Plant Systems Branch RAIs, R. Singh (NRC) to A. E. Scherer (C-E), dated January 24, 1990.
Dear Sirs:
The reference letter requested additional information for the NRC staff review of the Combustion Engineering Standard Safety Analysis-Report -1 Design Certification (CESSAR-DC).
Enclosure I to this letter provides our responses and Enclosure II provides.the corresponding revisions to CESSAR-DC. Responses to questions 410.67 and 480.8 will be provided separately.
Should you have any questions on the. enclosed material,_
please contact me or Mr. S. E. Ritterbusch of my staff at
-(203) 285-5206.
Very truly yours, COMBUSTION ENGINEERING, INC.
~
E. H. ,nne y Manage <
Nuclear Systems Licensing
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Enclosures:
- As Stated cc: P.-Lang (DOE - Germantown)
J. Trotter (EPRI)
T. Wambach (NRC)
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M Enclo2uro I to LD-91-014 RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION, PLANT SYSTEMS BRANCH I
I Onu tion 280 1 Provide the fire protection analysis and/or interface requirements to ensure that safe shutdown can be achieved, assuming that all equipment in any one fire area will be rendered inoperable by fire and that re-entry into the fire area for repairs and operator actions is not possible with exception of the control room. For the control room, provide the fire protection analyses and/or interface requirements having an independent alternative shutdown capability that is physically and electrically independent of the control room. Also, provide the fire protection requirements for redundant shutdown systems in the reactor containment building that will ensure, as much as practicable, that one shutdown division will be free of fire damage. Additionally, also ensure that wLe, Dot gases, or the fire suppressant will not migrate into other tire area; to the extent that they could adversely affect safe shutdown capabilitier, including operator actions.
Bnponse 280.1 A fire protection analysis of each fire area is conducted as part of the Fire Hazards Analysis. The System 80+ design basis, as stated in CESSAR-DC Section 9.5.1 (as revised in Amendment 1), is to assure the ability to achieve Safe Shutdown following fire in any fire area outside of containment. This includes loss of all equipment in any given area and effects of electrit 1 interaction which may disable equipment outside of the immediate area. The plant is at ranged so that Safe Cold Shutdown can be achieved following fire in any area outside of cor,tainment without need for repairs or extraordinary operator action. Emergency shutdown from outside the control room is described in CESSAR-DC Sections 7.4.1.1.10 and 7.4.2.5. Outside of containment, redundant divisions of safety related equipment are separated by three hour fire rated boundaries, in the control complex (and most locations in the Nuclear Annex) redundant safety related divisions of protective electrical channels are separated by three hour fire rated barriers so that loss of all equipment in these areas would not affect either division of safety related equipment required to achieve cold shutdown, inside containment Engineering Analysis conducted as part of the Fire Hazard Analysis assure that fire at any location which can disable more than one channel of cold shutdown ear 'ent will not affect the ability to achieve cold shutdown using equit which would not be affected by fire at that location.
Smoke conti s recognized as an important element of the Plant fire Protection c in features, in the subsphere area, containment, Fuel Pool building, Reactor Annex and Diesel Generator building the HVAC System has smoke control capability by allowing any area to be purged with 100% outside air. In the control complex, dedicated smoke exhaust fans are provided for the control room and TSC. In addition, a smoke exhaust system is provided for each channel of safety related equipment.
A connection to the normal HVAC system intake is used for fresh air supply. Smoke detectors are installed in return air ducts to alarm and annunciate in the control room. Smoke dampers are arranged for remote operation from the control room.
The System 80+ design does not have connections (door or ventilation openings) between redundant safety-related divisions. This further mitigates the possibility that smoke and products of combustion of fire suppression agents will affect redundant safety-related equipment.
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L 10uestion-410.63 ,
You have stated-in Section 9.1,1,1, Design Bases, that storage-is-provided for 166 new fuel assemblies. In Section 9.1.1.3.! .3, one of the stated methods of maintaining criticality safety margins is by " limiting the caoacity to 166 fuel assemblies." However, in Section 9.1.1.3.4, you .
have stated that " storage is provided;for at least-166 new fuel assemblies..." Clarify these inconsistencies. Also, provide maximum rated capacity of the new fuel storage racks'. Indicate how that-limit-will be . implemented _ (e.g., administrative controls, plugging of vacant cells).
Response'410.63 LTheidesign of new fuel' racks have been_ changed to provide _ storage for 121 fue11 assemblies in a-50% density array. Two 11 x 11 fuel racks are utilized with every other cell blocked by a cell -blocker.
The' cell blockers-restrict the' placement of a fuel assembly into the restricted cells during normal fuel handling operations. The racks have been analyzed based on a postulated accident condition of a fuel assembly located-in a restricted cavity and the results showithat the-rack k,,, is-less than. 98.
- Section 9.1.1.2:has been rewritten (in Amendmentfl) as noted below.
The new fuel racks are made up of two -(2) 11-x 11 individual rack modules (see Figures 9.1-la &-9.1-21), each module co..caining 121 storage' ceils..
A module is an array of fuel storage cells similar- to that shown in-LFigure 9.1-1. The storage racks are-stainless-steel honeycomb-structures
.with rectangular fuel storage cells. - A single. pitch of 9.78 inches is provided-for all1the racks.- Cell blockers are installed in the alternate cells:to limit new fuel storage'to 121 fuel assemblies.
The stainless steel construction of the stcrage racks is compatible with fuel assembly materials and the fuel storage ~ environment.
The clearance between the fuel rack module and-the walls of the storage
- cavity-is less; than the width of a . fuel assembly-to preclude the Linadvertent placement of a -fuel assembly outside of the rack module. The racks are bolted to embedmonts at the bottom.of the rack storage cavity
'to preclude tipping..
Sections 9.1.-l.3.l'.3 and 9.1.1.3.4.have-been modified to reflect the capacity limit-of 121 fuel assemblies.
e lh Qyestion 410.64 You have identified the different storage densities for regions I and 11 of the spent fuel pool (50% and 75%, respectively) in your submittal.
Provide pertinent information concerning the design criteria and anticipated controls to be implemented for the. storage of spent fool asse:/; lies in the above regions.
Resoonse 410.64 Both Region I and II storage areas are designed to accommodate fuel essemblies with initial enrichment up to 5 weight percent U-235. Region I has no restriction on burnup history of stored fuel assemblies. Region II is restricted for storage of fuel having a minimum cumulative burnup which is dependent on the initial enrichment for each fuel assembly. The burnup versus enrichment curve is internally documented. This restriction on fuel storage in Region II will be imposed by administrative controls developed and implemented by the Owner-0perator.
The following will be added in Section 9.1.2.2.2: "A fuel assembly may be stored in Region II only if it has the minimum burnup required for an assembly of its initial enrichment. The Owner-Operator will develop and implement administrative controls to permit storing a fuel assembly in Region II only if it meets established burnup versus initial enrichment requirements."
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- o. ,3 Onestion 410.61 You have stated in Section 9.1.2.3.1.3 that one of the accidents considered _in the design of the spent fuel pool storage racks is a fuel assembly and its handling tool " falling into a blocked-off fuel storage cavity." Supply additional information concerning the mechanical blocking assemblies to allow determination of the extent of penetration of a fuel assembly into a blocked cavity.
Response 410.65 The spent fuel racks provide storage for 363 fuel assemblies in Region I (50% density) and 544 fuel assemblies in Region II (75% density). The restricted rack cells contain cell blockers which prevent the placement of a fuel assembly into the restricted cells. The racks have been analyzed based on a postulated accident condition of a fuel assembly fully inserted into a restricted cell. Taking pool boron concentrations into consideration, the results show that the rack k,,, is less than .95.
The cell blockers cannot be inadvertently removed once installed as special tooling is required to unlock and remove them from the spent fuel storage racks.
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s .o Question 410.66 Your submittal does not provide information concerning the handling of heavy loads in the vicinity of the spent fuel pool. Provide an evaluation of the capability of the spent fuel loading pit to withstand a dropped heavy load. The evaluation should include a shipping cask drop without breach of the pit area or loss of spent fuel pool water.
Response 410.66 The spent fuel cask laydown area is separated from the spent fuel pool by a gate and a structurally reinforced concrete wall. The gate is closed, sealed, and locked during all cask handling operations. The floor in the laydown area has been designed to withstand the impact of a shipping cask dropped from a height of 30 feet without breaching the integrity-of the floor plate.
Any small water loss as a result of local damage to the laydowr area wall liner cannot be communicated to the spent fuel pool due to the closed gate and the integrity of the independent spent fuel pool liner. Damage to the gate is prevented during cask handling by stops on the bridge crane rail that limit cask travel and by the recessed gate design.
Design features to address the spent fuel cask drop accident are summarized in CESSAR-DC, Section 15.7.5, Amendment H.
- O Ouestion 410.68 You stated in your submittal under Section 9.1.3.3.1 that the SfPCS has no emergency function during an accident. General Design Certification (GDC) 44, however, requires that the system be able to perform its intended safety function under accident conditions Verify that the system will be capable of continued operation during all accident conditions.
Response 410.68 The Pool Cooling and Purification System (PCPS) is designed in accordance with General Design Criterion 44. Section 9.1.3.1.4 states that the system safety function, which is pool cooling, can be accomplished assuming a single active failure, during both normal and plant accident '
conditions. Redundant, independent cooling trains and associated components are provided to assure that if one cooling train is unavailable, the second train provides backup capability and continued spent fuel pool cooling.
The statement made in Section 9.1.3.3.1 has been corrected in Amendment I.
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'Ouestion'410.69 You have stated in your submittal under Section 9.1.4.2.1.2 that an-interlock prevents fuel carrier movement unless the transfer tube valve isLfully open. Verify that;an additional. interlock-is provided to prevent the transfer tube valve movement'while the fuel carrier is passing through the valve or verify that transfer tube valve movement
.will not cause.any fuel damage.-
Response 410.69 The transfer system winch overload interlock (9.1.4.'2.1.0. A) will terminate movement of the fuel-carriage through the transfer tube.in the 1 event the fuel carriage contacts a partially closed-transfer tube'. valve. L Testing has shown that there is no damage to the fuel carrier or the fuel-assembly: under this condition. Administrative controls prepared and implemented by the Owner-0perator will restrict operation of the fuel transfer tube valve,during fuel handling operations.
The'following will becadded in Section 9.1.4.2.1.2:- "The Owner-Operator-will prepare and~ implement administrative controls to restrict operation' of the fuel. transfer Lube valve during fuel handling operations.".
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.- o Qyestion 410.70 You have stated in yout submittal uider Section 9.1.4.2.1.3 that an
-operator is a backup t=' . lectrical zone interlocks. Provide information on translation speed, i.nerlock set point margin including obstruction or restricted area, free cabi length, etc., to verify that the operator action is a viable option concerning the identification of an interlock failure and follow-up with proper response.
Response 410.70 Paragraph 9.1.4.2.1.3.8 has been revised (Amendment 1) to read as follows:
"If these interlocks fail, the spent fuel handling machine mast will protect the fuel assembly from damage in the event of wall or gate contact."
Ouestion 410.71 Section 9.1.4.4, Testing and Inspection Requirements, discusses "preoperational checks." Verify that these checks include equipment testing before each use of fuel handling machines, overhead crane and polar crane. Also, verify that these checks include load testing and other testing designed to detect degradation due to wear or normal use for the above equipment.
Response 410.71
- The fuel handling -equipment plant operating procedures typically specify detailed preoperational checkouts that must be performed prior to equipment use to insure that the equipment is in proper working order.
These checkouts include the following: interlocks, brakes, hoisting cable, control circuitry, lubrication, and load testing. The Owner-Operator will prepare cod implement operatirig proceduies to accouplish '
these preoperational checkouts.
The following will be added to the end of Section 9.1.4.6: " Operating procedures prepared by the Owner-0perator will also require peroperational load testing and checkouts of interlocks, brakes, hoisting cables, control circuitry and lubrication of fuel handling equipment."
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- . 'a 10uestion 410.72 Provide the fuel building layout drawings wh_ich show the (1) overhead heavy load paths _ and (2) safety-related equipment locations in the vicinity of those paths susceptible to damage by failure of electrical
-inter 1ocks,-swinging of the 1oad, or:other mechanisms for causing damage.
Resoonse 410.72 The containment polar crane, the cask handling crane, and the fuel
-handling crane are designed to prevent the drop of a heavy load such as, the reactor yessel head and the spent fuel shipping cask. -In-addition,.
predetermined load paths for major lifts (see Figures 9.1-19 and 9.1-20),
operator training,'and regular crane maintenance minimize the possibility Lof load mishandling.
i.imit switches, electrical interlocks ad raachanical iriterldcks pruar,t- !
. improper crane operation which might result in a fuel handling accident.
-This is-also'discusced in Section 9.1.4.2.1.7. The spent fuel cask handling-hoist'is restricted from movement over the new and spent fuel ;
storage areas when the-fuel racks contain fuel assemblies. The new fuel
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handling hoist is restricted- from movement over the spent fuel storage area when-the--spent fuel racks contain4 fuel - a.isembl ies.
'In accordance with the regulatory position of Regulatory Guide 1.13 and General Design criteria 61 of Appendix A to 10 CFR 50, the hoists are also restricted from passing over. the spent fuel pool cooling system or ESF-~ systems which=could be damaged by dropping the load.
Set points for-the hoist' interlocks are set to preclude falling or tipping of the ' loads into the fuel- storage areas.
. ' Typically,1 administrative. controls prepared by'the-Owner-0perator
' preclude movement of heavy loads within:the containment building pool when the1 refueling machine contains a fuel assembly. During heavy load i novement, the fuel transfer tube valve is closed to avoid water level-changes in the fuel building during postulated accident conditions such
.as' dropping the heavy load on the reactor vessel pool seal.
THe'first sentence of the last paragraph of Section 9.1.4.3.1 has been
' modified to state: " Administrative controls prepared by the Owner-Operator..."
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e . .o question 410.73 Provide containment layout drawings showing the reactor vessel head storage location, the upper guide structure storage stand, the load paths from the reactor to those locations, and safety-related equipment in the vicinity of the load paths susceptible to daw.ge by load handling accidents.
Reinonse 410.73 Figure 9.1-19 depicts the load paths of the reactor vessel closure head, the core support barrel (CSB), and the upper guide structure (VGS) from the reactor vessel to their respective storage areas during the refueling outage. The designated load path for each component passes over the reactor vessel flange. An analysis has shown that in the event the reactor vessel head or the internals are dropped on the reactor vessel, the core will be maintained in a coolable condition.
Typically, operating procedures prepared by the Owner-0perator control the lift height of the VGS to minimize its clearance with the pool floor and the polar crane is positioned to insure direct travel from the reactor vesici to the UGS storage area. Additionally, the ICI holding frame is installed over the seal table at the operating floor level during fuel handling. This prevents the UGS from moving over the seal table. Therefore, operating procedures preclude the UGS from being lifted above the seal- table or being closer than approximately five feet to the seal table, thereby making seal table damage a remote possibility.
However, under a postulated = load drop on the seal table, the seal table would fail resulting in containment pool draindown to the reactor vessel flange area. Since ICI tubes are only restrained laterally and not vertically, the tubes would be bent down to the level of the reactor vessel flange. Any tube failure, therefore, would in all likelihood be at or near the reactor vessel flange level which would result in a-water-level within the reactor vessel similar to that prior to reactor vessel head removal. The accident condition would not be any more severe than that analyzed for the reactor vessel head drop on the reactor vessel fl ange .
There are no other unprotected safety-related components within the load paths of the reactor vessel head, VGS and CSB. An unprotected component is defined as a component that is not protected by the pool walls and/or operating floor.
The transfer tube valve will be closed during these handling evolutions to preclude water level changes in the fuel building.
The following sentence will be added_ to the end of the first paragraph of Section 9.1.4.3.1: "The Owner-0perator's' operating procedures will control the load paths and height of the reactor vessel closure head, the core support barrel and the upper guide structure above the pool floor.
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Ouestion 410.Zi Provide the following for the water systems described under Sections
-9.2.1 through 9.2.10:
(a) Tabulated equipment design parameters at least including anticipated normal and accident heat loads, system design flow rates, heat removal capacities, and tank capacities. .
(b) System piping and instrument diagrams'(P& ids) and system
- descriptions that contain sufficient information to determine the
-following:
(1)- Interfaces'between safety and non-safety. portions of systems (including changes in component safety classifications).-
(2) Containment penetrations and isolation capabilities.
(3)- Complete system flowpaths.
(4) Interfaces with other systems and system boundaries.
(5) -Isolation- capabilities .between essential and non-essential portions of the systems, and-
-(6) -Physical ~ division-between redundant portions of the systems, and (c) 'FMEA for the essential portions of water systems to verify that Lthese portions can withstand design basis-accidents concurrent with-a-single active failure.
Resoonse 410.74-
" P& ids and System Descriptio_ns have been completed for, CESSAR-DC Section
'9.2.1, Station Service Water System-(SSWS), and Section 9.2.2, Component-Cooling Water System (CCWS). - A conceptual description of the Ultimate l Heat. Sink-(UHS) is:provided in CESSAR-DC, Section 9.2.5. Details shall be provided.in.the site-specific SAR. The'other water systems described in'CESSAR-DC Sections 9.2.3, 9.2.4, 9.2.6, 9.2.7, 0.2.8, 9.2.9, and 79 2.10 are non-safety related. Flow diagrams are provided in the Lapplicable sections. Please see' Amendment ~ I of CESSAR-DC for the above revisions.
(a) CESSAR-DC Section 9.2.l. Station Service Water System Table 9.2.1-2 contains the system and component design parameters.
The design flow rate of the-SSW pumps are specified when piant-nspecific components'(UHS and CCW heat exchangers) are procured, The heat load on the SSWS for normal operation, shutdown cooling (initial and final), end for a design basis accident are given as the total heat load per division in Table 9.2.2-3.
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.CESSAR-DC Section 9.2.2. Component C.nglina Water Syst M Table 9.2.2-4 contains the system and component design parameters.
. Table 9.2.2-3 contains the heat loads and flow rates for each component and division.
(b) CESSAR-DC Section 9.2.1. Station Service Water System (1) The interfaces between safety and non-safety related portions of the SSWS are illustrated as piping class changes in Figure 9.2.1-1.
(2) The SSWS does not contain any containment penetrations.
Isolation valves are discussed in CESSAR-DC Section 9.2.1.2.1.8 and are illustrated in Figure 9.2.1-1.
(3) The system flowpaths are illustrated in Figure 9.2.1-1. The component vent and drain lines will be added after the equipment has been purchased.
(4) The SSWS interfaces with the CCWS, the VHS. and the Emergency Class lE Auxiliary Power System. This is discussed in the system description, CESSAR-DC Section 9.2.1.2 and is illustrated in Figure 9.2.1-1.
(5) The isolation capabilities between the essential and non-essential portions of the SSWS are illustrated in Figure 9.2.1-1.
(6) The two divisions are physically separated as discussed in CESSAR-DC Sections 9.2.1.1, 9.2.1.2, and 9.2.1.3.
CESSAR-DC Section 9.2.2. Component Coolina Water System (1) The interfaces between safety _ and non-safety related portions of the CCWS are illustrated as piping class changes in Figure 9.2.2-1.
(2) The CCWS's containment penetrations are illustrated in Figure 9.2.2-1 and discussed in CESSAR-DC Section 9.2.2.2.1.9.12. Component and header isolation valves are discussed in CESSAR-DC Section 9.2.2.2.1.9 and are illustrated in Figure 9.2.2-1.
(.y The system flowpaths are illustrated in Figure 9.2.2-1. The component vent and drain lines will be added after the equipment has been purchased.
(4) The CCWS interfaces with various systems through the heat exchangers listed in CESSAR-DC Section 9.2.2.2.2, the SSWS, and the Emergency Class lE Auxiliary Power System. The system interfaces are illustrated in Figure 9.2.2-1.
(5) The isolation capabilities between the essential and non-essential portions of the CCWS are illustrated in
q 6, .4 Figure 9.2.2-1 and discussed in CESSAR-DC Sections 9.2.2.2, 9.2.2.2.1.9, and 9.2.2.5.8.
(6) The two' divisions are physically separated as discussed in CESSAR-DC Sections 9.2.2.1, 9.2.2.2, and 9.2.2.3.
(c) .CESSAR-DC Section 9.2.1. Station Service Water Ey11gm Table 9.2.1-1 contains the Single Failure Analysis for the SSWS.
CESSAR-DC Section 9.2.2. Component Coolina Water System Table 9.2.2-2 contains the Sin 9 1e Failure Analysis for the CCWS, i
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'Ouestion 410.75 Provide the following for the station service water system (SSWS):
_(a) An analysis to demonstrate that the SSWS pumps will be protected from abnormally high levels-of the ultimate heat sink due -to.
flooding, t
'(b) Division cross-connect information including. valve positions and actuations,
-(c):- An: analysis concerning the effects'of high and moderate energy-line breaks on safety functions of the system,
- (d)- Provisions to preclude system failure due to water hammer events.
(e) Provisions: to prevent potential radioactive leakage from the component cooling water system to the SSWS, and
-(f) System evaluation concerning the SSWS dependency on compressed air to perform its safety function.
I Response 410.75 Detailed Piping and Instrumentation Diagrams (P& ids) and-the System Description have been completed for the Station Service Water System (see Section 9.2.1, Amendment I).
(a) . The location of the SSWS pumps-and design of the SSWS pump structure are site-specific. Therefore, flooding analysis to demonstrate that the SSWS aumps .are protected against- abnormally high levels of the ultimate 1 eat sink-is~a site-specific analysis. CESSAR-DC Sections.
9.2.1.l.l.D, 9.2.1.1.1.1, 9.2.1.3.D, and 9.2.1.3.1 provide-interface
- requirements by stating that SSWS pumps are located and protected from adverse environmental occurrences such as flooding.
(b) No' cross-connections exist between the two divisions of the SSWS.
This is stated in CESSAR-DC Section 9'2.1.2.1.3 and illustrateo in Figure 9.2.1-1.
(c) L : CESSAR-Dc Sections 9.2.1.1.1.C, 9.2.1.3.C, 9.2.1.1.1. I, and 9.2.1.3.1 address this concern and state that two separate 100%
redundant' systems-are provided and protected to prevent a high or moderate energy line break from affecting the- SSWS safety. func' ion.
When components-are procured and construction drawings'are prepared, a review will confirm compliance with these design base;.
.(d) The SSWS is designed-to prevent damage to components and piping due to water hammer by properly placing high point venting and by assuring proper filling of.the system. Valve open and closure times are set to preclude 'this problem. This issue is addressed in CESSAR-DC Sections 9.2.1.1.I'.J, 9.2.1.2.1.3 and 9.2.1.3.J.
(e) A radiation monitor is provided in each SSWS division downstream of the component cooling water heat exchangers. If radiation is detected at a preset level above the background radiation an alarm
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$ a is sounded in the control room. This is discussed in CESSAR-0C Section 9.2.1.5.5. Each division of component cooling water is also provided with a radiation monitor which alarms in the control room.
This is discussed in CESSAR-DC Sections 9.2.2,2, 9.2.2.2.1.7 and 9.2.2.5.5. If high radiation is detected the leaking component can be isolated to prevent further leakage in the CCWS. In addition the leaking CCWS heat exchanger can be isolated and the second CCWS heat exchanger aligned.
(f) The SSWS does not utilize the compressed air system. All safety-related valves are manually or motor cperated (CESSAR-DC Table 9.2.1-3 and Section 9.2.1.2.1.8),
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J Ouestion 410.76 Provide the following for the component cooling water system (CCWS):
(a) Information to determine that the CCWS surge. tank is sized for the.
maximum expected leakage from the system for seven days, or that Seismic Category I makeup can be aligned within the time allowed by surge tank capacity.
(b) Information concerning the effects of high and sedium energy line breaks on safety function (s) of the system.
(c) Provisions to prec6de system failure due to water hammer events, -
(d) Results of tests demonstrating the ability of the reactor coolant pumps (RCPs) to operate for 20 minutes without seal cooling or provide provisions for the RCP seal and buring cooling following the pcstulated accidents. (Section 9.2.2.2 ir.dicates that RCP cooling functions are nonsafety-related. Figure 9.2.2-1 shows these
-functions isolated on a safeguard signal), and (e) A system evaluation concerning the CCWSs dependency on compressed air to perform its safety fum: tion.
Response 410.76 Detailed Piping and Instrumentation Diagrams and the System Description have been completed for the Component Cooling Water System (see Section 9.2.2, Amendment I).
(a) A Seismic Category I makeup supply line from the SSWS is provided.
A removable spool piece is placed in this line to assure tnat inadvertent addition of station service water is prevented.- The surge tank is adequately sized to accommodate normal fluid losses to allow the operator time to incert the spool piece (CESSAR-0C Sections 9.2.2.2 and 9.2.2.2.1.3).
(b) Tables 3.6-3 and 3.6-4 of Amendment I identify system -locations, piping drawings, temperatures, and pressures for high energy lines inside and outside containment. Consistent with the approach in Appendix C of Branch Technical Position SPL8 3-1 (SRP Sectioa 3.6.1), emphasis is placed on-location of piping and physical separation to minimize the effects of high energy line breaks. When specific plant components are procured, the need for special features to protect that equipment is evaluated and, if necessary, protective measures are taken.
l With respect to the CCWS, CESSAR-DC Sections 9.2.2.1.1.C, 9.2.2.1.1.J 9.2.2.3.C, 9.2.2.3.J and 9.2.2.2 address this concern and state that the CCWS safety-related components are designed and protected such that this type of failure would not affect the safety performance of the CCWS. Also the CCWS consists of two 100%
redundant divisions. When ccmponents are procured and construction drawings prepared, a review will confirm compliance with these design bases.
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(c) The CCWS is oesigned to prevent damege to components and piping due to water hammer by properly placing high point venting and by assuring proper filling of the system. Valve open and closure times are selected to miaimize water han.mer eff0 cts. This assue is
' addressed in CESSAR-0C Sections 9.2.2.1.K, 9.2.2.2.1.4, and 9.2.2.3.X.
(d) Section 5.4.13 of CESSAR-DC states that tests wili be performed to -
verify that the RCP can operate 30 minutes uithout seal- cooling. -
This time limit is established in Topical Report CEtJPD-201-A. 'The containment isolation valves to the RCPs lA and 10 header and RCOs 2A and 2B header do not isolate cn a safeguard signal, they isolate i on their respective low surge tank level or by operator action if i leak is noted. Yhese isolation valves can also be opened from the control room. See CESSAR-DC Sections 9.2.2.2.1.9.12 and 9.2.2.2.2.5.
(e) The CCWS is designed such that all valves dependent on compressed air (instrument air) are provided with safety-grade operators and solenoid valves. The soknoid valve is vented such that loss of instrument air would result in a fail-stfe position. Travel stops are provided on the safety related pneumatically operated control valves to limit the maximum flow. This issue is discussed in Section 9.2.2.2.1.9 and Table 9.2.2-5.
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1 Lan four sty > ittal, you have stated that (1) the essential chilled water
- Syt tes (l.C0,i,$ receives mall-up from the condensate storage tank (CST) via a-SelamictCat, gory I line following a loss of offsite power and (2) the CST'iswa! sathty-related and is not designed as Seismic Category 1.
Provide your Justifications concerning the reliance on a nonsafety- 4 related CST to render the safety-related ECWS operable.
Baroonse 41Q421 The third sentence of Section 9.2.9.2.1 (Amendment 1) now reads as fo'. l ows : "In case of a loss of demineralized water, make-up is supplied frit the ttation service water system, via a Seismic Category I assured water line. A removable spool piece is placed in this line to prevent intrusion of raw w;ter into the clean, chemically treated syster; during (l normal operation." _
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QMt111on 410 18 (Section 9.2.7)
Identify the portions of the refueling water system that provide isolation from the safety-telated and non+ safety-related systems used to
-fill and drain the refueling pool (e.g., containment spray, shutdown cooling, chemical and volume cor. trol systems). Also, identify the provisions made to ins'Jre that failure of the isolati':n devices (either mechanical or through human error) will not adversely affect the operability of the safety-related systems.
Egiponse 41Q22B The Piping and Instrument Diagram provided in Amendment I for the PCPS (Figure 9.1-3) shows all interfacing isolation points between safety and non-safety related systems used to fill and drain the refueling pool.
Safety classifications of piping and components are also shown on figure 9.1-3, and on all referenced interf acing drawings. The design basis of the PCPS, which includes the refueling water system, meets all requirements-which ensure that non-safety related systems do not P jeopardize the operation of any safety related systems.
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O b Question 410.79 Provide an evalt.9 son concerning the conformance of the turb'ae bailding component cooling and service water systems with the guidar.:o of N sition C.2 of Regulatory Guide (RG) 1,29, llgip.pDse41h22 The turbine building component cooling water system is contained entirely in the turbine building and is only utilized for cooling nonsafety-related secondary-side components. The turbine building service water system only cools the turbine building component cooling water heat exchangers, is separated from the station st:rvice water system, and utilizes the power generation cooling towers as the heat sink.
Therefore, the turbine building component cooling water and service water systems do not interface or interact with any safety-related systems and do not require any Seismic Category I design.
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Question 410.80 Provide the fellowing for the chilled water systems:
(a) You have stated in Sectio 9 9.2.9.2 that: "Each division is totally.
independent... except for areas where it is physically impractical or unsafe." Provide a single f ailure analysis for the chilled water systems and assure that the lack of separation will not result.in %
potentikl single failures, active or passive, that could render the ECWS inoperable, (b) Provide ECWS design heat loads for normal and accident conditions.
Also provide design details including flow rates, heat removal rates, etc., for the ECWS and normal chilled water system, and (c) You have stated in your submittal that the ECWS serves primarily safety-related HVAC cooling loads. Identify the nonsafety loads served by ECWS. Describe the measures provided to ensure that these nonsafety loads do not impact the ability of the ECCW to perform its safety-related functions.
Response 410.8Q (a) The third sentence of Section 9.2.9.2 (Amendment 1) now reads as follows: "Each division is totally independent and separated both mechanically and electrically; therefore, sitgle failure or lack of separation cannot render ECW3 inoperable."
(b) Heat loads and flow recuirements for area coolers supplied by ECWS and NCWS for normal anc accident conditions are now provided in Table 9.4-2.
(c) The word "primarily" has been deleted from the second sentence of Section 9.2.9. The ECWS only serves safety-related HVAC 2aads.
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Question 410.81 l
< Provide the following for the compressed air systems:
(a)' Provide the P&lDs for the instrument and compressed air system .
showing portions classified as Quality Group C and Seismic Category )
I, (b) Provide a list of safety-related as well as nonsafety-related )
equipment serviced by the instrument air system. Identify equipment essential to safe shutdown or accident mitigation, include system air capacity versus equipment consumption data and provide an evaluation of the affects of loss of instrument air on this equipment.
(c) You have stated in Section 9.3.1.2.1 that in the event of low instrument air pressure, the station air system will automatically supply air to the instrument air system. Describe the connection between the two systems and the means for pro'.ecting the instrument air system from a failure in the station air system, and (d) In Section 9.3.1.3 of your submittal, you have stated that f ailure of the instrument air system during an accidelt er station blackout would cause all-pneumatically operated valves essential to safe shutdown to fail in the safe position. This implins that all pneumatic valves that actuate to mitigate ar, accident would actuate on loss of instrument air. Provide an analpis that shows that inadvertent actuation of safety-related valves due to failure of the instrument air system will not cause any unsafe conditions that preclude achieving and maintaining safe shutdown, Rea r nse 410 E (a) figure 9.3.1-1 " Instrument Air System," is incleded in Amendment 1.
, (b) Table 9.3.1-1, " Active Safety-Related Components Serv' ced by Instrument Air," is included in Amendment 1. Table 4.3.1-1 lists all active safety-related equi 3 ment with an instruwnt air supply which is essential to safety sautdown and accident mitigation. The list denotes each component's safety function as well as its loss-of-air, failed position and fail-safe podtion. A list of nonsafety-related-equipment serviced by the instrum t air system is
, not included in this response as a result of the issign described in l item (c) below.
Equipment air capacities will be based on expected consumption for both safety-related and nonsafety-related components, when procured.
An evaluation of the effects of loss of instrument air on systems essential to safe shut 6 % n or accident mitigation is provided in item (d) below.
l (c) The air supply systems have been arranged and grouped into separate
' and isolated syatems. To assure separation of the air systems, the systems have no interconnections with the other systems. Therefore, the instrument air system is independent of the station air system.
. i (d) All active safety-related valves having an instrument air supply are listed in Table 9.3.1-1. These valves are designed to fail in the safe position on a loss of instrument air to the valve actuator.
The fail safe position of these valves takes into account any inadvertent actuations which may occur from loss of instrument air to insure the ability to achieve and maintain safety shutdown. This is evident from a review of the safety function of the valves listed in Table 9.3.1-1 and the systems containing these valves. '
As listed in Table 9.3.1-1, the majority of the pneumatically operated valves with an active safety function are containment isolation valves. By design, the containment isolation valves are required to close following a design basis event. These valves were previously reviewed for determination of their role in achieving safe shutdown and accident mitigation. As a result of this review, the valves were determined to be nonessential for acMeving safe shutdown and are designed to fail in the closed positit,.'. Therefore, inadvertent closure'of the pneumatically operated contatiment isolation valves due to'a loss of instrument air would have no impact on the plant's-ability to achieve and maintain safa shutdown.
Inadverteht closure of the MSIVs or MFIVs from a loss of instrument
- air would cause a reactor trip. The emergency feedwater system and atmospheric dump-system would be available to bring the plant down to a shutdown cooling entry condition. Consequently, no unsafe conditions would result from inadvertent closure of these valves which would preclude achieving and maintaining safe shutdown.
The emergency feedwater system contains pneumatically operated valves with an active safety-related function. The steam supply isolation valves and steam supply bypass isolation valves permit steam flow to the emergency feedwater turbines. A loss of instrument-air will cause these normally closed valves to fail safe in the open position allowing steam flow to the EfW turbines.
Inadvertent opening of the. valves due to a loss of instrument air will start the turbine driven pumps and thus does not render the EfW system inoperable. With the-valves in a failed open position, initiation of the tTW system is still possible. Therefore, a loss of instrument air to these valves will not result in an unsafe condition or impede safe shutdown.
The active, pneumatically operated, component cooling water system valves can be divided into two groups to analyze the effects of inadvertent actuation. One group would include those valves which serve to isolate nonessential portions of the CCW system. As nonessential isolation valves, they isolate portions of the CCWS system which are not required for safe shutdown. Therefore, these valves are deemed nonessential for safe shutdown and are designed to fail closed. The second group of valves can be categorized as flow control valves. As such, they are required to= fail safe to the open position insuring system operability for safe shutdown and accident mitigation. Excessive flow is prevented by the use of travel stops on the valves should they fail open. Thus, it is apparent a loss of instrument air causing these valves to unintentionally fail to their safe position will not affect achieving or maintaining safe shutdown.
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p i The pressurizer spray control valves on the reactor coolant system are not required for safe shutdown. The safety depressurization system provides a safety grade means for depressurization to achieve safe shutdown.
The letdown isolation valve and letdown backup isolation valve on the chemical and volume control system are system isolation valves and are not essential to achieving safe shutdown. These valves fail closed on loss of air which is their safe position since this l prevents further letdown should there be a break in the letdown I line.
As previously stated in Section 9.3.1.3, the failure of the compressed air system will not render any safety system .quipment or its function inoperable. In addition, from the above ev61uation, it is evident an inadvertent actuation of the pneumatically operated l active safety-related valves will not create an unsafe condition which would preclude achieving or maintaining safe shutdown. i l
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l Ouestion 410.82 Provide the following for the equipment and floor drain system:
(a) Provide an analysis of postulated tank failures. This analysis should include the failure of the drain and waste tanks flooding safety-related equipment and/or release of radioactive contents to the environment. Also, these tanks should be classified Seismic Category I and Quality Group C.
(b) Provide an evaluation concerning the ability of the system to withstand active component failures, blockages, and probable maximum flood without inundating safety-related areas, and (c) Verify that contaminated liquid waste cannot be inadvertently routed to the noncontaminated drainage systems.
Response 410.82 (a) As summarized below, System 80+ has adequate design features to control flooding due to postulated tank failures. The System 80+
design also contains provisions to prevent flooding of safety-related equipment and control the release of radioactive materials to the environment.
An analysis of postulated radioactive releases due to liquid-containing tank failures was performed in Section 15.7.3 of CUSAR- l DC. This analysis evaluates the failure of the Boric Acid Storage Tank (BAST) which is the most limiting radioactive tank failure for the System 80+ design. The BAST is located in the yard area of the plant and is surrounded by a seismically designed dike, constructed in accordance with Regulatory Guide 1.143. The Chapter 15 analysis establishes the maximum allowable dilution factor for the radioactive liquid released from the failed BAST prior to reaching the potable water supply. The results of the analysis should be used to establish site acceptance criteria for the minimum dilution flow required to limit the concentration at the nearest potable water supply to less than 10 CFR Part 20, Appendix B, Table 11, Column 2 limits.
Of the drain and waste tanks located within buildings of the System 80+ plant, a postulated failure of the Equipment Drain Tank is considered the worst case due to the high activity of its contents.
The Equipment Drtin Tank (EDT) is located in the Chemical and Volume Control System (CVCS) equipment area of the Nuclear Annex. The Nuclear Annex is a Seismic Category I structure and is in compliance l with Regulatory Guide 1.1.43. .
In the event of tank failure, the contents of the EDT are contained in the CVCS equipment floor area and drain via floor drains to the CVCS equipment area floor drain sump. The floor area in this section of the Nuclear Annex coupled with the CVCS floor drain sump capacity is adequate to contain the contents of the Equipment Drain Tank. Six inch curbing prevents the leakage from flooding into areas containing safety-related equipment. Curbing is provided for
- r individual rooms throughout the CVCS equipment area to protect equipment in the CVCS area from flooding. The spilled contents of the EDT, collecting in the sump and in the floor area, would be pumped via the equipment and floor drainage system to the LWMS for proAssing.
Collection of the liquid waste in the Nuclear Annex and subsequent pumping to the LWMS preclude a direct liquid release to the surface water or groundwater since the liquid is contained within a controlled area. A direct release of the liquid, resulting in contamination of surface water or groundwater, would reqd re the i inadvertent discharge of the contaminated liquid from the Waste !
Monitor Tanks in the LWMS. Since this postulate requires operator errors and/or equipment fanures in addition to the original l accident, it is unlikely a direct release of contaminated liquid ,
will take plar.e. l Furthermore, an uncontrolled release of liquid contaminants to the environment resulting from a failure of the Equipment Drain Tank is bounded by the accident analysis performed in Section 15.7.3 of CESSAR-DC. Although the Nuclear Annex is a Seismic Category 1 l structure, no credit for liquid retention can be taken per Standard :
Review Plan (SRP) 15.7.3 since the structure's foundation is not lined. Consequently, the potential for leakage through cracks in the foundation must be addressed.
Since the failure of the EDT in the Nuclear Annex would not result in liquid contaminants being directly released to the surface water, groundwater modelling would be utilized to evaluate the movement of radionuclides through the groundwater pathway. Using groundwater modelling, due to the extremely slow movement of radionuclides through the hydrogeologic medium, only the relatively long-lived radionuclides would have to be evaluated as reaching the nearest potable water supply. This sloe migration of radionuclides through the soil provides ample time for interdiction to mitigate the consequences of the tank failure.
An analysis of the characteristics of movement of the radionuclides through the soil to the groundwater to a potable water source is site-specific requiring information on the site surface and groundwater hydrology as well as soil characteristics (e.g.,
permeability).
Since the EDT is normally vented to the Gas Waste Management System (GWMS), there should not be a significant buildup of radioactive gases in the EDT. Therefore, the consequences of a gaseous release from the EDT due to a tank failure is bounded by a failute of the GWMS, which is analyzed in CESSAR-DC Section 11.3.7.
The Equipment Waste Tanks, Floor Drain Waste Tanks, Chemical Waste Tank, and Detergent Waste Tank are located in the Radwaste Building.
l The System 80+ Radwaste Building is designed to contain leakage which could result from tank failures in the building. The Radwaste Building design meets the requirements of Regulatory Guide 1.143 and contains no safety-related equipment. The lower elevatirn of the Radwaste Building is intended to act as a seismically designed l
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- to contain postulated leakages and preclude the release of tank contents to the environment. Here again, the tank contento are contained in a controlled area (Radwaste Building). The plant is
- designed to reclaim the contaminated licuid and process it to remove and capture the activity via the floor crain sumps and LWMS, 4
As in the case of the Nuclear Annex, however, no credit for liquid retention is given to the Radwaste Building since linars are not utilized in the foundation. The Equipment Drain Tank analysis given above, bounds postulated tank failures in the Radwaste Building since the activity of the EDT contents is much higher than the expected activity of the contents of the tanks in the Radwaste Building.
The Reactor Drain Tank, located in the Containment Building, is located in a pit to establish proper tank elevation to facilitate drainage from the reactor vesse'. A failure of the tank will result in the contents collecting in the pit and draining, via floor drain, to the Containment Building floor drain sumps in the Holdup Volume Area. The floor drain located at the bottom of the pit is adequately designed to accommodate a tank failure and the approximate 60,000 gallon capacity of the Holdup Volume Area is capable of containing the contents of the Reactor Drain Tank. From here, the contents can be routed to the LWM$ for processing. The lined Containment Building is given credit for liquid retention accoMing to SRP 15.7.3 and a failure of the Reactor Drain Tank will not result in a release of contaminated liquid to the environment.
A failure of the Reactor Drain Tank will not result in the flooding of safety-related equipment.
With the exception of the Reactor Drain Tank, drain and waste tanks, as well as the BAST, are located in areas containing no safety-related equipment. This significantly reduces the possibility for flooding of safety-related equipment due to tank failures.
As demonstrated in the above analysis, with the exception of the Reactor Drain Tank, which is classified as Seismic Category I and Quality Group C, adequate design features exist in the System 80+
design to reduce the need for seismically designed waste and drain tanks.
(b) The System 80+ equipment and flood drainage system contains features to withstand active component failures, blockages, and probable flood without inundating safety-related areas.
Each quadrant of the System 80+ Reactor Building Subsphere is provided with an individual floor drain sump. Each sump is equipped with two full capacity sump pumps. Each quadmnt's sump pumps are powered from a separate electrical channel, in the event of power loss, the pumps are powered from the Diesel Generators.
The Containment Isolation valves (CIVs) in the Equipment and Floor Drainage System adhere to the requirements for CIVs as stated in Section 6.2.4 of CESSAR-DC and are designed to fail in the safe position,
o 6 The System 80+ design contains features to provide flood protection and accommodate blockages in the Equipment and floor Drainage System. The Nuclear Annex and Reactor Building Subsphere are physically separated divisionally by walls with no unsealed penetrations up to elevation 65+0. At higher elevations, six inch curbs are provided for flood control. This prevents a potential flood in one division from flooding into the other division. The Reactor Building Subsphere is further divided into quadrants up to elevation 65+0. The basement in each division has adequate volume to collect water from a break in any system without flooding the other division. The Component Cooling Water Heat Exchanger Building and Station Service Water Pump Structures are divisionally separated by walls such that a flood in one division cannot flood the other division.
Mechanical systems, including the Equipment and floor Drainage System, contain no cross connections between divisions such that a flood in one division cannot flood into the other division and disable safety-related equipment. Backwater valves in the Equipment and Floor Drainage System are provided to protect against backflow into safety-related areas.
Blockages in the Equipment and floor Drainage System are accounted for under the divisional separation design criteria of System 804.
Floor drains are separated by quadrants in the Reactor Building Subsphere and divisionally separated elsewhere. Blockages which may occur in the floor drainage system will not affect adjacent quadrants or divisions.
(c) To preclude the inadvertent release of radioactive waste to the environs, the drainage and collection systems used to handle radioactive or potentially radioactive liquid radwaste are completely separate and isolated from the systems used to handle strictly nonradioactive waste, Nonradioactive floor drain sumps with provisions to route contaminited contents to LWHS have sufficient safeguards to preclude inadsertent routing of contaminants to noncontaminated drainage systems. The Nuclear Annex nonradioactive floor drain sumps are equipped with radiation monitors in the discharge lines which actuate air-operated fail-safe valves upon detecting any radiation and divert the flow te the Floor Drain Waste Tanks. Normally, the nonradioactive sump contents are pumped to the Turbine Building sump which is also equipped with radiation monitors, a control room alarm, and air-operated valves to divert contaminated liquid waste to the Floor Drain Waste Tanks.
, This diversion does not mean that the liquid cannot be discharged to l_ the environment, it merely gives the Liquid Waste Management System t
a single point through which all radioactive water can be released according to administrative procedures.
Radiation monitors, backwater valves, and air-operated fail-safe valves throughout the equipment and floor drainage system provide a reliable means of preventing inadvertent mixing of contaminated and noncontaminated liquid wastes. Operator error is reduced through 1
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a A,_ m - < - - - -n x- a (liation monitor /afr operated valve control of 1ud a e lo 1
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e t Qgtstion 410.83 Provide the following for the heating, ventilation, and air conditioning (HVAC) systems described under Sections 9.4.1 through 9.4.8:
(a) Tabulated design equipment parameters at least including anticipated normal and accident heat loads, system design airflow rates, heat re oval capacities, (b) System P&lDs and system descriptions that contain sufficient information to determine:
(1) Interfaces between safety and non-safety portions of the systems (including changes in component safety classifications),
(2) Containment penetrations and isolation capabilities, (3) Complete system flow paths, (4) Interfaces with other systems and system boundaries, (5) Isolation capabilities between essential and non-essential portions of the systems, and (6) Physical division between redundant portions of the systems, (c) Provide an FMEA for the essential portions of each water system since essential portions of the HVAC systems must be able to withstand design basis accidents and a concurrent single active failure, and (d) P' ovide a tabulated summary of environmental design parameters for equipment cooled by the HVAC system in mild environments, ikippate 410,83 (a) Please see Tables 9.4-2-1, 9.4-2-2, 9.4-2-3, 9.4-2-3 and 9.4-2-5 for design equipment parameters.
(b) Reference the following flow diagrams: Figure 9.4.1-1 " Air Flow Diagram Nuclear Annex Control Building", figure 9.4.2-1 Air Flow Diagram fuel Building", Figure 9.4.5-1 " Air flow Diagram Cooling Reactor Subsphere", figure 9.4.5-2 " Air flow Diagram Ventilation Reactor Subsphere", figure 9.4.6-1 " Air Flow Diagram Containment Cooling Purge & Pressure Control", figure 6.2.3-1 " Air Flow Diagram Annulus Vent" and Figure 9.2.9-1 " Flow Diagram Chilled Water".
Reference the revised sections 9.4.1-9.4.9, Amendment 1, for the.
system descriptions.
(c) Each of these systems has two divi.!ons that are completely separated and 100% redundant. Each division is powered from a separate division of Class IE electrical power. Therefore, each of these systems is capable of withstanding a single active failure.
Reference the chilled water flow diagram figure 9.2.9-1 for the redundant chiller systems. Reference Section 9.2.9 for the chilled water system description. Reference section 9.2.1 for Service Water (
and Section 9.2.2 for Component Cooling Water redundancy.
(d) Please see Section 3.11 of CESSAR-DC.
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Qggstion 410.84 Concerning the control building ventilation system, verify that the intake monitors for the control room outside air supply alarm in the control room prior to or upon isolation of an intake. Verify that provisions will be mde for in service inspection of control room isolation dampers and other control room HVAC dampers.
Response 410.84 Each outside air intake location is monitored for 'io presence of radioactivity, toxic gases, chlorine and products of combustion.
Isolation of the outside air intake occurs automatically upon indication of high radiation level, high chlorine concentration or smoke concentration in the intake. Should both intakes close, the operator can override the intake monitors and by inspection of the control room readouts select the least contaminated intake. This will ensure pressurization of the control room. Please see figure 9.4.1-1, " Air flow Diagram Nuclear Annex Control Building" and Section 9.4.1, " Control Building Ventilation System".
- I Qgestion 410.J1 Provide the following for the fuel building ventilation system:
(a) Section 9.4.2.1, Design Bases, of your submittal states that the fuel building ventilation system will be in operation whenever irradiated fuel handling operations in the fuel pool are in progress. Provide the intended system configuration and its mode of operation during normal plant operations (other than fuel handling).
(b) Verify that the radiation detectors in the exhaust ducts are far enough upstream of the bypass dampers to ensure that the dampers )
will have completely actuated to direct exhaust flow through the filter trains before the first airborne mhterial reaches the bypass dampers, and (c) Provide assurances that system performance will not be affected due to pipe breaks and whip, jet impingement and other associated failures of nonseismic systems in the vicinity.
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(a) During normal power operation, the one 100% capacity ventilation supply air handling unit and associated dampers and ductwork will be continually operated unless it is put out of service for repairs and maintenance. At least one of the t.<o 100% capacity exhaust systems will be continually operated with the filter trains bypassed. The following has been added to the end of Section 9.4.2.2: "During normal power operation, the one 100% capacity ventilation supply air handling unit is continuously operated unless it is out of service for maintM ance along with at least one of the two 100% capacity exhaust systems with the filter trains in the bypass mode."
(b) The following senter.ce will be added to the second paragraph of Section 9.4.2.3: "The radiation detectors are located to ensure that the dampers will have completely actuated to direct exhaust flow through the filter trains before the first airborne radioactive material reaches the bypass dampers."
(c) The first sentence of the eighth n.ragraph of Section 9.4.2.1 has been revised as follows: "The fue ' ding Ventilation System is located completely within a nisir .egory I structure and all essential components (exha' n fih . rains, exhaust fans, exhaust ductwork) are fully protected from ,ioods, tornado missile damage, internal missiles, pipe breaks and whip, jet impingement and interaction with nonseismic systems in the vicinity."
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4 Ouestion 410.86 l
Concerning the auxiliary and radwaste building ventilation system, you ;
have stated in Section 9.4.3.2, System Description, that the general !
ventilation system is not safety-related and performs no function essential to safe shutdown or post-accident operation. Provide i additional information supporting a determination that the system is not '
required to control radioactive releases from the auxiliary building in the post-accident environment.
Response 410.86 Reference Section 9.4.5 for a description of the Subsphere Building Ventilation System. The subsphere ventilation is row safety-related and is credited for filtering releases for safe guard pump leakage in Chapter 15 LOCA Offsite & Control Room Dose Analysis. Reference section 9.4.9 for the Nuclear Annex Ventilation System. The Nuclear Annex Ventilation Systems are not engineered safety features and no credit has been taken
, for their cperation in analyzing the consequences of design basis :
accidents.
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Ouestion 410.87 ,
Concerning the station service water pump structure ventilation system, Section 9.4.8 does not provide information to preclude the accumulation of dirt and dust in electrical equipment, and makes no mention of filters. Provide information pertinent to the location of the system ;
intakes and/or other features (e.g., cabinet gaskets) designed to prevent the ingress of dust to electrical cabinets.
Response 410.87 The design of the station service water pump structure and associated ventilation system is site dependent. The fo'/ lowing has been added to Section 9.4.8.1 to assure this issue is adequately covered in the site-specific design: "The location of the station service water pump structure ventilation system intakes are such to minimize the ingress of dust. All electrical cabinets located in the station service water pump structure are provided with features (e.g., cabinet gaskets) to prevent dust ingress."
. I Ouestion 410.88 1
Provide the following for the diesel generator support system as '
described in Sections 9.5.4 through 9.5.9:
(a) Verify that the diesel generator building is not only Seismic Category I, but that is also provides protection against the affects of missiles and floods as required to meet the requirements of GDC 2 of Appendix A to 10CFR50, (b) Provide system layout diagrams for the diesel generator support systems with sufficient detail so that component location within (or )
outside) the diesel generator building can be determined and the ;
accessibility of equipment for test and maintenance can be )
evaluated. ;
(c) System P&lDs showing equipment classification and c1carly 1 identifying system boundaries and system hterfaces, (d) FMEAs, (e) Verify that the system components and structures, including isolation devices between essential and nonessential portions of-the system, are Seismic Category 1. Also, for those portions of the system not housed in the diesel generator unit structure, provide information verifying that these portions of the system are:
(1) Seismically qualified.
(2) Protected from the results of the seismic failure of nonsafety-related systems, (3) Tornado missile protection, and (4) Flood protected, and (f) Provide information concerning the diesel generator engine cooling water system (Section 9.5.5), starting system (Section 9.5.6), and lube oil system (Section 9.5.7) alarm and/or trip signals, including the lock-out circuit logic of these trip signals during EDG operation following a loss of coolant accident (LOCA) or loss of offsite power (LOSP) and conformance with the guidance of RG 1.9, Position C.7.
Response 410.88 (a) The following has been added to Sections 9.5.5.1, 9.5.6.1, 9.5.7.1, 9.5.8.1, and 9.5.9.1:
"All components and piping are located within a Seismic Category I structure (diesel generator building) and all essential components are fully protectad from floods, tornado missile damage, internal missiles, pipe breaks and whip, jet impingement and interaction with nonseismic systems in the vicinity."
The following has been added to Section 9.5.4.1:
l "All components and piping are located in a Seismic Catego.'y 1 structure (diesel generator building) except for the fuel oil
. 3 i storage tanks and a cortion of the pipias from the fuel oil storage tanks to the day tan (, which is seismically qualified and protected.
All essential-components and piping are fully protected from floods,
-tornado missile damage, internal missiles, pipe breaks and whip, jet impingement and interaction with nonseisnic systems in the vicinity."
(b)& General- otrangements including location of the diesel generators are (c) provided in Section 1.2 figures. Flow diagram and system descriptions for specific diesel generator support systen.s have been provided in- Sections 0.5.4 to 9.5.9 of CESSAR-DC, Amendment I.
-(d) A separate and complete fuel cil system, cooling water system.
starting-air system, lube oil system air intake and exhaust system and diesel building sump pump system is provided for each of the two emergency diesel generators. Therefore, a single failure in any one ,
of these systems which would render its associated diesel generator inoperable-is accommodated by starting and Punning the other' emergency diesel-generator and its associated support systems.
Therefore, . single failures are addressed without identifying the specific cause of the failure.
_(e) Sections 9.5.4.3, 9.5.5.3,-9.5.6.3, 9.5.7.3, 9.5.8.3, and 9.5.9.3 designate the portions of the systems which are ANSI Class 3 and the i portions o.f the-systems which are ANSI Class 4, if applicable.
Section 3.2.1 states that all components in Safety Class 1. 2, and 3 are Seismic Category I. The only components not located in the
. diesel building are the ful oil storage tahks. Protection of these tanks is discussec in item (a) of this response.
(f) Diesel generator protection systems are discussed in Section
-8.3.1.1.4.4.- Control.. Room indication of diesel generator operational status is discussed-in Section 8.3.1.1.4.5.
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Question 410.89 Provide the following for the diesel generator fuel oil system:
(a) Verify that the fuel oil storage tank fills and vents are located
- above the probable maximum flood level.
- (b)- -Verify that the fuel oil sampling and impressed current cathodic
- protection-system surveillances are in conformance with the guidance of RG 1.137, Position C.2, and '
(c) Verify that sufficient fuel oil storage capacity for seven days is provided, including consideration of periodic testing and the t unusable portion below the EDG suction location. Also, provide the fuel oil storage and day tank capacities, include flow capacity data from the storage tank to the day tank.
Response 410 E
- (a): The first sentence of the last paragra)h of Section 9.5.4.2.1 has been reviseo as follows: "The day tan < vent and fuel oil storage
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- tank ven'ts-and fill' connections which are exposed outdoors, are protected from tornado missiles through the construction of the
'-vents using heavy gage pipe end are located above the probable -
maximuin flood level."
(b) The following has been added to Section 9.5.4.3: "The fuel oil
- sampling and impressod current cathodic protection (if provided) system surveillances--are in conformance with guidance of RG 1.137, Position C.2."
(c) The'second and third sentences of'the third paragraph of Section 9,5.4.3 state, *The fuel oil storage _ capacity is= based on continuous 4
operation of the diesel generator engines at rated load for a period of seven days. A 10 percent margin in storage capacity is provided to preclude the necessity of refilling the tanks following routine performance testing."
The following has been added (Amendment I)_ after the first paragraph of Section 9.5.4.2.1: " Typically, this requires a combined usaale volume of 135,000 gallons. The site-specific SAR sha11' verify that- :
this is adequate for the diesel generators purchased."
The first two sentences of the second paragraph of Section 9.5.4.2.1 have been deleted and replaced (Amendment 1) with the following:
" Fuel Oil is transferred by the fuel oil transfer ) ump from the storage tanks to the day tank which is located wit 11n retaining walls inside the diesel generator building. The fuel oil transfer si pump is 'also located in the diesel generator building and is typically sized for 75 g)m. The day tank has a-sufficient capacity of fuel oil-to operate tie diesel generator engine in excess of 60 i minutes at full load. Typically,-this_ requires a day tank of 900 l gallons. The site specific SAR shall verify that fuel oil transfer pump flow and day tank capacity are adequate for the diesel j generators purchased."
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Question 410.90 Provide tabulated design data including design flow and heat removal requirements for the diesel generator engine cooling system.
Question 410.91 Provide tabulated design data including compressor capacity, power source and air receiver capacity for the diesel generator engine starting system.
Question 410.92 Provide power supply information for the motor-driven pro lube oil pump and system design data including pump flows, cooling system heat removal ,
capabilities and electric heater characteristics for the diesel generator engine lube oil system.
Ettrmlits 410.90. 410.91. and 410.92 The flow diagrams for diesel generator systems have been added to CESSAR-DC, Amendment !. These diagrams identify system components and arrangements. The specific data for these support systems will depend on as procured diesel generator data.
. 4 Question 410.93 Provide the following for the diesel generator engine air intake and exhaust system:
(a) Complete the second sentence of the second paragraph in Section 9.5.8.3 by providing the naeded information, (b) Verify that there are no active components in the air intake and exhaust system, and (c) Identify the restrictions used for the location of the diesel generator building in relation to possible onsite sources of gases which may be intentionally or accidentally released, to ensure that such releases do not result in degraded operation of the LDG, Euppnse 410.91 (a) The second sentence of the second paragraph of Section 9.5.8.3 has been revised as follows: "This fact and site-specific analysis of the diesel generator engine exhaust will establish that the rise of the exhaust gases is sufficient to preclude the possibility of recirculation to the point that system integrity is jeopardized."
The next sentence will start " Normal ventilation flowrate is 5%...".
(b) ThefollowingsentencehasbeenaddedattheendofSection 9.5.8.2.2: There are no active components in the air intake and exhaust system."
(c) The following has been added to the end of Section 9.5.8.3: "0nsito stortge of gases is discussed in Section 9.5.10. These gases are stored-at a distance from the diesel generator building such that there is no threat to the proper operation of the diesel engines."
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I Ouestion 410.94 Provide the following for the diesel generator sump pump system:
(a) Verify that this system is evaluated concerning the internal flood l protection for the diesel generator building. l l
(b) Provide interface information with the equipment and floor drain '
system, and I (c) Provide pumping design characteristics and identify the pipe rupture used to size the sump pumps. Also, provide the maximum expected flow rate data from the pipe rupture.
Response 410.94 (a) Section 9.5.9.1 has been revised as follows:
"The diesel generator building sump pump s,. stem is designed to remove 1cakage and equipment drainage from the diesel generator building and to protect the diesel generator units from internal flooding e-aused by the maximum credible pipe rupture in the diesel generator building."
(b) The following has been added to Section 9.5.9.2: "The sump pumps start automatically on high sump water level and transfer the water to the equipment and floor drain system."
(c) The following will be added to Section 9.5.9.2:
"The diesel generator building sumps and sump pumps are designed for a constant inflow rate of 75 gpm with a maximum pump cycle time of three starts per hour (one pump operating with 37.5 gpm inflow).
The maximum pumping flowrate with both pumps operating is 150 gpm.
This-pumping caucity constitutes a restriction on the detailed piping layouts developed for construction (af ter equipment procurement).
. T Ouestion 410.95 Provide the following for the Turbine Generator System (TGS):
(a) Additional information on the power / load unbalance circuit, including closure time for all control valves upon load / power
. unbalance condition, (b) - Verification of " excess -vibration" as one of the signals that leads.
to a turbine trip, (c) Justification to deviate-from the inservice inspection review guidance of SRP 10.2 for the turbine components (e.g., stop valves, intercept valves). You have stated in your submittal that 151 for these components All be approximately once every five years, during refueling while abn'e guidance provides the ISI once every three and a third years, during refueling or maintenance shutdown, (d)- Verification to assure thd IGS inservice inspection (ISI) testing
_ will be performed in arewdence with the requirements of ASME codes.
-and-
-(e) 'Information that will confirm that the extraction check valves will be capable of' closing within time limits required to maintain stable conditions.
Response 410.95'
-(a) Information on the power / load unbalance circuit depends on the specific turbine-generator procured. Procurement specifications will ensure that the specific equipment procured meets tne design requirements summarized in Section 10.2.
(b) Section 10.2 of CESSAR-DC will be modified to include a requirement .
'that the turbine generator include a turbine trip on excess '
vibration.
(c) CESSAR-DC Section 10.2.4 has been revised to reflect the requirements set forth in SRP 10.2 for inservice inspection of the main steam and reheat valves.
(d) CESSAR-DC Sect ta 10.2.4 has been revised to include the requirement
.for inservice Inspection testing-ta be performed in accordance with the applicable ASME codes.
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(e) Information to confirm the closure time of the extraction valves depends on the specific component procured. Com)11ance with this closure time requirement will be required througi equipment procurement specifications.
i Question 410.96 Verify that the main condenser design provides features to protect safoty-related systems from the effects of a flood due to its complete failure.
Response 410.96 The following has been added to the end of Section 10.4.1.3:
"A leak or failure in the condenser shell would allow condensate to drain out, but the pits located below the condenser will hold more water than the condensate hotwell volume. The flooding due to a loss of condenser water box or circulating water piping would be limited to the turbine building which contains no safety-related equipment.
A failure in the recirculating water system or the main condenser large enough to cause flooding will be detected by high level alarms in the turbine building sumps. The operator can isolate the appropriate equipment."
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Question 450.1 Provide the following information concerning the control room habitability system for conformance with the guidance of SRP 6.4, Revision 3.
(a) P&lDs of the control building ventilation system (CBVS),
(b) Drawings showing the control building emergency zone (CBEZ) and air intake locations and distances relative to potential radiological and hazardous and toxic chemical release points.
(c) Drawings showing the bo'andaries of the CBEZ, indicating how specific rooms and areas are enclosed in or inttrface with this zone,'
(d) FEMA of the CBVS which demonstrates that it will perform its emergLncy function with any single active failure prior to or during an emergency, and (e) Provide the following CBVS and CBEZ design data:
- 1) Normal ventilation flow rates,
- 2) Emergency mode flow rates for the CBEZ,
- 3) Design basis inleakage rates of potentially contaminated air for dampers, 6ucts, etc., associated with the CBEZ by type of damper, etc., and with its totals.
(4) Composition, type, and size of filters used in the emergency -
mode of the CBVS, (5) Free air volume of the CBEZ and total volume of the control room, (6) Maximum number of occupants expected in the CBEZ during an emergency, (7) The presence, quantity,- and location of toxic and hazardous substances within the site boundary (e.g., chlorine for water treatment, CO 2 for fire suppression), and (8) The location and quantity of bottled air which would be accessible to occupants of the CBEZ during an emergency.
Response 450.1 (a) Please see Figure 9.4-2, " Air Flow Diagram Nuclear Annex Control Building", which were provided in Amendment 1.
(b)& Please see figures 1.2-3, 1.2-7, and 1,2-8 which were provided in (c) Amendment H.
(d) The control building air-handling system consists of two independent, full capacity systems. Each system serve, the associated train of essential electrical equipment areas. Each system is powered from independent Class lE power sources and served
, from separate essential chilled water system:. Reference section 9.4.1.3 Safety Evaluation.
(e) (1),(2),(3) Reference Table 9.4-2 for CBVS and CBEZ for design data.
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(4) Reference Table 9.4-4, " Input for Release Analysis filter Efficiencies".
(5) Please see figures 1.2-3, 1.2-7, and 1.2-8 wnich were provided in Amendment H. The total volume of the control room is 100,000 cubic feet.
(6) The maximum number of occupants-expected in the CBEZ during an emergency is 8.
1 - Shift Supervisor 1 -
Control Room Supervisor ,
3 -
Operators 1 1 -
Technical Advisor 1 -
NRC Representative (observer) 1 -
Utilities Management (observer)
(7) Please see figure 1.2-1, Amendment H, for the location of oxygen, nitrogen, and hydrogen bulk storage tanks.
l (8) The control room habitability system consists of 6 compressed )
air cylinders located inside the control room.
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Questien 450m2 Provide information to demonstrate that th9 technical support center has the same radiological habitability as the .ontrol room under accident conditions, in accordance with th? guidan< e of NUREG-0696, " functional Criteria for Emergency Response F *cil: ties." ;
Resoonse (12.2 Section 13.3.3.1.6 discusses the habitability requirements for the
-technical support center. In addition, the ventilation system fer the technical support center is discussed in Section 9.4.1,1, the fourth paragraph of Section 9.4.1.2, and the second partgraph of Section 9.4.1.3.
Question 48L.2 In order to evaluate the performance capability of the containment spray system for conformance with GDC 38 and P 6.2.2 Revision 4, provide the fol hwing design parameters: ,
(a) Spray header location relative to containment inner structure.
(b) Spray nozzle arrangement on the spray headers and the expected spray pattern.
(c) Spray drop size spectrum and mean drop size emitted from the nozzle as a function of differential pressure across the nozzle.
(d) Average spray crop residence time in the containment atmosphere.
Response 480.9 (a)& The layout of spray headers has shown that adequate coverage can bc (b) obtained. The number and orientation of spray nozzles will be provided to ensure that spray patterns will cover 90% of the net free volume of containment.
(c) The spectrum of drop sizes is that spectrum emitted from a SPRACO Company 1713A nozzle operating with a differential pressure of 40 psi, The mass mean drop diameter produced at these conditions is 530 microns. This diameter is assumed to be 660 microns in the washout calculation to account for coalescence.
(d) The average spray droplet residenu time in the containment atmosphere is approximately 13 seconds. This value is calculated assuming a minimum fall 5eight of 82.5 feet, and taking the terminal velocity of the average drop as 6.3 feet /sec. This velocity is calculated using guidance from " Properties c/ Air-Steam Mixtures Containing $ mall Amounts of lodine," Knudsen, JG, BNWL-1326, April 1970.
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0 4 Ouestion 480.10 Provide 69 evaluation of the heat removal capability of the containment spray heat exchangers with surface fouling on the secondary side for conformance with GDC 38 and SRP 6.2.2 Revisian 4.
Response 480.10 The design of the System 80+ containment spray heat exchangers is based on a fouling resistance of 0.0005 hr-sq.ft.-F/Blu on both the tube side ,
and the shell side. These values are given in Amendment I to CESSAR-DC, (
Table 6.5-1. The containment analysis presented in Section 6.2.1.1 '
f models the containment spray heat exchangers using the design parameters from Table 6.5-1. The results of the analysis show that the SRP requiraments are met. The SRP acceptance criteria are that the peak containment pressure must be below the containment design pressure and that the containment pressure at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> must be less than halt' of the peak calculated pressure.
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Ouestion 480.11' Provide thr design features and provisions of the -containment, spray
= system for-periodic inspection and operability: testing for conformance .
-with GDCs 39,740, and SRP 6.2.2. Revision 4.- Also, outline the anticipated schedule and' extent ofL this inspection and testing. -
Responip 480 11:
- Designiprovisions for inservice inspection of containment spray system components are-described:in CESSAR-DC Sections 16.5.1.3.K.-12 and-6.5.1.3.K.13. =The prooram for inservice inspection and testing of the:
containment spray system is covered-by-the description provided in CESSAR-DC Section 6.6 for inservice-inspection;of Class 2 and 3 components. Addit'.0:a1 information on inservice inspection and testing of the contair. ment sr ray-system is found in Technical Specification 3.0.6. -
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1 Ouestion 480.12 Provice an evaluation of the effect of CCWS failures on the containment spray system (CSS) via the CSS heat exchanger as addressed in Table 6.5-3.
Response 480.12 Table 6.5-3 addresses the loss of component cooling water to a containment spray heat exchanger. This could occur for the following two single failures: (1) the containment spray heat exchanger isolation valve failed to open on a Containment Spray Actuation Signal (CSAS) and (2) failure of a diesel generator. The other possible failures are discussed in Table P.2.2-2 (Amendment 1) plus the attached future revision. The design basis criteria will still be met if a failure of cooling water (CCWS) to a single CSS heat exchanger occurs because two separate, independent, and 100% redundant divisions are provided (CESSAR-DC Section 6.5.2). ,
i Each division contains one containment spray heat exchanger. Each containment spray heat exchanger is sized to transfer 100% of the design heat load used in the containment peak pressure analyses (CESSAR-DC Table !
6.2.1-19). Only one containment spray heat exchanger is required to prevent over-pressurization of the steel containment vessel, lable 9.2.2-2 in CESSAR-DC Section 9.2.2 al;o addresses this failure.
Since both divisions of the CCWS are separate, independent, and 100%
redundant, the failure of one of the CSS heat exchanger isolation valves on the component cooling water systeia will leave the other division unaffected.
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l Ouestion 480.13 You have stated that secondary containment functional design will be provided later. Provide information to permit the staff to perform an integrated review of the secondary containment.
Response 480.13 The secondary containment function is to capture emissions.and the annulus ventilation system exhausts them out through a filter train before release to the unit vent. Please see figure 6.2.3-1 " Air Flow Diagram Annulus Vent" and Section 6.2.3, " Annulus Ventilation System".
l Question 480.14 Provide the following information to evaluate the containment isolation system (CIS) in conformance with GDCs 16, 54, 55, 56 and 57 guidance of RG 1.141 and SRP 6.2.4, Revision 2:
(a) Completed Table 6.2.4-1, " Containment Isolation System,"
(b) The number and physical location of containment isolation valves (CIVs),
(c) The actuation and control features for each valve, (d) The positions of each isolation valve for normal, post-accident, and valve operator power failure conditions,.
(e) falve actuation signals for each isolation valve, (f) The closure time and basis for each isolation valve, and (g) The mechanical redundancy for each isolation valve.
Response 480.14 (a) Table 6.2.4-1, " Containment Isolation System," and accompanying Figures 6.2.4-1 (8 sheets) " Containment Isobtion Valve Arrangement," have been added to CESSAR-DC (Amendment 1).
(b) The number of Containment Isolation Valves may be obtained directly from Table 6.2.4-1 and Figure 6.2.4-1. Most penetrations will be located with the outside containment isolation valve as close as possible to the containment wall in the Reactor Building Subsphere Pipe Chase, and the inside containment isolation valve located within containment between the crane wall and inside containment wall. The following outside containment isolation valves will be located in spaces other than the Reactor Building Subsphere Pipe Chase:
- 1) The four Safety injection and Containment Spray Pumps' Suction
- from the IRWST penetration lines' outside containment isolation valves (see CESSAR-DC Table 6.2.4-1, items 13,14,15, and 16) will be located in the Containment Subsphere ECCS Pumprooms.
- 2) The High Volume and Low Volume Containment outside Containment Isolation Valves will be located in the Annulus.
- 3) The four Main Steam lines and four Main Feedwater lines, due to the nature of their service, will have all of their containment isolation valves located within the Main Steam Isolation Valve House, with the exception of the Main Feedwater inside containment check valves.
(c) Table 6.2.4-1, gives actuator type and whether or not the valve is automatically, manually, or remotely actuated.
- u (d) The positions of each Containment isolation Valve for normal, post-accident, and valve operator power failure conditions are given in Table 6.2.4-1.
(e) The actuation signals of each Containment Isolation Valve are given in Table 6.2.4-1.
(f) The closure time and basis for each isolation valve are in accordance with ANSI /ANS 56.2-1984, which specifies a 5 second closure time per inch of nominal valve diameter, but no more than 15 seconds for valves 3 inches or smaller in diameter, or more than 60 seconds for valves greater than 12 inches in diameter, unless specifically justified. Maximum allowable actuation times are imposed or Containment Isolation Valves consistent with their required safety function, e.g. Main Steam Isolation Valves (MSIVs) must close within 5 seconds, since the safety function of Main Steam Isolation has this requirement. Requirements for low Volume Purge Subsystem isolation are clar.4 )d CESSAR-DC Section 6.2.4.3, Amendment I.
(g) Each penetration has redundant Containment isolation Valves (refer to Table 6.2.4-1 and Figures 6.2.4-1, except the following:
- 1) The four Safety Injection and Containment Spray Pumps' Suction from the IRWST penetration lines have only one valve each.
This is permissible since the IRWST itself provides a water seal against the release of fission products during an accident.
- 2) The Integrated Leak Rate Test pressurization line, which has a blind flange on its inside containment boundary used in conjunction with one outside containment manual locked closed valve. This line is in use only during an outage for the Integrated Leak Rate Test, and the blind flange will be scaled during normal operation.
- 3) The Fuel Transfer Tube Closure and Equipment Hatch, which have double seals, and are both considered to be actual boundaries of containment in and of themselves. These are both Type-B leak rate tested in accordance with 10 CFR 50 Appendix J, as is currently described in CESSAR-DC, Section 6.2.4.2.
- 4) Electrical penetrations, which have a sealing medium or double seal 0-Rings to provide containment integrity. A description of electrical penetrations and their required testing has been added to CESSAR-DC, .ection 6.2.4.
. s Question 480.15 Provide the containment isolation design provisions for all instrument lines that penetrate the containment in accordance with the guidance of RG 1.11.
Ennonse 480.15 The following will be-added in Section 6.2.4.1.2: " Instrumentation and control sensing lines which penetrate the containment are provided with containment isolation provisions which meet the intent of Regulatory Guide 1.11."
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Question 480.16 Provide correct maximum integrated radiation dose for the CIV design (see page 6.2-35 as 4.0E-7 rads), in addition, qualify this dose regarding whether it is solely due to gamma radiation. If so, provide design beta and neutron doses for the CIVs and associated valve operators.
Sesponse 480.16 Section 6.2.4.1.2 states the environmental qualification requirements for containment isolation valves. This statement references Section 3.11.
Environmental conditions for various zones are provided in Appendix 3.11A and the containment isolation valves will be qualified to the requirements for the zone in which they are located. The last two paragraphs of Section 6.2.4.2 will be deleted to avoid further confusion.
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- i Ouestion 480.17 Provide specific details on outside containment leak testing provisions for remote manual valves of engineered safety feature, engineered safety feature-related, or safe shutdown systems penetrating the containment for conformance with the guidance of SRP 6.2.4 Revision 2.
Response 480.17 All valves outside of containment will be Type-C leak tested in accordance with 10 CFR 50 Appendix J, unless otherwise specifically justified (see Table 6.2.4-1 and Figure 6.2.4-1). Valves will be pressurized in the accident direction (out of containment) to at least DBA pressure. In this_ condition, leakage will be measured and verified not to exceed allowable limits when summed with all other containment penetration leak tests in accordance with 10 CFR 50 Appendix J.
Appropriate test connections, drains, vents, and pressurizing means will be provided to allow penetrations' inside and outside containment isolation valves to be pressurized in the accident direction (out of containment). Testing requirements and methodology are described in CESSAR-DC, Section 6.2.4.4 (Amendment 1).
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, l 0ue'stion 480.18l l
' Provide a-system reliability evaluation-and outside containment = enclosure-design details for all-engineered safety features or engineered safety Lfeature-related system containment penetrations having only one isolation
' valve .for conformance with the guidance of SRP 6.2.4, Revision 2.
Response 480.18 ,
A single containment isolation velve will be provided in each in-containment refueling water storage tank (IRWST)fsuction line to the engineered safety-related (ESI) pumps. -The valve need not be enclosed in
' a leak-tight enclosure, since all- requirements of Note 56.1 in Appendix A of ANSI /ANS 56.2-1976 are met as outlined below:
(a) .The _ valves are attached to lines which are extensions of containment and enclosed in a pump room adjacent to the containment which has provisions for environmental control of any-fluid leakage.- The IRWST ruction lines are shown in Figure 6.8-1 (Amendment I). :The containment isolation valves for these lines are located in the reactor building subsphere along with the ESF pumps.and heat exchangers;as shown in Figure 1.2-4. The reactor building subsphere has provisions for environmental control of any fluid leakage (i.e.,
leakagec is- collected in.the sumps shown in Figure 1,2-4 and the reactor subsphere area is maintained at a negative pressure, and.all potential radioactive releases are filtered and monitored as discussed in'Section'9.4.5).
-(b) The lines from the IRWST are always submerged so no containment atmosphere can impinge upon'the valves.
(c) The-systems which the lines from the IRWST connect to outside containment are closed systems in-the preceding standard (N271-
- 1976), including 3.6.4 and 3.6.7 -(see Figures 6.3.2-1A and IB).
(d) -The valves provide a barrier outside containment to prevent loss-of
-IRWST water should a leak develop. The valves 'are closed remotely from the control room and the' operator can detect which one of;the four_ valves.to close by indication of high sump water level in the
-sump located'within-the quadrant in which the associated ESF equipment and valve is located (see Figure 1.2-4 showing quadrant separation, ESF- equipment location and sumps).
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y -w l0uestion 480.19-Provide the design details of any, sealed closed barriers that are used in
. place of-automatic isolation valves for containment penetrations for
.conformance with the guidance of SR 6.2.4 Revision 2.
Resoonse-480.19' The {quipment Hatch, Fuel Transfer Tube Closure, and Integrated Leak Rate Test Pressurization Line use blind flanges or are themselves part of the containment botadary. These penetrations are sealed during-normal o>eration-and-therefore do-not require an automatic closure function. i Tie Equipment Hatch and' Fuel Transfer Tube Cover have a double seal which is Type-B: leak--rate tested in accordance with 10 CFR 50 Appendix J. The ILRT--Pressurization Line blind flange (located inside containment) is-complimented by a-locked closed manual valve-outside of containment for containment isolation. Alsos electrical. penetrations are qualified as part of the-containment boundary and utilize either a sealing medium-or
. double seal .0-Rings,- both of which are Type-B leak rate tested in m accordance-with-10 CFR 50 Appendix-J.
The following: statement will be added to.Section 6.2.4.2 to incorporate the above-leak rate. testing information that'is not already in CESSAR-DC:
_- "The fuel transfer tube blind flange is type-B leak rate tested in l-l_.
l accordance with 10 CFR 50 Appendix.J."
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Ouestion 480.20 Provide the relief valve setpoints for valves which are used as CIVs.
Response 480.20 The following will be added to Section 6.2.4.1.2, item B: " Relief valves for providing overpressure protection from heat up between closed containment isolation valves have a relief set point equal to the design pressure of the piping." Item B already states that the design pressure of all piping and connected equipment comprising the isolated boundary is greater than the design pressure of the containment.
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Question ~480.21 ,
~ Provide'an' evaluation 1for selecting 5 psig as the; containment isolation .
actuation- signal (CIAS) setpoint.= (See Chapter.7). This evaluation j;, should follow;the guidance of SRP 6.2.4,' GDC 54, NUREG-0737, and NUREG- t 0718 for selecting the-CIAS"" minimum value= compatible with normal .
operating conditions."
- Response 480.21
-Amendment I'to CESSAR-DC changed to-CIAS-setpoint~from 5 psig to 2.7'.
>sig. The-guidance'of SRP 6.2.4, GDC 54, NUREG-0737, and NUREG-0718 has-seen.followed in selecting the CIAS " minimum value compatible with normal operating conditions." The 2.7.psig setpoint is based-on the following components:-(1) a 0.4 psig Technical Specification limit,-(2) a 1.2 psi
-maneuvering spike based on data from similar= plants,_and-(3) an assumed 1.1 psi channel instrumentation error.
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Question 480.21 SRP 6.2.4, Revision 2 guidance states that all lines that provide an open path between the containment and the outside containment environment should be equipped with radiation monitors. These lines should be isolated on receipt of a high radiation signal. This design feature is not addressed in Section 6.2.4. Provide the design details for these monitors for-all applicable containment penetrations.
Response 480.22 The following will be added to the end of the first paragraph of Section 6.2.4.5: " Penetrations which provide an open path between the containant and the outside containment environment are isolated on high containment radiation as well as a CIAS."
i Question 480.23 Provide an evaluation of the plant design- features that will minimize the time required to keep open the 4-inch on-line containment pressure control system (OLCPCS) CIVs. This evaluation should specifically address the areas of excess air leakage from pneumatic systems, heat loss from the primary and recondary systen, to the containment air, and excess airborne activity levels in the containment atmosphere. Also, provide the design basis for operation of the OLCPCS, including estimated integrated time during a year when this system is expected to be open.
Quantify the additional radiological consequences of this system being open (before isolation is completed) during a design basis accident (DBA) and evaluate these doses against 10CFR100 requirements to conform with the guidance of BTP CSB 6-4, Revision 2.
Response 480.23 Excess air leakat . rom the instrument air system is minimized by utilizing all-welded air supply headers and piping, and leak testing the system. The use of soft seat valves, and capping purge valves and safe vent points also minimize air leakage from the instrument air system.
Table 9.4-3 of CESSAR-DC references the RCS Insulation Heat Loads, heat load per component (BTU /hr) and heat load per plant (BTV/hr), for the reactor vessel closure head, reactor vessel and bottom head, pressurizer, steam generator (primary head only), reactor coolant pump casing, and reactor coolant system piping.
The Containment Cooling and Ventilation System discussed in CESSAR-DC Section 9.4.6 is designed to maintain acceptable temperature limits inside containment to ensure proper operation of equipment and controls during normal plant operation, normal shutdown and for personal access during inspection, testing and maintenance. The containment recirculation cooling subsystem functions during normal plar.t operation to maintain a suitable ambient temperature for equipment located within the containment. The control element drive mechanism cooling subsystem functions during normal plant operation to maintain a suitable air temperature around the rod drive mechanisms. The containment air cleanup
-subsystem operates before and during personnel entries to reduce airborne radioactivity. The cavity cooling subsystem functions to maintain a suitable air temperature in close-ended cavities. The low-volume purge is a pressure relief system used to relieve containment pressure during start-up or shutdown. The low-volume purge and recirculation filters reduce radioactivity during normal operation. Utilization of the recirculation filters to reduce airborne activity levels during normal operations minimizes the number of releases required from the low-volume purge. The chilled water system is utilized instead of station service water for cooling thus lower containment temperatures and more effective heat removal capability are obtained thus minimizing the releases required from the low-volume purge for pressure control.
The estimated gaseous releases from plant sources during normal operation, including anticipated operational occurrences are shown in Table 11.3.6-1. This includes the release from the low-volume purge l which assumes a continuous low-purge exhaust rate of 12.5 SCFM or 1250 SCFM release rate with the system operating 1% of the time as referenced
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71niCESSAR-DC 11.3.6-1.: Gaseous effluents are released through the common
. plant:yent with-the-exception of relatively small quantities released through the turbine-building: vents. ' Atmospheric- dilution; and relative-
' disposition factors-at the worst case locations for the respective-pathways are assumed in this analysis =and will become minimum design criteria 1to be verified during the site selection and documented in the Environmental- Report. = The offsite doses associated with the annual
' releases assuming the dilution factors in ll.3.6.3 resultiin postulated- ,
maximum individual. doses shown in = Table 11.3.6-2. Population doses-resulting from radioactive gaseous releases are related to site characteristics such as-population distribution and wind rose-data; therefore, population dose projections are deferred.to site specific environmental reports.
The total doses to~ an individual offsite, following a LOCA, are given-in
-Table 15.6.5-1 of CESSAR-DC. The total doses'to a control room operator are given in Table 15.6.5-2. The limiting criteria for offsite doses ~ are given in;10CFR100. These are 25 rem to the wholebody and.300 rem to the thyroid.1-These-apply to '+cth the two-hour dose at the exclusion area
- boundaryf and .to.the _ thirty-day dose at the -low population zone. Table 15.6.5-1 shows that the.offsite. doses are within the criteria limits.
The limiting criteria for control. room. doses are given in NUREG-0800, SRP:
15.<4 Rev. 2, " Control Room Habitability System", 1981. Those limits are 5-rem gamma to the wholebody,.30 rem to-the thyroid and 30 rem betaLto.the skin.. Table 15.6.5-2 shows that the control room doses are within the criteria limits.
The low-volume' purge containment isolation valves have a closure time of 30 seconds post-LOCA. . This is referenced in,CESSAR-DC Chapter 15 Offsite and Control Room Dose Analysis Post-LOCA. Reopening of containment isolation valves will require deliberate operator action.
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- Question 480.24-
- Provide =CIS design-information<which addresses an< inadvertent: reopening >
- of CIVs and ensures a high reliability-against reopening to conform with the guidance =of GDC 54, SRP 6.2.4, Revision 2, NUREG-0737, and NUREG ' '
0718.- It should be noted that ganged reopening of CIVs and administrative controls for manual isolation valve closure before.
resetting the. isolation: signal may not be acceptable.
Response 480.24-
-The following will be added to the end.of Section 6.2.4.5: . "The design-of instrumentation and ' control 1 systems for. automatic containment
- isolation valves is such that resetting _ the isolation l signal does not.
result-in the automatic-reopening of containment isolation' valves.
Reopening _of containment isolation valves requires deliberate o)erator action to open valves on an individual containment penetration Jasis."
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Question 480.25 You have stated that the combustible gas control in containment (CGCC) section will be provided later. Provide CGCC information to permit the staff to perform an integrated review.
Response 480.25 CESSAR-DC Section 6.2.5 has been provided in Amendment 1.
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Question-480.26 *
- Provide "a -comparison of the design- basis containment-leak rate '
listed.in Section 6.2.6 (0.5 weight per cent per day) with~the
, containment-leak rate assumed by CE in performing;the LOCA site boundary doses in accordance with the requirements of 10 CFR 100. -
Justify any non-conservative differences.
Resoonse 480.26 The design basis containment leak rate was changed in Amendment I to 0.34 volume percent-)er day-to be consistent with leak rate assumed in performing tie LOCA site boundary doses in CESSAR-DC Section 15.6.5.-
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- 0 Q Ouestion 480.27 You have committed under Section 6.2.6 to conduct Type A and B tests in accordance with 10 CFR 50 Appendix J. However, you have only stated that Type-C tests are " described in 10 CFR 50 Appendix J." Clarify that Type C tests will be conducted in accordance with Appendix J or provide justification for any deviations from the requirements of Appendix J.
Response 480.27 Part C of Section 6.2.6 has been modified to read as follows:
"C. Type C Tests Tests to detect and measure containment isolation valve leakage in accordance with Appendix J of 10 CFR 50."
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Ouestion 480.28
' Provide an evaluation for conformance with the review guidance of SRP--
6.2.6: Revision 2, concerning the instrument lines that penetrate containment during the containment integrated leak rate test (CILRT). ;
Response 480.28' >
During the containment -integrated-leak rate test (CILRT) the following instrument lines'will be open to containment for the following reasons.
A, Instrument' penetrations for all four channels of safety related-pressure measurement will beiopen to containment with the spray actuation signals administratively defeated, since pressurization of :t containment for the test would normally generate a containment spray -
signal. This will function to leak test these' lines and their associated instruments, which by design remain open to containment during and after an accident. During outages in which no CILRT is performed'these lines and associated instruments will be-Type C tested =for integrity utilizing the instrument penetration valves.
Following the Type C tests, these valves will-be returned to normal open position by administrative controls.
B.- The ILRT Pressure Sensing Line will be utilized for pressure - 4
- measurement during a containment-integrated 11eakirate test. This cwill function to perform a' Type A test on this instrument
. penetration. During outages-in which no CILRT is performed, a Type C test-will be performed on this instrument penetration.
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Ouestion 480.29 Identify all systems that will not be vented and drained during the Type A CILRT for conformance with the guidance of SRP 6.2.6, Revision 2.
Provide justification for not venting or draining these systems.
Response 480.29 Table 6.2.4-1 has been added to CESSAR to provide a listing of all containment penetrations. In the table, all penetrations are addressed with respect to testing and venting / draining requirements. Justification for not testing or not venting and draining specific penetrations is I provided in the notes accompanying the table.
I II
o i Qysition 480.30 List all containment penetrations which will be subjected to Type B tests for conformance with the guidance of SRP 6.2.6, Revision 2. Justify the a exclusion of any penetrations from such testing.
Response 480.30 Table 6.2.4-1 has been added to CESSAR to provide a listing of all containment penetrations, in the table, all penetrations that require a Type B test are identified under the aegis of NOTE 11. In addition to these penetrations, all electrical penetrations will be Type-B leak rate tested.
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Question 480.31 List all containment isolation valves that will be subjected to Type C tests for conformance with the guidance of SRP 6.2.6, Revision 2.
Justify the exclusion of any isolation valves from such testing.
Response 480.31 Table 6.2.4-1 has been added to CESSAR to provide a listing of all containment penetrations which are Type C tested. Justification for not tecting or not venting and draining specific penetrations is provided in the notes accompanying the table.
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Ouestion 480.31 Clarify whether external hydrogen recombiners or equivalent are included in th6 containment design. If so, include them specifically in the CILRT for conformance with SRP 6.2.6, Revision 2 guidance.
Response 480.32 Two containment penetrations have been included in revised Table 6.2.4-1, which has been added to CESSAR-DC, for addition of external hydrogen recombiners. These penetrations will be included in the Containment Integrated Leak Rate Test (CILRT).
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Ecestion 480.33 Identify any potential bypass leakage pathways and include them l specifically in the Type C local leak rate testing program for 1 conformance with the SRP 6.2.6, Revision 2 and BTP CSB 6-3 guidance.
Resoonse 480.33 A-review of containment penetrations and potential bypass leakage pathways was conducted in formulating Table 6.2.4-1 of CESSAR-DC, Amendment 1.
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Enclo3uro II to
<< LD-91-014 PROPOSED REVISIONS TO 7tfE COMBUST 1^W E!1GINEERING SCHEDULE SAFEi'1 ANALYSIS REP (XW - DP (IGN CERTIFICATION
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CESSARin an* -- #
q Nl0 *b Y 9.1.2 SPENT FUEL STORAGE 9.1.2.1 Design Bases Tho - following - design bases aro imposed on the storage of fuel-within the spent-fuel pool:
A. Accidental' criticality' shall be prevented for the mont i l
reactive' arrangement of funl stored with optimum moderation E by avoiding a K- greater than 0.95. This design basis shall be-met und8[fany normal or accident conditions.
B. The requirements of' Regulatory Guide'1.13 shall be met.
-C. The ' storage racks and facilities shall be Solsmic category I, D. Storage shall be provided for up to 907 spent fuel y assemblios.
9.1.2.2 F_acility Description 9.1.2.2.1 Spent Fuel Pool E
The spent-fue1 pool is a stainless steel lined, concrete walled pool that is-an-integral ~part of the fuel building.
9.1.2;2.'2 Spont Fuel Pool Storage Racks The spent fuel pool storage racks are mado up of twelve 11x11 individual' modules _containing 121 storage cells each (sco Figures I
9.1-21 and 9.1-22). A module is an array of fuel storage' cells similar to that- shown in Figuro 9.'l-1. The storage racks are
-stainless l steel honeycomb structures with rectangular fuel storage-cells._ The stainless stcol construction of the racks is compatible with ' fuel _ assembly materials and.- the spent fuel. E borated water environment.
A single pitch of 9.780-inches is provided for all of the racks it, the pool. Tho' spent- fuel is stored in two regions of the pool. Region I provides -- core off-load capability for 363 spent fuel assemblics-(equivalent to one and one-third cores). Thisislg' . l achieved with_50% density storage in a checkerboard. array using "L" inserts-in the usablo cells (Figure 9.1-2)._ _ The "L" insert is a non-poisoned- stainloss steel insert which provides the E needed flux trap water gap. Region II provides 75% density storago-for 544 spent fuel assemblics. The cells that are not used are blocked to prevent improper storage. A total 'o f 907 I I- usable spacen for spent fuel storage is thus provided.
'f 9.l.'2..*L.'1- .
Amendment I 9.1-5 December 21, 1990
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'.Id e rt 9.1. 2. 2. 2 A fuel assembly may be stored in Region 11 only if it has the minimum burnup required for an assembly of its initial enrichment. The Owner-Operator will develop and implement administrative controls to permit storing a fuel
, assembly in Region 11 only if it meets established burnup versus initial enrichment requirements. -
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CESSAR mRicmon
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- c. Transfer ::yatom Upendor Rotation Interlock l This ;nterlock prevents rotation cf the re ender while the ref.neling machine and spent feel handling machine (SFitM) are 1 I
at their upender stations.
Failu.o of this interlock while the machi.ios tro at the upending station-will allow the transfer equipment operator to initiato rotation of the fuel carrier. In the event that ,
this signal is erroneously initiated while the fuel assembly .
is being lowered from or raised into the refueling machino, a bonding load would be applied to the fuel bundlo.
D. Transfer System Upondor Interlock This interlock provents rotation the upondor unicas th I fuel c.strier.is correctly located ar uponding.
Failure of this interlock will:
- 1. With the fuel carrior in the transfer tubo allow tho upondor to rotato with no offect on the carrior or fuel ,
bundlo. '
'2. With the fuel carrior partially in.the upondor, attempt to but not be successful in, rotating the carrior sinco a mechanical lock provents prematuro carrior rotation, i
'E. Puol N rrier Rotational Interlock This incerlock provents rotation of the fuol carrior unless ;
. the fuel carrior is correctly locatod in the upondor. ,
Tallure of-this interlock may cause contact between the fuel carrior and the transfer tube. assembly which will result in an overload signal and termination of motion of the transfer "
carriago. No dan: age to -the fuel assembly will result since the fuel assembly is onclosed in the carrior.
9.1.4.2.1.3 Spent Puol liandling Machino C The spent fuel. handling machine will be a refueling machino adapted for uso in tho spent fuel pool aroa'. It will contain tho l same interlock featuros as described in Section 9.1.4.2.1,1, except 301 noted below for the Spont Fuol llandling Mrchino Translation Zono Interlock
.A. Zono interlocks protect against running the load into walls
.or the gate of the storago area. I h'
Amendment I 9.1-23 December-21, 1990
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CESSAR E!n%mou 6 W '
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B. The equipment will be assembled and chocked for proper functional and running operation at the shop and prior to using the equipment. E C. Inspection and maintenance will be performod in accordance with plant maintenanco procedures.
9.1.4.5 Instrumentation iloquiremonta The refueling system instrumentation and controls are described in Section 9.1.4.2. No credit is taken for instrumentation or interlocks on components of the fuel handling equipment to either prevent or mitigate the consequences of the postulated accident.
Thus. safety-related interlocks are not provided.
9.1.4.6 Operating Proceduro Guidelinos Sito-specific guidelines will be established for component '
handling procndures and plant operating procedurco. Component ;
handling proceduro guidelinos will requiro-the owner-operator to-establish the safe load path for lifting heavy lands and to perform opocial handling component inspections prioc to lift.
plant operating proceduro guidelines will require appropriato -
operation training and crano inspections. f)4 gg
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i Amendment I 9.1-40 December 21, 1990
CESSAR Ela% mon G W 'R
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fuel racks contain fuel assemblics. The now fuel handling holot is rectricted from movement over the spent fuel storage aron when the spent fuel racks contain fuel assemblico.
In accordanco with the regulatory position of Itagulatory Guido 1.13 and General Design Critorion 61 of Appendix A to 10 Crit 50, the hoints are also rostricted from panning over tho spent fuel i g pool cooling cystem for 1:SF systemo which could be damaged by dropping the load.
Set pointo for the hoist interlocks are act to provent falling or tipping of the loads into the fuel storage areas.
Administrative } trolu
. he con &Vproclude Ay AAJ-movement OA- Oh6 of heavy loads within the containment building pool when the refueling machino contains a fuel assembly. During heavy load movement, the fuel trancfor tube valvo is closed to avoid water level changes in the fuel building during postulated accident conditions such no dropping the heavy load on the reactor vencol pool seal.
9.1.4.3.2 Puol llandling f A failure modes and effecto analynis is doncribed in Tabic 9.1-2.
Direct voice communication between the control room and the refueling machino concolo in available whenever changen in core geometry are taking place. This provision allows the control room operator to inform the refueling machine operator of any impending unsafe condition detected from the main control board indicators during fuel movement.
operability of the fuel handling equipment including T.h c bridge and trolley, the lif ting mechanicmo, the upending nachincs, the transfer carriage, and the annociated instrumentation and controls in accured through the implomontation of preoperational n testa and routines. Prior to the first actual fuel loading, the equipment la cycled through its operationn using a dummy fuel assembly. In addition to the interlocks described in Section 9.1.4.2.1, the equipment has the following opecial featuren:
A. The major cyntoms of the fuel handling cyctem are electrically interlocked with each other to accist the operator in properly conducting the fuel handling operation.
Failure of any of these interlocks in the event of operator error will not result in damage to more than one fuel assembly.
B. Miscellaneous special deuigr. featuren which facilitate i handling operations include:
e Amendment I 9.1-37 December 21, 1990
CESSARnnL m Q %.n
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are required, they may be introduced into the upper guido structure at this time. The exponded CEAs are moved to the CEA ,
elevator, adjacent to the upper guido structure storage aton, j whero the upper CEA casting is removed from the CEA rods ;
utilizing special tooling. Each rod is picked up individually y' and placed into the transport container where the lower 15-foot l section is cut off utilizing the portable underwater hydraulic CEA cutter. The upper 5-foot section of the CEA rod is then placed into the transport container and the operation is repeated until all rods have boon cut. The transport container is then moved to the transfer carriage where it is transported to the spent fuel building for CEA rod disposal.
At the completion of the refueling operation, the fuel transfer tubo valvo is closed. The upper guido structure is reinserted in the reactor vessel, the CEDM extension shaf t assemblics and CEAs i
aro loworod into position, and the lift rig is removed. The water in the refueling pool is loworod to the top of the extension shafts. The reactor vessel head is then loworod until the CEDM extension shaf t assemblies are engaged by the control clomont drive mechanism nozzle funnels. Lowering of the head and the water level is continued until the head is scated. The remainder of the refueling pool water is then removed. Than tho /
studo are installed, the head is bolted down, and the transfer tubo penetration sloovo is scaled. The ICIs are reinserted into ("
the core region and reconnected to their cabling.
The head area cable tray is replaced, CEDM and ilJTC cabling is connected, cooling ducts are reconnected to the CEDM cooling manifold, and the vessel vont piping is installed.
9.1.4.3 Safety INaluation 9.1.4.3.1 Puol Building Overhead Crancs and Containment Polar Crano The containment polar ctano, the cask handling hoist, and tha fuel handling hoist are designed to provent the drop of a heavy load such as the reactor vossol head or the spent fuel shipping cask. In addition, prodotermined load paths for major lifts (sco Figuros 9.1-19 anri 9.1-20), operator training, and regula maintonar.co minimize the possibility of load mishandling.(@r M crano 9./.'t.3.1)
Limit switches, electrical interlocks and mechanical interlocks pre" mt improper crano operations which might result in a fuel handling accident. This is also discussed in Section 9.1.4.2.1.7. The spent fuel cask handling hoist is rostricted from movement over the now and spent fuel storage areas when the Amendment I 9.1-36 December 21, 1990
Insert 9.1.4.3.1 The Owner-Operator's operating procedures will control the load paths and height of the reactor vessel closure head, the core support barrel and the upper guide structure above the pool floor.
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CESSARHnh a G y / o,77
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9.2.9.2 System Description The CWS are closed loop, chilled water systems, divided into two E
. subsystems: the safety-related Essential Chilled Water System (ECWS) and the non-safoty- related Normal Chilled Water System (NCWS). The ECWS subsystem is made up of two equally sited divisions. Each division is totally indopondant and separated both mechanically and electrically; thorofore, singlo failuro or '
lack of separation can not render ECWS inoperablo. The ECWS is 3 an engincorod saaoty featuro (ESP) which can be actuated manually or automatically te furnish essential chilled water. The heat exchanger pump or the chiller pump from one Division is normally in operation. The essential chillor is enorgized when required by a riso in the ECWS water temperature. -The NCWS subsystem is comprised of two 100% capacity divisions. Each NCWS division is comprised of two 50% capacity chillors. Figuro 9.2.9-1
' illustrates the chilled water system configuration.
9.2.9.2.1 ECWS y, pad L, 1
Each 100% capacity division is comprised of a chilled water refrigeration unit, a circulating chilled water) pump, control valves, instrumentation, and piping. A makeup ator line to the C ECWS is connected to the domineralized water system, the normal source of makeup. In caso of a loss of emineralized water, makeup is supplied from the station servic water system, via a Seismic Category I assured water lino. A 1. ...u/ S spool picco is placed in this line to provent intrusion-of raw water into the y clean, chemically treated system during normal operation.
L The ECWS equipment design roquirements are as-follows:
A. The system is designed to provido - a suf ficient quantity of chilled water to moot the cooling load domands of the full load of the essential HVAC chilled water coils at a normal 45'F water temperature from the refrigeration unit and a maximum of 10'F AT across the refrigeration unit.
B. The ovaporator tubos and the condensor tubes of the refrigeration unit are designed to inplude an allowance for E tube fouling of 0.0005 hr-ft *F/ Btu and 0.002 hr-ft'7 *F/Dtu, respectively.
C. Components of tho-system are designed in accordance with the Soismic Category I and Class 1E-requirements.
l D. Each refrigeration unit along with its pump, compression tank, and control valvos is physically seperated from the 3 m
( other refrigeration unit (s)
Amendment I 9.2-67 December 21, 1990
- s o CESSARneibuos g gjg, g g-1 9.4.2.3 Safety INaluation g
The Fuel Building Exhaust System is an engincered safety teaturo.
Each redundant filter train (two 100% capacity) , fan, and motor operated damper is served from a separato train of the omorgency l1 Class lE standby power. This assures the integrity and availability of the Exhaust System in the event of any single activo failure.
Air exhausted from the fuel handling area is monitored by a h radioactivo gaseous detector sampling the air in the exhaust duct
[ header betwoon the fuel handling area and the inlet to the filter trainTl4 Indication of radioactivity ubove allowable limits will i',
automatically divert the flow of air through the filter trains prior to discharge into the atmosphoro through the unit vent.
Additional monitoring of exhauct air in provided in the unit vent.
The 100% exhaust air syntom is manually set to the filtered mode 3
during all fuel handling operations.
The Fuel Building Ventilation Exhaunt ayatom is available '
following a loco of offsite power; however, fuel building supply l g will not be available.
9.4.2.4 Inspection and Tenting lloquirementu o
Performance characteristics of the Fuel Building Ventilation li System will be verified through qualification testing of %
components as follows: , i A. Encontial equipment, fans, dampers, coils and ductwor); will ,g be tested in accordance with ASME/A11S1 AG-1-1988. ,/
One fuel building supply fan and one fuel building exhaust B. I fan in tented in accordance with AMCA standards to assure fan characteristic performance curves, one of each type of essential cooling fan will also be tested in accordance with AMCA.
C. licating and cooling coils are le a);od- te c ted with air, or hydrostatically, to assure integrity. Coils are rated in accordance with ARI standards.
D. !! EPA filters are manufactured and teated prior to installation in accordance with MIL-F-51068. !! EPA filters will be tested in place after initial installation and periodically thereafter to verify filter integrity.
O Amendment I 9.4-13 December 21, 1990 W .
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Insert 9.4.2.3 The radiation detectors are located to ensure that dampers will have completely actuated + direct exhaust flow through the filter trains before the first airborne , vioactive material reaches the bypass dampers.
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- e CESSAR8n h ou g Qjg, c79 9.5.9 DIESEL GENERATOR DUILDING SUltP PUMP SYSTIM 9.S.9.1 Denign 11anen The Diosol Generator Building Sump Pump System is doaigned to remove leakage and equipment drainage from the diosol generator building and to protect the diocol generator units from internal flooding caused by the maximum credible pipo rupture in the Dional Gonorator Building.
1 All components and piping are located within a Soismic Category I otructuro (diocol generator building) and all ossential componento are fully protected from floods, tornado missilo damago, internal missiles, pipe breaks and whip, jet impingement and interaction with non-seismic syntoma in the vicinit/.
9.5.9.2 _Syntnm_Dencription l
Two cump pumps are provided in each diesel generator building.
The pumps are located in the pit below the lubo oil cump tank. ;
The sump pumpa start automatically on high nump water level and l[ transfer the water to the equipment and floor drain syntom.
The diocol generator building numps and sump pumps are designed for a constant inflow rate of 75 gpm with a maximum pump cycle j time of throo ntarts per hour (one pump operating with 37.5 gpm inflow). The maximum pumping flowrate with both pumps operating is 150 gpm. -T'^ rite-reccific "/ " chn11 v&if y that thic '
adequai.u ior tne,mw4 mum-4eahge er m -:irum cred4b4W wruptury 1~""
in-the dicaul generate h"i1 ding.
9.S.9.3 Safety Evaluation The Diccol Generator Building Sump pump System is an ANSI Class 3 piping system and the pumps and cystem components are designed in accc rdance with the requirements of the ASME Doller and Precouro Vensel Codo,Section III, Clans 3, E 9.5.9.4 Inspection and Tenting Requiremonta System components and piping are tested to pressures decignated by appropriate codos. Incpection and functiona.' tenting are performed prior to initial operation; thereafter, equipment not in continuous use in subject to periodic testing and vinual inspection.
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Anendment I 9.5-77 December 21, 1990
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CESSAR En!i6os pg , y
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11 . After the reheat pressure has bled down, the unit is running a few rpm above rated speed ready to be synchronized. t In case of malfunction of any portion of the first line of defense against overspeed (speed control tn control and intercept valves) when load is lost, the turbine will accelerate to the r -ir speed where the overspeed trip will activate. This will directly trip the main and intermediate stop valves (second line of defensei, and the disc " imp valves of the Control and Intercept Valve actuators will also be tripped. Subsequently, the turbine will coast down to zero speed.
The Trip and Monitoring System will initiato appropriate action on abnormal operating conditions and indicate the existence of dose conditions to the operator.
In addition to the inte' amted turbine trips, any of the following externally trip inputs will result in remeving the hydraul e .luid pressure from the emergency trip sya cm (ETS). The removal of this pressure will result in rapia c)osure of all turbine valves.
Externally generated trip inputs are
( A. Low Condenser Vacuum.
B. Thrust Bearing Failure.
C. Low Bearing Oil Pressure.
D. Internal Fault in Generator.
E. Generator Breaker Failure.
F. Reactor Trip.
G. Loss of Generator Stator Coolant h'ithout EllC Runback.
11 . Steam Cenerator Hi-lii Level.
I. Safety Injection.
J. Both Main Feedwater pumps Tripped.
K. liigh Exhaust flood Temperature.
L. Manual Turbine Trip.
!1. Turbine Oil Fire Trip.
Amendment E 10.2-11 December 30, 1988
1 TABLE 9.2.2-2 (Cont'd)
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(Sheet 2 of 2)
COMPONENT COOLING SYSTEM SINGLE FAILURE ANALYSIS Inherent Ccepensating Components Failure Mode /Cause Effect on System Method of Detection Provision i Piping loss of pump None- redundant icop/ Flow indication and Redundant loop /dtvision J (pipe discharge / header / division is available pressure alarm in is provided breaks) Ifnebreak or the control room
, c:echanical damage 4
Non-essential Valve to pump None-redundant Flow itidication and Redundant division is j
header suction header division is available pressure alars in provided stays closed / the control room operator error Piping Loss of return None--redundant loop / Flow indication and Redundant loop / division
- (pipe heade'r/linebreal or division is available pressure alarm in is provided breaks) crechanical damage the control room l
Non-essential Valve to header None pumps are Flow indication in Equipment sized to prevent
- and fuel falls to close/ sized to prevent control room flew degradation pool header mechanical or electrical failure flow degradation p y
c- , - ., v..- w.,, .ei n- ~ua a s.wu-aw na m ~c.u ao. ,' ,. .
Spray He.d McMai<*1 or din h fec ck cs (% pdic ,Hcm in 13 - ,ed 6 % e.s b A Erk m ,.*g e v,r r e eleciritsl Elu t r E*cd egyr is sited g ,,gg g g, m gq ,,m 'p o.ess
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Heat Excka<,er- WaectowtcaJ di=ision flaa in/fcefica la i5 P' "'
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- o frol e s V&s % turs k <- a v - _ - - _ _ . _ _ _ - - - _ . - - - _ _ _ _ . _ _ . - _ - _ - _ _ _ _ _ _ - _ _ _ _ - _ _ . - - _ ____J
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"CESSAR 8l'%nem Q y g d. I 6'IS0,28
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6.2.4.1.2 Design Features '
The following is a summary of Containment Isolation System design features. Incorporation of those feature's into the Containment Isolation System results in a design where the design critoria for containment isolation barriors given above are mot:
A. Containment isolation valvos and interconnecting piping are designed and constructed to Safety Class 2 and Soismic Category I standards as defined in ANSI N18.2-1973 and Regulatory Guide 1.29, respectively.
B. The design pressure and temperature of all piping and connected equipment comprising the isolated boundary is than nd temperature of the greatercontainment. ,the design
.gwt (. .:tpressuro 9 . /, ;L B C. Containment isolation valves and interconnecting piping are protected against missiles.
D. Containment isolation valvos and interconnecting piping are protected against the offecte of pipe whip and jet impingement.
E. The maximum allowable particle sizo entrained in water taken from the containment sump is limited. This ensures that the proper operation of ESF systems and CIS valves will not be inhibited by debris introduced into the containment g following a LOCA.
F. Containment isolation valvos are designed to operato under normal environmental conditions and to fulfill their safety related function under post-accident environmental conditions, consistent with the requirements of Section 3.11.
G. Containment isolation valvo and associated penetratior.
piping are qualified in Section III of the ASME Code, as Class 2 components, as described in Section 3.9.3.
H. Maximum allowable actuation timos are imposed on containment isolation valvos consistent with their required safety h function and ANSI /ANS 56.2-1984.
I. Valve operators and power sources are selected for containment isolation valves consistent with their required g safety function.
3 nSe d fg,L,f,g f Amendment I 6.2-37 December 21, 1990 L .- . . - _ -
- n Insert 6.2.4.1.2.8
- Relief valves for providing overpressore protection from heat up between closed containment isolation valves have a relief set point equal to the design pressure of the piping."
Insert 6.2.4.1.2.J Instrumentation and control sensing lines which aenetrate the containment are provided with containment isolation provisions watch meet the intent of Regulatory Guide 1.11.
l
.. n l
CESSAR !Mncuio C Qpg,/ g *
. i
(
Fluid lines which must remain open subsequent to a design basis accident, such as lines serving ESP systems, do not have containment isolation valves that are automatically closed by the CIS or-MSIS. Each of these penetrations has a minimum of one remote-manual operated isolation valvo outsido containment.
Provisions nado to assure the operability of the isolation valvo system under an accident environment satisfy the requirements for _ redundancy, independenco, and testability. The valving system is designed for pressures equal to or greater than the containment design pressure. A comprehensive testing and inspection program assures that thoso components will operato for the timo period required in the post-design basis accident conditions of temperature, pressure, humidity, radiation, or solomic phonomena. The proper doign basis accident environmental conditions are listed in the design specifications for all components that are part of the Containment Isolation System.
Vendor factory testing is performed on a prototype of those components to assure their adequacy under those conditions.
Air or motor-operated valves are used for the automatic isolation valves. Air-operated valves are designed to assumo the position of greater safety upon loss of air. Motor-operated valves are
( poworod from the emergency power sources.
Remoto manual control of the automatically actuated cc.7tainment isolatir.n valvos is-provided.
Automatic valvos are installed in lines that must be immediately isolated after an accident. Those lines which must remain in service after an accident have at least one remoto manual valvo.
The intogrity of the isolation valvos system and connecting lines, under the dynamic forces resulting from inadvertent closure whilo at operating conditions (e.g., main steam lines) is assured by the performance of staria and dynamic analysis on tho
. piping, valvos and restraints. c The supports and rostraints are applied such that integrity is assured and pipe stresses and support reactions are within allowablo limits. Valves, in nonsafoty-related systems where function permits, are normally positioned closed to minimizo any release following a - design basis : cvont are equipped with valvo operators to move the valve rapidly.
Contain. t isolation val o and operato , are design. to withs id a maximum in Jrated radiatio dose of 4.0FL rads dur g the life of th- )lant. v k_
l Amendment E
,- 6.2-40 December 30, 1988
- 4. r CESSAR En!%- /D, //,g g ,/ g
(
Containment olation valve that are loegdd inside the*
containman are desi .
to funct, ion under ho pressuro-_t mporaturo cond ons of both norpd1 operation a hat during .the design bp a ovent. Tp(' pressure-to aturo condi on used for v Ivo doign under ormal operatio in 14.7 pai and 130*F. pressure-te ature conditio. used for v o design und accident co tions is give in Section
.2.1.
6.2.4.3 Safety Evaluation The containment structure and the containment penetratir .c form an ossentially leat tight barrior. Allowablo leak rates from the containment under design pressure condition are discussed in Section 6.2.1. Testing provisions and performance are also discussed in Section 6.2.1. Whenever practicable, isolation valvos outside containment wh.ch are normally open and required to clouc on a signal to isolate the containment are deolgned to fail clocod.
In order that no single, credible failure or malfunction will result in loss of isolation capability, the closed piping systems, both inside and outside the containment, and various types of isolation valves provide a doubic barrier.
(..
The isolation valve and actuators are located as close as practical to the containment and protected from missilo damage.
This minimizco the potential hazards that could be experienced by c the system.
The integrity of the isolation valve system and connecting lines under the dynamic forces resulting from inadvertent closure under operating conditions is assured, based upon required static and dynamic analysis.
The supports and restrainto are applied such that pipe stresses and support reactions are within allowable limits as defined in Section 3.9.3.
Although the liigh Volume Containment Purgo Sub-system isolation valves (supply and exhaust) are required by Technical Specifications to be closed when the Reactor Coolant System exceeds hot shutdown conditions, a CIAS is provided to them to further assure closure. The Low Volume Containment Purgo I
Sub-System and Containment Pressure Control System isolation valves also receive a CIAS signal. Diversity of the CIAS signal is provided sinco a CIAS occurs on containment high pressure or upon roccipt of a SIAS. In addition, penetrations which provide
, an open path betwoon the containment and the outside containment i
Amendment I 6.2-41 December 21, 1990
(
2 n CESSAE Ufecamu 9#"
- A h Ah $o - 6
( S && Ad 1Ocfte R 4,% A y 7.
The isolation arrangement of the fuel transfer tube consists of a transfer _ tubs / closure and a blind flange, enclosing the transfer tube. (The b1i.id. flange contains two 0-ring grooves and a pressure tap which runs through the blind flango to the annulus between the two 0-rings. When assembled preparatory to reactor operation, the blind flange is bolted to the transfer tube closure and tM annulus between the seals is pressurized to ensure tha' ch seals are functioning. The seal is further tested whea 9t pressure is introduced into the containment.
When these .ests have been satisf actorily completed, the fuel
' transfer tube is isolated from the containment. The transfer tube closure and the blind flango are considered to be the g containment boundary and, therefore,_ General Design critorion 56 does not apply to the transfer tube penetration and an isolation valve is not required.
A normally locked-closed manual valve will be provided on the transfer tube outside the containment. liowever, its basic function is not to provide containment isolation. At the beginning of refueling, during filling of the refueling pool, ,
this valve is maintained closed until a common water lovel is reached in the refueling pool and the spent fuel pool. Then the valve is opened to allow the transfer of fuel.
C.
The equipment hatch consists of an arrangement similar to the fuel transfer tube closure. It has a double seal which is Type B leak rate tested in accordance with 10 CFR 50 Appendix J. Since it is considered _part of the containment boundary, General Design Critorion 56 does not apply and an isolation valvo is not required.
The Integrated Leak Rate Test pressurization ponctration concists of an insido containment blind flango and an outsido containment manual isolation valvo. This line is utilized to pressurize containment to perform the Containment Integrated Leak Rate Test.
During outages in which no Containment Integrated Leak Rate Test 7 is performed, this ponctration is Type C Icak rate tasted in accordance with 10 CFR 50 Appendix J. The blind flange is scaled' with the manual isolation valve locked closed during- normal cporation.
, Electrical per.*trations consist of a gas-scaled or a double l
O-Ring sealed electrical penetration assembly. Both types are
- l. Typo B_ leak rate tested in accordance with 10 CFR 50 Appendix J
! testing _ requirements by pressurizing the test volume to DBA t
pressure and ' measuring olther pressure decay or volumetric flow rate required to maintain test DBA pressure. Electrical
/.
i penetrations are further tested when containment is pressurized for the_ Containment Integrated Leak Rate Test.
Amendment I 6.2-39 December 21, 1990
, _ _ _ ~ , _ . _ . . _ _ _ _ ,_ ,_ . _ , . _ _ _ . . _ . _ _ _ _ . - _ . _ _ _ .
G r CESSAR Eininema ggp/gy or by computation. The test volume is pressurized to P or Design Basis Accident pressure. The test volume press 0fo, I is recorded at 15-minuto intervals for a minimum of one hour. The leakage rate is computed using the following equation:
(At) P 3
Where: Ly is leakage rato (ft /sec)
VT is the test volutto (ft )3 At is the timo interval over which the pressure decay is recorded (sec)
/. P is the change in pressure from the initial pressure at the start of time interval At (psig)
P is the initial pressure at the start of time interval At (psig)
(( This method assumes that the temperature of the test volume g remains constant throughout the test.
B. Method 2, Air Flow The tnst volume is established by closing the appropriato isolation valves. This method does not require the determination of the volume to be tested. The test volume is pressurized to P or Design Basic Accident pressure, y using an air flow al,egulator until pressure is stable.
Pressure and air flow are then measured and recorded.
6.2.4.5 Instrumentation Itequirements Containment isolation will be initiated by means of a Containment
! Isolation Actuation Signal (CIAS). A CIAS occurs on containn.cnt high pressure as sensed by two out of fo r containment high pressure censors or upon roccipt of an SIAS. se d 4 2.t./ E The instrumentation circuits that generate 1AS are described in Section 7.3. The inclusion of the Main Steam Isolation Signal as a CIAS is not necessary as the radiological releases due to a steam line break are within acceptable guidelines. Main steam and main feedwater isolation are initiated by the Main Steam Isolation Signal as described in Chapter 7. )
1, 1
.Tnserh dc24.D R Amendment I 1
l l 6.2-43 December 21, 1990 l l
e- .c l
Insert 6.2.4.5 i Penetrations which provide an open path between the containment and the outside containment environment are isolated on high containment radiation es well as a CIAS."
Insert 6.2.4.5.X The design of instrumentation and control systems for automatic containment isolation valves is such that resetting the isolation signal does not result in the automatic reopening of containment isolation valves. Roopening of containment isolation valves requires deliberate operator action to open valves on an individual containment penetration basis.
4 l
tc c L CESSAR !!an. 4 y eo.2.s i C. Type C Tests L,
l Tests to detect and measure containment isolation valvo i leakage in accordance with Appendix J of 10 CPR 50. ,
- 6.2.6.1 Containment Integrated Imak Rate Test 1 1 The containment leakage rato for the System 00+ containment must 1
be proven lower than-0.34% volumo por day. During the test, the containment is isolated _ and pressurized in accordanco with
, Appendix J _of -10 CFR 50. When test . pressure is reachod, the containment is _ isolated from its pressure source and. the following parameters are recorded at periodic intervals:
! A. Containment absoluto pressure.
B. . Dry bulb' temperatures, t
C. ' Water vapor pressures.
D. Outsido containment weather conditions.
During the test, vontilation insido the containment is operated - .
as necessary to _ enhanco an even air temperature distribution.
The test data are processed at periodic intervals during the test. E %{..
to dotormino test status and leak-tight confidence level. .If it appears that tho - leakage is excessivo,. the - prosauro plateau is
-either maintained on the _ test _ or aborted to perform repairs.
After a proscribo.timo period and assurance of leak test rato, tho .prosauro is slowly blod off to verify tho loak- rato monsuromont.- This is accomplished by preciso measuromont of - a flow which causos a- change in the weight of air in the containment that is in the same order of magnitudo as the
- allowablo leakage rato. Formulas used .in computing the integrated Icak rate are -based on the formulas frnd in ANSI 1 N45.4, " Standard for Loakago Rato Testing of- Containment Structures for Nuclear Roactors."
Tho test methods for'_ the. parlodic Type A tests are essentially the samo as those used for the preoperational Type A tests.. Any dif ferences in the methods aro ; due- only to minor dif ferencos -in post operational-system alignmonts, o.g.,- the piping betwoon-tho-
- refueling cavity-and the auction of the RW pump cannot be drained with-water in the refueling cavity.
6.2.6.2 Containiment Ponotration Lankago Hato Test
.t Type B Icakage rate tests .aro -performed on all oicctrical, equipment, Land personnel hatch penetrations in accordance with 10 CFR 50 Appendix J.- The test pressure, test frequencies and A / . 7. . d . I Amendment I 6.2-62 Docomber 21, 1990
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~
o.) n Insert 6.2.6.1 During the containment integrated leak rate test (CILRT) the following instrument lines will be open to containment for the following reatons.
A. Instrument penetrations for all four channels of safety related pressure measurement will be open to containment with the spray actuation signals administratively defeated, since pressurization of containment for the test would normally generate a containment spray signal. This will function to leak test these lines and their associated instruments, which by design remain open to containment during and after an accident.
During outages in which no CILRT is performed these lines and associated instruments will be Type C tested for integrity utilizing the instrument penetration valves.- following the Type C tests, these valves will be returned to normal open position by administrative controls.
B. The ILRT Pressure Sensing Line will be utilized for pressure measurement during a containment integrated leak rate test. This will function to perform a Type A test on this instrument penetration. During outages in which no ClLR1 is performed, a Type C test will be performed on this instrument penetration.
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