ML20067D622
| ML20067D622 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 12/20/1982 |
| From: | Longenecker J ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT |
| To: | Check P Office of Nuclear Reactor Regulation |
| References | |
| HQ:S:82:148, NUDOCS 8212210050 | |
| Download: ML20067D622 (101) | |
Text
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Department of Energy Washington, D.C. 20545 Docket No. 50-537 HQ:S:82:148 DEC 2 o 1982 Mr. Paul S. Check, Director CRBR Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Mr. Check:
ADDITIONAL INFORMATION RESULTING FROM DECEMBER 15, 1982, MEETING ON PLANT AUXILIARY SYSTEMS, PRELIMINARY SAFETY ANALYSIS REPORT (PSAR) CHAPTER 9 In accordance with agreements reached between our respective staffs at the December 15, 1982, auxiliary systems meeting, enclosed are responses to questions (Enclosure 1) and associated amended PSAR pages (Enclosure 2) for the following PSAR Sections of Chapter 9, " Plant Auxiliary Systems."
Section 9.1 See responses in Enclosure 1, amended PSAR Section 9.1 pages (Enclosure 2).
Section 9.2 See responses in Enclosure 1, amended PSAR Section 15.7.3.7 (Enclosure 2).
Section 9.4 See responses in Enclosure 1, amended PSAR Section 9.4 (Enclosure 2).
Section 9.5 See responses in Enclosure 1.
Section 9.6 See responses in Enclosure 1, amended PSAR pg. 9.6-1, 4, and 23.
Section 9.7 See amended PSAR Q/R's CS410.18 and 19 (Enclosure 2).
Section 9.9 See amended PSAR Section 9.9 J O
(Enclosure 2). DO Section 9.10 See amended PSAR pg. 9.10-1 I (Enclosure 2).
Section 9.15 See amended PSAR pg. 9.15-2
/ @
(Enclosure 2)
Section 9.16 See responses in Enclosure 1, amended PSAR pg. 4.2-256 (Enclosure 2).
8212210050 821220 PDR ADOCK 05000537 A PDR
2 The amended PSAR pages of Enclosure 2 will be incorporated into Amendment 75 of the PSAR scheduled in January. Additional informa-tion associated with Sections 9.3 and 9.13 will be submitted under separate cover later in December.
Questions regarding these responses may be addressed to D. Robinson (FTS 626-6098) or D. Hornstra-(FTS 626-6110) of the Project Office Oak Ridge staff.
Sincerely, bO tb Jo n R. Longene Acting Director,UOffice of Breeder Demonstration Projects Office of Nuclear Energy 2 Enclosures cc: Service List Standard Distribution Licensing Distribution
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, RESPONSES 10 NRC COMENTS
- 1. Dumnant: Equipment with active cooling (i.e., EVST, EVTM, I FHC, and fuel transf ar port cooling Insert) should include diesel power provisions or otherwise settsfy clad temperature limits for loss of offsite (normal) power as an anticipated t event; the PSAR is unclear with respect to applicability of such a requirement.
Rarponse: Fuel Clad Fallure and subsequcnt fission product release will result in site boundary doses well below ,
established limits as discussed in PSAR Chapter 15.5. !
Cooling loops supplied with backup electrical power by diesel l generator are provided for EVST sodita and FHC argon cooling. [
The fcrced conyoctIon cooling system for the EVTM Is supplied with normal electrical power but is backed by a natural i convection cooling system which can maintain the cladding [
temperature within its limits. The FHC cooling grapple !
I blowers are supplied with normal electrical power. In the l event of an extended loss of power while handling a bare fuel !
a:sembly in the FHC, the cladding might be heated to the i point of f ailure. Fission products released would be !
contained in the FHC because the diesel power-supplled argon j
, circulation system would maintain the FHC pressure negative relative to surrounding areas. Diesel power (from one diesel) is provided to the FHC cooling systems to minimize exposure to operators on a loss of of fsite power. However, the FHC boundary is not considered safety-related and credit is taken only for the safety-related RSB confinement to limit j the release. The reactor, EVST, and FHC f uel transf er ports l have cooling capability provided by blowers supplied with l l
normal electrical power. In the event of a loss of this l power and immobilization of a fuel assembly-containing core component pot (the EVTM grapple drive is also supplied with l normal electrical power), the peak cladding temperature i remains below the clad temperature limit for anticipated (
events. In any case, emergency power is not required because l In case of power f ailure, the manual drive capability of the !
EVTM can be used without electrical power to raise or lower a !
core component pot to a location In which it is passively l cooled.
- The enclosed markup of the PSAR revises Section 9.1 to clarify the type of electrical power supplied for each situation in which cooling is needed and the consequences of loss of normal power. The revision consists of a new Table '
l 9.1-2A to Iist the peak fuel assembly ciadding temperature for loss-of-power cases and text in the description of each applicable f aceilty to describe the power supplied and to reference the new table.
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s Equivalent data for other components can be fcond as folIous:
EVST, paragrsph 9.1.3.1.3 EVTM, paragraph 9.1.4.3.3 Fuel Transfer Port, peregraph 9.1.4.7.3 -
- 2. Commen.t: A 2-h station blackout should be assessed (assiming 20-kW fuel and/or blanket In FTPs a bare 15-kW assembly in FHC) with the ensuing radiological evaluation including seal degradation, due to either high temperature or loss of pressure, as appropriate.
Resnonse: A two-hour station blackout while handling a bare fuel assembly during normal fuel handling cell (FHC) operations (assembly decay power less than or equal to 6 kWt) could result in release of fission products to the environment. The potential radiation doses at the site boundary and low population zone distances resulting from such a release would be less than the established limits.
Doses calculated for long-term station blackouts are listed in Table 1. The site boundary doses are Integrated for a l 2-hour period fof IowIng ioss of cooling and thus ere !
applicable for a postuiatod 2-hour statton blackout. The '
low-population-zone doses are Integrated for a 30-day period, and thus envelope those of a 2-hour station blackout. It is noted thut station blackout is not a design event for the f uel handling system.
The loss of all station power (both of f-site and on-site) could result in release of fission products from a fuel assembly being handled in the FHC. It would disable all FHC equipment, including the cooling grapple (which provides forced cooling of a bare fuel assembly during handlir.g), and the FHC Ire-celi crane (which transports the fuel assembly between locations where long-term passive cooling is provided). If a fuel assembly were being handled at the time of such a power loss, the assembly would be Immobilized in the FHC atmosphers. Cooling would be prcvided only be natural convection of the FHC argon atmosphere. The heat removal rate by this mechanism is lower than the heat t
generation rate in the fuel and the fuel assembly' temperature would begin to rise. It has been calcu6ated that the peak cladding temperature would rise to 1500 F in 33 min. At this temperature It is asstmed that the cladding would f all, l releasing fission products into the FHC atmosphere.
A conservative analysis of the postulated event assized that.
fission products from the high-temperature fuel assembly would be released directly to the environment, except for plateout on building surf aces as described in the next .
paragraph. The two systems which normally operate to eliminate or reduce release to the environment of fission products from the FHC would be disibled by a station bl ackout. The first of these systems is the FHC argon
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. circulation system (ACS), which removes host from the FHC l argon afmosphere to maintain the colI pressure negative l relative to the pressure in surrounding cells. The redundant t
ACS loops are supplled with power from a standby diesel i generator In the event of loss of off-site power. The -
j postulated two-hour station blackout includes loss of this backup diesel power. The ACS would no longer operate to l
remove fuel essembly decay heat, and the temperature of the FHC atmosphere would rise. The FHC pressure would become positive relative to the pressure of surrounding cells. The FHC liner would remain intact; however, there would be some leakage from the FHC to the atmospheres of adjacent cells.
The conservative assumption is made that the FHC liner would provids no holdup of fission products.
The second system which normally operates to minimize i radiation releases from the FHC is the reactor service building (RSB) ventilation sytem, which provides RSB confinement in the event of a radiation release. in a station blackout the ventilation f ans would be Inoperative and the RSB pressure would no longer be maintained negative relative to atmospheric pressure. Fission products released i from the FHC Into the RSB Interior are conservatively assmed to be released directly to the atmosphere.
The building structure is assmed to provide no holdup of fission products released during a station blackout from a fuel assembly in the FHC. All noble gas fission products would thus be released directly to the environment. There would, however, be plateout of volatlie fission products on '
the relatively cold surf aces of the FHC and the R$8 Interior.
It is assmed that.50% of volatile fission products released from a fuel assembly would be plated out before reisese to the environment. This factor is consistent with the guideline value for Iodine releases from LWR design basis accidents used in NRC Regulatory Guide 1.4, "Asseptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors."
The 50% factor is conservative in that it does not consider formation of Csl in the oxygen-depleted atmosphere of the FPC. This reaction would lead to a higher rate of removal for cesium and Iodine particulates penetrating the FHC liner.
The release of particulate forms of these isotopes would be expected to be reduced to less than 10% of the total enount released from a fuel a,ssembly Instead of the 50% assmed.
The analyses were carried out using the SIPOCX) serosol generatf or. code and the procedure in NRC Regulatory Guide 1.25 to determine the integrated radiation doses at of fsite locations. The Integrated doses to the whole body and to ,
designated body organs are listed in Table 1.
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An evaluation of a 2-hour station blackout during handling of a 15kW fuel assembly in the fuel handling cell has not been analyzed. The expected occurrence of handling a 15kW (i.e.,
fuel assembly greater than 6kW) fuel assembly in the FHC is 3 times in the life of the plant. The combination of this withf a 2-hour atatton blackout Is considered to be suf fIctentiy leprobable that evaluation of that situation is not required.
TABLE 1 INTEGRATED OFF-SITE RADIATION DOSE STATION BLAO(00T WITH BARE 6kWt FUEL ASSE2LY IN FHC Intmorated Radiation Dome (Ram) location Whole Body Thvrold Luna Bana Site Boundary 0.27 0.46 3.2 (Total)
(0-2 hr dose) 0.55 (Total) 0.0 (Pu) 0.0 (Pu)
Low Population 1.5 1.0 17 (Total)
Zone Boundary 4.0 (Total) 0.004 (Pu) 0.6 (Pu)
, (0-30 day dose)
Site Boundary 20 150 7.5 (Pu)
Construction 15 (Pu)
Permit Revlow Limit
- 3. Cnement: For inflatable or double purged seals, the project should demonstrate that loss of of fsite power and loss of purge or inflating gas does not exceed ar.ticipated event guidelines.
Response: There are three types of elastomer seals used on fuel handling machines and f acilities: static, dynamic, and infl atable. All of these seals are provided in redundant pairs and have essentially zero leakage (i.e., leakage is almost entirely due to permeation through the seat material).
The dynamic and inflatable seals have slightly larger leakage than the static seals on a comparable basis. All three types of seals have a buf fer' space between seal pairs. The buffer
' space for static seals is used primarily for periodic leak testing. The of fectiveness of these seals does not depend upn the presence of a buffer gas. Dynamic and inflatabia seals are provided continuously with a buf for pressure 4 between the double seals. The purpose of this buffor pressure is for leak detection and is not required to prevent seal leakage although it would mitigate an Inner seat leak.
The , Inflatable seals are the only ones which depend upon a
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4 continuous source of eloctrIcal powor and InfIetIon ges to i porf arm tholr function. In case of Ioss of offsito power, the seei InfIetIon system yelvos f alI open, providing the l esels with a continuous source of inflation gas from the t normal supply system. In the case of the EVTM, which moves _
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from one location to another this gas supply is from two separate ges bottles, two separate piping lines to separate [
seals and is Independent of the loss of plant gas supply. .
The argon bottles are standard high pressure bottles that meet DOT requirements and the piping from the bottles to the EVTM InfIstabie seals Is ANSI B31.1. The InfIstable seals on the EVTM are on the closure valve. [
- 4. .Commani Acceptance crIterie and results for of fnormal events involving blanket or control assemblies are required i (to date, only fuel ciad has been addressed).
The acceptance criteria for blanket assemblies, !
Response
control assemblies, and radial shield assemblies are less l stringent than those for fuel assemblies. To be consoryative, the Iimits for fueI assembiles have been used for alI other types of core assembilos.
- 5. Comment: The project should commit to perform (back up) ,
neutron monitoring to a technical specification during fuel i l
loeding (to include number of required operable detectors, calibration frequency, etc.).
Response: A technical spectfIcation (PSAR Sect!on '
16.3.10.3.3) Is defined by the reactor system for monitoring neutron flux level during refueling. (Calibration PSAR l SectIon 16.6.3)
I
- 6. Canment: The Project should address the IVTM grapple impact on assembly flow.
I Response: Calculations have been performed to establish the percentage of the fuel assembly outlet flow area which could be blocked without causing more than a negligible increase In fuel assembly exit temperature under ref ueling conditions.
These calculations resulted in Interf ace requirements placed on the IVTM grapple and holddown sleeve to limit their cross-sectional area in the region of the core assembly outget. The IVTM grapple is required to provide et least 1.2 I n, of fIow orsa through the outiet of the fueI assembiy;
. the actual grapple provides 1.6$ 1n.p The IVTM holddown sleeve, which rests on the six surrounding core assemblies, may not block any scre than 50% of any assembly's flow area. .
Because the holddown sleeve is a simple cylinder,. It provides substantially greater flow area than 50%. 4 The enclosed markup of the PSAR revises Sections 9.1.4.4.2 and 9.1.4.4.3 to describe the IVTM grapple input on coolar.t flow through core assemblies.
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'asert A Similarly, the AHM also has two sets of inflatable seals on the closure valve, supplied by separate gas bottles and gas systems.
(Note: any failure of the AHM inflatable seal system is enveloped by the accident described in PSAR 15.5.2.4)
The floor valves, located at the reactor, EVST and FHC during operation of refueling equipment, receive electrical power and Prior gas to inflate seals from the EVTM or AHM as appropriate.
to motion of the respective machine from the floor valves, the inflation gas is locked into the seals by the respective control valves in the floor valve. A single failure to one inflation system, i.e. failure of the control valve, will only disable one of two redundant seals.
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- 7. Qumant: The Project should specify design temperature and pressure of EVST, DTM, FHC, and their anals: consistency with normal and of f-normal accidents should be demonstrated.
Ramponse: The enclosed PSAR markup includes a new Table -
9.1-2B to Iist Ilaiting and maximum calculated pressures and temperatures for the EVST, EVTN, and FHC.
- 8. Damiani: Instroentation to verify adequate cooling of EVTN and fuel transfer port cooling inserts should be provided.
Response: Instroentation is provided to verify adequate cooling of the EVTN end fueI transfer ports. Thermoccupies are 1ocated along the Iength of the EVTN cold valI and at the cooling air Inlets and outlets; the cold well thermocouples and the alr outiet thermocoupie wIII verify adequacy of cooling. Thermocoueles are also attached at two places (near the outlet and near the seals) on the reactor fuel transfer port adapter cooling inserts. The thermocoupies wIII Indicate the need for cooling and will verify the adequacy of cooling if the adapter blowor is in oporatIon. The EYST adapter contains a thermocouple on the inner wall to serve the same function as the reactor fuel transfer port adapter thermocoupl es. The FHC spent fuel transfer port does not contain instr a entation. The decay power of core assembiles transferred is suf ficiently lower than the other ports, that overheating will not occur.
The enclosed markup of the PSAR revises Sections 0.1.4.3.2 and 9.1.4.7.2 to include the above Information.
- 9. Comment: The Project should identify and justify deviations from MS 57.1 and 57.2 (these standards are required by the SRP f or 9.1.2 and 9.1.4).
Response: The MS Standards 57.1 and 57.2 have been reviewed for applIcabtiIty to the retuelIng system. Those require-monts which were judged to bpapplicable have been incorporatod Into the design tystam a equipment.
41arifIcation b_.y. ,E M.t as foilows:
- a. Pressure Boundarv to Radioactive Environ (ANS 57.1.
Parmaraoh 6.1 ). The referenced paragraph refers to refueling equipment which is part of the primary reactor containment. There is no (RBRP fuel handling equipment
. In that category. The equipment hatch, which is open for refueling and is the applicable eaulpment, is part of the containment vessel (referenced PSAR Section 3.8.2.1).
- b. Raou t rad interlocks ( ANS 57.1. Table 6.2.1). The refueling equipment does not require any safety '
Interlocks as identified in PSAR Section 7.7.1.9. The design of the refueling equipment is consistent with ANS 57.1, Tabl e 6.2.1.
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- c. aridna Traval Annunelated f ANS 57.1. Paranranh 8.2.1.d).
The KVTM design includes motion alarm horns.
- d. t aan of Electric Passer Ranutta in Safa Onnfinuratlan f aut 57.1. Paranranh 6.2.1.5) . The Information is added to -
the PSAR by the enclosed markup of Sections 9.1.4.3.2 (EVTM), 9.1.4.4.2 (lVTM), and 9.1.4.5.2 (MM).
- e. CanabilItv for Emernancy Pomar Diwannect (ANS 57.1.
Parmaraoh 6.2.1.6) .' The information is added to the PSAR by the enciosed markup of Sections 9.1.4 A2 (EVTM),
9.1.4.4.2 (IVTM), and 9.1.4.5.2 (MM).
- f. Manual Motor t'=nabilltv ( ANS 57.1. Parmaranh 6.2.3.01 The Inf ormation is added to the PSAR by the enclosed markup of Sections 9.1.3.2.2 (FHC) and 9.1.4.3.2 (EVTM).
The MM has no manual motion capability. It does not handle fuel assemblies or control components, so the referonced requirament Is not applIcabie.
- g. Release-Proof Grannia f ANS 57.1. Parmaraoh 6.2.3.17).
The enclosed markup of PSAR Section 9.1.3.2.2 states that the grapple fingers are prevented from operating when supporting a core assembly.
- h. Adaeuata coollna ( ANS 57.1. Paranraoh 6.2.4.1.13). The referenced requirenent refers to an LWR fuel assembly transfer tube, for which the (RBRP counterpart is the EVTM (including transfers to and from it). Cooling of fuel assemblies in the EVTI' is described in PSAR Section 9.1.4.3.2.
I. Position indication ( ANS 57.1. Paracraoh 6.2.4.1.14).
Position Indication is displayed on equipment control panels. The MM will be positioned visually by movement of the R(B crane.
J. Svstem Naaded fo' Extramalv IJnlikelv Accident Acc.- - dation Shall Do So Assumina a Sinale Failure (ANS 57.2. Paracranh 4.2.4.5).- The EVTM is designed to accommodate a singte f alIure. Thore Is no other equipment in the refueling system which is required to limit the release of radioactivity.
- k. 1E N .- for Anv Svstam Keaning Radioactive Gas from the
- - Envirc.= nt or for Decav Heat R==nval ( ANS 57.2.
. - Parmaranha 4.3. d.3.1 (1 ) . and 4.3.1(21). Response
- in comment #1 and f3.
- j. No Draina ( ANS 57.2. Parnaraoh 5.1.1.11 The design provlsions to avoid draining sodlun to lower the level below the fuel level are described in Section 9.1.3.1.3, which is included in the enclosed markup.
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- m. Eatmale i Mekana Evatan f ANS M 2. Parmarash B.1.1.21 The EVST system is prwided with design features which prevent excessive loss of sodle through drain lines (see .
previous item) and to maintain a minimum safe sodlum (PSAR level even if the EVST ltself should rupture.
Section 9.1.3.1.2 is revised in the enclosed markup to Include the EVST guard tank function in the latter case.)
Therefore, a makeup system is not needed to accomplish the purpose of the referenced requirment.
- n. PrGvision for inanaction of Ffwad Ab w M.a FAN 3 U 2.
Parmaraoh 5.1.17.3). The flxad absoroers in the EVST can be removed using a special tool. This provides the capability for inspection.
EVTM
- o. Rad Monitor on FHM f ANE M 2. Parmaranh 5.a.11 permanent shielding is provided. As stated in PSAR Section 12.1, personnel are excluded from proximity of transf er port during raising and lowering of fuel assembilos. No permanent monitoring provided for refueling equipment. Radiation monitoring will be conducted by HPs. Section 7.3 of the PSAR discusses radiation montforIng for Contalment Isolatton. Section 12 of the PSAR, Figures 12.1-1 through 12.1-14 Identify the location of area and mobile monitor for the Reactor Service Building and Reactor Containment Building.
- 10. Cement: Confirmatory monitoring should be provided for the EVST during startup (for example, using a temporary neutron detector) since the calculated 2-signa upper-bound value of 0.947 for k-ef fective is close to the established limit of 0.95.
Resnonse: The value of k-of foctive quoted in the PSAR is for an upper-limit loading of the EYST, assisning the entire 650 storage positions of the EVST are loaded with This new case fuel of alsothe highest enrichment, which is the worst e.ase.
assumes that LWR recycle fuel of the highest plutonim and fissile content is being used. Initial Icedings of the EVST will be approximately 167 fuel assemblies. Thus, there will be substantial margin relative to criticality. A capability is available, however, to add a temporary neutron detection The system for conf truatory monitoring during EYST loading.
availability of monitoring capability is added to PSAR Section 9.1.2.1.2 In the enclosed markup.
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9.1-11 Comment: Can a single failure cause the EVTM to move away from a floor valve without the fuel assembly fully raised and the appropriate floor valves and closure valves closed.
9.1-11 Motion of the EVTM requires that (1) an Interlock be satisfled to energize the drive motors and (2) that the operator unlock thy seismic locks based upon electro-mechanical Indication. The Interlock and the Imput to permit operator disengaging selsmic locks are Independent. Hence no single failure will cause the EVTM to be moved until the F/A is fully raised and EYTM is properly sealed.
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MR Se(4g q,1 9.2-2 Comment: The appropriate selsmic classification of the ,
Nuclear Island General Purpose Maintenance System should be spectfled.
Response: The seismic category of the decontamination facility, the primary sodlum removal and decontamination system, the handling containers and the remote viewing equipment are established consistent with the impact on plant safety related equipment and witte their impact of their failure on public safety.
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9.2-4 Clarification is required on the analysis of sodium-water Purpose reactions Maintenance System. in the Nuclear Island General t ..
Response: F.arised PSAR section 15.7.3.7 Is attached. -
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1 9.2-5 Comment: Coustic-Induced stress corrosion cracking of ~ ,
4 the process system and the components to be cleaned f' ,
l should be evaluated. -
Response: For the stainless steel materials used, the stress ,
cracking regime due to NaOH has a lower temperature boundary of approximately 2400F as shown in attached Figure 2 of article' by R.E. .?wandby 'Correston Cha;1st Guides to Material Selection", Chemical Engineerlag 69:22, 186-201 (November 12, 1962). Expected process temperatures are below 200 0F. Hence, stress corrosion. ,
cracking is not considered to be a problem f or either l the process system or the components cleanet.
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75AR %Cahew % 9 Commentsr Additional Information is required on the trace heating of the following safety related equipments intermediate Heat Transport System Drain Lines Sodlum Water Reaction Product Rollef System (Piping, Valves and Tanks) -
Primary Cover Gas Equalization Line Overflow Heat Exchanger and associated Sodlum and Nak piping Ex-Vessel Sodium Loops Active saf ety-related valves C== M aqt3.S.2-4c, Response: Revised PSAR Section 9.4 gis attached (Eul#
- D.
Technical Specifications will be provided in the FSAR for the operation, as appropriate, of trace heated saf ety related equipment to consider thermocouple and heater malfunctions.
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The portions of the argon system which are extensions of the primary system boundary, the WST system boundary or the inert gas supplies to any sodium containing component or system should be designed to seismic Category I requirements up to and including the first isolation valve.
Response
Seismic Category I design is provided at least up to (and including) the first isolation valve in the argon supplies to and, vents from, all liquid metal ts gas e OM4M*2 Question 9.5-2 A backup argon supply (such as gas bottles) should be provided for the reactor and EVST to maintain the inert atmosphere on these systems in the event of a loss of normal argon. The capacity of the backup supply should be based upon a reasonable estimate of the duration of a loss of normal argon supply.
Response
Due to the low pressure of the cover gas over the reactor and EVST, and the heavy nature of argon, a continuoys supply of argon is not required for purposes of maintainine an inert atmosphere for either.
In both cases, the flowing supply of argon is provided primarily for purging of gaseous impurities. For the reactor, the normal supply is from recycled argon with on-line back-up supply of fresh argon supplied from liquid argon storage via the RCB gaseous argon header. For the EVST, an immediate on-line back-up is not required. The RSB header (normally supplied f rom its own liquid argon storage) can manually be supplied for back-up purposes through a valved cross-connect to the SGB header. "KP Q a 9 ,s L L o c-W4 -4. wa s 9 M M vm.4- veuA %dL" *I
' Question 9.5-3 a ,m. bcosT a. . we.d -b as k A us .A esc 4--
L- ,9 Jn ~A,+ .b W , c- <-- g . e e y. W.t4. Argy The desigi. code to be applied in the design of the piping " '-'l ^' V needs to be specified. Also the material for the nitrogen distribution system piping needs to be specified. & " v< p' d"
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,L.bco <-4 Safety classes and safety class changes are shown on PSAR e '
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= -- = = = - NY .It
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.. .. 1 . . . . . . . _ _ _ _ _ _ . . . . -
Figures 9.5-1 through 9.5-10. Desi directly (e.g., "sc-2" design code !n scodes ASME-III correspond Class 2;
'SC-None" design code is ANSI-B31.1). The material for the ntt rugen distribution system piping is substantially carbon steal. Stainless steel is specified, however, for those.
portions of the system containing liquid nitrogen, and up through the first valves on the outlets of the nitrogen ,
l v9.riurs.
Ouestion 9.5-4 The nitrogen samp1'ing and analysis units should be designed to be seismic II and should be capable of periodic testing and calibration. In addition, the radiation sensor, control and valves which divert the exhaust from the ex-containment inerted cells to CAPS upon detection of high radioactivity should be Seismic I.
Response
The nitrogen sampling and analysis units perform no nuclear safety functions and thus are specified as high quality commercial equipment. Features to facilitate periodic testing and calibration are provided in the design and include connections to introduce known gas and vapor test mixtures.
The potential for direct radiation release thru H&V venting of the ex-containment inerted cells (without diversion to CAPS) is well within 10CFR20 limits. Automatic and manual diversion features to CAPS are provided on the basis of maintaining releases As low As Baasonably &chievable.
Seismic I requirements are not warranted. Radiation detection, in addition to that provided by the nitrogen sampling and analysis unit, is also provided at least twice along.any of the potential exit paths thynugh CAPS or the H&V systems. The potential for release i$ discussed in PSAR Section 11.3.3.3.
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P54R w4 % 9.I, RSB RAPS & CAPS HVAC CHANGES (BECOMES CAPS HVAC)
- 1. RAPS, cells removed f rom RSB and placed in RCB.
- 2. Redundent RAPS & CAPS exhaust f ans removed, along with.
essociated Independent exhaust structure. -
- 3. . CAPS cell provided with exhaust (hense negative presure) via R5B Filtered Exhaust System, connected upstream of first cl ass 3 Isol ation damper.
- 4. Three unit coolers for CAPS cells rather than one.
- 5. Rad Monitors indicate high radiation from groups of cell.
High radiation will Initiate closure of quick closing isol ation dampers.
t
, . , _ _ _ , , _ . - , = , --e-v-
Quasileu 9.6.5 Diesel Generator Rooms HVAC System
- 1. Describe the design for ventilation of the r.ooms and associated equipment for the thlrd emergency diesel generator.
- 2. Provide justification for a f all closed mode for the diesel generator rooms emergency supply f ans e
outside air intake dampers and discharge dampers.
Verify that there is no common mode failure source that could result in loss of air flow to the diesel generator rooms when all dampers fall to opea. A fall open mode appears to be the proper f all safe position for the above dampers.
BERDQD&D
- 1. The ventilation system for each of the three diesel generator cells is typical as follows:
During normal plant operation (i.e. diesel is not operating) is provided by an H&V unit to remove heating the heat and load. venti wNde.
{l attonag During diesel operation, cooling is provided by two 50%
capacity f ans whichacirculate the room air through cooling coil s served by the emergency pl ant service water.
The DGB HV AC System has been revised to a recircul ating system cooled by emergency pl ant service water. Outside air 2.
is no longer the cooling medium f or the HV AC system serving the diesel generators. Accordlingly, the outside air intake
_ dampers and discharas dampertrhave been deleted.
6 4 JiTMis** glad 4+t %4Jeese % d % 4 (* M %
to ne noted that the HVAU system serving Division 1 &
2 switchgears remains as described in SSG119D1 9.6.1.1.2. ..
The two diesel generator cel l s (511 and 512) above the 816' elevstion have been removed and relocated to the new diesel generator buildings. The cells below el. 816' remain (with
- the exception of fuel oil transfer pump cells 526, 527) and are redesignated as being In the electrical equipment building (EEB).
- ~
The switchgear for Division 3 IE power will be located in the SGB and the cooling system serving the area wil-1 ese outside air and will be powered for the 3rd Division IE in the event of loss of normal power. Intake and exhaust air ducts will be Tornado and missile protected.
TS AR Sed % i.i t.
Eastien 2 16a-Quss11en 1 1
The design temperatures and pressures of the subsystems should be I made the same as those of the cells which they serve In order to ensure that a sodium or a NaK leak In a cell will not rupture the ~
gas cooling system, even assuming that an isolation valve.f alls to close. Added assurance of cooltag system Integrity will preclude opening a path for combustion product release of air In-leakage to the ilquid metal. j 1
BR&DQant j The design pressure of the subsystems is at least equal to the maximum cell design pressure due to Na or NaK leak.
The piping and system components are located outside the cells cooled, and thus are not directly exposed to the cell environment unless an isol ation valve f ail s to close. The piping design temperatures calcul ated, used are based upon consider the maximum cell temperature due to Na/NaK leak, the piping and component location is related to cell, whether natural or forced circulation is present, thermal inertia of the system, and thermal conductance of the piping system. In all cases the oesign temperature will be equal to or greater than the maximum expected temperature.
Sas11en 2.16._Quas11sn 2 If control rod cooling from subsystem CR is required to ensure a safety function, then that subsystem should be safety class 3, and a minimum of seismic 11 (also ASME code lil, class 3).
88599D3R Primary Control Rod Drive Mechanisms are cooled by nitrogen gas, supplied by Subsystem CR of the Recircul ating Gas cooling System.
The ef f ect of a f ailure in any part of these systems to supply this cooling gas has been investigated by a series of tests at W-ARD. The results of these tests were presented to NRC (R.
Stark, D. Moran) In a meeting on 10/14/82 and officially transmitted to NRC by DOE letter HQ: S:82:107, J. R. Longenecker to P. S. Check. A summary of these tests and results is presented below.
The.PCRDM Loss of Stator Coolant Flow tests were conducted with pr.ototypic hardware in 10000F sodium flowing at the design flow rate of 45,000 lbs/hr. The PCRCM stator temperature is normally measured by radundant thermocouples located in the outlet of the stator coolant flow. For these tests, additional thermocouples were located in the stator winding to measure the maximum stator winding temperature as a function of coolant flow.
Normal stator coolant flow is 157 scfm N2 at 95 psig. For these l
l l~.- -- - . . - . _ - . . . . - .
~ -_
I tests the coolant flow was reduced in a sortes of steps until and coolant flow was zero. At each flow rate, the stator temperature was measured as a f unction of time until the temperature reached an asymptotic value.
During the stator heatup, the PCRDM was placed in a hold-condition with the PCA withdrawn 36 inches. When the stator temperature reached its asymptotic value, the PCA was driven in eleven inches and then withdrawn to 36 inches a total of five times to demonstrate that the mechanism would operate properly with the stator at an elevated temperature. The mechanism was thee, scrammed and the unlatch time and scram insertion time recorded. ,
During this test, the stator was held at the design value of 17 5 volts D.C. As the stator winding temperature increased, the resistenace also increased and the current decreased. Thus power to the stator decreased and eventually an asymptotic condif f on was reached. This process was aided by radiant heat transf er from the surface of the stator.
At the worst case condition, i.e. complete loss of coolant flow, the maximum stator winding temperature reached 6580 F in approximately 260 miinutes. In this condition of the PCRDM would run, hold and scram properly. However, once the mechanism had scrammed, the stator winding resistance was too high to latch immediately. There t:ss no loss in scram insertion speed, and at higher temperatures the time to unleich became shorter.
The only negative ef f ect of high stator temperature on the operation of the Primary Control Rod System was the errette behavior of the absolute rod position Indication (ARPI). This occurs in a narrow band from 12.5 inches to 15 inches when the coolant outlet temperature reached 380 0F, approximately 170 minutes after complete loss of coolant flow. Continued operation of a PCROM without cool ant flow woul d cause the af f ected zone in the ARPI to increase and potentially cause permanent damage to the ARPl. However, plant procedures call for a consideration to remove power from a PCRDM, which has reached 3000F, and subsequently shutdown the plant. This procedure is Intended to prevent permanent damage to the ARPl.
As a result of the above described tests, it is concluded that the Primary Control Rod System can perform its required safety f unction to shutdown the pl ant with no loss in perf ormance in the event that there is a complete loss of coolant gas to the Primary Co,ntr ol Rod Drive Mechanisms. Consequently there is no need to upgrade the saf ety cl ass of Subsystem CR of the Recirculating Gas Coolant System, which supplies cooling for the PCRDMs.
PSAR pg. 4.2-256 (attached) documents that cooling gas'is not required to maintain the safety function of the PCRDM.
Sas11sn 1 Quasiinn I m
If the ability of the primary and Intermediate heat transport system loops to offect decay heat removal could be adversely impacted by e loss of cell cooling and subsequent, consequent impacts on freeze vent or. freeze seal integrity, then the associated subsysters should be updated to saf ety class 3, -
seismic catetory 1, and should have redundant cooling loops with the capability of being powered by lE power.
B2129n12 Loss of cell cooling has been analyzed for its affect on the freeze vent or freers seal Integrity.
As a result, once the primary heat transport system has been filled with sodium, the cover gas linas to primary heat transport systen freeze vents are capped external to the cell to preclude ingestion of cover gas in the event of a freeze vent melt.
Santien 9 1Ga-Qusstien s A description of the Instrumentation providing signals to the isol atin valves, causing them to close, is required, the description shoul d incl ude l ocation, seismic category, safety class and power supply. The staf f s position is that the safety class should be the same as that of the cooling system It serves and that all instrumentation should be seismic I and powered by IE power.
BAS 2enia PSAR Section 7.6.6 provides the detall description and figures of the Instrumentat ton and controls provided in the Recirculating Gas Cooling System. This section contains all the lief ormation required above. It should be pointed out that the Project has recently changed the failure position of the isolation valves to fail-as-Is. Chapter 9.16 and 7.6.6 of the PSAR will be revised in a future ammendment to reflect this.
. Santion SaLka-Quastien 5 The power supply and the failure mode of all isolatio valves should be specified.
Rainenis
~
PEAR Section 7.6.6, provides the Information regarding power supply and failure mode of the isol ation val ves. It should be noted 1 hat the Project has recently changed the f ailure position of the rectreulation gas cooling system isolation valve to fail-as-is. This will be incorporated into the PSAR in asfuture ammendment.
Sastien SalGa-Qunstlan Ik
I The rationale for automatic closure of isolation valves in RGCS cooling systems (e.g. MA, ME, EA, EB, and FC) where cooling may be required to support decay heat removal should be provided.
Alternately, remote manual operation may afford operator -
flexibility to better ensure safety.
Bannenns The Project has recently changed the f alloro position of the recirculation gas cooling system isolation valves to fail-as-is.
The valves will close automatically only upon receipt of a leak in the cooling coil or a high return gas temperature. Automatic operation is preferred over remote manual operation in order to provide an additional barrier in the unlikely event either a sodium or water leak to assure no sodium / water reactions can occur.
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... - . . . , ..-.~,--~.en . - ~ , - - - . - . - - - -
E m b re 2.
a Each systein consists of a control assembly, a control rod drive-I line and a control rod driv'e mechanism. The control assembly, located in the core, consists of a movable absorber pin bundle called the control rod and an outer duct assembly. The control rod driveline provides a linkage between the control rod and the control rod drive mechanism, located above the reactor closure head, which positions the control rod at appropriate axial core positions.
The control rod systems operate on the principal of varying the neutron absorption in the core by movement of the control rods in and out of the core. The primary system provides a means for starting up the reactor, regulating the power level of the reactor and compensa-ting the reactivity loss due to fuel bu"nup as well as functioning as the primary shutdown system. The secondary system provides reactor shutdown in the extremely unlikely event af failure of the primary system to shutdown the reactor.
4.2.3.2.1 Primary Control Rod System The 9 core locations comprising the Primary Control Rod System (PCRS) are shown in Figure 4.3-1. The three major components of primary control rod system are shown schematically in Figure 4.2-100. The PCRS
. design is essentially the same as the FFTF CRS design in order to maximize the use of the FFTF design, analysis, and testing experience.
C~.,"
. Figures 4.2-101 through 4.2-104 show the principal design features of 51 the primary control rod system.
4.2.3.2.1.1 Primary Control Rod Drive Hechanisms Principal features of the Control Rod Drive Mechanism (PCRDM) 51 l are shown in Figures 4.2-101,102 and 103.
The PCRDM is an electro-mechanical actuating device which utilizes a collapsible rotor roller nut drive,and is actuated by signals from the reactor control system. These signals cause the stator to be energized and magnetically actuate the rotor assembly ams, causing the roller nuts to engage the threaded portion of the leadscrew. Ro-tation of the electrical field of the stator causes rotation of the roller nuts with respect to the leadscrew which is rotationally restrained. This rotation raises or lowers the leadscrew whereas, stopping the rotation causes the rotor assembly to hold the leadscrew at any desired position. De-energizing the stator causes the roller nut to disengage the leadscrew, causing the leadscrew, driveline and the control rod absorber to drop into the core at a rapid rate of insertion (scram).4 Two independent control rod position indicating pystemsarenn.urpuratedineachcontrolroddrivemechanism.
Ne*A Irwa -4Ls. e4mbe is wm.veJ ley Ws. w.e,es.b4,g gas seeb43 s,ysh (ses "?swL se.W 8l.l4.F.2). Less .S hlNa w'au w.4 bk' d h, 4La. aVil Ay mE Te.RDM 4=
?* h 56 Amend. 51 Sept. 1979 O
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l 5.2.1.5 Reactor Vammal Prahmat l l
The Reactor Vessel Preheat System vilI control the dry heat-up and cool down I of the Guard Vessel, Reactor Vessel and internals between ambient (700F) and I 4000F and if required will provide make-up heat for that lost to the Reector Cavity during prolonged shutdowns.
The heat will be provided by tubular electrical heaters mounted between the j Guard Vessel and Insulation. These heaters will be arranged circumferentially around the Guard Vessel and will be grouped and controlled in zones of uniform j heat output. Temperature sensing devices will monitor the Guard Vessel temperature In each of these zones and provide the necessary feedback for power level adjustments in the heaters.
The heaters wilI be mounted to the same frenework which supports the Guard ,
VesseI insulation. Attachment ciips wilI offset the heaters from the Guard Vessel surface. Convective barriers, reflective sheaths and Guard Vessel Insulation will be used to optimize heat input to the Guard Vessel and minimize losses to the Reactor Cavity.
Preliminary preheat, startup, shutdown analyses have been performed on the Reactor Vessel and Guard Vessel to determine the temperature differences which will result In opening and/or closure of the annular gap between the two vessels. By necessity the preheat analysis is very preliminary since no firm preheat procedure has yet been developed. Figures 5.2-4 through 5.2-6 show the temperature dif ferences between the Reactor Vessel and Guard Vessel In the inlet and outlet plenum regions for the three transients in question. As shown the largest positive temperature dif ference between the Reactor Vessel and the Guard Vessel occurs in the outlet plenum region during startup (3350F) while the largest negative temperature dif ference occurs in the outlet plenum region during shutdown (-2140F). The nominal radial gap between the reactor vessel and guard vessel is 8 Inches at assembly and at the end of preheat.
This gap decreases to approximately 7.6 inches minimum during start-up and increases to approximately 8.3 inches maximum during shutdown. During preheat the gap also increases but to a lesser value than during shutdown due to the smaller maximun temperature dif ference.
Verlations in the axial gap between the bottom of the reactor vessel and the Inner surf ace of the guard vessel are noted between the states shown in the table. Thus the largest axial gap is 11.0 inches at the dry cold condition and the smallest gap is 6.2 inches at the end of the heating phase of preheat.
! 4 Im.,+ Sech m 5.2 .1.b 5.2.2 Design Paramatars Overall schematic views of the reactor vessel, closure head assembly, Inlet and outlet piping, and guard vessel are shown in Figures 5.2-1,1 A and 18. l The top view is given in Figure 5.2-2. i
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b 5.2-4c hnend. 72 Oct. 1982 i
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. . . . . = .: . .
...........a.
1 fnsect 5.2.1.6 closure Head Beating The closure Head Beating temperature control system consists of a single master temperature set point device which is used as the set point reference by the individual heater zone controllers. j The' individual heater zone controllers control the temperature of the closure head and reactor vessel l support ring based on the temperature reference ;
indicated and the individual zone temperatures as indicated by embedded temperature sensors located within !
the individual zones. Each sone has multiple temperature sensors. Failure of a single temperature
,,ensor s will be detected and alarmed.
The heaters are placed within individual zones to account for heat sinks, i.e., risers, nozzles and other head mounted equipment.
All elements of the head heating and control syntam are classified'as Seismic Category II.
Sufficient redundancy has been incorporated into the design to assure that failures of individual components shall not cause degradation of system performance to the point where the thermal requirements are not met.
A technical specification will be provided to address -
operations with failed heaters and temperature sensors
~
h I
@g dh a floor valve is mated to the EVST, and extends from the striker plate top to the cooling insert. The cooling sleeve, shield collar, and ficor valve adaptor reduce the transient dose rate from a spent fuel assembly being transferred into the EVST from the EVTM to less than 200 aren/hr at the surface. The port penetrations through the closure head are stepped to limit radiation streaming through the gaps. In order to allow sufficient time for Inspection and maintenance of the main bearings and seals, shielding is provided to attenuate the direct and scattered radiation levels to less than 125 mram/qtr.
The EVST Internals, storage vessel, and guard tank thicknesses are based on structural considerations, but also attenuate radiation in the radial and downward directions. However, the bulk of shielding to reduce radiation levels in adjacent vaults is provided by the concrete vault walls which are discussed in Sections 3.8 and 12.1.
The fuel transfer port plugs in the EVST head have double, static elastomer seals. Large diameter metallic seals are between the storage vessel and the closure head. The operating floor striker plate has a seal at its mating surf ace with the side wall vault lining. The turntable driveshaf ts have double, dynamic electemer seat s. All seals in the EVST are double with repability for convenient leak testing by pressurizing the buf fer space between seals in a pair. The ef fectiveness of the seals does not depend on 2
the presence of a buf fer gas, although it would mitigate an inner seal leak.
The EVST is designed with sensors and interlocks to prevent any unscheduled movement of the turntable while the EVTM is mounted on the EYST. The Interlock allows the turntable to rotate only when the EVTM grapple is in the full up position. The EVST is designed to prevent excessive relative motion between the head and turntable during an SSE.
Temperature instrumentation and sodlum level sensing probes will monitor cooling capability. High EVST sodium outlet temperature, and high or low sodium levels will sound an alarm. Other monitors will bo provided in the EVST cooling system (see Section 9.1.3). Sodium leak detectors will monitor the space between the storage vessel and the guard tank. An argon cover gas activity monitor will be provided. An area monitor of the gamma scintillation type will measure the gamma radiation on the RSB operating floor above the EV ST. The EVST design also includes the capability to add a temporary neutron
- detection system for confirmatory monitoring during EVST loading.
9.1.2.1.3 Safetv Evaluation The minimum center-to-center separation distance between storage tubes and the 9 storage positions permanently filled with BaC will keep the storage array
, subcritical even if the EVST were completely Toaded with new fuel assemblie; of the highest reactivity. The Ba C neutron absorbers are designed such that they cannot be removed inadvertenfly, i.e. cannot be removed with the normal refueling equipment. Based on the calculations reported below the K of this array, either with sodlum or void of sodlum, will be less than b5, as required.
I 9, t -1
= == --
I The system provides the capability to maintain the oxygen content of the sodium in the EVST at, or below, 5 ppe. The cold trap used for this service is separate from those used for reactor and primary loop sodium purification.
The system, working in conjunction with the Primary Sodium Storage and Processing System described in Section 9.3-2, provides a means of removing reactor decay heat in the event of loss of normal heat removal paths. These two systems, operating together, provide the Direct Heat Removal Service (DHRS). The DHRS is sized to I fait the average bulk primary sodium temperature to approximately 1140 F when the DHRS is initiated one-hat f hour after reactor shutdown. IJnder this condition, all primary pump pony motors are assumed operational. When the DHRS is initiated twenty-four hours after shutdown, the average bulk primary sodium tempe ature is maintained belor 900 F, assuming operation of a single primary pump pony motor. Total heat rejection capability of the EYS Sodium Processing System is based on removal of the required reactor decay heat in addition to the heat generated by spent fuel within the EVST. The maximum simultaneous EYST and reactor decay heat load is approximately 11-1/2 W, with DHRS Initiated one-half hour af ter reactor shutdown.
9.1.3.1.2 Deston Desertotton The EVST design and operating decay heat loads and sodium coolant outlet temperatures are given in Table 9.1-1.
The major assembt les of the EYST important to decay heat removal, other than the cooling system itself, are the storage vessel, the guard tank and the internals. The Internals, spectfIcally the turntabie, separate and support the spent f uel assemblies (contained in sodium-filled CCPs) permitting them to be satisfactorily cooled. The structural design of the turntable has already been discussed in 9.1.2.1.
The storage vessel has been classified as Saf ety Class 2 and is designed, f abricated and inspected in conformance with the appropriate codes and standards (see Section 3.2) to provide a leak-proof containment for the sodium coolant. The sodium level is maintained at a high enough elevation so that normal fluctuations due to changes in temperature or number of stored components do not uncover the top of the CCPs in which the spent f uel is stored. During of f-normal conditions, such as a leak or rupture in either the vessel or the cooling system, the vessel sodium outside the CCPs cannot fall below the minimum safe level. This level is defined as that below which fuel cladding temperatures would exceed the limits specified in Table 9.1-2 for the f uel assembly stored at the highest possible location within the storage vessel. The sodium nozzles in the vessel are located in the upper elevations of the vessel wall (see Figure 9.1-6). The EVST sodium inlet lines contain antisyphon devices which prevent a cooling system leak from lowering the
~
vessel sodium below the minimum safe level. The EVST guard tank is sized to contain sodium leaked by the storage vessel and meintain the sodium level above the minimum safe level.
D
..v.............. .
The EVS Processing System includes two Independent forced convection cooling circuits, designated circuit Nos. I and 2, each of which con remove the required EVST heat loads.
During normal operation, one forced convection circuit is used for EVST cooling and the other is on standby. Each of the circuits is composed of two loops, one a sodium loop and the other a NaK loop. The sodium loop circulates sodium from the EVST through a sodlueto-NaK heat exchanger and back to the EVST. The NaK loop circulates NaK through the exchanger where it picks up EVST heat, to a forced-draf t airblast heat exchanger, for dissipetion of heat to the atmosphere, and back to the sodium-to-NaK heat exchanger. The system also includes a cold trap to provide purification of the EVST sodium.
The two forced convection cooling circuits are supplied with Class 1E electrical power. Standby electrical power is provided for both circuits in the event of loss of normal power (see Section 8.3.1.1.1). Standby power is supplled to the two circJits by dif forent diesel generators.
in addition, the EVS Processing System includes a third independent natural convection backup cooling circuit designated No. 3 which can also remove the required EYST heat loads. In the extremely unlikely event of loss of both normal cooling circuits, the backup natural convection cooling circuit is used to remove the required EVST heat loads. Sodium circulates from the EVST through a backup sodium-to-NaK heat exchanger and back to the EVST. The NaK Icop circulates NaK through the exchanger where it picks up EVST heat, to a natural-draf t heat exchanger, for dissipatf or. ** heat M the atmosphere, and back to the sodium-to-NaK heat exchanger.
The EVST sodium outlet downcomers within the EVST terminate at dif forent elevations above the stored f uel . Loop #2 (forced circulation) has two outiets; the hlghest outIet used for normal operation, and a second outiet at a lower elevation such that any sodium leakage from Loop #1 (forced circulation) will not uncover the Loop #2 outlet. Loop #1 has one outlet nozzle located at an elevation between the Loop #2 nozzles. The lower Loop #2 nozzle would be used only in of f-normal conditions when both Loop #1 and the higher Loop #2 flow paths will not function. The third (backup) cooling circuit (Loop #3) has one outlet located bslow all Loop #1 and Loop #2 nozzles such that the Loop #3 outlet wilI not be uncovered by a leak in either Loop #1 or Loop #2. A leak in the Loop #3 piping wilI not uncover any of the loop outlets because it is entirely elevated above the minimum enfe level in the EVST.
The entire EVS processing system includes the following components:
l EVST Sodium Pumps (2)
~
,. EVST Sodium Coolers (2)
EYST Backup Sodium Cooler (1)
EVST NaK Pumps (2) 41-1 1. l I
i expension tank Is teolated and the EVST NaK pump is increased to 400 gpm each. The cover gas space in the two EVST NaK expansion tanks is cross-connected to equalize tank NaK levels.
9.1.3.1.3 safetv Evaluattan ~
The EVST cooling capability can be provided by either of two identtcal, forced convection cooling circuits, each of which can remove 1800 kW while maintainlog a maximum EVST sodium outlet temperature of In the extremely unlikely event that the normai circuits are unavailable, heat will be removed through a third independent (backup) natural convection cooling circuit. At 1800 kW this backup 0 cooling circuit will maintain sodium temperatures within the EVST below 775 F.
The critical temperature in a fuel assembly, from the standpoint of safety, is the peak f uel cladding temperature. The normal and emergency limits are given in Table 9.1-2.
The peak fuel cladding temperatures shown in Table 9.1-2a, are within the limits. Hence, no damage to the stored fuel assemblies will occur.
The codes and standards to which the EVST vessel and the surrounding guard tank are designed and f abricated assure that leakage of sodium will be a very low probabitIty event. At the minimum level, adequate cooling is usintained with no temperature Increases from those shown in Table 9.1-1.
Each of the three sodium coolln! loops is designed against the possibility of common-mode f at ture. Two pump ,uction lines are provided witnin the EVST for normal sodium circuit No. 2. T1e open end elevation of each is dif ferent, one high, one low. Each of the tw a lines is separately valved externally to the EVET. Af ter the inittal fif ' of the loop, the isolation valve in the low suction line is locked closed and remains closed (except for periodic testing) throughout the plant lif e. This low suction IIne is used only In the event of a major loop or vessel rupture. One pump suction line is provided within the EVST for normal cooling circuit No.1. The open end elevation of this line Ic between those for circuit No. 2. This line is valved externally to the EVST, and is called a "high" pump suction line. During normal systou operation, one of the normal cooling loops is operated using the "hlgh" pump suction line.
The suction line(s) in the standby normal loops are closed. In the event of a major f aiiure (rupture) of the operating normal sodium cooling ioop, the isolation valve in the pump suction line is closed by operator action from the control room, signalled by concurrent alarms, Indicating low level in the EVST
~
4.H 0 4
and a sodium leek within the cooling loop colI. If the Isolatton valve should not be closed the EVST sodium level could only be siphoned to the (*.lgh) pump suction outlet within the tank. Siphoning from the return If a f ailureline laofprevented normal I
I by en antisiphon vent in this line within the EVST.
cooling loop occurs, as described previously, the standby n increasing pump flow to the design rate of 400 gpm.
In the extremely unlikely event that the second normal loop cannot be activated af ter the f trst loop has experienced One suction a fellure, iIne Isthe third (backup) provided wIthin circutt wIII be brought into operatfon. The open end elevation of this the EVST for the backup cooling circuit.
suction line is below the lower suction line of normalSiphoning cooling circutt from this No. 2.
Flow back to the EVST is through the fill / drain line.
return line is prevented because the entire backup loop is elevated above the The drain line for the Na heat exchanger in the sodium level in the EVST.circrit has a removable pipe spool at an elevation above ;
to prevent siphoning through this circuit. ,
l Faili,re of any component, in any of the sodium or NaK loops, can cause lo only the circuit in which it is located. circuit can then be put into operati The potential radiological continuous cooling of the EYST sodium.
consequences of an extremely unlikely release of EVST sodlum to an inerted cell is described in Section 15.
All components of the norme.1 sodium and NaK loops which require electrical power are on the Class IEInpower the event system, to ensure of complete loss ofcontinuous external power EVST cooling a reactor decay heat removal.
to the plant, power to both of the normal cooling circuits is provided by the i
Imediate activation of the diesel-powered supply is not plant diesels.necessary for the EYST sodium pumps since the sodium volume within provides a heat sink to minimize sodium temperature rise during loss of Sodium circulation can be lost for approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> before circul ation.
the maximum sodium temperature in the upper portion of the EVST reaches 6 Activation of the emergency power supply to the NaK pumps and a1rblest f ans is required within 1/2 hour, however, to ensure the availability of DHRS for reactor decay heat removal .
The only " active" component in the backup loop is the damper on the natural i
draft heat exchanger. It is operated manually and, therefore, does not requiro connection to the seergency power system.
Isolation of all of the cooling circuits (sodium plus .ie associated NaK loop)
In separately shielded, inerted cells precludes both ;: dioactive sodium f the fire
, and the possibility of any failure in one loop imparing the operability o pther.
I 1
l 9.i-17 1
is used for transfer and will normally be empty or contain an empty CCP.
Normally, the decay power of fuel assemblies handled in FHC will be limited to 56 kW each. However, under unusual conditions, it may be desired to examine a short-cooled assembly, i.e., with a decay power greater than 6 kW, but less than 15 kW. Under this condition, no more than a single fuel assembly shall be permitted in the FHC. The transfer station will be designed to cool a single fuel assembly of up to 15 kW decay heat without exceeding the normal ciadding temperature if ait.
The gas cooling grapple will have sufficient cooling capacity to maintain the cladding temperature of a fuel assembly below the normal cladding temperature limit with a decay heat load of 15 kW.
9.1.3.2.2 Desion DeserietIon The spent f uel transfer station shown in Figure 9.1-8 consists of a lazy susan assembly containing three transfer locations for core component pots (CCPs), a bearing and drive system for the lazy susan, a structural support frame and bracketry, heaters, insulation, and seismic restraints for the lazy susan.
The transfer station is designed to ASE III/Sub Nf3 and Setsmic Category 1 requirements.
The spacing between the storage locations is determined such that adequate natural convection cooling is provided.
The lower portion of each storage location is a tapered cylindrical socket to support the CCP while providing a catch basin for sodium drippage. The cylindrical socket houses heaters on its outside which prevent sodium freezing wVn storing core assembiles with Iittie or no decay heat.
The decay heat will be removed by natural convection to the FHC argon atmosphere, which in turn is cooled by the redundant argon circulation system.
Under the worst case conditions the cladding temperature will not exceed 11000 F.
Cooling of the FHC argon atmosphere is provided by the Argon Circulation System, which has two loops, each consisting of a f an, gas heat exchanger and a piped distribution system. The heat exchanger removes heat from the argon gas and rejects it to the recirculating Dowtherm J System which rejects it to the Chilled Water System (Normal or Emergency as applicable) which in turn rejects the heat to the ambient air through the Emergency Cooling Tower inThethe emergency mode and through the Normal Cooling Tower in the Ncrual mode.
argon circulation system and supporting heat removal systems operate during normal plant operation, accident conditions, and periods of normal electrical power failure. The Argon Circulation System and Recirculating Dowtherm J System are Non-Class 1E systems supplled with standby electrical power by the jseme 'dlesel generator (see Section 8.3.1.1.1). The chilled water system loops are Class IE systems supplied with standby electrical power by diesel generators. One generator serves the argon circulation system loops, also (see Section 8.3.1.1.1). The low-pressuro perystem, including the shell of the cooler, is designed to ANSI B31.1, fnd Section Vill of the ASE Boller hwe,Ms.O
. 9.u-v f
i and Pressure Vessel Code, Solamic Category 1, and is located within a hardened I structure.
'The crane handled gas cooling grapple, shown schematically in Figure 9.1-9, is j mainly used to transfer bare fuel assemblies from the spent fuel transfer I station to the spent fuel shipping cask. Design of the grapple finger actuation mechanism prevents actuation of the fingers to release a coreThe assembly while the fingers are supporting the weight of the assembly.
crane hook includes a latch to prevent inadvertent disengagement of a cooling grapple from the hook. In the event of a loss of electric power, the crane Design of the ;
will stop at its position at the time of the power failure. Access to the crane for l crane includes the capability for manual operation.
manual operation is through ports in the wall and roof closure. l Two redundant argca gas-cooling blowers are mounted on the upper end of the These blowers draw argon gas from the surrounding cell gas-cooling grapple. environment and blow it through the grapple and fuel assembly, dischar '
back into the cell through the nozzles at the bottom of the fuel assembly. ,
The argon gas flow rate will be large enough to maintain the cladding temperature of a fuel assembly below the normal cladding temperature limit for l The blowers are suppf led with normal electrical decay heat loads up to 15 kW.
power.
1 I
9.1.3.2.3 safetv Evaluation A CCP containing a fuel assembly is cooled sufficiently by natural convection of the adjacent FHC atmosphere to maintain the peak fuel cladding temperature below the limits given in Table 9.1-2. The peak temperatures, given in Table 9.1-2A for normal operations in which the FHC atmosphere temperature is maintained by the argon circulation system and for the unlikely event of loss of cooling of the FHC ntmosphere, are within the limits.
The argon cooling gas flow rate through the spent fuel assemblies while being handled by the gas cooling grapple is suf ficient to maintain the maximum 0 In steady-state cladding temperature of a 15 kW fuel assembly below 600 F.
the event of loss of argon cooling gas, suf ficient time exists for the assembly to be transferred back to a Na-filled CCP in the spent f uel transfer station within the FHC before the fuel cladding reaches 15000F.
Adequate cooling of a spent fuel assembly suspended from the cooling grapple is maintained by the following means:
- 1) The grapple blowers are redundant to protect against loss of cooling capability by failure of one blower.
- 2) Each blower will be tested before beginning FHC spent fuel shipping operations to ensure its operability.
Evaluation of the loss of power for cooling systems for fuel assemblies in the FHC shows that the consequences are acceptable. In the event of loss of normal of fsite power, operatton of the cooling blowers The would loss stop and the of power would temperature of a suspended fuel assembly would rise.
also prevent movement of the FHC in-cell crane to return the assembly to a sodium-filled CCP. The extent of the temperature rise would depend on the
- i' 4.f-3( ,
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decay power of the assembly and the duration of the power loss. During normal operations with the maximum powered 6-kWt fuel assembly, there would be about 33 minutes before the peak cladding temperature would reach 15000F. If normal power were not restored before the temperature limit was reached, It. Is assumed that fission products would be released into the FHC. They would normally be retained within the FHC, however, because the argon circulation system and supporting heat removal systems (supplied with electrical power from an onsite diesel generator), would continue operating and maintain the FHC atmosphere at a negative pressure relative to the surrounding cells. In the unlikely event of failure of the diesel generator supplying the argon circulation system, the FHC pressure would become positive relative to surrounding areas and fission products would leak to the building. No credit is taken in ccident analysis for the seals of the FHC. This event is enveloped by the event discussed in PSAR Section 15.5.2.3.
31')L u-
Af ter removal from the reactor core, the sodium wetted core special assemblies are first brought to the EVST, and are later transferred into the FHC. 162 of these core special assemblles simulate fuel assembiles in the core and have full-flow filters. Some of these assemblies are partially disassembled in the FHC and made ready for sodium removal performed in the large component cleaning vessel. All other core special assemblies are only inspected.
Whenever core special assemblies are handled by refueling equipment, they are accounted for using the same inventory control system as 'real" core assemblies. Before entering and af ter leaving the reactor core lattice they are electromechanically identified by the IVTM using identification notches (see 9.1.4.4.2). In the FHC, core assemblies are identified and dif ferentiated both visually and elociromechanically. The core special assemblies leave the FHC in a polyfilm wrapped transfer rock. The outer surf ace of this polyfilm wrap is checked for radioactivity immediately af ter sealing and leaving the FHC port. The core special assemblies are transferred f rom the FHC to the Large Component Cleaning Vessel (ICCV) located in the R 2 for sodium removal . Cleaned core special assemblies are packaged in polyethylene bags, loaded into holding transfer racks, and transferred to a ste-age area.
The physical dif ference of identification marks between special and real core assemblies, the positive identification of core assemblies at two locations, and the radiological monitoring of core special assemblies before cleaning them are regarded as sufficient safeguards to insure that no real fuel assemblies are mistaken as special ones, and stored in a storage f acility not designed to receive them.
The maximum pressures and temperatures of fuel handling equipment during normal operations and of f-normal design basis events are listed in Table 9.1-28. The values are within the limits in the same table.
9.1.4.2 Deleted 9.1.4.3 Safety Asnects of the Ex-Vassel Transfer Machine MVTM)
The primary function of the EVTM is to transfer core assemblies between the reactor, EVST, new fuel unloading station, and FHC. The EVTM is designed to handle both new and Irradleted core assemblies in sodium-filled CCP's and bare new core assemblies. The EVTM has the following capabilities:
- 1) Grapple and release core assemblies, CCPs, and port plugs
- 2) Raise and lower core assembiles, CCPs and port plugs
- 3) Provide containment of radioactive cover gas
- 4) Maintain an argon environment
- 5) Maintain preheat temperature for new core assemblies
., 6) Provide up to 20 kw cooling for spent f uel assemblies
- 7) Provide radiation shielding
- 4. l - b ~l u
The EVTM is a shielded, inerted, single-barrel fuel handling machine. The EVTM is mounted on a trolley, which, in turn, is positioned on rails on top of the gentry. The gentry moves on crane rails between the Reactor Containment Building (RG) and the Reactor Service Building (RSB). The trolley rails are perpendicular to the gentry rails, ailowing complate indexing of the EVTM.
The EVTM mounted on its gantry is depicted in Figure 9.1-13.
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9.1.4.3.1 Danton aasts Adequate shielding for radiation protection is provided in the design of the EVTM to meet the radiation protection requirements of 10CFR20.
The activity released from a damaged or leaking spent fuel essembly while in the EVTM is contained in the EVTM by proper sealing or welding of penetretions and openings. Radioective leakage and dif fusion through seals are well below
+he limits specified in 10CFR100.
Suf ficient cooling capacity is provided in the EVTM to cool spent fuel assemblies with up to 20 kw of decay beat in sodium-filled CCPs and to ensure that fuel cladding temperatures do not exceed the values given in Table 9.1-2.
Mechanical damage to fuel assemblies could potentially be caused by the EVTM due to dropping of a grappled CCP or new fuel assembly or by tipping over of the EVTM. Dropping of new fuel assembiles or of (X:Ps containing new or spent fuel assemblies is prevented by the design of the grapple special-locking fingers and by suitable interlocks. The EV1N gantry and trolley are designed with anti-lif t-of f restraints and rail stops to prevent deral!!ng ofMechanical the gantry and trolley under combired normal operating and SSE loads.
collision between the EVTM on its gentry and other equipment, especially the control rod drive Iines, IVTM and equipment hatch between the RCB and RSB are prevented by a combination of stops, interlocks, and procedures.
If a seismic event occurs while the EVTM is mated to a floor valve at the reactor, or other location, the design limits the transmitted structural loads, such that the reactor head or other mating f acility is not danaged.
Motion of the EVTM relative to the mated f acility is limited to prevent contact with, or damage to a CCP, if a CCP happens to pass through the floor valve or the mating f acility and the EVTM at the time of the seismic event.
Cover gas release is prevented by maintaining seating of the EVTM to the mating f acility during a seismic event.
9.1.4.3.2 DasIon Deserietton Axial and radial shielding is provided in the EVTM to limit the dose rate to less than the criteria given in Sections 12.1.1 and 12.1.2 at the surf ace of the cask body. Shielding is provided over the entire length of the EVTM and is graduated in thickness, being thinnest at the upper end, where the radiation source from the spend fuel assembly being handled is least.
Approximately 11 in of lead shielding is provided at the lower end of the EVTM.
The pressure boundary of the EVTM is sealed using metallic and elastomer seals. -The metallic seals are single seals which serve as backup to two of the pairs of elastomer seals. There are All of three seals these types are of elastomer provided seats:
in redundant static, dynamic, and inflatable.
pairs and have essentially zero leakage (i.e., leakage is almost entirely due to permeation through the seat material). The dynamic and inflatable seatsAll have slightly larger leakage than the static seats on a comparable. basis.The three types of elastomer seals have a buf fer space between seal pairs.
buf fer space for static sealspdoes not depend on the presence of a buf fer gas.
Dynamic and inflatable sealfsre provided continuously with a buffer pressure q Q p.u.elg -% 9dske s \ea Ao*% ?%eenen A 44.n. osait q l-51
....z.=~..:.'...n..
between the double seals. The purpose of this buffor pressure is for leak detection and is not required to prevent seat leakage, although it would mitigato an inner seal leek. The inflatable seals are the only ones which depend on a continuous source of electrical power and inflation ges for operation. In case of loss of of fsite power, the seal inflation system valves would f all open, providing the seals with a continuous source of inflation gas from the normal supply system. (The valves are closed during normal operation to provide more sensitive seat leak detection.) The gas supply is from two separate gas bottles and is independent of loss of plant gas supply. Because the supply valves f ail open, loss of of fsite power would not af fect seal infl ation. The piping and valves from the gas bottles to the inflatable seals are ANSI B31.1. The seal inflation system and controls have been investigated to ensure that there are no common cause f ailures which would disable both inner and outer seals.
6 The EVTM is hermetically sealed to a refueling station by lowering the closure valve which mates with a floor valve. The actual sealing at this interface is accomplished by elastomer double seals, which are periodically leak checked.
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, , ,,,,7 The space between the closure valve and the floor valve is purged by the plant argon supply and vent system, through hose connections made af ter the EVTM has been mated to a refuel tag station. -
When the EVTM is mated to a floor valve at the reactor or other location, the large bending moments and shear loads In the combined vertical structure due to a seismic event sre relieved by structurally decoupling the EVTM from the floor valve at the joint Interf ace. The joint between the extender and closure valve Is designed with two sliding surf aces. One of these can experience limited horizontal motion It horizontal earthquake loads exceed a predetsrmined value, while retaining Its vertical toed carrying capability.
Similarly, the second surf ace can experience limited vertical movement during a seismic event but retains horizontal restraint capability. All sliding joint surf aces are sealed against each ot.ier to provide cover gas containment under normal and seismic conditions.
Cooling of a CCP within the EVTM is accomplished by heat transf er to a cold \s wall system consisting of an about 8 in. ID sealed cold wall having an array of axial fins attached to the outside. The cooling concept is li tustrated in s Figure 9.1-15. Heat from the 3-ft high fueled region of the spent fuel assembly is distributed over the 15-f t length of the CCP by natural convection ' -
of the sodium in the CCP. The heat is transferred from the surf ace of the CCP -
to the cold wall primarily by thermal radiation and secondarily by conduction ,
across a stagnant argon-filled gap. The cold well is cooled by forced convection of ambient air circulated past the axial cold wall fins. Forced s2 air convection is provided by a blower with the capacity to circulate '
suf ficient air to maintain fuel cladding temperature to less than the normal Ilmit (see Table 9.1-2). In case of f ailure of the blower, or complete loss of all power, natural convection of air is initiated by automatic opening of butterfly valves just upstream of the blower. Natural convection air flow is sufficient to maintain fuel cladding temperature to less than the limits for unlikely and extremely unlikely events in Table 9.1-2. The EVTM cold wall is part of the EVTM containment boundary and is designed and f abricated to quality and inspection standards corresponding to Safety Class 3 (see Section 3.2).
Instrumentation is provlJed to verify adequate cooling of the EVTM.
Thermocouples are located along the length of the EVTM cold wall and at the cooling air inlets and outlets. The temperatures measured by these thermocoupias w1Ii verify adequacy of cooling.
A cooling system for a CCP during normal transfers to or from the EVTM is not necessary. Although there is a region in the transfer path in which the ef f ective cooling is less than needed to prevent f uel assembly heating, the time spent traversing this region is short enough that heating is Insignificant. In the event that a CCP became immobilized in this region, 9.l- @
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however, there would be significant heatleg of bothla thefuefuel essemb Thwefrre, the capability material surrounding the CCP in the port.
provided for removal of host from en leechfilzedEYFi CCP fuel. at tech 3,
,j'
- of th trNitter ports:, reactor fuel transfer port (se'r. Section 9.1.d.7).At one part '
transfer port, and me spent fuel transfer port. ,
N, pitadref to a cooling duct in the port, 'io circulete P wereimtiding str.?n i
- pedi's./ The blower is norme!IV of f buf would be 4pene'd on if\ a CC 'q Twilovers are supplied ,WQh norral ~.slectrical '
lasrobilized in the port. ,,,
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<, s sentatica includes disptey cf the graprig c The EVTM grapple drive Thesystensystem. insirje.
litludes %e capability for senual opersilon, .
wrtic8 M esition.
th a hand'srank to allev raising or f owaris$ a CCP to o region of pass uooling In the event of a. power f ailure. , fApower bra fellure. If a CCP th^
.the grapple at its positthy at the $lme o Immo'cllized in a reglon offeduced ef fectIVe cooling i sive ering a trb '
from the EV D , alth the tre?.stpapower ' ' . f ailvra, ,
the CCP could,lse +
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cooling.
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i Electric power to9 the LVTE can be ?.g manually dlaco
'EYTW power. ,,
e q.
_ CCP The EYTM ctP grapple has sn;1nterlocking finger design such be \ -
thrt'i w engaged with and supported by ihe grapple fingers, tne fingers [ cann ,
retracted even if the entire weight is supported by the fingeFactuating
. chain. Redundant support chains are utilized to Insure component s
, s.
- ihe event of a single chain f at ture. TI.e gantry, In turn, trarrels on rat a
,t *
< by a trolley, traveling on a gantry. Trol. ley and gan*'ry 3 '
tracks (see Figure 9.1=13) secured to the R$8/RCB floor. + i ts. .
s I
s-wheel truck structure's incorporate anti-lif t-of f and overtGrAlegres ra n Trolley and gantry rails are equipped with rail stops plys.shoch positive trevel limitations. N '
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adequate assurance for both an CBE ond $$E that: (a) the composite component will reseat from much greater than maximum anticipated vertical motions (b).
the clmps will prevent disengagement of the extender from the closv:,o valves and (c) the lip on the closure valve will limit horizontal motion to one Inch.
The latter is more than adequate to prevent contact with or damage to a CCP that might be in transit through the plane of the slip joint. The seals between the extender and the closure valve and between the closure valve andy ^
the floor valve ensure cover gas containment under normal and seismic
conditions.
The EVTM cooling capacity of 20 kW is adequate to provide a substantial margin '
I above the maximum normal heat load expected, which is 15 kW. The active portion of the cooling system, the bicwor, is capable of providing the speelfled cooIIng without exceeding normal temperature limits. In case of f ailure of the blower or loss of all AC power, completely passive cooling Is autcoatically provided by natural convection. In this case, cladding temperature Is maintained to less than the limit for unlikely events. Tne ,
peak f uel cladding temperatures, shown In Table 9.1-2A, are within the limits given in Table 9.1-2. Therefore, no damage to the fuel assembly will occur.
The temperature of a fuel assembly in a CCP will Increase If the CCP becomes Imobilized during a transf er to or f rcr.n the EVTM. The peak fuel cladding temperatures are listed in Table 9.1-2A for Immobilization in the EVTM valve <
stack assemblies (the assemblies below the cask body, see Figure 9.1-14) or the f uel transfer ports for the reactor vessel, EYST, and FHC. In each case, the peak temperature is less than the limits given In Table 9.1-2 for the frequency class of the event and no radioactivity will be released from the fuel assembly.
The design heat removal capability of the EVTM has been experimentally verifled in EVTM heat removal tests. These tests were planned early in the CRBRP program; theIr purpose and outlIne ate described !n SectIon 1.5.2.7 of the PSAR.
The tests have been successfully performed, and the test data have been analyzed. References 7 and 8 of Section 1.6 document test evaluations, test descriptions, and experimental data. Following the review of test data, the EVTM heat transfer computer model was modified to consolidete the model predictions with the experimental data.
The main conclusion from these tests is that the EVTM has heat transf er a pability adequate to meet its design conditions for both forced and natural air convection modes.
A summary of the tests and major findings is provided below.
Full-Scale Heat Transfer Test _s Full-scale tests (Reference 7 of Section 1.6) were performed in a HEDL test f acility design to simulate the cooling systems of the CLEM for the FFTF and the EVTN for the CRBRP. The fuel assembly was simulated by a full scale, 217-pin, electrically heated " fuel" bundle in a hexagonal duct. The fuel 4f-C 11
/
essembly was contained In a sodium-filled .ve component pot (CCP), surrounded by en inert gas-fi1ied annulus, and cooled by the concentric cold walI. The test f acilIty and test articie design assured th9 accurate extrapolation could be applied to test data for either refueling achine. Major test results showed the following:
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9.1.d.4.2 Dasign nosertpttan The steel shleid column of h IVTM hos a verleble thickness radiation shield (from 2.75 to 1.5 In. above the small rotating plug nozzle) to attenuate gamma radiation from radioactive sodlum. The amount of radfoective sodlus that wilI i be in the upper part of h IVTM's housing Is Iimited to h surf ace fila remaining on the grapple stem af ter It is raised from the sodlum pool during the transfer operations. The active fuel zone of the core assembly which Is under approximately 9 ft of sodium during transfer operations, will also contr!bute to the external dose rate, but to a 1 esser degres. There wIIl also be a slight contribution to the dose rate from the radioactive cover gas inside the annulus between the grapple stem and the shield column.
The IVTM is sealed to the SRP by 3 elastomer 0-ring seals with a buf fer region between them. Between each set of seals, a leak detector port is provided to enable connection of a sensor for monttoring the seals Integrity before !
refueling. This arrangement is typical also for most seal sets that involve j dynamic motion, such as the grapple stem and holddown drive shaf ts. Static j seals or dynamic seals with very small displacements, such as those for the l Identif ter pawl shaf t, are double, with capability for convenient periodic leak testing by pressurizing the buf fer region between the seals. The ef fectiveness of the seals does not depend upon presence of a buf fer gas.
The IVTM drive mechanism has been designed to exert an upward or downward load of 5000 lb maximum. Incorporated into the design Is a pneumatic load control system and a load cell system that limits the load exerted on a core assembly to 4300 lb pull and 3000 lb push. The load and the grapple vertical position are displayed on the IVTM control console. In the event of a loss of electrical power, a braking system automatically stops the grapple at its position at the time of the power f ailure. The grapple drive system includes the capability for manual operation to allow raising or lowering a grappled core assembly. Repositioning would not be needed for cooling the assembly.
It would be immersed in sodlum and thus passively cooled, whenever handled by the IVTM. Electric power to the IVTM can be manually disconnected at olther the IVTM control console or the substation supplying IVTM power.
The IVTM is pcsitioned above core assembly locations by rotation of the reactor rotating plugs. The positions of the plugs are displayed on the IVTM control console to define the IVTM horizontal position.
The load in the pneumatic load control system will be set to provide a normal push or pull force on a core assembly of 1000 lb. The pressure of the load control system may be adjusted to provide higher load capability up to the push or pull leed limits.
The design provides for the loed to be limited in two ways.
- 1. The primary limitation is provided by the pneumatic load control sy stem. The pressure in the pneumatic load control cylinders limits the load applied by the electromechanical actuator to the driven core assembly. When the core assembly insertion reststance load exceeds the pressure setting in the system, the core assembly stops moving, but the electromechanical actuator continues driving the pistons for about 0.25 inches. This dif ferential travel trips a set of limit
- 4. f *fi 43-
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l switches which automatically stop the electromechanical drive. Self-contained hydraulic deshpots prevent sudden actuator movements due to sudden changes of the frictional resistance at the load. .
J
- 2. Load cells are used as backup to the pneumatic system to shut of f the
, electromechanical drive when the preset load limits are exceeded.
Actuation of the grapple fingers for pickup or release of a core assembly is possible only when all the grapple finger actuation Interlocks have been satisf fed, the grapple is pushing on the core assembly (f.o., core assembly Is in full down position), and the load control limit switches that shut of f the electromechanical drive are tilpped.
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released from the IVTM does not exceed the limits set forth in Sections 12.1.1 and 12.1.2. Leek detectors will be used to continuously monttor the ef fectiveness of the dynamic seats. Periodic leek checks will be carrie.1 out for the static seals. In addition, the R(B will have radioactivity monitors to detect accidental releases and sound alarms.- _
The drIya mechenIam end grapple design, together wIth Iosd controf s, IIntt switches, and interlocks will prevent exertion of excessive forces on core essemblies, unscheduled vertical movement of the grapple or disengagement of the grappio fingers except in the full-down position of the grapple. Limit switches end interlocks (see Section 7.6.2) will also prevent Inadvertent rotation of the reactor plugs. The extremely unlikely accident of a fuel assembly dropped from the IVTM is discussed in Section 15.5.2.1.
The combination of position Interlock switches and the switches of the Iced control system in conjunction with the design of the discriminator post and matching receptacle ensures proper seating of core assemblies. The fuel and Inner blanket assemblies are divided into 8 groups, each having a dif ferent configuration of discrimination post inner and outer diameters fitting into corresponding receptacles. This ensures that a fuel assembly of one group can only fit into its corresponding receptacle, see also Section 4.2.1.2.3.
The IVTM grapple and holddown sleeve, when positioned for pickup of a core assembly, partially cover core assembly outlets, reducing coolant flow Through af f ected assemblies. The calculated minimum flow area to ensure a negligible increasg in fuel assemgly outlet temperature during refueling conditions is 1.2 in. . The 1.6-In. grappled assembly flow area provided by the actual grapple provides for adequate coolant fIow through the assembiy. A substantially greater flow area is maintained for assemblies af fect.',d by the simpie cyiInder of the hotddown sieeve.
Core assembly identification by the IVTM Identification and orientation system prior to core assembly insertion into the core, and the position Indicator system of the reacter rotating plug, vilI ensure core assembly Insertion into a correct core location.
If, however, a core assembly is erroneoutly Inserted into a core location belonging to another core assembly group, the core assembly discriminator post will bottom against the top of the receptable, thus resulting In an Improper sosting. The length of the core assembly discrimination post is sufficient to preclude tripping of the grapple finger actuation interlock switches, thereby preventing core essembly release even though the preset force of the loed control system has been exceeded. This condition and the IVTN grapple <
position Indication system signal vilI warn the operator to Initiate a corrective action.
Jamediately af ter core assembly renovel from the core, the IVTM Identification systsm Identifies the core assembly serial number, in the automatic control mode, this serial number is con $ored by the control system with tho' serial number designated by the ref ueling program that must be in that particular core location,1hus ensuring that the correct core assembly is removed from the core. In the manual control mode, the operator compares the calculated serial number to the serial number designated in the refueling plan.
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If, however, a core essembly la erroneously removed from the core, the sertal number discrepancy wilI warn the operefor to return the core essembly lado the core position last serviced and to initiate a corrective action.
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provide a second, redundant load path from the handling ball to h polar crano hook. When not In use, the ##4 la stored at a parking station located In the northeast quadrant of the imIIdIng.
The parking station Is designed for the $$E solemic loads which are carried -
into the R3 structure.
In the event of a loss of electric power, a braking system automatically stops the grapple at its position at the time of power f atture.
Elvetric power to the AHM can be manually disconnected at either the MM console or the substation supplying the floor service station from which the MM is being supplied.
The vertical position of the AHM grapple is displayed on the AHM control console.
When the AHM is in position at the reactor, only the extender mating flange is resting on the floor valve, which in turn is supported from the small rotating plug (SRP) by an adaptor. If the two components were fIrely attached to each other, the resulting combined structure, in ef fect, would represent a tall, vertical cantilever rising from the SRP, attached at its upper end to the polar crane. The large bending moments and shear loads in this combined structure, resulting from horizontal excitation due to an OBE or SSE, are relieved by structurally decoupling the AHM from the floor valve at the extender / floor valve joint interf ace. At a predetermined horfzontal ground acceleration, complete severance of the AHM from the floor valve (" breakaway" concept) eliminates the cantilever beam ef fect and significantly reduces all seismic loads.
The joint between lower extender flange and floor valve is designed Thiswith shear enables pins which f all upon reaching a predetermined horizontal load.
the AHM to separate from the floor valve during a seismic event. The design incorporates a pneumatic reservoir which Initiate raising of the AHM extender following the shearing-off of the shear pins. The actuators can ratse the extender by about 3 Inches in less time than it takes for the extender to clear the floor valve during the horizontal movement due to an OBE or SSE.
9.1.4.5.3 safetv Evaluation The radial and axial shielding provided by the AHM limits the Integrated dose to personnel to less than the maximum allowable dose rate during the Installation or removal of the components handled by the AHM. As with the EV1H (see 9.1.4.3) the radiation source in the machine is intermittent and short term.
The AHM has adequate seals to prevent radioactive emissions to the R2 operating floor. Radioactivity released does not exceed the limits as set forth in Sections 12.1.1 and 12.1.2.
a 11-5"7 14 -
\ . -
i The design of the grapple release mechanism, Interlocks, and the redundancy of chains will limit the potential for dropping grapple loads. The interlocks are similar to those of the EVTM, and slutter to those discussed in Section 7.6.2. The structural support and restraint of the AHM storage f acility, and the rigging of the AHM when attached to the polar crane prevent toppling or dropping of the AHM due to an SSE or other loeds.
9 I
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In. diameter opening. The incrossed thickness of the EVTM floor valve in radial and axial direction provide the additional shielding required for the much hig5er radiation source which passes through an EVTM floor valve (spent fuel assembly) compared to en AHM floor valve (IVTM port plug).
The stepped upper and lower steel plates of the floor valves, concentric to the valve port, (see Figure 9.1-18) prevent dif fusion and radiation streaming through the minimal mating surface gaps. These design features limit the transient dose rate at the surf ace to less than 200 rem /hr durlag transfer of radioactive comporents, and 5 wom/hr when closed over the reactor ports.
The floor valve is sealed to the fuel transfer aort adaptor by double seals, and bolted to the adaptor flange. The movable circular disk which closes of f the port opening in the valves is also sealed by double seals.
The rotor driveshaf t is sealed by dynamic seals. All seals are double buf f ered and of elastomer material . The discussion of EVTM elastomer seals in Section 9.1.4.3.2 is applicable to the floor valve seals also.
The position of the floor valve gate (open or closed) is displayed on the floor valve control panel and, If the EVTM is mated to the floor valve, on the EVTm control console.
9.1.4.6.3 Safetv Evaluation The radial shielding limits the dose rate on the floor valve surf ace to less than the criteria in Sections 12.1.1 and 12.1.2 during transfer of the highest powered spent fuel assembly (for the EVTN floor valve). The floor valve is considered a piece of equipment whose main function is to permit transfer of radioactive components, both fueled and non-fueled, between a machine and a facility. The radiation source is transient and short term (less than 1 min per transfer) in nature. Hence, it results in a low Integrated dose.
Another f unction of the floor valves Is to provide axial shleiding to replace that normally provided by the port plugs. T M axial shielding limits the dose rate to personnel to 5 wom/hr when plac.ed over a reactor port and to 0.2 mrem /hr when placed over EVST or FH(: ports. PersonneI cannot receIye a direct axial dose because of the large vlameter of the floor valve. In addition, the valve is covered by a mat.ng machine much of the time. In all cases, suf ficient axlei and radial shleiding for the EVTN and AHM floor valves is provided to limit the Integrated dose to less than 125 wom/ quarter, and dose rates to the zone criteria of Section 12.1.
The floor valve has adequate seals to prevent excessive radioactive release to the RW and RSB operating floors. Accidental cover gas release through Inadvertent opening of a floor valve in the absence of a mating fuel handling machine (EVTM, AHM) on top of the floor valve is prevented by interlocks. One Interlock prevents energizing the valve operating motor unless a mating machine is on top of the floor valve. (Electrical power to the floor valve motor is supplied by connection to the mating machine.) Other Interlocks prevent (1) depressurtzing the buf for gas zone, and (2) raising the closure valve extender, unless both the closure valve and the ficor valve are in their closed positions.
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l As discussed in Section 15.5.2.4, an unlikely accident releasing radioactive I cover gas from the reactor leads to a site boundary done well below the guideline value of 10CFR20.
9.1.4.7 safety Asnects of the Damefor Fuel Trensfer Port Admator and Fuel Transnort Port CoolIna insarts The reactor f uel transfer port adaptor (see Figure 9.1-19) Is positioned on top of the reactor fuel transfer port and extends from the reactor head to the bottom of the floor valve which is located at the elevation of the RCB operating floor. it serves as an extension of the reactor cover gas containment and provides shielding when irradiated core assemblies are removed from the reactor. The aceptor also guidos cooling air from an air blower-to a cooling insert inside and below the adaptor.
l The function of the cooling inserts located around the EVST and FHC fuel transfer ports as well as the reactor port (see Figure 9.1-19), is to remove decay heat should an Irradiated core assembly in a sodium-filled CCP become immobilized in a fuel transfer port during transfer between the reactor l vessel, EVST or FHC and the EVTM.
9.1.4.7.1 Destan Basis The design bases for shielding and radioactive release of the fuel transfer port adaptor are the same as for the EVTM (see 9.1.4.3.1). The reactor, EVST, and FHC fuel transfer port cooling inserts have the capacity to remove decay heat from 20 KW f rradiated core assemblies in sodium-filled CCPs to prevent 0
exceeding the 1500 F spent fuel cladding temperature limit specified for unlikely or extremely unlikely events (Table 9.1-2).
9.1.4.7.2 DasIon DeserIntion The reactor fuel transfer port adaptor extends from the upper surface of the ,
f uel transfer port in the reactor head to the operating floor, see Figure !
9.1-19. The upper surf ace of the reactor fuel transfer port adaptor consists ;
of a flange to which is bolted to the flange of the cooling insert. Shielding is provided by a thick, annular lead cylinder surrounding the adaptor cover ;
! gas containment tube over its entire length to limit the dose rate at the shield surf ace to less than the limits given in Sections 12.1.1 and 12.1.2.
The lower part of the adaptor is bolted to the ,r actor head permanently, while l
theupperpartisInstallegringfueling ni l The reactor fuel transfer port cooling insert extends from the top flange of l the adaptor to the fuel transfer port nozzle. The cooling insert uses a cold l
wall cooling concept, similar to the EVTM. The CCP containing a spent fuel assembly is cooled by thermal radiation and conduction across the argon gas l gap to the cold wall which forms the confinement barrier for the reactor cover l gas. Ambient air is blown down the outside annulus of the cooling insert, and discharges into the reactor head access area. Air flow from the blower is
' adequate0to limit the cladding temperature of a 20 KW fuel assembly to less than 1500 F. The. mocouples are attached at two places (near the cooling air outlet and near the seals) on the reactor fuel transfer port adapter cooling insert. The thermocouples Indicate the need for cooling and will verify i adequacy of cooling If the adapter blower is in operation.
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The EVST adapter contains a thermocouple on its inner well to serve the same The FHC function as the reactor fuel transfor port adapter thermocoupies. The decay powers spent fuel transfer port does not contain Instrumentation.
of core assemblies transferred are lower then for the other ports.
The cooling insert is sealed to the rotating guide tube and the EVTM floor valve to form a continuous reactor cover gas pressure boundary through the Sealing to the rotating gulde tube is by three adapter during refueling. static elastomer seals with a continuously monitored buf fer r center seal and each outboard seal. Sealing to the floor valve Is by two sets of double static alastomer seals for which the floor valve provides the sealing surf ace. The buf fer region of each seal pair is continuously monitored to detect a loss of seal Integrity, s
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9.1.4.7.3 safetv Evaluattan The translent dose rate from the highest powered spent fuel assembly is less than the criteria in Sections 12.1.1 and 12.1.2 at the surf ace of the adaptor body. This significant dose rate exists only during the short time (a few minutes) when a spent fuel assembly travels through the adaptor and floor valve into the EVTM. The closest location where personnel can be exposed to the radiation source is more than 10 ft. from the adaptor surf ace for normal oporation, end nore than 2 ft. from infraquent malntenance operations. Both locations are on the RW operating floor, above the adaptor . Therefore, Integrated exposures are low.
Cooling of spent core assemblies in the reactor or EVST fuel transfer port is adequate to maintain the assembly cladding temperature below the 12500F limit for normal operations and anticipated events. During normal operations, the transit time of the core assembly through the port is short (a few minutes) so that there is no significant heatup. In the event that an assembly becomes immobilized in the port, design provisions to maintain the cladding temperature below 1250 0 F will be used. If Immobilization of an assembly is the result of a mechanical f ailure of the EVTM grapple drive system, the backup cooling system for the port may be turned on to provide the necessary cocling. If Immobilization is the result of a loss of power (which would also distbie the backup cooling system), the EVTM grapple drive system may be operated manually to raise or lower the core assembly to a location (EVTM or sodium pool) where adequate passive cooling is provided to maintain the cladding temperature within the 12500F limit. However, those operator actions are required only to maintain fuel temperatures within normal limits. In the unlikely event that a core assembly becomes immobilized in the port for a longer time by coincident drive system mechanical failure and loss of power or f ailure of the operator to respond to this condition the cladding temperature would exceed 1250 0 F but would remain below the 15000F limit for unlikely and extremely uniIkely events.
The seals of the reactor fuel transfer port cooling insert are adequate to prevent excessive radioactive release due to cover gas leakage into the Rm from the reactor fuel transfer port adapter portion of the equipment stackup at the port. Radioactivity released from the cooling insert does not exceed the limits set f orth in Sections 12.1.1 and 12.1.2. The pressure in the seal buf fer regions is continuously monitored to ensure continued seal Integrity, in the unlikely (mobilized fuel assembly event seal temperatures increase but remain below the seal limit. One of the sets of cooling insert-to-floor valve seals is in a cooler region than the other set and serves as a backup to the Inboard pair.
9.1.4.8 snent Fuel Shlontna Cask
- Ihe integrity of the SFSC design will ensure suf ficient margins to meet all requirements stipulated in the applicable regulations, especially 10 CFR 71.
The shipping cask is discussed in this section only to the extent that conditions to which it is subjected inside the RSB are potentially more severe than those design conditions specified in 10 CFR 71.
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l Regulation 10 CFR 71, paragraph 71.56, states that the cask design shall withstand a hypothetical accident characterized by a 30-f t drop onto a flat, essentially unyloidir.g, horizontal surf ace without exceeding a specified reduction in shielding and containment of radioactive material. The LMFBR spent fuel shipping cask will be designed to withstand, with no release of radioactivity, a maximum dsceleretion of 123 g if dropped 30 ft onto an unylalding surf ace. The largest height for a potential SFSC drop In the CRBRP is the 72-f t vertical distance of the SFSC handling shaf t.
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the Integrated dose at the same level as the remainder of the reactor head (see Chapter 5.2.1.3).
Activity in the reactor cover gas is contained by plug and cap seals during ;
reactor operation and by adapter and floor valve seals during refueling.
l Under all conditions, redlooctive leakage and diffusion through seals are in conformance with the limits listed in Chapter 5.2.1.3.
Mechanical damage to core assemblies is prevented by control Interlocks :
governing RGT positioning during refueling and the RGT cap locking the RGT in '
place during reactor operation.
9.1.4.9.2 aggIan DesertotIon The shield plug is so designed as to limit the total radiation dose rate at the upper end of the RGT to less than 2.0 mr/hr at a distance of 3 feet from the ciosest accessible surface.
Hermetic sealing is provided by both plug seals and seals In the RGT cap. A means to purge the cap-plug interf ace volume before removal of the cap is also provided. The pressure boundary consists of metallic seals between the RGT and the reactor fuel transfer port nozzle, elastomer seals for the RGT drive shaf ts (static during reactor operation, dynamic during RGT movement for refueling), and static elastomer seals in the RGT gear housing which seal to the RGT cap during reactor operation and are not part of the pressure boundary during refueling. The RGT is sealed to the reactor fuel transfer port adapter cooling insert during refueling as described in Section 9.1.4.7.2. All RGT seals are double wl1L a pressurized, continuously monitored buf fer region between each pair of seals. The buffer region pressure is for leak detection and is not necessa.y for seal offectiveness.
Control logic Interlocks prevent improper sequences of core assembly-RGT movement whenever the RGT is in use. During reactor operation, the RGT end cap locks the RGT in position and prevents all tube movement. Also, no electrical power is provided to the RGT during reactor operation.
9.1.4.9.3 Safety Evaluation The RGT, RGT plug, and RGT cap are so designed that refueling and/or operating personnel wilI never receive a total dose greater than 125 mrem / quarter.
(Actual allowed dose and leakage levels are shown In Chapter 5.2.1.3.)
Double seals and a capability of purging the cap-plug Interf ace volume ensure that gaseous radioisotope leakage from all sources to the head area will never I cause a dose rate in excess of that given in Section 12.
pontrol Interlocks are designed to prevent mechanical damage to core assemblies contained in core component pots (CCP) and reactor components by preventing the following actions:
- 1) Inadvertent attempt to insert an assembly in an occupied location.
- 2) Motion of the RGT with a CCP or grapple extending below the base of the RGT.
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- 3) Any motion of the RGT during reactor operation.
- 4) Positioning of the RGT over any position except one of the storage / transfer locations.
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- 1. FIC Mormal Fuel HandlIna In a typical spent fuel handling sequence, a spent fuel assembly in a core component pot is lowered through the fuel transfer port (see Figure 9.1-7) by the EVTM, into the spent fuel transfer station directly below the port. A lazy susan assembly, with three transfer positions supported by a stainless steel gridwork, provides the storrge locations. Each position holds one fuel assembly, in a sodlue-filled core component pot. Decay heat is removed by natural convection to the FHC atmosphere.
The spent fuel assembly is removed from the core component pot by the In-cell crane, using a gas-cooling grepple, and allowed to drip dry. If for some reason not identified as a part of normal procedures, it is deemed necessary i to remove a sodium film from the exterior surf aces, the exterior surf aces will be wiped with alcohol wetted swabs.
Then the spent f uel assembly is lowered into the spent fuel shipping cask located in a shaf t below the cell floor. The sequence is repeated for the number of assemblies necessary to fill the shipping cask. The above functions within the FHC are performed remotely by operators in the adjacent operating gallery, and can be observed through the viewing windows.
Normal core assembly handling operations in the FHC are conducted with assembiles having a decay haat of 6 kW or less. Infrequently, it may be necessary to (1) examine a f uel ~' s. ; 7,.1. In this event a single assembly may be handled in the FHC with a decay heat not to exceed 15 kW.
- 2. Snent Fuel Examination Spent fuel examination in the FHC is limited to inspecting the exterior surfaces of fuel assemblies to determine their geometrical condition before loading into the spent fuel shipping cask. Spent f uel assemblies will not be disassembled or sectioned in the FHC.
It is planned that only a few selected spent f uel assemblies will be examined, af ter the plant operation has reached its equilibrium. During the first few ref uelings, it is expected that more spent fuel assemblies may be inspected.
The extent of the spent fuel examination covers the following operations, all of which will be performed in the f uel examination fixture:
- 1) Visual inspection of all exterior surfaces
- 2) Determination of axial and radial dllatton of fuel assembly by measuring its lengih and distences across flats
, 3) Measurement of the fuel assembly bow q,l- (,4 W
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I Radioactive releases and con + amination from spent fuel assemblies that are being prepared for shipment in the FHC are contained within the FHC by proper sealing or closure welding of penetrations. Radloective leakage and dif fusion through seals, In the unlikely event of release of the entire fission gas inventory of a fuel assembly, are limited to well below the criteria of 10CFR100.
Criticality of fuel assemblies in the spent fuel transfer station in the FHC is not possible because only three fuel assembly locations are provided.
The spent fuel transfer station design considers all normal loadings in combination with the loeds from en SSE in maintaining the necessary physical separation. The FHC roof closure is designed to absorb the load of the heaviest equipment handled by the RSB bridge crane over the FHC: (a) for the main hook, lowered at the maximum crane speed (5 fpm), and (b) for the auxillary hook, accidentally dropping from the maximum handling height to which it is raised, onto the center of the roof closure without af fecting the integrity of the fuel separation lattice. The FHC is located such that heavy equirment not belonging to the fuel handling and storage system Is not carried ovec It by the RSB bridge crane.
The spent fuel transfer station within the FHC is designed so that movement of the lazy susan will not occur while a CCP is being inserted or withdrawn.
This design condition prevents mechanical damage to the CCP or its contents.
Monitoring instrumentation is provided for the FHC for conditions that might result in a loss of the capability to remove decay heat, and to detect excessive radiation levels.
9.1.4.10.2 Deston Desertation The top of the FHC is located at the operating floor of the RSB, as shown in Figure 9.1-7. Suf ficient shielding is provided so that the radiation level above the FHC does not exceed the radiation Zone I criteria, see Section 12.1.
This shielding is provided by the cell's roof closure assembly, a lood-bearing structure which is part of the RSB operating floor. The FHC side well facing s
the operating gallery is shielded by high density concrete to protect the operating gallery against radiation dose rates exceeding the radiation Zone I criteria. The other walls and the floor are shielded by conventional concrete to protect the neighboring vaults and the spent fuel shipping cask handling corridor against radiation, see Section 12.1. All windows, and port penetrations through the roof, walls, and floor are stepped to limit radiation streaming in the gaps. The main source of radiation in the FHC is spent fuel assemblies in the spent fuel transfer station.
The pressure boundary of the FHC is sealed using weided seams and elastomer seals. There are three types of elastomer seals used: static, dynamic, and infl atabl e. All of these seals are provided in redundant pairs and have essentially zero leakage (l.o., leakage is almost entirely due to permeation through the seal material). The dynamic and inflatable seals have slightly larger leakage than the static seals on a comparable basis. All three types of elastomer seals have a buf fer space between seal pairs. The buffer space for static seals is used for periodic leak testing. Effectiveness of the seals does not depend on presence of a buffer gas. Dynamic and inflatable 43 6 20-
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seals are provided continuously with a buf fer pressure between the double seats. This pressure is monitored continuously for leak detection and is not required to prevent semi leakage, although it would mitigate an inner seat leak. The (nflatable seals are the only ones which depend on a continuous source of eloctrIcal powor and InfIation gas for operation. In case of Ioss of of fsIto power or gas supply, the seal InfIatIon systam valves would f alI open, provid{ng the seals wIth a continuous source of {nfIat1on gas pressuro.
(The valves are closed during normal operation to provide more sensitive seat Isak detection.) Since the supply valvos f atI open, Ioss of of fsito power would not affect seal inflation.
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l The 1Iner seams on the col l Interfor walIs, roof and fIoor, and welded penetrations through the FHC walls, roof, and floor are alphe-tight welded and Inspected. Fuel transfer ports, the 28-In. maintenance floor valve, window seals, utility penetratons, and slave manipulator penetrations each have double, static elastomeric seals. The 28-in. maintenance port FWOR valve also has double, dynamic elastomeric seals. Sealed ccver glasses are provided on the interior side of the window penetrations.
The 28-In. maintenance port floor valve and equipment transfer drawer are sealed by double inflatable seals. The floor valve inflatable seats are active only during periods of valve use. At other times the maintenance port is sealed by a plug with double static seals. The equipment transf er drawer inflatable seals are on the closure doors on each side of the penetration and are pressurized except when a door is open.
The spent f uel transf er station within the FHC is shown in Figure 9.1-8. A maximum of 3 spent fuel assemblies in CCPs can be stored in this interim storage; however, in normal operation, a maximum of 2 fuel assemblies will be stored (1 storage position lef t empty). The storage positions within the transf er station consist of shcrt, tapered, cylindrical inserts at the bottom, the middle, and the top of a substantial supporting steel grid structure.
Each CCP is held in place by the three cylindrical sections. The conter-to-center distance of the three storage positions is about 25 in. Each position can hold only one spent fuel assembly in a CCP.
During loading of the spent fuel shipping cask (SFSC), a cask-FHC seal assembly forms a gas-tight extension of the FHC containment to ilie cask interior. The SFSC is, therefore, connected to the FHC atmosphere and is separated from the air atmosphere of the cask corridor.
The FHC roof closure assembly consists of a large north closure plug (21-ft by 18.5-f t) and a smaller south closure plug (8-f t by 18.5-f t), both joined by a cross beam assembly.
The two closure plugs are composite structures consisting of 34.5-in. thick reinforced concrete, enclosed on four sides by a steel liner, and resting on five 8.5-In. thick steel shield plates. The entire 43-in. thick composite structura provides suf ficient radiation shielding and is designed to support normal structural loads as well as the accidental impact loads given In the design bases.
The heaviest load carried over the FHC roof is the EVTM floor valve (9 tons).
The lif t height of the EVTM floor valve is limited to 2 f t. by administrative centrols. All heavy maintenance equipment is transported by the Large Component Transporter (LCT) between the RSB and RC8. Maintenance equipment welghing more than 25 tons and the LCT itself are handled only by the double i d-657t
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1 reeved main hook of the RS8 bridge crane. Caly maintenance equipment weighing less than 25 tons may be handled by the single reeved auxillary hook. A load dropped from the auxillary crane could Impact the LCT when It is stationed above the FHC roof. Most or all of the impoet energy would be absorbed by the LCT. Selsmic restraints prevent equipment leaded on top of the LCT from toppling onto the RSB operating floor or FFC roof during an earthquake.
Normal operating procedures require large maintenance loeds to be f astened to the LCT seismic restraints before disconnecting them from the RSB bridge crane.
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TABLE 9.1-2A SPENT FUEL ASSEELY CLAD 0 LNG TEMPERATURES (Sheet 1 of 2)
Peak Fuel Location of Fraquency Assembty Cladding Fuel Assembly Cl ass Temperature (OF )
CCP in EVST storage location 20 kW Assembly Normal operation Normal h II WO "D Natural convection loop Unlikely MI) W cooling CCP in FHC (15-kWt assembly)
Argm circulation system Unlikely 1060II) operative Argon circulation system Extremely 1265 Inoperative Unlikely CCP in FHC (6kW assembly)
~1 10 ACS operative Normal ? ^,^^ lL Z. 7 " ^ _ : ! )
ACS Inoperative UniIkely 9000 CCP in EVTM cold wall 20 kW Assembly Cold walI blower on Normal 1240( II Cold wall blower of f
< 30 min. Anticipated 935
> 30 min. UniIkely 1350(II CCP Immobilized in EVTM stack assembly (assembiles below the cask body assembly, see Figure 9.1-14) 20 kW Asseebly
< 30 min. Anticipated 1230
> 30 min. Unlikely 1415(1)
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22
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TABLE 9.1-2A SPENT FUEL ASSE W LY CLADDING TEMPERATURES (Sheet 2 of 2)
Peek Fuel Location of Frequency Assembly Cledding Fuel Assembly Cless Temperature (OF)
CCP Immobil fred in reactor fuel i transfer port 20 kW Assembly Blower on
< 30 min. Anticipated 950
> 30 min. Unlikely 1444(1)
Blower off (< 30 min.)(2) anticinated 970 Blower off (> 30 min.) extremes.p UniIkely 1482 CCP Imobilized in EVST fuel transfer port 20 kW Assembly Biower on
< 30 min. Anticipated <1250 8'4
> 30 min. Unlikely 1389(1)
BIower off (< 30 min.)(2) AntIetented <1250 BIower off (> 30 min.) Extreme @ uni Ikely 1482 CCP Imobilized in FHC spent fuel transfer port 15 kW Assembly Blower on
< 30 min. Anticipated <1225 6
> 30 min. UniIkely 1249(')
. Blower off (< 30 min.)(2) Anticipated 1225 Blower off (> 30 min.) Extremely Unlikely 1275(ab3 (1) Steady-state temperature (2) For a stopped CCP with the f uel transfer port blower inoperative, the operator is required to take action as described in PSAR Section 2.1.4.3.2 within 30 minutes to raise or lower the CCP.
- q. A%
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TABLE 9.1-2B TEMPERATURE AND PRES $URES FOR FUEL HANDLING EQU; PENT Maximum Calculsted Values Design for Normal Operations and Limits Of f-Normal Design Events Pres. Temperature Pressure Temperature (Psig) (OF) (Psig) (DF)
Normal Off Norm. Normal Of f Normal EYST 15 775 10" WG 8 535 6250 (off normal)
EVTM 15 435 10" WG 11.5 40 0 4350 FHC +2.5 225 -3" WG +1.44 110 1850 Inflatable Seals 50 2 50 30 50 ambient
- 2 500 **
EVST Cooling Sleeve -
50 0 - -
110 5000 Seals EVST Port Plug -
250 - -
142 183c Seals EVTM Static Seals -
250 - -
80 2500 RFTP Seals -
50 0 - -
182 2420 FHC Static Seals -
250 - -
110 1850
.* Closure valve and floor valve transfer of a 20KW assembly could raise the seal temperature to 1500F.
- Closure valve and floor valve steady state (more then 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 0 2700F.
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3,4 pipruc Ann monipum smim nun n--;ggL 9.4.1 Demian namin The Piping and Equipment Electrical Beating and Control System provides the electrical heaters, electrical heater mounting hardware, heater power controllers, related temperature measuring and controlling instrumentation, and equipment required to heat the following sodium containing systems and components:
Reactor Enclosure
- Reactor Refueling (Storage. Tank)
Reactor Beat Transport (Primary and Intermediate)
Systems Steam Generation System (Dump Tanks and Sodium Water Reaction Products Tanks)
Auxiliary Liquid Metal System Inert Gas Receiving and Processing System Sodium Impurity Monitoring System This heat is required to preheat those sodium process
. systems prior to fill, to prevent sodium freezing when systems heat sources such as reactor decay heat and pumping heat becomes insufficient, and to maintain pre-established temperature differences in
- The trace heating system which services the reactor vessel head, control rod drive lines, and vessel support area is discussed in Section 5.2.1.6.
-3 the system.
To perform the dry heat-up function, the electrical heating system shall be capable of preheating the sodium process systems from ambient tesperature to any temperature between ambient and a maximum of .
approximately 450oF before the system is filled with sodium. The heating requirements for each trace heated component in the above systems will be determined by the particular sodium process system.
The-electrical 1. eating system shall also be capable of providing the applicable heatup rate for the particular system or components when filled with .
sodium and of holding process system temperatures when filled with sodium. Heat provided by this system can be used to melt frozen sodium in piping or components. Freezi'.g af sodium in major systems or components is consioered unlikely and is an abnormal ,
event.
The temperatures of all measured points shall be indicated locally and in the Main Control Room. The thermocouple used for monitoring shall be separate l
from the theosocouple used for concrol.
! Thermocouples leads as a matter of good design p'ractice shall use different paths from heater power conductors. The electrical heating, temperature monitoring, and temperature control lines are all non-safety related. As such, the electrical heating, monitoring, and control lines shall be separated from all Class IE lines per IEEE std. 384-1981.
All piping and equipment in inaccessible areas shall be provided with spare heaters and thermocouples.
The spares shall be accessible in a junction box in a man-safe location. All heater circuits shall be provided with ground fault protection. All instruments and controls shall be testable in place.
All components of the Piping and Equipment Electrical Heating and Control System shall remain operational for an OBE (i.e. , qualified Seismic Category II). In addition, no component of this system shall, as a result of an SSE, impact in any way the performance
. of the safety function of the piping and equipment heated by this system. The heaters and thermoccuples 4
shall withstand the vibration forces of the
I . -
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l i.
components to which they are attached without failure or inpact on any safety related function.
The heater physical mounting arrangement and the electrical protection of the heater circuitry shall be designed to preclude damage to the components .
being heated. The Piping and Equipment Beating and .
Control System has been designed and applied such that it is not a system which is important to safety.
Safety related components which require the application of heat from the Piping and Equipment Electrrical Heating and Control System have also been designated such that any' combination of failures, 1
single or gross, of the non-safety related Trace Beating System will not compromise the safety related function of the equipment. (e.g., cover gas equalization lines, overflow heat exchangers, ex-vessel storage tank).
9,4.2 system nenerintion The electrical heating and control system provides power to the tubular heaters or mineral insulated
.- (Mi) heating cable mounted on the piping and/or components of the systems indicated in Section 9.4.1 MN ) f d ** h g oc er<4 I.3 4 i.u w,e p .,.,#
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The' heat rates, required b'y different components are
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' controlled by using therwocouples to monitor piping a'nd component tempermtures and to adjust the power ~
m a
supplied to 'the heaters by means of three mode ,
s t\
n proportional temperature controllers and solid state i ,
rela'ys., " "
~~
- 1 ,1 s
e Alarm is provided to the Main Control Room in the /
, e " ,
?
event of thejfollowing:
. s
- 1) Setpoint Deviation (Control thermocouple compared to temperatuse setpoint) '
High Temperature l
- 2) (Monitor thermocouple ccmpared to a high i
- 3) Low Temperature '
- 4) Open Thermocouple (Both control and moniter thermocouples) ?
- 5) Open Circuit Breakers
- 6) Load Loss over 10% (Detects a single heater failure)
- 7) Data'Trananlasion Failure '
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x s ,
i , . .
Tubular heaters apply heat via 'a shiral wourid nickel-chromium alloy resistance wire insulated fron its containing metal tubular sheath by tightly packed magnesia (Mgo) powder. Several inches on each end of i
l each heater are unheated having a heavy electrical conductor to the electrical termination. In certain
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l cases, i.e., selected piping smaller than eight inches 0.D., the heat is applied by mineral insulated heating cable that consists of a metal sheath drawn down over a Ngo insulated single heating element.
Separate Chromel-alumel thermocouples are used throughout the systems for the feedback signal to control the operation of the electric heaters and for monitoring the temperature of the metal boundary of the sodium taining
,,pi p. 44. is grepiping
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equipment. Tar &or d a y 4.
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TIiermocoup e compensation is provided for all thermocouples. -
Thermocouples on piping are locate 6 at a point on the pipe to enable control of the averLge temperature of the pipe within specified limits. On equipment, the thermocouples are located in the spaces between heaters for both monitoring and control purposes. '
Control of any heater or bank of heaters is by automatic control. This control provides for continuous and automatic adjustment ofe heat based on an error signal generated from the difference between the temperature setpoint, as set by the plant 1
. . :..: : :X=:. 7 7-operator, and the temperature feadback signal from the thermocouple.
The controller compares the temperature control setting (ramp rate in heat-up mode and setpoint in hold mode) as set by the plant operator to the actual temperature of the sodium process metal as measured The by a thermocouple and generates an error signal.
error signal is converted into a corresponding "on" to "off" ratio of voltage which is applied to a solid state relay which controls the AC power to the yh:
heaters.
The required poser is controlled by conducting a For fraction of the time over a 17 second period.
example, 50 percent power would be conducting for 11.5 seconds and off 11.5 seconds, 90 percent power would be on for 15.3 seconds, and 10 percent power would be conducting for 1.7 seconds.
Heaters are arranged in a particular control circuit according to the uniformity of heating required by a This type of heat application is bank of heaters.
called zoning. A heater zone is an area that can be
...._....r... . . . . . ... . . . . .
-3 heated with the same unit input and can be controlled from a single temperature indicating point that is representative of the sone.
The temperature feedback thermocouple is located in a representative position within the pipe run or area within the heated zone. The monitoring thermocouples are located l in different areas of the sone from the feedback thermocouple to provide independent checks on the zone temperature.
All heaters are in operation continuously during dry l heat-up (system completely empty). Some heaters will j be in operation continuously for the occasional fill and. drain situations in some piping and components such as cold traps, dump' tanks, ex-containment storage tank guard vessel, gas equalization lines, i
and other components.
For all other normal 1
operations (start-up, hot standby, and shutdown) the heaters will be in operation only intermittently to make up for the heat loss through the insulation.
l l
A dedicated, pre-programmed, direct digital control system is provided for the Reactor Containment Building, Steam Generator Building, an,d Reactor Serivce Building. The system is modular te permit physical distribution of the various functional t
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h D M l% L D,N\MD waTEst sysm Question CEA10.18 (9.7 M 1he normal chilled water systen is a non-safety related system with see sei mic Category I piping. Verify that the normal chilled water syste piping and equipment which is located in cells containing sodia or llaK piping is
- designed to Selenic Category I critaria.
Response
with the exception of the SGB loop cells, all normal chilled water piping and equipnent are located outside the cells containing sodia or llaK piping. .
il a g;t_7 :-'i- wa .._d millisif~~ [ piping is rently liirb i
i o Seismi Category III. s ification currenti ing i evalua to determine Seimnic Ca ry II the ocp cells is o te. She r of this tion will be ovi e I
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.Fof the SGB !~p ~"" .pfhx gog,,,J d,y,j w.sc 9 %
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3 components and to facilitate future expansion and the upgrading.
The panels and Operator Control Centers are located in these three areas of the plant. An additional Opertor Control Center (Naster) is located in the l Main Control Room. The system arrangement is shown in Figure 9.4-1.
9.4.3 saferv Evaluation Since the trace beating system is not important to safety, the heating system components are non-class lE.
Therefore, as required by Federal Regulations, the trace heating system design features which make operation highly reliable are excluded from safe shutdown scenarios. In fact, the safe shutdown scenarios must assume gross combinations of heating system failures. This is required without '
consideration to the design reliability of the non-safety electrical trace heating components.
Under these exteme conditions, the apparent failure mechanisms of the trace heating and control system are lack of heat when heat is required, heat when
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D none is required, and current flow through piping and other non-wiring components due to shorts concurrent with multiple failures of the over current protection components. The effects of these potential failures on the safe shutdown of the plant is discussed in this section.
As discussed in PSAR Section 3.f.3, the Piping and Equipment Electrical Beating and Control System is not safety-related. The heating system is not essential for the safe shutdown of the reactor, nor will failure of the system result in a release of radioactive material. In those cases where heaters are applied to safety related components, the heaters are not required for the components or the associated systems to perform their safety function.
The loss of vent and drain line trace heating does not compromise the safe shutdown function. In the event of a large sodium / water reaction (SWR), the water side of each evaporator module in the affected
~, loop is dumped and the superheater ste.am side is vented from the superheater outlet safety relief valves; therefore, no drain of sodium through the 7 ;; y -y, = u ; e.3_ .;,;;t. ,.. .p .7 y;. ' _y p_ ; ., 3;e. ;: y 7 y,v i ' . % :: . Q
- ci-
. 11 IBTS drain lines is necessary to mitigate the SitR, and these drain lines need not be maintained hot for l
s'afe shatdown af ter this event. If the gas feed and vent lines close, the range of sodium level variation in the PBTS puncps is limited by the sodium level sensing and gas isolation system (qualified for SSE and in-containment sodium fire DBE and provided with ,
Class lE power) to a safe cperating level for shutdown.
O
~....--.-
In additibn. heat tr.wesport systemi temperatures remain above the temperature at which tracc heating 1s Ired to prevent sediana solidiffcatfon during tha time when the Shutoown Heat a1 System operability is required.
normally)' The unwanted additional heating of sodinan Ifnes (sensed and controlled due to sultipie faflures in the trace heating and control systen which ti on trece heaters (which should be off) is less than five percent of the lung term siesystem heat removal capability per loop. The utmeanted heat is less than one s percent of the short terw subsysten heat removal capability per loop, and it is less than one-half of one percent of the total plant power removal capability.
These percentages arv sufficiently senll in tenus of heat transprt system ,
capability that the occurrence of this failure rechanism would not cograntse the sat'n shutdown function.
The third potentfal failure mechanism is a short to a non-wiring
' component occurring with concurrent failures of the ground heater sheath, the eround fault detectors, and the over current protectiva devices. This mechanism would not compromise tM safe shutdesm ef tha plaretr~ Tiin yea 11est pipe where a short could occur is greater than ten times the cross-i,vetional diameter of the electrical wiring. Therefore, for the smallest pipe, the conductivity of the electrical wiring is one-helf the conductivity of the pipe, and the conductivity of the pipe is over forty times higher than the heater wire. In either a short or an arcing situation. the pipe would not fail.
Operellonally, the failure mechanism requires the fatture of the tae-perature sensing system and one of the following: (1) excess current application.
(?) cross-over in sounting of adjacent heaters, and (3) improper setting of pro-tective devices. For design related failures, the failure mechanism can be caused by impropor heating wire design fissures in the magnesium oxide, and bends less than the minimune bend radif. The effect of the failure will not cause failure of the sodium containment.
9.4.4 Desien Rv1,1,9b111ty J. valuation .
In order to prevent the effects of heater failure from propagating to the piping or equipment to which heaters are attached, the following oper-ational criteria are used:
(I) For normal operatton, the heaters are ope sted at less than 1/2 rated power. For abnormal operation. each heater control circuit is protected against overcurrent by i,hennal overload circuit breaker and temperature sensors on the heated compenent.
Ground fault laterrupters (GFI) will be used for protection against ground currents.
(2) Migh and low tamperature alarms arv provided for all control oul monitor thennocouples in all heater tontrol zones.
- (3) The cold ends of the heaters are bent 90* and brought out from the component. A spacing is maintained between adjacent heaters to prevent crwssover of heaters and significant mutual heating by radiation.
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(4) The prepar setting of the Sri units ul11 he set at installation.
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) Tu prevent tweater failure from dm. lyn considerations, the heaters are designed to a high quality standard. The use of um standard etquires that heaters be radiographed. In addition, the technical, anchanical, electrical, matarla1 fabrication, and quality assurance requirements are specified. ,
9.4.5 Tests and Inspections The de: inn of the electrical heatino system permits periodic testing
- tu confirm the operation of (fie gesunJ fault detection system and heater con-trol system. The heater control system will be tutec and inspected at in-sta11ation innd at refueling pc.riods as dictated by application. Inspection et the h6aters in accessible areas following shutJonn mill be performed accord-inJ to the swinterwace requirements of each procen system. Redundant heaters
- wirwi to accessible ters.inal blocks will be provided in inacr.e.xsible areas as Imq requ1rud.p '
9.4.6 leistricientation Applitation Instrumentetton dpplication is discussed in Sectinn g.4.2.
Redundant heaters are nonoperating installed spares which can be made to operate in place of tailed heaters.
L 3.4-F i
6 .
1
- l INSERT C In addition, the trace heating system will be tested prior to startup following an earthquake of intensity greater than or equal to the OBE.
f l
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Y l l
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.. l Table 9.4-1 .-
46
{
SAFETY-RELATED COMP 0NENTS REQUIRING STAND-0FF ELECTRIC HEATERS .
common t . .',. j
! Reactor rd Vessel Ex-Vessel St age Tank Guard Vessel Primary Heat Tr sport System Main Piping Pump Guard Vesse Intermediate, Heat changer Gucrd Vessels
, Appendage Piping Intermediate-Heat Transpo System Main Piping Pumps l ,
Expansion Tanks l Appendage Piping Steam Generator Sy nem Appendage Piping Auxiliary Liquid Metal System -
Overflow Heat Exchanger Primary Sodium Overflow V sel
' In-Containment Sodium S rage Vessel Primary and EVST Cold aps Overflow Line :
, Piping in Reactor d EVST Cavities 46 Piping in Primary nd EVST Cold Trap C' ells Inert Gas Proce ing System i
All Safety lass 1, 2, 3 Components, Piping and Equipment Impurity onitoring and Analysis System All afety Class 1, 2, 3 Components, Piping and Equipment Amend. 46 7 Aug. 1978 9.4-4
....___-___.._---- =
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- e I s etASTER ANNUNCIATORS CONTROL CENTEg MAIN CONTROL A00M MAIN r.t*TRik k_M4 OPC4ATOR CONTROL .---,
g ggig CFMTER (RSS) '
(Qty. 3)
OPERATOR CON N R.
CENTER (SGft) M LCC.s (Qty.11)
OPCRATOR CONTROL CENTER (RCB) TYPICAL OF -
7 LCC's .
3.otAL CONTROL Ct.HTER
'. 2f .
< m. ...
tEATERS AND T/C (TYPICAL)
"s' Figure 9.4-1 Power and Cont.rel Sys. tem Arrangement
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9.6 HEATING, VENTILATING, AND AIR CONDITIONING SYSTEM 9.6.1 Control Building HVAC System 9.6.1.1 Design Basis 9.6.1.1.1 Control Room HVAC System The Coktrol Room HVAC System is a safety related system de-
- signed to provide filtered and conditioned air to the Control Room as required to pennit continuous occupancy of the Control Room and to ensure the operability of all Control Room Equipment and instrumenta-tion under all conditions. The Control Room HVAC System is designed to:
a) Maintain a positive pressure in the Control Room to mini-mize the infiltration of radioactive or chemical contami-
- nation.
b) Prevent the increase of internal airborne radioactivity over the limitations set forth in 10CFR20.
c) Pennit continuous occupancy of the Contro1 Room during all operating modes in compliance with the CRBRP General Design Criterion 17 (defined in Section 3.1).
d) Maintain space tgmperatures within the Control Room at approximately 75 F during all modes of operation.
e) Permit purging of the Control Room following a fire.
f) Comply with the single failure criterion.
g) Operate from the Class IE AC power supply dur~ing loss of offsite power.
r &
9.6.1.1.2 Control and Diesel Generator Building Emergency HVAC Systems .
The Control and Diesel Generator Building Emergency HVAC System is a safety related system designed to provide filtered and conditioned air under all conditions to the Control Building, Battery Cells, Battery Maintenance Area. Upper and Lower Cable Spreading Rooms, Vital AC/DC Rooms and the Diesel Generator Building Class IE Switchgear Rooms. The
/ system proyides the required environment to permit personnel access
.f - during nonnal plant operation and to ensure operability of the equipment i
under all conditions. The HVAC system serving these areas is designed 49 to CA 8 !*
') b d N. p' 0
.N'I
\
. ... .. ~..- - ...
HEPA filters are capable of removing a minimum of 99.97 oercent mennally generated dioctylphthalate particulate of uniform 0.3u t'roplet size at the design flow rate of 8,500 CFM. -
The charcoal filter bed is assumed to remove 95 percent of airborne radioactive elemental iodine and 95 percent of methyl iodine at
- relative humidities below 70% at the design flow rate of 8,500 CFM. The actual tested efficiency of the charcoal bed in removing elemental iodine is 99.9% and 99.5% in removing methyl iodine.
The Filter Ur)it Supply Fans are connected to their respective filter units by a flexible connection. The supply fans are V-belt driven centrifugal fans provided with automatic inlet vanes. The discharge side of each fan is connected to the supply ductwork by a flexible connection followed by an automatic isolation damper. Each supply duct is connected 59 to the corresponding CR air conditioning units.
The 100% redundant return fans are located in their respective A/C unit cells. Two (2) sound absorbers are located upstream of the return 1 fans and the fans are connected to a common plenum by a gravity damper 59 followed by a flexible connection and automatic inlet vanes. The discharge side of each fan is connected by a flexible connection to a discharge plenum which is connected to three (3) branch ducts.
One duct connects with the Control Building missile protected exhaust structure and is provided with an opposed blade damper and two (2) redundant automatic isolation valves connected in series. The second duct connects with the return air damper of the Control Room air conditioning unit.
The third duct connects with the Control Room filter unit. The 59 return fans are V-belt driven centrifugal fans provided with automatically adjustable inlet vanes.
The toilet, janitors closet, and the kitchen are continuously exhausted to the outside of the building by a toilet exhaust fan and a kitchen exhaust fan during normal operation. The discharge of the kitchen exhaust fan and the toilet exhaust fan, each with gravity dampers are joined together into a common exhaust duct and provided with two(2) re-dundant au mt' a,n)pers and is connected to a missile protected exhaust structure. Upo a containment isolation signal, a high radiation signal from the redundant radiation monitors or high levels of toxic chenical or smoke in the main or remote intake ducts, will close the automatic
, dampers./ The toilet and kitchen exhaust fans will te stopped manually.
All cells and corridors served by the Control Room (CR) System maintained at a 1/4 inch water gauge positive pressure relative to the outdoor atmosphere during the normal and accident modes of operation. j
~
Two separate outside air intakes , one main and one remote, are provided for the Control Room. The main intake is located at the SW corner of the Control Building roof at approximately elevation 880'. The remote 49 intake is located at the NE corner of the Steam Generator Building Auxiliary Bay et approximately elevation 858'. Instrumentation is provided to measure b
$ ym)ok cykM4 CYam(
y/ 'd)C. scoui, c anJ c onInHd /E 9.6-4 Eseb als gdg CenEro/s - - - -- - - - - ---
.. . ; L:z.:.:: = = - ~ ~
- pressure sensing device is provided with an alarm set-point to indicate that the differential pressure across the filter is higher than normal. This alarm set-point is selected on the basis that after initiation of the alarm, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> are available to correct the failure without significantly deviating from the system design parameters.
The Containment Isolation Valves and their instrumentation and controls are redundant. The isolation valves are provided with remote position indicators and manual opening devices.
The Below Operating Floor Air Conditioning Unit automatic dampers are provided with remote position indicators and manual operators.
The failure of any damper can be detected, identified and corrected within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. During this time the affected space temperature v'll be maintained below 120,F.
The Dome Recirculating Fans are not required for the safe shutdown of the reactor and maintenance of the safe shutdown condition.
- 2. Loss of Normal Chilled Water Supply The Unit Coolers serving the safety related equipment in the EI&C cubicles are provided with Emergency Chilled Water. During loss of Nomal Chilled Water, the Upset Design Temperature shall be main-tained in these cells to satisfy the Operational Requirements.
- 3. Loss of Normal Power During loss of Normal Power supply, the EI&C cubicle unit coolers are automatically switched to the on-site emergency Class IE AC powersuppig 4 4.
Radioactive Contamination Protection of the RCB Aress The ventilation air quantities for the above and below operating floor areas of the RCB are selected to maintain the radioactive gas concentrations under the 10CFR20 limits.
The source of radioactive concentrations for the below operating floor areas are the probable outleakage of inert gases from the normally inertad cells.
Since the inerted gases are continuously purified by the Cell Atmosphere Processing System (CAPS), the initial airborne radioactivity in these cells is low. The cells are designed with steel liners leak tight penetrations and sealed doors to withstand the pressure,s resulting from accident conditions.
The pressure differential during normal operation is very low, therefore the outleakage is minimal. The ventilation system air quantities for these areas are selected to maintain the acceptable airborne radiation concentration, resulting from the simultaneous design cells. basis pre sure differential and outleakage from all inerted 49 design.
The selection provides sufficient conservatism in the Amend. 49 Apr. 1979
Insert A Atleast one EI &C cubicle shall be capable of operation to perform ,
safe shutdown requirements, and the temperature in all three EI &C cubicles )
will remain below 120 F, even if any one 1E Division of Emergency Power l or one Emergency Chilled Water train is inoperative. Analysis has been I performed that shows that these conditions are met.
1
. , . . . _ _ _ T_ . . _ . _ .
_7 ,
is annunciated in the Control Room. A logic circuit is available to automatically start the standby pap when the operating pmp motor trips or is (
Inadvertently stopped.
9.9.2 Emergenev Plant Service Water System 9.9.2.1 Design Basis The Emergency Plant Service Water System is designed to provide suf ficient cooling water to permit the safe shutdown and the maintenance of the saf e shutdown condition of the plant in the event of an accident resulting in the loss of the Normal Plant Service Water System or the loss of the plant AC power supply and all of f site AC power supplies. The Emergency Plant Service Water System is not used during normal plant operation. The system provides the Emergency Chilled Water System chiller condensers and the Standby Diesel Generators with coc ing water. Additionally, this system provides fire fighting water for the seismically quellfled fire pumps of the nonsodium fire protection system. The Emergency Plant Service Water System includes the ,
Emergency Cooling Towers and Emergency Cooling Tower Basin, as described in Section 9.9.4.
The Emergency Plant Service Water System is designed to Seismic Category I requirements as defined in Section 3.2. Paps, valving and piping required for the safe shutdown of the plant are designed to ASNE Section 111, Class 3 requirements, as def ined in Section 3.9.2. All electric motors serving the system are connected to the Class 1E onsite power supply. In case of loss of plant and of f site power, these motors are switched autcrnatically to the Standby Diesel Generator. The piping and equipment for each redundant loop of the system is physically separated or protected with a barr!er to conform to common mode f ail ure criterion. System piping is below ground between the Seismic Category l Emergency Cooling Tcwer and Diesel Generator Bullding The Emergency Cooling Tower structure is tornado missile hardened as described in Section 9.9.4.1.
j Elahal 9.9.2.2 System Descriotion dAMb- f .ge.d 8mh The E.nergency Plant Service Water Sy em (EPSW) consists of two 100 percent capacity fully redundant cooling loops Each cooling loop includes one circul ating pump, x --M; pr;, one emergency cool Ing tower and associated piping, valves, Instruentation and control s. Figure 9.9-2 shows the various equipments and represents the system component configuration and relationship.
The components served by the Emergency Plant Service Water System are listed in Tabl e 9.9.-3. Design data on the major system components is I isted in Table 9.9-4.
"Upon loss of Normal Chilled Water or upon start of the Standby Diesel Generators, the EPSW pumps, 47,o .i y n , _,- and Cool ing Tower Fans w 11 I autcrnatically start and provide cooling water at 90 F maximum to the I'
9.9-2 knend. 73 Nov. 1982
- ~ ~ *
. . . . . .. ~.
en Or Emergency Chiller Conde rs in the SG8 nd the Standby Diesel Generators in the DGB. The EPSW pump take suction f om the Emergency Cooling Tower
, he :;r:t' ;; basing which located d,;:# h the Emergency Cooling Tower.
Dur)ng sy em ope ation Tn trsir ~eup p s wil transfy waty t.om Jhe ca(nmon orage sin to e redu ant ting asins ft com pnsate y J 6vano tive a d drift sses f r the t ers.
Cooled water from the Emergency Cooling Tower ; Y; basin)( Is pumped via underground supply mains to the emergency loads in the DGB and SGB. After cooling the emergency chillers and the standby diesel generators, warm water 1 is returned, also through underground mains, to the Emergency Cooling Towers.
To account for seasonal temperature variations, temperature control valves served by electro-hydraulic operators bypass a portion of the returnir.g water '
i back to the pump suction. A temperature indicator controller automatically l adjusts the valves as required to maintain supply temperature above 55"F, the
, minimum required for chiller operation.
In addition to cooling the Emergency Chilled Water chit lors and the standby Diesel Generators, each loop of the EPSW System provides a connection to supply water to the Non-Sodium Fire Protection System. The EPSW pumps and the Emergency Cooling Tower Basin are designed to allow fire protection operation while maintaining the capability for supplying 100 percent cooling to the emergency loads. The fire protection pumps are provided with instrumentation that will automatically terminate operation when a prescribed amount of water has been used (see Section 9.13). This ensures that the guaranteed 30 day supply of water for EPSW system operation will not be comprantsed. In addition, this system is connected to the EPSW loops in such a manner as to precl ude a single f ail ure from compromising the capabil Ity of the EPSW system to perf orm its required f unction.
1 9.9.2.3 Safetv Evaluation koo C is su&d % ckss % bwas 3
/ 11 ,
The EPSW system is a Ismic Category I, safety related system designed to have 100% redundan in both active and passive components. The system is provided with AC ower f rom the Cl ass 1E power sources. EPSW Loop "A" is
- supplled from ass 1E Division 1 and Loop "B" is supplled from Class 1E Division 2. This arrangement assures that 100 percent cooling capability will be available even if one of the Standby Diesel Generators or one of the EPSW loops should f all.
j The EPSW system is a f ully automatic system, normally controlled f rom the Main Controf PaneI in the Controf Room. Redundant controf s have been provided that wIlI alIow fulI operation of the system from a controf panel in the Diesel Generator Buil ding.
, Pipe break analysis for this moderate energy fluid system will be provided in the FSAR.
I 9.9-3 Amend. 73 Nov. 1982
_ . .. . c. . ; .- .. a._ i _5 -- .- = . - - - - -- -
.--~ .
bio-i[9 During the initial phase of recovery from an accident, one O Emergency Plant Service Watir loop satisfies the cooling of the Standby Diesel Generators and the Emergency Chilled Water Chiller Condensers.
The Emergency Plan't Service Water System is capable of accomo-
. dating any single component failure without affecting the overall system capability of providing cooling water to achieve a safe shutdown con-di tion. A single failure an'alysis of the Emergency Plant Service Water 59 System is given in Table 9.9-6.
15 9.9.2.4 Tests and Inspections The system components will be tested at the manufacturer's facili-ties, and a complete system t'est will be accomplished prior to plant operation.
The EPSW System does not operate during nonnal plant operations. However, the system, including all active components will be operated periodically during the year in conjunction with the Standby Diesel Generator testing program as outlined in USNRC Regulatory Gaide 1.108. The system can be proven operable at any time by manual initiation. Inservice inspections will be conducted according to ASME Section XI, as described in Section 9.7.2.1.g. In addition, isolation valves and pressure test connections on the supply and return headers in the pumphouses and the DGB permit inservice inspection of the buried piping by hydrostatic testing.
50 9.9.2.5 Instrumentation Application Instrumentation will be provided for local and/or remote (Control Room) indication of the following parameters as indicated:
pump discharge pressure (local / remote) 59l -
diesel generator / emergency chilled water chillers supply temperature (local / remote) 50 -
storage basin level (local / remote) diesel generatqr and emergency chiller flow rate (remote) diesel generator and emergency chiller supply temperature (local) diesel generator and emergency chiller return temperature (local / remote) diesel generator and emergency chiller supply and return pressure (loca1) operating basin level (local / remote) v h.:p-h low)
A flow switch, l'ocated in the return line from each diesel g'enerator and emergency chiller will detect an abnormal low flow condition 43 33 and energize an annunciator in the Control Room.
15l9.9.3 Secondary Service Closed Cooling Water System The objective of the Secondary Sarvice Closed Cooling Water (SSCCW) System is to provide cooling to auxiliary equipment located in A the turbine building.
Amend.MTf Dec. 1980
. 9.9-4
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The Emergency Cool ing Towersgpumphouses, :;;.-d; ., L..... and storage basin are designed to withstand the most severe natural phenomena (e.g., Safe Shutdown Earthquake, tornado, tornado missiles, wind, Probable Maximum Flood '
or drought). The design has the necessary redunda gpone .
Electrical power for the Emergency Cooling Tower fans, pump , and control equipment is provided from the Class IE AC power supply. loop is provided with electrical power from System Class 1E Division 1 ^h: ^' r- im S;c^ -- 0; - -
,2.g3 repM .
9.9.4.2 Desfon Descrintion he The Emergency Cool ing Tower Structure consists 4[e of pumphouses (containing the pumps and piping of the EPSW System, Section 9.9.2) ated directly above the :;:- d'..,__:_. storage basin. The cooling iovers phouses en&
--- ^ ' - ' - ' - ar e ' ^ ^" _ 2_- _ _ . . Selsmic Category l Tornado protected structures. The common storage basin is a Seismic Category I, flood and tornado protected structure. The storage basin has suf ficient storage capacity for 30 days of operation, including 30,000 gallons of water storage for the non-sodium Fire Protection System plus adequate allowance for drif t and evaporation losses. Each cooling tower is designed to achieve pe required heat dissipation rate at any time, approximately 2.36 x 10 BTU /HR at the maximum Emergency Plant Service Water Flow of approximately g gpm4 The change in water chemistry due to the absence of blow-down fran the cooling towers has minimal of f act on operation of the Emergency Plant Service Water
, Sy stem. Proper selection of the Emergency Plant Service Water components, appiled blocide additives, and maintainence of proper water chemistry wilI provide compensation for the Increased tube fouling. The maximum makeup water f required af ter 30 days of operation is approximately 100,000 gallons per day. -
In case the make-up water is not available af ter 30 days, make-up water can be suppi led by either truck, rail or temporary piping f rom the Clinch River or purchased under agreements with the Department of Energy, Oak Ridge Operations.
The top elevation of the Emergency Cooling Tower Basin is 818 ft. which is 9
, l ft. above the probable maximum flood level. The entire basin and the cooling tower supports are founded on slitstone. The basin is a beim grade reinforced concrete structure. For f urther detail s on the basin, ref er to Section 3.8.4.1.5.
Each Emergency Cooling Tower consists of a single cell, provided with an Induced draf t f an system. Each cooling twer is enclosed in a Seismic Category 1, ' tornado missile protecteo structure. The water intake and
& MS -lawers Q 1sQ Bwlg,, et} g a & m h e plQ Ser.hce Gftr Nov l risin 960 Cv'm & hatt C.
9.9-7 Amend. 73 Nov. 1982
c w ,. - w ~ . .
discharge piping are located within the tower or safely beim the ground.for tornado missile protection. The water intake and discharge piping and the
- internal distribution piping are Seismic Category I, ASE Section 111, Class 3 design. -E
- iE-- ; ry T^^* ' ;ie er hee e i;;;!; h nt: 46%
The Emergency Cooling Towers are of a counter-flow, wet-type, mechanically induced draf t design. The Internal distribution piping distributes the intake water evenly over the fill area so that suf ficient water area Is exposed to the counter air flow to provide evaporation for the required heat removal.
The counter air fim is provided by the Induced draf t f ans.
Drlf t eliminators are located above the internal water distribution piping and below the Induced draf t f ans. The drlf t el iminators are a z igzag pattern of channel s which prevent water carryover through the f an stack.
The Emergency Cooling Towers are supported by the reinforced concrete storage basin. The top of the cool ing twers is approximately 44 ft. above the maximum water level of the storage basin.
The Emergency Cooling Tcwer Basin is filled with potable grade water which is treated f or bacteria control. The quality of the stored water is analyzed at regular Intervals and the required blocide additive is injected manually in quantitles required to control seasonal variations of the bacteria growth.
The Emergency Cool ing Towers and Emergency Cool ing Tower Basin will be seismically analyzed as described in Section 3.7.
9.9.4.3 safetv Evaluatfon g j %g M lme i
i The Emergency Cooling Tower steucture j consists of two 100 percent capacity cool ing towers,pumphouses, andAc;enth; 5 :h: and one 100 percent capacity below grade cool Ing water storage basin. The entire __ structure is Seismic Category 1, tornado, and flood protected.[ Piping, associated with the ,
Emergency Cooling Tower is designed to ASE Section lil, Class 3 requirements. '
The structure can withstand the most severe natural phenomena expected, and i other site related events, such that the Emergency Cooling Tower cooling 1 capability is assured under required conditions. The method of analysis is ;
similar to that used for other Seismic Category I structures. The entire i structuro is designed to withstand the Safe Shutdown Earthquake. The fIl1, I drif t el iminatcrs, motors, mechanical drives, piping, electrical conduit, f cables and supports will be seismically analyzed in accordance with the #
procedures discussed in Section 3.7. _.
Nodrss ar e- *NC few/r fcaer Bath
. hsed 5 9.9-8 Amend. 73 Nov. 1982
MSN Freezing of the basin water will not af f ect the operation of the l emergency cooling tower and the emergency pl ant service water I
system. This is because the suction elevation for the emergency plant service water pumps are located at the bottom of the basin which is 39 ft. f rom the surf ace of the colling tower basin water body.
- . . . - . . . . . n.. . .a .
b(0-151 1 The Emergency Cooling Towers e.J epe n t'r; ' '.: are above the probable maximum flood level. The flood level considerations are discussed in Section 3.4. ,.
f- 1
. The Emergency Cooling Tower pumphouses. ::::;t f:r the c.'-vup p; p pite ktch ov+.na anm- +n .1= = tic. 771' 0", are also above the
_ probable maximum flood level. unmou r,+he E.g.oe cy canuma u3ter 50 Wh "; 9 pt are : i.~ aibl: thereby "-^"idia" sy?+a= fland Pa+ :ti0n. - -
2 The Emergency Cooling Tower structure is designed to withstand
( tornado windforces and tornado missiles and the cooling tower internals
, 50 are protected by the enclosing structure. The tornado and wind loadings and the Missile Protection are discussed in Sections 3.3 and 3.5 respectively. .
All materials use' dfor the Emergency Cooling Tower Structure 43 are designed to be non-flammable in order to negate the possibility of
- 50l loss of tr.e cooling function due to fire.
- In order to evaluate the capability of the Emergency Cooling Towers and Emergency Cooling Tower Basin to act as an ultimate heat sink for the Emergency Plant Service Water System for a minimum period of 30 days, a detailed analysis will be done using the following conserva-
- tive assumptions
i 1. The Emergency Cooling Tower Structure is subjected to the maximum
- probable heat load. This load corresponds to the heat removal duty i of the Emergency Plant Service Water System to controi a postulated design basis accident and 's listed on Table 9.9-3. During all other modes of operation the Normal Plant Service Water System removes the heat loads
I 2. The postulated design basis accident is assumed to occur under conditions that minimile the heat removal rate, and maximize the L water usage as follows:
6 L a. Meteorological Condition for Minimum Heat Removal Rate.
The meteorclogical condition for minimizing heat removal rate
- is the highest wet bulb temperature that may occur at the
- inlet to the cooling tower. Wet bulb temperature is the only meteorological condition significantly affecting the
. water temperature produced by mechanical draft cooling towers.
1 Each Emergency Cooling Tower is designed to dissipate the maximum expected heat load during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a design basis accident assuming average wet bulb temperature l for the worst day of record.
. 43 Amend. M -
! '3 -
June 1979 4
9.9-9
.F _ _ _ _
J J * . -1
. 1 15 -
TABLE 9.9-3 COMPONENTS SERVED BY EMERGENCY PLANT SERVICE WATER SYSTEM -
Component Location Component Service Requirements I
Flow btu /nR :1 Bldg. *5wT Component Cell Elev. GPM F (X106 ) ll do S f*f -
Standby Diesel Generator A 816'-G" DGB -M00 TED 90" 43-? TAD I[
Standby Diesel Generator B DGB 816'-0" itrJgTBb 90 M.-e-Tab !
1 rdsdby Diesef Gener4w C- DG6 T69 914Lo" qsp 90' . rap .. . I F Emergency Chilled Water .
System Chiller A SGB 216 733'-0" 2100 90" Max. 10.5 :q Emergency Chilled Wat:..-
50 System Chiller B SGB 217 733'-0" 2100 90" Max. 10.5 IN b . ? -
366 T60 7476~ Tpp 9,. p t
bic$ W Geoim, Okir S '
e+1!o
St D oc n % % t.we o.orr c.
- Entering Water Temp. p 5
L \
S a
n?R P
P
$N m-
. . . . . - . . = ._. . ,
(
15 TABLE 9.9-4 EMERGENCY PLANT SERVICE WATER SYSTEM MAJOR COMPONENTS Design Data For Description Quantity Each Component 59l Emergency Plant Service .2 20^^ 07;; TSO 6pf 51 Water System circulating pump 4%-hv t:t:1 'OM 76D N. 7iE/
4"M d48 Q 3 33 Emergency Cooling Tower W8 2 -3600 GPM: 779D 6Pn1
- .. . ,m ,7- *hM -'- - h -
n; Or, SE" iter ".:ke-Up " ;-3= 92 e , ;;;;', 5 : s Emeqcae l % f Serance l yy gpm O'* R.75fnl Y
.p?mp}-Loo,oC. % ,
\
43 33 .
<5me w Cooling78tur- O [ TBb GPA1 Amend. 59 9.9-16 Dec. 1980
_ _ _ _ _ _ _ _ _ _ _ _ _-- --------------- _-------_--_ -- ---- d
l 9.10 COMoRESSED GAS SYSTEM The Cmpressed Gas Systems discussed in this section are those which supply l Instrument air, service air, hydrogen for generator coollng, and carbon dioxide for generator purging. Instrument and Service Air Systems are depicted in Figure 9.10-1. The Compressed Air System Saf ety Class Ill Instrument Air Supply is shown in Figure 9.10-2. A final drawing of this system will be provided upon completion of design. Too Hydrogen System is shown In Figure 9.10-3, and carbon dioxide system in Figure 9.10-4.
9.10.1 Service Air and Instrument Air Svstame 9.10.1.1 DesIon Basis The Service Air System is designed to provl<$e clean, oil - free, compressed air which will be used to:
- 1) Provide air necessary for various maintenance functions.
- 2) Provide air to required stations for personnel breathing where respiratory protection is required.
- 3) Provide air for the Instrument Air System.
To f ulfill these requirments, the system components and piping are designed,
. f abricated and Inspected in accordance with applicable codes as f ollow's:
o Air receiver tanks, filter bodies, drying chambers, moisture separators, Intercoolers/af tercoolers meet ASME Boller and Pressure Vessel Code, Section Vill, Division 1.
o Piping (except containment penetration piping and Isolation valves, Instrument air piping and accumulators serving safety related components) meets ANSI B31.1.
o Containment penetration piping and Isolation valves meet ASE ",ection li t, Class 2 and Seismic Category 1.
o Instrument air piping and accumulators serving safety related components meet ASE Sect!on iII, Class 3, Setsmic Category 1.
o in addition, service and instrument air equipment piping and components meet WARD-D-0037 (Appendix 3.7-A), Sels.mic Design Criteria for Clinch River Breeder Reactor Plant".
G h9+s ame & $~ne' S Sbbnv MMIS AUSS NC 'll. /-
En*vironmental design reqtt'reme ts will meet those for saf ety related equipment discussed in Section 3.11.
(
9.10-1 Amend. 66
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process 1ag. Non-radicactIye dralnege Ie pumped to the eqvellaatIon ponds of the Wastewater Treatment System. A power f ailure to the redlaiton monitor or [
diversion valves will cause recirculation back to it e sep to prevent radiceetIye dralna enteriog the non-radtosct!ye wastavster 1rsetment stem. A wuan mee..By-c.hssJ vehm. k leW alema4<uAm of dg *(# -
e, 4 Wars 4a ved d.. % amey # he me plsJ yvie. 4 p=g03 4e Treated water end other process water treatment wastes which do not have the N %*F #'
potential to be radioectively contaminated, are routed to seperate seps for teensport to the weste waicr trsatment syates. .
)
Where there is a pof antial for oil spills, the dralpage is routed to the oil separation system prior to discharge into the waste water disposa; system.
Oli sp1 tis are not alicwed to drale, in areas that contain radioactively contaminated equipuent or fIutds, in thIs case, the oII sp1II Is contamineted !
with curbs and dikes and removed manually. 0!! routed to the oil separation t system is collected in a vaste oli tank and removed f rom the site for l subsequent disposal.
I 9 .15.3 Irfety I' valuation The plant equipmerit and floor drainage system is designed so that It Is not reasonably possible for any radioactive drainage in these systems to be discharged out of the plant wIthout undergoing the required treatment or processing.
Evaluations of radiological considerations f or normal operation and postulated -
spills and accidents are presented in Sections 11.2.5 and 15.0 respectively.
The plant Equipment and FIocr Dralnage Systems Is net sof ety rolated except for the piping and valves regulred for containment isolation (Section 6.2.4).
EFDS piping within areas contalning saf ety related equipment is supported with Selsmic Category I supports.
Thero are no drains in cells where soditm piping or aquipment cortaining soditsn is located, accordingly sod!Lan leaks cannot enter the equipment and fIoor dealnage system.
A water pipe break or fire protection system drainage load cannot enter cells or ccrnpartments containing sodlum from drain system backf tcw because these cells do not have any drains. The QERP design criterla requires that three passive barriers (or two passive and one active barrier) exist between alI nodlin and water boundarics. Accordingly, leak detectors located in ibe drainage system are not required.
$afety related systenscontaining water have instrumontatf or to detect leakage.
9.15.4 Tests and Insnactions EFDS pipes embedded in concrete are leak tested prlw to the pourtn3 of ccncrete. All EFDS piping is tested ior le.aks af ter Installation. Ail leaking pipes or joints are repaired before the concrete is placed. The piping wIlI be cleened out to insure that construct ton debris wilI not cause a .
blockege or reduction in the ficw. All pumps are tested to ensure ttsat their j gg ~4,, Amend. 73
.___---__-.-_----..-__-_.____-.__._-A
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ATTAOtENT I l PSAR REVISION: S0DIUM WATER REACTION IN LCCW 15.7.3.7 Sodlum-water Reaction in t.orge th aonent cleaning vammai 15.7.3.7.1 Identification of r'annen and AceIdent DeserIetIon A sodim-water reaction accident in the Large Component Cleaning Vessel (LCCV) would be caused by unplanned introduction of liquid water which would react with bulk sodia prior to completion of the WN phase of the sodim renoval process. The consequences of the accident would depend on the amount of soditsn on the component in the LCCV and the geometry of the component. This analysis assumes that the component having the largest initial sodium inventory is being cleaned. The frequency of soditsn removal from components having enough soditsn to make possible a serious sodium-water reaction is very low. This, together with the design features which prevent such a reaction, makes this an extremely unlikely event.
in the normal soditan removal process, all soditan except anall enounts isolated j in crevices is removed during the WN cycle by the following reaction:
( Na + H2 O = NaOi + 1/2 H2 l
Monitoring to determine completion of the sodita-water vapor reaction in the WN cycle is accompiIshed by measuring hydrogen concentration In the gas leaving the LCCV. The reaction rate Is controlled by establishing a water vapor concentration In the WN entering the LCCV to limit the exhaust gas hydrogen concentration to less than 45. As sodium is removed, the reaction rate and resulting hydrogen concentration decrease for a flxed inlet water vapor concentration. To maintain the reaction rate, the water vapor concentration is gradually raised to a maxistan of 155. The reaction is considered complete when the hydrogen concentration in the exit gas then f alls below 100 ppm.
- The rinse cycle in the normal sodium removal process removes the Inert reaction products of the WN cycle. This will not normally involve significant chenical reactions. Presence of the enounts of sodita necessary for a significant chemical reaction could occur only as a result of initiating the rinse cycle without performing the WN cycle. The design includes an interlock to prevent this error by preventing opening of the water supply valve until 24 hr af ter opening of the steen supply valve for the WN cycle.
The interlock can be bypassed by use of a key switch whose key is kept under supervisory control. The accidental addition of water while all sodita remains on the component is the worst possible case and is analyzed for the soditsn-water reaction accident.
I Sodita in the LCCV prior to the WN cycle would react with water during tho accidental rinse cycle by the same reaction as in the WN cycle. The hydrogen and heat generated would result in high pressure and temperature in the 1
-7 ,
........-._ -- . . a.
vessel. This would promote the additional reactions lIsted below; however, the reaction of the above equation would be predominant and was used in the analysts of this event.
Na + Na(H = 2Nat 1/2 H2 2Na + Na(H = Na0 + nan NaH + H2 O = NaGI + Hy Na0 + H20 = NaGI The sodium of Interest for this analysis is in the form of frost deposited on parts which have been in the cover gas space above the r it was assumed that the reaction is Instantaneous when water reaches ;
Many components will use the LCCV for sodim removal; however, all except two components, the Intermediate rotating plug (IRP) and the small rotating plug (SRP), contain a quantity of sodia for which complete Instantaneous reaction l with water would result in an LCCV Internal pressure less than the 15 psig ;
design pressure. The design of the SRP is similar to that of the IRP described in the next paragraph. The event for the SRP would be the same as for the IRP, but the enount of sodlum involved would be less by a f acter of about six. Also, the SRP is expected to be cleaned only once per 30 yr. the Therefore, the sodlum-water reaction with the same f requency as f or the IRP.
IRP is an e' . eloping event and was the case analyzed. !
The IRP consists of a series of horizontal plates supported by four colens :
supporied by the rotatable plug which is part of the reactor vessel closureit and the ;
head. The suppressor plate is the lowest plate, ;
Its support colens are immersed in the reactor sodlum pool during operation The and will have a 0.003-in.-thick film gf sodlum when removed for cleaning. i area of the plate is about 47,000 in. , giving a sodle content of 4.5 lb. l The lower 36 In. of the support colans will contain another 2 lb. of sodium l film. The next plate, 48.7 In. above the suppressor plate, is the lowest h separated by 1/2of ,
the reflector plates. There are 20 reflector ple Each has a coating of l In. and having a surf ace area of about 36,000 In.jos,9 eac .
frost deposits consisting principally of sodlm but also containing Na 0 and NaH.
The thicknesses of these coatings range frcen 0.0445 in. for the Bottom ;
plate and the upper section of the support columns to 0.0005 in, for the top In this analysis, it is assumed that these are the l reflector plate. The lowest reflector plate contains about thicknesses of solid sodlum film.The next two higher platos containThe total40 lb.
51 and '
65 lb. of sodim.
sodium on the IRP is about 350 lb.
The sodim-water reaction event A flow would begin of nitrogen with at 50 cfm thewouldaddition of water to the be maintained LCCV at a rate of 125 gpm.
through the water into the LCCV and out through the vent to maintain a purge of the system. The nitrogen would carry over water droplets which, together with the water vapor above the water surf it isace, assmed, wouldhowever, react withthatthe no sodle at a rate comparable to that in the WVN cycle, l sodle is removed by this reaction and that it all remains until the water l
reaches it. When the water reaches the suppressor plate, the 4.5 lb. of 2
I
,.~.a..-- i. -
sodim on it will react instantaneously. The resulting pressure in the LCCV will be less than the LCCV design pressure of 15 psig. The hydrogen concentration in the LCCV nitrogen will be 2.5%, which is less then the 45 ennunciator and Interlock setpoint. It is assumed that water addition will continue at 125 gpm. The water level will rise at about 1-1/2 in, per min so that about 30 min will be required to reach the lower reflector plate. During this time, the hydrogen from the suppressor plate reaction and the slow I roection with support colan sodle will be purged fran the LCCV.
Water and the 65 lb. of sodium on the lower reflector plate will react when the water level has risen to the plate elevation. The pressure and hydrogen concentration in the LCCV gas space wil.1 Increase. At a pressure of 8 psig, an Interlock is activated to close the rinse-water Inlet valve. At a pressure of 16.5 psig the LCCV pressure relief valve will open to vent the gas into the Large Component Cleaning Cell. The maximum pressure which would be reached without venting would be 89 psig. This is lower than the burst pressure of all components of the system, so that the hydrogen-nigrogen mixture will be contained except for venting through the pressure relief valve and the normal system vent to the H&V System. The hydrogen concen* cation in the LCCV gas w il l be 225. The Increase will be detected within a few seconds by the hydrogen analyzer in the LCCV vent line. When the detected level exceeds 4%
an interlock will be activated to close the valve In the water inlet line.
This Interlock provides backup f or the high pressure interlock which closes the same valve.
The hydrogen-nitrogen mixture which is vented through the LCCV pressyre relief valve is mixed with the air at ahnospheric pressure in the 67,000 f t cell.
The pressure resulting from adiabatic expansion of the mixture into the cell is about 2 psig. The hydrogen concentration in the cell af ter mixing with the air is about 2.5%
15.7.3.7.2 Analysis of Effects and Consecuences The sodlun-water reaction described in the above section is an extremely unlikely event because the two omnponents with which it could occur are each cleaned only once In 30 years, and because of the number of f ailures which must occur to pennit the event. The principal f ailure would be in not completing the WVN cycle before adding water in the rinse cycle. An Interlock requires that the WVN cycle must be started by opening the steam valve and must proceed for 24 hrs bef ore the water Inlet valve may be opened without using the key switch Interlock bypass. Control of the key by supervisory personnel will evold leproper use of the bypass. Once the WVN cycle Is begun, i f ailure of a second Interlock would be required to tenninate it bef ore the i hydrogen concentration in the exhaust was less than 100 ppm. This low hydrogen concentration ensures that much of the sodium is reacted even if the Inlet water vapor concentration is not raised to the normal 15%.
l i
Analysis of the event hypothesized the Instantaneous reaction of the 65 lb of sodium q the lowest reflector plate. The reaction releases hydrogen into the 2,100-fr nitrogen gas space above the water level and releases heat. It is assumed that all heat from the reaction goes to heating the nitrogen and the reaction products (NaOH and H ). Due to this heating, the gas space above the water would be pressurized to2a maximum of 89 psig, which is less than the static rupture pressure of all components in the system. It Is assumed f or 3
Cell.
the analysis that all of the gas is released adlebatically into the LCCV The resulting cell pressure of about 2 psig is less than the cell design pressure of 10 psig. It is also assumed that there is complete mixing of the vented LCCV gas and the LCCV cells air atmosphere. The resulting hydrogen concentration of 2.55 is less than the 45 explosive limit of hydrogen in air.
Since there is no designed vent between Cell 125 and the RCB atmosphere, the eerosol would be confined in the colI. Since the pressure in the colI is only 2 psig, there would be minimal leakage past the cell penetrations seats and by the tipthis leakage works its way up to the R(B atmosphere and through the system , 4 CB filtration system, the Impact on the site boundary would be negligible.
15.7.3.7.3 conclusion Based on the analysis described In the preceding sections, It is concluded that the vessel and system design is adequate to protect the plant and the public, and that there are no adverse consequences to the health and safety of the public which would result from this accident. Specifically:
I o An uncontrolled sodlun-water reaction In the LCCV is an extremely uniIkely event.
) o The LCCV pressure relief valve is set to vent the gas to the cell at e
! pressure 10% above the design pressure of the vessel.
o Failure of the relief valve to oper. will result in a maximum pressure of 89 psig. This Is less than the calculated burst pressure for the LCCV and conr.ected process equipment.
o Release to the LCCY cell of all reactior. products will pressurize the cell to only 20% of the cell design pressure of 10 psig, o There will be an Imperceptible Impact on the site boundary dose.
l
~
i
- ~ - - . ~ - . . . . . . . . . ,
i 15.7.3.7.4 Enveloping other Sodium-Water Reactions l To envelope the site boundary dose of all other sodium-water reactions in any of the cleaning vessels, calculations were made assuming 100% of the radioactivity deposited on the IRP to be released via a hypothetical vent from the LCCC to the RCB HVAC and thus to the environment.
Activity content of the assumed release was derived from information in PSAR Table 11.1-7 decayed for 10 days. Such a release would isolate the RCB and the postulated effluent will pass through the filter system before release to the outside environment. A decontamination factor of 20 for iodine and 100 for particulates was assumed in the analysis. The activity is conservatively assumed to be in the form of a " puff."
Table 15.7.3.7-1 provides the resultant doses from this set of conservative assumptions and event. All doses are well within the appropriate requirements and guideline values of 10CFR100.
[
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Question C$410.19 (0.7.3)
! . )
The normal and emergency chilled water systems provide cooling for plant WAC systems. WAC units serv;ng areas containing sodium or NaK are provided with drains to carry away chliled water leakage to prevent moisture carry-over In the HVAC ducting. Lesk detectors are provided in the drains to detect chilled water system cell fat ture. Activation of the detector results in automatic closure of the chilled water coil Isolation valves. Justify h e use of non-saf ety related normal chilled water system piping and valves in HVAC units serving areas containing sodia and NaK. .
Response
With the exception of the SGB loop cells, the HVAC units provided with normal chilled water and serving areas containing sodium and NaK are located outside the sodium and NaK cells. These cells do not require saf ety-related cooling.
Accordingly, thelr associated HVAC untts are ciassifled as non-saf sty-related.
For the SGB loop colis. Three barriers between the sodium and water are l provided as follows:
a) Chilled water piping walls i b) HVAC equipment walls which serve as spray shields c) Sodium piping walls
< The fety ci ssificat n of these ba lors is rently being eydluated to d ermine I they pro de adequate otectio ainst a sodlum/ water reacti .
e resul of this valuation wt be provi ed in a futurs amendment.
/
QCS410.19-1 Amend. 69
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