ML20064J603
| ML20064J603 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 01/12/1983 |
| From: | Longenecker J ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT |
| To: | Check P Office of Nuclear Reactor Regulation |
| References | |
| HQ:S:83:183, NUDOCS 8301180136 | |
| Download: ML20064J603 (37) | |
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Department of Energy Washington, D.C. 20545 Docket No. 50-537 HQ:S:83:183 JAn 1 a 1983 2
Mr. Paul S. Check, Director CRBR Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Ccuck:
ADDITIONAL AND REVISED INFORMATION REGARDING THE PLANT AUXILIARY SYSTEMS, PRELIMINARY SAFETY ANALYSIS REPORT (PSAR) CHAPTER 9
Reference:
Letter HQ:S:82:148, J. R. Longenecker to P. S. Check,
" Additional Infonnation Resulting from December 15, 1982, Meeting on Plant Auxiliary Systems, Preltminary Safety Analysis Report (PSAR) Chapter 9," dated December 20, 1982 Ir. response to comments from the staff reviewer of PSAR Chapter 9, the enclosed pages should be inserted as new or replacement pages to the reference letter. The pages are separated and iderttified as attachments to the enclosure with appropriate instructions.
Questions regarding this submittal may be directed to Mr. D. Robinson 1
(FTS 626-6098) of the Project Office Oak Ridge staff or Mr. W. Murphie (353-5313) of my staff.
Sincerely, J n R. Longene r
Acting Director, Office of Breeder Demonstration Projects Office of Nuclear Energy Enclosure O I h
cc: Service List Standard Distribution Licensing 3istribution qO l
l 8301180136 830112 PDR ADOCK 05000537 l
A PDR
ENCLOSURE ATTACHMENT 1 Replacement pages for the first, third, and sixth pages of Section 9.1, Enclosure 1 i
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a Tsar se.ew ci.1 RESPONSES TO NRC COMMENTS 1.
Cor,,ent:
Equipment with active cooling (i.e., EVST, EVTM, FHC, and f uel transf er port cooling insert) should include diesel power provisions or otherwise satisfy clad temperature limits f or loss of of f site (normal) power as an anticipated event; the PSAR is unclear with respect to applicability of such a requirment.
Resoonse:
Fuel Clad Failure er.d subsequent f ission product release will result in site bounda y doses well below estabi;shed limits as discussed in F-SAR Chapter 15.5.
Cooling loops supplied with backup electrical power by diesel generator are provided f er EYST sect um and FHC argon cooling.
The forced convection cooling system for the EVTM is supplied with normal electrical power but is backed by a natural convection cooling system which can maintain the cladding i
temperature within its limits.
The FHC cooling grapyh '
["
blowers are supplied with norma! O!00+r!ce8 p^eer. ' h-ths event-cf a eWeMed 10=> vi pc-er wh!!e her.d!!ng : barc fuci essembly ln Gu FHC, ihe ci addi ng migh i isa hemied is--the-poht-of-fel-hsre.
F+ssJea produc-ts-rs sesed wewl d be-
.ccate! ed 'n the - FMC-becevce +he d!ess!-pawer-supp14ed-arg>n circulatJon syste woul d mainteir. ths ".lC pressure negethe-rel-ethe-to-sum >uMI+g-erees;- Diesel power (from one t
diesel) is provided to the FHC cooling systems to minimize l
exposure to operators on a loss of of fsite power.
- However, l
the FHC boundary is not considered safety-related and credit is taken only for the safety-related RSB confinement io llelt the release.
The reactor, EYST, and FHC f uel transfer ports have cooling capability provided by blowers supp!!ed with normal electrical power.
In the event of a Ioss of this power and Immobilization of a f uel assembly-containing.: ore component pot (the EVTM grapple drive is also supplied with normal electrical power), the peak cladding tamperature remains below the clad temperature limit for anticipated events.
In any case, emergency power is not required because i
in case of power f ailure, the manual drive capability of the l
EVTM can be used without electrical power to raise or lower a l
core cceponent pot to a location in which it is passively cooled.
The enclosed markup of the PSAR revises Section 9.1 to l
ciarIfy the type of electrleal power supplled for each situatloa in which cooling is needed and the consequences of loss of nonnel power.
The revision consists of a new Teble 9.1-2A to list the peak f uel assembly cl edding temperature f or loss-of-power cases and text in the descriptinn of each applicable f acii tty to describe the power supplied and to ref erence the new table.
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% c "144C. c.oel3 circut ation system (ACS), which removes heat f rom the FHC 0
[h argon atmosphere to maintain the cell pressure negative 6"*h m, i,t. a relative to the pressure in surrounding cells.. The redundant
-6 Re. W C cMccrb ACS loops are supplied with power from a standby diesel -
- risc.,c. Acs cve <h generator in the event of loss of of f-site power.rThe postul ated two-hour. station bl agkout l'ncludes loss of this s g. a,~, 9, p-back~up diesel power.
The ACS'IroufdTo IUnidf operate to
%,,. c m,q wa remove f uel assembly decay heat, and the temperature of the
.g c, gg FHC atmosphere would rise. The FHC pressure would become
.;,gg,4 g, positive relative to the pressure of surrounding cells. The FHC liner would remain intact; however, there would be some leakage f rom the FHC to the atmospheres of adjacent cells.
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. The conservative assumption is made that the FHC liner would provide r.o holdup of fission products.
The second system which norr.at ly operates to minimize radiation re! eases f rcrn the FHC is the reactor service buil di ng (RSB) ventil ation 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 f rcrn the FHC Into the RSS Interior are conservatively assumed to be released directly to the atmosphere.
The building structure is assumed to provide no holdup of fission products released during a station blackout from a f uel assembly in the FHC.
All noble gas fission products would. thus be released directly to the environment. There woutd, however, be plateout of volatife fission products on the reic+Ively cold surf aces of the FHC and the RSB Interior.
it is assumed that b0% of volatile fission products released f rcan a f uel assembly would be plated out bef ore release to t
I the envirorrent. This f actor is consistent with the guideline value for lod!ne releases frcrn LWR design basis accidents used in NRC Regulatory Guide 1,.4, " Assumptions Used f or Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident f or Pressurized Water Reactors."
The 50% f actor is conservative in that it does not censider f ormation of Csl in the oxygen-depleted atmosphere cf the FHC.
This reaction would lead to a higher rate of removal for ceslum and lodine particulates penetrating the FHC liner.
The release of particulate f orms of these isotopes would be expected to be reduced to less than 10% of the totM crnount released f rom a f uel a,ssembly Instead of the 50% asstuned.
The analyses were carried out using the SIROCO eerosol l
generation code and the procedure in NRC Regulatory Guide 1.25 to determine the Integrated radiation doses at of f site t
locations.
The Integrated doses to the whole body and to designated body organs are listed in Table 1.
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Insert A The AHM also has inflatable seals on the closure valve, one on top of the gate and two on the bottom of the gate, supplied by gas bottles or I
the AHM, The gas supply is adequate to maintain the seal inflation for at least two hours.
(Note: any failure of the AHM inflatable seal system is enveloped by the accident described in PSAR 15.5.2.4)
The AHM and EVTM floor valves, located at the reactor, EVST and FHC i
during operation of refueling equipment, receive electrical power from the EVTM or AHM as appropriate.
Inflation gas is supplied from the inert gas receiving and processing system. Prior to motion of the respective machine from tne floor valves, the inflation gas is locked into the seals by the respective control valves in the floor valve.
Upon loss of gas supply, the inflation gas is locked into the seals i
by a check valve until the control valves are closed. Leakage rates will be low enough to maintain the seals adequately inflated for two hours. A single failure to one inflatien system, i.e. failure of the control valve, will only disable one of two redundant seals.
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ATTACHMENT 2 Insert pages at the end of Section 9.1, Enclosure 1 i
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Enclosure 9.1 - 12 Comment Provide additional justification why the Fuel Handling Cell cooling system and boundary are not safety related.
Response
Off-site doses from a combined argon cooling system failure and cooling grapple blower failure with a bare core assembly in the FHC are enveloped by the accident in PSAR Section 15.5.2.3 and the RSB fuel handling accident margin source term.
Therefore safety related FHC equipment is not required to support the safety analysis of PSAR Chapter 15.
The RSB HVAC cystem described in PSAR 9.6.3 is designed to mitigate the consequences of an RSB fuel handling accident margin source term.
This margin source term is a 20kw fuel assembly with a release of 100% of fission product inert gases, 100% of halogons, 1% of other fission products and 1% of Pu.
The off-site doses from this fuel handling accident margin source term are:
site Boundary Dose 0-2 hr Whole Body 1.27 Rem Thyroid 64 Rem Lung 2.4 Rem Bone 1.3 Rem j
Low Population Zone Dose 0-30 day Whole Body
.57 Rem Thyroid 25 Rem Lung 1.1 Rem Bone
.81 Rem On site doses from the above FHC cooling system and grapple blower failure accident have been evaluated to confirm that operator action can be taken in the RSB to further mitigate this accident and to operate other equipment.
For a 15kW bare assembly in the FHC, the RSB would have to be evacuated or breathing apparatus donned approximately 10 min after a loss of argon cooling system plus failure of the FHC cooling grapple blower.
Dose to an operator with breathing 1
. apparatus in the FHC gallery would be.43 Rem in the first hour, 1.2 Rem in the second hour and the dose rate will not exced 1.65 Rem /hr.
For a 6kW bare assembly in the FHC, the RSB would have to be evacuated or breathing apparatus donned approximately 30 min after a loss of the argon cooling system plus failure of the FHC cooling grapple blower.
Dose rate to an operator with breathing apparatus in the FHC operating gallery will not exceed 1 mrem /hr.
Even upon the loss of offsite power, actions could be taken to return the fuel assembly to the core component pot by remote-manual operation of the FHC crane.
The above dose rates will allow this effort to continue until the fuel assembly is in a core component pot where it can be lef t unattended.
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Coment 9.1-13:
Provide seismic classification of brakes on hoists of the EVTM, IVTM, AHM and FHC crane.
Response
l The brakes on the hoists for the EVTM and FHC crane are Seismic I.
The brakes on the hoists for the AHM and IVTM are Seismic II, but have been analyzed to confirm that brakes will engage during an SSE.
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r-l ATTACHMENT 3 Insect page at the ered of Section 9.5, Enclosure 1 O
e-.
Question 9.5-5 Where in the HVAC system does the nitrogen bleed go to from inerted cells in the RSB?
Response
The N distribution subsystem exhaust is sent to HVAC unstream of the safety-p related radiation detectors and the RSB cleanup system as discussed in Section 9.6.3.
Upon defection of high radiation, the RSB HVAC system switches into the " recirculation" mode.
RSB cells not within the confinement boundary are automatically isolated from the RSB cells located within the confinement bounda ry.
The nitrogen filled RSB cells will be automatically isolated from i
the HVAC exhaust system in this recirculation mode.
Thus, safety related diversion of the cell N2 exhaust system to CAPS is not required.
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ATTACHMENT 4 Replacement page for the first page of Section 9.16, l
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5 closure l hiFff Seck/dY)
N Sns11on 9 16 -Quss1Lon 1 The design temperatures and pressures of the subsystems should be made the same as those of the cells which they serve in order to onsure that a sodium or a NaK leak in a cell will not rupture the gas cooling system, even assuming that an isol ation valve f all s to close.
Added assurance of cooling system integrity will preclude opening a path for combustion product release of air in-leakage to the liquid metal.
Essnonsa The design pressure of the subsystems is at least equal to the maximum cell design pressure.
The piping and system ccmponents are located outside the cells cooled, and thus are not tirectly ex.cosed to the cell environment unless an isoletion valvo faits to close.
The piping design temperatures _
used are based upon
"~~~- the maximum pipi h i omponent cult temperature due to Na/NsK leak, the ioca! Ion is rela 1ed to cell, whether natural or forced circulation i s p r s. s e n t, thermal inertia of the system, and thermal conductance of the piptog system.
In all cases the design temperature will be equal to or greater than the maxtmum sypecied temperature.
Ses112n 2.16 -Quss119n_2 If co n t r' o l red cooling from sut system CR is required to ensure a safety function 15en ihet subsystem should be safety class 3, and a minimum of seismic 11 (also ASME code Ill, class 3).
Bukaansa Primary Control Rod Drive Mechanisms are cooled by nitrogen gas, supplied by Subsystem CR of the Rect rcul ating Gas cooling System.
The effect of a failure 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 letier HQ: S:62:107, 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 prototypic hardware in 10000F sodium flowirig at the design flow rate of 45,000 lbs/hr.
The PCRCM stator temperature is normally measured by redundant thermocouples located in the outlet of the stator coolant flow.
For these tests, additional thermocouples acre located in the st at or winding to measure the maximum stator winding temperature as a function of coolant flow.
Normal stator coolent flow is 157 scfm N at 95 psig.
For these 2
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ATTACHMENT 5 i
Insert page at the end of Section 9.16, Enclosure i l
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j Section 9.16, Question 7 The discussion of recirculating gas cooling system shutdown on leak detection is inconsistent between PSAR Section 9.16 and 7.6.6.
Response
Automatic shutdown of Recirculating Gas Cooling Subsystem due to moisture and t
leak detection signals have been deleted from PSAR Section 9.16.3 to avoid automatic shutdown due to spurious signals. The attached change to PSAR Section 7.6.6.2.1.1.2 deleted the discussion of shutdown on high water vapor, j
but adds a shutdown due to high temperature in the return piping. Automatic shutdown of the Recirculating Gas Cooling Subsystems due to high water level
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l in the cooler is retained in the discussion of PSAR Section 7.6.6.2.1.1.2 and 9.16.3 (pg 7.6-12, 9.16-7).
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Encksare. L
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a)
A discrimination system shall be provided to insure against a refueling error which could result in a significant reactivity error or undercooling of the control assemblies.
This system shall prevent the following situations:
Insertion of any assembly other than a control assembly into a control assembly position.
Insertion of a control assembly into a fuel assembly position or a wrong control assembly position (i.e. interchange of a primary and secondary control assembly).
b)
The relative location of the absorber pellet colunn within the pin shall be maintained uncer shipping, handling and scram j
arrest loadings by a properly designed axial spring suoport system.
51 4.2.3.1.6 Environmental Requirements The control rod system shall provide safe reliabic. shutdown and control capability when subjected to the following enviror. mental condi tions.w A ec. aMdA
-b w % sAmad W e. -bled \\=ss uS %
a-loss -C C-M n eM/
4 " Won-l c..t'$ w & A l
The external surfaces of the Cor. trol Pod Drive Mechanisms are exposed to the head access area environment at temperatures between 70 and 150 F during full power operation.
The internal atmgsphere of the CRDM is inert gas at temperatures ranging from 70 to 400 F.
Normal primary mechanism internal pressures range from 15 to 30 psia.
The control rod drivelfnes are exposgd to an environment of liquid sodium containing 2 (>B00 F) to 5 (<800 F) PPM oxygen over the lower 60 percent of its length (approx. ).
The remaining upper portion is exposed to an atmosphere of inert cover gas and sodium vapor.
The control assemblies are exposed to an environment of g
U liquid sodium containing 2 (>S00 F) to 5 (<800 F) PPM oxygen.
51 l The design of the CRBRP Primary CRDM/CRD is conceptually the same as that used on the FFTF program.
This design employs some proven 531 design concepts, (References 145,146 and 147) that inhibit the upward movement of sodium vapor laden cover gas by reducing the annulus width while increasing the length thereby minimizing the effect of natural convection.
To further insure the reliability of the Primary CRDM/CRD a continuous purge system consisting of recycled cover gas will be utilized to provide a constant downward flow through the annuli into the reactor.
4.2-247 Amend. 53 Jan. 1980
ATTACHMENT 6
!r. sert. page immediately prior to page 9.1-8, Section G.I, Replacement pages for pages 9.1-31, 39a, 57, 65a, 65da, 68a, 58b of Section 9.1, Enclosure 2 i
(d) Motor thermal overload.
The bypass identified in item (b) above is initiated by a Key Operated Selector Switch located on the Local Control Panel and administratively controlled.
(5) The f an start /stop indication is provided on the local control panel as well as on the back panel.
" Fan stopped" is alarmed on the local control panel, and alarmed as " Fan trouble" in the computer.
Bypass switch status I's indicated on the local control panel.
" Fan stopped" is alarmed as " Inoperable Status (IS)" in the computer located in the control room.
7.6.6.2.1.1.2 Automatic isolation Valve Oceration (Figures 7.6-39, 40 & 42)
- as -is The AutcnnatIc isolatlen valves are f al1 epea yalves.
(1) The Automatic isolation valves can be operated from en "open-auto-close" swlich (spring return to auto) located on the back panel and the local control panel.
(2)
When the switch is in the "open" poslilon, valves will open vi,eo all of the following conditions are satisfied:
4c)
N's h gh.eter s;per ir, th; sepp l7 go; st r: ;
( ) No high water level in the cooler.
s (3) When the switch is in " auto" position, valves will open when f an start demand switch signal is received and all of the followi~ng conditions are satisfied:
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L's h!gh.ater vrpnr in thn ennniv n--
- t c ;;7,,
cv (p9 No high water level in the cool er.
(4) The valves can be closed manually by placing the switch in the close position.
l (5)
When the switch is in " auto" position, valves will automatically close under any of the following conditions:
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- igi..;tcr ;;;;r '- +'e supp!y ;;; et r, prev!d
- d r.c...: us!
typ;;; f s high water ;;pc- !: '-'+!sted.
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(Y) High water level in the cooler.
e(g) The f an has stoppe<j and there it no f an stert demand.
(. c.) thfk femferdure. in rehrt, p'p n%
(6) The Automatic isolation valve open/close Tndication is provided on the back panel and local control panel.
Closure of the valve is
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alarmed as " inoperable Status (IS)" in the computer located in the control room.
7.6-12 Amend. 71
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'.> tween the double seals. The purpose of this buffer pressure is for leak detection and is not required to prevent seat leakage, although it would mitigate an inner seal leak. The inflatable seals are the only ones whic.h depend on a continuous source of electrical power an,d inflation gas for operation.
in case of loss of of f site power, the seal inflation sfstem valves would f all open, providing the seals with a continuous source of inflation gas f rom the normal supply system.
(The valves are closed during normal operation to provide more sensitive seal 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 all open, loss of of f site power would not ef fect seal inf l ation. The piping and va!ves from the gas bottles to the inflatable seats 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 seal s.
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l The EVTM is herr.,etically sealed to a ref ueling station by Itvering the closure valve vhich maies with a floor velve.
The actaa! seeling at this interface is accomplished by elvTomer double seat s, which a/e periodically leak checked.
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and Pressure Vessel Code, Seismic Category I, and is located within a hardened structure.
"The. crane handlec gas cooling grapple, shown schematically in Figure 9.1-9, is mainly used to transf er bare f uel assemblies f rom the spent f uel transfer station to the spent f uel shipping cask.
Design of the grapple finger actuation mechanism prevents actuation of the fingers to release a core assembly while the fingers are supporting the weight of the assembly. The 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 will stop at its position at the time of the power failure.
Design of the crane includes the capability for manual operation. Access to the trene for manual operation is through ports ir the wall and roof closure.
Two redundant argon gas-cooling blowers are mounted on +h, tpoer e.id of the I
ges-cool ing greppl e.
These blowers draw arcon gas from the kurrounding cell enviromnent and blow it through the grapple enc f uel assembly, dischargir g it teck into the cell through the nozzles at ibe bottczn of the fuel assembly.
The argon gas flow rate will be larr,e anough to maintain the cladding temterature of a f uel assembly belcw 1ho no mal cl adding tnmperature limit f or decay I.est loads up to 15 kW.
The blowers are ap;1 N 4%-ecet-etee*r4eal pows, w - cic. ss _i c s pws w@d A Any e le c. W..\\ p c< h&
g,,c d.ca ry nAw M LA V.3 L.LI>-
9.1.3.2.3 Sefety Evaluation I f CCP containing a f uel assembly is cooled suf ficiently by natural convection of the adjacent FHC atmosphere to maintain the peak f uel ciadding temperature below the limits given in Table 9.1-2.
The peak temperatures, given in Table 9.1-2A f or normal operations in which the FHC atmosphere temperature is maintained by the argon cl.culation system 'and for the unlikely event of loss l
of cooling of the FHC atmosphere, are within the limits.
The argon cooling gas flow rate through the spent fuel assembliss while being handled by the gas cooling grapple is suf ficient to maintain the maximum 0
steady-state cladding temperature of a 15 kW f uel assembly below 600 F.
In the event of loss of argon cooling gas, suf f!cient time exists for the assembly to be transf erred back To a Na-filled CCP in the spent f uel transf er station within the FHC before the f uel cladding reaches 15000F.
Adequate cooling of a spent f uel assembly suspended f rom 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 biceer will be tested before beginning FHC spent f uel shipping operations to ensure its operability.
Evaluat ',on of the loss of power f or cout ing syste-s f or f uel assemblies in the FHC shows that the_ consequences are acceptable, in the event of loss of normal of f site power, operctton of the cooling blowers wout d stop and the temperature of a suspended f uel assembly would rise.
The loss of power would 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 h
)
PS & See G 9.1 1.
FHy Normal Fuel Handling In.o typical spent f uel handling sequence, a spent f uel assembly in a core compenent pot is lowered through the f uel transfer port (scc Figure 9.1-7) by the EVTM, into the spent f uel transfer station directly below the port. A l
lazy susan assem' bly, with three transfer positions supported by a stainless steel gridwork, prov! des the storage locations.
Each position hcids one fuel assembly, in a sodlun-filled core componont pot.
Decay heat is removed by natural convection to the FHC atmosphere.
The spent f uel assembly 'Is removed f rom the core component pot by tho in-celi crene, using a ges-cooling grapple, and allowed to drip dry.
If f or some reason not identif ied as a part of normal procedures, it is decmed necessary to rcoove a sodlum film frcan the extorio surfe:es, the exterior surf aces will be wiped with alcohol wetted swabs.
Then the spent f uel assembly Is lowered into the spent f uel shipping cask located In a shaf t below the cell floor.
The sequenes Is repeated f er the i
- m. $ber of assembiles necessary to fill the shippt r.g cask.
The ebove fenet!ons althin the RiC are performed remotely by operators in tho adjacent oportting gallery, and can be observed through the viewing vindows, flormal core essembly handling operations in the FHC ar.e conducted with assemblies having a decay heat of 6 kW or loss, infrequently, it may he neesssary to examine a high-powered Core assemcly. Th!s will be dora only when (1) It is necessary to complete ref ueling er to cossence reactor starttp, or (2) use of the RfC is necessary to recover from en EYTN grapple mal function occurring while grappled to a high-powered core assembly.
During these operations, special pre aut!ons shall be observed, including renoval of all other spent core assemblies from the R4C prior to introduction -
of the high-powe' ed core assembly.
In addition, only one core assembly-r greater than 6 kW shall be permitted in the FHC at any time.
2.
Sagnt Fuel Examination Spent f uel exanination in the FHC is limited to inspecting tho exterior surf aces of fuel assemblies to determine their geemetrical condition bef ore loadirg Into the spent f uel shipping cesk.
Spent f uel assemblies will not be disassembled or sectioned in the M4C.
It is planned that only a few selected spent f uel assemblics will be exemlned, 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 exient of the spent f uel examina Ion covers the following operations.,all of which will be perf ormed in the f uel examination fixturo:
- 1) Visual inspection of all exterior surf aces
- 2) Determination of axial and radial dilation of fuel assombly by measuring its length and distances across flats
- 3) Measurement of the f uel assembly bow
- 9. / - 4 b
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.gl,o y provide a second, redundant load path from the handling ball to the polar crane hook. When not in use, the AHM is stored at a parking station located in the northeast quadrant of the bullding.
The parking station is designed for the SSE seismic loads which are carried into the RW 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 ailure.
Electric power to the AHM can be manually disconnected at either the AHM console or the st:bstation supplying the floor service station f rom which the AHM is being supplied.
i i The vertical pcsition of the AHM grapple is displayed on the AHM control console.
When the AHM is in position at the reactor, only the extender matir.g fienge is resting on the flocr valve, which in turn is supported from the srtall rotatir.g plug (SRP) by an adaptor.
If the two components were firmly attached +o eac5 other, the resol ting combined structure, in effect, would represent a tali, vertical cantilever rising f rce the SRP, attached at its upper end to the I
pol ar crane. The large bending moments and shear loads in this cont.ted struciure, resulting from horizontal excitation due te an ODE or SSE, are ret leved by structurally decoupling the AHM from the +1oor valye at the extender / floor valve joint interf ace.
At a predetermined horizontal ground acceleration, complete severance of the AHM from the floor valve (" breakaway" concept) eliminates the cantilever beam effect and significantly reduces all seismic loads.
The joint between lower extender flange and floor valve is designed with shear pins which f all upon reaching a predetermined horizontal load. This enables 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 fofIowIng the shearing-off of the shear pins. The actuators can ralse the extender by about 3 inches in less time than it takes f or the extender to l
clear the floor valve during the horizontal movernent due to an OBE or SSE.
1 9.1.4.5.3 Safetv Evaluation l
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 EVTM (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 RG operating floor.
Radioactivity released does not exceed the limits as set forth in Sections 12.1.1 and 12.1.2.
9.I - 57 A
J a#'
,9
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seals are provided continuously with a butfor pressure between the double seals. This pressure is monitored continuously for leak detection and is not required _fq prevent seal leakage, although it would mitigate an inner seat
'l'eaCThe Inflatable seals are the only ones which depend on a continuous source of electrical power and inflation gas for operation.
In case of f oss of of f stte power or gas supply, th; :::! infhrtion systemmFves-vouf4-f ail open,-prwidfog-themats-with-e-coatlewees ega c. ci inHetion gas pressur+q (The valves are closed during normal operation to provide more sensitive seal leak detection.)
Since the supply valves f all open, loss of of f sita power
)
would not af f ect seal Inflation.
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TMLE 9.1-2A SPENT FUEL ASSE16LY CLADDING TEMPERATURES (Sheet 1 of 2)
( g)
Peak Fuel Location of Frequency Assembly Cladding Fuel Assembly Cl ass Temperature (O )
F CCP in EVST stcrage location 70 kW Assembly b 1) UlD* (**D
$,ormal oporathan llormal Neturel convection loop Ualikely b1) W cooltrq CCP H. FiC (1.'kWt asse'nbly)
Arp circulertion system Unlikely 1060(I) operative Argon circulation system Extremely 1265 Inoperative Unlikely CCP in FHC'(6kW assembly)
~1 10 ACS operative Normal 70 7 Q : 6.7:C f_:!)
ACS Inoperative Unlikely 9000 CCP in EYTM col d wal i 20 kW Assembly Cold walI blover on Normal 1240(1)
Cold wall blower of f Anticipated (4) 935 UniIkely 1350(II CCP Immobilized in EVTM stack
' assembly (essembiles below
- the cask body essembly, see Figure 9.1-14) 20 kW Assembly O)
Anticipated 1230 Unlikely 1415(1)
A
g...,,....,......,
TELE 9.1-2A SPENT FUEL ASSEELY CLADDIN3 TEMPERATURES (Sheet 2 of 2)
Peak Fuel Location of Frequency Assembty Cladding Fuel Assembly Class Temperature (OF)
CCP Imobilized in reactor fuel transf er port 20 kW Assembly Blover on Anticipeted 950 l
Unlikely 1444(1)
Bicee off AntIcIned ed 970 Blower of f Extredsty UniII'oly 1482 CCP IrnmobilIzed in EYST f uel transfer port t
20 kW Assembly Blower on Anticipated (.2)
(1250 9)
UniIkely 1389(1 Antict osted (2)(4) l l
BIover off
<1250 BIower of f Extremerp unt ikely 1482 CCP Imobilized in FHC spent fueI trans.fer port 15 kW Assembly Blower on Anticipated (2)
<1225 Unlikely 1249( )
l Blower off Anticipated 1225 I
Blower of f Extremely Unlikely IDS (ad (1) Steady-state temperature (2) For a stopped CCP with the f uel transf er port blower inoperative, the operator is required to take action as described in PSAR Sectiong.1.4.3.2 within 30 minutes to raise or lower the CCP.
9.l-6@)
Footnotes to Table 9.1-2A (3) Loss of off-site electrical power during a fuel handling operation is classified as an unlikely event. This classification is based uoon the number of hours per year that this fuel handling equipment is operating and the conditional probability of loss of off-site power at such a time.
(For the probability of loss of off-site power, see question response 222.24.)
(4) Upon failure of power to the EVTM, the operator is required to connect a temporary power cable to a Floor Service Station.
i 4
- 9. l-bklo) cN1
ATTACHMENT 7 i
Replacement page for page 6 (top), secticn 9.4, Enclosure 2
~
h, M ~
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 Mgo insulated single heating element.
Separate Chromel-alumel thermocouples are used throughout the systems for the feedback signal to i
control the operation of the electric heaters and for l
monitoring the temperature of the metal boundary of l
the sodium containing pipina and eauipment. b b c +t. p h.
,,yt p,
.k4s ts yee.,.4 e3 &< o. k wh t s.4 m.o.:6 +n Ac,
as,
TIlermocoup e compensation is provided for all thermocouples.
Thermocouples on piping are located at a point on tiie pipe to enable control of the average temperature of the pipe within specified limits.
On equipment, the thermocouples are located in the spaces between heaters for both monitering and control purposes. Id P " C"
oQ \\Swyme coagle s cebined W4h kaad -4raosb ebME8E6 j
yve.cL.hw wodeAe. 4exL co \\1 s poks.
Control of any heater or bank of heaters is by automatic control.
This control provides for continuous and automatic adjustment of. heat based on an error signal generated from the difference between the tempe rature setpoint, as set by the plant
4 ATTACHMENT 8 Replacement pages for pages 10,11 (top) and 9.4-5 of Section 9.4,, with new pages 9.4-12,13, and insert c
.-...:...:. a.-..
({ L 10 -
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.p.3, the Piping and Equipment Electrical Heating 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.
However, considerations fqr trace heating of selected safety related components have been taken.
t i
G - II -
A Technical specification will be provided to address f ailures of trace heating in the primary equalization line, the IHTS cover gas equalization line, the argon lines leading to the primary argon relief cystem, and the overflow heat exchanger.
Relief protection f or the IHTS is provided by the rupture discs to either the dump tank or the Sodium Water Reactions Products tank.
The large eighteen-inch lines downstream of the rupture discs are normally empty and are not trace heated.
The eighteen-inch line up to the rupture discs (approximately five foot long) has trace heating installed, but during normal operation the heat transfer from the flowing IHTS sodium is sufficient to maintain the temperature in the line above the heater setpoints.
The six-inch gas line between the IHTS expansion tank, with its associated rupture discs, is trace heated to reduce sodium frosting.
Should this six-inch line become plugged due to trace heater failure, the effect of a sodium water reaction would be neutralized by blow down of the water side of the SG modules and, if needed by rupture of the large rupture discs in the eighteen-inch line.
Thus, failure of l
the trace heating of the six-inch line will not cause a safety problem.
l t
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4.4.It The EVST is cooled by three cooling circuits, two re4 Ment "r.cmal" forced Each of the circulation circuits, ar.d a b:ckup natural circulation circuit.
circuits contains a sodium loop, circulating EVST sodium, and a NaK loop, which transfers the heat to the atmosphere. All of the sodium and NaK loops are electrically preheated. The preheat is segmented into multiple control renet.
each zone having at least one control thartr,0 couple. Preheat tenperatures higher or lower than the setpoint band are alarmed. The preheaters are not redundant and the preheat system is not on amergency power.
During EVST cooling operation, the forced circulation circuit in operation dces not require electrical preheat; the heat is provided by the fuel within the The natural circulation loop also does not require electrical preheat, as EYST.
long as a heat source is present in the DST, since a sr.,all flow of sodius and The standby forced curculation loop does NaK is always maintained in this loop.
Should a heater circuit, or circuits, require preheat to maintain temperature.
fail in the standby cooling circuit (or in any circuit) the cooling circuit can be isolated and the bester repaired while still maintaining two means of EYST Consequently, failure of individual heater circu' ts is not cor,Sidarred cooling.
In the event of loss of plant power or a total a significant safety probl=.
loss of the preheat systen, the normally operating forced cooling circuit and the natural circulation circuit would be unaffected from a stand;cint of DST In the standby circuit. NaK temperature can be maintained by pump heat coolleg.
(pump on emergency power); only the stagnant soflum loop is susceptible to decreasing terrerature (and ultimately freezing) and this siv tion could be monitored by tbe safety-)related themocouples on the EYST sodM o (pst-accid?nt monitors.
ically switching EYST cooling from one circuit to another in order to use EYST In any cycat,isss of heat to maintain sodium temperature well above freezing.
pc,.er or prehet can cause lois of no more than one of three available cooling circuits, if no operator action is taken, and will result in no loss of cooling circuits if EVST cooling is alternated betwcen circuits. Consequently, neither redundant heater circuits nor safety-related heating is considered required to I
ensure continued EYST cooling.
l The DHRS circuit consists of the norra11y operating reactor cycrft:r,ej, a6p circuit, the comily stvet On and NaK " crossover' piping, ar.d the EVST
.taK fur ced circulation loops previously discusscd. Redur. dant heaters are l
currently used on the overflow / makeup circuit and OHX and a change is being failure of heater circuits made to add redundancy to the HaK " crossover" pipir.b on any of these components (including the standby EVST hP. circuit) will be Where redundant heaters are installed. the redundant signaled by an alars. heater can be connected and energized from accessible a cells) within an hour, well within the tir.e when cooldown of an isolated circuit As previously noted, the preheat on the EYST NaK could have significant impact.
loop can either be repaired or temperatures vintained by pap heat, at the f
9.y- /s operator's option. Consequently loss of individual prehent circuits should impact neither reactor op2 ration nor capability to perform the DHAS function.
If the entire preheat systm (or substantial portions of it) is Icst due to some electrical'or occhanical failure, then the reactor should be shut down. With i
the reactor shut down and prirary sodium temperature reduced, the.DhRS function is not compromised even without preheat, as discussed balcW.
An evaluation of DHR$ capability following loss of plant power was mede to ensure that DHR$ is not cocpr=fssd during the plant event. Uponlossofgewer, the reactor is tripped and sodium temperature automttcally reduced to 600 F.
Both EVST NaK flow and overflow / makeup sodium flow are continued after a short interruotion since these pumps are on er:ergency poder. The stagnant OHX and NaK
' crossover
- piping slowly begin to cool since the al'ectrical preh2at is out of service without normal plant power. This cooling will increase the startup thermal transient strat??s. wht:h thG pieficating is designed to minimize, should unu.5 be invoked 1/2 hour or more after scram.
The increase in these stresses in the controlling components involved in DHRS (e.g.. overflow heat exchenger) was determined to have minor irrpact on their structural integrity. The worst case thermal condition in the DHRS startup time from 1/2 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after scram was analyzed tu find that these cocponents are designed to withstar.d many cycles
(>100) of this event. The conclusion is that the additional themel stresses due to ccoldown in this event are not controlling and DHRS can be successfully initiated and carried out.
If the loss of power is long enough, ultimately the OHX will freeze; however. calculations indicate that this will take approximately 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, an extremely unlikely situation.
If the loss of pesar does Appear to be extensive, for example, if not restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. then gt that time the reactor sodium could be reduced to refueling torperature (400 F) and the trokeup sodium stream diverted such that it flows through the 0ltX. Under these circumstances, freezing would be precluded and DHR$ could be initiated at any I
tire, indefinitely, af ter power loss.
dering the above capability, a safety-related, IE powered p.ehant' system is i
not considerad necessary.
l 1
l l
l l
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Inwd O
A number of sodium valves are active components, used either for EYST cooling o DHRS initiation. The valves are prehcated by separate preheat circuits. The preheat syste:h is not safety-related ard is not on IE per. Safety-reistM preheat is net considcred necessary for these valves, for the same reasons f
a discusses w since these valves are part of the circuits involved in that discussion.
The redundancy noted also applies to the valve heater circuits.
feh:126
normally)' The unwanted addttional heatine of sodfim ifnes (sensed and controll en trece heaters (which should be offdue to multiple failures in the trace he subsystem heat removal capability per) loop.is less than five percent of the long tem s persent of the short term subsystAm heat removal capability per loop, hse pers:ent40es are suf ficiently small in terus of hea capability that the occurrence of this rallure mechanise would not cogroelse
~
the sat'n shutouwn functfun.
The third potenttal failure mechanism is a short to a non-wiring camponent accurring with concurrent failures of the ground heater sheath, the ground fault detectors, and the over Current protectiva devices.
would nut cen9tweise
?** safe shutdowr..w6 tha"plaret.- h slurtlest pipe where a This mechanism short could oct.ur is grwar '..an ten ilmes the cross-s,vetional disaseter of the electrical wiring.
1hereture. for the smaller,t pipe, the conductivity of the
(
electrical wiring is one-helf the conductivity of the pipe, and the conductivi I
of the pfpe is over fort or an arcing situation y times higher than the heater wire.
the pfpv would not fail.
In either a short OpereLlonally, the failure mechantnas requires the failure of the tan-perature sensing syst.m and one of the following:
(1) excess current application
(?) cross-over in sounting of adjacent heat,ars, and (3) improper setting of pr tective devices.
For desfgn related failures, the failure mechanism can be crused by impropor heating wire design, fissures in the magnesium oxide, and bend less than the minimum bend radii.
failure of the sodtun containment. The effect of the failure will not cause 9.4.4 Desten Reliebt11tv 1. 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:
(1) For norinal operatton, the heaters are operated at less than 1/2 rated power. For abnormal operation, each heater control circuit. is protected against overcurrent by thermal overload circuit breater and temperature sensors on the heated enspenent.
Ground fault interrupters (GFI) will be used for protection against ground currents.
(2)
Migh and low temperature alarms ans provided for all control and monitor thennocouples in all heater control zones.
(3)
Tne cold ends of the heaters are bent 90' and brought out from the component.
A spacing is maintained between adjacent heaters to prevent crossover of heaters and significant mutual heating by radiation.
9.4-#
~
e ATTACHMENT 9 Insert pages to follow page 9.15-2 Section 9.15 Enclosure 2
In order to maintain the integrity of the cell liner, the piping from the cell liner up to and including isolation valves for non-safety
(
related subsystems which serve cells containing Na or Nak is designed to the requirements of Seismic Category I.
All the safety related subsystems are designed in accordance with the requirements of ASME Section III Class 3, Seismic Category I, supplied Class IE electrical power and emergency chilled water.
Safety related subsystems f!A and MB serve the two primary Na makeup pumps.
Since the primary fla makeup pumps are redundant to one another, no further redundancy in components is provided for subsystems MA and MB.
Safety related subsystems EA and EB serve the two EVS lia pumps in the two active EVS lla cooling loops.
No redundancy in components is provided for subsystems EA and EB since there is redundancy in the EVS fia cooling loops themselves.
All the subsystems using water as the coolant and serving areas containing fia or flak are provided redundant water leakage detection sensors.
On detection of water leakage in a subsystem, the operation of the subsystem is stopped automatically, the automatic isolation valves in the gas lines and chilled water system lines are closed and the redundant drain valves on the cooler are opened automatically.
In 4e case of a. seelm ce %k spill er leak in c cell> M
-b ::::
- 2 redfe" er Mab spill er 1:J ' a cell, th; escatfor
, (
- of a sub5ym.m scr A g th; cell h :tepp;d and th: =te~ tic ire'atte
--m-
, _ 3,. m,. u,. a a iptw & [> rod ed' " lee 6A A K / c/c/ceS M ti cd5c'd5
'^ E'dd^ 7 5 5-gg 9.10 4 Tests and Inspection Each individual component of the system is tested at the factory and, before the plant startup, entire system is tested and the gas flow l
- 7. If.3 rate is balanced and set at design flow conditions.
Periodic inspection l
of the components is scheduled to ensure proper system operation.
In-l service inspection will be conducted according to ASME Section XI for l
safety related subsystems as described in detail in Section 3.1.
9.16.5 Instrumentation and Control The RGCS instrumentation is designed to provide for measurements, controls and alarms of system parameters.
Each subsystem is provided with a control panel located near the fan.
The panels include the control switches, monitors and system alarms.
All non-safety related subsystems are provided with local monitoring and alarm and a remote alarm in the main control room.
All safety related subsystems are provided with local monitoring 47 and alarm and remote monitoring and alarms in the main control room.
A list of compressed air operated safety related valvefwhich
(
- 59. Table 9.16-3.
are provided with safety class 3 air accumulators is contained in Amend. 59 l
9.16-7 l
l
rN gerk hr
/H
_a 14< ell Cooling tos_s fecacts on DHR$
The loss of cell cooling should not prevent the capability of acceptable DHRS operation.
in containmnt, the cells containing the DHRS components are separated into two groups of cells. each gmup being cooled by a safety-related cooler.
Ea h of the coolers cools one of the two primary sodium makeup pu@s. The coolers are on emergency power and each cooler has redundant fans sized such that a fan failure does not result in unacceptable' cell tegeratures.
Consequently, neither loss-of-plant pnwr nor fan failure has any i@act on DHRS. Shoyld an entire cooler unit be lost for other reasons, the nakeup pum (direct gas cooled from the cooler) associated with that cooler would be lost, hever. perforcance tests of the makeup pu@s have shown that a single pug alone can provide the l
required DiiRS flow.
In addition, it is a plant raquire.Nnt that all components r.aintain pressure' boundary under loss of cell cooling for an indefinite period of time.
In the reactor service building, the only DHR$ components in inerted cells cooled by cell coolers are piping and the EVS sodium coolers (these coolers are not "used" for DHRS; however, the DHR5 NaK flows through theo; in this sense only, they are called "DHRS components"). Neither the piping nor the coolers are impacted by loss of cell cooling.
The remaining DHR$ cove, ants (Abiys, NK pumps. piping) are located in two air-atmosphere cells (one for each of the two NaK Icops cogrising the DHRS haat sink) cooled by the plant H!,y systc= cociers.
The ecolcrs are safety-related on ecergency power.
Even with loss of cooling, the large cell coupled with relatively lod heat input rake it unlikely that even lengthy loss of cell cooling would interfere with DHRS operation.
Even if cooling for one of the NaK cells was lost and te@cratures did reach unacceptable limits, a thernal analysis conducted as part cf a "OttRS with single failure" study has sho.m that a single Nak loop can by Itself pMvide sufficient,DhRS hnt rejeci, ion to ensure adequate core ceoling.
feh:126