ML20205K772
| ML20205K772 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 10/25/1988 |
| From: | Cockfield D PORTLAND GENERAL ELECTRIC CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NUDOCS 8811010342 | |
| Download: ML20205K772 (22) | |
Text
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October 25, 1988 E
8'EC.
David W. Cockfield Vice President, Nuclear f
License NPF-1 U.S. Nuclear Regulatory Commission ATTN: Document Control Dock Washington DC 20555
Dear Sir:
i TROJAN NUCLEAR PLANT Safety Evaluation for Puol Assembly Repair As previously discussed with representativos from the Office of Nuclear Reactor Regulation Portland Conoral Elect ric (PCE) intends to repair 18 nucioar fuoi assemblies at the Trojan Nuclear Plans. The assemblies to be repaired were damaged in the reactor by a phenomenon referred to as "baf flo gap jetting" during the 1981-82 fuel cycle. The maximum number of fuel rods to be replaced por assembly is 11, including broken rods, rods with cladding damage, and fuel rods with suspoeted damage.
One assembly will be disassembled and reassembled using the samo fuel rods in a new fuel l
assembly skeleton due to spacer grid damage.
The repair activity, scheduled to begin November 7, 1988, is being performed within the provisions of Title 10, Codo of Federal Regulations, Part 50.59 (10 CFR 50.59).
As noted in the enclosed safety ovaluation, the conceivable occurrences and accidents relating to the reassembly and reconstitution of fuel assemblics have been evaluated, and PCE has con-cluded that this activity will not havo an adverso offect on Plant or per-l I
sonnot safety.
It will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Trojan Final Safety Evaluation Report (FSAR),
L nor have any mechanisms for an accident or malfunction, which has not been previously evaluated, been identiflod.
Also, the fuel repair activity does not decreano the margin of safoty as defined in tho bases for any Trojan Technical Spoeification. Thorofore, the fuel repair activity doos not constitute an unroviewed safety question as definod in 10 CFR 50.59.
Tho enclosed sofoty evaluation is submitted for information and review if desired.
Noto that the rad.8ological anal} sis in the sect.lon entitled "Radiological Consequences of a ruol Handling Accident" in tho enclosed safety evaluation states that "All ol' the fuel rods in one assembly at o essumed to be damaged as a result of the (postulated) handling accident". An additional analy-sis, performed by PCE has concluded that the consequencas of a fuel hand-ling accident during fuel repair operations are bounded by the FSAR r
8311010342 83102'3 PDR ADOCK 05000344 pf P
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Poriland General BoctricCompany U.S. Nuclear Regulatory Commission October 25, 1988 Page 2 analysis even if up to 52 replacement fuel rods are damaged in additicn to all the rods in one assembly. The replacement fuel rods will be located in a fuel rod storage container in the fuel elevator basket adjacent to the fuel assembly being repaired.
Refer to the Description of Operation section of the enclosed safety evaluation for more detail.
Please do not hesitate to contact Mr. T. D. Walt at (503) 226-8120 if further information is required.
Sincerely, bf Enclosures c:
Mr. John B. Martin Regional Administrator, Region V U.S. Nuclear Regulatory Commission Mr. William T. Dixon State of Oregon Department of Energy Mr. R.
J.
Bare NRC Resident Inspector Trojan Nuclear Plant
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SECL.88 476. Rev. O Customer erenceNo(s).
Westinghouse Reference No(Applicable) s).
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NUCLEAR SAFETY EVALUATION CHECK LIST
- 1) NUCLEARPLANT(5)
TROJAN NUCLEAR PLANT
- 2) CHECK LIST APPLICABLE 70: FUEL ASSEMBLY REPAIR
- 3) The safety evaluation of the revised procedure, design change or modification required by 10CFR50.59 has been prepared to the extent l
- equired and is attached.
If a safety evaluation is not required or is incomplete for any reason, explain on Page 2.
l Parts A and B of this Safety Evaluation Check List are to be completed I
only on the basis of the safety evaluation performed.
CHECK LIST PART A l
I;. p Yes_
No A change to the plant as described in the FSAR7 t
(. p Yes No A change to procedures as described in the FSAR?
J Yes,_ No.)L A test or experiment e... described in the FSAR?
(I,. J Yes No_1 A change to the plant..chnical specifications (AppendixAtotieOperatingLicense)?
ii
- 4) CHECK LIST - PART B (Justification for Part B answers must be included onpage2.)
(4.1) Yes _ No_1 Will the probability of an accident previously evaluated in the FSAR be increased?
(4.2) Yes No X Will the consequences of an accident previously evaluated in the FSAR be increased?
(4.3) Yes No 1 May the possibility of an accident which is i
different than any already evaluated in the FSAR be created?
(4.4) Yes No X_ Will the probability of a malfunction of equiment important to safety previously evaluated in tie FSAR be increased?
(4.5) Yes No 1. Will the consequences of a malfunction of equip.
ment important to safety previously evaluated in the FSAA be increased?
(4.6) Yes No 1 May the possibflf ty of a malfunction of equipment important to safety different than any already evaluated in the F$AR be created?
(4.7) Yes No.l Will the margin of safety as defined in the bases to any technical specification be reduced?
PAGE 1 0F 20
W.Pd If the answers to any of the above questions are unknown, indicate under 5) REMARKS and explain below.
If the answer to any of the above questions in 4) cannot be answered in the negative, based on written safety evaluation, the change cannot be approved without an application for license amendment submitted to the NRC pursuant to 10CFR50.59,
- 5) REMARKS:
The following sumsrizes the justification upon the written safety evaluation, (*) for answers given in Part 8 of the Safety Evaluation Check list:
$EE ATTACHED SAFETY EVAWA*.10N
(*)Referencetodocument(s)containingwrittensafetyevaluation:
FOR FSAR UPDATE Section:
Pages:
Tables:
Figures:
Reason for / Description of Change:
Date:/#//9b Preparedby(Nuclear $4fety):-
is CoordinatedwithEngineer(s): oN F/4 8 /###9 ^!w 0 ate /8 A V//r CoordinatedGroupManager(s): a
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Date: <$/U/r r
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CoordinatedwithEngineer(s):
lhater/4b8/#1 I/6 astc.-
Coordinated Group Manager (s )
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/ C.GER$T8EMEA Nuclear Safety Group Mana Date:#2s//8 dMW
/fff PAGE 2 0F 20
NS-RCSCL-C\\l 88-748, Rev 0 NS SECL 88 476, Rev. O PAGE 3 0F to TROJAN NOCLEAR PLAm FUEL ASSEMBLY REPAIR SAFETY EVALUATION INTRODUCT10N.
This safety evaluation addresses the nassembly/ reconstitution of Westinghouse 17 X 17 standard fuel assemblies which is to begin during late 1988 at the Portland General Electric Trojan Nuclear Plant. The evaluation assesses the potential Jafety impact of the fuel reassembly / reconstitution procedure to be implemented at the Trojan site.
The evaluation is completed in accordance with Title 10 of the Code of Federal Regulations Part 50.59 (a)(t) (10 CFR 50.59 (a)(2)) cri',oria.
This safety evaluation includes a description of the equipment to be used in the repair procedure, a description of the procedure to be used, and an f
i evaluation of postulated events / conditions related to 4
j reassembly / reconstitution of nuclear fuel assemblies.
Applicable design bases for spent fuel storage given in the Trojan Final Safety Analysis Report (F5AR), Section 9.1.2.1 are maintained as shown t
repair process, uithout soluble boron prlr, t 0.95 during the fuel herein. The capability of maintaining k i
66nt in the pool, has been 4
verified. However, to protect against postulated accidents, abnormal l
conditions are assumed as in fuel handling operations, and a so M le boren t
oncentration of 2,000 ppm will be saints' ned. Dose rates at the surface tf the spent fuel poci (SFP) and at the outsido surface of the walls of i
l the SFP will not exceed the design bases limits and a drop accident resulting from fuel repair activities will not result in offsite radiation j
i dtses to the public exceeding the values given in the Trojan FSAA.
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l EVALUATIONt l
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l Fuel Repair Equipment Description The Multifunction Fuel Repair System (MFRS) provides for nuclear fuel assembly reconstitution by the re'noval and replacement of failed fuel rods j
or the reassembly of fuel assemblies having damaged skeletons. The system 4
j is transportable by truck and can be set up in two days, t
l j
The MFRS consists of an elevator, a fuel assembly basket with a rotator l
i for inverting the fuel assembly, a fuel rod handling tool and the associated tools required t3 remove the bottom nozz e.
The "footprint" of i
the assembled equipment is small, having approximately a 3 foot by 3 foot cross sectit.n, allowing convenient setup in the transfer canal, cask load i
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pit area or elsewhere tii the spent fuel pool. The removal of the bottom j
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MS RCSCL C\\L-88 748, Rev. 0 NS SECL-88 476, Rev. O PAGE 4 0F 20 nozzle is necessary to provide access to the fuel rods. To facilitate nozzle removal, the elevator can ratie the fuel assembly from a fuel rack loading depth to the highest elevation permitted by plant procedures (81 ft. 2 in..
The rotatable basket, which will contain an assembly to be repaired )and a new skebMn (if required) is mounted on the elevator.
Operations may be conducted from th9 bridge or from the pool deck, as preferred, to suit the fuel building geometry and refueling equipment opert.sions.
The MFRS is classified as non nuclear safety equipment.
It consists of virtually the same equipment as is used for spent fuel consolidation.
American National Standard, No. AN$1/ANS 57.10 87, "Design Criteria for Consolidation of LWR Spent Fuel" specifically states that such equipment shall be designated non-nuclear safety (NNS). The design of the equipment meets objective B.b of !!fety Guide 13 (Regulatory Guide 1.13), "Fuel Storage Facility Design Basis", dated 3/10/71, which is to, "Protect the fuel from mechanical damage'. The criteria in the design and manufacture l
of the equipment are in accordance with Westinghouse Quality Assurance Program Plan, WCAP 9245, Rev 9.
Specific equipment design requirements include:
m All submerged materials are corrosion resistant and compatible with the spent fuel pool water chemistry. All materials in direct contact with the fuel components are stainless steel. Tile i
MFRS frame and elevator structure is aluminum.
Primary fuel lifting components enleyed in the fuel assembly l
reassembly / reconstitution process inve been designed per NUREG 0612, "Control of Heavy 1.oads at Nuclear Power Plants,
July,1980, giving che primary lifting components a minimum safety factor of 5 for static loading.
r Upward travel of the fuel above the designated elevation is prevented by fixed mechanical stops on the elevator.
j t,oad testing of the fuel basket and elevator was performed with a test load of 3,000 lbs in the basket.
l Load testing of the MFRS lifting rig was performed at 7500 i
pounds.
i Qualification testing of all equipmnt was performed under simulated site conditions.
Two postulated dynamic load scenarios that occur during normal operations o
have been evaluated.
l 1.
The start upward of the elevator carriage and fuel basket from a full stop position.
2.
The downward stop of the carriage and fuel basket by the rechanical stops.
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NS RCSCL C L.88 748, Rev. 0 NS SECL 88 476, Rev. O PAGE 5 0F to The forces in these two casen have been evaluated and were found to be less than 1.0 g.
They are not included in the determination of the factor of safety of 5 datermined under pure stath: conditions.
Partinent information is as follows:
i Motor Speed:
1750 RPM Gear Reduction: 473:1 Cable Speed: 0.172 Ft/Sec Applicable Weight:
$500 lbs ForCase(1):
Worst Case:
10 ft long cables l
Wire rope spring rate = 30,000 lb/in System Frequency: 48 Radians /Sec 1
Maximum Acceleration: 0.24 4's Resulting Dynamic load:
5179 lbs ForCase(2):
2 Stop Spring Rate: 67,630lbs/ inch System Frequency: 71.8 Radians /Sec 4
Maximun Acceleration: 0.38 G's j
At the Trojan site, the MFRS will be supported on the floor at the bottom i
l j
of the spent fuel shipping cask load pit enclosure and stabilized at the top by attachment to a working platform which will span the enclosure at j
the operatirg deck level. The aligerent of the MFRS will be maintained by a horizontal standoff which will contact the west wall of the enclosure l
near the bottom of the fraae.
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Since acess to the fuel rods for repair will be through the bottom end of the fuel assemblies, the assemblies must be rotated for removal of the bottom nozzle a the extraction / insertion of fuel rods.
In order to l
obtain sufficient space for rotation of the assemblies within the limited confines of the cask land pit enclosure, the MFRS will be positiored next i,
to the door to the enclosure such that the bottom portion of the elevator t
basket may pass through the doorway during rotation.
l j
The MFRS will be assembled using the 17 ft, upper section, the 2 ft and 4 i
ft, middle extensions, the 17 ft leder section and a 4 ft, base for an overall length of 44 ft. The upper stop for the elevator will be set to maintain a minimum de>th of 10.5 feet 16 in, of water above the active fuel region when in tie up position and pool water level is maintained at l
I a nominal elevation of approximately 91 ft. 8 in. At this elevation the t
[
fuel assembly will be approximately 2 ft. above the top of the fuel racks.
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i NSRCSCLC\\t,88-748,Rev.O i
NS SECL-68-476. Rev. O i
PAGE 6 0F to The major compcnents to be used at the Trojan site are:
Elevator Upper Assembly 17 ft. - 2560 lb l
Elevator Middle Extensions 2 ft.- 90 lbs 4 ft.- 200 lb.
i Elevator Lower Assembly - 17 ft.
720 lb i
Elevator 8ase - 4 ft.
480 lb Fuel Basket - 14 ft. - 1550 lb Fuel Rod Handling Tool Bottom Nozzle Removal Tools i
3 Fuel Rod Storage Canister 1
p Debris Canister Debris Vacuum System TV Viewing System Tool Rack i
Miscellaneous Long Handled Tools (The maximum weight of any single tool is 350 lbs.)
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Description o/ 0peration The equipsei', will be set up in the cask load pit enclosure in the fuel i
butiding. ' he cask load pit is at.tached to, but saparated by a door from, the spent f.lel pool. Two types oI repair can be performed using the MFRS:
reassembly and reconstitution. Re unstitution refers to replacement of a few failed rods within an i
j assembly t ith other rods, either fuel rods or solid stainless stul rods.
l In the ca.e of reassembly, all sound rods in a damaged fuel assubly are transferred to a new skeleton. This is a much more extensive operation since it involves movement of all unfailed rods rather than only those I
that are failed, as in the case of reconstitution. With the elevator down at the fuel rack depth, the fuel assembly to be repaired will be loaded into the fuel elevator basket with the spent fuel landling tool.
In the case of reconstitution, a fuel rod storaes container is located in the basket adjacent to the fuel assembly to 6e repaired and the transfers of i
i rods are made between the assembly and this container.
In the case of I
reassembly, a new skeleton is placed into one of the MFR$ work station i
baskets and the damaged assembly or fuel rod storage basket is located in the adjacent basket for executing the rod exchances required. At no time are two adjacent cells vacated in an assembly being repaired. This avoids j
the potential of a fuel rod being cross-loaded between cellt within an i
assembly.
I l
The fuel will then be raised by the elevator to a minimum water depth of 10 ft. The fuel basket will then be rotated to gain accass to the bottom nozzle.
The bottom nozzle tooling indexes on the nozzle using a guida plate that engages the alignment pin holes. The thimble screws are removed by i
unthreading after loosening the locking device. Welded lock bars are cut using a hollow grinding cup. Crimp locking devices are loosened using a l
long handled screw driver. After the thimble screws are removed and i
placed in a debris container, the Mzzle is pulled free of the fuel j
assembly using the guide plate.
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l N$.RCSCL C\\L 88 748, Rev 0 1
NS-52CL.88-476, Rev. O PAGE 7 Of to After nozzle removal, the fuel basket is lowered and the fuel rod handling tool is positioned over a fuel rod for removal through a guide block. The fuel rod is grasped by a mechanical collet actuated by an air cylinder with fati-safe gripping. The rod is removed and depending on its disposition, either placed in a rod storage canister or transferred to a new skeleton. The options for replacement of individual failed rods, which has been determined by nuclear design, is as follows:
1.
Use of replacount pins from another spent assembly.
2.
Leaving the failed fuel rod location vacant.
3.
Use of an inert stainless steel rod, Realacement rods are loaded into the skeletons on 6 one for one basis (tiereby occupying the location from which the failed rod was extracted) and a new botton nozzle is fastened to the thimbles with new thimble screws. A crimp type locking device is used for the replacement screws.
Nozzle ositioning is checked by visual monitoring verification. The assemb1 is then rotated to its upricht position by rotating the basket.
y The fue assembly is lifted out of the fuel basket with the spent fuel handling tool and returned to the storage rack after complete visual scan of all four sides.
Westinghouse personnel operating the equipment are forsk11y trained and qualified. They have also workeent fuel pool,beyond the center to extend the travel of the fuel suilding crane of the cask load pit for MFR$ assembly. To accommodate this requirement, temporary removal of the mechanical and relocation of the electrical rail stops by1 feet, allowing the crane hook centerline to come near the edge of the cask load pit wall will be necessary. This temporary removal / relocation of the rail stops will be maintained only during fuel repair equipment setup and teardown and strict administrattve procedures will be enforced to assure safe crane operation. The rati stops will be reinstalled issediately following the fuel repair activity.
Decontaminatier, of Fuel Repair Equipment Prior to Shipment Upon completion of the fuel reassembly / reconstitution effort, aM j
equipment will require decontamination prior to leaving the l
Trojan site so that the equipment may be shipped Low Specific by Portland General Electric. This equipment has Activny (LSA)ix different sites prior to use at Trojan ano some been used at s decontamination has been required after each fuel repair activity. No problems have been encountered in an I
decontamination procedures in reaching LSA levels.y of these The deconning has generally recuired only the use of respirators by the l
personnel engaget. The number of fuel assembites being processed at Trojan is not significantly greater than that processed at other plants. However, the extent of fuel damage at Trojan will require additional precautions.
Precautionary measures taken with respect to discrete radioactive particles (DRP) are discussed later in this evaluation.
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l Fuel Repair Process Moving Fuel l
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Dropped Equipment During Operation I
The fuel repair process will be performed in the cask load pit area and the consequences of any dropped tools in this area, i
should this occur, are such that no contact will be made with the spent fuel racks. During the fuel repair process there are two i
operations that will involve movement of equipment over the spent fuel pool.
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NS RCSCL C\\L 88 748 Rev. O NS 5ECL-88 476, Rev. O l
PAGE 10 0F 20 1
1.
Transfer of Failed Fuel Rods to the Failed Rod $torage tasket The failed rod storage basket will rest in a cell of a spent fuel storage rack. As a failed rod is withdrawn from the fuel assembly being repaired, it will be transferred from i
the fuel assembly to tse failed rod storage basket. This i
will reouire underwater transfer of the failed rod using the fuel rod handling tool from the fuel repair station in the l
t J
cask load pit area to the failed rod baskst in the fuel l
pool. This transfer will be accomplished usin; the spent fuel pool bridge and its hoint. Tie fuel rod sandling tool is approximately 30 ft. long and weiflhs 350 lbs. As the
's moved from the repair tool, in which the rod is enclosed, levation of the lower station to the storage basket, the e end of the tool will be maintained at an elevation of aI> proximately one foot above the top of the fuel racks.
Ttis elevation is assured by the configuration of the j
hoisting equipment and by administrative controls. A drop of the fuel rod handling tool is therefore bounded by the L,,.,
dropped fuel assembly accident reported in the FSAR.
i 2.
Transfer of the failed Rod Storage Basket Within the Spent l
Fuel Pool l
1
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The failed rod storage basket holds up to 52 fuel rods and l
1s transferred within the spent fuel pool using the Trojan spent fuel handling tool. The combined weight of the basket filled with fuel rods is less than a single fuel assembly.
Therefore, the drop of the filled basket and handlinti tool is no more severe than the dropped fuel assembly ace'dont reported in the FSAR.
l Oropped Fuel Assembly Since the operation involves the handling of only one fuel assembly at a time, any handling accident will be bounded by the fuel handling accident described in the Trojan FSAR.
Collision of the Bridge Crane With the MFR$
The NFRS will be set up in the cask load pit area. The work platform section which supports the MFRS is equipped with l
removable handrails and ji) crane to allow the spent fuel bridge to pass over the MFRS. Operator action, with supervisory l
attention will prevent the spent fuel bridge from impacting on the installed handrails or tie jib crane. All personnel are l
thoroughly trained in the operation of the spent fuel bridge L
crane. Using these precautions, collision of the bridge crane j
and the MFRS will be prevented.
i Loss of Spent ^ 31 Pool Cooling I
The loss of sp f feci pool cooling system accident is adequately j
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N$.RCSCL-C\\L 88 748, Rev. 0 NS SECL 88 476, Rev. O PAGE 11 0F 20 discussed in the Trojan FSAR, Section 9.1.3.
There are no conditions resulting from the fuel reassembly / reconstitution process that would cause the heat load in the spent fuel pool to a incre: sed beyond that which is reported in the FSAR, nor are there any conditions existent which would invalidate the results of the analysis reported in the FSAR. Therefore, loss of spent fuel pool cooling during fuel reassembly / reconstitution is not a safety concern.
Loss of Fuel Building Exhaust System The spent fuel pool (5FP) exhaust system AB 4 is described in detail in the Trojan FSAR, Section 9.4.1.
The AB 4 system exhausts the SFP, the cask load pit and fuel transfer canal I
areas. This system consists of exhaust inlets located around the l
perimeter of the SFP, cask load pit and transfer canal which are connected by ductwork to two parallel exhaust plenums, each of which is designed to exhaust 19,375 cfa. Each plenum contains an exhaust fen, a bank of carbon absorbers, two banks of HEPA
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filters, an automatic roll filter, and motorized isolation dampers at each of the filter trains. Backdraft dampers are installed in each plenue to prevent reverse flow through an idle plenus. The system is designed for 100 percent capacity operation with one exhaust plenum in service while the other serves as a standby.
The Trojan plant Technical Specifications require that at least one spent fuel pool exhaust systes be operating and discharging l
through the High Efficiency Particulate Air (HEPA) filters and I
chemical absorbers whenever irradiated fuel is stored,
the spent fuel pool, and either fuel is being moved in the pool or a
other loads are being moved over the poo with either fuel L
buildingcrane(includingfuelhandlingtools). Meeting this requirement assures that fuel reconstitution work will be undertaken only if the AB 4 system is operating. The surveillance requirements of the Technical Specifications requires verification of system operation at least once per twelve hours under the above conditions.
With the plant's Technical Specification requirements being met with regard to the operation of the AB.4 system, repair of spent fuel in the cask area is acceptable. The consequences of an accident involving a fuel assembly involved in the reassembly / reconstitution procedure are less severe than with a fuel assembly being discharged from the reactor and being placed in storage. The fuel assemblies to be repsired were subjected to low burnup and have decayed for several years.
Fuel Repair Process - Inverting Fuel in MFRS Loss of Pellsts From Damaged Fuel Rods
NS RCSCL C\\l 88 748, Rev. O NS SECT. 88 476, Rev. O PAGE 12 0F 20 The rotatable basket is closed on its four sides and provides a clearance of approximately 0.060 inches on each side. This configuration, and the fact that both the top and bottom nozzles are in place during rotation, prevents the loss of fuel pellets during rotation of the fuel assembly. After reassembly / repair of the fuel assembly, but srior to opening the door, the floor of the cask load pit will se searched for fuel pellets and debris.
These pellets and debris will be retrieved using a debris suction system and air operated vice grips and placed in a d;bris canister. Therefore, loose fuel pellets and debth are not a safety concern because of the capability for recc.ary, and an isolated work location (lost pellets would be retained in the cask load pit). The debris canister will be stored in the spent fuel pool.
DiscreteRadioactiveParticles(ORPs)
Strict precautionary meas. ares are taken to avoid personnel contact with DRPs. All personnel involved in the fuel repair work will receive training in this area. The Radiation Work assure the safety of personnel. lists specific actions required to Permit (RWP),aPGEprocedure A zonal control system will be used with three zones for control of DRPs:
Theinner(red)zonewhereDRPsareexpected(orsuspected),
a buffer zone, and a clear zone In addition, highly sensitive portal radiation monitors will be used to prevent the spread of contamination.
All material removed from the spent fuel pool will be washed, wiped and decontaminated under the supervision and surveillance ofRadiationProtection(RP) personnel. Personnel will be required to wear two layers of protective clothing within areas potentially contaminated by DRPs, and to use respirators (if during equi sment removal.
RP personnel will thoroughly required)l equipmentseing removed from the water.
survey al Pcssible Damage to Old Fuel Assembly Inverting the fuel in the MFRS basket is a procedure that has been conductcJ on a large number of fuel assemblies without any resulting damage. The basket totally encloses the fuel assembly, with small clearances between the assembly and the basket, and the rotational acceleration and deceleration as well as the rotational speed are kept to a minimum, thereby precluding serious impact loads between the fuel assembly and the basket.
Potential for Adverse Stack Relocation Consequences
NS RCSCL C\\l 88 748, Rev. 0 NS-SECL 88 476 Rev. O PAGE 13 0F 20 There is extensive experience in both Europe and the United States with repair of fuel by inverting the assembly and removing the bottom nozzle. There is no evidence that this has resulted in any deleterious effects on the performance of the fuel after return to operation and subsequent refueling.
The equipment has been desigrad such that the fuel assembly is rigidly supported and that the inverting process is relatively slow, precluding any sharp loads to the assembly that would be any more significant than the loads during routine iuel transfers to and from the spent fuel pool area during refueling. These precautions, plus proper implementation of the ramp rate start-up procedures, are sufficient to conclude that thers is no significant increase in the potential for stack relocation or adverse operational consequences resulting from the inverting of fuel.
Fuel Repair Process Removal of Bottom Nozzle Noted here is the fact that the entire fuel repair process is conducted under water.
Debris Generation The fuel assembly bottom nozzle of assemblies to be repaired consists of a box like structure with an approximately 0.75 inch thick end plate. This and plate butts against, and is attached thimbles in to, the bottom end of the Red Cluster Control (RCC)imilar to the assembly. The attachment is made with screws s threaded into the ends of the thimbles,gh the nozzle end plate. Locking is shoulder bolts which are inserted throu l
and locked.
accomplishedwithlockingbars(wires)thatfitintotheslots across the screw heads and are welded at the ends to the counterbores around the screw heads.
To unlock the thimble screws, the welds are cut by grinding an annular groove around the head of the screw with a cutting tool i
)
having a diamond coated tip.
Fines generated by the 'grinding are removed from the cutting area with a suction system w11ch has a 25 micron strainer in the exhaust end. The strainer entraps all cutting debris which is large enough to be of concern. A small volume of grinding fines of minute size may pass throJgh the strainer, but neither the volume nor the size of these fines 4
constitute a safety concern because the minute size of these fines precludes the fouling of any pumps or valves and will not cause blockage of any flow areas required for cooling. At the completion of work, the ttrainer used to catch the cutting debris will b4 properly disposed of, or stored, by Portland General Electric.
Once the locking bars are cut. the bottom nozzle is removed by loosening each of the thimble.,ccm and backing then out of each thinible tube. During the no' le tiansfer, the screws are
NS RCSCL C\\L 88 748 Rev. 0 N$.SECL 88 476, Rev. O PA3E 14 0F 20 captured at all times. With the thimble screws resting in their respective botton nozzle counterbores, the bottcm nozzle is then removed from the fuel assembly and transferred to a submerged work platform. The screws are then removed from the nozzle and i
collected into a container. All screws from all fuel assemblies will be collected into containers. Portland General Electric will properly dispose of, or store, the screws upon completion of i
the fuel repair effort.
Since all the bottom nozzles from this fuel reconstitution effort utilize the welded type locking bar, the screws and nozzles are not reusable.
Each completed fuel assembly will have a new i
bottom nozzle installed with crimp lock thimble screws. Those bottom nozzles from fuel assemblies being reassembled will be left attached to the old remaining near-empty skeletons. The old skeletons will still contain the failed fuel rods from the reconstituted fuel assembly. Those bottom nozzles resulting from fuel assembites being reconstituted will not be attached and will require separate storage.
In all cases, the old skeletons and/or nozzles remaining when the fuel repair effort is completed will be properly disposed of, or stored, by Portland General Electric.
Following is a list of the anticipated debris resulting frem this fuel reassembly / reconstitution effort One or more containers of thimble screws and lock bars.
One strainer assembly with fines from the grinding process.
A very small volume of Lncollected fines of less than 25 microns.
Irradiated skeletons and or norries from renstambled or reconstituted fuel assemblies.
Tool tips used in the grinding process.
(Based on previous experience, these may be handled by personnel after removal fromthetooling.)
Loose fuel sellets, damaged fuel rods, and other debris will be placed in tie debris canister which will be stored in the spent fuel pool.
The uncollected fines of 25 microns or less will be collected in the filters of the domineralizer system for the spent fuel pool. These uncollected fines will not have a signi'icant impact on the period of service nor the radiation levels for the filters.
Potential for Damage to Fuel Rods The probability of damage to fuel rods during bottom nozzle removal is negligible. Because of the bottom nozzle construction and the method of attachment to the fuel assembly, the nozzle end
NS RCSCL C\\L 88 748, Rev. O NS SECL 88 476. Rev. O PAGE 15 0F 20 plate stands as a protective barrier between the lock bar cutter i
and the fuel rods.
The thimble screw lock bar grinding process uses a diamond coated tip that produces an annulus around each screw head. The depth of this annulus is approximately 0.030 inch deep and is self limiting in that this tip is not capable of making deep cuts. To cut through the bottom nozzle and expose the fuel rods would require a cut depth of about 0.75 inch. Thus there is no danger of the cutting tool contacting a fuel rod.
The sequence of removing the nozzle is such that the nozzle does not bear against the fuel rods at any time. As noted earlier, the nottle is attached to the fuel assembly with screws which mate with the RCC thimbles.
In the inverted position, the ends i
of the thimbles provide a surface to support the nozzle and preclude it from impacting the fuel rods.
i Once the screws are loosened, the nozzle is pulled from the fuel assembly with a fixture that is either lifted by hand or with a s
hoist.
f Potential for Dropping Equipment on a fuel Assembly i
The most serious consequence that could result from dro> ping a i
tool or a piece of equipment ou a fuel assembly while tle fuel t
assembly is in the basket would be the rupture of all of the rods in that fuel assembly. Such an accident is bounded by the fuel dron accident evaluation reported in the Trojan FSAR.
l t
Fuel Repair Process Fuel Rod Exchange Damaged fuel Rods Due to Operational Errors of NFRS In fuel rod exchanges, the rods are handled one at a time using manual operator dependent processes. Two pieces of equipment utllized in concert are required to accomplish the exchange.
a 1
These are:
1
(
(1) A fuel rod handling tool for extracting the fuel rods i
individually from the fuel assembly and inserting rods back 1
l into the fuel assembly.
I (2) Anindexingfixtura(eitherguideblocksorXYpositioner) 1, for locating the bottom end of the fuel rod handling tool at the correct cell location in the fuel assembly.
The fuel rod handling tool is comprised of an air actuated gripper for engaging the fuel rods and a drive system for raising ard lowering the gripper. The drive system is powered by an electric motor with speed and torque controls and transmits axial l
motion to the gripper by a sprocket and chain arrangement through i
. _ _ _ _. - ~ _. - _ _ _ -,
.. = - _ -.
17 1
NS-RC5CL-C\\l 88-748, Rev. 0 NS 5ECL 88 476, Rev. O PAGE 16 0F to a drive rod to which the cri>per is attached. The motor speed i
control limits the rate al witch fuel rods are extracted or inserted and the torque control limits the axial force exerted on the rods.
Torque limits are preset on the drive to limit the maximum pull or insertion forces imposed on the fuel rods. Also, the gri pper a
and fuel rod are constrained within an suter guide tube on tie t
handling tool over the full extracted rod height. The outer tube limits the lateral deflection of the fuel rod, thereby precluding excessive bending stresses.
During extraction of the fuel rod from the assembly, the motor torque is set to limit the pull force to a maximum of 100 lb.
In a small number of cases, the force recuired to initiate travel i
may exceed 100 lb.
In such cases, a ceviation in the force limit may be permitted. Under such conditions, extreme care is exercised and the operation is closely supervised. A manual handwheel is also available and may be used in place of the motor i
i 3 whre manual sensitivity is desired to initiate rod travel.
l3 Insertion of fuel rods into a fuel assembly is done very slowly i
and cautiously, particularly until the end of the fuel rod has entered and passed through the first support grid. Television i
camera verification is generally used as the rod enters the first i
grid.
In addition, the motor torque is generally reduced below the 100 lb. limit to the minimum necessary to seat the rods in i
i the assembly. After passage through the first grid, the l
insertion speed may b>e increased, f
j The tools, equipment, techniques and precautions described above, have been used in prior fuel assembly repairs tc transfer in i-excess of 1000 fuel rods without incurring damage.
I criticality l
The Westinghouse Nuclear Fuel Division has evaluated the I
l criticality effects of placing fuel assemblies into the MFRS i
transfer basktt. The assumptions used in this evaluation are as l
follows:
i f
All fuel rg contain uranium dioxide at an enrichment of l
3.55w/oU tver the infinite length of each rod l
)
No credit is taken for any U234 or U236 in the fuel, nor l7 is any credit taken for the buildup of fission product poison material, j
The moderator is pure water at g torperature of 68 F, A
0 i
conservative value of 1.0 gm/cm is used for the density j
of the water, l
l l
1 1
b
NSRCSCLC\\t88748,Rev.0 NS SECL 88 476, Rev. 0 PAGE 17 0F 20 l
No credit is taken for any spacer grids or spacer sleeves.
The array is infinite in axial extent which precludes any neutron leakage from the ends of the array.
5 The minimum poison material leading of 0.025 grams Gdg03 per square centimeter is used in tae MFRS.
The results of this evaluation show that it is acceptable to load i
two 17 X 17 standard fuel assemblies having up to 3.5 w/o i
enrichment, without burnup restrictiens, into the basket while i
yet maintaining a X eff of less than 0.g5.
Therefore the design basesgivenintheTrojanFSAR(Section9.1.2.1 are maintained.
Documentation of this tysluation is maintained )n the Fuel i
Division's engineering files.
I The Westinghouse Nuclear Fuel Division also perforced a criticality analysis of the fuel rod storage canister which showed that in its most reactive configuration (ly loaded with fullyloaded),it
{
i is much less reactive than an intact fuel assemb l
fuel rods of the same enrichment. Therefore, the handling and storage requirements for the fuel rod storage canistar are no nore restrictive than for any intact fuel assembly that has been licensed for storage in the fuel pool.
If the fuel from which 1
the rods were remeved was not allowed to be placed in a certain i
section of the fuel racks, then the fuel rod storage canister.
l 1esdod with the rods, is subject to this name restriction.
Documentation of this analysis is also maintained in the Fuel Division's engineering filos.
l Mislocattag Rods Within an Assembly The probability of misleadin!.ed such that all fuel movements arefuel rodsi low.
Procedures are formela individually identified by coordinates as to location rods are pulled from and subsequently relocated.
In the case of multiple rod movements between essemblies, multiple independent mis orientations would have to occur for the mistake to go undetected, e.g., if a rod were mis loaded, this would be detected later when either the rod to be moved is not available or a cell location to receive a red is determined to be l
occupied.
The detailed precedures involving both operator and f
independent checking of rod movement makes the probability for such errors low.
In the unlikely event that a fuel rod is mislocated, it is anticipsted that the nuclear design impact would be acceptable.
t Special Nuclear Material Accountablity l
}
Two failed fuel storage canisters will be used for the storage of
NS RC$CL C\\L 88 748. Rev. 0 NS-SECL 88 476 Rev. O PAGE 18 0F 20 replaced fuel rods from reconstituted assemblies. These fuel rods will be placed into specific locations within the canisters as s pecified by the procedures. Verification of proper plac uent in tie canisters will be accomplished using an underwater camera.
Reassembled fuel assemblies will normally retain their failed fuel rods with the old skeleton. New skeletons will be stemped with the letter 'R' to identify them as repaired assemblies.
Fuel Repair Process Returning Assemblies to Spent Fuel Pool Fuel Handling Accident In the unlikely event of a fuel handitag accident while returning the repaired fuel assembly to the spent fuel pool, such an occurrence would be no more severs than that of an accident involving the n.ovement of fuel as it is discharged from the reactor to the spent fuel pool. As such, the consequences of such an accident are enveloped by the analyses for fuel handling i
accidentsreportedintheTrojenFSAR.
Integrity of Old fuel Assembly f
This concern refers to the movement of the old skeleton back to i
the spent fuel pool following a reassembly. Reassembly involves
[
transferrir.g all intact fuel rods from a damaged skoleton to a new skeleton.
Following a reassembly, the bottoe nozzle is t
reinstalled onto the ole skeleton so that a top and bottos nozzle i
remain on the old skeleton. The old skeleton may then be handled l
with the spent fuel handling teel in the safne manner as any fuel assembly in the sMnt fuel pool.
An intact failed fuel rod or the segment of a broken fuel rod say be left in the skeleton tf there is reasonable assurance that they are held in place by grids.
$ttuations such as these are handled on a case by case basis as they arise and will be subject l
to approval by the Trojan Reactor Engineer, t
After the skeleton is removed from
- t fuel repair basket, it I
will be visually inspected and any ioose debris will be removed from the skeleton by mechanical means or by vacuusing.
i Radiological Consequences of a Fuel Handitng Accident j
An analysis was perfors.ed to determine the radiological consequences of 5
the limiting fuel handling accident during the fuel reassembly / reconstitution process and to compare the results with previously analyzed accident conditions. The results show that the site boundary doses from this analysis are auch lower than those presented in the Trojan Final Safety Analysis Report (FSAR) and are well within the 10 CFR 100 guidelines.
',}
NS ACSCL-C\\L 88 748 Rev. 0 NS SECL 88 476, Rev. O PAGE 19 0F to Since fuel repair is to be performed in the spent fuel pool area (cask load >it) of the fuel building, the postulated accident was analyzed only for tie fuel building (not containment) and the results compared with the FSAR accident analysis in the fuel building.
The radiological analysis was performed using the following assumptions:
l 1.
All of the fuel rods in one assembly are assumed to be damaged as a result of the handling accident.
All of the pgd pyvity in the damaged rods is released to the 2.
ac pool (cask
) water and conststs of 30 percent cf the total Krypton el activity. Note: only Kr al need be considered because of decay per assumption 5.
3.
Assembly fission product inventories are based on full power oper.ation at the end of life immediately precteding shutdown as i
given in F5AR Table 15.0 5 and a radial osaking factor of 1.65.
From FSAR Table 4.11, the full core contains 193 assemblies, I nl' 4.
The 0 I hour site boundary accident X/Q of 4.26 X 3 is used. This in the same X/Q used in the fuel 10'4sec/m j
handling accident described in Section 16.7.4 of the F5AR.
i 5.
A decay period of 5 years is assumed.
l 6.
Whole body dose models from FSAR Section 15.0.11 are used.
7.
Retention of noble gases in the spent fuel pool (cask load pit) 1: negligible (i.e.,DF=1.0) l 8.
All radioactivity that escapes from the spent fuel pool to the building is released from tse building over a I hour period.
9 g.
One spent fuel pool p h ust system is operating and discharging through the HEPA filters and chemical absorbers.
I 3 curies of l
Under these assueptions, the activity release of 1.86 X 10 Krypton 85 cccurs (all other noble gases and halogens decay to negitgible 1evelsin5 years). This activity release results in the following doses at the site boundary:
Thyroid dose 0
Beta skin dese 40.6 millirse G m a body dose 0.42 millirem a
Total dose 41.0 millirem These results are compared with the results of the fuel handling accident in the fuel building as presented in the FSAR in the following table.
y f,' e NS RCSCL C\\L-88 748. Rev. O NS 5ECL 88 476, Rev. O PAGE 20 0F 20 SITE SOUNDARY DOSE COMPARISON Whole Body Thyroid Analvn's Done (RDQ Dose fREM) 1 Assembly 2.03 18.1 (FSAR 96hr Decay) 1 Assembly (kapair) 0.041 0
(5yeardecay) 10 CFR 100 Guidelinas 25.0 300 NUREG 0800 Dese Limit 6.25 7f-(2516 of 10 CFR 100) 4 As can be seen in the comparison table, the radielo;tc0 consequences of tne postulated fuel handling accident during fuel feppir operations are substantially lower than those presented in the FSM nne PN well bounded by the FSAR analysis.
In addition, all of the dotn are wall within the 10CFR100guidelinesandtheStandardReviewPlan(V;&GC800) limits.
Conclusion Based on the foregoins it is concluded that performin the fuel rapair operations in the cask, load pit gren of the spent fuel;'ht1 ding et the Trojan Nuclear Plant does not involve an unreviewed safeb question. All postulated occurrences relating to the installation, op6 ration, teardown, decontamination and shipment of equipment have been consWired.
Coeplying with the requirements of the structural / seismic evaluatier. tusults assures the adequacy of the equipment to withstand the loads impostd by a suistic eventwithoutcausingdamagetospentfuelassembliesbeyondth6t accounted for in the plant s design basis. The equipment ar4 its installation and remov;l procedures have been tried and proven to be safe on several occasions at other nuclear sites.
All conceivable occurrences / accidents relating to the reassembly ad reconstitution of fuel assemblies have been evaluatea, and it is shown herein that this repair operation will not have an adverse effect on plant or personnel safety.
It will not increase the probability of occurrence or the consequences of an accident or malfunction of equipent important to safety previously evaluated in the FSAR. Nor has any mechanism for sn accident or malfunction, which has not been previously evaluated, been identified. Neither does the fuel repair operation decrease the margin o' safety as defined in the basis for any Technical Specification.
Therefore, conducting fuel repair operations, as discussed in this safety evaluation, does not represent an unreviewed safety question as defined in 10 CFR 50.59.