ML20203D650
| ML20203D650 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 12/12/1997 |
| From: | Dugger C ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| W3F1-97-0270, W3F1-97-270, NUDOCS 9712160254 | |
| Download: ML20203D650 (31) | |
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Charles M. Dugger s
nt. ore atu s W3F1-97-0270 A4,05-PR December 12,1997 U.S. Nucleat Regulatory Commission Attn.: Document Control Det,k Washington, DC 20555
Subject:
Waterford 3 SES Docket No. 50-382 License No. NPF-38 l
Request For Additional Information (RAl) Regarding Technical Specification Change Request NPF-38-193 l
l Gentlemen:
1 By letter dated March 27,1997, Waterford 3 proposed to amend Operating License l
NPF-38 to increase the Spent Fuel Pool storage capacity and increase the maximum l
fuol enrichment. The NRC review starf requested additional information, in their i
letter dated November 19,1997, regarding the proposed changes. This information is included in the enclosure entitled " Additional Information Regarding Technical Specification Change Request NPF-38-193." This additionalinformation has no effect on the previously providod no significant hazards determination.
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'Q; i Request For Additional information (RAI) R,garding
' Technical Specification Change Request NPF-38-1g3
- W3F1-97-0270:
- Page 2
December 12,1997 Should you have any questions or comments concerning the addnlonal information,-
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please contact Roy Prados at (504) 7394632.
Very truly yours, y-C.M. Dug 0er Vice President, Operations Waterford 3-r CMD/RWP/tmm
Enclosures:
Affidavit Attachments -
i cc:
_ E.W. Merschoff, NRC Region IV 3
C.P. Patel, NRC NRR NRC Resident inspectors Office (w/o attachments)
J. Smith L
N.S. Reynolds 4-Administrator Radiation Protection Division (State of t ouisiana)
American Nuclear Insurers k
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UNITED STATES OF AMERICA -
NUCLEAR REGULATORY COMMISSION in the matter of -
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Entergy _ Operations, incorporated
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Docket No. 50-382 Waterford 3 Steam Electric Station
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AFFIDAVIT Theodore Roy Leonard, being duly sworn, hereby deposes and says that he is i
General Manager Plant Operations - Waterford 3 of Entergy Operations, incorporated; that he is duly authorized to sign and file with the Nuclear Regulatory Commission the attached Additional Information Regarding Technical Specification Change Request NPF-38-193; that he is familiar with the content thereof; and that
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the matter:, set forth therein are true and correct to the best of his knowledge, information and belief.-
'i Theodore Roy Leonard General Manager Plant Operations - Waterford 3 STATE OF LOUISIANA
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) ss PARISH OF ST. CHARLES
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Subscribed and sworn to before me, a Nptary Public in and for the Parish and State above named this 2 > - day of 4 ~ ~...d-,
,1997.
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. Notary Public -
My Commission expires d 4Ml L
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4 ADDITIONAL INFORMATION REGARDING TECHNICAL SPECIFICATION CHANGE REQUEST NPF-38-193 ittm.1 Entergy Operations Inc. (EOI) proposes to increase the Spent Fuel Pool (SFP) storage capacity from the currently licensed capacity of 1088 fuel assemblies to 2398 fuel assemblies. This increase of SFP storage capacity will be achieved by reracking the SFP with new high-density racks of 1849 cells, installing additional high-density racks of 255 storage cells in the Cask Storage Pit, and after permanent shutdown installing additional high-density racks of 294 cells in the Refueling Canal. Please provide the following information:
Decay heat generation rates from the spent fuel assemblies stored in the Cask Storage Pit and the Refueling Canal as a function of time.
Cask Storage Pit and Refueling Canal water temperature as a function of time.
Detailed description of how the decay heat generated from the spent fuel assemblies stored in the Cask Storage Pit and the Refueling Canal will be removed.
Information should include: cooling system design parameters, equipment redundancy, seismic category, etc., and drawings to show cooling system configuration.
Detailed description to demonstrate how the drains in the Cask Storage Pit and the Refueling Canal will be plugged to preclude any water loss through the drains.
Discussion of the probability that the water level in the Cask Storage Pit and the Refueling Canal will be inadvertently drained belov a point approximately 10 feet above the top of the spent fuel.
In the event of complete loss of cooling in the Cask Storage Pit and the Refueling Canal, how long will it take the water to boil. Discuss the means for providing make-up water to these areas.
Res_ognse 1: first asterisk The time variation of the decay heat from the fuel assemblies stored in the Cask Storage Pit is shown in the figure provided in Attachment 1. The decay heat load in the Cask Storage Pit consists of the decay heat from 217 freshly discharged assemblies and 38 "old" or previously discharged assemblies. The corresponding curve for the Refueling Canalis presented in the figure provided in Attachment 2. The decay heat 1
i load in the Refueling Canal consists of the decay heat from 294 previously discharged assemblies. These heat generation rates have been computed using Auxiliary Systems Branch procedures, (ASB 9.2), which as stated in Holtec Position Paper WS-101, included as Attachment 3, overstates the heat load by a considerable margin.
ReSDOnSM1: Second asterisk Figures 5.8.1 and 5.8.2 of Holtec International Report HI 371628 (previously submitted on March 27,1997 as part of Technical Specification Change Request NPF-38-193) provide the average bulk pool temperature for the aggregate of the water mass in the Spent Fuel Pool, Cask Pit, and Refueling Canal. The spatial average temperature (i.e.,
bulk temperature) in each of the three bodies of water deviates from the overall bulk average by a small amount, because of the extensive interface (interconnection) between the three pools. The bulk average temperature within each body of water can be computed by utili2.ing the computational fluid dynamic (CFD) solution which provides a complete articulation of the temperature field in each " computation cell" throughout the three regions. Using the CFD solution, the peak bulk temperature in the Spent Fuel Pool, Cask Pit, and Refueling Canal for the limiting case (fu!I core offload with 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> hold time followed by fuel transfer at the rate of four assemblies per hour) was calculated. The following results were obtained:
Aggregate bulk pool temperature (Figure 5.8.2):
151.6*F Cask Pit average bulk temperature:
155.5'F (temperature difference of 3.9'F relative to the bulk pool temperature)
Refueling Canal average bulk temperature:
158.5'F (temperature difference of 6.9 F relative to the bulk pool temperature)
The above temperature differences can be applied to Figure 5.8.2 to obtain the Cask Pit and Refueling Canal average bulk temperatures as a function of time. The differential between the bulk temperatures in the three bodies will become smaller at non-peak instants. Even at the point in time when the transient effect is most pronounced (i.e., the instant of peak aggregate bulk temperature), the inter-body temperature differentials are quite small. Similarly, the Cask Pit and the Refueling Canal average bulk temperatures for the normal discharge scenario will be slightly higher than the temperature shown in Figure 5.8.1 (i.e., peak temperature slightly.
above 140*F). The aggregate bulk pool temperature, as shown in Figure 5.8.1, remains less than 140 F. The above data confirms the conclusions drawn in previous dockets that the thermosiphon effect in wet storage tends to homogenize the temperature field.
2
i Response 1: third esterisk The spent fuel decay heat in all three pools will be removed by a combination of the spent fuel pool cooling system and heat loss to the pool surroundings. The latter process consists primarily of evaporative heat loss. Therefore, the method of heat removal remains unchanged from those previously relied upon prior to the proposed change.
The tamperatures in the three pools are expected to be relatively equalized, as stated above. This equalization will take place primarily through interchange of water mass between the three regions. This water mass exchange process is driven by natural convection which takes place from the changes in water density due to small variations in temperatures, it is well documented that this process of natural convection also takes place within storage cells and adjacent pool walls forming convection cells. The warmer water rises in the cell, cools at the top of the racks, and falls along the racks cuiside perimeter in what is referred to as the "downcomer."
Convection cells will also be formed at the gate openings separating each of the regions. Through this convection process, the water masses in each of the regions will be constantly exchanged. Heat will flow from the warmer to cooler areas of the regions producing mixing and a nearly homogeneous fluid temperature throughout all three regions. Figures 5.8.9 and 5.8.10 of Holtec International Report HI-971620 depict the results of the full core discharge CFD analysis, which included the dimensional characteristics of the three regions. Figure 5,8.10 provides velocity vector plots of the fluid mass in all three regions. This figure clearly indicates the inter-mixing of the regions.
Administrative controls will be implemented to ensure that the gates are not installed when spent fuel is present in the Cask Storage Pit and/or the Refueling Canal.
Consistent with the practice over the past 10 years, the sparger lines in the Spent Fuel Pool will be truncated. In-depth analyses of temperature fields in fuel pools has shown that spargers do little to smear the temperature differentials. The remainder of the spent fuel pool cooling system remains essentially unchanged from its condition prior to the proposed change.
The cooling system is described in detail in Section 5.2 of Holtec International Report HI-971628 (included as part of the March 27,1997 submittal) and in much greater detail in Section 9.1.3 of the Wateriord 3 FSAR. The following specific highlights are extracted from Holtec International Report HI-971628 and the FSAR:
The cooling system is a closed loop consisting of two half capacity pumps and one full capacity heat exchanger. A backup fuel pool heat exchanger is also available, when the primary exchanger is out of 3
service. These components are all designated Nuclear Safety Related - Scismic Category 1.
A drawing showing the Spent Fuel Pool cooling system configuration is provided here3a as Attachment 4. The pump head / flow curves are provided herein as Attachment 5.
The data sheets for the primary and backup heat exchangers are provided herein respectively as Attachments 6 hnd 7. Component cooling water at a maximum temperature of 90 F is supplied to the shell side of each heat exchanger. The component cooling water total maximum flow, for fuel pool cooling, is 5000 gpm.
Response 1: fourth asterisk The 4' diameter drain holes in the Cask Storage Pit and Refueling Canal will be sealed by being covered with a plate measuring approximbtely 6*X6'X %". The Refueling Canal drain cover plate will not be installed until just prior to the installation of the racks in the Refueling Canal. An all around fillet weld will secure the plate to the %' liner surrounding the hole.
Response 1: fifth asterisk Draining of the Spent Fuel Pool was not a previously postulated event because the Spent Fuel Pool was designed to preclude draining. The construction of the Cask Storage Pit and Refueling Canal is very similar in nature to the Spent Fuel Pool, using reinforced concrete and the same liner thickness. Th9 drains at the bottom of inese two regions represent the only new significant difference between these regions and the Spent Fuel Pool with regard to the possibility of draining. Therefore, the drains will be welded shut, as discussed above in Response 1: fourth asterisk, to eliminate this possible drain path. The closure design provides the same level of leakage protection as that afforded by the liner welds and the underlying leak chase system previously existing in all three regions (liner leakage would be detected by the leak chase system; any leakage past the cover plate is not a credible event, but would be contained within the piping downstream of the welded closed drains; the isolation valves downstream of the welded closed drains will be maintained closed through administrative controls).
The probability that the water level in the Cask Storage Pit and Refueling Canal will be lowered due to draining is negligible because of the same level of drainage preclusion as is presently provided for the Spent Fuel Pool. In addition, if leakage did occur it would be detected by the Spent Fuel Pool low level monitor / alarm. Leakage from the Refueling Canal (into containment) through the fuel transfer tube is not a credible event, during normal operaticn, because the containment side of the fuel transfer tube is sealed with a bolted on cover (which is pressure tested each outage). Administrative controls in place for the installation and testing of the cover are procedures RF-006-001 ' Reactor Vessel Head and internals Installation' and STA-001-004 ' Local Leak Rate Test (LLRT) respectively. Leakage from the Cask Storage Pit into the Cask 4
(
Decontamination Pit through gate 3a is also not a credible event because this gate is seal welded closed.' This sew weld will be in place until just prior to removal of spent fuel from the Fuel Handling Building. L Therefore, drainage of the Cask Storage Pit and r
Refueling Canal is not a postulated event. Drainage of the Spent Fuel Pool is also currently not a postulated event. The Cask Storage Pit, and Refueling Canal were-originally designed to accommodate short term storage er movement of spent fuel.
Therefore, the proposed long term storage of spent fuel in these pools represents little, if any, change from the standpoint of protecting fuel from the possibility of being left with limited cooling / shielding coverage. Since all of the fuel in the proposed increased -
storage configuration is located at the same elevation and the pool gates will remain open at all times, lowering of the water level after the proposed modification represents the same consequence as prior to the modification However, this point is moot, since drainage of these two regions will be precluded.
i Resnonse 1: sixth asterisk The time-to-boil in the aftermath of loss of all forced cooling paths to the water mass represented by the three regions (SFP, Cask Pit, and Refueling Canal) is presented in Section 5.8 of Holtec International Report Hi-971628. Data on water level change subsequent to bulk boiling conditions with and without make-up is provided in graphical
. form in Figures 5.8.5 through 5.8.8 for the four postulated discharge scenarios. Since three regions of the pool are connected through large interfaces and the bulk temperatures in the three regions are very close to each other, the tinw-to-boil and post-boiling information presented in the above mentioned figures is applicable to all three bodies of water. -it is not possible to produce bulk boiling (with regions connected.
and administrative controls in place) in any one. region while maintaining a sub-boiling condition in others, because of extensive gravity induced heet and mass transfer fluxes.
82H12 With regard to the calculated decay heat loads following the proposed pool expansion, provide the decay heat generation rate as a function of decay time for both the routine o
refueling discharge and unplanned full-core offload conditions (information should clearly show the decay heat generation rate from each batch of the previously discharged spent fuel assemblies and the freshly discharged full core in the Spent Fuel Pool).
Response 2 The ' equested data is provided in Attachments 8A through BC.
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htDL1 Roracking of the SFP and installing additional racks in the Cask Storage Pit would only provide an increase in storage capacity which would maintain the plant's capacity to accommodate a full-core discharge through the end of Cycle 19 in 2018. However, EOl plans to operate for two additional cycles until 2022 without a full-core offload capability. Provide detailed justifications for the deviation to the guidance described in
- SRP Section 9.1.2 regarding spent fuel storage capacity.
Response 3 e
Entergy Operations, Inc. (EOl) plans to operate Waterford 3 until at least the year 2024, the last year of the current operating license. The proposed raracking provides full core discharge' capability until the year 2018, based on projected fuel cycle data.
L Therefore, the proposed reracking does not provide end of plant life spent fuel storage capability. Also, EOl is not proposing to operate Waterford 3 without full core discharge capability. The proposed reracking does provide Waterford 3 with en additional eighteen years of spent fuel storage capacity. It is hoped that this a(ditional time will allow the Department of Energy time to fulfill its obligation to take p. session of and store the Waterford 3 spent fuel. EOl plans to monitor and assess the Waterford 3 spent fuel storage situation as time progresses and will take the appropriste actions to safely store the Waterford 3 spent fuel and also maintain fuil core discharge capability.
Section 9.1.2.lli.1. of NUREG-0800 (SRP) states that "... for a single unit facility the 4
storage capacity shall equal or exceed one full core discharge plus the maximum normal fuel discharga cycle...." This is the minimum recommended spent fuel pool storage capacity and is a hold over from the time when reprocessing of commerr. al spent nuclear fuel was planned in the United States. Part I of Section 9.1.2 of NUREG-0787 (Safety Evaluation Report related to the operation of Waterford Steam Electric Station, Unit No. 3) states that the current Waterford 3 spent fuel pool "... facility 3 -
- provides high density underwater storage for 1088 fuel ass?mblies or approximately 5 full core loads."' The proposed reracking will provide storage for 2398 fuel assemolies or approximately 11 full core loads; therefore, the proposed reracking does not deviate
' from the guidance given in SRP Section 9.1.2.
BSDL4 Discuss the procedure to be utilized by the W6terford staff to monitor and control the SFP water temperature and decay heat load 'so as to remain within the design basis 1
limit values for routine refueling and planned or unplanned full-core offload events.
include discussion of the location of needed instrumentation, means of monitoring it 6
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1 and integration of operation staff activities with engineering staff activities in order to implement the procedure (s).
Response 4 Spent Fuel Pool temperature is monitored by a temperature probe which provides a Spont Fuel Pool temperaturo high alarm at 135'F in the Control Room. This alarm setting is conservative in relation to the normal (partial core) refueling limit of 140*F and the Waterford 3 self ir iposed abnormal (full core discharge) refueling limit of 155'F.
The thermal-hydraulic analyses were performed using limiting heat loads for two postulated normal and two postulated abnormal heat load conditions. These limiting heat loads were calculated using parameters that bound the actual conditions under the worst cases for each of the postulated scenarios, in the thermal-hydraulic analysis of the system (Chapter 5 of Holtec International Report HI 971628) the total number of stored fuel assemblies is conservatively assumed to be greater than the installed storage capacity (i.e.,2485 assemblies vs. the proposed 2398 assemblies). Other parameters such as fuel burnup and radial and axial peaking factors were routinely assigned values that conservatively bound the expected as-built parameters. The thermal-hydraulle analyses for raracking are routinely performed in this manner to alleviate any need to perform outage specific heat load calculations or monitoring.
Engineering review and input to the Reload Fuel Safety Analysis Groundrules Document (Groundrules) will ensure that these limiting parameters are not exceeded.
This review occurs during the design process for each new batch of reload fuel.
The first barrier for ensuring that spent fuel pool decay heat load and water temperature Ilmits are not exceeded is an engineering review of the Groundrules. The second barrier is system opera?.inq procedure OP 002-006, ' Fuel Pool Cooling and Purification." This procedure directs the oprators to maintain fuel pool temperature loss than 130'F. The third barrier is the actual monitoring of the Spent Fuel Pool temperature high alarm by operations personnel, in the event that the alarm did sound due to high fuel pool temperature, procedure OP-901-513 antiuod " Spent Fuel Pool Cooling Malfunction," and the Waterford 3 corrective action process, Site Directive W2.501, entitled " Corrective Action," would ensure that the appropriate corrective actions are taken.
7
LIST OF ATTACHMENTS ATTACHMENT 1:
Waterford 3 Cask Pit Decay Heat Variation with Time ATTACHMENT 2:
Waterford 3 Refueling Canal Decay Heat Variation with Time ATTACHMENT 3:
Holtec International Position Paper WS-101, Spent Fuel Pool Heat Loads and Pool Bulk Temperatures ATTACHMENT 4:
Waterford 3 drawing G-169, Flow Diagram Fuel Pool System ATTACHMENT 5:
Waterford 3 drawing 1564-1275R1, Fuel Pool Pump Curves ATTACHMENT 6:
Primary Heat Exchanger Specification Sheet ATTACHMENT 7:
Backup Heat Exchanger Specification Sheet ATTACHMENT 8A:
Decay Heat input to Waterford 3 Fuel Pool From A Freshly Discharged Full Core ATTACHMENT 88:
Decay Heat input to Waterford 3 Fuel Pool From a Normal (Partial Core) Discharge ATTACHMENT 8C:
Decay Heat Generation Rate From Previously Discharged Assemblies
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HOLTEC IN'IERNATIONAL POSITION PAPER WS 101 SPENT FUEL POOL HEAT LOADS AND POOL BULK TEMPERATURES l
Revision 0: December 21,1995 i
It is common knowleds: in the nuclear power industry that the heat load burden on the spent fbel pool cooling system is grossly overestimated. 'Ibere are two sources of the overestimate:
(i) heat loss through evaporation of pool water, (ii) huge conservatism in the decay heat load calculations. 'Ibe net result of these overestimations is an overly conservative assessment of the magnitude of the pool bulk temperature. Mei== pool water temperatures are routinely overpredicted by as much as 20*F to 30*F, because of the at,c,m. M conservatisms in T
the calculations.
In 1989, this deficiency-in the state +f-the-art was partially reniedied when Holtec International performed a series of pool evaporative heat loss mea..urements at Millstone Unit 3 under the sponsorship of Northeast Utilities. These so<:alled " pool evaporative loss" e@ur.ia, carried out under Holtec's 10CFR50 Appendix B Q.A. progrsm, were correlated with a theoretical formuistion, resulting in a reliable formalism for estimating heat loss in spent fbel pools.
The second, and mort l@ test, source of error arises from the inaccuracy in the NRC's ASB 9.2 and ANS' standard 5.1, customarily used to compute decay heat loads. Neither of these two calculational methods correlated well with benchmark data. 'Ihe attached figure shows a comparison between the ASB 9.2 and Oak Ridge National Labs' code ORIGEN 2.
ORIGEN is known to have <*d<tauble built-in conservatism, the ASB 9.2 result is even more conservative.
Fortunately, this limitation, too, is a tidng of the past.
Holtec International has recently implemented a fully 14r4=wked decay beat code - named DECOR
- based on work done at the Oak Ridge National Laboratory. This program, along with the evaporative heat loss, results in pool bulk temperature predictions which are projected to be Inodestly <maarvative (by roughly 2'F - 5'F compared to the actual pool water temperature).
Implementation of these enhancements in the pool bulk temperature evaluations would help utilities plan the reactor decay time (time for passive decay of fuel before transfer to the pool commences) by utilizing a reliable pool temper 3nne predictive vehicle, instead of the coarse, groasly conservative methodology in use today.
i Response to Questions on Technical Specification Change
- Request, NPF 38193-ATTACHMENT 3
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DECAY HEAT COMPARISON BE# WEEN ASB9-2 and ORIGEN2 Millstone Unit 1, 2011 MW(t), 3.97% enrichment, 46000 MWD /MTU 4000
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ATTACHMENT 6
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QTLQS INDUSTOICL MF00 CO o' '
e CLIFTON, NEW JED9EY EXCHANGER SPECIFICATION SHEET Je6 N9 CUST9MER Combustion Engineering Comphny PETERENCE NO.
02 ADDRESS Windsor, Connecticut IN0VIRY NO.
A 70 PLANT LOCATION Taft, Loui siana DATE T.29 71
~
669-71 SERV!CE OF UNIT Fuel Pool Heat Exchanger ITFM No.
SIZE 44 2 6 te TYPE CEN CONNECTED IN SURFACE PER UNIT 4313 SHELLS PER UNIT l
SURFACE PER SHELL 4313 Alternate Condition Rev.#4 SHELL SIDE TUPE SIDE FLUID CIRCULATED Inhibited Water Boric Acid Solution TOTAL FLUID ENTERING a/HP.
249P924 438 1000000 000 VAPOR
- /HR.
LIQUID a/HR.
STEAM
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NON-CONDENSADLES
- /HR.
i FLUID VAPORIZED OR CONDENSED STEAM CONDENSED M9LECULAR WEIGHT-VAPORS LATENT HEAT-VAPORS a.T.U./a SPECIFIC HEAT B.T.U./8 1 000 0 999 DENSITY
- /CU.FT.
61 950 61 800 VISCOSITY CP.
0 711 0 630 THERMAL CONDUCTIVITY 0 364 0 365 TEMPERATURE IN F
100 000 120 000 l
TEMPEPATURE SUT F
104 960 107 600 l
OPERATING PRESSUPE
- /SO.!N.
IR5 000 50 000 FOULING RESISTANCE 0 00050 0 00057E l
NUMBER OF PASEES I
2 VELOCITY.
FT./SEC.
4 463 3.R79 i
PRESSURE DROP
- /SO.IN.
14 706 2 694 NEAT EXCHANGED-R.T.U./HR.
12389335 625 MTD(CORPECTED) 9 856 TRANSFER RATE-SERVICE 300 817 CLEAN 444 587 I
CONSTRUCTION PER SHELL i
DESIGN PRESSURE
- /SO.IN.
150 000 75 000 TEST PRESSURE 8/SO.!N.
225 000 112 500 DESIGN TEMPERATURE F
250 000 250 000 TUPES SS-304, SA-249 0.D.
0 750 P!TCH 0 9375 TRI NUMPER OF TURES 1000 PWG.
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RAFFLES-SA 34 TH1CKNESS 0.$00 PITCH 37" TYPE SEGMENTAL BAFFLE-LONG THICXNESS TYPE GASKETS CONNECTIONS-SHELL IN 18 000 OUT 18 000 SERIES R. W.
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C9 PROS!ON ALLOWANCE-SHELL SIDE 1/8" TUPE SIDE None CODE REQtJIREMENIS SEc{D u.A 197,1 E 0 TEMA CLASS R
WEIGHTS-EACH SHELL 20 590 60NDLE FULL OF WATEn 32,382 REMARMStVl8 PAT 10N ANALY$1S Response to Questions on STRt RAT 10s a.747 SPANS 20 900 Technical Specification Change
'N0Zt RAT 108 5 401 SPANS jh.933 Request NPF 38-193 ATTACHMENT 6
4 ATTACHMENT 7
7 YCDA HEAT TOANSFER CORPORA 780']
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ATTACHMENT 8A l
i
ll 0
ATTACHMENT SA: DECAY HEAT INPUT TO WATERFORD-3 FUEL POOL FROM A FRESHLY DISCHARGED FULL CORE Time After Reseter $butdows (Hrs)
Decay Heat (Mullom Btu /Br)
DURING FUEL TRANSFER OPERATIONS 72 0
100 24.09 127 42.05 POST FUEL TRANSFER 151 39.11
-175 36.79 201
?4.81 250 32.01 299 29.93 347 28.30 404 26.66 453 25.44 509 24.20 595 22.53 697 20.86 i
Response to Questions on Technical Specification Change Request NPF-38193.
l ATTACHMENT 8A
s' b
4 1
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ATTACHMENT 8B l
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ATTACHMENT 88: DECAY HEAT INPUT TO WATERFORD-3 FUEL POOL i
FROM A NORMAL (PARTIAL CORE) DISCHARGE i
Time After Reactor Ehetdows (Hrs)
Decay Best (Millies Stu/Br)
DUk!NO FUEL TRANSFER OPERATIONS 72 0.00 101 25.08 POST FUEL TRANSFER 3
149 21.25 196 19.02 249 17.37 311 15.99 371 14.97 419 14.27 503-13.23 602 12.20 680 11.51 5
Response to Questions on Technical Specification Changs Request NPF-36193 ATTACHMENT SB
-~.
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ATTACHMENT 8C i
I t
l 0
ATTACHMENT 80: DECAY HEA' GENERATION RATE FROM PREVIOUSLY DISCHARGED FUEL ASSEMBLIES Cycle Number Batch Size Years Since Non Dimensional Discharge Decay Power 1
92 39.4 0.0041 2
84 38.0 0.0039 3
84 36.6 0.0040 4
84 35.1 0.0041 5
84 33.6 0.0043 6
92 32.1 0.0049 7
96 30.6 0.0053 8
96 29.1 0.0055 9
88 27.5 0.0052 10 92 25.7 0.0057 11 100 24.0 0.0064 12 116 22.0 0.0079 1:i l16 20.0 0.0082 14 116 18.0 0.0086 15 116 16.0 0.0091 16 116 14.0 0.0095 17 116 12.0 0.0100 18 116 10.0 0.0103 19 116 8.0 0.0110 20 116 6.0 0.0117 21 116 4.0 0.0134 22 116-2.0 0.0213 o
(TOTAL) 0.1746 SPECIFIC ASSEMBLY POWER = 57.64 Million 5tu/Hr TOTAL BACEGROCMD = 0.1746 s $7.64 Millloa = 10.M Million Btu /Hr Response to Questions on Technical Specification Change Request NPF 38-193 ATTACHMENT BC
OVERSIZE DOCUMENT PAGE(S) PULLED SEE APERTURE CARD FILES
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