ML19312B996
| ML19312B996 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 07/09/1975 |
| From: | Parker W DUKE POWER CO. |
| To: | Giambusso A Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19312B997 | List: |
| References | |
| NUDOCS 7911270775 | |
| Download: ML19312B996 (34) | |
Text
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, NRC Did(RIBUTION FOR PART 50 DOCKET .mATERI AL
, , , (TEMPORARY FORM)
FILE:
FROM: Buke Power Company DATE OF DOC DATE REC'D LTR TWX RPT OTHER Charlotte, NC 7-9-75 7-14-75 XX W O Parker .fr TO: ORIG CC OTHER SENT NRC PDR xx Mr Giambusso one signed SENT LOCAL PDR YY CLASS UNCLASS PROPINFO INPUT NO CYS REC'D DOCKET NO:
XXXXX 1 M 70/287 65SCRIPTION: ENCLOSURES:
Ler te our 12-27-75 order for modification Proposed andt to OL/ Change to Tech Specs:
.....trans she,,foiw" - M STE D C naisting f rem sed ECCS acceptance Criteria j\[n;.U 4 &-addl & supporting info for the change. . . . . .
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PLANT NAME: oconeee 1-3 $
FOR ACTION /INFORMATION 7-14-75 ehf BUTLER (L) SCHWENCER (L) ZIEMANN (L) REG AN (E)
W/ Copies W/ Copies W/ Copies W/ Copies CLARK (L) STOLZ (L) DICKER (E) LEAR (L)
W/ Copies 'N/ Copies W/ Copies W/ Copies PARR (L) VASSALLO (L) KNIGHTON (E) S?IES W/ Copies W/ Copies W/ Copies W/ Copies KNIEL (L) PURP,LE (L) YOUNGBLOOD (E) LFM W/ Copies / W/fCopies W/ Copies W/ Gopies g INTERNAL DISTRIBUTION EG F TECH REVIEW DENTON LIC ASST A/T IN D f,u n L ro R SCHROEDER GRIMES R. DIGGS (L) BRAITMAN OGC. ROOM P SCO,4 MACCARY GAMMILL H. GEARIN (L) SALTZMAN
/ GOSSICK/ST AF F KNIGHT K ASTN E R E. GOU LBOURNE (L) MELTZ
/ CASE PAWLICKI BALLARD P. KREUTZER (E)
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P. COLLINS /TEDESCO YOUNGBLOOD H. SMITH (L) HARTFIELD (2)
DENISE J. COLLINS REGAN S. TEETS (L) KLECKER R_E G O P R 7 LAIN AS PROJECT LDR G. WILLI AMS (E) EISENHUT
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Mr.AngeloGiambusso,Directorly] V db <' y 5 i-j Division of Reactor Licensing - ' '
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Re: Oconee Nuclear Station GA 2 .x Docket Nos. 50-269, 50-270, 50-287 s (.h/ *
Dear Mr. Giambusso:
Pursuant to the Commission's December 27, 1974 Order for Modification of Licesse for Oconee Nuclear Station, Units 1, 2 and 3, a re-evaluation of ECCS cooling performance has been performed. This re-evaluation utilized a calculational model conforming to the provisions of 10 CFR 50, Section 50.46. Based on this re-evaluation, Duke Power Company is sub-mitting herewith proposed changes to Oconee Technical Specifications 3.1.3.5, 3.1.7, 3.5.2.3 and 3.5.2.5. The proposed revisions are indicated in the attached replacement pages (Attachment 1). The pro-posed changes to Specifications 3.1.3.5 and 3.5.2.3, which are necessary to incorporate minimum ejected rod worth criteria into the rod withdrawal limits, are comparable to those submitted on May 9, 1975 and are included here for completeness.
The evaluation model utilized in performing the re-evaluation of'ECCS cooling performance is described in Babcock and Wilcox (B&W) non-proprietary Topical Report BAW-10104, "B&W's ECCS Evaluation Model." Non-proprietary Topical Report BAW-10103, "ECCS Evaluation of B&W's 177 FA Lowered Loop NSS," describes the results of the re-evaluation for a generic B&W unit of the Oconee class. Since the generic parameters used in the re-evaluation are conservative for units of this type, and the parameters associated with Oconee 1, 2 and 3 are bounded by those utilized in the generic analysis, BAW-10103 provides a conservative evaluation of ECCS performance for the Oconee units. In the case of Oconee 1, portions of the analysis presented in BAW-10103 were also performed utilizing specific Oconee 1 parameters. This analysis, and the results thereof, are described in Attachment 2.
The results presented in BAW-10103 and in Attachment 2 demonstrate the conformance of the Oconee units to the criteria of 10CFR50, Section 50.46 7 R1;'
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Mr. Angelo Giambusso Page 2 July 9, 1975 under the operating conditions specified in the attached proposed technical specifications.
In addition to the above, and in accordance with Mr. Robert A. Purple's letter of June 13, 1975 to Mr. A. C. Thies, the following information is also provided:
- 1. Break Spectrum and Partial Loop Operation The requested information for operation with four reactor coolant pumps is presented in BAW-10103 and in Attachment 2. For operation with three or two reactor coolant pumps, the proposed Technical Specifications are based on ECCS limits for four pump operation and on ejected rod worth criteria. Additional information concerning the limits for three and two pump operation will be provided by August 1, 1975.
- 2. Potential Boron Precipitation The requested information was provided by Mr. A. C. Thies' letters of April 16, 1975 and day 30, 1975 in response to Mr. Purple's letter of March 14, 1975.
- 3. Single Failure Analysis A single failure analysis for manually-controlled, electrically-operated ECCS valves has been performed and is provided as Attach-ment 3 hereto. Valves 1CF-1, 2CF-1 and 3CF-1, and 1CF-2, 2CF-2 and 3CF-2, core flood tank discharge isolation valves, identified in Attachment 3 are ECCS valves which are currently required by Technical Specifications to have power disconnected during operation
- see Technical Specification 3.3.3(c) . Based on the information
- provided in Attachment 3, it is concluded that no station modifi-cations or changes to the Technical Specifications are necessary in
, order to protect against a single failure of a manually-controlled, electrically-operated ECCS valve.
- 4. Submerged Valves A review of equipment arrangement to determine if any valve motors within the Reactor Building will become submerged following a LOCA has been performed and those valves which may be affected are identified in Attachment 4. As can be seen, flooding of the valve motors identified would have no effect on short term or long term ECCS functions, or on containment isolation.
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Mr. Angelo Giambusso Page 3 July 9, 1975
- 5. Containment Pressure The containment pressure used to evaluate the performance capability of the ECCS has been calculated in accordance with the methods described in BAW-10104 and the results thereof are presented in .
BAW-10103.
Also, as requested in the staff's Safety Evaluation Report which accompanied the Commission's December 27, 1974 Order, and in ac-cordance with Mr. Purple's letter of February 10, 1975 to Mr. A. C.
Thies, as-built passive containment heat-sink data have been compiled and are given in Attachment 5.
Timely approval of the attached proposed Technical Specifications is requested. In the interim, in accordance with the Commission's December .
27, 1974 Order for Modification of License, operation of the Oconee units l is continuing within the limits of.
l l
(a) The requirements of the Interim Acceptance Criteria, the Technical Specifications, and license conditions imposed by the Commission in accordance with the requirements of the Interim Acceptance Criteria, and '
(b) The limits of the proposed Technical Specifications submitted by the licensee on September 20, 1974 and August 5, 1974, as modified by the further restrictions set forth in Appendix A to the Order.
Forty copies of this letter and attachments are enclosed.
Ver' truly yours,
/
N -
hWilliam u iO. &. n Parker, Jr.
DCH:vr Enclosures
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g b I Mr. Angelo Giambusso Page 4 July 9, 1975 ,
WILLIAM 0. PARKER, JR., being duly sworn, states that he is Vice President of Duke Power Company; that he is authorized on the part of said Company to sign and file with the Nuclear Regulatory Commission this request for amendment of the Oconee Nuclear Station Technical Specifications, Appendix A to F ility Operating Licenses DPR-38, DPR-47 and DPR-55, and that all stat nts and matters set forth therein are true and correct to the best of s knowledge.
[/WilliamwL & . M w, O. Parker, Jr. ,
Vice President ATTEST:
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Dorothea B. Stroupe /
Assistant Secretary l
Subscribed and sworn to before me this 9th day of July, 1975. l
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Notary Public My Commission Expires:
D e:wl u A . 197 *l
Attachment 2 Oconee 1 ECCS Evaluation i
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. OCONEE I, CYCLE 2 ,
LOCA LIMITS
- 1. Introduction In Babcock and Wilcox Topical Report BAW-10103 , generic LOCA limits ap-plicable to the 177 FA class plants with lowered-loop arrangement (Category 1 plants) were developed. LOCA limits specific to Oconee 1, Cycle 2 fuel are developed in this attachment as were developed in the Oconee 1, Cycle 2 reload report, BAW-1409. The analysis is performed using the B&W evaluation model documented in BAW-101042 ,
- 2. Summary and Conclusions The LOCA limits developed herein are specific to Oconee I, Cycle 2 fuel.
The analysis performed to generate these limits meet all the requirements of 10CFR50, Section 50.46. Section 4 of this attachment gives the allowable peak linear heat rates as a function of elevation within the core. The results of the analysis are tabulated below and are shown graphically on Figure 1.
Elevation From Allowable Linear Local Bottom of Core, Heat Rate, Peak Cladding Metal-Water ft kw/ft Temperature,F Reaction, %
2.0 16.0 1930 3.40 4.0 17.5 1978 3.17
- 6.0 *18.0 *2146 *5.46
- 8.0 *17.0 *2110 *5.19
- 10.0 *16.0 *1931 *2.93
- These numbers are from the generic LOCA limits analysis presented in BAW-10103.
- 3. Input 3 4 5 The method of selection of the input for the CRAFT , REFLOOD , CONTEMPT ,
andTHETA6codesforthegenericevgluationoftheCategory1plantsis discussed in Section 4 of BAW-10103 . Also presented in that section is the procedure used in the evaluation of a LOCA. This section will describe the deviations from the input chapter in BAW-10103. To facilitate the ECCS evaluation specific for the Oconee 1, Cycle 2 fuel the following assumptions are made.
(a) Power Level - Prior to a postulated pipe rupture, the plant is as-sumed to be operating in a steady-state condition at 102 percent of rated power, 2568 MWt.
(b) Stored Energy - The initial average fuel temperature takes into
. account the densification phenomenon. The Batch 4 fuel is used to generate the LOCA limits as described in the Cyc c 2 Reload Report, BAW-1409.
(c) Minimum Capacity - The performance of the LPI system is modeled l conservatively at 90 percent of the design flow rate. Also, the CFT l resistance used is 6.5 which is a higher value than that obtained by l testing. The HPI has not been modeled for the purposes of core l cooling in the analysis.
(d) Backpressure - The determination ofminimum backpressure during the reflood phase of a LOCA is based on conservative assumptions. The actual input parameters and calculational methods are discussed in Section 3.3.
3.1 CRAFT The input description and calculation procedures outlined in Section 4.2 of BAW-10103, with the exception of the value quoted for the core flooding line resistance, remains valid. The core flooding line resistance used is 6.5 which is a value that is higher than the measured value for Oconee I. Use of a higher value for the core flooding line resistance is conservative as this will lead to a longer refill period and lower flooding rates and will result in a higher peak cladding temperature than expected.
3.2 REFLOOD Since the value for the core flooding line resistance has changed from that reported in RAW-10103, the effect of the hot wall time delay must be reexamined. As shown in Section 4.3 of BAW-10103, 3 The therateofavailablewaterstoragevglumeis119ft/sec.
injection rate of ECC water is 108 ft /sec. Since the storage volume rate exceeds the injection rate, there will be no additional ECC water spillage due to the hot wall phenomena. Therefore, the hot wall time delay has no effect on the length of refill.
The remainder of the input selection description given in Section 4.3 of BAW-10103 is unchanged by the specific analysis of Oconee I.
- 3. 3 CONTEMPT The CONTEMPT input description in Section 4.4 of BAW-10103 was formulated on a generic basis. The entire section, with the exception of the procedure used to generate mass and energy releases, is revised herein. The following assumptions were used for the actual CONTEMPT analysis:
(1) Initially, the Reactor Building is at 110 F, 13.7 psia, and 100 percent relative humidity. Data taken from the Oconee 1 Reactor Building from December 1973 to June 1974 indicate that
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the pressure is nearly always -1.0 psig and varies (though rarely) up to +1.0 psig. The temperatures averaged over the same period were 940F for the steam generator cavities and 129 F for the dome space. Since tne dome makes up over half the building free volume, the temperature for CONTEMPT was weighted equally between these two numbers and rounded down.
Values for relative humidity are not readily available, so the conservative value of 100 percent was assumed.
(2) The outside air temperature is 40 F.
i (3) The containment free volume, including'a 5 percent conservatism, is 2,010,000 ft3 Calculations consisting of subtracting the volumes of the internal structures and installed equipment from the volume determined from the building walls and ceiling inside dimensions show the SAR value is less than 3 percent low. Thus, 5 percent is a suitably conservative value for ECCS evaluation.
(4) All heat removal systems are actuated. The delay times are based on an assumption of no loss of off-site power (see Section 4.6 of BAW-10103 for single failure position). This results in no delay for the fan coolers. The time delay to the sprays is 65 seconds from the start of the accident. A breakdown of the starting time which yields the quickest activation time for the spray systems, is shown below.
0 2.4 12.4 ESF Delay MOVs Open 8.4 68.4 Accident l
BS Pumps BS Headers Full )
at 100% Speed l f
6s 60s BS System l Operstional Time, s :
l The coolant water for these systems is assumed to be at 40 F l 6
and the sprays are considered 100 percent efficient. The 40 F temperature is derived from a lower Technical Specification limit for the conditions in the BWST. The total fan cooler heat removal (HR) rate as a function of Reactor Building at-mosphere temperature is HR (BTU /s) = 3.0 (0.9T - 76.9T + 1670)
q 1
' where 3.0 represents the number of fan coolers in the contain-ment. The fan coolers are the largest units for the Category 1 Reactor Buildings. The coolant water is assumed to be 400 F, and the coolers are taken as clear and unfouled. The spray flow rate is 1800 gpm for each of the two spray systems. The design flow for the sprays is 1500 gpm. The spray efficiency is taken as one, so that the water is completely heated to the contain-ment vapor temperature as it falls.
(5) Complete mixing of the spilled ECCS water with the containment atmosphere is considered. The HPI injecting in the broken loop is considered part of the spilled ECCS vater.
1 (6)Rainoutofsuspendedwaterigassumed,andahighheattransfer coefficient of 1000 BTU /h-ft - F between the liquid and vapor regions is used.
(7) The building is modeled with five heat sinks:
(a) The Reactor Building walls and dome including the concrete wall, steel liner, and anchors:
Exposed area, ft 82,950 Paint thickness, ft 0.00083
- Steel thickness, ft 0.04384 Concrete thickness, ft 3.25 (b) Painted internal steel
Exposed area, ft 218,272 Paint thickness, ft 0.00083 Steel thickness,'ft 0.03125 (c) Unpainted internal steel:
I Exposed area, ft 31,728 Steel thickness, ft 0.03125 (d) Unpainted stainless steel: ,
i Exposed areas, ft 10,000 Steel thickness, ft 0.03125 The total area of sinks b, c, and d is calculated from Figure 4-14 of BAW-10104 using the (5 percent margin factor not ap-plied) building free volume. The surface area of slab d was obtained by selecting the maximum stainless steel area from a survey of six Category 1 plants.
1 The unpainted steel is obtained by applying the same percentage '
i used in BAW-10103 and is given in slab c. The area of the painted steel is found by subtracting the areas of sinks e and d from the total area found from Figure 4-14 of BAW-10104. Ap-propriate stainless steel properties are used for slab d.
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' (e) Internal concrete:
Exposed area, ft 160,000 Paint thickness, ft 0.00C83 Concrete thickness, ft 1.0 (8) The following thermophysical properties are used.
Thermal Conductivity HeatCagacity, Material BTU /h-ft2-F BTU /jt - F Concrete 0.92 22.62 Steel 27.0 5d.8 Stainless Steel 9.1836 54.263 Paint 0.6215 40.42 (9) The condensing heat transfer coefficients given in Section 4.3.
l 6.1 of BAW-10104 are used:
(a) At the end of blowdown, as.=ume a maximum heat transfer co-efficient four times higher than that calculated by Aerojets' Tagami correlation:
hm,x = 72.5 (Q/Vtp )0.62 where h = maximum heat transfer coefficient, BTU /h-ft2 _ p, Q = primary coolant energy deposit to containment at end of blowdown, BTU, V = net free containment volume, ft ,
t = time interval to end of blowdown, s.
p Before the end of blowdown, assume a linear increase from h
31
= 8 BTU /h-ft F to the peak value specified above.
(b) During the long-term stagnation phase of the accident, characterized by low turbulence in the containment atmos-phere, assume condensing heat transfer coefficients equal to 1.2 times the one obtained from the Uchida data. The Uchida heat transfer coefficients are shown in Table A-1 of BAW-10095.
(c) During the transition in phase of the accident between the end of blowdown and the long-term, post-blowdown phase, a reasonably conservative exponential transition with a decay constant of 0.0255 is used.
^
1 c-3.4 T5 ETA 1-B Section 4.5 of BAW-10103 gives the input description for the genn.ric analysis. The writeup given in that section remains applicable with the exception of the quoted TAFY7 parameters. The revised parameters specific to the Oconee I, Batch 4 fuel are shown below.
Fuel Parameters-THETA Inout Fuel Diameter, in Cladding ID, in 0.370753(*))
0.377673(*
Cladding OD, in 0.430140 Cladding Surface Roughness, in 0.00003 Fuel Surface Roughness, in 0.00007 Poisson's Ratio 0.35 Gas Composition, moles Air 0.0027555 Helium 0.04681285 l Krypton 0.00000041 Xenon 0.00000041 (a) 18 kw/f t linear heat rate 3.5 Single Fa13ure The single failure used in this analysis is identical to that presented in Section 4.6 of BAW-10103.
! 4. LOCA Limits i
4.1 Analytical Method The allowable linear heat rate (kw/ft) limits as a function of elevation within the core are developed using the method and input assumptions outlined in Section 4 of BAW-10104 and Section 3 of this attachment. The results of the sensitivity analysis performed in Section 5 of BAW-10103 are used as outlined in Section 7.1 of that report. Since BAW-10103 was a generic report applicable to all Category 1 plants, the results of the sensitivity studies in Section 5 remain valid for the Oconee I NSS.
The LOCA limits analysis is performed by analyzing the worst break with different axial power distributions. In Section 6 of BAW-10103,theyorstbreakfortheCategory1plantsisshowntobeThe the 8.55 ft DE break at the pump discharge with a Cp = 1.0.
power shape is varied in CRAFT so that the maximum axial peaking factor (1.7) is located at the point of interest. The pcwer shapes used are shown in Figure 7-1 of BAW-10103. Separate CF. AFT analyses are run for each power shape studied. The peak linear heat rates analyzed in CRAFT are greater than or equal to the allowable linear heat rates that are developed. Tgis approach is conservative as noted in Section 7.1 of BAW-10091
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U'aing the CRAFT predictions at the end of blowdown, separate REFIDOD runs are made for each axial location studied. The resultant flooding rates are used to develop heat transfer coefficients during the reflood portion of the accident. For the peak power locations at or below 6 feet, the modified FLECHT correlation is used as dis 9 cussed in Section 4.3.6.5 of BAW-10104. Above 6 feet, the REFLECHT correlation is used.
THETA is run for each axial location using a model representative of the hot spot modeled in CRAFT. The power level in THETA is lowered from the CRAFT peak power level until the criteria of 10CFR50.46 are met.
4.2 Allowable kw/ft Figure 1 shows the results of the LOCA limits analysis for the Oconee I fuel. For elevations below 6 feet, specific limits based on Oconee I, Batch 4 fuel were generated. For elevations 6 feet and a'bove, the generic limits are used. This is conservative because, as shown on Figure 1, the specific analysis would generate higher limits.
Figures 2 through 9 show the results of the LOCA limits analysis at the 2 and 4 foot elevations. For elevations above 4 feet, the parameters of interest for the LOCA limits analysis is presented in Figures 7-10 and 7-21 of RAW-10103. The results of the LGCA limits analysis are tabled below.
Core Elevation, Ft Allowable Feak 2 4 6* 8* 10*
Linear Heat Rate, kw/ft 16.0 17.5 18.0 17.0 16.0 Peak Cladding temp of unrupt node,/ time,F/s 1930/43.5 1978/43.0 2146/61.0 2110/132.0 1931/135.0 Peak cladding temp of rupt node / Time,F/s 1882/43 1975/43 2066/45.0 1743/68.0 1642/45.0 Initial Pin Pressure PSIA 1910 1725 1300 1450 760 Rupture Time, s 10.90 12.39 15.55 15.01 39.2
, Local Metal-Water Reaction, % 3.40 3.17 5.46 5.19 2.93
- Numbers from Table 7-1 of BAW-10103
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- 5. Compliance to l'0CFR50.46 The analysis presented in this attachment and the pertinent sections of BAW-10103 demonstrate compliance to the five requirements of 10CFR50.46:
(a) Peak Cladding Temperature - The maximum peak cladding temperature calculated is 2146 F.
(b) Maximum Cladding Oxidation - The maximum local cladding oxidation calculated is 5.46 percent.
(c) Maximum Hydrogen Generation - Compliance to the 1 percent whole-core chemical reaction limit is demonstrated in Section 8 of BAW-10103, where the whole-core reaction is conservatively calculated to be 0.557 percent. This is unaltered by the specific analysis presented in Section 4 of this attachment for elevations below 6 feet. The method of calculating whole-core chemical reaction is performed at the elevation which yields the highest local metal-water reaction for a given kw/ft. This evaluation shows that only elevations of 6 feet or higher are used.
(d) Coolable Geometry - The discussion given in Section 9 of BAW-10103 is applicable to the Oconee I NSS and shows that the core geometry will remain amenable to cooling.
(e) Lona-Term Coolinz - Section 10 of RAW-10103 and Mr. A. C. Thies' letters of April 16 and May 30, 1975 in response to Mr. R. A. Purple's letter of March 14, 1975 demonstrate that long-term cooling is ensured because of the redundant pumped injection systems used on the Oconee I plant, and that precipitation of salts will not occur.
- 6. References R. C. Jones, J. R. Biller, and B. M. Dunn, ECCS Analysis of B&W's 177-FA Lcwered-Loop NSS, BAW-10103, Babcock & Wilcox, June,1975.
B. M. Dunn et al., B&W's ECCS Evaluation Model, BAW-10104, Babcock &
Wilcox, May, 1975.
R. A. Hedrick, J. J. Cudlin, and R. C. Foltz, CRAFT 2-Fortran Program for Digital Simulation of a Multinode Reactor Plant During Loss of Coolant, BAW-10092, Rev. 2, Babcock & Wilcox, April,1975.
4 B. E. Bingham, and K. C. Shieh, REFLOOD-Description of Model for Multinode Core Reflood Analysis, BAW-10093, Babcock & Wilcox, March, 1974.
5 Y. H. Hsii, CONTEMPT-Computer Program for Predicting Containment Pressure-Temperature Response to LOCA- B&W's Revised Version of Phillips Petroleum Company Program (L. C. Richardson, et al. , June, 1967), BAW-10095, Babcock & Wilcox, July,1974.
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6R'. H. Stoudt and K. C. Heck, THETA-B-Computer Code for Nuclear Reactor Core Thermal Analysis- B&W Revisions to IN-1445 (Idaho Nuclear, C. J.
Hocevar and T. W. Wineinger), BAW-10094, Rev. 1, Babcock & Wilcox, April, 1975.
C. D. Morgan, and H. S. Kao, TAFY-Fuel Pin Temperature and Gas Pressure Analysis, BAW-10044, Babcock & Wilcox, April,1972.
B. M. Dunn, et al., B&W's ECCS Evaluation Model Report With Specific Application to 177-FA Class Plants With Lowered Loop Arrangement, BAW-10091, Babcock & Wilcox, August, 1974.
B. M. Dunn, et al., REFLECHT Correlation, BAW-10091P, Appendix B, Babcock & Wilcox, August, 1974, (PROPRIETARY) .
- 7. Computer Data on Figures Figure Version No. Version Name Date Run Name Run Date 2 THETA 1B VERSION 6F 01/23/75 T1410UL 07/02/75 3 CRAFT 2 Version SPP 04/17/75 FC141QF 06/30/75 4 REFLOOD 2 NO LOOP VERSION 12/20/74 RF1411V 07/02/75 5 THETA 1B Version 6F 01/23/75 T1410UL 07/02/75 6 THETA 1B Versien 6F 01/23/75 T1430WK 07/03/75 i 7 CRAFT 2 Version SPP 04/17/75 FC1431B 07/02/75 8 REFLOOD (MIMIC) Version 1 12/20/74 RF143SY 07/02/75 9 THETA 1B Version 6F 01/23/75 T1430WK 07/03/75
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- ELEVATION IN CORE, ft ALLOWABLE KW/FT VS ELEVATION IN CORE - 177 FA GENERIC ,ANO OCONEE 1, BATCH 4 RESULTS Figure 1
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0 40 80 120 160 TIME S HOT SPOT HEAT TRANSFER COEFFICIENT-LOCA LIMITS-0CONEE I, BATCH 4 16.0 KW/FT AT 2 FT ELEVATION FIGURE 5
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0 20 40 60 80 100 120 140 160 180 TIME, S RUPTURED AND UNRUPTURED NODE CLADDING TEMPERATURE - LOCA LIMITS - OCONEE 1, BATCH 4-17.5 KW/FT AT 4 FT ELEVATION Figure 6
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Attachment 3 Manually-Controlled, Electrically-Operated ECCS Valves Single Failure Analysis
_. _ _ .- - _ - . _. - = -- -
OCONEE NUCLEAR STATIO74 SINGLE FAILURE ANALYSIS OF MANUALI.Y-CONTROLLED, ELECTRICALLY-OPERATED VALVES IN EMERGENCY CORE COOLING SYSTEM Valve Identification Valve Position
- ""## E " " "" "
s t on Fa ue -
A. High Pressure Injection System ,
lilP-2 3 lip Pump Normal Suction (from 1. DST) Open Closes No effect on ECCS capability. HP pump
" suction from LDST is not required for 211P-23
" ECCS performance.
3tiP-23 lilP-98 IIP Pump "A" & "B" Suction Open Closes Would prevent suction tolW pump "B." ,
2ilP-98 Ileadere isolation But suction to the other two HP pumps )
311P-98 would not be affected.
lilP-ll 5 IIP Pump "A" & "B" Discharge Open Closes Would prevent IIP injection by Pump "B."
2ilP-ll5 lleader isolation But ilP injection by "A" and "C" would 311P-il5 not be affected, liiP-120 Normal RC makeup control valve in Throttled Will Not Would prevent throttling of IIP in-2ilP-120 parallel with IIP Injection Valve Close jection flow rate through "A" string.
311P-120 llP-26 But throttling capability is not a requirement for ECCS performance.
IllP-31 RC pump seal flow control Throttled Will Not Operability of this valve is not a re-211P- 31 Close quirement for ECCS performance. Also,.
311P-31 the flow through this line is negligible compared to the available ,
ECCS flow. -')
B. Low Pressure Injection System ILP-19 RB emergency sump isolation (Line A) Closed 1. Opens If the valve opened immediately subse-2LP-19 quent to a LOCA, flok through one LPI-3LP-19 string could possibly be reduced since or part of the BWST water entering that LP ILP-20 RB emergency sump isolation (Line B) pump suction would be diverted to the i 21.P-2C RB emergency sump. This would have no 31.P-2 0 effect on the HPI strings or on the other LPI string.
_j-Valve Identification Valve Position
~
No. Description g y "f* Evaluation
- 2. falls to Other line admits necessary flow.
open after -
a LOCA ILP-6 1.P Pump "C" Suction "A" lleader 2LP-6 Closed Opens ,'
3tP-6 or _ No effect on LPI capability.
ILP-7 2LP-7 I.P Pump "C" Suction "B" lleader Closed Opens '
3LP-7 -
)
ILP-5 2LP-5 LP Pump "A" Suction Open Closes, 3LP-5 Causes loss of suction to one LP pump, or - But suction to the other LP pump is ILP-8 not affected.
21.P-8 LP Pump "B" Suction Open Closes
2LP-9 31.P-9 ILP-10 LP "C" pump discharge to "B" header Closed Opens No effect on LPI capability.
2LP-10 3LP-10 , )
ll.P- 11 Cooler "A" inlet Open Closes ,
21.P-11 ,
Stops flow in that LPI string; but or other LPI string is not affected.
ILP-13 Cooler "B" inlet Open Closes s 2LP-13 ILP-12 Cooler "A" outlet Open Closes s Stops flow in that LPI string; but 2LP-12 other LPI string is not affected.
or .
ILP-14 Cooler "B" outlet Open Closes s 2LP-14
3-Valve _ Identification Valve Position ,
""* ""' E'" os on a ti e 3LP-12 Cooler "A" outlet Open Closes, Would cause loss of cooling in one or -
string. (The bypass line would provide 31.P- 14 Cooler "B" outlet Open Closes' the LPI flow rate in that string.) No effect on the other LPI string.
11.P- 69 I.P1 Cooler "B" bypass Closed Opens No effect on LPI capability. I f "C
2LP-69 pump was being used as an LP pump, would cause partial loss of cooling in "B" string.
3LP-92 LP1 Cooler "A" bypass Closed Opens ,'
No effect on LPI capability. Would , T "B" bypass cause partial loss of cooling in one 31.1 - 93 LPI Cooler Closed Opens LPI string.
ILP-3 RC return (Dil line) isolation valve Closed Opens No effect on LPI capability because 21.P- 3 of the normally closed manual valve LP-4 31.P-3 RC return (Dit line) isolation valve Closed Opens No effect on LPI capability since the pressure at the LP suction line is less than the relief valve LP-25 setpoint pressure of 388 psig.
II.P- 15 1.P discharge to RB spray and llP Closed Opens No effect on LPI capability.
2LP-15 (A loop) 3LP-15 or ILP-16 LP discharge to RB spray and HP Closed Opens No effect on LPI capability.
2LP-16 (B loop) 31.P-16 C. Core Flooding System ICF-1 Not 2CF-L Tank "A" Outlet Open Credible Technical Specification 3.3.3(c) requires 3C F- 1 breaker for valve operator to be locked or open.
ICF-2 Not 2CF-2 Tank "B" Outlet Open Credible 3CF-2
__I
Valve Identification Valve Position
- """# E " #" "" " ~
os on a ue ICF-3 Tank "A" sample and drain Closed Opens ,' No effect on ECCS performance since-2CF-3 there is a normally closed manual valve 3CF-3 (CF-19) in series with the power
~
operated valves, or ICF-4 Tank "B" sample and drain Closed Opens '
2CF-4 1 3CF-4 ICF-5 Tank "A" vent Closed Opens This line is utilized during operation 2CF-5 for CF tank venting to the quench tank.
3CF-5 If the valve opened during CF tank discharge, the tank discharge flow rate could be affected. However, the proba-bility of a valve single-failure during or CF tank discharge is very low due to the I CF-6 Tank "B" vent Closed Opens short time interval necessary for CF tank 2CF-6 discharge and due to the functional use 3CF-6 of these valves during operation.
I s
m m, Attachment 4 Post-LOCA Submerged Valve Motor Analysis t
i 1
1 I
OCONEE NUCLEAR STATION ANALYSIS OF VALVE MOTORS WilICll MAY BECOME SUBMERGED FOLLOWING A LOCA Valve Identification No. Description Evaluation ,
ICF-1 "A" Core Flood Tank Discharge Valve No effect on ECCS capability or containment integrity.
2CF-1 Valve is locked open during operation and is not 3CF-1 operated subsequent to a LOCA. Valve is not a containment isolation valve.
ICS-5* Reactor Building Isolation Valve for Quench No effect on ECCS capability or containment integrity.
2CS-5* Tank Drain Valve is normally closed. Redundant isolation valve 3CS-5* on outside of containment is not affected. s J
lilP-1 Reactor Coolant Inlet to "A" Letdoun Cooler No effect on ECCS capability or containment integrity.
2HP-1 Letdown coolers are not used following a LOCA, Valve 311P- L is not a containment isolation valve.
lilP-2 Reactor Coolant Inlet to "B" Letdown Cooler No effect on ECCS capability or containment integrity.
2ilP-2 Letdown coolers are not used following a LOCA. Valve 311P-2 is not a containment isolation valve, lilP-3
- Reactor Coolant Outlet from "A" Letdown No effect on ECCS capability or containment integrity.
211P-3* Cooler and Reactor Building Isolation Valve Letdown coolers are not used following a LOCA.
3HP-3* Redundant isolation valve on cutside of containment is not affected.
, l!!P-4* Reactor Coolant Outlet from "B" Letdown No effect on ECCS capability or containment integrity.
2 tiP-4
- Cooler and Reactor Building Isolation Valve Letdown coolers are not used following a LOCA.
3HP-4* Redundant isolation valve on outside of containment ,)
is not affected.
ICC-1 Component Cooling Water Inlet to "A" No effect on ECCS capability or containment integrity.
2CC-1 Letdown Coolers Letdown coolers are not used following a LOCA. Valve 3CC-1 is not a containment isolation valve.
ICC-2 Component Cooling Water Inlet to "B" No effect on ECCS capability or containment integrity.
2CC-2 Letdown Coolers Letdown coolers are not used following a LOCA. Valve 3CC-2 is not a containment isolation valve.
i s
Attachment 5 As-Built Passive Containment Heat-Sink Data
s .-
OCONEE NUCLEAR STATION CONTAINMENT VOLUME AND HEAT SINK DATA Parameter As-Built 6
Reactor Building Free Volume (ft ) 1.836 x 10 Building Cylinder Surface Area (ft2) 61,353 Thickness Concrete (ft) 3.75 (nominal)/3.8 (equivalent)
Steel (ft) 0.0208 Paint (mils) 7-10 Mass 6 Concrete (lb) 35.2 x 10 6 Steel (1b) 0.59 x 10 Building Dome 2 Surface Area (f t ) 16,230 Thickness Concrete (ft) 3.25 Steel (ft) 0.0208 Paint (mils) 7 Mass 6 Concrete (1b) 7.91 x 10 6 Steel (lb) 0.166 x 10 Building Base 2
Surface Area (f t ) 8890 Thickness Concrete (ft) 8.5 Steel (ft) 0.0208 Paint (mils) 7 Mass Concrete (lb) 11.3 x 10 6 6 Steel (lb) 0.091 x 10 i
Internal Concrete 2 Surface Area (ft ) 66,318 Thickness concrete (ft) 1.76 Paint (mils) 7-10 Mass 6
Concrete (1b) 17.5 x 10 Internal Steel (Painted)
Surface Area (ft2) 159,746 Thickness Steel (ft) 0.0316 Paint (mils) 7 Mass 6
Steel (1b) 2.47 x 10 l
- es u.
Parameter As-Built InternalSteel(Ugpainted) 54,153 Surface Area (f t )
Thickness (ft) 0.0078 Mass (1b) 0.21 x 10 6 Stainless Steel Surface Area (f t ) 17,553 Thickness (ft) 0.026 6 Mass (1b) 0.229 x 10 AS-BUILT HEAT SINK THERMOPHYSICAL PROPERTIES Specific Thermal Heat Conductivgt Material Densit'f)
(1b/ft (BTU /lb- F) (BTU /hr-ft gF Concrete 150 0.213 1.05 Carbon Steel 490 0.12- 32.0 Stainless Steel 501 0.11 9.4 l
1
- -.