ML19210A622
| ML19210A622 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 08/08/1975 |
| From: | Arnold R METROPOLITAN EDISON CO. |
| To: | |
| Shared Package | |
| ML19210A621 | List: |
| References | |
| NUDOCS 7910300656 | |
| Download: ML19210A622 (16) | |
Text
.
B.8-?S METROPOLITAN EDISON COMPAhi JERSEY CENTRAL POWER & LI6tlT COMPANY AND PENNSYLVANIA ELECTRIC COMPANY THREE MILE ISLAND NUCLEAR STATION UNIT 1 Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request No.17 This Technical Specificatf on Change Request is submitted in support of Licensee's request to change Appendix A to Operating License No. DPR-50 for Three Mile Island Nuclear Station Unit 1. As a part of this request, proposed replacement pages for Appendix A are also included.
METROPOLITAN EDISON COMPANY
/1/ T' ',\ ,/ ,i
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By (N U {i , 'f V d \.
Vice President-Generation Sworn and subscribed to me this day of C M , 1975 GM ,
/ Notary Public RICH ARO 1. RUTH v n PW c 'Ir. - 'cr; i.vp., Ecrks Co.
My C mm.ss.ca Exp:ss u; tem:ar 23,1978 1492 191 7810soo g g
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.j UNITED STATES OF AMERICA j, ' ~.
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- NUCLEAh REGULATORY COMMISSION b, "'#"'"
IN THE MATTER OF DOCKET NO. 50-289 OPERATING LICENSE NO. DPR-50 METROPOLITAN EDISON COMPANY This is to certify that a copy of Technical Specification Change Request No. 17 to Appendix A of the Operating License for Three Mile Island Nuclear Station, Unit 1, dated August 8,1975, and filed with the U.S. Nuclear Regulatory Commission August 8,1975, has this 8th day August, 1975, been served on the chief executives of Londonderry Township, Dauphin County, Pennsylvania, and of Dauphin County, Pennsylvania, by deposit in the United States Mail, addressed as follows:
Mr. Weldon B . Arehart, Chairmar Mr. Charles P. Hoy, Chairman Board of Supervisors of Board of County Commissioners of Londonderry Township Dauphin County R.D. #1, Ceyers Church Road Dauphin County Courthouse Middletown, Pennsylvania 17057 Harrisburg, Pennsylvania 17120 METROPOLITAN EDISON COMPANY
/
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Vice President-Generation 1492 192
Containment Pressure has been calculated using TMI-1 Specific Parameters.
As built containment passive heat sink data has been compiled and is shown in Attachment 2.
B&W has completed a computer calculation utilizing the TMI-l Specific Parameters with the generic method to determine containment back pressure, in the event of a LOCA. Attachment 3 gives the results of the analysis and shows that the TMI-l building yields a higher containment pressure than that calculated using the generic containment model described in Section k.h of BAW-10103. Therefore the minimum contain=ent back pressure transient used in the ECCS ane. lysis (BAW-10103) is conservative relative to the actual results for the TMI-1 building. The generic back pressure was also used in the development of the proposed Technical Specifications therefore in this regard they are conservative for TMI-1.
In some cases these proposed Technical Specifications are more restrictive and in others less restrictive than the present Technical Specifications.
As stated in our submittal of July 9, 1975 ve vill continue to operate within the most restrictive and conservative limits of proposed and existing Technical Specifications until the NRC Staff has had time to review and act upon this change request.
In su= mary, this change vill assure adequate protection for the health and safety of th'e public in that i' was derived from approved NRC guidelines.
1492 193 h
i
THREE MILE ISLAND NUCLEAR STATION UNIT I (TMI-I)
Operating License NO. DPR-50 Docket NO. 50-289 8-8-?S Technical Suecification Change Request NO.17 The licensee requests that the attached changed page a replace pages 3-16, 3-34, 3-35, 3-35a, 3-36, Figure 3 5-2A, Figure 3 5-2B, Figure 3 5-2C, Figure 3.5-2D, Figure 3 5-2E, and Figure 3 5-2F of the existing Technical Specifications.
Reasons for Prorosed Change These proposed Technical Specifications are pursuant to 10CFR50 Appendix K, your order for Modification of License dated Dece=ber 27, 1974, and the concerns of your letter of June 18, 1975 and are necessary in order to ensure compliance with the Final Acceptance Criteria (FAC) of 10 CFR50.h6.
Safety Analysis Justifying Change The proposed Technical Specifications are based on the Babcock & Wilcox (B&W) ECCS evaluation model for B&W 177 fuel assembly class plants with lowered loop arrangement (BAW-10104), that has been determined by the NRC Staff to meet the require =ents of Appendix K of 10CFR Part 50, and also based en the evaluation for this class of plants BAW-10103.
Further it should be noted that the analyses were conducted to satisfy NRC concerns as more explicitly expressed in May 1975 regarding partial pump operation and in this regard further information is provided in Attachment 1 to this change request.
Since the proposed Technical Specification rod position limits are based on ECCS power peaking, minimum shutdown margin, and ejected rod worth, the proposed rod position limits are based on the most limiting of the three criteria, which for TMI-1 are ECCS uover reaking and ejected rod worth. The current TMI-l Technical Specifications include restrictions on msxi=un ejected rod worth at hot zero and rated power in Spec. 3 5 2.3 Inclusion of the ejected rod worth criterion into the rod position limits makes the existing Technical Specification 3 5.2.3 redundant. However, Technical Specificction 3.5.2 5 on Rod position limits allows the positions limits to be exceeded for a period of up to h hours before a violation of the~ Technical Specification limiting conditions for operation is considered to exist. This time has been deter =1ned on the basis that the LOCA has a very low probability of occurrence that deviations from the limits should be permitted for short periods. We consider the ejected rod accident to be on the same order of probability as the LOCA and the same deviations from limits should also be allowed. (i.e. operation in Restricted Region)
Other ECCS issues (Boron Precipitation, Single Failure Analysis and Submerged Valves) related to these proposed Technical Specifications have been addressed in our submittal of July 9, 1975 Therefore, other than B?eak_ Spectrum and Partial-Loop Operation, which is addressed [
above, the only re=aining issue is Containment Pressure. T
}kh2
3.17 MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY Apolicability Applies to maximum positive moderator te=rcrature coefficient of reactivity at full power conditions.
Objective To assure thac the moderator temperature coefficient stays within the limits calculated for safe operation of the reactor.
Suecification 3.1.7.1 The moderator temperature coefficient shall not be positive at power levels above 955 of rated power.
Bas es A non-positive moderator coefficient at power levels above 95% of rated power is specified such that the maximum clad temperatures vill not exceed the Final Acceptance Criteria based on LOCA analyses. Belov 95% of .ated power the Final Acceptance Criteria vill got be exceeded with a positive moderator te=perature coefficient of +0.5 x 10- AK/K/F. All other accident analyses as reported in the FSAR have been performed for a range of moderator temperature coefficients including +0 5 x 10-h AK/K/F.'
The experinentalvalue of the coderatcr coefficient vill be corrected to obtain the hot full power moderator coefficient. The correction factor will be verified dut i ng startup' testing on earlier E&W reactors.
The Final Acceptance Criteria states that post-LOCA clad temperature vill not exceed 2200 F.
REFERENCES (1) FSAR, Section 1k (2) FSAR, Section 3 5
\h92 \9I 3-16
- f. If a control rod in the regulating or axial power shaping groups is declared inoperable per Specification h.T.l.2., operation may continue provided the rods in ihe group are positioned such that the rod that was declared inoperable is maintained within allow-able group average position li=its of Specification h.7.1.2.
- g. If the inoperable rod in Paragraph "e" above is in groups 5, 6, 7, or 8, the other rods in the group shall be tricted to the same position. Normal operation of 100 percent of the ther=al power allovable for the reactor coolant pu=p combination may then continue provided that the rod that was declared inoperable is maintained within allowable group average position limits in 3.5 2 5 3.5.2.3 The vorth of single inserted control rods during criticality are limited by the restrictions of Specification 3.1.3 5 and the control Rod Position Limits defined in Specification 3.5.2.5 3.5.2.h Quadrant tilt:
- a. Except for physics tests if quadrant tilt exceeds k percent, power shall be reduced icnediately to below the power level cutoff (see Figures 3.5-2A and 3 5-2B). Moreover, the power level cutoff value shall be reduced 2 percent for each 1 percent tilt in excess of h percent tilt. For less than four pump operation, thermal power shall be reduced 2 percent of the thermal power allovable for the reactor coolant pump combination for each 1 percent tilt in excess of 4 percent.
- b. Within a period of h hours , the quadrant power tilt shall be reduced to less than h percent except for physics tests, or the following adjustments in setpoints and limits shall be made:
- 1. The protectien system reactor power / imbalance envelope trip setpoints shall be reduced 2 percent in power for each 1 percent tilt.
- 2. The control rod group withdrawal limit's (Figures 3.5-2A 3.5-23, and 3.5-2C) shall be reduced 2 percent in power for each 1 percent tilt in excess of k percent.
- 3. The operational imbalance limits (Figure 3.5-2D) shall be reduced 2 percent in power for each 1 percent tilt in excess of h percent.
3-3h 1492 196
3.5.2 5 Control rod Positions:
- a. Operating rod group overlap shall not exceed 25 percent,
+ 5 percent, between two sequential groups except for physics tests.
- b. Except for physics tests or exercising control rods, the centrol red insertion /vithdrawal limits are specified on Figure 3 5-2A (from control rod interchange up to hhD full power days of operation), Figure 3.5-2B (for after hh0 full power days of operation) for four pu=p operation, and Figure 3 5-2C for three or two pu=p operation. If the control rod position limits are exceeded, corrective measures shall be taken immediately to achieve an acceptable control rod positian. Acceptable control rod positions shall be attained within four hours.
- c. Except for physics tests, power shall not be increased above the power level cutoff (See Figures 3 5-2A and 3.5-2B) l ualess the xenon reactivity is within 10 percent of the equilibrium value for operation at rated power and asymptotically approaching stability.
- d. Core imbalance shall be monitored on a minimum frequency of once every two hours during power operation above 40 percent of rated power. Except for Physics tests, corrective l measures (reduction of imbalance by AFSR movements _and/or reduction in reactor pover) shall be taken to maintain operation within the envelope defined by Figure 3.5-2D. If the imbalance is not within the envelope defined by Figure 3 5-2D, corrective measures shall be taken to achieve an acceptable imbalance. If e.n acceptable imbalance is not achieved within four hours, reactor power shall be reduced until imbslance limits are met.
- e. Safety rod limits are given in 3.1.3.5 3.5 2.6 The control rod drive patch panels shall be locked at all times with limited access to be authorized by the superintendent.
3 5 2.7 A power map shall be taken to verify the expected power distri-bution at periodic intervals of approximately 10 full power days using the incore instrumentation detection system.
Bases The power-imbalance envelope defined in Figure 3.5-2D is based on LOCA analyses which have defined the maximum linear heat rate (see Figure 3.5-2E) such that the maximum clad temperature vill not exceed the Final Acceptance Criteria (2200F).
Operation outside of the power imbalance envelope alone does not constitute a situation that would cause the Final Acceptance Criteria to be exceeded should a LOCA occur. The power imbalance envelope represents the boundar;r of operation v t 3-35 992 d}
limited by the Final Acceptance Criteria only if the control rods are at the withdrawal / insertion limits as defined by figures 3,5-2A, and 3 5-23, and if a L percent quadrant power tilt exists. Additional conservatirm is intro-ducted by application of:
- a. Nuclear uncertainty factors.
- b. Ther=al calibration uncertainty.
- c. Fuel densification effects,
- d. Hot rod manufacturing tolerance factors.
The Rod Index versus Allowable Power curves of Figures 3.5-2A, 3.5-2B and 3 5-2C describe three regions. These three regions are:
- 1. Permissible operating Region
- 2. Restricted Regions
- 3. Prohibited Region (Operation in this region is not allowed)
Note: Inadvertant operation within the Restricted Region for a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not considered a violation of a limiting condition for operation. The limiting criteria within the Restricted Region are potential ejected rad worth and ECCS power peaking and since the probability of these accidents is very lov especially in a k hour time frame, inadvertant operation within the Restricted Region for a period of k hours is allowed.
1 9
1 3-35" 3
1492 11%
The 25+5 percent overlap between successive control rod groups is allowed since l the vorth of a rod is lower at the upper and lover part of the stroke. Control rods are arranged in groups or banks defined as follows:
Groun Function 1 Safety 2 Safety 3 Safety b Safety 5 Regulating 6 Regulating 7 Xenon translent override 8 APSR (axial power shaping bank)
Centrol rod groups are withdrawn in sequence beginning with group 1. Groups 5, 6, and 7 are overlapped 25 percent. The normal position at power is for groups 6 and 7 to be partially inserted.
The rod position limits are based on the most limiting of the following three criteria: ECCS power peaking, shutdown =argin, and potential ejected rod worth. As discussed above, compliance with the ECCS power peaking criterion is ensured by the rod position limits. The minimu. tvailable rod vorth, consistent with the rod position limits, provides for achieving hot shutdown by reactor trip at any tite, assuming the highest worth control rod that is withdrawn re=cins in the full out position (1). The rod position limits also ensure that inserted rod groups vill not contain single rod worths greater than: 0.65% Ak/k at rated power. Tl se values have been shown to be safe by the safety analysis (2) of the hypothatical rod ejection accident. A ma>itum single inserted control rod worth of 1.05 Ak/k is allowed by the rod position limits at hot zero power.
A single inserted control rod vorth of 1.0% Ak/k at beginning of life, hot ,
zero power vould result in a lover transient peak ther=al power and, therefore, less severe environmental consequeraes than a 0.65% Ak/k ejected red worth at rated power.
The plant computer vill scan for tilt and inbalance and vill satisfy the technical specification requirements. If the computer is out of service, then =anual calculation for tilt above 15 percent power end icbalance above 40 percent power must be performed at least every two hours until the ec=puter is returned to service.
The quadrant power tilt limits set forth in Specification 3.5.2.h have ceen established within the ther=al analysis design base using the definition of quadrant power tilt given in Technical Specifications, Section 1.6.
During the physics testing program, the high flux trip setpoints are ad=ini-stratively set as follows to assure an additional safety targin is provided:
Test Power Trin Setnoint 0 <55 15 505 50 75 0
855
}k
>TS 105.55 3-26
REFERE!*CES (1) PSAR, Section 3 2.2.1.2 (2) FSAR, Section ik.2.2.2 e
1492 200 3-36a
RCD POSITIO.' _IMITS FOR 4 PUMP OPERATIL APPLICASLE OURING THE PERIC FRCM 253110 EFFO TO 4401. EFPD.
(
(177.43 o2)
(152g IC2) -
i (222.3 302) 100 -
(** h '9 ) ,
PO'.1ER LEVEL CUT 90 -
0PERATION IN Tills REGION 0F F (222.3,90) -
(172.3,87)
IS NOT ALLO'AE0 80 -
(300,78) g S
g 70 - C -
=' &
M @
g 60 - @ PERMISSIBLE OPERATING k
o REGION
- i(162,50) 50 -
RESTRICTED %
5 40 S REGION
(
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30 -
20 - g ~
(7h'5) (i21,is) -
10 i (119.5,0) 0 0 50 100 150 200 250 300 Rod index. 5 Witndrawn 0 25 50 75 100 0 25 50 75 100 t f i ! I I I f 1 t
Group 5 Group 7 75 100 2.ol 25 50 0
Group 6
}d2
- c. . . . .
- g. , Y '- l had index is the percentage sum O! the withdrawl of Groups 5.6 and 7 4'j #
UNIT 1 ROU POSITICN LINITS Figure 3.5-2A
- ~
,7 * ROD POSITION LIM 11S FOR 4 FUW U.PtHAIJ'IN AFFLICABLE DURING 'lHi:.
(1623 :02) (270,102) 100 -
.. . (173.8,'90) 90 .-
OPERATION Ill THIS REGION . .- .
I2 "*')' I IS NOT ALLOWED .
80 -
POWER LEVEL y j CUT OFF g
g 70 - w
= .
. . . 3g 60 -
PERMISSIBLE OPERATING E '
REGION
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, 50 -
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20 -
4 -
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(izi,it) - -
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' ' ' '- I ' (ii9. s,0) 0 0
50 -
100 150 200 250 .. '300 Rap Index. 5 Withdrawn 0 25 50 75 100 0 25 50 75 100 t t t i I t t I i t Group 5 Group 7 .
0 25 50 75 100
. . I f I f I Group 6 e .
Rod index is the percentage sura of the withdrawl of Groups 5.8 and 7.
UNIT 1 RCD T SITION LIMITS Fic.,ure 3.5-28
(162,102) 100 g (17a3 :02) 4 90 -
OPERATl0?i 111 THIS REGIO!I 3- IS liOT ALLOWE0 0 80 -
5 . D -
8 o 70 - v 8 .
&., 3 . .
- a. .
. o 60 -
cc PERMISSIBLE:0PERATitiG .
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' REGION E (1s2,50) -
3 50 -
g[ . . .,
-co 3
$ 40 .
. g RESTR MTED s - .
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s 30 -
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(si9.50) 0 ' ' ' ' ' I ' ' ' ' '
0 50 100 150 200 250 . .300 -
Rod Index. '4 Withdrawn .
0 25 50 75 100 0 25 50 75 100 t ' ' ' J ' ' ' ' t Group 5 . Group 7 . .
- g. .
0 25 50 75 100 '20# -
Group E g
Rod index is the percentage sum of the withdrawl of Groups 5.G and 7.
- r
, UNIT 1 C
. '*? -
RCD POSITION LIMITS Figure 3.5-2C
em
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P 0 YiE R L E W L 1025
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49 - - .
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CORE IMSALANCE, % .
OPERATIONAL P0Y!ER
, IMSALANCE ENVELOPE
' Figure 3.5-2D .
r 1492 204
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20
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Axial 1.acation of Peak Power From Bottam of ' tare. it .
s LOCA LIMITED MAXIMUM ALLOWABLE LINEAR HEAT RATE Figure 3.5-2E 1492 205 .
Attachment 1 partial Loon LOCA Analysis This study shows that in the event of a LOCA during partial loop operation, peak cladding temperature and Metal Water reactions are significantly lover than during four pump operation. The partial loop analysis was performed assuming the vorst case break (8.55ft DE, 2 C =1) reported in BAW-10103 at the maximus kW/ft D
limits allowed under Technical Specifications. The maximum cladding temperature for the partial loop LOCA analysis, is 1766F, which is 313 F less than the sa=e break at fu11 power and flow conditions.
There are 5 possible break configurations at the pump discharge for partial loop operation:
- 1. 3 - pump operation
- a. break in down loop (loop vith idle pump), down cold leg (cold leg with idle pump)
- b. break in down loop, up cold leg
- c. break in up loop, up cold leg
- 2. 2 - pu=p operation, one pump up in each loop 1492 206
- a. break in down cold leg
- b. break in up cold leg Analysis of the 3-pump operation instead of 2-pump operation was chosen for the following reasons. First, 3-pump operation is the more probable partial loop operational code. Second, the rated power level for 3-pump operating is 775 of full power rating compared to 51% of full power rating for 2-pumps operating. The ref1 coding rate vill be lower for higher core power, thus a greater cladding temperature rise after the End of Blovdown (E03) is expected for 3 pu=ps operating.
Due to the nature of core flow which results during a cold leg break, two break locations for 3-pu=p operation vere examined. Typically, eore flow retains highly positive during the initial phase of the blowdown transient. As the head of the RC pumps degrades , due to 2-phase effects, of core flow, the magnitude of the positive core flow diminishes. Core flov then becc=es negative for the re=ainder of the blevdown transient. The two phases of core flov, positive and negative are effected by the choice of break location. placement of the break at the pu=p discharge of the
, o I
idle pump (dovn leg) vill induce a greater driving force from the intact cold legs. This will yield high positive flows and low negative flows.
A break at the pump discharge of the down loop, up cold . Leg will cause a loss in positive flow during the first half of the {
transient. Analyzing both break locations will ensure that the {
most conservative assu=ptions. effecting core flow during the blow-down transient have been considered.
The parameters used in the partial loop CRAFT and THETA models are consistent with the spectrum analysis reported in section 6 of ,
BAW-10103, except for the following:
- 1. The total plant power for both cases analyzed is reduced to 77% of rated power for 3-pump operation. . The peak linear heat rate for the hot bundle is the maximu:a kW/f t LOCA limit allowed under Tech Spec at the 6 f t elevation for this mode .
of operation.
- 2. Since there is a power imbalance between the loop with 2 RC pumps and the loop with 1 RC pump, the loud ratio between the steam generators is changed to 2.27:1 by control in the feedwater flow to each steam generator. I
- 3. The flow and pressure distribution was modeled to ref1cet I
the imbalance caused by the idle pump and the reduction l in the RC flow to .75% of normal 4 pu=p operation. At steady l state conditions the idle pu=p is locked in position because l
flow is reversed in that cold leg.
- The flow proceeds from the idle pu=p to the lower plenum of the steam generator where it mixes and proceeds back to the reactor vessel through the RC pump in the down loop, up cold leg. About 147. of the RC flow, f rom the downcomer plenum is directed back .in the cold leg. If the flow reverses to the positive direction during l the transient the idic pu=p would act as a free spinning rotor with no power.
Tabic 1 sumaries the results of the partial loop analysis and compares those results to the worst break reported in BAW-10103. Figures l' and 2 show respectively the hotspot and rupture node cladding temperature and the core flow for 3 pu=ps running with the break located at the up leg of the down loop. The maximum cladding temperature is 1766F at 98,5 seconds. Figures 3and 4 show respectively the hot spot and ruptured 1492 207
l node cladding temperature and the core flow for 3 pump running with i the break located in the down Icg of the down loop. The maximum cladding tempt.rature is 1751 F at 91 seconds. Examination of the core flow for both cases reveals a distinct difference in the flow transient. With the break at the idle pump, core flow is similar to the 4-pump operation shown ..n figure 6-2 of BAW-10103. When ,
I the break is placed at the pump discharge of the up leg-down loop, l the positive phase of the core flow is sharply reduced and the !
l transition from positive to negative flow occurs earlier, approximately 11 seconds compared to approwmately 14 seconds for the- 4-pump case.
The negative flow is increased due to the decrease from 3 to 2 active f pu=ps trying to force the flew into the vessel. The flooding rates j calcalated using the REFLOOD code are slightly higher than those predicted for the 4 pump operation case because of the lower average core power. The hot pin cladding temperature response calculated with the THETA code are shown in Figures 1 and 3 for the two cases mnMned.
Rupture for both cases occurs just af ter the E0B. The ruptured node cladding temperature decreases rapidly af ter rupture because of the reduced gap heat transfer from the fuel to the cladding and the increase in the surface area for cooling. The reflooding heat transfer coef-ficients are high enough to prevent a rise in the ruptured cladding te=perature after rupture. The containment building pressure calcu- ;
I lated by the CONTEMPT code is similar to the worst case shown in Figure 6-10 of BAW-10103.
The low tecperatures experienced for the partial loop cases analysed are considerably lower than those for h-pu=p operation reported in BAW-10103 The maximum cladding temperature for the partial loop LOCA analysis is only 1766F compared to 2079F for the worst h pump operation break as reported in Section 6 of BA'.7-10103 The proposed Technical Specifications for partial pump operaticn are calculated in a manner consistent with these resultc , ejected rod worth, and minicus shutdown margin.
1492 208
_ TABLE 1 Comparison of 8.55-ft DE break at pump discharge, DC = 1.0, with 4 and 3 pumpa operating.
3-p d ps, break 3-pu=ps, break in . in down loop, 4-puros (BAW-10103) down loop. up lee down lee
~
Case Nd ber FC ll2(IL) PP102(Yl) PP101(lB)
~
Per Cent Power 102 77 77 -
(100% Power =2772) -
Peak Cladding Temp 2079/61.5 1766/98.5' 1751/91.0 t.nrupt/ time, F/s Peak Cladding Temp 1916/43.5 1674.4/11.5 ' 1569/42.0 rupt/tice, F/s Cont Pressure at Peak Cladding 36.4 35.37 35.48 Temp, psia Rupture Time / blockage 13.8/63.14 25.39/65.04 25.9/64.78 S/%
CFT actuation time,s 16.7 16.6 17.2 End of bypass, s 24.4 24.8 25.2 End of Blowdown,c 24.4 24.8 25.2 End of adiabatic heatup,s 35.4 35.8 36.4 Water mass in reactor at end of 1532.0 1824 1623.0.
blowdown, Ibn -
Loesi metal-water reaction,% 4.2923 2.86' 2.738 Full-power seconds at end of blowdown 1.949 1.874 1.959 1492 209 7 .
2000 .
FIGURE l HOT SPOT AND RUPTRUED NODE CLADDING TEMPERATURE VS TIME FOR 8.55 FT2 DE BREAK AT PL'EP DISCHARGE OF DOWN LOOP UP CO' .D LEG,
,.C D =1.0,18KW/FT AT 6 FT. ELEVATION.
1766 F0 1800 -
. .~ %
- 1674 F -
' N 1600 -
e, s
L f{ -
i
/ \
\
iy
'd \
(s s%~_-',-_
1400 -1 i m
y g
~~s~~~, s
- 2 '
~
- N f ,
cn 1200 M I .
a l O I '
d 1 1000 .
A UNRUPTURED NGDE N . .._.
RUPTURED NODE ,
N 800 '-
O .
' I ' ' I ' '
600 O 20 40 , 60 80 100 120 140 160
-- n ,- c
300 .
CORE FLOW VS TIME FOR 8.55 FT 2 DE FIGURE 2 BREAK AT PUMP DISCHARGE OF DOWN 250 -
LOOP-UP COLD LEG, CD = 1.0, AT q
- 6 FT ELEVATION 200 -
150 -
.o a .
@ 100 -
I e
o o 50 -
A '
0 -
- = .
< -50 i-N -
l N
~
j l t l' 1 1 I I I I I
-100 14 16 18 20 22 24 26 2 4 6 8 10 12 0
' Time, s
2200 -
FIGURE 3 HOT SPOT AND RUPTURED NODE CLADDING ~
TEMPERATURE VS TIME FOR 8.55 FT 2 DE
~
BREAK AT PUMP DISCHARGE OF D0hN LOOP-DOWN COLO LEG, CD = 1.0, AT 6 FT ELEVATION 1800 -
1751F .
Hot Spot .
- u. 1600 .
1586F
- ,m 0 / \ '
B / \
a / g g1400 - f> \ Ruptured .
-~~~,
'p
,) ' 's u
E 1200 -
s i .
d' 1000 -
1 m
'800 -
a .
N N
~ '
I I I i 1 m 600 I I 0 20 40 60 80 100 120 140 160
,' Time, s 9
s-
~
. 350 .
FIGURE 4 CORE FLO?l VS TIME FOR 8.55 FT 2 DE BREAK AT PLFP 300 _
DISCHARGE OF DOWN LOOP-D0%N COLD LEG, CD = 1.0, AT 6 FT ELEVATION 250 -
200 -
N ~
150 -
i .
3 h
S o
100 -
4 o N N
50 }
- C i
-~ -
0.; ~
L
' I I 1 i i i I i i i I I
-50 16 18 20 22 25 26 2 4 6 8 10 12 14.
0 Timo, s e
wttach=ent 2 A Ccmparison of Key Parameters of the Genecic Evaluntion Model to Individual Plant Parameters Parameter Generic TMI-1 Model Reactor Building Free Volume ft3 2.205 x 10 6 2.126 x 10 6
- a. The reactor building valls including the concrete vall, steel liner, and anchors: .
Exposed area, ft 2 67,410 63,300 Paint thickness, ft 0.00083 .00083 Steel thickness, ft 0.0550h .029h6 Concrete thickness, ft 4.0 3.5 The surface Area is 5% larger than the largest values reported for Category 1 plants.
- b. The reactor buildingdome including c cncrete, steel liner, and anchors:
Exposed area, ft 2 18,375 18,h00 Paint thickness, ft 0.00083 .00083 Steel thickness, ft 0.065h6 .029k6 Concrete, ft 3.0 3.0 The surface area is 5% larger than the largest value reported for Category 1 plants.
- c. Painted internal steel:
Exposed area, ft2 2h9,000 311,599 Paint thickness, ft 0.00083 .00083 Steel thickness, ft 0.03125 .0227
- d. Unpainted internal steel:
Exposed area, ft 2 36,000 94 Steel thickness, ft 0.03125 .006h
- e. Unpainted stainless steel:
Exposed area, ft 2 10,000 h2,151 Steel thickness, ft 0.03125 .0108
- f. Internal concrete:
Exposed area, ft 2 160,000 118,000 Paint thickness, ft 0.00083 .00083 Concrete thickness, ft . 1.0 1.5h L492 214
ATTACIEEIIT 3 CONTAIIi!E!iT - PRESSURE CCMPARTSON: TMI-1 & GEITERIC Time Pressure (usic) Pressure (usic; (sec) TMI-1 Generic (sec) TMI-1 Generic 2.0 13 32 12.29 72 21.55 21.k1 h.7 18.11 17 52 76 21.h2 21.31 6.7 22.19 21.46 78 21.38 21.28 B.9 25.67 24.75 82 21.27 21.21 99 27.00 26.00 88 21.18 21.1h 12 29.24 28.11 92 21.10 21.07 13 30.10 28.59 94 21.07 21.06 15 31.34 29.87 96 21.03 21.03 17 32.03 30.41 98 21.00 20 99 19 32.h6 30.68 105 20.91 20.30 21 32.49 30.59 115 20.82 20.79 23 31.87 29.5h 125 20.69 20.69 25 30.67 28.71 135 20.60 20 52 27 29.h3 27.69 1h5 20.47 20.38 29 28.38 26.61 155 20.36 20.23 31 27.53 25.77 165 20.25 20.09 33 26.Th 25.11 175 20.14 19 24 35 26.09 24.53 185 20.01 19.79 37 25 53 24.08 195 19.89 19.63 39 25.0h 23.69 42 2h. :.2 23.22 h6 23.70 22.T7 48 23 32 22.50 52 22.81 22.10 56 22 38 21.8h 1492 215