ST-HL-AE-3118, Application for Amend to License NPF-76,allowing Use of New Power Distribution Limits Described in Encl FSAR Pages
| ML20247Q706 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 06/01/1989 |
| From: | Rosen S HOUSTON LIGHTING & POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| ST-HL-AE-3118, NUDOCS 8906070017 | |
| Download: ML20247Q706 (110) | |
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The Light c c omp a ny wt h Texas hoject Dectdc Generating Station M. Box 289 Wadmonh, Texas m83
! Houston Lighting & Power June 1, 1989 ST-HL-AE-3118 File No.: G9.01, R5 10CFR50 U. S. Nuclear Regulatory Commission
. Attention: Document Control Desk Washington, DC 20555 South Texas Project Electric Generating Station Unit 1 Docket No. STN 50-498 Proposed License Amendment Concerning the Unit 1 Cycle 1 Licensing Basis Reference 1: Letter ST-HL-AE-2683 from J.N. Bailey to USNRC Document Control Desk dated June 15, 1988.
2: Letter ST-HL-AE-3021 from M. A. McBurnett to USNRC Document Control Desk, dated March 30, 1989
[ Errata submitted May 05, 1989]
The purpose of this letter is to apply for amendment of the South Texas Project Unit i license to allow use of new power distribution limits as described in the attached FSAR changes. Approval of this application is required to support the STPEGS Unit 1 Cycle 2 and future core safety analyses.
The proposed amendment will also be applicable to the present Unit 1 core.
Initially, the core design and associated safety analysis for both Units 1 and 2 were the same. To support a longer operating cycle, the Unit 2 core uas redesigned. The safety analysis for the revised Unit 2 fuel design identified power distributions more limiting than the power distributions assumed for the Unit 1 safety analysis. These new power distributions resulted in changes to several key parameters of some Chapter 15 accidents.
The safety analysis also considered additional conservatism that provide margin for future fuel reloads. The results of this safety analysis were submitted by Reference 1 and approved by the NRC in SER Supplement 6. For schedular reasons, the revised safety analysis was not applied to the Unit 1 licensing basis at that time; and as a result, the licensing bases for Units 1
.and 2 are currently different for several Chapter 15 accidents.
This submittal is to apply the revised Unit 2 safety analysis assumptions to Unit 1. The design and core characteristics for the redesigned STP Unit 2 are similar to the corresponding values of Unit 1. We anticipate that the Unit I reanalysis will demonstrate that the Unit 1 Cycle 1 safety analysis assumptions are bounded by the Unit 2 Cycle 1 safety analysis assumptions.
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'8906070017 890601
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ADOCK 05000498 N - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
. ST-HL-AE-3118 Houston Lighting & Power Company . File No.: G9.01, RS 1 South Texas Proj,ect Electric Generating Station Page 2 HL&P does not believe that incorporating the Unit 2 analysis into the Unit 1 licensing basis will affect the staff's conclusions stated in the STP SER (NUREG 0781). However, it meets the criterion of 10CFR50.59 for being a
" change" where the consequences of an accident have increased, although I slightly, and is therefore considered an unreviewed safety question.
The annotated FSAR pages, description of the changes, and safety l evaluation are attached. HL&P will incorporate these changes into the STP Updated FSAR subsequent to their approval by NRC staff.
.HL&P has reviewed the proposed changes in accordance with the requirements of 10CER50.59 and 10CFR50.92. The results indicate that an unreviewed safety question (USQ) is involved. HL&P has reviewed the USQ and concluded there are no resulting significant hazard considerations.
The South Texas Project Nuclear Safety Review Board has reviewad and approved the attached proposed revision and concurs with the 10CFR50.59 determination.
I In accordance with 10CFR50.91(b) HL&P is providing the State of Texas i with a copy of this proposed amendment. j i
If you should have any questions on this matter, please contact Mr. l A. W. Harrison at (512) 972-7298. l M I S. L. Rosen l Vice President, Nuclear I Engineering and Construction l SDP/hg Attachments: 1. Description of Change l
- 2. Safety Evaluation
- 3. Annotated revisions to FSAR l
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ST-HL-AE-3118' N' - Houston Lighting & Power Company File No.: G9.1, RS
- g. ' South Texas Project Electric Generating Station cc: i
.1 Regional Administrator Region IV Rufus S., Scott
]
Nuclear Regulatory Commission. Associate General Counsel' ]
611 Ryan Plaza' Drive, Suite 1000- Houston Lighting &' Power Company '
Arlington, TX 76011 P. O. Box 1700 i
. Houston, TX 77001 !
George Dick, Project Manager l U.S. Nuclear Regulatory Commission INPO.
.]
Washington,1DC 20555 Records Center ,
1 1100 circle 75 Parkway i Jack E. Bess Atlanta, GA 30339-3064 ]
Senior Resident Inspector / Unit 1 .
j c/o U.S. Nuclear Regulatory Dr. Joseph M. Hendrie I
' Commission .
50 Be11 port Lane P. O. Box 910- Be11 port,.NY 11713 Bay City, TX .77414 -l I
J I. Tapia l Senior Resident Inspector / Unit 2 c/o U.S.3 Nuclear Regulatory.
.. Co.amission !
P. O. Box 910 '
Bay City, TX L77414-J. R. Newman, Esquire Newman & Holtzinger, P.C.
1615 L Street, N.W.
Washington, DC 20036 R. L. Range /R. P..Verret Central Power & Light Company P. O. Box 2121 Corpus Christi, TX 78403
'R. John Miner (2 copies)
Chief Operating Officer City of Austin Electric Utility
.. 721 Barton Springs Road !
Austin,.TX 78704 R. J. Costello/M. T. Hardt City Public Service Board i P. O. Box 1771 l San Antonio, TX 78296- i i
L 1
Revised 12/21/88 ,
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Ifouston Lighting & Power Company ST-HL-AE-3118 File No.: G09.01, R5 South Texu Project Electric Generating Station UNITED STATES OF AMERICA NUCLEAR REGUIATORY COMMISSION In the Matter )
)
Houston Lighting & Power ) Docket No. 50-498 Company, et al., )
)
South Texas Project )
-Unit 1 )
AFFIDAVIT S.L. Rosen, being duly sworn, hereby deposes and says that he is Vice President, Nuclear Engineering and Construction of Houston Lighting &
Power Company; that he is duly authorized to sign and file with the Nuclear Regulatory Commission the attached proposed revision to the Final Safety .
Analysis Report to incorporate the Unit 2 Cycle 1 safety analysis into the Unit 1 Licensing Basis; is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge and belief.
S. L. Rosen Vice President, Nuclear Engineering and Construction STATE OF TEXAS )
)
)
Subscribed and sworn to before me, a Notary Public in and for the State of Texas this /Y day of e , 1989.
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Attachmsnt 1 ST-HL-AE-3118 Page 1 of 3 Description of Change The following South Texas licensing basis transients were reanalyzed for the Unit 2, Cycle 1 loading pattern; are applicable for the present Unit 1 core; and form the bases for the Unit 1, Cycle 2 reload evaluation and that of future cycles.
Hot Zero Power (HZP) steamline break event (both the credible break in FSAR 15.1.4 and the hypothetical break in FSAR 15.1.5)
Loss of forced reactor coolant flow events (partial (1/4 loops) loss of flow in FSAR 15.3.1, and complete (4/4 loops) loss of flow in FSAR 15.3.2)
Locked rotor events (for the effects on pressure and peak clad temperature with and without coincident LOOP, and for the effect on the percentage of rods in DNB in FSAR 15.3.3)
Spectrum of rod withdrawal at power events in FSAR 15.4.2 Dropped rod and statically misaligned rod events in FSAR 15.4.3 Spectrum of rod ejection events in FSAR 15.4.8 No other licensing basis transients have been reanalyzed. For those events not reanalyzed, the analysis casumptions, results and conclusions presently in the South Texas FSAR remain applicable to both Units 1 and 2.
The HZP steamline break events (the credible break in FSAR 15.1.4 and the hypothetical break in FSAR 15.1.5) were reanalyzed with revised stuck rod reactivity coefficients (moderator density and boron coefficients) to confirm that the DNB design criteria are met. The coefficients used in the reanalysis reflect the discussions in the FSAR regarding the end-of-life rodded negative moderator coefficient with the most reactive RCCA stuck in the fully withdrawn j position remain valid. The changes in these assumptions used in the analysis are reflected in the revised Keff vs temperature curve in Figure 15.1-11.
The revised analysis includes the flow rates associated with a reduction in the safety injection (SI) capability as identified in 1987. The revised analysis also assumes the current value for the thermal design flowrate 1 (95,400 gpm/ loop).
The minimum plant shutdown margin value (1.75%) is unchanged and still used in the revised analysis. The results of the analysis show that the DNB design limit of 1.45 using the W-3 correlation is met.
The loss of forced reactor coolant flow events (partial (1/4 loops), and complete (4/4 loops)) were reanalyzed with the more limiting trip reactivity shape shown in revised FSAR Figures 15.0-4 and 15.0-5. The total minimum negative reactivity insertion (4.0%) as indicated in FSAR Section 15.0.5 is unchanged. In addition, the RCCA insertion time to dashpot L4\NRC\fi
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c Attachmznt 1-ST-HL-AE-3118 Page 2 of 3 entry'(safety analysis value for the rod drop time) as provided in FSAR Section 15.0.5 is assumed in the revised analysis.
The revised analysis also includes the effect of Veritrak pressure transmitter uncertainties on the initial pressure uncertainty and the effect of the
-increeted coolant average temperature uncertainty due to the Temperature Averaging System. It also assumes the current value for the thermal design flowrate (95,400 gpm/ loop). The DNB safety analysis limit is met using the WRB-1 correlation.
The locked rotor events (with and without coincident LOOP for the effects on pressure and peak clad temperature) were reanalyzed with the more limiting trip reactivity shape shown in revised FSAR Figures 15.0-4 and 15.0-5. For each event the hot spot peaking factor (F ) is assumed to be 2.65 times the average rod power. The results of the ana sis show that the limits on RCS pressure (2750 psia), and clad temperature at the hot spot (2700 F) are not exceeded.
The locked rotor event (for the effect on the percentage of rods in DNB) was reanalyzed with the more limiting trip reactivity shape shown in revised FSAR
. Figures 15.0-4 and 15.0-5.
The DNBR evaluation for this event assumes the Technical Specification moderator temperature coefficient (MTC) value of 0 pcm/ F.
For all locked rotor events, the revised analysis also includes the effect of Veritrak pressure transmitter uncertainties on the initial pressure uncertainty and the effect of the increased coolant average temperature uncertainty due to the Temperature Averaging System, i
The spectrum of rod withdrawal at power events were reanalyzed with the more limiting trip reactivity shape shown in revised FSAR Figures 15.0-4 and j 15.0-5. The revised analysis also includes the effect of the increased coolant average temperature uncertainty due to the Temperature Averaging System. The DNB safety analysis limit is met using the WRB-1 correlation.
The dropped rod and statically misaligned rod events are analyzed on a
- cycle-specific basis to confirm that the DNB safety analysis limits and
( dropped rod limit lines (DRLLs) are met using the current DNB correlation.
The cycle-specific analysis includes the effect of Veritrak pressure transmitter uncertainties on the initial pressure uncertainty and the effect I
of the increased coolant average temperature uncertainty due to the Temperature Averaging System. The DNB safety analysis limit is met using the WRB-1 correlation.
The spectrum of rod election events were reanalyzed with the (1) more limiting trip reactivity shape shown in revised FSAR Figures 15.0-4 and 15.0-5, (2) an increased transient hot channel factor for the hot zero power end-of-life case, (3) with an initial hot spot peaking factor (F g) of 2.65 times the l~ average rod power for the hot full power beginning-or-life and end-cf-life l cases, and (4) more conservative values for the reactivity weighting feedback.
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-ST-HL-AE-3118 Page 3 of 3 l The total minimum negative reactivity insertion for the HFP cases (4.0%) as indicated in FSAR Section 15.0.5 is used in the revised analysis. The HZP cases continue to assume a minimum negative reactivity insertion of 2.0%. In addition, the moderator and doppler coefficients and delayed neutron fraction values used have not changed. The revised analysis also includes the effect of Veritrak pressure transmitter uncertainties on the initial pressure uncertainty and the effect of the increased coolant average temperature uncertainty due to the Temperature Averaging System. It also assumes the current value for the thermal design flowrate (95,400 gpm/ loop). The results of the analysis show that the limits on peak clad temperature (2700 F),
maximum fuel stored energy (200 cal /gm), and percent fuel melting (10%) are not exceeded.
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d, Attachment 2 1 ST-HL-AE-3118 Page 1 of 4 SAFETY EVALUATION FOR-SIGNIFICANT HAZARDS CONSIDERATION Does the proposed change:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated: ,
. I' The FSAR changes are a description of changes to previously analyzed accidents. Therefore, the change does not increase the probability of an accident. However, the change describes an increase in consequences. .
The attached Table 1, Summary of Accidents & Key Parameters Affected by Proposed FSAR Changes, provides additional detail. l l
- 2. Create the possibility of a new or different kind of accident from any accident previously evaluated:
The FSAR changes describe changes to previously analyzed accidents.
Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated. l i
- 3. Involve a significant reduction in the margin to safety:
The proposed safety analysis contains assumptions more limiting than the current safety analysis. These new assumptions resulted in changes to several key parameters of some Chapter-15 accidents. Table 1 describes a decrease in margin of safety for the Locked Rotor and Rod Ejection accidents. The result of this safety analysis was reviewed by HL&P as part of the Unit 2 submittal. That review concluded that the analysis-does not involve a significant reduction in the margin to safety. The result was also reviewed and accepted by the NRC in SER Supplement 6.
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NOTES TO TABLE 1 i
Note 1. Table 15.3-?a and Figure 15.3-18 show that the maximum reactor ,
coolant pressure increases from 2589 psia to 2615 psia for the with offsite power case. Table 15.3-2b and Figure 15.3-18 show that the ;
msximum reactor coolant pressure increases from 2589 psia to 2616 yrla for the without offsite power case. This is an increase over the previous results. Therefore, this change results in a decrease in margin of safety as defined by Technical Specification Bases Section 2.1.2. However, this pressure is below the 2735 psig design limit identified in Technical Specification Bases. Therefore, this change does not represent a significant increase in hazards.
Note 2. Table 15.3-2a and Figure 15.3-20 show that the maximum clad temperature increases from 1675 F to 1713 F for the with offsite power case. Table 15.3-2b and Figure 15.3-20 show that the maximum clad temperature increases from 1680 F to 1717 F for the without offsite power case. This is an increase over the previous results.
Therefore, this change is an increase in consequences of an accident previously analyzed in the FSAR. Since the temperature is significantly below the design limit of 2700 F, it loes not represent a significant increase in hazards.
Note 3. The text and Table 15.3-3 show that the number of rods in DNB ine: ceases from 7% to 100. This results in the increase in dose shcwn on Table 15.3-4. This is an increase over the previous results. Therefore, this change is an increase in consequences of an accident previously analyzed in the FSAR. However, this increase is the same as that evaluated by HIAP for Unit 2 and is below the limiting case which considers 15% of all rods in DNB. Therefore, it does not represent a significant increase in hazards. This result was also reviewed by the NRC and accepted in SER Supplement 6.
Note 4. This change is an increase in consequences of an accident previously analyzed in the FSAR. [The peak clad temperature for non-LOCA transients is not discussed in the Technical Specification Bases.)
The temperature is below the design limit of 2700 F and the change does net represent a decrease in the margin of safety as defined in the Technical Specification Bases. Therefore, it does not represent a significant increase in hazards.
Note 5. Table 15.4-3 presents the results of the maximum fuel stored energy calculation for the four cases analyzed. This table shows that the stored energy increases are below the design limit of 200 calories per gram stated in FSAR Section 15.4.8.1.2. The stored energy limit is not discussed in the Technical Specification Bases. Therefore, the change does not represent a decrease in the margin of safety as defined in the Technical Specification Bases and does not represent a significant increase in hazards.
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ATTACHMENT .5 STP FSAR ST-HL-AE-PAGE 7 0F 3138@
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- 3. Isakage flow from the vessel inlet nozzle directly to the vessel outlet nozzle through the gap between the vessel and the barrel.
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X 4 Flow introduced between the baffle and the barrel for the purpose of cooling these components and which is not considered available for core b, cooling.
M 5. Flow in the gaps between the fuel assemblies on the more periphery and the adjacent baffle wall.
N The abeve contributions are av sted to confirm that the design value of the k Y core bypass flow is met.
percent of design value of core bypass flow is agual to 4.5 tal vos.s. . flow.
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' Of the total ow e cent sociated with the intarnals (stems 1, 3, 4 and 5 above) and 2 0 percent for the core. A Calculations have been per-formed using drawing to erances on a worst case basis and accounting for uncertainties in pressure losses.1 Based on these calcu (ens flow is < 4.5 percent.
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}{. 4 pass Flow model test results for' the f16w path throu h the reactor are discussed 1A Subsection 4.4.2.7.2. g%
4.4.4.2.2 Inlet Flow Distributions: Data has been considered from several 1/7 scale hydraulic reactor model tests, Referenr:es 4.4-23, 4.4-24 and 4.4-62, in arriving at the core inlet flow maldistribution criteria used in the THINC analyses (see Subsection 4.4.4.5.1). THINC-I analyses ande, using this data, have incicated that a conservative design basis is to consider 5 percent redue-tion in the flow to the het assembly Reference 4.4-63. The same design basis of 5 percent reduction to the hot assembly inlet is used in THIEC IV analyses.
The experimental error estimated in the inlet velocity distribution has been considered as outlined in Reference 4.4-18 where the sensitivity of changes in inlet velocity distribyttons te hot channel thermal performance is shown to be small. Studiesl4*'*16J ande with the improved THINC model (TEINC-IV) show that it is adequate to use the 5 percent reduction in inlet flov to the het assembly for a loop out of service based on the experimental data in References 4.4-23 and 4.4-24.
The effect of the total flow rate on the inlet velocity distribution ums studied in the experiments of Reference 4.4-23. As was expected, on the basis of the theoretical analysis, no significant variation could be found in inlet velocity distribution with reduced flow rate.
4.4.4.2.3 Empirical Friction Factor Corralstions: Two empirical friction '
.farter correlations are used in the TRINC-IV computer code (described in Sub-est G 4.4.4.5.1).
fuel rod axis, is The friction factor in the axial direction, para 11elgog evalusted using the Novendstern-Sandberg correlation . This correlation consists of the following: -
4.4-23 CHANGE NOTICE M L
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An inlet tempentans of 560.4*r used in the fanoving analyses Fg Qsa lacety,4, flow of 95,400 @ asse have been
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Figare 15A5. 8 Normalized MCCA Sank Rosetivity Worth versus Normehred Drop Time l
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- 2. A negative moderator coefficient corresponding to the end. f life rodded \ 57 core with the most reactive rod cluster control assembly ii the fully withdrawn position. The variation of the coefficient wit erature and pressure is included. The k versus temperature at psi
] correspondingtothenegativemoNatortemperaturecoeffi nt used is shown in Figure 15.1-11. The effect of power generation in the core on overall reactivity with the most reactive RCCA in the fully withdrawn N position is shown on Figure 15.1-14 for aninal reactor coolant flow.
t-g The core properties associated with the sector nearest the affected steam generator and those associated with the remaining sector were conserva.
tively combined to obtain average core properties for reactivity feedback calculations. Further, it was conservatively assumed that the core power distribution was uniform. These two conditions cause underprediction of the reactivity feedback in the high power region naar the stuck rod. To verify the conservatism of this method, the reactivity, as well as the power distribution, was checked for the liatting conditions for the cases analyzed. This core analysis considered the Doppler reactivity from the 8 high fuel towperature near the stuck RCCA, moderator feedback from the high water enthalpy near the stuck RCCA, power redistribution and non-
. uniform core inlet temperature effects. For cases in which steam gener-ation occurs in the high flux regions of the core, the effect of void formation was also included. It was determined that the reactivity employed in the kinetics analysis was always larger than the reactivity calculated, including the above. local effecta for the conditions. These results verify conservatism (i.e., underprediction of negative reactivity feedback from power generation).
- 3. Minimum capability for injection of high concentration beric acid solution corresponding to the most restrictive single failure in the 2 safety injection system. Iow concentration boric acid must be swept fro: ..
the safety injaction lines downstream of the isolation valves prior to
the delivery of high concentratfoz. boric acid (2,500 ppsi) to the reactor coolant loops. This effect has been allowed for in the analysis.
- 4. . The case studied is a steam flow of 292 pounds per second at 1300 psia 2 from one steam generator with offsite power available. This is the maxi.
mua capacity of any single steam dump, relief, or safety valve. Initial hot shutdown conditions at time scro are assumed since this represents the most conservative initial condition.
Should the reactor be just critical or operating at power at the time of a steam release, the reactor will be tripped by the normal overpower protection when power level reaches a trip point. Following a trip at power. the RCS.contains more stored energy than at no load, and there is appreciable energy stored in the fuel. Thus, the additional stored ener-l gy is removed via the cooldown caused by the steam release before the i no-load conditions of RCS temperature and shutdown margin assumed in the l enalyses are reached. After the additional stored energy has been removed, the cooldown and reactivity insertions proceed in the same man-mer as in the analysis which assumes no load condition at time sero.
However, since the initial steam generator water inventory is greatest at no load, the magnitude and duration of the RCS cooldown are less for steam release, occurring at power. 43 CHANGE NOTICE M 7 15.1-10 Amendment 57
ATTACHMENT 3 ST-HL-AE- 37io on PAGE U OF T1 s P TSAR --
- c. High nogetive stea=line pressure rate signal (tvo ou: of three in any loop) below the F 11 setpoint. 57 Tas: se:ing isolation valves are provided in each steam line that will fully close within 10 seconds of a large break in the steam line. For breaks down.
stream of the isolation valves, closure of all valves would completely terzi.
cate the blevdown. , A description of steam line isolation is included in 33 Chapter 10.
.'Desi5 n criteria and methods of protection of safe:y-related equipment from the dyna =ic effects of postulated pipin5 ruptures are provided in Section 3.6.
A block diagram summarizing various protection sequences for safety actions required to mitiga:e the consequences of 'is .
^ - prov de vent is in T gure 43 15.1 9. $a@s;:> / TA*aw/ R /ro-f -- =/
& C W-e-15.1.5.2 Analvsis of !.fects ar.d cense uenees. f' .
.ve :hed of Analvsis ne analysis of the stea= pi;e rupture has been perfor:ed to deter =ine:
- 1. B e cere hea flux and F. s temperature and pressure resulting f;cm the cooldevn following the stes: line break. Be IM:';.AN code (Ref. 15.1.') .
has been used.
- 2. ne ther:a1 and hydrau".ic behavior of the core fc11oving a stea: line break. A detailed ther:41 and hydraulic digital cc=puter code, '"M!SC.
has been uses to deter:ine if ON3 occurs fe the core condi:icts cc ;uted in 1:em 1 abeve.
B e etthedclegv e=pleved is censis: n: with that used in the stea: line r.::- ,-
A hA&
- re elee. : pics: rep :: ("R ' 13.1 4)
' w e %d M tw *s. 6 ,/c/ tr.ce, M sEassLp:ofe ex[s:4 e,v a:n %...e ti=e of a main stea: s .es M-.%
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. break accident:
, 1. End of life shutdown cargin a: no load, et,uilibriu= xenon ccrditicus, and with the mes: reactive ICIA stuck in 1:s fully withdravn posi:icn. 0;ers-tien of the control rod banks during core burnup is restricted in such a way that addition of positive reac:ivity in a stea= line break acciden: .
vill no: lead to a more adverse condition than cl.e case ar.aly:ed.
. A negative mode.rator coefficient cc espending to the end.of life reddec core with the ses: reactive ROCA in the fu*1y withdrawn positten. ne variation cf the coefficien with te=peratu- nd pressure has been included. De k , 1000 psi,corres;ending to the negative sedera:I!;f versus te=perature temperature coefficien: a: ed is shown on Tigure 15.1 11. ne'effect cf power generation 1. the coreNon overall reac-tivi:y with the mes: reactive ROCA in the ully viihdrav6 pcsiticn is 2 3 show: on Figure 15.1-14 for ne:inal reac er coolant flev. ,
l I k
CHANGE NOTICE / 7- N kMi f0 p &c U Q L p& ' '
11.1 13 -
A=endmen: 5'
\ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ -
ATTACHMENTS ST-HL-A - 311o PAGE / OF b STP FSAR steam release from only one steam generator. Should the core be critical at i 1
near zero power low steam when the line pressure willrupture trip the occurs, reactor.the initiation of safety Steam release frominjection more thanby l 2 one steam generator will be prevented by automatic closure of the fast-acting isolation valves in the steam lines via the low steam line pressure signal. 57 Even with the failure of one valve, release is limited to no more than 10 seconds from the other steam generators while the one generator blows down. 43 The steam line stop valves are designed to be fully closed in less than 5 seconds from receipt of a closure signal.
As shown in Figure 15.1-17, the core attains criticality with the RCCAs inserted (with the design shutdown assuming one stuck RCCA) shortly before !18 boron solution at 2,500 ppm enters the RCS. A peak core power less than the [2' nominal full power value is attained.
The calculation assumes the boric acid is mixed with and diluted by the water flowing in the RCS prior to entering the reactor core. The concentration after mixing depends upon the relative flow rates in the RCS and in the SIS.
The variation of mass flow rate in the RCS due to water density changes is included in the calculation as is the variation of flow rate in the SIS due to changes in the reactor coolant system pressure. The SIS flow calculation 2 includes the line losses in the system as well as the pump head curva.
Figures 15.1-18 through 15.1-20 show the response of the salient parameters for case b which corresponds to the case discussed above with additional loss of offsite power at the time the safety injection signal is generated. The 18 safety injection delay time includes 10 seconds to start the standby Diesel generator and in 12 seconds the pump is assumed to be at full speed. l43 ,
Criticality is achieved later and the core power increase is slower than in the similar case with offsite power available. The ability of the emptying i steam generator to extract heat from the RCS is reduced by the decreased flow 12 in the RCS. [For thq D Re lu tiort @ powe o powe hap anal is con 1 - I ten th tli uid itiorja as urseiP. The power ns wel lo 118 the qinal fy powe alue.f l It should be noted that following a steam line break only one steam generator blows down completely. Thus, the remaining steam generators are still avail-able for dissipation '.,f decey 1. eat after the initial transient is over. In the case of loss of offsite p wer, this heat is removed to the atmosphere vic the steam line safety valves.
l Margin to Critical Heat Flux l
A DNB analysis was performed for both of these cases. It was found that the DNB design basis as stated in Sm. tion 4.4 was met for all enses.
15.1.5.3 Radiological Consequences. The postulated accidents involving l 43 release of steam from the secondary system do no result in a release of radio-sactivity unless there is leakage from the RCS to the secondary system in the steam generators (SG's). A conservative analysis of the potential offsite doses resulting from a steamline break outside Containment upstream of the main steam isolation valve (MSIV) is presented using the Technical Specifica-tion limit secondary coolant concentrations. Parameters used in the analysis 43 are listed in Table 15.1-2.
15.1-16 Amendment 57
ST-H L- -3ll no PAGE OF 7 I m r. ao TAB!.E 15.1 1 (Centirmaed)
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ATTACHMENT l ST-HL-AjjL-d" ? I 6go PAGE 4R ' 0F I I STP TSAR 15.3.1.2 Analysis of Effects and Consequences.
Method of Analysis 54 The case analyzed is the loss of one pump with four loops in operation.
This transient is analyzed by three di ital 5 computer codes. First the LOTTRAN code (Reference 15.3 1) is used to calculate the loop and core flow during the transient, the time of reactor trip based on the calculated flows, the nuclear power transient, and the primary system prsssure and temperature transients.
The FACTRAN code (Reference 15.3-2) is then used to calculate the heet flux transient based on the nuclear power and flow from 1477RAN. Finally, the THINC code (Section 4.4) is used to calculate the DNBR during the transient based on the heat flux from TACTRAN and flow from 14FTRAN. The departure from nucleate boiling ratio (DNBR) trahsients presented represent the minimum of the typical or rhimble cell.
Initial conditions %
Plant characteristics and initial condit s are diseu ed in Section 15.0.3.
l Initial operating conditions assumed to this event a e the most adverse with l respect to the margin to DNB; i.e., ma imum steady a te power level, minimum steady state pressure, and maximum st 'ady state co ant average temperature.
The pressure uncertainty used in th analysis is psi and the coolant 54 average temperature uncertainty is v, Reactivity Coefficients +* ,
The most negative Doppler only power coe ficient is used (see Figure 15.0 2).
This is the equivalent of a total integr ted Doppler reactivity rom 0 to 100 percent of 0.016 percent Ak.
The "least negative moderator temperature coe ficient (see Figure 15.0 6) is assumed since this results in the maximum core power during the initial part of 2.he transient when the minimum DNBR is reached.
_Tle v Coastdown The flow coastdown analysis is based on a momentum balance around each reactor coolant loop and across the reactor core. This momentum balance is combined with the continuity equation, a pump somentum balance and the pump character-istics and is based on high estimates of system pressure losses.
l Plant systems and equipment which are available to mitigate the effects of the l accident are discussed in Section 15.0.8 and listed in Table 15.0 6. No sin-I gle active failure in any of these systems or equipment will adversely affect the consequences of the accident. ,
CHANGE NOTICE N 15.3 2 Amendment 54
ST-HL-AE- 3 l 3
nn d
(t PAGE 34 0F 7 'T l, I-h undervoltaae or underfrequenc h 's M [Me N a'riat on bytweKn tlyis ana ys 4 and that
- 2. seco evi us
.Thi et n is tha the saco . rvativ CA nse t/onti o dashp entry is
. ' time er the r- ced flew
- h. c tio at ex harhen ene C are ins rte for s trans ent.
f Results h=Aw y Figures 15.3 9 through 15.3 12 show the transient response for the less of g power to all reactor coolant pumps with four loops in operation. The reactor is again assumed to be tripped on undervoltage signal. Figure 15.3 12 shows f .f l the DNBR to be always greater than 1.30. l I
/ Since DNB does not occur, the ability of the primary coolant to remove heat g ,
from the fuel rod is not greatly reduced. Thus, the average fuel and clad g
temperatures do not increase significantly above their respective initial o ,
values. .
d f&
The calculated sequence of events is shown in Table 15.31. The reactor coolant pumps will continue to coastdown, and natural circulation flow will eventually be established, as demonstrated in section 15.2.6. With the reactor tripped, a stable plant condition will be attained. Normal plant
' shutdown may then proceed.
15.3.2.3 Radiological Consequences. A complete loss of reactor coolant flow from full load results in a reactor and turbine trip. Assuming, in addi-tion, that the condenser is not available, atmospheric steam dump would be required. The quantity of steam released would be the same as for a loss of offsite power.
There are only minimal radiological consequences associated with this event.
Therefore, this event is not limiting. Since fuel damage is not postulated.
the radiological consequences resulting from atmospheric steam dump are less severe than the steam line break, discussed in Section 15.1.5.
15.3.2.4 Conclusions. The analysis performed has demonstrated that for the complete loss of forced reactor coolant flow, the DNBR does not decrease below 1.30 at any time during the transient. Thus, the DN3 design basis as {
described in Section 4.4 is set.
15.3.3 Reactor Coolant Pump Shaft Seizure (Locked Rotor) 15.3.3.1 Identification of Causes and Accident Description. The acci-dent postulated is an instantaneous seizure of a reactor coolant pump rotor such as is discussed in Section 5.4 Flev through the affected reactor cool-ant loop is rapidly reduced, leading to an initiation of a reactor trip on a low reactor coolant flow signal. l Following initiation of the reactor trip, heat stored in the fuel rods contin-ves to be transferred to the coolant causing the coolant to expand. At the same time, heat transfer to the shell side of the steam generators is reduced, first because the reduced flow results in a decreased tube side film coeffi-cient and then because the reactor coolant in the tubes cools down while the shell side temperature increases (turbine steam flow is reduced to zero upon turbine trip). The rapid expansion of the coolant in the reactor core. com-bined with reduced heat transfer in the steam generators causes an insurge
- CHANGE NOTICE d7' 15.3 5 Amenement 55
ATTACHME STP FSAR GE OF N into the pressurizer and a pressure increase' throughout the RCS. The insurge h into the pressurizer compresses the steam volume, actuates the automatic spray hj system, opeas the power operated relief valves, and opens the pressurizer 8- safety valves, in that sequence. The two power-operated relief valves are o designed for reliable operation and would be expected to function properly
\ during the accident. However, for conservatism, their pressure reducing effect as well as the pressure reducing effect of the spray is not included in the
.EL. d 4anal
.- ysis.
.-e,.
This event is classified as an ANS Condition IV incident (a limiting fault) as defined in Section 15.0.1.
15.3.3.2 Analysis of Kffects and consequences.
Method of Analysis .
Three digital computer codes are used to analyze this transient. The 1.0FTRAN code (Reference 15.31) is used to calculate the resulting loop and core flow transients following the pump seizure, the time of reactor trip based on the loop flow transients, the nuclear power following reactor trip, and to deter-eine the peak pressure. The thermal behavior of the fuel located at the core hot , spot is investigated using the FACTRAN code (Reference 15.3 2), using the core flow and the nuclear power calculated by 14FTRAN. The FACTRAN code includes the use of a film boiling heat transfer coefficient. The FACTRAN code is also used to calculate the heat flux transient based on the nuclear power and flow from IDFTRAN. Finally, the THINC code (Section 4.4) is used to calculate the DNAR distribution in the core during the transient based on the heat flux from FACTRAN and flow from 14FTRAN. The DNBR distribution is
.ised to calculate the number of rods in DNB.
Two cases are analyzed:
- 1. Four looes operating, one locked rotor
- 2. Tour loops operating, one locked rotor, loss of power to the other reacter ecolant pumps g g
At the beginnin5 of the postulat locked rotor ac dent, i.e., at the time the shaft in one of the reactor oolant pumps is ssumed to seize, the plant is assumed to be in operation der the most adv rse steady state operating condition (i.e., naximum stes state power leve' , maximum steady state pres-sure, and maximum steady stat coolant average temperature). Plant charac-teristics and initial condit ons are further dinussed in Section 15.0.3. The pressure uncertainty used i hese analyses is @ psi and *ka ecolant averag temperature uncert F e g kh:.
For the case withe power available, power is lost to the unaffected pumps 2 seconds after reactor trip. (Note: Crid stability analyses show that the grid will remain stable and that offsite power will not be lost because of a unit trip from 100 percent power. The 2 second delay is a conservative * $5 assumptionbasedongridstabilityanalyses/.)
When the peak ressure is evaluated, the initial pressure is conservatively estimated as psi above nominal pressure (2250 psia) to allev for errors in g pi[d .4 15.3 6 g3 Amendment 55 l CHANGE NOTICh MA 7 dN i 1
a
STP FSAR NhWOF the pressurizer pressure measurement and control channels. To obtain the l maximum pressure in the primary side, conservatively high loop pressure drops are added to the calculated pressurizer pressure. The pressure responses shown on. Figure 15.3-18 are the responses at the point in the RCS having the l*
anximum pressure.
haluation of the Pressure Transient
' Afterpumpseizure,theneutronfluxisrapidkyreducedbycontrolrodinser.
tion. Rod motion begins one second after the flow in the affected loop reaches 87 percent of nominal flow. No credit is taken for the pressure -
reducing pffect of the pressurizer power-operated relief valves, pressurizer spray, ataan dump or controlled feedwater flow after reactor trip. Although these are expected to occur and would result in a lower peak pressure, an additional degree of conservatism is provided by ignoring their effects.
The pressurizer safety valves are full open at 2575 psia and their capacity for steam relief is as described in Section 5.4 haluation of DNB in the Core Lurint the Accident s
tf For this accident, DNB is assumed to occur in the core, and therefore, an
- n. evaluation of the consequences with respect to fuel rod thermal transients is o
g performed,--Results obtained from analysis of this " hot spot
- condition repre-sent the uppe with respect to clad temperature and zirconi
. reacti MA L,@
In thea a untion, rod power a e hot spot is assumed to be timee the erste rod never (i.e. F -
O at the initial core enwar novel fgg .t.k w=w9p.e4 v* pfe 9 e. we t,pa *a w '
6+ *n
Film Boilina Coefficient ( -
- D-y_
Cg ce
- t
- e. v tA .
..s %J l y,
' The film boiling coefficient is calculated in the FACTRAN code using the Bishop Sandberg Tong film boiling correlation. m fluid properties are eval-usted at film temperature (average between vall and bulk temperatures). The progran calculates the film coefficient at every time step based upon the 1 sceusi heat transfer conditions at the. time. The neutron flux, system pres- i sure, bulk density and mass flow rate as a function of time are used as ;
program input. i For this analysis, the initial values of the pressure and the bulk density are used throut,hout the transient since they are the most conservative with i respect to clad temperature response. For conservatism, DNB was assumed to j start at the b6 ginning of the accident. I Fuel clad Cao Coefficient l
The magnitude and time dependence of the heat transfer coefficient between fuel and clad (gep coefficient) has pronounced influence on the thermal results. The larger the value of the Sap coefficient, the more heat is trans-ferred between pellet and clad. Based on investigations on the effect of the gap coefficient upon the maximum clad temperature during the transient, the gap coefficient was assumed to increase from a steady state value consistent with initial fuel temperature to 10,000 Stu/hr ft 8 *F at the initiation of the CHANGE NOTICE N' 15 ^**"'**"*"
ST-HL4AE- 3 11 %
PAGE D 0F 'T9 STP FSAR transient.: Dus the large amount of energy stored in the fuel because of the
-)D small initial value is released to the clad at the initiation of the tran-1
% sient.
11rconium Steam Reaction
\(
The airconium steam reaction can become significant above 1800'F (clad temperature). The saker Just parabolic rate equation shown below is used to define the rate of the zirconium steam reaction. ,
d(w') , ,gg ,,,-(45,500) de 1.986T where:
w - amount reacted, ag/cm8 t - time, see T -
temperature,'/ (
The reaction heat is 1510 cal /ga.
The effect of zirconius steam reaction is included in the calculation of the
" hot spot
- clad temperature transient.
Plant systeps.and equipment which are available to mitigate the effects of the accident are disevssed in Section 15.0.8 and listed in Table 15.0 6. No sin-gle active failurm in any of these systems or equipment will adversely affect the consequences of the accident.
Results =
14 eked Rotor with Tour 1. cops Operating W 0D j The transient results for this case are shown n Figures 15.3 17 throur,h I I 15.3 20. The results of these calculations - e also summarized in Table 15.3-2a. Thp peak RCS pressure reached dur n5 the transient is less than that h which would cause stresses to exceed the f ulted condition stress limits.
Also, the peak clad surface temperature i considerably less than 2700'F. I should be noted that the clad temperature was conservatively calculated as-suming that.PNB occurs at the initiatioy f the transient. The number of r in DNB,yas conservatively calculated asQypercent of the total rods in the O core (o g io $jtr w t W h v em e w N w A Zhfar4 ] '
1ecked Rotor with Tour leons Operatina. Imss of Power to the Remaining Purps 1
The transient results for this case are shown on Figures 15.3 17 through 15.3 20. The results of these calculations are also summarized in Table 15.3 2b. The peak RCS pressure reached during the transient is less than that i which would cause stresses to exceed the faulted condition stress limits.
Also, the peak clad surface temperature is considerably less than 2700'F.
Both the peak.RCS pressure and the peak clad surface temperature for this cass 54 are similar to the 4 loop transient with power available as discussed above.
i The total percentage of fuel cladding damaged is the same as the with power case, thus the concluatons of Section 15.3.3.3 are applicable to both events. l d7 -
- # ' ^**"'**"* "
CHANGE NOTICE
- m w ,g e g 1 ST-H L-M- 3 llY g PAGE I 7 0F /7 STP FSAR The 15.3-1. calculated sequence of events for the two cases analyzed is shown in Tame Figure 15.3-17 shows that with offsite power available. the core fiv.- 54 reaches a new equilibrium value by 10 seconds. k'ith the reactor trippec. a stable plant condition will eventually be attained.
15.3.3.3 Radiological Consequences.
release of steam from the secondary system do not result in a release ofThe postulate l43 radioactivity unless there is leakage from the Reactor Coolant System (RCS) to the secondary system in the steam generators (SGs). A conservative analysis of the potential offsite doses resulting from a reactor coolant pump shaft seizure dary accident coolant is presented using the Technical Specification limit.secon-concentrations.
Table 15.3-3. Parameters used in the analysis are listed in l43 The conservative assumptions and parameters used to calculate the activity released loving: and offsite doses for a pump shaft seizure accident are the fol-
- 1. '
Prior to the accident, the primary coolant concentrations are assumed to be equal to the technical specification limit for full power operation 43 fo11owin5 an iodine spike (I-131 equivalent of 60 pCi/g). These concen-trations are presented in Table 15.A-4 Q
- 2. Prior to the accident, the secondary cool spec ie activity is equal to the technical specification limit of .10 pCi/gm dose equivalent 1-131. This dose equivalent spteific activity is presented in 48 Table 15.A-5. g g 3.
~
e percent of'the total core fuel cladd n i amaged, which results I inventory of thetocor in t e release the reactor coolant of se n percent of the total gap i crivity 57
" is asjumed xed in the v;;1512:.19 2 , w ,, s y G : n Nni ormm. 4M 4
e prim'ary-to-secondary leakage of 1 gal / min (Technical Specification limit) is assumed to continue for 8 hrs fo11owin5 the accident. i 5.
Offsite power is lost; MS condensers are not available for steam dump.
6.
Zight hours after the accident, the Residual Heat Removal System (RHP.5) starts operation to cool down the plant. No further steam or activity ir released to the environment.
- 7. The iodir.e partition factor in the S h is equal to 0.01. 1
- 3 The steam releases and meteorological parsmeters are given in Table 15.3-3.
The thyroid, gamma and bets doses for the reactor coo'. ant pump shaft seizure accident are given in Table 15.3 4 for the Exclusion ene Boundary (EIB) of 43 1430 meters and the Low Population Zone (LPZ) of 4800 meters. .
15.3.3.4 Conclusions. Since the peak RCS pressure reached during any of the transients is less than that which would cause stresses to exceed the faulted i,
condition stress limi:s. the 1:: ecrity of the primar. ene. ant s~w
.: ,,e n . : e.4 CHANGE NOTICE M 7 ~
ns s-g +
'h~TTACHMENT 3kg j ST-HL- - ,
str TSAR PAGE e X op QQ TABLE 15.3-1 TIME SEQUENCE OF EVENTS FOR INCIDENTS WICH RESULT
, , IN A DECREASE IN A REACTOR COOLANT SYSTEM TLO'J L Accident Event Time (see)
Partial 141$.5fForced Reactor Coolant Flow %\ /
r Four loops operating.
one pump coasting Coastdown begins now Lau reactor coolant flow trip 4
\1. - '
.% ' i. n
'i down s. Jods begin to drop 2. 8 11 1i. Einimus DNBR occurs 3 Os "3 f o Complete less of Forced y' Reactor Coolant Flev f
gC Four{ le[p All operating purcps \0! C. 5 lose power and begin coasting down s
, Reactor coolant pump 0 o undervoltage trip point reached i
~ k.* -
Rods begin to drop 1 l 5~
- ;- Minieurn DNBR occurs .I 3. f Reactor Coolant Pump Shaft seizure (Locked Rotor) (With offsite power)
Rotor on one pump 0 C locks
- ;- 14w reactor coolant flow setpoint reached 0.07 ok E
CHANGE NOTICE M/7 l
, 15.3 12 Amendment 54
_ ____ ______.___________----2--
ATTACHMENT F STP FSAR OF i TA3LE 15.3 1 (Continued)
TIME SEQUENCE or EVENTS FOR INCIDENTS VHICH RESULT IN A DECREASE IN A REACTOR C001 ANT SYSTEM FLOW
.. A N , ,, Jja* (**e)
Accident - Event Ope' ratio M 2.
~~ w.# .
~~T Rods begin to drop .. Of [s.o7 1 ) ll8 Maximus RCS pres- .
. 32 l56 sure occurs * ~
Maximum clad temper- .7 1b IB ture occurs Id f Reactor Coolant Pump 18 Shaft Seizure (!acked Rotor) (Without 5 t.
offsite power) Rotor.on one pump locks 0 Low reactor coolant 0.07 0' \
flow setpoint
(.
Rods begin to drop 1.07 1.0 RCFs lose power, 3.07 i 4 7 coastdown begins Maximum RCS pressure 3 .
occurs '6 Maximum clad temperature 3,9 occurs b
CHANGE NOTICE Od7 4
15.3 13 Amendment 56
ATTACHMENT T ST HL-hE- 3118 Da PAGE C/O 0F I/
STP TSAR Tall.E 15.3 2a StMMARY OF RESULTS FOR lbCKED ROTOR 71tANSIENTS (With offsite power) q\,. E 3
4 Loops operating \
+ -
Initially - 54 Maximum Keactor Coolant System f 9%'?_ _ _ _ fW,%.)
. Pressute (psia) f 2 9 fa bir
, /
Maximus'CInd Temperacura at 17 i70 Cora Hot' Spot (*T) 1
' 2r H 0 rea_ction.at Core Hot C.A Spot 2(4 by weight) .
I 1
I l
CHANGE NOTICE /887 15.3 14 Amendment 54
TTTTChkENT ST HL-AE- 3[8 PAGE H l 0F STF_FSAR TABLI 15.3 2b SLMMARY OF RESUI.TS FOR IDCKED ROTOR TRANSIENTS (Without offsite Power) g i
4 14eps Operating Initially /
M.+4,
' gw ,
- Maxidan dsastor Coolant 23 1 System Pressure (psia) ~ /a bs4 e (
Maximum. Glad Temperature at Core Hot Spot (*F) 16 [ OnJ ldh n7 2r H O reaction at Core Hot g -
go #
Spot 2(t by iTeight) .1 8 c, , 3 l
l .
l CHANGE NOTICE M/7 1
25.3 15 Amendment 5-
ATTACHMENT
- , ST-H L-M- 3lTl$
PAGE C42 0F sir rSAx l TABLE 15.3 3 .-
PARAMETERS USED IN RC PUMP SRA)'I SEIZURE ACCIDENT ANALYSIS Farameters Core thermal power, MWt 3.800 SG tube leak rate prior to accident 1.0 go and initial 8 hrs following accident CUPS operating prior to accident go Offsite power 14st Fuel defects 4 1.0%
Frimary coolant concentrations Table 15.A-4 Secondary coolant concentrations T + - 15.A 5 Failed fuel (following accident) ! f fuel' rod ore h %g
- Activity released to reactor
//]h 1)j 10Ie al gap-coolant from failed fuel and '
inventor'f of noble available for release gases and iodines lodine partition factor in SC's o,01 during accident Steam release from four 614,000* (0-2 hr' SCs, Ib 1,264,000 (2-8hrL Meteorology 5 percentile Table 15.B 1 Dose model Appendix 15.B
- Condensers assumed unavailable for steam durep.
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DOSES REStd7ING FROM RC PUMP SHAFT SEIZURE ACCIDDG he.A 'le ~YdC (30 Lhhk)
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ATTACHME 3 1
?f"$?fTth %
STP FSAR Dg8f Flant characteristics and initial conditions are disc sed in Section 15.0.3.
In order to obtain
- \ power accident, theconservative to results for an uncont olled rod withdrawal at g assusnetons are mader 5.0 G;s/N#< bd('3X* M M D M d
- 1. Initial condit o maximus' era power anc reactor coolant average temperature ( F uncertain and minimum reactor coolant pressure 57 1 (M r..
- 4)mresultina M ,'; in"MP the minimum initial martin
&*3 W. A<diiMauP to y Dh 4
- 2. Reactivity Coefficients ' Two~ cases are analyzed. 4
- a. Minimus Reactivity Feedback: A least negati . LMM ;.m r temperature
- coefficient of reactivity and a least negative Doppler only power
/]
coefficient of reactivity (See Fig.15.0 2) are assumed correspond.
ing to the beginning of core life.
p, b. Maximus Reactivity Feedback: A conservatively large negative moder-3 ' ator temperature coefficient and a most negative Doppler only power g coefficient are assumed.
N 3. Tha reactor trip on high neutron flux is assumed to be actuated at a d, conservative veNe of 118 percent of nominal full power. The AT trips include all adverse instrumentation and setpoint errors, while the delays for the trip signal actuation are assumed at their maximum values.
4 D e RCCA trip insertion characteristic is based on the assumption that the highest worth assembly is stuck in its fully withdrawn position.
- 5. The maximum positive reactivity insertion rate is greater than that for the simultaneous withdrawal of the combinations of the two control banks having the maximum combined worth at maximum speed.
6.- The effect of RCCA movement on the axial core power distribution is accounted for by causing a decrease in overtemperature AT setpoint pro- 1 portional to a decrease in margin to DNB.
A block diagraa summarizing various protection sequences for safety actions 2'
- equired to mitigets the consequences of this event is provided in Figure Q2 15.0 15. 6 Flant systems and equipment which are available to mitigate the effects of the ace!dont are discussed in Section 15.0.8 and listed in Table 15.0 6. No sin-gle active failure in any of these systems or equipment will adversely affect the consequences of the accident. A discussion of anticipated transients without trip (ATVT) considerations is presented in Reference 15.4-4.
Results .
Figures 15.4-4 through 15.4 6 show the transient response for a rapid RCCA withdrawal incident starting from full power. 2.eactor trip on high neutron flux occurs shortly after the start of the accident. Since this is rapid with respect pressureto the thermal result timetoconstants and margin of the plant, small changes in T,yI and DNS is maintained.
t CHANGE NOTICE M7 15.4-7 Amendment 57
+ ~6 V ATTACHMENT ST-HL- E- 3 }la,g
.g The insertion limits in the Technical Specifications may vary from time .
- ~ to time depending on a number of limiting criteria. It is preferable, N therefore, to analyze the misaligned RCCA case at full power for a posi.
tion of the control bank as deeply inserted as the criterf a on minimum DNBR and power peaking factor will allow. The full power insertion limits on control bank D aust then be chosen to be above that position and will usually be dictated by other criteria. Detailed results will vary from cycle to cycle depending on fuel arrangements.
4 For this RCCA misalignment with bank D inserted to its full po er insert. 53 ion limit and one RCCA fully withdrawn, DNBR does not fall be ow the limit value. This case is analyzed assuming the initial rea tor power, pressure and RCS temperature are at their nominal values (as given in 57 Table 15.0-2) including a +2 reent power uncertaf.nty, a - psia pressure uncertaint and a F temperature uncertaint u with the increased radial p ing facto asso ed with the misa gned RCCA.
DNB calculations have not been perfo specifically for RCCAs missing l 53 from other banks; however, power shape calculations have been performed, as required, for the RCCA ejection analysis. Inspection of the power shapes shows that the DNB and peak kW/ft situation is less severe than the bank D case discussed above assuming insertion limits on the other banks equivalent to a bank D full-in insertion limit.
'Y For RCCA misalignments with one RCCA fully inserted, the DNBR does not I 53 I
fall below the limit value. This case is analyzed assuming the initial reactor power, pressure and RCS temperature are at their nominal values given in Table 15.0 2) including +2 percent power uncertainty, a ()
psia pressure uncertaint and a F temperature uncertainty but 57 Le with the increased radial pe ing fact r associated with the misa gned RCCA.
b DNB does not occur for the RCCA misalignment incident and thus the ability of the primary coolant to remove heat from the fuel rod is not reduced.
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Trio Rosetivity Insertion h The trip reactivity insertion assumed is given in Table 15.4 3 and includes the effect of one stuck RCCA adjacent to the ejected rod. These values are reduced by the ejected rod reactivity. The shutdown reactivity was simulated by dropping a rod of the required worth into the core. The start of rod action occurred 0.5 seconds after the high neutron flux trip point is reached.
This delay is assumed to consist of 0.2 seconds for the instrument channel to produce a signal. 0.15 seconds for the trip breaker to open and 0.15 seconds
.for the coil to release the rods. A curve of trip rod insertion versus time
'was used which assumed that insertion to the dashpot does not occur until 2.6 seconds after the start of fall. The choice of such a conservative insertion rate sesans that there is over one second after the trip point is reached before significant shutdown reactivity is inserted into the core. This con-servatism is particularly important for hot full power accidents.
The minimum t.asign shutdown sargin available for this plant at hot zero power (HZP) may be reached only at end of life in the equilibrium cycle. This value includes an allowance for the worst stuck rod, adverse menon distribution, conservative Doppler and moderator defects, and an allevance for calculational uncertainties. Physics calculations have shown that the effect of two stuck RCCAs (one of which is the worst ejseted roc) is to reduce the shutdown by about an additional one percent ak. Therefore, following a reactor trip resulting from an RCCA ejection accident, the reactor will be suberitical when the core returns to HIP.
Rosetor Protection As discussed in Section 15.4.8.1.1, reactor protection for a rod ejection is provided by high neutron flux trip (high and low setting) and high positive !'3 neutron flux rate trip. These protection functions are part of the Reactor Trip System (RTS). No single failure of the RTS will negate the protection l57 functions required for the rod ejection accident, or adversely affset the consequences of the accident.
Results Ca6ses are presented for both beginning and end of life at zero and full power.
I- Reginning of cycle. Full Fower c4 Control bank D was assumed to inserted to its insertion limit. ne worst ejected rod worth and t channel factor were conservatively caleu-lated to be 0.20 percent nd 7.10 respectively. The peak hot spot clad l18 average resperature was 2 9 h e peak hot spot fuel center temperature reached melt g, conservatively assumed at 4,900'F. However, melting was restricted to less than_ tan er ellet.
A ei /i ,.ntQ'q .of the AME
- 2. Seginning of Cycle Eero Power k Q ST-For this condition, control bank D was assumed to be fully inserted and banks B and C were at their insertion limits. The worst ejected rod is located in control bank D and has a worth.of 0.86 percent Ak and a hot channel f or of 3.0. The peak hot spot clad av a e temperature 18 reached .s *F he fuel center temperature was 3 F3 D*%
- \
15.4-30
@ g (t' endment 57 l \ 2 & b k3 . /
Q HANGENQTICE AM
5 ST-H L4E- 3ll g- PAGE 103 0F .
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- 3. End of Cycle. Full Power Control bank D was assumed to b inserted to its insertion limit. The ejected rod worth and hot cha 1 factors were conservatively calculated to be 0.20 percent Ak and 7 spectively. This resulted in a peak clad average temperature of l18 Fr The peak hot spot fuel center ter:-
perature reached selting at F. However, selting was restricted to less than ten percent of the pellet.
4 End of Cycle, Zero Power S
{
S, i
i; 46%>o wed.W The ejected rod worth and hot channel factor for this case era obtained
~ ,
assuming control bank D to be fully inserted and banks C at their pt insertion limits. The results were 1.0 percent Ak and 2 0 --.C..
The peak clad average and fuel center temperatures v e 2.
- and 4 4F ,
The Doppler weighting factor for this case is si .i cantly h g er than for the other cases due to the very large trans nt hot chan- M h nel factor.
9 g24 g.p4qp.qg l A summary of the cases presented above is given in Table 15.4 3. e . ar power and hot spot fuel and clad temperature transients for the worst cases !
(beginning of life full power and and of life zero power) are presented on Figures 15.4 26 through 15.4 29.
} The calculated sequence of events for the worst case rod ejection accidents.
}
as shown on Figures 15.4 26 through 15.4 29, is presented in Table 15.4-1. !
For all cases, reactor trip occurs very early in the transient, after which '
the nuclear power excursion is terminated. The reactor will remain suberitical following reactor trip. l The ejection of a RCCA constitutes a break in the RCS, located in the reactor pressure vessel head. The effects and consequences of loss of coolant acci-i dents are discussed in Section 15.6.5. Following the RCCA ejection, the oper-ator would follow the same emergency instructions as for any other loss of. ,
j coolant accident to recover from the event. '
Fission Product Release It is assumed that fission products are released from the gaps of all rods entering DNB. In all cases considered, less than 10 r rcent of the rods 3 entered DNB based on a detailed three dimensional THIM aalysis (Ref. '
15.4 10). Althou full power cases,gh limited it is fuel highly meltingthat unlikely at the hot spot selting was predicted vill occur since the for the analysir conservatively assumed that the het spots before and after ejection were coincident. -
1 Pressure surne A detailed calculation of the pressure surge for an ejection worth of one dollar at beginning of life, hot full power, indicates that the peak pressure does not exceed that which would cause stress to exceed the faulted condition stress limits (Ref. 15.4-10). Since the severity of the present analysis does not exceed the " worst case" analysis, the accident for this plant will not result in an excessive pressure rise e,t further damage to the RCS.
CHANGE NOTICE Md7 l
ATTACHNI5NT 3~
3llY STF TSAR ST-HL ,A{-
PAGE U OF N Table 15.4 1 1 TIME SEQUENCE OF TVENTS FOR INCIDENTS WHICH CAL'SE REACTIVITY AND PC'='ER DISTRIBUTION ANOMALIES Accident Event T se e Uncontrolled Rod Initiationofuncontrolged v . o
Cluster Control rod withdrawal from 10 C Assembly Bank of nominal power Withdrawal from a Suberitical or Low Fever Startup Condition Power range high neutron 13.7 7 flux low setpoint reached i
\
Feak nuclear power occurs 13.8 L 5' Rods begin to fall into 14.2 1 core Minisua DNBR occurs 15.6 j 8 (
\
Feak average clad temperature 15.6 1 ,L occurs Peak heat flux occurs 15.6 I I' Feak average fuel temperature 15.8 1 S . C. '
occurs Uncontrolled RCCA Sank Withdrawal at Power
- 1. Case A Initiation of uncontrolled 0 RCCA withdrawal at a high rsactivity insertion rate (70 pea /sec)
Power range hi5 h neutron flux 1.7 high trip point reached
- ~
Rods begin to fall into core 2.2 Minimus DNBR occurs 34 st
~
5.h i
M's CHANGE NOTICE M87 33,5 37 4aendment 37 l
l
R ATTACfiliEliT 3/
ST-HL-AE5 PAGE lo 0F 9 STP FSAA Table 15.4 1 (Continued)
TIME SEQUENCE OF EVENTS FOR INCIDENTS VHICH CAUSE REACTIVITY AND POWER DISTRIBUTION ANOMA1.IES Accident Event Time (see.s V a-
~
- 2. Case B Initiation of uncontrolled N RCCA withdrawal at a small reactivity insertion rate (5 pem/sec) ph Overtemperature AT reactor I ll.7 trip signal initiated Rods begin to fall into core 13.5 '
i Minimum DNBR occurs Startup of an Inactive Initiation of pump 0 Reactor Coolant Loop Startup l
Power reached P 8 interlock 10.2 A setpoint, coincident with v low reactor coolant flev
/\
Rods begin to drop 11.2 s.E Minimum DNBR occurs 12.0 I .C 1 Uncontrolled Boron '
j Dilution f
- 1. Dilution during Power range hi5 h neutron flux. O startup lov setpoint reactor trip due to dilution Shutdown margin lost (if -1200 -l i dilution continues after trip) l- .
de\. s=.h CHANGE NOTICE M/ 7 4
l 15.4 38 Amendment 57 1
= - - _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
d- 1 ATTACHMENT ST-HL-AE,- 3lE l STF TSAR PAGEf I tf 0F k -l
]
Table 15.4 1 (Continued)
TIME SEQUENCE OF EVENTS FOR INCIDENTS VHICH CAUSE REACTIVITY hre, AND POWER DISTRIBUTION ANOMALIES
)
Accident Event Tire (sec.)i 2 Dilution during N)'T~
/ T full power operation
- a. Automatie 09erator receives low.lov' 0 l reactor control rod insertion limit, alarm due to dilution j
s shutdown margin lost -1620 ~
- b. Manual reactor Reactor trip on 0 C q control overtemperature AT / j due to dilution j 1
Shutdown margin lost (if dilu. -1020 f tica continues after trip) 4 '
Rod Cluster Contro?
Assembly Ejection'
, 1. Beginning of Initiation of rod ejection 0.C '
Life. Tull Power Power range high neutron Ap6 0. 0 I '
flux setpoint reached / p Feak nuclear power occurs O'. $ 0.I3 Rods begin to fall into core .56. 8. O f e . 9 E I' l
Feak heat flux occurs .59 el. S$
- 2. End of Life. Initiation of rod ejection .0 c.C Zero Power
~
Fower range high neutron .
flux low setpoint reach l bb-L l
l l.
CHANGE NOTICE OO 15.4 39 Amendment 57 l
ATTACHMENT ST-HL ,\ - 3I STP TSAR PAGE OF 07 Table 15.4 1 (Continued)
.t-TIME SEQUENCE OF EVENTS FOR INCIDENTS WHICH CAUSE REACTIVITY
_AND POWER DISTRIBUTION ANOMALIES Accident Event Time (see.)
Peak nuclear power occurs Rods begin to fall into core .4 0. b'l
- n. lad +e ntwo Feak '--- w-- occurs 46 I . Il I Peak kas.4
- Ed --S\J;- y : n occurs 4 (,E f
aferu tt-Peak fuel4temperature occurs 1.97 l , g, Y
CHANGE NOTICE M 7 15.4 40 Amendment 57
ATTACHMENT 3Il
$%P TSAR ST-HL-SE PAGE to E OF k9 TABLE 15.4-3[
G47) PARAMETERS ,USED IN THE ANALYSIS OF Tite ROD CLUSTER
- COATROL ASSDiBLY EJECTION ACCIDENT I
Begi ng Reginning d End of cy le of eye 3e cycle of eyel Power evel, % 102 0 102 0 Ejected rod worth, AK 0.20 0.86 0.20 1.
Delayed n utron fr crion, % 0.55 0.5 0.44 44 Feedback r criv y weighting 1.60 2. 0 1.30 4.5 Trip reactiv ty %aK 5.0 .0 4.0 2.0 Fq before rod jection 2.50 -
2.50 -
r after rod je tion q
7.10 2 .D 7.10 0.00 Wumber of op rati nal pumps 4 2 4
!!ax. fuel p 11et av rage temperat re. *F 4090 2856 3900 3453 flax. fuel center tempe sture,
'T 490 3476 800 4154 Max. c1 average tempera ure.
'F 2 19 2001 2096 2422 Hex. f el stored energy, cal /gm 179 117 169 146 e_
M k t fE._ LE=
CHANGE NOTICE M/7 54 % f 15.4-d Amendment )1I. "S*C
l TACHMENT g AGE OF STP TSAR -
TABLE 15.4-3 ,
ARAMETERS USED IN THE ANALYSIS OF THE ROD CLUSTER CONTROL ASSEMBLY EJECTION ACCIDENT L
8eginning Seginning End End of cycle of cycle of cycle of eyele Power level, % 102 0 102 0 ,
Ejeetad rod worth, %AK 0.20 0.86 0.20 1.0 Delayed neutron fraction 1 0.55 0.55 0.44 0.44 Feedback reactivity weighting l.303W5 1.2 M 1.30 ArL 3.r5 -
Trip reactivity *.aK 4.0 Duc 2.0 4.0 2 . .,
Fg before rod ejection 2.67 W -\ 2. 6J.*.>90 -
Fq after rod ejection 7.10 13.0 7.10 M 23.00 Number of operational pu.ps 4 2 4 2 Hax. fuel pellet average temperature, 'T 41% 3DQ 34. fi 2DG llCitHC 3mC 1537 <
Sax. fuel center temperature,
'T 4900 4113 3H1 4800 CHC 4012. j Hax. : lad average te=perature,
'T Z,1.9 2.25%, L606 M 10iL2D96 M 2.67i Hax. fuel stored energy, cal /gm PPC 2M 26% 3m(, i ilY l[$ lb5 l[t P+w4 ful w.+ co o cre e I l
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SOUTH TEXAS PROJECT UNIT.' 1 CHANGE NOTIM /JO s.1%ro.i,A#',i$;,1,,,,,w,7,,n,l e e7 EOL MZP Rod Eiection Accident
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ATTACHMENT h
'\ kGE t) 0F N STP TSAR Response (Continued)
As shown in Table Q440.47N- , the minimum DNBR and rod action for most of the Condition II events occur n less than two seconds after reactor trip.
a LOOP occur two seco Shoul:1 after reactor trip, the RCPs would coast down at the same rate as the comp see loss of flow analysis (Section 15.3.2). Since the coastdown is occurr decreasing rapidly g after the time of minimum DNBR and the reactor power is
" facted. For the us to rod action, the minimum DNBR is not adversely af-assumed two seco, ses where the minisua DNBR occurs after the conservatively power occur two a delay, it is easy to show that should a loss of offsite
[ the complete loss econds of flow.
af ter reactor trip, the results would be bounded by For example, in case b) of the uncontrolled rod cluster control a ssembly (RCCA) withdrawal at power analysis, che ainfram DNBR occuTs#C nM -#
tri r n " - --Sp and rod motion W seconds after reactr.:
. Should a LOOP occur at the time of rod motion the nuclear power wo '* fy,
/ rease rapidly percent of thermal due to the rod design flowmotion, while the flow would be approxi=ately ese core conditions are less severe tha--
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those of the cesplete loss of flow at the time of minimum DNBR. Thus, the ur.-
controlled RCCA .tithdrawal at power eve ich th0P is bounded by the cot:ple:e loss of flow event. A similar argument ca be made for the inadvertent open-
, ing of a pressurizer safety or relief valve event.
N All of the design basis events are analyzed with and without offsite power available.
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ATI5CHMENT T ST-HL-AEr STP FSAR PAGE it' I 3d n9 0F 7 TABLE OL40.47 1N (Continued)
Time of Time of Time of FSAR Reactor Rod minimum section Accident Trip Motion DNBR Comments 15.2.6 Loss of nonemergency ac power to the plant This is a LOOP auxiliaries 15.2.7 Loss of normal feed-water flow Sec. 15.2.6 is case with LOOP 15.2.8 Feedwater system pipe break Done with ar.d without LOCP 15.3 Decrease in reactor coolant system flowrate Bounded by Complete Loss of Flow 15.4.1 Uncontrolled RCCS 10.2 10.7 12.0 withdrawal from a suberitical or low '
power startup condition 6
15.4.2 Uncontrolled RCCA withdrawal at power a) 98E M i
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W87 15.4.3 RCCA misalignment
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f t. 7 yg 1.Vuet. FSAR l being updated 1, H.$. // / 3. 3 ty 13.$ to incorporate negative flux rato trip thodology per 10297 P.A 15.4.4. Startup of an inactive 10.2 11.2 12.0 reactor coolant loop \n loop at an incorrect temperature 15.4.5 Failure of a BWR Not applicable flow controller to South Texas Vol. 3 Q&R 15.0-4N Arendment~49 CHANGE NOTICE /387
ATTACHMENT '
ST-HL-Ak 3ll ,.
S'IP FSAR , PAGE M 0F' i l i
Quescion 213.51 For Chapter 15 accident events, provide the number of fuel rods calculated to .
I be in DNB.
.. h-For Condition II events it is d strated that DRIR runnins greater than the limiting value; thus, the a m r of rods calculated to be in DN8 corresponds to the criteria set forth action 4.4.1.
For large and.small LOC . uncovering of the core results in DN2 for all rods. For the steamline break events (Section 15.1.5). the feetwater Igne l break events (Section 15.2.8), and the complete loss of forced reactor coclant l4 flow events (Section 15.3.2), the DNBR does not f all below the 1.isiting value as indicated in the appropriate sections of Chapter 15. Therefere. the criteria for rods in DNB' presented in Chapter 4 applies to these evente also.
As stated in Section 15.4.3. the number of fuel rods with DNBR less than the
' limiting value for the single RCCA withdrawal event is less than 51 of the rods in the core. For an it7 toper fuel loadig event, undetected errors will cause sufficiently small perturbations to be acceptable within calculational uncertainties, as stated in Sect on 15.4.7; akus. the effect due to impreper
!. loading on rods in DNB f tra sient t will be le.
f L be. E thAs \Q%
The RCCA ejection analystsspzase ed etto 5.4. co ervativelv ossuses l.
that 10% of the rods in the core o into DNS d fail. For the locked rotor event presented in Section 15 3.3 ths max number of fuel rods in 11Q is [.
conservatively calculated to t 3 9 : :P '
of the rods in the corehd)As stated in section 15.3.4. the consequences e an reactor cooling pump shait break will be less severe than those for a locked rotor event.
Evaluation of the steam generator tube rupture indicates that no clad damap would be espected in this transient. The RCS depressurisation due to flow out -
of the tube rupture presents the possibility of obtaining a low DNBR.
Bowever, the depressurisation in a tube rupture is such less severe than the depressurisation tramstent analysed la section 15.6.1. In this accident it was determined that the DNSR is always greater than the limiting value, and thus me alad damage is supected. From this, it is concluded that so clad damsge is espected is the staan generator tube rupture accident. For all other events diossased is Chapter 15. DNat rematas above the limiting value. .
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CHANGE NOTICE M7 ,,, 33,,_3, ,,,,,,,,, ,3
ST-HL-AE 3 llEoc PAGE F 0F 7
-section 15.3.3.3 contains conflicting statements as to whether or not fuel failure ' occurs. Clarify this discrepancy and, if fuel failure is not assumed for some conditions, provide a justification, with bases, since DNB is assumed to occur.
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Response
i-The lockad rotor transient for South Texas presented in Sectica 15.3.3 of ;
the FSAR is performed in several different parts.
Part 1 Calculation of Peak Puel Clad Temperature
< To anximise the fuel clad temperatures, DNB is assumed to occur at the start of the transient. The analysis showed the peak clad temperature is approximately '1800*T, well belos the limit value of 2700'F. Thus, no clad' failures are calculated to occur.
1 Part 2 The Number of Rods in DNB g f
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The number of fuel rods calculated to experiebce DNB was .-
U percent A rod experiencing DNB does not necessarily rean it Part 3 "12 -
Dose Release www#pe hMg ]
i for the purpose of calculating dose releases it aus ;
conservatively assumed that fuel experiencing DNB fails, even though the peak clad temperature sus not high enough tu cause l fuel failure. The South Texas dose release evaluation for a l locked rotor conservatively assumed 7 ercent of the fuel failed.
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-l CHANGE NOTICE /88 7 1
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