ML20151N014
| ML20151N014 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 07/01/1988 |
| From: | Gottula R SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML14191A345 | List: |
| References | |
| ANF-88-094, ANF-88-94, NUDOCS 8808080074 | |
| Download: ML20151N014 (43) | |
Text
{{#Wiki_filter:- - - - [ plCA__._c. A N F-8 8-0 9 4 W s J \\ f(%Q 7
- g, ADVANCED NUCLEAR FUELS CORPORATION H.B. ROBINSON UNIT 2
CH APTER 15 OVERTEMPER ATURE AT TRIP EVENT AN ALYSIS FOR ELIMIN ATION OF i RTD BYP ASS PIPING [ JULY 1988 --w 2
ADVANCEDNUCLEARFUELS CORPORATION ANF-88-094 Issue Date: 7/1/83 H. B. ROBINSON UNIT 2 CHAPTER 15 OVERTEMPERATURE AT TRIP EVENT ANALYSIS FOR ELIMINATION OF RTD BYPASS PIPING By
- k. l. /
h R. C. Gottula, Team Leader PWR Safety Analysis Licensing & Safety Analysis Fuel Engineering & Technical Services June 1988 gf t
CUSTOMER DISCLAIMER NPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY Advanced Nuctaar Fuels Corporacon's warranties and representatens con. coming the subloct matter of the document are those set forth in the Agreement between Advanced Nuclear Fuels Corporaten and the Customer pursuant to wtuch this document is issued. Accordingfy, except as otherwise expressty pro-viced in such Agreement, nocer Advanced Nuclear Fuels Corporaton nor any person acting on its behatt makes any warranty or reoresentation, expressed or implied, with resoect to the accuracy, completeness, or usefulness of the infor. maten contained in this documet, or that the use of any informaton, apparstus, method or process disclosed it; this document will not intnnge pnvately owned nghts; or assurnes any liabilities with respect to the use of any information, ao-parstus, method or process disclosed in this docurnent. The informadon contamed herewt is for the sole use of Customer. In order to avod impaarment of nghts of Advanced Nuclear Fuels Corporation in patents or wtventens wnich rnay De included in the mformaten contained in this document, the rectoient, by its acceptance of this doeurnent, agrees not to publish or make puotic use (in the patent use of the term) of such informaten until so authonzed in wntog by Advanced Nuclear Fuels Corporaten or untd after six (6) months following terminaton or expiration of the aforesaid Agreement and any extensen thereof, unless omerwise expressty provided in the Agreement. No ngnis or licenses in or to any patents are imphed by the fumishing of this docu-ment. XN NF F00 765 Ol8D
ANF-88-094 Issue Date: 7/1/88 H. B. ROBINSON UNIT 2 CHAPTER 15 OVERTEMPERATURE AT TRIP EVENT ANALYSIS FOR ELIMINATION OF RTD BYPASS PIPING Distribution GJ Russelman LJ Federico RC Gottula JS Holm DC Kilian TR Lindquist LD O' Dell GL Ritter l FR Skogen l 1Z Stone BD Webb HE Williamson CP&L/HG Shaw (10) Document Control (5)
i ANF-88-094 Page i TABLE OF CONTENTS Section Paae
1.0 INTRODUCTION
1 2.0
SUMMARY
1 3.0 ANALYSIS OF PLANT TRANSIENTS.............. 5 1 15.2.1 LOSS OF EXTERNAL LOAD 15.2.1-1 15.4.2 UNCONTROLLED CONTROL R0D ASSEMBLY WITHDRAWAL AT POWER 15.4.2-1 15.4.3 CONTROL R0D MIS 0PERATION (SYSTEM MALFUNCTION OR OPERATOR ERROR) 15.4.3-1
4.0 REFERENCES
6
ANF-88-094 Page ii LIST OF TABLES Table Eagg ( 2.1 Overtemperature AT Reactor Trip Delay Times and Lag Constants 3 2.2 Summary of Event Analysis Results 4 15.2.1-1 Loss of External Load - Summary of Initial Operating Conditions.............. 15.2.1-5 15.2.1-2 Loss of External Load Event Sequence.......... 15.2.1-6 15.4.2-1 Uncontrolled Control Rod Assembly Withdrawal - Summary of Initial Operating Conditions 15.4.2-4 15.4.2-2 Uncontrolled Control Rod Assembly Withdrawal l Event Sequence..................... 15.4.2-5 15.4.3-1 Dropped Full Length RCCA (Manual) - Summary of Initial Conditions 15.4.3-4 15.4.3-2 Dropped Full Length RCCA (Manual) Event Sequence.... 15.4.3-5 l l l
ANF-88-094 Page lii LIST OF FIGURES Fiaure Elat 15.2.1-1 Loss of Load, Reactor Power Level 15.2.1-7 15.2.1-2 Loss of Load, Core Inlet, Average, Cold and Hot Leg Temperatures 15.2.1-8 15.2.1-3 Loss of Load, Pressurizer Pressure........... 15.2.1-9 15.2.1-4 Loss of Load, Pressurizer Liquid Volume 15.2.1-10 15.2.1-5 Loss of Load, Steam Generator Dome Pressure 15.2.1-11 15.4.2-1 Slow Rod Withdrawal,100% Rated Power, Reactor Power 15.4.2-6 15.4.2-2 Slow Rod Withdrawal,100% Rated Power, Primary Coolant Core Inlet, Average, Cold and Hot Leg Temperatures 15.4.2-7 15.4.2-3 Slow Rod Withdrawal, 100% Rated Power, Pressurizer Pressure.................. 15.4.2-8 15.4.2-4 Slow Rod Withdrawal, 100% Rated Power, Pressurizer Liquid Volume 15.4.2-9 15.4.2-5 Slow Rod Withdrawal,100% Rated Power, Steam Generator Dome Pressure 15.4.2-10 15.4.2-6 Slow Rod Withdrawal, 100% Rated Power, Reactivity Additions.................. 15.4.2-11 15.4.3-1 RCCA Drop, Reactor Power................ 15.4.3-6 15.4.3-2 RCCA Drop, Core Average and Inlet Coolant Temperatures...................... 15.4.3-7 15.4.3-3 RCCA Drop, Pressurizer Pressure 15.4.3-8
ANF-88 094 Page 1
1.0 INTRODUCTION
Presented in this report are the results of Standard Review Plan (SRP)(l) Chapter 15 event analyses performed for H.B. Robinson Unit 2 to support a revised RTD installation design which eliminates the bypass piping. The analyses include a simulation of the processing of temperature signals input to the Overtemperature 6T trip. The associated lag constant and delay times are shown in Table 2.1. The analyses are structured to support a Technical Specification Fag limit of 1.65. Basic assumptions used in the analyses include (a) precluding the withdrawal function of automatic rod control, (b) a non-positive moderator temperature coefficient above 50% power for the rod drop transient, and (c) a K1 value in the OTAT trip function of 1.24 including uncertainties. The results of the analyses are summarized in Section 2.0 of this report. The detailed event case descriptions and results are provided in Section 3.0. References for this report are listed in Section 4.0. 2.0 SUtHARY The DNBR limiting cases of the loss of external load, uncontrolled control rod withdrawal, and control rod drop events were simulated to determine the validity of the Overtemperature AT reactor trip function for the new RTD installation which eliminates the bypass piping. These events were identified in the analysis of record (2) as the most limiting DNBR events which also trip on the Overtemp, ture AT reactor protection failure. The calculations employed stand J ANF thermal hydraulic and Chapter 15 event calculation methodology (3 MDNBRs were calculated with the XNB critical heat flux correlation (5), which is applicable to ANF fuel in the H.B. Robinson Unit 2 reactor (6). Operating conditions and event acceptance criteria are as given in Reference 2 unless otherwise noted. Event results are summarized in Table 2.2. Calculated MDNBRs for the events analyzed are above the XNB critical heat flux correlation safety limit
ANF-88-094 Page 2 of 1.17. Based on these results, it is concluded that applicable acceptance w criteria are met with the current Technical Specification limit on F f 1.65 AH and a Technical Specification Overtemperature AT trip function with a K1 value I of 1.24 including uncertainties. l 1 1 e 4 ~
ANF-88-094 Page 3 TABLE 2.1 OVERTEMPERATURE AT REACTOR TRIP DELAY TIMES AND LAG CONSTANTS [.Qmponent Thermal lag representing both thermal transport 4.0 see through the thermowell and the RTD response time Electronics delay representing electronic signal 0.75 sec processing, trip breaker operation, and control rod drive shaft gripper release i
ANF-88-094 Page 4 TABLE 2.2
SUMMARY
OF EVENT ANALYSIS RESULTS MDNBR Transient Event gg) 15.2.1 Loss of External Load 1.19 15.4.2 Uncontrolled Control Rod Assembly 1.19 Withdrawal at Power 15.4.3 Full length RCCA Drop (Manual) 1.23 I
ANF-88-094 Page 5 3.0 ANALYSIS OF PLANT TRANSIENTS The FSAR Chapter 15 events were reviewed to determine the limiting DNBR events which also trip on the Overtemperature AT reactor protection feature. The events identified were: (1) Loss of Load - MDNBR Case; (2) Uncontrolled Control Rod Bank Withdrawal From Full Power; and (3) Dropped Full Length RCCA. These events were analyzed using the PTSPWR2 plant transient simulation . code (7) The version of PTSPWR2 used incorporates a model which simulates the time lags and delays resulting from the RTD installation in a thermowell as well as electronic signal processing, trip breaker operation, and control rod drive shaft gripper release. This section provides the results of the analysis of the three limiting events. Subsections in the report are enumerated in accordance with the SRP in order to facilitate review. Single failures were considered for each of the three limiting events.
ANF-88-094 Page 6
4.0 REFERENCES
(1) ' Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," NUREG-0800, U.S. Nuclear Regulatory Commission, July 1981. (2) XN-NF-84-74. Rev. 1, "Plant T at 2300 MWt With Increased Fgansient Analysis for H.B. Robinson Unit 2 " Exxon Nuclear Company, Richland, WA, AH, April 1986. (3) XN-NF-82-21(A), "Application Of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," Exxon Nuclear Company, Richland, WA, September 1983. (4) XN-NF-84-73(P), "Exxon Nuclear Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," Exxon Nuclear Company, Richland, WA, December 1984. (5) XN-NF-621(A1. Rev. 1, "Exxon Nuclear UNB Correlation for PWR Fuel Designs," Exxon Nuclear Company, Richland, WA, September 1983. (6) XN-NF-711(P1, "Extension of XNB Correlation to PWR Fuel Assembly Designs With Spacer Pitch Greater Than 22 Inches," Exxon Nuclear Company, Richlar:d, WA, May 1983. (7) XN NF-74-5(A) & Sucos. 1-6. Rev. 2, "Description of the Exxon Nuclear Flant Transient Simulation Model for Pressurized Wter Reactors (PTS-PWR)," Exxon Nuclear Company, Richland, WA, October 486. (8) File Number NF-2484/Nr.1077. Rev. 3, "Single Failure Analysis, H.B. Robinsen Unit 2," Carolina Power and Light Company, April 1985. (9) XN-CC-28. Rev. 5, "XTG: A Two Group Three-Dimensional Reactor Simulator Utilizing Coarse-Mesh Spacing (PWR Version)," Exxon Nuclear Company, Richland, WA, July 1979. n
ANF-88-094 15.2.1-1 15.2.1 LOSS OF EXTERNAL LOAD The analysis addresses the DNBR branch of the Loss of External Load event. That event was shown to result in an Overtemperature AT reactor trip in che analysis of record.(2) The event is analyzed to verify the OTAT trip function with the new RTD installation which eliminates the bypass piping. 15.2.1.1 Identification of Causes and Event Descriotion A loss of external load can result from loss of the generator Jue to an electrical system disturbance. Offsite electrical power is available to operate the reactor coolant system pumps and other station auxiliaries. Following the loss of generator load, the turbine stop valves close, terminating the steam flow and causing the secondary system temperature and pressure to increase. The primary-to-secondary heat transfer decreases as the secondary system temperature increases. Two event sequences may be postulated, one leading to a challenge of the vessel pressurization criterion, the second leading to a challenge to the DNBR limit. Only the second event sequence resulted in an Overtemperature AT reactor trip in the analysis of record.(2) Only the event challenging the ONBR limit is analyzed here. If the reactor is not tripped when the turbine is tripped, the primary system temperature continues to rise. The primary liquid will expand and the pressurizer steam space is compressed, causing the pressurizer pressure to rise. In the event sequence considered, the mitigative features of the pressurizer spray and prersurizer relief valves are assumed to function. This minimizes the pressurization of the primary
- system, resulting in a
conservative evaluation of the MONBR for this event. Energy is removed during the early phase of the transient through the steam generator safety valves when the steam generator pressure exceeds the safety valve opening setpoint. )
9 ANF-88-094 15.2.1-2 The challenge to the specified acceptable fuel design limit (SAFDL) on DNBR is evaluated because of the increasing core inlet temperature and the potential for the reactor core power to increase prior to reactor trip. Reactor control is assumed to be in the manual mode so the reactor power will not be reduced when the primary system average temperature begins to increase. This event is a moderate frequency (Condition II) event. The acceptance criterion for this event sequence is that the MDNBR during the transient must be above 1.17. The cited (8) single failure for this event does not affect the results of the analysis of the event since the Engineered Safety Features (ESF) are not challenged within the time period of interest. 15.2.1.2 Analysis Method This event is analyzad with the PTSPWR2 computer program.(7) The core tharmal-hydraulic boundary conditions from the PTSPWR2 calculation are used as input to the XCOBRA-IIIC methodology (3) to predict the minimum DNBR for the event. 15.2.1.3 Definition of Bents Analyzed and Boundiro Inout This event is analyzad to ensure that reactor protection systems are properly set to prevent penetration of the SAFDLs. The analysis takes no credit for the turbine bypass system or for direct reactor trip on turbine trip. The inpuc ;arameters are biased to maximize the increase in reactor power during the transient consistent with inimizing event MDNBR. Also, the parameters and the equMient operational states are selected to reduce the prita y system pressurization to provide a conservative estimatio:: of the minimum DNBR during the transient. The bounding operating mode for this event is full power initial conditions with the reactor control system in the manual mode. ~
ANF-88-094 15.2.1-3 15.2.1.4 Analysis of Results { The event initiates with closure cf the turbine control valves. Steam line pressure increases until the secondary side safety valves open at 6.1 sec. The maximum pressure in the steam dome of the steam generators is not achieved until 19.9 sec. The pressurization of the secondary side results in decraased primary to secondary heat transfer aM a substantial rise in cold g leg temperature. The rapid increase in primary side temperatures result in a large insurge into the pressurizer, compressing the steam space and pressurizing the primary system. The pressurizer PORVs open at 4.7 sec. and the pressurizer safety valves are opened at 8.8 sec. The valves have enough capacity to ( mitigate the pressure transient and limit the pree -e to the safety valve setpoint value. Pressurizer PORVs and spray limit the pressure rise, preventing a reactor scram on high pressure. Reactor scram occurs on Overtemperature AT, with rod insertion commencing at 15.5 sec. The reactor power reaches about 117.5%. The steam generator safety valve flow limits the core average temperature rise to 24.0*F at 16.7 sec. The DNBR challenge results from the core power and primary coolant temperature increase. The challenge is exacerbated by the action of the pressure control systems. Plant initial operating conditions assumed in the analyses are summarized in Table 15.2.1-1. The transient response to this event is shown in Figures 15.2.1-1 to 15.2.1-5. An event summary is shown in Table 15.2.1-2. The minimum DNBR was computed to be 1.19. This is above the DNBR limit of 1.17 for the XNB correlation.
ANF-88-094 15.2.1-4 15.2.1.5 Conclusion The minimum DNBR is greater than the XNB DNB correlation safety limit. Therefore, the DNBR acceptance criterion is met.
ANF-88-094 15.2.1-5 TABLE 15.2.1-1 LOSS OF EXTERNAL LOAD -
SUMMARY
OF INITIAL OPERATING CONDITIONS Parameter Value Power (MWt) 2346 Core Inlet Temperature (*F) 550.2 Pressurizer Pressure (psia) 2220 W l R2 actor Coolant System Flow Rate 97.29 x 106 (lbm/hr) Steam Dome Pressure (psia) 828.3
ANF-88-094 15.2.1-6 TABLE 15.2.1-2 LOSS OF EXTERNAL LOAD EVENT SEQUENCE Event Time (sec) Turbine Trip 0.0 Pressurizer PORVs Open 4.7 Steam Line Safety Valves Open 6.1 Pressurizer Safety Valves Open 8.8 Peak Pressure 15.0 Reactor Scram (Begin Rod Insertion) 15.5 Peak Power 15.5 Minimum DNBR 16.3 Peak Core Average Temperature 16.7 Peak Steam Dome Pressure 19.9
ANF-88-094 15.2.1-7 ON A c .c m I O W ....s.'...........:........ ....)..............>............ 4..... M w W a m w 3 C. o .. <...........t.... .g..............:..............<..............N c, w m o h U b 0 w> CQ W F-< cc .O = 0 w k d O O L.- o c .g g (H m O c J m O ..). ..........;...........<............c i l l l <N. m O .....t.............. y .g ). a e-O O O O O O O O O O O O O O O O O O O O l7) N N M M (EUi) JOA0d lumaql, I
Loss of Load, MDNBR Case 675 l \\ i TCIO i 650- .t--- -- -------+---------- ---t. -----~~-------t----- --------- --------- -- - +- -------- III21
- W~
m 625- ------- - -. ----- -----:----------- CD D i f i i d v c) 0 600~
-- ~~-- =.------ - -- =. ~--- ---------:.--------~~----- - ---------- =.----
--=. --------------.-- .es 4 Q D.
.----------i--------------i---------+-:r---._-:..-+,
t o 575 - r - - - - - - - - - _ :.- : = ~ -,. : - - - - - - - - - - - - - E-+ X .a-650
.~c"-------:-~=----------t.------------'-------------.--------------+.--------------------------
l \\ m.n l 525 .m z o> a i 5 I s i i w 03 0 2.5 5 7.5 10 12.5 15 17.5 20 &6 e Time, sec u fIGURL 15.2.1-? LOSS OF LOAD, CORE INLET, AVERAGE, COLD AND 110T LEG TEMPERATURES r I ' l-s a
Loss of Load, MDNBR Case 2450 P - ... P. R _. 2400-4- +- -i.-- - - A: - = --- q 2350-- + - + - i---- - - + - - - -i---- ---- i- - - - -- - .. e C1 a 2300.. 4 O L U m +- - + -- -- -- --i - m 2250- + --+ o4 A % 2200-- +- --- 5 -- -i - o N .a 0 2150 - .. 7.. .s. m m o u % 2100-4- + i 2050- - -i-- -t-- -t------- v-N mz l 71 N i l 2000. i - m ~m l 0 2.5 5 7.5 10 12.5 15 17.5 20 e a eo Time, sec g FIGURE 15.2.1-3 10SS OF l.0AD, PRESSURIZER PRESSURE
Loss of I oad, MDNBR Case 1200 CFWPR
- l..Z.. '.l.T_
~ ~ ' Os m 1:00 .g.. O o s a 1000- - - - - --> -- - -- i-- -- b- -+- ~ O> l ..a y . o,' A l 900 ......b. 1 k t) .s, u 3 v1 7, U 800-4--- --b-i- --- - + - b- - 4 3. l l mi m 700- -m 0 2.5 5 7.5 10 12.5 15 17.5 20 0 E> l Oe ,l,im O, S CC l FIGURE 15.2.1-4 1 OSS Of 10AD, PRESSURIZER I.IQUID V01.UME
Loss of Load, MDNBR Case 1300 l I i PD01_ 1200-n - - - 1. --- -- - t - -- - - -?- - - - ---.- -- - ------ r--- - - -- --- ! -- ----- --- as I l i y) i I CL v O L Q 1100- -- --- -~ ~.>- ---- ----- -. ~~~ - - ----t- ------ - .t---- --~ ~ --> - - ----~~^-----~~---t.--------~~ m M O 5-e g O G 1000- -----------b
5.~~-------:--------}--------[-----~~--{.---~~~~--{.-~~-----
O ~ .s. 900- -- - -- -----:- -- - ------.- - ~ ~ ~- ~ ~ ~.~ - ----- --.-- -~ ~ - ----: - --- -- --- ~ r. - ~ ~~ ~--- -:.-- ----------- 800 ~> .m =, r N e i i e i e i m' 0 2.5 5 7.5 10 12.5 15 17.5 20 ~ m. e -o T1rne, sec - 10a FIGURE 15.2.1-5 LOSS OF LOAD, STEAM GENERATOR DOME PRESSURE
ANF-88-094 15.4.2-1 15.4.2 UNCONTROLLED CONTROL R0D ASSEMBLY WITHDRAWAL AT POWER The analysis addresses the limiting uncontrolled rod withdrawal transient resulting in reactor trip on the Overtemperature AT reactor trip. That case was determined by review of the reference anz1ysis(2) to be a rated power case at B0C conditions. The reactivity insertion rate is that resulting in minimum DNBR, and results in simultaneous Overtemperature AT and power range high flux trips. The event is discussed below. 15.4.2.1 Identification of Causes and Event Description This event is defined to result from an uncontrolled control bank withdrawal at full power. The event could be caused by misoperation of the most reactive control rod banks wired in common withdrawing at up to the maximum rate. The reactor protection system is designed and set to preclude penetration of the SAFDLs. The Overtemperature AT and power range (high setting) high flux trips are principally challenged. Both trip setpoints include allowance for process variable measurement, processing channel drift, and operating variances from that indicated. The Overtemperature AT function is designed and set to protect against DNB. Principal DNB parameters such as power (measured as core coolant temperature rise), core coolant temperature, primary pressure and core power distribution are measured, and the function decreases margin to trip setpoint when process variables indicate a decrease in operating margin. This function is established based on the core protection boundaries, operation within which assures protection of the SAFDLs. Reactivity insertion rates are large enough at less than the maximum rate that core temperature rise lags behind nuclear power. The power range reactor trip protects the system from these events.
ANF-88-094 15.4.2-2 A broad range of reactivity insertion rates and initial operating conditions are possible. The range of reactivity insertion is from vary slow, as would be associated with a gradual boron dilution, and bounded o.1 the fast end of the range by bank withdrawal. The objective of the analysis is to demonstrate the adequacy of the trip setpoints to assure meeting the acceptance criteria. To assure this objective, the limiting rod bank withdrawal transient was analyzed to assess the impact of the new RTD installation which eliminates the bypass piping. This event is classified as a Condition II event. The acceptance criterion is that the SAFDLs must not be penetrated. This will be assured if the minimum DNBR is above 1.17. The systems challenged in this event are redundant and no single active failure will adversely affect the consequences of the event.(8) 15.4.2.2 Analysis Method The analysis is performed using the PTSPWR2(7) code and the XCOBRA-IIIC methodology.(3) The PTSPWR2 code models the salient system components and calculates neutron power, fuel thermal response, and fluid conditions. The fluid conditions and rod surface heat transport at the time of MDNBR are transposed to the XCOBRA-IIIC methodology (3) for calculation of the MDNBR. 15.4.2.3 Definition of Events Analyzed and Boundina Inout The limiting rod bank withdrawal event is from full power initial conditions with an insertion ramp of 2 pcm/sec and positive reactivity feedback. Additional cases were analyzed to verify the trends of the reference analysis (2) plot of MDNBR versus reactivity insertion rate. This ensured that I the limiting case was selected for analysis. l
ANF-88-094 15.4.2-3 15.4.2.4 Analysis of Results The limiting rod bank withdrawal transient was analyzed using a modified version of PTSPWR2 which simulates the new RTO installation. The limiting ( event was a reactivity insertion ramp of 2 pcm/sec from full power initial conditions with positive reactivity feedback. Initial conditions for the event are summarized in Table 15.4.2-1. Thermal power increases steadily throughout the transient in response to ( the reactivity insertion until the occurrence of reactor scram. Coolant temperatures also increase steadily due to the primary-to-secondary system power mismatch. The pressure increase due to coolant expansion and insurge flow to the pressurizer is limited to a maximum of 2274 psia by the primary PORVs. Increasing core power and temperature result in a reactor trip on the Overtemperature AT reactor trip at 27.2 sec. The MDNBR occurs shortly after the beginning of scram. Figures 15.4.2-1 through 15.4.2-6 show the plant responses for the limiting rod bank withdrawal transient. Table 15.4.2-2 presents the sequence of events for this event. The calculated MDNBR is 1.19. Calculations were performed for various reactivity insertion rates to verify that the limiting reactivity insertion rate, which results in a trip on the Overtemperature AT trip function, determined in the reference analysis (2) (2 pcm/sec) is still the limiting reactivity insertion rate. 15.4.2.5 Conclusion The MDNBR is greater than the XNB correlation safety limit. Therefore, the DNBR acceptance criterion is met.
ANF-88-094 15.4.2-4 l l Table 15.4.2-1 UNCONTROLLED CONTROL R00 ASSEMBLY WITHDRAWAL -
SUMMARY
OF INITIAL OPERATING CONDITIONS Parameter Value Power (MWt) 2346 Core Inlet Temperature ('F) 550.2 Pressurizer Pressure (psia) 2220 Reactor Coolant System Flow Rate 97.29 v (lbm/hr) l
ANF-88-094 15.4.2-5 s Table 15.4.2-2 UNCONTROLLED CONTROL R00 ASSEMBLY WITH0RAWAL EVENT SEQUENCE Event Time (sec) Uncontrolled RCCA Bank Withdrawal begins 0.0 Overtemperature AT Setpoint reached 26.5 ~ Scram Results in Rod Motion 27.2 Minimum DNBR occurs 27.4 . l \\
l l RCCA Bank Withdrawal, Pos. Feedback,2.0E-5 dk sec 3000 PL i 2500- - - - i--- - -i-
4--
i-- - -i---------- g $ 2000- - -- --- -+- - v 4 g s - -?- - - - - - - { 1500- -;- ----- - = - d 6 4 ) 1000-i-----i---- E* l = l 500-i-- --- i-------- --- = - - - ----- ~ ~ - 1 l l l l l U1." l 'T1 A e l O . co l 0 5 10 15 20 25 30 35 40 ?? cn o T1rrie, see o* l FIGURE 15.4.2-1 St.0W ROD WITilDRAWAL, 100% RATED POWER, RLACIOR POWER
RCCA Bank Withdrawal, Pos. Feedback,2.0E-5 dk sec 620 X i I TCIO i i i i _ _TCA _ _ .- W. 610- --- --------. . -------------r.--------------:.--------------:--------------.--------------:.--------- 600- ---- --- ----->--- ------ --- +------- ----
- t. ------------- +. ---- --------.>--------------<.------- -------l.-
b0 O
b--------------k.--------------:'--------------.5--------------+-------------b.-------------\\.-5-------------
590-O .s > s 1 i g. g 580- :====.==u=u----+-~"===_+-----------+------------.>-----------'g-4.--------------+.---- y g. \\ g. l .N O .\\ . g . s 570 - -------------.--------------.- ------------ :.------------ --:.--------- ----.------- ------r-- v ------ --:-------- ---- s N 560- - ----------- + ------ ---- -<----- ------- + --------- --- + ---- --------->------ -------<----- -------i- ----------- e.-* 2> - / 4 e mz - m - m 550 su, m, i i i e i i i 0 5 10 15 20 25 30 35 40 No e Time, see FIGURE 15.4.2-2 SLOW R0D WITilDRAWAL,100% RATED POWER, PRIMARY COOLANT CORE INLET, AVERAGE, COLD AND ll0T LEG TEMPERATURES \\ b T
RCCA Bank Withdrawal, Pos. Feedback,2.0E-5 dk see 2300 2250- -- ---- - --- ? ------ --- - +- - ---- ---~~
l s*"*%
?-- ------- --?. --- - ---- T. - ----- --- .d,4 M V O< 2200- --------------. q) g, U3 03 q) k gg5o.............. r............. :..............:..............................:.........................:.............. A L .N q) L U 2100- --- ----- - + - -- --------<- ---
l.--- ---------- +. --~~--------- +.--------------;.--------
l.
g3 M G) L A l 2050- --------------r--------------+-------- -----* -- ------- -- ' ---- ---------:--------------+-------- -- --=.- 2 m. n ae 2000 - m a ru co 0 5 10 15 20 25 30 35 40 e i co o Time, sec FIGURE 15.4.2-3 SLOW ROD WIIllDRAWAL,100% RATED POWER, PRESSURIZER PRESSURE I 1 1 1
\\ l RCCA Bank Withdrawal, Pos. Feedback,2.0E-5 dk sec 850 .i i CFWPR e.>,5 - \\ m. w aco. ............................<..............s................>........... p O v 775_.................................................. o p .O yso..............,. ~ l -o l 725. - - - - - - - - - - - - - - - - - - - - - -. - - - - - - - - - - - -. - - - - - - - - - - - - - -. - - - - - - - - - - - - - - - - - - - - - - - - - -. - - - - - - - - - -.---------- --- i \\ .er a 700- -------- ----*- --
t.-------------
o s., .N 4 675-
-- -- -;----- -- ----- =. -- --------- ~~ -------------~; ------- ------;---- --------- *. --------- - --;.--------------
a y3 m o es0-3..............:...............;............................ 3............... 4 fl. 625-- --- --------.>------ ------ +. ------- ---- +. --------------; ---------- --.>--- --------- +. ------ ----- +. -- --------- 600- ~>
- m. z
.u. m mm 0 5 10 15 20 25 30 35 40 a6 = D Time, sec a flGURL 15.4.2-4 SLOW R0D WilllDRAWAL, 100% RATED POWER, PRESSURIZER LIQUID VOLUME cy 's m A
RCCA Bank Withdrawal, Pos. Feedback,2.0E-5 dk/see 1200 PD01 1150-i-- - - - - - - - ]- - i-. 3 1100-- 5- - - m O< .i v k 1050- -f----- - - l -- --- 3 ~ m M o L. 1000- =- - = - ----- - - - - i ---- - -- -- 04 U bo 950~ '?' -'? *-** ca (7 i i i yi 900- --i- ----- -i-4- - -i- - - - = 850-i- 5-
f---
~> ui z T1 aa 800- - m co ca 0 5 10 15 20 25 30 35 40 s : -o Time, see o'e l l l l fIGllRL 15.4.2-5 SI OW ROD WIIllDRAWAl.,100% RATED POWER, i SILAM GEf1LRATOR DOME PRESSilRE l \\
1 RCCA Bank Withdrawal, Pos. Feedback,2.0E-5 dk sec 3 p -t-2- --- ---- ---.
7.------------.-----
_ _Dh,}illE3_ _ _U.r- - - h- - ^ 1_.............;..........................g...............g...............>.............<..... \\ l i i O ., M. ~ n* 4
- W.
---~t.--- i d O i .s-e l I I g g...............>.........................;.............. 3..............>........... ..t.............. I l ss &,4 i N \\ _2 .4 p c} d, g i 1
.>-----------+.-------------l.--------------.s------------.>----------~+.----------+.------------
_4_....................................................... . --------------:r----- ------. ---- -------- 4 -- --- ----- ?. -- ------ --.>--- -- ---- -i-------------- ;-- ---- - -- 3 m =- a,n . o3 0 5 10 15 20 25 30 35 40 '? ? -0 T1I110, Sec -eu FIGURE 15.4.2-6 SLOW R0D WlIllDRAWAi.,100% RATED POWER, REACTIVITY ADDITIONS \\ O s .*9. y 4 M A emdU i * ' 5 _-
I l ANF-88-094 15.4.3-1 15.4.3 CONTROL R00 MIS 0PERATION (SYSTEM MALFUNCTION OR OPERATOR ERROR) The analysis addresses the limiting rod drop transient resulting in reactor trip on the Overtemperature AT reactor trip. The limiting case assumes manual rod control, a non-positive moderator temperature coefficient above 50% power, and active turbine runback. The event is analyzed to verify the OTAT trip function with the new RTD installation which eliminates the l bypass piping. 15.4.3.1 Identification of Causes and Event Descriotion The event is defined to be initiated by a dropped RCCA. The dropped RCCA l promptly inserts negative reactivity which reduces reactor power and disturbs the power distribution, resulting in increased local power peaking. The rod bottom and negative flux rate signals can independently initiate turbine runback to 70% of full power. With turbine runback, the reduction in load initially results in a load mismatch if the dropped rod reactivity does not match that required for the runback power level. If reactivity insertion is greater than that required to match the runback power level, T initially avg decreases. If reactivity is less, T initially increases. The reactor avg protection system will limit consequences should conditions approach setpoint values. This event is classified as a Condition II event. The acceptance criteria for th;s event is that the MDNBR is greater than 1.17. 15.4.3.2 Analysis Method The analyses are performed by coupling a conservative power peak to transient response and DNBR calculations. The power peak associated with each event is characterized through an augmentation factor which relates the maximum power ceak to the steady state power peak. The steady state power distributions and augmentation factors are calculated with the XTGPWR reactor
4 ANF-88-094 15.4.3-2 simulator.I9) Standard neutronic methodology is used to calculate neutronics parameters such as control rod worth and power peaking. The system response to a single dropped RCCA is analyzed with the N) PTSPWR2 code. The DNB analysis is performed t.si ng the XCOBRA methodology (3) using the operating conditions from the PTS calculation. Local power redistribution effects due to the dropped rod are input to the XCOBRA methodology by a local power augmentation factor. The Technical Specification value of the allowed F is multiplied by this augmentation 3g factor. 15.4.3.3 Definition of Events Analyzed and Boundina Inout For control rod misoperation events, the maximization of power peaking results in a reduction in the DNBR. To assure that bounding values are determined for the radial power peaking, the following approach is used. The increase in power peaking above that associated with equilibrium steady state conditions is determined for a spectrum of cycle nposures and applicable control rod configurations. based on these
- results, a
conservative au3 mentation factor is derived. The limiting Condition II event analyzed is a drcpped full length RCCA of low worth with turbine runback. No single failure assumption is required since manual rod control is assumed. 15.4.3.4 Analysis of Results The event initiates with a step negative reactivity insertion repre-senting a single minimum ' worth dropped rod. At event initiation, a turbine runback signal is assumed. Turbine load demand reaches its programmed value at 9 sec. Average coolant temperature at first decreases in response to the power reduction caused by the dropped rod, but later increases due to the reduced secondary load demand. The temperature increase causes insurge to the pressurizer, resulting in a pressure increase sufficient to open the PORVs at
ANF-88-094 15.4.3-3 16.1 sec. Reactor scram on Overtemperature AT occurs at 61.1 seconds. The reactor power continues to decrease after the rod is dropped. The MDNBR as calculated by the XCOBRA methodology is 1.23, greater than the 1.17 XNB DNBR limit. The minimum DNBR occurs at 61.2 seconds. A summary of initial conditions is presented in Table 15.4.3-1. The sequence of events for the limiting case is presented in Table 15.4.3-2. 15.4.3.5 Conclusion The minimum DNBR is greater than the XNB correlation safety limit of 1.17. Therefore, the event acceptance criterion on DNBR is met. I
ANF-88-094 15.4.3-4 Table 15.4.3-1 DROPPED FULL LENGTH RCCA (MANUAL) -
SUMMARY
OF INITIAL CONDITIONS Condition Value Power, MWt 2346 Core Inlet Temperature, 'F 550.2 Pressurizer Pressure, psia 2220 6 Reactor Coolant System Flow Rato, Ib/hr 97.29 x 10 l
ANF-88-094 15.4.3-5 Table 15.4.3-2 DROPPED FULL LENGTH RCCA (HANUAL) EVENT SEQUENCE With Turbine Runback Event Time, sec Dropped RCCA Fully in 0.0 Turbine Runback Begins 0.0 Turbine Runback Reaches Low Load Limit 9.0 Pressurizer PORVs Open 16.1 Peak Pressurizer Pressure 17.0 Reactor Scram and Rods begin to Fall 61.1 (0vertemperature AT) Peak Core Power Level 0.0 Steam Generator Safety Valves Open 57.5 Minimum DNBR Occurs 61.2
RCCA Drop, 0.0 MTC, no ARC, w Turbine Runback 2500 2250-...................................................................................................... 200C- -------------. ---- -------t.----- -- - -- e.- - -- - - - +. - ~ ~ ~ --.-------------. ------------ n. - ----------- 1750_..................................................................................................................... g, D
- F 1500- ----------+--------------t.-----------t.-------------t--------------+-----~~-t.---
- ~ ~
t.---------------
O fl ~d 1250- - - - --- - -.------------- ~ - - ----------.-- - ------ - - ~ ~ -- - ---- - - --- - -. -- - ~ ~~ ~ --.---------- --- p D 1000- -------------=-- ~ ~ ~-- ~ t.----------
- ---------------t-------------- *-- - ----- - -t. - -----------:.- - ------ - -
4 .Le E--* 750- ----------- ~.;------ - - - - o. -- ------- ~ ;.- - --- - -----;-------------- 4. ------------4. 500-3.----- 5----------- - - ---------------: - - --- - - 3.----- -
t---- ----------:----------- --
250- ---- - -------? - - - - - --- ?. --- - - - ---i. - -------- - -i.-------- ~ ~ -? ------- - - - t. ------- -- n_. m~ wn 0 . m i e i i e i i wm l 0 10 20 30 40 50 60 70 80 &6 e T1me, sec a FIGURE 15.4.3-1 RCCA DROP, REAC10R POWLR
l RCCA Drop, 0.0 MTC, no ARC, w. Thrbine Runback 610 -- TCA s g g 600-
?-------------- ?. ------ ------ *. --------------t---------------?.*^------ --- *-- r ------ T. -------------
g g g m t l s [x, 5 9 0 - ------------ :. s 1 CnD .-------------:------------>:*--------------:-------------:-------------:---------v----:-------------- 1 O s I v , r. s,. ..s O k M 580- ?*--~==ui."'----------?.-------------i.------------+.-----------+.-----------t.--------M.----------- p g3 h Q. O O 570 - ------------- + -------------t.---------------l.--- -----------5.-----
+--------------t.---------------l--------------
E-* t 560- ------------->--------------+---------:..-------------.t--------------->-------------'.--------------:.--------------- 1 1 .u, 2 3,n 3 4 5 I I I 5 CD 0 10 20 30 40 50 60 70 80 YP ~o Time, see u> u FIGURL 15.4.3-2 RCCA DROP, CORE AVERAGE AND INLET C00LAHI TEMPERATURES
RCCA Drop, 0.0 MTC, no' ARC, w Turbine Runback 2350 2300_........................................ ^ 2250-
+ --- -------<--- ---- ---i- ------- --.l -------------->--- ---------- +------------- l.---
.a3 g3 2200_.............................,............................... o g., d p1 on 2150- ------------+------------+.------------4.---------------l-------------.>--------------4.--- ---------t.--------------- o Q A ..4 1 .s o 4 k U 2050-h------- ----- + -------------i.---------------l--------------.>--------------4.
.--t.--------------
g) n2 i g se 4 2000_............................................................... 1g50_...............>............s...............;...............s...............>...........+........... z................ O's _,.. a,i . m 1900 ym 0 10 20 30 40 50 60 70 80 moe Time, sec 4 4 FIGURE 15.1.3-3 RCCA DROP, PRESSURIZEh PRESSURE t}}