ML20214K165

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Forwards Addl Info on SER Backfit Issue 4 Re Steam Generator Level Control & Protection,In Response to 860630 Request. Draft marked-up FSAR Changes Also Encl
ML20214K165
Person / Time
Site: Beaver Valley
Issue date: 08/12/1986
From: Carey J
DUQUESNE LIGHT CO.
To: Harold Denton, Tam P
Office of Nuclear Reactor Regulation
References
2NRC-6-087, 2NRC-6-87, NUDOCS 8608180178
Download: ML20214K165 (20)
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{{#Wiki_filter:* 'Af 2NRC-6-087 Be Vall No. 2 Unit Project Organization Telecopy ( 3-Ext.160 P.O. Box 328 August 12, 1986 Shippingport, PA 15077 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation United States Nuclear Regulatory Comission Washington, DC 20555 ATTENTION: Mr. Peter Tam, Project Manager Division of PWR Licensing - A Office of Nuclear Reactor Regulation

SUBJECT:

Beaver Valley Power Station - Unit No. 2 Docket No. 50-412 Beaver Valley Unit 2 - SER Backfit Issue 4 Steam Generator Level Control and Protection Gentlemen: In a letter dated June 30, 1986, the Nuclear Regulatory Commission requested additional infonnation on Backfit Issue 4 for Beaver Valley Unit 2. Duquesne Light Company responses are provided in Attachment 1. DUQUESNE LIGHT COMPANY By J. J. Carey Vice President RWF/ijr Attachment NR/ STEAM / GEN cc: Mr. P. Tam, NRC Project Manager - w/ attachment Mr. L. Prividy, NRC Resident Inspector - w/ attachment 0608100178 960812 d[ a PDR ADOCK 05000412 E PDR \\

O - Request for Additional Information NRC Request: 1. Table 15.0-4 includes the high stean generator level as a trip function assuned in the accident analyses. This should be renoved from this table and the appropriate accidents reanalyzed. 2. In Section 15.0.8 on page 15.0-11 the following statement is made: .the nonnally operating systems and camponents listed in Table 15.0-6 will be available for mitigation of the events dis-cussed in Chapter 15". Table 15.0-6 includes the high stean generator level as available for transient and accident conditions related to a feedwater system malfunction causing an increase in feedwater flow. The high stean generator level actuation should be completely renoved from Table 15.0-6 or it should be footnoted that no credit is taken for this trip in the accident analysis. If credit is taken, these accidents should be reanalyzed without this trip. 3. Section 15.1.2, "Feedwater Systen Malfunctions Causing an Increase in Feedwater Flow," contains statenents such as: " Continuous ad-dition of excessive feedwater is prevented by the stean genertor hi-hi level trip. " and "the feedwater resulting from a fully open control valve is tenninated by a stean generator hi-hi level trip signal. Statenents referring to the stean generator l hi-hi lcvel trip signal should be renoved from the FSAR, and an-alysis perfonned unoer Section 15.1.2.2, which currrently assunes that the high stean generator level signal tenninates this trans-j' ient, should be replaced by a new analysis that accurately de-scribes the assumptions used including those relating to tennination of the transient. The analysis should state explicitly l which safety-related equipment is taken credit for in the i mitigation of this transient. If tennination of the transient is not required for safety, this should be so stated and bases pro-vided. Also, the time sequence of events contained in Table 15.1-1 and all related FSAR Figures must also be revised accordingly for the new analysis.

Response

As per References (a), (b) and (c), Duquesne Light Company (DLC) explained that the high steam generator water level function was used in the excess Feeowater Event analysis in FSAR Section 15.1.2, but it was used only as a convenient (and historically typical) ending point for the Westinghouse analyses. No credit was taken for the high steam generator water level systen to perfonn a safety-related function. As explained in the Backfit meeting on May 9,1985 (Reference c), the point of minimum DNBR occurs well before a high stean generator water level occurs. Thus, even if another analysis was performed without this function, the conclusions currently shown in FSAR Section 15.1.2.4 would renain unchanged. The high stean generator level function was originally included in the Chapter 15 analyses based on the precedent set up by Westinghouse in the past. FSARs for Westinghouse-NSSS plants typically included this function (including the Beaver Valley Unit 1 FSAR) even though it is a non-safety related systen used for turbine protection. This was acceptable to the NRC for over 25 other Westinghouse-NSSS plants. This does not include the FSARs for the Westinghouse-NSSS plants which do not contain _a_ny automatic n high stean generator water level functions. No commitment was made by DLC nor was any conmitment recanmended by NRC Management at the May 9 Backfit Meeting (Reference c) to reanalyze the Excess Feedwater Event already provided in FSAR Section 15.1.2 without the high stean generator water level function since Westinghouse stated that no credit was taken for this function performing any safety-related action and hence no conclusions would change. DLC only conmitted to correct Table 15.0-6 to correctly show the high stean generator water level function as non-ESF, as per Reference (a) and as per the Backfit Meeting. Thus, as ;>er Reference (b), DLC let the high stean generator water level function renain in FSAR Tables 15.0-4 and 15.0-6 since it was discussed in the analyses, even though no credit for its use was assuned. However, in accordance with your new request, Tables 15.0-4 and 15.0-6 will be footnoted in the next Beaver Valley Unit 2 FSAR amendment to show its non-safety significance. A paragraph will also be added in the next FSAR anendment to Section 15.1.2 pointing out that the high stean generator water level function is non-safety related and its use (or potential lack of use) in the analyses does not alter any conclusions provided in Section 15.1.2.4. Draft narked-up FSAR changes are attached

Reference:

(a) DLC letter 2NRC-5-052, dated March 27, 1985 (b) DLC letter 2NRC-5-150, dated Decenber 20, 1985 (c) NRC letter, dated June 25,1985, " Meeting Summary and Transcript"

NRC Request 4. Section 7.3.1.1 provides the systen description for ESFAS and refers (see page 7.3-2) to Table 7.3-1 and 7.3-2 for additional infonnation pertaining to ESFAS logic and function. In Table 7.3-2 under feedwater isolation, the logic for stean generator high-high water level is discussed. This functional unit should be renoved from Table 7.3-2. 5. Section 7.3.1.1 also refers (see page 7.3-2) to Table 7.3-3 for interlocks associated with ESFAS. In Table 7.3-3 high stean generator level is included as "P-14". This interlock should be renoved from this table. 6. A functional description of the main feedwater isolation on stean generator high-high water level should be included in the appropriate Chapter 7 section of the FSAR following the guidance of Standard Review Plan, NUREG-0800, and Regulatory Guide 1.70.

Response

Westinghouse provided the original infonnation for Tables 7.3-2 and 7.3-3 in the Beaver Valley Unit 2 FSAR. Westinghouse typically included the high stean generator water level function even though it is not considered to perfonn any ESFAS function nor any safety related function (also shown in the Beaver Valley Unit 1 FSAR). Reference to the high stean generator water level function will be renoved from 7.3-2 and 7.3-3 and a short description will be added to Section 7.7 in the next FSAR anendment in accordance with your request. Draft marked-up FSAR changes are attached. 4

'BVPS-2 FSAR TABLE 7.3-2 INSTRUt!ENT OPERATING CC!!DITIONS FOR ISOLATION FUNCTIONS No. of Channels Functional Unit Channels Needed to Trip Containment Isolation

1. Automatic safety injection (Phase A)
a. Containment pressure (Hi-1) 3 2
b. Low compensated steam line 3/ steam 2/ steam pressure line***

line*** (lead-lag coavensated) any steam line

c. Pressurizer lov pressure
  • 3 2

2

2. Containment pressure (Phase B) a.

1.' l - 3 4 2

3. Manut.1
a. Phase A 2

1

b. Phase B**

4 2 Steam Line Isolation l

1. High steam pressure rate 3/ steam 2/ steam line line any eteam line
2. Containment pressure (Ill-2) 3 2
3. Low steam line pressure 3/ steam 2/ steam line***

line*** any steam line

4. Manual 1 loop ****

1/ loop Feedwater Line Isolation

1. Safety injection
a. !!anual 2

1

b. Containment pressure (111-1) 3 2
c. Low compensated steam line 3/ steam 2/ steam line***

pressure line*** any steam line (lead-lag compensated)

d. Pressurizer low pressure
  • 3 2

Amendment 4 1 of 2 December 1983

w-BVPS-2 FSAR TABLE 7.3-2 (Cont) flo. of Channels Functional Unit Channels fleeded to Trip \\ enerator high-high At " water leve

a. 2/3 on a 2/ loop

.rator (P-14) !!OTES :

  • Permissible bypass if reactor coolant pressure is less than 2,000 psig.
    • lianual actuation of containment spray is accomplished by actuating either of two sets (two switches per set). Both switches in a set must be actuated to obtain a manut.lly-initiated containment depressurization signal per train.
      • Signal is automatically blocked from isolated loops.
        • Additionally there will be two sets of control devices (two momentary controls per set) on the main control board.

Operating either set will actuate all three main steam line stop and bypass valves at the system level. T 2 of 2

BVPS-2 FSAR TABLE 7.3-3 INTERLOCKS FOR ENGINEERED SAFETY FEATURES ACTUATION SYSTEM Designation Input Function Performed P-4 Reactor tripped Presence of P-4 signal actuates turbine trip Presence of P-4 signal allows manual reset / block of the automatic reactuation of safety injection Absence of P-4 signal defeats the manual re-set / block preventing /' automatic reactuation of safety injection ~ Presence of P-4 signal closes main feedwater valves on T below setpe$nE. Pres-cnce of P-4 signal pre-vents opening of main feedwater valves which were closed by safety injection high-high steam generator water level P-11 2/3 pressurizer pressure Allows manual block of below setpoint (Presence safety injection on low signal permits functions pressurizer pressure shown. Absence of signal signal defeats functions shown) Allows manual block of safety injection actu-ation on low compensated steamline pressure signal Permits steamline iso-lation via high steam pressure rate if low pressure signal manu- [ nlly blocked 1 of 2

7 t 1 j i-BVPS-2 FSAR TABLE 7.3-3 (Cont) Designation Input Function Performed Blocks opening of pressurizer power-operated relief valves P-12 2/3 T below set-P-11 pointhresenceofP-12 signal performs or per-Blocks steam dump except mits functions shown. for cooldown condenser Absence of signal de-dump valves feats function shown) 38 2/3 steam generator level Allows manual by of g above setpoint on any steam steam water p block erator (Presence of for ooldown valves sign rforms or permits y functier,s s Closes all feedwater isolation valves Trip water pumps Actuates turbine / 2 of 2

BVPS-2 FSAR - information on this system, see the summary report titled Westinghouse Reactor Vessel Level Instrumentation System for tionitoring Inadequate Core Cooling (Westinghouse 1980). Reactor vessel level is also utilized to indicate the need to vent non-condensible gases from the reactor vessel head. The RVLIS uses two sets of differential pressure cells to measure rea.ctor vessel level. The narrow range RVLIS instrument provides an indication of reactor vessel water level from the bottom of the reactor vessel to the top of the reactor vessel when zero or one reactor coolant pump (RCP) is operating. The narrow range instrument also measures the reactor core and internals pressure drop, and therefore provides an indication of the relative void content or density of the circulating fluid when only one RCP is operating. When more than one RCP is operating, the narrow range instrument reading will be off scale. The wide range RVLIS instrument provides an indication of reactor core, internals, and outlet nozzle pressure drops for any combination of operating RCPs. Comparison of the measured pressure drop with the normal, single-phase pressure drop provides an approximate indication of the relative void content or density of the circulating fluid. The wide range instrument monitors vessel level on a continuous basis. The upper-range RVLIS instrument provides an indication of reac vessel water level from the hot leg pipe to the top of the re vessel when the RCP in the loop uith the hot icg connecti operating. The measurement provides an accurate indica for guidance when operating the reactor vessel head vent. When t e pump in the loop with the hot leg connection is operating, the differential pressure would be greater than the transmitter span, and the transmitter output would be disregarded. 7.7.2.12 Plant Safety !!onitoring System The plant safety monitoring system (PS!!S) is used to process and output the inadequate core cooling (ICC) variables in proper format to internal plasma displays, and external indicators, displays, cabinets and other equipment. The PS!!S consists of three types of modular components: the remote processing unit (RPU), the display processing unit (DPU), and the plasma display. These components perform the data acquisition and processing, the data base consolidation and comparison, and the data selection and display, respectively. The system is seismically and environmentally qualified, is configured to address single-failure criteria, and qualification details are availabic in Section 3.10 and 3.11. In addition, the PSils has the capability for on-line testing without affecting reactor protection and control. Amendment 2 7.7-20 July 1983

BVPS-2 FSAR The plasma display modules are redundant, qualified, graphic / alphanumeric modules for displaying reactor vessel level core cooling margin (T saturation ), and the core exit thermocounles on demand. These displays will be used to detect the approach to inadequate core cooling. Sections 3.10 and 3.11 provided details of the seismic and A.dd environmental qualification.

  1. (

Y 7.7.3 References for Section 7.7 he (C Lipchak, J.B. and Stokes, R.A. 1974. Pluclear Instrumentation System. WCAP-8255 (for background information only). Shopsky, W.E. 1977. Failure flodes and Effects Analysis of the Solid State Full Length Rod Control System. WCAP-8976. U.S. Department of De fense 1982. Reliability Prediction of Electronic Equipment. !!IL-ilDBK-217D. Westinghouse 1980. Westinghouse Reactor Vessel Level Instrumentation System for !!onitoring Inadequate Core Cooling. December 1980. T Amendment 2 7.1-28a July 1983

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BVPS-2 FSAR TABLE 7.7-1 BVPS-2 CONTROL SYSTEM INTERLOCKS Designation Derivation Function C-1 1/2 Neutron flux Blocks automatic and (intermediate range) above manual control rod set point withdrawal C-2 1/4 Neutron flux (power Blocks automatic and range) above set point manual control rod withdrawal C-3 2/3 Overtemperature AT Blocks automatic and above set point manual control rod withdrawal Actuates turbine runback via load reference C-4 2/3 Overpower AT Blocks automatic and above set point manual control rod withdrawal Actuates turbine runback via load reference C-5 1/1 Turbine impulse chamber Blocks automatic control pressure below set point rod withdrawal C-7 1/1 Time derivative !!akes steam dump valves (absolute value) of available for either turbine impulse chamber tripping or modulation pressure (decrease only) above set point P-4 Reactor trip breakers open Blocks steam dump control via load rejection Tavg controller llakes half of the steam dump valves available for either tripping or modulation The following condition Blocks steam dump control exists when P-4 is via reactor trip Tnyg not active controller (this func-tion is provided by absence of P-4) 1 of 2

BVPS-2 FSAR TABLE 7.7-1 (Cont) Designation Derivation Function C-9 Any condenser pressure Blocks steam dump to above set point, or condenser all circulation water pump breakers open C-11 1/1 Bank D control rod Blocks automatic rod position above set point withdrawt.1 P'IA 2/3 Steen peut mkv leveI C lc 5t s en il [o rd w<kY c&c,g se4p :,,o en u y slenm (0'< A Vo I Ve 5 yt>4m4cr(l'te,1C+1Ce#f Tr6 h [cc<dwrItr p.yJr f a, y& qu oder,,, s e PC '"

  • IS 4 ' +"4 5 N. < wL. se cm.)

6 6ta 9 k 2 of 2

BVPS-2 FSAR TABLE 15.0-4 TRIP FOINTS A!!D TI!!E DELAYS TO TRIP ASSUllED IN ACCIDE!!T A!!ALYSES Limiting Trip Point Assumed Time Delays Trip Function in Analysis (sec) Power range high neutron 118 percent 0.5 flux, high setting Power range high neutron 35 percent 0.5 flux, low setting High neutron flux, P-8 79 percent 0.5 Overtemperature and over-Variable, (Figures 6.0* power AT 15.0-1 and 15.0-la) High pressurizer pressure 2,410 psig 2.0 i l Low pressurizer pressure 1.920 psig 2.0 Low reactor coolant flow 87 percent loop flow 1.0 (from loop flow detec-l tors l RCP underfrequency 57 Hz 0.9 l l Lo-lo steam generator level O percent of narrow 2.0 l range level span l High steam generator leve1## 80.0 percent of 2.0 narrow produces feedwater isulation range level span and turbine trip il0TE :

  • Total time delay (including RID time response and trip circuit channel electronics delay) from the time the temperature in the coolant loops exceed the trip set point until the rods are free to fall.

1 Nk ferG.ws sufe ty - r els k e f u sis k no Amendment 11 1 of 1 January 1986 l

BVPS-2 FSAR TABLE 15.0-6 PLANT SYSTEMS AND EQUIPMENT AVAltABLE FOR TRANSIENT AND ACCIDEtti CONDITIONS Cther Actuation g Incide_n3 Reactor Trio Ftmetions ESF Actuation Fametions femc tions an<f E<2uioment ESF Eauioment i 15.1 Increa se in heat removed by the secon-ca ry sys tem feedwater sys-Power range hign flux, NA Feedwater isolation NA tem malfunc-turbane trip-reactor valves, trip of tion causing trip, manual, OTAT turbine from high an increase steam generator in feedwater levelF flow Excessive in-Power rant,e high flux. NA NA NA crease secon-manual, OTAT, OPaT dary steam flow Accidental de-Low pressurizer pressure, Low pressurizer pressure, Feedwater isolation Auxiliary feed p ressuriza tion manual, SIS low compensated steam line valves, steamline system; safety of the main pressure Hi-1 containment isolation valves (HI-2 injection system steam system pressure, s anua l containment pressure) Steam system SIS, low pressuri'.er pres-Low pressurizer pressure, Feecwater isolation Aux i l ia ry feed piping failure sure, manual low compensated steamline valves, steamline system; safety pressure, Ni-1 containment isolation valves (HI-2 injection system; pressure, manual containment pressure) Containment heat removal system (HI-3 containment pressure) 15.2 Decrease in heat removal by the secon-dary system Loss of ex-High pressurizer pressure, Pressurizer safety Auxilia ry feed ternal elec-C' T. steam generator valves, steam gen-system trical load / lo-lo level, manual erator safety valves turbine trip Amendment 11 1 of 4 / January 1986 h r Puru.,, 9-nas Fa.u ,,e

BVPS-2 FSAR 15.1.1.3 Radiological Consequences There are no radiological consequences associated with a decrease in feedwater temperature event. The activity is contained within the fuel rods and the-RCS radionuclide concentrations remain within Technical Specification limits. 15.1.1.4 Conclusions The decrease in feedwater temperature transient is less severe than the increase in secondary steam flow event (Section 15.1.3). Based on results presented in Section 15.1.3, the applicable acceptance criteria for the decrease in feedwater temperature event have been met. There are no radiological consequences for this event. 15.1.2 Feedwater System !!alfunctions Causing an Increase in Feedwater Flow 15.1.2.1 Identification of Causes and Accident Description Additions of excessive feedwater will cause an increase in core power by decreasing reactor coolant temperature. Such transients are attenuated by the thermal capacity of the secondary plant and of the RCS. The high neutron flux trip, overtemperature AT trip, and overpower AT trip prevent any power increase which could lead to a DNBR less than the limit value. An exemple of encemmive feeddales flow would be a full opening of a feedwater control valve due to a feedwater control system malfunction or an operator error. At power, this excess flow causes a greater load demand on the RCS due to increased subcooling in the steam generator. With the plant at no-load conditions, the addition of an excess of feedwater may cause a decrease in RCS temperature and,

thus, a reactivity insertion due to the effects of the negative moderator coefficient of reactivity.

m ee ,,,,.. xM3 reldJ Continuous addition of excessive feedwater )( prevented by the steam generator hi-hi level trip, which closes all feedwater contro1 and isolation valves, trips the main feedwater pumps, and trips the main turbine. An increase in normal feedvater flow is classified as an ANS Condition II event, a fault of moderate frequency (Section 15.0.1). Plant systems and equipment which are available to mitigate the effects of the accident are discussed in Section 15.0.8 and listed in Table 15.0-6. 9 15.1-3

BVPS-2 FSAR 15.1.2.2 Analysis of Effects and Consequences Method of Analysis The excessive heat removal due to a feedwater system malfunction transient is analyzed by using the detailed digital computer code LOFTRAN (Burnett 1972). This code situlates a multi-loop system, the neutron kinetics, the pressurizer, pressurizer relief and safety valves, pressurizer spray, steam generator, and steam generator safety valves. The code computes pertinent plant variables including temperatures, pressures, and power level. A control system malfunction or operator error is assumed to cause a feedwater control valve to open fully. Two cases are analyzed as follows: 1. Accidental opening of one feedwater control valve with the reactor just critical at zero load conditions assuming a conservatively large negative moderator temperature coefficient. 2. Accidental opening of one feedwater control valve with the reactor in automatic control at full power. Both of these cases are analyzed for operation with three loops in service and for operation with two loops in service. The reactivity insertion rate following a feedwater malfunction is calculated with the following assumptions: 9@$ 1 1. For the feedwater control valve accident at full power, one feedwater control valve is assumed to malfunction resulting in a step increase to 160 percent of nominal feedwater flow to one steam generator (220 percent for N-1 loop operation). 2. For the feedwater control valve accident at zero load conditions, a feedwater control valve malfunction occurs which results in an increase in flow to one steam generator from zero to 170 percent of the nominal full load value. 3. For the zero load condition, feedwater temperature is at a conservatively low value of 32'F. 4. No credit is taken for the heat capacity of the RCS and steam generator thick metal in attenuating the resulting plant cooldown. 5. The feedwater flow resulting from a fully open control valve is terminated by a steam generator hi-hi level trip signal which closes all feedwater control and isolation valves, trips the main feedwater pumps, and trips the turbine. 15.1-4 I

BVPS-2 FSAR Initial operating conditions are assumed at values consistent with steady-state N and N-1 loop operation. Plant characteristics and initial conditions are further discussed in Section 15.0.3. Normal reactor control systems and engineered safety feature (ESF) systems are not required to function. The reactor protection system (RPS) will function to trip the reactor due to overpower or high steam generator water level conditions. Ho single active. failure will prevent operation of the RPS. A discussion of anticipated transients without trip (ATWT) considerations is presented by Westinghouse Electric Corporation (1974). Results The calculated sequence of events for this accident are shown in Table 15.1-1. In the case of an accidental full opening of one feedwater control valve with the reactor at zero power and the preceding assumptions, the maximum reactivity insertion rate is less than the maximum reactivity insertion rate analyzed in Section 15.4.1 and therefore, the results of the analysis are not presented here. It should be noted that if the incident occurs with the unit just critical at no-load conditions, the reactor may be tripped by the power range high neutron flux trip (low setting) set at approximately 25 percent of nominal full power. The full power case (with rod control) gives the largest reactivity feedback and results in the greatest power increase. Assuming the reactor to be in the manual control mode results in a slightly less 1 severe transient. The rod control system is not required to function 46e a cw Adt 4 ( k' k 'l '5 " * "" for an excessive feedwater flow event. 7 % p n W m L) ^I n d M

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When the steam generator water level in tLe faulted loop reaches the hi-hilevelsetpoint,allf,eepy[tercontrylandisolationvalvesand the main feedwater pumps anau tripped. This prevents continuous addition of feedwater,--I., adjition,.a reactor trip and turbine

trip,

,m $7,gsg g <M wi h eM <, Transient results (Figures 15.1-1 and 15.1-2) show the core heat flux, pressurizer pressure, T avg and DNBR, as well as the increase in nuclear power and loop AT associated with the increased thermal load on the reactor. The DNBR ter 7;t :'-^n h.1nu W I M W 1ue. Ebpwes--l+rl-la-end-101 ?= shou the trJnnient results-with-two-loops in-eperatien.*%4 a4W here Following the reactor trip and feedwater isolation, Beaver Valley Power Station - Unit 2 (DVPS-2) will approach a stabili:ed condition at hot standby. Normal plant operating procedures may then be followed. The operating procedures would call for operator action to 15.1-5

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s i wp M # BVPS-2 FSAR 4eo control RCS boron concentration and pressu izer level using the chemical and volume control system and to mai tain steam generator level Anyactionrequiredoftheoperator,g/omaintainBVPS-2in through control of the main or au iliary feedwater system (AFWS). t a stabilizedconditionpl be in a time frame in excess of ten minutes. fell:rir.; n ::ter 4 f.ite.,,% ree tv-;p Since the power level rises during the excessive feedwater flow incident, the fuel temperatures will also rise until after reactor trip occurs. The core heat flux lags behind the neutron flux response due to the fuel rod thermal time constant, hence the peak value does not exceed 118 percent of its normal value (that=is, the assumed high neutron flux trip point). The peak fuel temperature will thus remain well below the fuel melting temperature. The transient results show that departure from nucleate boiling (DN8) does not occur at any time during the excessive feedwater flow incident;

thus, the ability of the primary coolant to remove heat from the fuel rod is not reduced.

The fuel cladding temperature, therefore, does not rise significantly above its initial value during the transient. 15.1.2.3 Radiological Consequences There are only minimal radiological consequences from feedwater system malfunctions causing an increase in feedwater flow. The reactor trip causes a turbine trip and heat is removed from the secondary system through the stea.n generator power relief or safety valves. Since no fuel damage it postulated to occur from this transient, the radiological consequences are less severe than those of the loss of non-emergency ac power to the station auxiliaries T analyzed in Section 15.2.6. 15.1.2.4 Conclusions The results of the analysis show that the DNBRs encountered for an excessive feedwater addition at power is at all times above the limit value; hence, the DNB design basis, as described in Section 4.4, is met. Additionally, it has been shown that the reactivity insertion rate which occurs at no-load conditions following excessive feedwater addition is less than the maximum value considered in the analysis of the rod withdrawal from a suberitical condition. The radiological consequences of this event are not limiting. 15.1.3 Excessive Increase in Secondary Steam Flow 15.1.3.1 Identification of Causes and Accident Description An excessive increase in secondary system steam flow (excessive load increase incident) is defined as a rapid increase in steam flow that 15.1-6}}

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