ML20003G673

Testimony on Behalf of Util on Doherty Contention 7 Re LPCI Cold Slug.Prof Qualifications Encl

ML20003G673
Person / Time
Site:	Allens Creek File:Houston Lighting and Power Company icon.png
Issue date:	04/20/1981
From:	Eckert E GENERAL ELECTRIC CO.
To:
Shared Package
ML20003G672	List: ML20003G671 ML20003G673 ML20003G674 ML20003G675 ML20003G676 ML20003G677 ML20003G678
References
NUDOCS 8104300476
Download: ML20003G673 (11)

Text

,

4-20-81 O

1 UNITED STATES OF AMERICA 4 # ut .

NUCLEAR REGULATORY COMMISSION -

$ ggg\F L ,

BEFORE THE ATOMIC SAFETY AND LICENSING BOA 4 In the Matter of )

)

5 HOUSTON LIGHTING & POWER COMPANY ) Docket No. 50-466

, )

6 (Allens Creek Nuclear Generating )

Station, Unit No. 1) ) .

7 )

8 9 TESTIMONY OF EUGENE C. ECKERT ON BEHALF OF HOUSTON LIGHTING & POWER CO. ON DOHERTY 10 CONTENTION 7 - LPCI COLD SLUG 11 Q. Please state your name and job position.

12 A. My name is Eugene C. Eckert. I am employed by General 13 Electric as Manager of NSSS Transient Performance Engineering.

14 My business address is 175 Curtner Avenue, San Jose, California.

is Q. Please state your educational background, work experience 16 and professional qualifications.

17 A. A statement of my qualifications is attached to this 18 testimony as Exhibit ECE-1.

19 Q. What is the purpose of your testimony?

20 A. The purpose of my testimony is to address Mr. Doherty's 21 Contention 7, which alleges that:

22 The design of obtaining Low Pressure Coolant Injection (LPCI) core spray water from the suppression 23 pool following exhaustion of the condensate storage tank during Loss of Coolant Accident (LOCA),

24 Reactivity Insertion Accident (RIA), or Transient Without Scram (ATWS) is an unnecessarily high risk 25 to Petitioner's safety and environment interests because suppression pool water is coldar than reactor 26 coolant; hence when sprayed in the core it will increase core reactivity causing high temperature 27 and increase possibility or actuality of fuel melt and formation of a critical mass.

l

^

810430047s _

t 1 2 Q. What is your understanding of the real concern expressed 3 in this contention?

4 A. My understanding of Mr. Doherty's concern is that 5 during either a Loss of Coolant Accident (LOCA), a Reactivity 6 Insertion Accident (RIA) , or an Anticipated Transient Without i

7 Scram (ATWS) the following are postulated to take i

8 place:

9 1. Reactor vessel pressure is reduced while the 10 High Pressure Core Spray (HPCS) sprays condensate 11 storage water onto the c' ore.

I 12 2. The reactor pressure will be reduced sufficiently I

13 to allow the low pressure Emergency Core Cooling 14 Systems (ECCS) to inject and/or spray on the core 15 cold water taken from the suppression pool. .

16 3. Core reactivity increases because the suppression l

17 pool water is colder than the reactor vessel water.

18 Reactor power, in turn, increases causing high 19 enough fuel temperatures to produce fuel melting.

20 0 As a preliminary matter, will you briefly describe 21 the Condensate Storage System?

22 A. The Condensate Storage and Transfer System (CSIS) mainly 23 functions as a storage and transfer medium for the reactor 24 ar.d turbine generator primary fluid during normal, abnormal 25 and shutdown operating conditions. During abnormal operating 26 conditions, which are of interest in this contention, the 27 system provides total reserve condensate capacity required 28 for operation of the Reactor Core Isolation Cooling (RCIC)

1 2 system and the HPCS system and to provide adequate suction l 3 pressure for the RCIC, HPCS, and Control Rod Drive (CRD) 4 pumps. The system is designed to operate in the temperature 5 range of 40'F to 120*F. The temperature of the water in the 6 system is chiefly determined by outside ambient temperature.

7 o. Will you also describe the function of the Suppression 8 Pool System during abnorma) accident conditions?

9 A. The suppression pool contains a sufficient supply of 10 water to satisfy the cooling water requirements of the low 11 pressure ECC systems and the Residual Heat Removal System F

12 (RHR) during all modes of operations. The suppression pool 13 also serves as a heat sink for steam release from the Nuclear 14 Pressure Relief System (NPRS) . The in-place volume of water 15 in the suppression pool is designed to provide the short 16 term energy sink for a LOCA.

17 Under normal plant operating conditions, the suppression 18 pool water temperature will be in the range of 90'F, approxi-19 mately the same as the containment vessel atmospheric tempera-20 ture. During a LOCA, the pool water temperature may rise to 21 170*F shortly af ter the initial release of energy to the 22 pool from a postulated large circumferential break of a line l

23 inside the drywell. The long term release of core decay 24 heat after this initial release of energy may result in pool 25 temperatures as high as 185'F.

26 Q. When the low pressure ECC Systems are called upon to 3

27 inject water in the vessel, is the suppression pool water i

28 hotter than the condensate storage water?

l I

e

1 2 A. The temperature range of operation for the two water 3 sources overlap extensively. However, since the suppression 4 pool serves as a heat sink for steam release before the low-5 pressure ECC systems operate,' the suppression pool would 6 generally be above the temperature of the condensate storage 7 water at the time the low pressure ECC system operates.

8 Q. Will you explain how ECCS injection affects reactor 9 physics?

10 A. An understanding of how water injection changes reactor 11 physics must begin with the concepts of reactivity and 12 criticality. For the fissioning process to be self-sustaining, 13 as a minimum, the number of neutrons born from each fission 14 reaction and surviving to cause another fission must be 15 constant with time. When this minimum condition is achieved 16 the reactor is said to be critical with an effective multipli-17 cation factor, Keff, equal to one. When Keff>l the reactor 18 is said to be supercritical and the neutron population is 19 increasing with time. For Keff 41, the reactor is said to be 20 suberitical with the neutron population decreasing.

21 The shutdown margin of the core is derived from the f 22 term Keff. Shutdown margin (SDM) =1-Keff and is used as 23 a measure of how suberitical the core is. Shutdown margin 24 can also be expressed in terms of how nuch reactivity must 25 be added to reach criticality. Reactivity is a measure of 26 the effect on the neutron population of any change to the 27 reactor core. The addition of reactivity increases neutron 28 population. The removal of reactivity decreases neutron

II 2 population.

3 For a BWR, a value is placed on the minimum amount of 4 shutdown margin which must be maintained when the reactor is 5 subcritical. This minimum is determined once the final core 6 enrichments are determined and is based on the core remaining 7 suberitical at cold shutdown even with the control rod 8 producing the greatest reactivity addition completely withdrawn.

9 Q. Please explain how the temperature of the water being injected during ECCS operation effects reactivity.

~

10 11 A. The four paramecers which affect reactivity in a BWR 12 are the position of the control rods, the temperature of the 13 core coolant water, the temperature of the fuel, and the 14 steam void concentration of the core. The water being 15 injected by ECCS operation is mixed at its injection location 16 and subsequently affects the temperature of the core coolant 17 water and the void concentration in the core. These effects 18 result in either a reactivity incr-ase or decrease depending 19 on reactor conditions. For most reactor operating conditions, 20 initiation of the ECCS will cause reactivity to decrease due 21 to a reduction in steam leaving the vess21, resulting in a 22 small reactor pressure decrease, an increase in core voids, and a power reduction. 'For a narrow range of low power 23 24 reactor conditions, ECCS injection can cause a reactivity 25 increase due to a decreace in reactor coolant la'.et tempera-26 ture. This increases the density of the steam-water mixture, 27 generally referred to as the moderator, and thus results in 28 an increasing neutron population by reducing the number of

W2 1

2 neutrons which leak from the fuel cell and undergo non-3 ,

fissioning absorption in a control rod or some core structure 4 material. The result is a reactivi'.* increase.

0 When is moderator density maximum?

5 A.

Moderator density is maximum wuen the reactor is in the 6

7 cold shutdown mode of operation (Temperature = 70*F, Pressure 8 = 1 atmosphere). Any ECCS injection prior to normal cooldown 9

to cold shutdown and after scram only serves to accelerate The reactor is 10 the rate at which cold shutdown is reached.

11 designed to maintain suberiticality (Keffel) even when ECCS 12 injection accelerates the cooldown process. For those 13 events in which ECCS operates without the reactor being 14 scrammed, a analysis must be done to quantify the reactivity 15 insertion rate and the effect on fuel temperature.

16 0 Has GE performed such an analysis?

17 A. Yes, GE has done such a detailed analysis.

18 0 For three events listed in the contention, would you 19 describe the sequence of events affecting reactor operation 20 and fuel temperature?

21 A. For LOCA, the core is brought to a suberitical state 22

1) by insertion of the control rods before any ECCS (Keff 23 injection starts. 7.he addition of cold ECCS water aids in 24 reflooding the core and makes core reactivity less negative However, the modera-25 since it increases moderator density.

26" tor density is still less than its value at cold shutdown, 27 thus the shutdown margin is maintained and the core remains 28 suberitical. The reactor is brought to a cold shutdown state I

11 i 2l af ter core reflood using the Residual Heat Removal (RHR) 3 System. This System circulates water from the suppression 4 pool to the vessel and maintains water level while removing 5 decay heat in the.IUIR heat exchangers.

6 For RIA, which is any rapid increase in reactivity 7 other than those increases due to expected plant conditions, 8 there is no ECCS operation. The most limiting RIA for 9 Allens Creek is the Rod Drop Accident.. For this accident 10 the reactor will either be in a cold shutdown condition, in 11 which case the reactor will be subcritical with sufficient 12 shutdown margin to remain so even if the maximum worth 13 control rod is rapidly withdrawn from the core, or there 14 will be sufficient forced circulation from the recirculation 15 system and water supply from the feedwater system to handle 16 any reactivity insertion without the aid of any cold water 17 ECCS injection.

18 For ATWS, which is an event which combines the failure 19 of the normal scram function of the control rod drive system 20 with an anticipated transient, the reactor is shutdown by 21 tripping the recirculation pumps in combination with either 22 an alternate method of control rod insertion or by the 23 injection of boron into the vessel via the standby-liquid 24 control line. Only in the case of boron injection would the 20 ECCS operate before the reactor was brought to a subcritical 26 state and then only the HPCS would be utilized since the 27 vessel pressure is well above the pressure injection point 20 for the low pressure ECC Systems. General Electric has e ,--

1 2 analyzed such a case with boron injection and found that the 3 reactivity increase due to HPCS injection is negligible.

4 (See NEDO 24222.) For HPCS injection after the reactor has 5 been brought to a suberitical state, the amount of reactivity 6 increase would be insufficient to cause criticality for the 7 same reasons discussed above concerning LOCA. Long term 8 cooling and eventual cold shutdown is provided by the RHR 9 System. In this mode, the System withdraws reactor water 10 from the vessel, passes it through heat exchangers to 11 remove core decay heat, and then returns it to the vessel.

12 This mode of RHR operation is not part of ECCS and does not 13 involve any cold water injection.

14 Q. Is there any event in which injection of ECCS water 15 occurs without the reactor being scrammed other than an.

16 ATWS7 17 A. There is no plausible event which would cause this to 18 happen. However, General Electric has done an analysis in 19 which they have considered inadvertent initiation of ECCS 20 during normal low power operation. Having the reactivity 21 insertion at low power maximizes the reactivity insertion 22 rate and thus the rate of fuel temperature increases. (ECCS 23 injection at high power'will result in a reactivity decrease.)

24 This injection results in gradual drop in the core coolant 25 inlet temperature and a simultaneous increase in core power.

26 When the power level reaches the Intermediate Range Monitor 27 (IRM) Scram setpoint, a scram is initiated which terminates 28 the power increase and prevents fuel damage. This event is

1 2 much less severe than the design basis rod drop accident 3 (See PSAR Section 15.1.35 and previous discussion on RIA) 4 since the rate of reactivity insertion due to the cold water 5 injection is much less than that due to a rod drop event.

6 0 What are your conclusions?

7 A. The ECC Systems do inject water which is colder than 8 the reactor coolant and therefore in some cases can increase 9 reactivity by increasing moderator density. However, in no 10 case will core reactivity be increased in an amount to cause 11 fuel damage.

12 13 14 15 -

16 17 18 19 20 .y_

21 22 23 24 25 26 27 28 w

~

1 Exhibit ECE-1 2 EDUCATION AND PROFESSIONAL QUALIFICATIONS 3 Eugene C. Eckert 4

Mr. Eckert received a Bachelor of Science Degree in 5 Electrical Engineering from Valparaiso University in Indiana 6 in 1958. During the next year, he a,ttended Standford University 7 under an Oak Ridge Fellowship and received the Master of Science 8 Degree in Engineering Science in August, 1959.

9 Immediately upon joining General Electric Company in 10 September, 1959, Mr. Eckert participated in a company-wide engineering training program. His work assignments in this 11 12 program included large jet engine control design, aircraft 13 nuclear propulsion control analysis,. nuclear submarine kinetic 14 and control analysis, and industrial control simulation analysis at GE's Research and Development Center. After completing this 15 16 program in 1962, Mr. Eckert joined General Electric's Nuclear 17 Energy Division to work on Boiling Water Reactor (BWR) simulation and dynamic analysis. He has been responsible for 18 i 19 design anc licensing documentation of the dynamic analysis for several GE BWR's and has participated in initi startup 20 21 testing of many of the units. He led the dynamic design efforts l 22 which estr.blished the BWR/4 product line, culminated in 1974 23 by the startups of the Browns Ferry (TVA) , Peach Bottom (PECO) 24 and Fukushima-2 (Japan) units. Since then, his design and 25 analysis capabilities have been applied in all BWR product 26 lines. He has been lead total p'lant design engineer and, 27 since 1971, manager of transient analysis for BWR's.

c 28 In his current position, he is responsible for establishing

~

., -~ -- - - , . . - . - - - - - - - - ..

i

~

I the simulation requirements of the computer models needed to 2 perform transient analyses, development of design procedures 3 evaluation of BNR stability, and evaluation and specification 4 of the functional protection systems required for reactor abnormal transient protection. Included is the analysis 5

6 and mitigation of transients with postulated failure of reactor 7 scram (ATWS). Plans for new product lines are evaluate.d, and 8 all projects are carried through plant startup until turnover 9 to the utility.

10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 l 25 26 27

, 28

, . -, - + ~~