ML20008D654

Forwards Response to Three 780818 NRC Questions Release to Operate Subj Reactor in Fluctuating Mode for Test Purposes at Less than 70% of Rated Thermal Pwr

ML20008D654
Person / Time
Site:	Fort Saint Vrain
Issue date:	09/06/1978
From:	Justin Fuller PUBLIC SERVICE CO. OF COLORADO
To:	Gammill W Office of Nuclear Reactor Regulation
References
FOIA-81-127 P-78146, NUDOCS 7809190190
Download: ML20008D654 (7)

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public service company e cenemaIo P. O. Bet 361, Platteville, CO 80651 September 6, 1978 Fort St. Vrain Unit No. 1 P-78146 Mr. Villiam Gammill Asst. Director for Advanced Reactors _ ,

United States Nuclear Regul,atory Commission  ?-

Division of Project Management '

Washington, D.C. 20555 Docket No. 50-267 Gentlemen:

Subject:

Fort St. Vrain Operations and Oscillations Testing Ref: NRC Letter of August 18, 1978, Speis to Fuller Attached please find the response to the three (3) questions forwarded to PSC as Enclosure 1 to the referenced NRC correspondence. The attached response should provida all the information required by the Staff to form the bcsis fer a release to operate the Fort St. Vrain reactor in the fluctuating mode for test purposes at less than 70% of rated thermal power.

PSC, therefore, requests the NRC provide a release in writing to perform fluctuation testing at < 70% of rated thermal power no later than Sept-ember 15, 1978.

The response to the questions forwarded'in the referenced correspondence as Eacle.wce 2 are presently scheduled to be ready in draft form for review on or about the week of October 9, 1978.

Very truly yours, J. .. ruller

Vice President l Engineering and Planning JKF:il l Attachment ~

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Response to NRC Question 1 for Less Than 70% Power Question l 1. With respect to' monitoring for potential incra ses in the amplitude f

b of fluctuating signals over those heretofore experienced:

a. provide the explicit limits that will be used to determine whether a significant change has occurred (e.g. , Helium temperature response, displacement probe response, neutron detector response), and,

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b. state the remedial action that will be taken should such limits be

! reached or exceeded.

Response

The reacto'r is not normaily operated'in the fluctuating mode and the 3

operators have specific instructions to i==ediately terminate any unplanned

fluctuations which may occur. During fluctection tests, the parameters that reflects the core outlet temperature fluctuations are primarily core

! region helium outlet temperature' indications, as pointed out in the recent j submittal (PSC Letter P-78137, dated 8/11/78). The temperature changes are observed on core region thermocouples and on steam generator helium inlet temperature thermocouples. Fluctuations are also observed on nuclear flux channels relating to both small localized reactivity changes and larger fluctuations due to neutron streaming through reflector gaps.

The FSV Plant Protective System settings and Limiting Conditions for 2

Operation were originally established in the FSAR and Technical Specifications to protect core and other primary system components from damage. The basic philosophy utilized was that core fuel temperature is a function of power level and helium temperature. Helium temperature, is directly reflected b2 steam temperatures, and steam temperatures relate to steam generator metal component temperatures. Therefore, Plant Protection System limits were l

established based on steam temperature and power parameters to provide appropriate protection. This logic still applies under conditions of core outlet temperature fluctuations. In addition to the limits given in the Technical Specification for normal operatica, limits for steam temperature variations were established and are being utilized in the fluctuation testing as indicated below (see RT-500 and RT-502):

Throughout the test, the intent will be to minimize the time spent in fluctuation except when necessary to record FM data. When fluctuations are present, the following limits apply:

i a) A temperature fluctuation of module main steam (MS) temperature (about its mean) up to 1 10*F is acceptable with no specific i time limit.

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2 b) A te=perature fluctuation of module MS te perature (about its mean) greater than i 10*F but less than - 30*F is tolerable for a maxi =um of one hour duration per event.

c) A te=perature fluctuation of module MS te=perature (about its mean) which reaches 30*F amplitude is cause to take i= mediate corrective action.

Test procedures call for a reduction in reactor power to stop fluctua-tions should these limits be exceeded.

The observed large changes in individual nuclear flux channels were determined to not be indicative of changes in core reactivity and therefore have not been constrained in the fluctuation test operating limits. The history of events to date-sh'ow that the maximum fluctuation changes occur on Channel VI and have been less than i 10% of rated at power levels up to 70%. Since it is desired to monitor future fluctuations at power levels up to 70% for similarity to pr'evious fluctuations, a limiting condition will be introduced into the test plan requiring a power reduction if the

_ fluctuations on Channel VI exceed i 10%. 3ased upon trends in data observed at less than 70% power, it is expected that the fluctuation change on Channel VI will exceed i 10% when fluctuation testing is conducted at power levels above 70%. Accordingly, appropriate 14-4 s on Channel VI fluctuations must be established for testing at power levels above 70%.

Some indication of increased PCRV resonant activity has been observed on the ?CRV displacement probes during fluctuations, but the correlation of the activity to te=perature and flux changes is not strong. Four = ore of these state-of-the-art probes have been attached to the PCRV for the next 7 test series, and an i=preved calibration method has been e= ployed in order to determine core accurately the relationship of increased PCRV resonant activity and fluctuating parameters. The use of the PCRV displacerents as an indication of amplitude of fluctuations is not recommended at this time since the observed displacement seems to decrease as power increases during fluctuating operation and to increase with a power increase during non-fluctuating operation. A better understanding of the relationship of the PCRV displacement to fluctuations is required before this parameter is considered as a limiting parameter.

A number of plant alarms would be activated if fluctuations increased in amplitude over previous conditions:

a) A core region high outlet temperature alarm is set at 1520*F for any one of the 37 regions.

b) The rod withdrawal prohibit alarm is currently set at 80% power.  ;

c) The linear channel high will alarm when indicated power exceeds the current setting of 90% power.

d) Main steam :odule outlet temperature mis =atch of i 30*F from the average is alar =ed.

e) Hot reheat steam odule outlet te=perature =is=atch of i 30*?

from average is alarmed.

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l Response to NRC Question 2 for Less Than 70% Power

. Question

2. Identify the role, if any, that the rer.ctor control system or the reactor protection system would play in protecting against or com-pensating for .a divergent or high amplitude signal response during oscillation testing.

Resoonse The reactor control sys3em is designed to respond to variations in system average parameters it rates much faster than the fluctuation period.

As discussed in the fluctuation status report (PSC letter P-78137, dated 8/11/78), average parameters have shown little or no variation during periods of fluctuation testing. In addition, flucruation tests have in-dicated no tendency for system parameters to diverge during fluctuations from their intended settings, the reactor control system would act to compensate for the change just as it would during operation without fluctuations.

A review of the plant protective system (PPS) operating limits described in Technical Specifications LCO 4.4-1 and LSSS 3.3 indicates two protective fluctuations which can act to protect the system should

, unexpectedly high a=plitude signals occur. These are:

Functienal Unit Setting & Full Power Linear Power Channels - High j[ 140%

Reheat Steam Temperature - High jL 1075'T While the plant is restricted to operation at j[ 70% power, the linear power channel trip setting has been reduced to 90% of raced power. Thus, La the event of high amplitude signals during fluctuation testing, trip signals rrom this PPS unit would be provided before the high reheat steam temperature unit would function. These high power trip signals, even if they did not occur in the proper combination to provide reactor trip, would alert the operator to the existence of unexpected large fluctuations.

However, it is expected that monitoring of steam generator module main steam temperatures, as discussed in the response to Question 1, will provide indications of divergent or high amplitude fluctuations prior to PPS action.

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Response to NRC Question 3 for Below'70% Power Question

3. Discuss the actions to be taken in the event coolant blockage were to occur to a fuel colu=n during oscillation testing with consequent

' fission product release to the primary system. Your response should address:

a) the time response of the primary coolant fission product monitor; b) the action that would be taken if a specified level or rate of increase in fission product cencentration were measured; and c) the potential degree of fission product leakage from the primary system and associated auxiliary systems during the course of subsequent coolant jurification activities; d) any other potential consequences including the effects of high gas temperature if coolant flow is reinstated to overheated fuel colu=ns.

- Response Although the possibility of coolant blockage and subsequent fission product release is extremely remote, appropriate protection against increases in the primary coolant activity is pre.'ided by the primary coolant fission product continuous monitor. The response time of the monitor is approxi-mately 2.5 minutes. The output of the fission product monitor can be ob-served by testing personnel during periods of fluctuation testing to deter-mine whether any significant change of the level of fission product activity has occurred.

The present levels of primary coolant circulating activity are well known and are far below FSAR " Design" activity limits; for exa=ple, the circulat-ing activity measured in startup test B-13 at 65% power is about a factor of 32 below the " Design" activity. The variation among readings taken at this power level is less than 10%.

During fluctuation testing, the reactor power will be reduced to stop the fluctuations if the circulating activity increases by 25%. If the pri-

, mary coolant activity level increases by a factor of five, the reactor will be shut down in an orderly manner. This factor of five increase would re-sult in less than 5000 Ci at or below 70% and no more than 10,000 Ci of circulating activity at 100% power based on present projections. The act-ivieles are well within the " Design" activity limit of 30,900 C1.

.Though considered highly unlikely, complete blockage of coolant flow to a fuel column during power operation would result in a rapid increase in column temperatures over the next several minutes. Scoping calculations indicate that sufficiently high temperatures would be reached to fail the fuel particle coatings and release some of the inventory of fission pro-ducts contained in the colu=n to the primary system. As an upper bound, e

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the total inventory could be up to 35x108 curies for a conservatively high peaking factor column.

High primary system activity alarm would alert the operator to such an extraordinary situation. Operator action would trip the plant and proceed with the removal of residual heat. At this time, most of the fission prod-

, ucts would plate out on the colder surfaces of the primary system. The l remaining circulating activity would continue to be processed through the helium purification system. Under these conditions, helium purification system parameters would remain within operational limits with the high te=perature filter adsorber retaining particulates, metals and iodine and the low temperature adsorber retaining the xenon and krypton.

For purposes of a bounding analy' sis, however, it was assumed that the helium purification system'is shut down and the primary system is isolated 4

with the entire inventory initially contained within the PCRV. losses i

from the PCRV would be less than one PCRV volume per year (22 lbs/ day),

i the maximum total design leakage rate for the outer (secondary) penetration i closures (FSAR Section D1.3.4.2). Althcugh 22 lbs/ day was utilized in this analysis, losses from the PCRV vould be =uch lower if purified buffer helium flow is maintained. In addition, pri=ary system pressure would drop to about half the design value during the first 15 minutes of cool-down following reactor trip, thus reducing the driving differential for PCRV leakage.

l In performing this bounding calculation, the NRC staff assumption that all of the fission products in the blocked column are instantaneously released to the primary coolant was used. It was further assumed that all noble gases, 25" of the halogens and 1% of the strontium associated with the blocked column were available for release from the PCRV per Table 4.3 of the Safety Evaluation for Fort St. Vrain (January 20, 1972). Accordingly,

for the hypothetical flow blockage case the calculated 180 day doses at the boundary of the low population

one are 0.002 rem whole body, 0.1

! rem thyroid, and 0.02 rem bone. These are much less than the 0.33 rem whole body, 8 rem thyroid, and 5 rem bone doses calculated by the NRC I

staff for Dssign Basis Accident No. 1 at Fort St. Vrain. Therefore, the consequences of the incredible column flow blockage scenario are well within those previously accepted by the NRC.

With respect to the potential consequences of reinstating coolant flow i to a fuel column, high coolant temperatures would be attained. Using the l sublimation temperature of graphite as a conservative upper limit, the maximum gas temperature would be on the order of 6000*F. The high tempera-ture gas would mix with cooler gas from the other fuel colu=ns resulting in roughly a 2600*F core support block outlet gas temperature. This gas

! would mix with lower plenun gas prior to contact with the ther=al barrier

( and result in steam generator inlet duct thermal barrier temperatures less than 1900*F. If the flow blockage conditions were severe enough to result in significant fuel failure, it would be obseried on region exit thermocouples readings when flow is re-established and would result in a high region exit temperature alarm as well as being observed on the primary coolant activity monitor described previously.

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3 In sumary, as we have stated in the previous submittal, we believe that the probability of coolant channel blockage is incredibly s.:all.. If coolant channel blockage were to occur, havever, the controls described above provide sufficient protection against greater than design fission product activity or esoponent temperatures, a e9 '

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