ML20011E786

Forwards Overview of Evaluation on Compliance W/Environ Qualification Regulation (10CFR50.49) During Defueling of Facility.During Defueling,Pcrv Will Be Maintained at Nearly Atmospheric Pressures & Rapid Depressurization Not Possible

ML20011E786
Person / Time
Site:	Fort Saint Vrain
Issue date:	02/07/1990
From:	Crawford A PUBLIC SERVICE CO. OF COLORADO
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
P-90043, NUDOCS 9002220471
Download: ML20011E786 (9)

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Public Service' l0.*,,-a ,

P.O. Box 840 i Denver CO 80201 0840 ,

3 2420 W. 26th Avenue Suite 100D, Denver, Colorado 80211 A. Clegg Crawford '

Vice President Nuclear Operations  ;

February ), 1990 ,

Fort St. Vrain i

F Unit No. 1 I l P-90043  !

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U. S. Nuclear Regulatory Commission .

ATTN: Document Control Desk  !

Washington, D.C. 20555 'I L 1

, Docket No. 50-267 i'

SUBJECT:

10 CFR 50.49 Compliance During Defueling

REFERENCE:

PSC Letter Crawford to Weiss, ,

dated November 21,1989(P-89452) .i Gentlemen:

Public Service Company of Colorado (PSC) announced the end of nuclear operations of the Fort St. Vrain Nuclear Generating Station (FSV) on  :

August _ 28, 1989. The Fort St. Vrain plant has been permanently shutdown and,-as stipulated in the referenced letter, PSC 'has made .

the decision that the FSV reactor will never be operated at any power level again.-

Following a 100 day cooldown, defueling of the active core began on November 27, 1989. Preceding the commencement of this activity, PSC submitted several License Amendments and a Safety Analysis Report to support defueling. Nume'rous other analyses were undertaken to determine the extent of applicabili+.1 of the 10 CFR Part, 50 requirements as a result of the_permanc:{ shutdown mode of operation. ,

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, .Public Service Company hereby submits in the enclosed attachment for your information, an overview of the evaluation on compliance -with *

.the Environmental Qualification Regulation (10 CFR 50.49) during the defueling of FSV. 1 l

The evaluation concluded, that upon completion of modifications to ht

./h,g' l limit the mass flow rate of the auxiliary steam system, no accident condition could result which creates e narsh environment requiring fgl6 l= equipment important to safety to be environmentally qualified. ,

- 9002220471 900207 "I

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. February 7,-1990 Should you have any questions regarding this infomation, please l contact Mr. M. H. Holmes at (303) 480-6960.

'Very truly yours, f

l' Gh A. Clegg Crawford

,. Vice President I- Nuclear Operations Attachment; ACC/JCS:tmk cc: . Regional Administrator, Region IV

,' ATTN: Mr. T. F. k'esterman, Chief Projects Sectior. B Mr.. Robert Farrell 1 -Senior Resident Inspector Fort St. Vrain 4

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Attachment to P-90043 Page 1 i

EVALUATION OF COMPLIANCE WITH ENVIRONMENTAL QUALIFICATION DURING THE I

DEFUELING OF F$Y The goal .of the evaluation is to show the method of compliance with

-10 CFR 50.49 during defueling. The approach and methodology used -in  :

i- the existing EQ program development were followed. The following  ;

overview presents excerpts from the

Engineering Evaluation (EE-EQ-0070) on Environmental Qualification during defueling.

j c The first- step is to define the various accidents considered in the

! EQ program, then to determine if a harsh environment exists from the accident conditions. The environmental conditions evaluated as listed in 10 CFR 50.49(d)(3) are as follows: -

1. Temperature

2. Pressure .

3. Chemical Effects

4. Humidity

5. Radiation

6. Submergence I. ACCIDENT EVALUATION l This section presents a separate evaluation for each of the accidents detennined to be applicable during defueling. These accidents are DBA-1, DBA-2, MCA and HELB's.

Design Basis Accident No. 1 DBA-1 is the permanent loss of forced circulation cooling as described in FSAR Section 14.10. This accident does not result in elevated building temperatures, increased building pressure or '

L humidity, release of chemical or water sprays, nor induce I

submergence.

As evaluated in the existing EQ program for reactor operation, DBA-1 only resulted in a radiation environmental conditien of 497 rads which is not a harsh environment.

This accident has been reanalyzed for the conditions associeted -

with defueling. This analysis was based on occurrence o' DBA-1 100 days following shutdown from 83.2% power operation. It was conservatively assumed that the one loop of PCRV liner cooling did not start until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the initial loss of forced circulation cooling. The analysis results indicate that, with .

one PCRV liner cooling loop operating with flow redistributed to the. top head liner, maximum fuel temperatures remain within those experienced during normal reactor operation. The above radiation condition resulting from reactor operation would bound that which would occur during defueling. Therefore, DBA-1 is considered a

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mild environment accident and qualification for this accident in accordance with 10 CFR 50.49 is not required.

Design Basis Accident No. 2 DBA-2 as described in FSAR Section 14.11, is the instantaneous failure of both the primary and secondary closures in one of the PCRV penetrations resulting in a rapid depressurization of the PCRV and release of the primary coolant.

During 'defueling, the PCRV_ will be maintained at nearly atmospheric pressures, and a rapid depressurization is not possible. Therefore, DBA-2 is no longer considered possible, and environmental qualification is not required.

Maximum Credible Accident MCA, as described in FSAR Section 14.8, is a break in. the helium purification system resulting in a slow depressurization of the PCRV.

l As discussed previously, the PCRV will. Le maintained at nearly L atmospheric pressures during defueling. Depressurization from high pressures is not possible. In addition, the helium purification system is no longer in service.

MCA is no longer considered possible, and- environmental L qualification is not required.

j High Energy Line Breaks HELBs originally evaluated for the Fort St. Vrain EQ program included breaks in the main steam, hot reheat steam, cold reheat 3

stene, auxiliary steam, extraction steam, feedwater, and condensate lines. The breaks were assumed to occur in either the Resctor or Turbine building.

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During defueling, the het reheat steam, cold reheat steam, and extraction steani systemt will not be in service. A break in one of these lines will not result in any accident conditions.

L The condensate,.feedwater ard main steam piping will be used for decay heat removal. Peadirge (temperature and pressure) were

l. taken to determine the chnditions for these lines during decay l' ,

heat removal. The conditions of these lines are below the saturatioti- point of steem and will not flash to steam for a line L rupture. l h

The auxiliary steam system will be used to supply building heat, L for maintaining PCRV liner cooling temperature requirements, and L possibly for tw decontamination system. Because of the l

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Attachment to P-90043 Page 3 possibility of elevated temperatures occurring following a break in this steam system, each of the environmental parameters is evaluated in detail in the following section for this event.

II. ENVIRONMENTAL CONDITION EVALUATION

, This section presents an evaluation of each of the environmental vMitions (as listed above) as a result of a rupture in the aux 111ary steam system.

i' Temperature The original conditions for a break in the auxiliary steam system were that of.an offset rupture in an 8 inch line, 150 psig, 600 degrees F, a 145,000 lbm/hr steam release with SLRDIS detection (280 seconds) and manual isolation at 940 seconds (see FSAR Section 7.3.10). This line break resulted in a temperature profile with a peak temperature of 168 degrees F at 941 seconds.

In order to reduce the building temperetures, the amount of steam release must also be reduced. The system conditions for a reanalyzed auxiliary steam system line break were changed to: an offset rupture in an 8 inch line, 162 psia (150 psig), 650 degrees F (the higher temperature rating of the backup euxiliary boiler), a' mass flow rate of 15,000 lbm/hr, leak manually detected and isolated at one hour with 5,000 lbm of residual steam in the system at one hour. Termination is accomplished by either valve closure or o> tripping the auxiliary boiler. Manual detection and isolation uter one hour is considered acceptable,  !

as discussed in FSAR Section 1.4.6.8.1.

The temperature profiles developed for these reduced system conditions are shown in Figures 1 and 2. The profiles show a peak temperature of 119.9 degrees F in the reactor building and a peak temperature of 116.5 degrees in the turbine building, j These temperatures are not significantly higher than the maximum ambient temperatures experienced in these buildings. Therefore, there would be no harsn temperature environment caused by this line break, and qualification for elevated temperatures would not be required.

In order to meet these analyzed conditions, the auxiliary steam system will be modified to limit the steam flow rate to 15,000 lbm/hr.

Pressure The peak pressure for the current EQ program is less that one psig. This peak pressure was based on profiles developed for all HELBi s included in the current EQ program. P se pressure

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P-90043 i Page 4 s 1

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! profiles bound the pressure that would accompany any auxiliary  !

steam, condensate, feedwater, and main steam line break that would occur during defueling.

The peak pressure of less than 1 psig does not constitute a harsh .

pressure environment. Qualification for pressure is not  :

7 required.

Humidity

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The' humidity profiles are also shown in Figures 1 and P., The humidity reached levels that are higher than normal operation for

- a short duration, however, the humidity never reaches 100%. As a result, the humidity levels are not considered to be significant.

Qualification for humidity is not required.

Radiation 4

The worst case radiation level for HELB's during normal operation is 12 rads as discussed in FSAR Section 1.4.6.3.1. The auxiliary -

steam,. condensate, feedwater, and main steam line breaks were t included in the HELB's evaluated. As a result, the 12 rad radiation level is considered to bound the radiation levels occurring following these HELB breaks during defueling.

The 12 rad level is significantly less than the 497 rad level ,

resulting from DBA-1 (discussed above) which is considered a mild environment. Therefore, qualification for radiation is not required.

C_hemical Effects / Fire Suppression System Sprays Fort St. Vrain does not utilize chemical spray systems for safe shutdown. Therefore no analysis is necessary for chemical -

effects.

There is the possibility that the fire suppression spray systems may be actuated by a steam line break. A review of the fire spray zones and the temperature rating of the actuation devices

. was made. A minimum temperature rating for the actuation devices is 165 degrees F. This is above the maximum bulk building volume -

temperature (120 degrees F) for the auxiliary steam line break.

Therefore, the actuation of the fire spray systems is not possible due to steam line breaks.

Qualification for chemical sprays / fire suppression system sprays is'not required. .

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Submergence As discussed in FSAR Section 1.4.6.7, an evaluation was made which determined that a break in the condensate lines presented the worst case flooding scenario. The results of that evaluation showed that all safe shutdown equipment was located above the expected flood level, i Based en that evaluation, no equipment will be submerged as the result..of an accident during defueling. Qualification for submergence is not required.

111. CONCLUSIONS i There are four possible high energy line breaks that must be considered during defueling, breaks. in the auxiliary steam, condensate, feedwater, and main steam systems. None of these

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line breaks result in harsh environment conditions for any of the parametersidentifiedin10CFR50.49(d)(3)asdiscussedabove.

The auxiliary steam system must be limited to a 15,000 lbm/hr

. steam capacity to support the above evaluation and conclusions.

Public Service Company will continue to implement EQ requirements until such time that the auxiliary steam system is modified to limit a steam release to less than or equal to 15,000 lbm/hr for a postulated line break.

Upon completion of the required modifications and return to service of the auxiliary steam system, the present EQ program will be modified. Full compliance with the requirements of 10 CFR 50.49 will continue to be met in that there are no accidents during defueling that would result in a harsh environment as defined in 10 CFR 50.49 which would require equipment important '

to safety to be environmentally qualified, v

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