ML20081K389
| ML20081K389 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 06/21/1991 |
| From: | Rhodes F WOLF CREEK NUCLEAR OPERATING CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20081K394 | List: |
| References | |
| ET-91-0109, ET-91-109, NUDOCS 9106270131 | |
| Download: ML20081K389 (18) | |
Text
,_ V.
l W6LF CREEK ' NUCLEAR OPERATING CORPORATION Forrest i Rhodes vee e. .n,mi t<,n nno a t.cnnes s nm..
June 21. 1991 ET 91-0109 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station P1-137 Washington, D C. 20555
Subject:
Docket No. 50-482: Revision to Technical Specification 4.6.2.3 - Containment Cooling System Gentlemen: ,
The purpose of this letter is to transmit an application for amendment to '
Facility Operating Licease No, U?F-42 for Wolf Creek Generating Station (WCGS), Unit _ No. 1._ This proposed license amendment revises Technical 3 Specification 4.6.2.3, " Containment Cooling System ' and affected Technical i Specification Bases to reduce the minimum required cooling water flow to the containment cooling- units during acciden*, conditions. The proposed amendment also deletes a surveillance requirement for periodic measurement of cooling water flow to the containment cooling units in -the norni operating alignment. Plant specific analyses have been performed to demonstrate continued adequate performance of the containment cooling-system with the proposed reduced minimum flow requirement.
Attachments I through III provide the Safety Evaluation, Significant Hazards Consideration Determination, and Environmental Impact Determination supporting the requested change. Attichment IV provides the revised Technical Specification pages, ,
In accordance with 10 CFR 50.91, a copy of this application, 4th attachmente, is being provided to the designated Kansas State Official.
This' proposed amendment is needed to support planned modifications and-flow balancing of the service- water _ system during the u}. coming fifth WCGS refueling outage. Therefore, Wolf Creek Nuclear Operating Corporation request approval of this proposed amendment prior to the fifth -refueling outage which is currently scheduled for September of 1991.
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- 1 ET 91-0109 l Page 2 of 2- -
If you have a", "%stions concerning this matter, please contact me or Mr. H. K. Che...ott of my staff. >
t Very truly yours, h
/n '(Y /
Forrest T. Rhodes
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Vice President ;
Engineering & Technical Services
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FTR/jra Attachments: I - Safety Evaluation !
II - Significant Hazards Consideration Determination III - Environmental Impact Determination IV - Proposed Technical Specification Changes -
f cc G. W. Allen (KDnE), w/a L. L. Gundrum (NRC), w/a A. T.11ovell (NRC), w/a R. D. Martin (NRC), w/a D. V. Pickett (NRC), w/a t
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4 STATE OF KANSAS )
) SS COUNTY OF COFFIN )
Forrest T. Rhodes, of lawful age, being first duly swocn upon oath says that he is Vice President Engineering and Technical Services of k'olf Creek Nuclear Operating Corporations that he has read the foregoing document and knows the content thereof; that he has executed that same for and en behalf of saio corporation with full power and authority to do so and that the facts therein stated are true and correct to the best of his knowledge, information and belief.
g, p u 1 Forrest T. Rhodes Vice President Engineering & Technical Services SUh5CRISED and sworn t o before me this 4/ day oft 'incJ , 1991.
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Attachment I to ET 91-0109 Page 1 of Io i
ATTACIMENT I j
SAFETY EVALUATION i
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At t a chtnent I to ET 91-0109 Page 1 of 10 ATTACI&iENT 1 SAFETY EVAL,UATION am h... _ . . _ . _ .. .. . .
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Attachment I to ET 91-0109 Fage 2 of 10 l SAFETY EVALUATION Introduction and Dencription of Proposed Change Major modifications to the Service Water System (SWS) and Essential Service Water System (ESWS) were implemented in 1990 during tne fourth Wolf Creek Generating Station (WCGS) refueling outage to address concerns relative to erosion and corrosion. The modifications increased and redistributed the flow supplied by the SWS and increased the backpressure in the system to reduce the potential for erosion. Follouing subsequent flow balancing, measurements indicated the pressure drop across the system had increased.
This resulted in a reduction in the margins between the available cooling water flow rate and the design cooling requirements for various components.
To provide sufficient flow rate margins to accommodate future flow balancing and assure adequate flow is available to all components, a reduction in the required cooling flow to the Containment Cooling System (CCS) is proposed in thi s amendment request. The pcoposed reduction in the required ESWS flow to the CCS is possible due to conservatisms inherent in the original safety analysis.
Two changes are proposed to the existing Technical Specification 4.6.2.3.
Technical Specification 4.6.2.3.b is revised to reduce the required cooling water flow rate to the CCS with the ESWS in its post-accident alignment from 4000 to 2000 gallons per minute (gpm) for each containment cooler group, t This surveillance is conducted at least once per 18 months. As detailed below, reanalysis using plant specific data and more contemporary computer codes demonstrates adequate CCS performance at the proposed reduced ESWS flow rate.
The second proposed change is to Technical Specification 4.6.2.3.a. This specification requires measurement of the cooling water flow rate to each cooler group at 31 day intervals with the ESWS in its normal alignment. As discussed below, flow measurements in the normal operating mode do not provide a consistent , or reliable indication of the ability of the cooler units to perform their safety function. Detection of potential flow degradation is more reliably detected by an ongoing heat exchanger performance monitoring program. Therefore, this surveillance requirement is being deleted.
Conforming changes are included for various sections of the Technical Specification Bases.
System Description
Technical Specification 4.6.2.3, " Cont ai nc.ent Cooling System,' contains surveillance requirements for the CCS. During normal operation the CCS works together with the containment Heating. Ventilation and Air Conditioning (HVAC) system to maintain suitable conditions for equipment located within the containment and maintain the containment conditions within the bounds of the initial conditions assumed in the safety analysis.
Following a Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB),
the CCS operates together with the Containment Spray System to maintain the I
containment temperature and pressure within design limits. The CCS also serves to limit offsite radiation exposure by reducing the pressure differential between the containment and the external environment following a LOCA.
I
Attachment I to ET 91-0109 i
Page 3 of 10 The CCS is r4ade up of two independent groups of two cooling units each.
. Each cooling unit cons:.sts of an air-to-water heat exchanger and a two speed fan. During normal plant operation the ESWS, and subsequently the CCS, is supplied with coolirg water from the non-safety related SWS via a cross connection. Under post-accident canditions the ESWS is isolated from the 4
SWS and cooling watt.r is supplied from the safety related ESWS pumps. The ESWS provides heat cemoval for safety-related equipment during and following design basis acciGents. Higher cooling water flow rates are provided to the CCS in this mode of operation. The CCS fans are operated at high speed with a nominal cooling water flow of approximately 1850 gpm per cooler group supplied by the SWS. Under accident conditions the CCS fans are automatically switched to the low speed setting to prevent overload of the fan motors and the flow of cooling water is increased to approximately 4000 gpm per cocier group as the ESWS is aligned for post-accident service. The safety analysis currently contained in the Updated Safety Analysis Report 1 (USAR) assumes an ESWS flow rate to each containment cooler group in the post-accident mode is at least 4000 gpm.
Evaluation The containment pressure and temperature response following a postulated LOCA and MSLB accidents was reanalyzed to evaluate the impact of the proposed reduction in the minimum allowable CCS cooling water flow rate.
These analyses showed no change in the peak containment pressure following a LOCA and only a small increase in the pesk containment pressure following an MSLB. The revised MSLB peak pressure remains well below the design pressure of the WCGS containment. The peak containment temperature following an MSLB also increased slightly but r emair. ,d within an acceptable range. The results of these analyses are detailed below along with a description of the analytical methods.
The current USAR analyses for the containment pressure and temperature response was performed using the g0PATTA computer code.1 The revised analyses utilized the CONTEMPT-LT/28 computer code. In addition, the mass and energy release to the containment was reanalyzed using more plant j specific data and a more contemporary computer code. The mass and energy release data used in the current analyses is based on generi plant data and was developed using the MARVEL computer code.g For theWestinghouse MSLB event only the first 300- seconds of mass and energy release data was developed using the MARVEL code. This data was extrapolated to the end of 4
the MSLB. transient. For the revised analyses WCGS specific data was used l
BN-TOP-3 Revision 4.- -* Performance and Sizing of- Dry Pressure Containments,' Bechtel Corporation October 1977
'NUREG/CR-0225, ' CONTEMPT-LT/28 - A Computer Program for Predicting Containment Pressure-Temperature Response to a Loss-of-Coolant-Accident,'
Hart: )ves, D. W., et al., Idaho National Engineering Laboratory, March. 1979 WCAP-8312-P-A, Revision 1 (Proprietary) and WCAP-8312-A (Hon-Proprietary),
' Westinghouse Mass and Energy Release Data for Containment Design,'
Shepard, R. M., et al., August 1975 v t-- we ,e-ms. -,r-e---ea-m-r-wr*www-+,w-ess--w,--+w-3-m sre- mm-se-- -w-s-
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. Attachment 1 to ET 91-0109
, Page 4 of 10 and the LOFTRAN computer code was used to develop the mass and energy release data. The new analyses included development of mass and energy release data for the complete duration of the HSLB event. The revised analyser use performed using the same assumptions and initial conditions as spea.i' led in the USAR with the following exceptions. An outside air temperature of 120"F was assumed for the new analyses vs. a previously assumed value of 95 F. The assumed Containment Spray flow was l' reduced by five (5) percent to 1.458E6 lbs/hr. These differences are conservative (i.e., would tend to result in higher values of calculated containment pressure and temperature).
The effect of reducing the cooling water flow to each cooler group from 4000 gpm to 2000 gpm was accounted for by reducing the assumed heat transfer rate of a containment cooler group by approximately 552. This value was derived by taking the design heat transfer rate for a single cooler unit at 2000 gpm and applying an additional 10Z reduction. This is conservative since the 2000 gpm required flow would be split between the two cooler units of each cooler group, resulting in a higher total heat transfer rate than for a single cooler unit with the same cooling water flow rate. Reducing ;
the single unit heat transfer rate by 10Z provides additional conservatism.
Using the methodology described above, the spectrum of LOCA and MSLB accidents described in USAR Section 6.2, " Containment Systems,' was ;
reanalyzed. The peak conta'nment oressure following a LOCA was calculated to be 47.3 psig. This value, obtained for a double ended reactor coolant pump suction guillotine break with minimum safety injection (Figure 1), is the same as the current USAR analysis. The peak pressure following a LOCA r en.ains below the peak pressure of 48 psig specified in Technical Specification 3/4.6.1.2, ' Containment Leakage,' for Local Leak Rate Testing and Integrated Leak Ratte Testing required by Appendix J to 10 CFR 50. As shown by Figure 1, the post-LOCA containment pressure falls to below 50% of the peak yalue within 2c hours as specified by Standard Review Plan Section 6.2.1.1.A.' Therefore, no change to this section of Technical Specifications is require. and there is no impact on the offsite radiological dose consequenews due to a postulated LOCA as described in the USAR.
6 ForanySLB, the peak pressure was calculated to be 48.9 psig for Case 9 ,,
0.80 ft split rupture at 502 power (Figure 2). This is slightly higher (0.8 psi) than thg previously calculated peak pressure of 48.1 psig for Case 12, a 0.66 ft split at 252 power, but remains well below the ,
containment design pressure of 60 psig. The peak containment pressure following an MSLB is not contained in the Technical Specifiestion Limiting VCAP-7907-p-A, 'LOFTRAN Code Description," Gurnett, T. W., et al.,
April, 1984 l NUREG 800, ' Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,' Revision 2 July 1981 i
Case numbers refer to Table 6.2.2-56 of the WCGS USAR
.. Attachment I to ET 91-0109 Page 5 of 10 Conditions for Operations or Surveillance requirements and therefore, no i changes are required as a result of this increase. The revised peak l contsinment pressure following an MSL1 is reflected in proposed changes to Technical Specification Bases sections 3/4.6.1.4, ' Internal Pressure,' and 3/4.6.1.6
- Containment Vessel Structural Integrity.'
The results nf the revised analyses of containment peak pressure are summarized below.
Containment Pressure (psig)
Current USAR New Analyses LOCA 47.3 47.3 MSLB 48.1 48.9 Containment Design Pressure 60.0 The revised peak containment temperature analyses showed that the LOCA events continue to be bounded by the MSLB events as described in the current USAR analyses. For an MSLB the calculated peak containment temperature was 386.5 F which occurred for Case 7, a full dauble ended rupture at 50Z power (Figure 3). This is a slight increase g (1.6 F) from the previously calculated peak temperature of 384.9 F for Case 6, a 0.84 ft split at 752 power.
Review of environmental qualification documentation for equipment located inside the containment has concluded that the equipment remains fully qualified for the revised containment environmental conditions.
In addition to the proposed reduction in required cooling water flow in the post-accident mode, this proposed change includes deletion of the current '
requirement for periodic verification of a specified minimum cooling water flow in the normal CCS operational alignment. In its normal operating mode the cooling water flow to the ESWS is supplied from the non-safety reisted SWS. In this mode the cooling water flow rate to components supplied by the ESWS are dependent on the pressure in the SWS, the position of throttle valves in the ESWS return line to the SWS, and specific component configurations within the ESWS. These parameters may vary greatly due seasonal temperature changes. As a result, the cooling. water flow rate to the containment coolers in their normal alignment also varies significantly. Cooling water flow rate variations of up to 300. gpm .per cooler group are frequently observed as a result.of such variations.in operating conditions. Therefore, the measurement of the normal cooling water flow rate to the containment cooling u:its~ does not provide an appropriate indication of the capability of the coolers to perform their safety function or allow meaningful data trending to detect potential flow degradation.
As discussed above, the capability of the containment coolers to perform their safety function is adequately verified by c.he surveillance requirements of Technical Specification 4.6.2.3.b. This test, which is performed at least once per 18 months, measures the cooling water flow rate to .the containment coolers with the ESWS and its supplied components in their post-accident alignment. This test configuration provides a constant ESWS backpressure and allows an accurate assessment of the tiow capability ,
of the containment cooling units.
Attachment 1 to ET 91-0109 Page 6 of 10 In addition, the performance of the containment cooling units, along with other safety related heat exchangers, is routinely monitored to detect any flow degradation due to pipe f ouling or tube plugging. This monitoring program includes periodic mesaurement of flow and pressure differential as well as periodic testing to measure the heat transfer capability of the cooling units. The surveillance testing required by Technical Specification 4.6.2.3.b together with the routine performance monitoring of these_ heat exchangers provides assurance that the containment coolers will function as required following any postulated accident.
During normal operations the containment coolers also serve to maintain the containment conditions consistent with the assumptions of the safety analyses and suitable for safety related equipment located in the containment. This is adequately verified by the requirements of Technical Specification 3/4.6.1.5, ' Containment Systems - Air Temperature," which requires verification that the average containment air temperature does not exceed 120 F at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Based on the above discussion, both the normal and post-accident functions of the CCS are adequately verified by other testing. Therefore, since the testing currently required by Technical Specification 4.6.2.3.a.2 does not provide an appropriate indication of capability of the CCS. this testing requirement is being deleted.
Conclusions The effect of the proposed reduction in the minimum required cooling water flow rate to the containment coolsrs has been evaluated and shown to have no significant effect on the consequences of postulated accidents. For a LOCA, the calculated peak containment pressure is unchanged and the containment pressure response remains consistent with the assumptions used to evaluate the potential offsite radiological consequences of a LOCA due to leakage from the containment. For an MSLB event, the containment pressure remains well below the design pressure of the containment and the safety related equipment located within the containment has been evaluated and found to remain fully qualified for the revised containment conditions. The deletion of the surveillance requirement for flow measrrement during normal operations does not adversely affect the reliability of the CCS since ongoing performance monitoring of these heat exchangers provides a more appropriate indication of potential degradation. On these _ bases, it is concluded that the probability of occurrence or consequences of equipment malfunctions or accidents previously evaluated in the USAR are not increased, i
The CC3 serves to mitigate the consequences of postulated accidents, but is not associated with the initiation of design basis events and has no direct impact on the Reactor Coolant System or other structures or systems associated with the initiation of postulated accidents. Therefore, the possibility of a new accident that is different from any already evaluated in the USAR is not created.
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. Attachment I to ET 91 0109 l page 7 of 10 j l
The proposed reduction in required cooling water flow is acceptable based on revised accident analyses. For a LOCA the containment peak pressure and temperature were not increased. Although a slight increase in the MSLB peak containment pressure does occur as a result of this proposed change, the revised value remains well below the containment design pressure. The margin of safety between the containment design pressure and the pressure at which the containment would ultimately fail is unchanged by this proposed amendment. Snerefore, it is concluded that there is no reduction in the margin of safety as described in the bases of any technical specification.
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CONTAINMENT PRESSURE RESPONSE TO POSTULATED j DEPSG LOCA WITH MINIMUM SAFETY INJECTION 2,3 50 ' - .. s
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. Attachment 11 to ET 91 0109 Page 1 of 3 ATTACIMENT 11 NII'IMT llAZARDS CONSIDERATION DETLM1NAyjogg I
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. At tactanent II to ET 91-0109 page 2 of 3 SICNIFICANT llAZARDS CONSIDERATION DETKFMINATION This proposed change has been reviewed per the standards provided in 30 CFR 50.92. Each standard is discussed separately below, i Standard 1 - Involves a Significant Increase in the Probability or Consequencen of an Accident Previously Evaluated.
This amendment request proposes to reduce the minimum cooling water flow requirement for the containment cooling system (CCS) for post-accident conditions and delete a surveillance requirement to verify a specified minimum flow rate in the normal operating mode of the CCS. The effect of the proposed reduction in the minimum required cooling water flow rate to the CCS has been evaluated and shown to have no significant effect on the consequences of postulated accidents. For a postulated Loss of Coolant Accident (LOCA), the calculated peak containment pressure is unchanged and the containment pressure response remains consistent with the assumptions used to evaluate the potential offsite radiological consequences of a LOCA due to leakage from the containment. For a postulated Main Steam Line Break (MSLB) event, the containment pressure remains well below the design pressure of the containment and safety related equipment located within the containment has been evaluated and found to remain fully qualified for the revised containment conditions. The deletion of the surveillance requirement for flow measurement during normal operations does not adversely affect the reliability of the CCS since ongoing performance monitoring of these heat exchangers provides more reliable indication of potential heat exchanger degradation. On these bases, it is concluded that there will be no significant increase in the probability of occurrence or consequences of previously evaluated accidents.
Standard II - Create the Possibility of a New or Different Kind of Accident From any Previously Evaluated.
The CCS serves to mitigate the consequences of postulated accidente, but is not associated with the initiation of design basis events and has no direct impact on the Reactor Coolant System or other structures or systems associated with the initiation of postulated accidents. Therefore, this proposed technical specification revision does not create the possibility of a new or different kind of accident from any previously evaluated.
I Standard III - Involve a Significant Reduction in the Margin of Safety.
l The proposed reduction in required cooling water flow has been shown to be acceptable based on revised accident analyses. For a LOCA tne containment I
peak pressure and temperature were not increased. Although a slight increase in the MSLB peak containt. ant pressure does occur as a result of
( this proposed change, the revised value remains well below the containment design pressure. The margin of safety between the containment design pressure and the pressute at which the containment would ultimately fail is unchanged by this proposed amendment. Therefore, it is concluded that there is no significant reduction in the margin of safety as described in the bases of any technical specification.
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. Attachment 11 to ET 91-0109 Page 3 of 3 i
Based on the above, the requested technical specification change does not involve a significant incresse in the probability or consequences of a i previously evaluated accident, create the possibility of a new or different kind of accident, or involve a significant reduction in the margin of safety. Therefore, the requested license amendment does not involve a significant hazards consideration in accordance with 10 CFR 50.92.
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- a. Attachment III t o ET 91 0109
- Page 1 of 2 l
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ATTACIMENT III ENVIRONMENTAL 1MPACT DETERMINATION l
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, Attachment III to ET 91-0109
+ Page 2 of 2 ENVIRONMENTAL IMPACT DETERMINATION This amendment request meets the criteria specified in 10 CFR 51.22(c)(9).
Specific criteria contained in this section are discussed below.
(1) the amendment involves no significant hazards consideration, As demonstrated in Attachment II, this proposed amendment does not involve any $1gnificant hatsrds conAiderations.
(11) there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.
The proposed change reduces the minimwn flow requirements for cooling water to the contaimnent cooling system undor post-accident conditions and deletes an unnecessary requirement to verify a specified minimum flow under normal i system alignment. The reduction in the minimum required flow to the l containment cooling system will increase the available service water system i flow rate margin. This will provide adequate margins for future flow balancing and redistribution to assure adequate flow is provided to the various safety related components served by the service water s ystem. There are no changes to the methods of operations as s result of this amendment and the system will continue to provide adequate flow for the required cooling of necessary systems and components.
(iii) there is no significant increase in individual or cumulative occupational radiat. ion exposure.
Other than the proposed reduction in minimum required flow to the containment cooling system, the proposed amendment does not affect the method of operation of the service water system or containment cooing system. The proposed amendment does not involve systems which contain radioactive materials and will have no affect on levels of radiation normally present in the facility. Therefore, there will be no significant
, increase in individual or cumulative radiation exposure associated with this
( proposed amendment.
1 i Based on the above, there will be no significant impact on the environment resulting from this change and the change meets the criteria specified in 10 CFR 51.22(c)(9) for a categorical exclusion from the requirements of 10 CFR 51.21 relative to a specific environmental assessment by the Commission.
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