ML19208B022: Difference between revisions
StriderTol (talk | contribs) (Created page by program invented by StriderTol) |
StriderTol (talk | contribs) (Created page by program invented by StriderTol) |
||
| (2 intermediate revisions by the same user not shown) | |||
| Line 14: | Line 14: | ||
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE, UTILITY TO NRC | | document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE, UTILITY TO NRC | ||
| page count = 3 | | page count = 3 | ||
| project = | |||
| stage = Request | |||
}} | }} | ||
=Text= | =Text= | ||
{{#Wiki_filter:.. | {{#Wiki_filter:. . | ||
20555 Re ference: Docket No. 50-285 Gentlemen : | n | ||
-,_L J_ | |||
. f- ' | |||
Omaha Public Power District 1623 HARNEY OMAHA, NESRASKA 66102 e TELEPHONE 536-4000 AREA CODE 402 Septe=ber 6, 1979 Director of Nuclear Reactor Regulation ATTN: Mr. Robert W. Reid, Chief Operating Leactors Branch No. 4 U. S. Nuclear Regulatory Co==ission Washington, D. C. 20555 Re ference: Docket No. 50-285 Gentlemen : | |||
The Omaha Public Pover District submitted an Application for A=end- | The Omaha Public Pover District submitted an Application for A=end- | ||
=ent of Facility Operating License, dated March 13, 1978, requesting the Cc==ission to revise Technical Specifications for the Fort Calhoun Staticn to per=it use of reduced air flow rates in the spent fuel pocl area and safety injecticn pu=p roc =. | =ent of Facility Operating License, dated March 13, 1978, requesting the Cc==ission to revise Technical Specifications for the Fort Calhoun Staticn to per=it use of reduced air flow rates in the spent fuel pocl area and safety injecticn pu=p roc =. On May 16, 1978, the District =et with men-bers of your staff to discuss this Application. As a result of the meet-ing, en March 6,1979, additional information was submitted to support our Application. The additional information provided resulte of analytical modeling and field testing which demonstrated the acceptability of pro-posed fiev rates. Subsequently, the District received further questicus frc= the staff via telephone co==unicatien which indicated that the analytical =odel provided by the District did not use a technique which was reviewed and approved by the Cc==ission and, as such, was not accept-able. In respcase to the staff's concerns, the District authorized Cc=bustion Engineering, Inc. to redo the analytical =odeling using an approved code. The results of this analysis are attached. | ||
On May 16, 1978, the District =et with men-bers of your staff to discuss this Application. As a result of the meet-ing, en March 6,1979, additional information was submitted to support our Application. The additional information provided resulte of analytical modeling and field testing which demonstrated the acceptability of pro-posed fiev rates. Subsequently, the District received further questicus frc= the staff via telephone co==unicatien which indicated that the analytical =odel provided by the District did not use a technique which was reviewed and approved by the Cc==ission and, as such, was not accept-able. In respcase to the staff's concerns, the District authorized Cc=bustion Engineering, Inc. to redo the analytical =odeling using an approved code. The results of this analysis are attached. | |||
Based on the results shown in the attached analysis , it is con-cluded that the safety injection pump roc = te=perature vill not exceed the 1170F peak reached at 55 =inutes for at least the first 27 hour period folleving a IDCA. By this time, the shutdevn ecoling system is put in operation to cool the centainment su=p water. Even in the worst case, assuming that the su=p te=perature does not decrease after 100,000 seconds and assu=ing zero ventilation flow, it is estimated that the roc = te=persture vill not exceed 122cF for at least 72 hours folleving the LOCA (based upcn linear extrapolation of Figure 2). | Based on the results shown in the attached analysis , it is con-cluded that the safety injection pump roc = te=perature vill not exceed the 1170F peak reached at 55 =inutes for at least the first 27 hour period folleving a IDCA. By this time, the shutdevn ecoling system is put in operation to cool the centainment su=p water. Even in the worst case, assuming that the su=p te=perature does not decrease after 100,000 seconds and assu=ing zero ventilation flow, it is estimated that the roc = te=persture vill not exceed 122cF for at least 72 hours folleving the LOCA (based upcn linear extrapolation of Figure 2). | ||
A reducticn in the centainment su=p water te=perature results in a reduction in the safety injection purp rec ='s heat loading, thereby centinuously reducing the rate of temperature increase. Additional assurance is provided by the conservative assu=ption that the centain-ment building air cooling and filtratien system is not in service | A reducticn in the centainment su=p water te=perature results in a reduction in the safety injection purp rec ='s heat loading, thereby centinuously reducing the rate of temperature increase. Additional assurance is provided by the conservative assu=ption that the centain-ment building air cooling and filtratien system is not in service \ | ||
during this interval. Normally this system would be in service to 0 O | |||
perform its designed engineered safety features functions, which would also reduce the containment building su=p te=perature. \ | |||
< s = f,)h q909180 h D | < s = f, ) h q909180 h D [ | ||
Director of Nuclear Reactor Regulation September 6, 1979 Page Two A remaining area of possible concern is an increase in rocm temperature due to pu=p motor heat. The analysis discussed above, which was performed to confirm the results of an analysis performed by the Omaha Public Pcwer District, clearly de=enstrated that room te=perature vill not exceed 11ToF for at least 27 hours after a LOCA. The results of both of these analyses are supported by tk results of an extensive testing program performed by the Distri and previously reported to the Nuclear Regulator / Cc==ission. All of these analyses and testing were performed assuming zero ventila-ticn air ficv through the pump roc =s. | |||
The analysis discussed above, which was performed to confirm the results of an analysis performed by the Omaha Public Pcwer District, clearly de=enstrated that room te=perature vill not exceed 11ToF for at least 27 hours after a LOCA. The results of both of these analyses are supported by tk results of an extensive testing program performed by the Distri and previously reported to the Nuclear Regulator / Cc==ission. | |||
All of these analyses and testing were performed assuming zero ventila-ticn air ficv through the pump roc =s. | |||
Obviously the operation of the safety injection pu=ps must be assured for periods of time much longer than the 27 hours discussed in the preceding paragraphs. This assurance is provided and de=on-strated by the folleving discussion. | Obviously the operation of the safety injection pu=ps must be assured for periods of time much longer than the 27 hours discussed in the preceding paragraphs. This assurance is provided and de=on-strated by the folleving discussion. | ||
A review of the attached Figure 2, " Pump Room Atmosphere Temper-ature (o?) vs. Ti=c", clearly shows that the pu=p rocm temperature reaches a peak of 'lToF at about 3500 seconds after the LOCA, rapidly decreases, then gradually increases, reaches a second peak at 70,000 seccnds, and then begins another decrease. By this time, a single safety injection pu=p is sufficient to cool the core. | A review of the attached Figure 2, " Pump Room Atmosphere Temper-ature (o?) vs. Ti=c", clearly shows that the pu=p rocm temperature reaches a peak of 'lToF at about 3500 seconds after the LOCA, rapidly decreases, then gradually increases, reaches a second peak at 70,000 seccnds, and then begins another decrease. By this time, a single safety injection pu=p is sufficient to cool the core. In addition , | ||
In addition , shutdown cooling is available to re=cve sump water heat. A steady-state calculatica has been performed (see Enclosure 2) using the conservative assumption that su=p water temperature does not de-crease below 165CF. Even under these conditions,1500 cfm of ventila-tion (less than 1/2 design flow) is sufficient to maintain the room temperature below 122cF. With zero ventilation flow available,122oF will not be reached until several days after the LOCA, allowing a=ple time for ventilation to be restored. | shutdown cooling is available to re=cve sump water heat. A steady-state calculatica has been performed (see Enclosure 2) using the conservative assumption that su=p water temperature does not de-crease below 165CF. Even under these conditions,1500 cfm of ventila-tion (less than 1/2 design flow) is sufficient to maintain the room temperature below 122cF. With zero ventilation flow available,122oF will not be reached until several days after the LOCA, allowing a=ple time for ventilation to be restored. | ||
The auxiliar/ building ventilation system consists of two supply fans , three exhaust fans , ducting, da=pers , and controls and instrument-ation.The equipment can be centrolled from the main centrol room. | The auxiliar/ building ventilation system consists of two supply fans , three exhaust fans , ducting, da=pers , and controls and instrument-ation. The equipment can be centrolled from the main centrol room. | ||
The fans are pcwered from the station's normal !.80 V power supply busses. However, under emergency conditions it is pcssible to power these units from the e=ergency power sources. The fans and ventila-tien ducting are designed and installed as seismic category 1 equip-ment."he power cabling for the fans is of the same quality as the plant's engineered safeguards cabling. Ecvever, this cabling has not Men designated nor identified as being engineered safeguards quality. | The fans are pcwered from the station's normal !.80 V power supply busses. However, under emergency conditions it is pcssible to power these units from the e=ergency power sources. The fans and ventila-tien ducting are designed and installed as seismic category 1 equip-ment. "he power cabling for the fans is of the same quality as the plant's engineered safeguards cabling. Ecvever, this cabling has not Men designated nor identified as being engineered safeguards quality. | ||
These features do provide a high quality ventilation system for safety injecticn 1.cp room cabling. | These features do provide a high quality ventilation system for safety injecticn 1.cp room cabling. The quality of this system has been de=cnstrated by over five years of reliable service. | ||
The quality of this system has been de=cnstrated by over five years of reliable service. | Based on the auxiliary building ventilation system design features and performance, there is an extremely high probability that at least one of the two supply fans and at least one of the three exhaust fans, r h f, L\ | ||
Based on the auxiliary building ventilation system design features and performance, there is an extremely high probability that at least one of the two supply fans and at least one of the three exhaust fans, h | |||
Director of Nuclear Reactor Regulation September 6, 1979 Page Three with asscelated ducting and controls, would be available to provide cooling to the safety injection.pu=p rooms. The analyses and testing discussed in preceding paragraphs clearly de=onstrate that adequate time is even available to effect repairs on the ventilat ion system and have equipment operable in a reasonable time period. The avail-ability of this system provides yet additional assurance that the safety injecticn pu=ps vill not be rendered incperable due to pump overheating. | |||
The analyses and testing discussed in preceding paragraphs clearly de=onstrate that adequate time is even available to effect repairs on the ventilat ion system and have equipment operable in a reasonable time period. The avail-ability of this system provides yet additional assurance that the safety injecticn pu=ps vill not be rendered incperable due to pump overheating. | |||
During the District's reviews and analyses of the safety in-jection pump roc ='s te=peratures following a LOCA, we identified a few items which would cptimize operation of the safety injection pu=ps and the auxiliary building ventilation system. We therefore propose to make the following modifications: | During the District's reviews and analyses of the safety in-jection pump roc ='s te=peratures following a LOCA, we identified a few items which would cptimize operation of the safety injection pu=ps and the auxiliary building ventilation system. We therefore propose to make the following modifications: | ||
(1)Install te=perature detectors , with readout and alarr.s , in the centrol rocm to =onitor safety injection pump room te=peratures. | (1) Install te=perature detectors , with readout and alarr.s , | ||
(2)Ensure that electrical pcVer and control cables to re-dundant supply and exhaust fans are physically and electrically independent of their redundant counter-p arts .( 3)Perfor= additional reviews and modify that portion of the safety injection pu7 roc ='s ventilation system as required to further assure the adequacy of that system to isolate or filter lodine which may be released to the pu=p roc =s after a LOCA. | in the centrol rocm to =onitor safety injection pump room te=peratures. | ||
(2) Ensure that electrical pcVer and control cables to re-dundant supply and exhaust fans are physically and electrically independent of their redundant counter-p arts . | |||
( 3) Perfor= additional reviews and modify that portion of the safety injection pu7 roc ='s ventilation system as required to further assure the adequacy of that system to isolate or filter lodine which may be released to the pu=p roc =s after a LOCA. | |||
The details of the modification proposed in item (3) vill be sub- | The details of the modification proposed in item (3) vill be sub- | ||
=itted to the Nuclear Regulatory Co==issicn for review prior to modi-ficaticns. | =itted to the Nuclear Regulatory Co==issicn for review prior to modi-ficaticns. | ||
The preceding discussions and attached analyses, plus previous infor=ation presented to the Nuclear Regulatory Co==ission on this issue, clearly demonstrate the adequacy of the existing equipment to perform the necessary functions in a post-LOCA situation. The proposed modifications will provide additional assurance of this adequacy.The Application for A=end=ent of Facility Operating License D?R-40, dated March 13, 1978, as amended by our sub=ittal dated March 6,1979, is consistent with the above-referenced dis-cussions and analyses. | The preceding discussions and attached analyses, plus previous infor=ation presented to the Nuclear Regulatory Co==ission on this issue, clearly demonstrate the adequacy of the existing equipment to perform the necessary functions in a post-LOCA situation. The proposed modifications will provide additional assurance of this adequacy. The Application for A=end=ent of Facility Operating License D?R-40, dated March 13, 1978, as amended by our sub=ittal dated March 6,1979, is consistent with the above-referenced dis-cussions and analyses. The District therefore respectfully requests that the Nuclear Regulatory Co==issica continue its review of that a=end=ent. | ||
The District therefore respectfully requests that the Nuclear Regulatory Co==issica continue its review of that a=end=ent. | Sincerely, W. C. Jcnes Division Manager g} h Production Operaticns ' | ||
Sincerely, | WCJ/KJM/BJH:j== | ||
Attach. | |||
Attach.ec: LeBeeuf, Lamb, Ieiby & MacRae | ec: LeBeeuf, Lamb, Ieiby & MacRae | ||
.__}} | . __}} | ||
Latest revision as of 07:02, 2 February 2020
| ML19208B022 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 09/06/1979 |
| From: | William Jones OMAHA PUBLIC POWER DISTRICT |
| To: | Reid R Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19208B023 | List: |
| References | |
| NUDOCS 7909180482 | |
| Download: ML19208B022 (3) | |
Text
. .
n
-,_L J_
. f- '
Omaha Public Power District 1623 HARNEY OMAHA, NESRASKA 66102 e TELEPHONE 536-4000 AREA CODE 402 Septe=ber 6, 1979 Director of Nuclear Reactor Regulation ATTN: Mr. Robert W. Reid, Chief Operating Leactors Branch No. 4 U. S. Nuclear Regulatory Co==ission Washington, D. C. 20555 Re ference: Docket No. 50-285 Gentlemen :
The Omaha Public Pover District submitted an Application for A=end-
=ent of Facility Operating License, dated March 13, 1978, requesting the Cc==ission to revise Technical Specifications for the Fort Calhoun Staticn to per=it use of reduced air flow rates in the spent fuel pocl area and safety injecticn pu=p roc =. On May 16, 1978, the District =et with men-bers of your staff to discuss this Application. As a result of the meet-ing, en March 6,1979, additional information was submitted to support our Application. The additional information provided resulte of analytical modeling and field testing which demonstrated the acceptability of pro-posed fiev rates. Subsequently, the District received further questicus frc= the staff via telephone co==unicatien which indicated that the analytical =odel provided by the District did not use a technique which was reviewed and approved by the Cc==ission and, as such, was not accept-able. In respcase to the staff's concerns, the District authorized Cc=bustion Engineering, Inc. to redo the analytical =odeling using an approved code. The results of this analysis are attached.
Based on the results shown in the attached analysis , it is con-cluded that the safety injection pump roc = te=perature vill not exceed the 1170F peak reached at 55 =inutes for at least the first 27 hour3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br /> period folleving a IDCA. By this time, the shutdevn ecoling system is put in operation to cool the centainment su=p water. Even in the worst case, assuming that the su=p te=perature does not decrease after 100,000 seconds and assu=ing zero ventilation flow, it is estimated that the roc = te=persture vill not exceed 122cF for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> folleving the LOCA (based upcn linear extrapolation of Figure 2).
A reducticn in the centainment su=p water te=perature results in a reduction in the safety injection purp rec ='s heat loading, thereby centinuously reducing the rate of temperature increase. Additional assurance is provided by the conservative assu=ption that the centain-ment building air cooling and filtratien system is not in service \
during this interval. Normally this system would be in service to 0 O
perform its designed engineered safety features functions, which would also reduce the containment building su=p te=perature. \
< s = f, ) h q909180 h D [
Director of Nuclear Reactor Regulation September 6, 1979 Page Two A remaining area of possible concern is an increase in rocm temperature due to pu=p motor heat. The analysis discussed above, which was performed to confirm the results of an analysis performed by the Omaha Public Pcwer District, clearly de=enstrated that room te=perature vill not exceed 11ToF for at least 27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br /> after a LOCA. The results of both of these analyses are supported by tk results of an extensive testing program performed by the Distri and previously reported to the Nuclear Regulator / Cc==ission. All of these analyses and testing were performed assuming zero ventila-ticn air ficv through the pump roc =s.
Obviously the operation of the safety injection pu=ps must be assured for periods of time much longer than the 27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br /> discussed in the preceding paragraphs. This assurance is provided and de=on-strated by the folleving discussion.
A review of the attached Figure 2, " Pump Room Atmosphere Temper-ature (o?) vs. Ti=c", clearly shows that the pu=p rocm temperature reaches a peak of 'lToF at about 3500 seconds after the LOCA, rapidly decreases, then gradually increases, reaches a second peak at 70,000 seccnds, and then begins another decrease. By this time, a single safety injection pu=p is sufficient to cool the core. In addition ,
shutdown cooling is available to re=cve sump water heat. A steady-state calculatica has been performed (see Enclosure 2) using the conservative assumption that su=p water temperature does not de-crease below 165CF. Even under these conditions,1500 cfm of ventila-tion (less than 1/2 design flow) is sufficient to maintain the room temperature below 122cF. With zero ventilation flow available,122oF will not be reached until several days after the LOCA, allowing a=ple time for ventilation to be restored.
The auxiliar/ building ventilation system consists of two supply fans , three exhaust fans , ducting, da=pers , and controls and instrument-ation. The equipment can be centrolled from the main centrol room.
The fans are pcwered from the station's normal !.80 V power supply busses. However, under emergency conditions it is pcssible to power these units from the e=ergency power sources. The fans and ventila-tien ducting are designed and installed as seismic category 1 equip-ment. "he power cabling for the fans is of the same quality as the plant's engineered safeguards cabling. Ecvever, this cabling has not Men designated nor identified as being engineered safeguards quality.
These features do provide a high quality ventilation system for safety injecticn 1.cp room cabling. The quality of this system has been de=cnstrated by over five years of reliable service.
Based on the auxiliary building ventilation system design features and performance, there is an extremely high probability that at least one of the two supply fans and at least one of the three exhaust fans, r h f, L\
Director of Nuclear Reactor Regulation September 6, 1979 Page Three with asscelated ducting and controls, would be available to provide cooling to the safety injection.pu=p rooms. The analyses and testing discussed in preceding paragraphs clearly de=onstrate that adequate time is even available to effect repairs on the ventilat ion system and have equipment operable in a reasonable time period. The avail-ability of this system provides yet additional assurance that the safety injecticn pu=ps vill not be rendered incperable due to pump overheating.
During the District's reviews and analyses of the safety in-jection pump roc ='s te=peratures following a LOCA, we identified a few items which would cptimize operation of the safety injection pu=ps and the auxiliary building ventilation system. We therefore propose to make the following modifications:
(1) Install te=perature detectors , with readout and alarr.s ,
in the centrol rocm to =onitor safety injection pump room te=peratures.
(2) Ensure that electrical pcVer and control cables to re-dundant supply and exhaust fans are physically and electrically independent of their redundant counter-p arts .
( 3) Perfor= additional reviews and modify that portion of the safety injection pu7 roc ='s ventilation system as required to further assure the adequacy of that system to isolate or filter lodine which may be released to the pu=p roc =s after a LOCA.
The details of the modification proposed in item (3) vill be sub-
=itted to the Nuclear Regulatory Co==issicn for review prior to modi-ficaticns.
The preceding discussions and attached analyses, plus previous infor=ation presented to the Nuclear Regulatory Co==ission on this issue, clearly demonstrate the adequacy of the existing equipment to perform the necessary functions in a post-LOCA situation. The proposed modifications will provide additional assurance of this adequacy. The Application for A=end=ent of Facility Operating License D?R-40, dated March 13, 1978, as amended by our sub=ittal dated March 6,1979, is consistent with the above-referenced dis-cussions and analyses. The District therefore respectfully requests that the Nuclear Regulatory Co==issica continue its review of that a=end=ent.
Sincerely, W. C. Jcnes Division Manager g} h Production Operaticns '
WCJ/KJM/BJH:j==
Attach.
ec: LeBeeuf, Lamb, Ieiby & MacRae
. __