ML20249C327
| ML20249C327 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 06/22/1998 |
| From: | Krich R COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-97-04, GL-97-4, NUDOCS 9806290059 | |
| Download: ML20249C327 (5) | |
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1-Commonweahh Ediwn Company 14c3 Opus Place Downers Grove, IL 60515-5701 l
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June 22,1998 U. S. Nuclear Regulatory Commission Washington, D.C 20555 ATTN: Document Control Desk
Subject:
Response to Request for Additional Information Regarding Generic Letter 97-04
" Assurance of Suflicient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pumps"
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Byron Nuclear Power Station Units 1 and 2 l
Facility Operating License Nos. NPF-37 and NPF-66 I
NRC Docket Nos. 50-454 and 50-455 Braidwood Nuclear Power Station Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. 50-456 and 50-457
References:
1.
J. Hosmer (Comed) letter to NRC, dated January 5,1998, transmitting the "90-day Response to NRC Generic Letter 97-0, 2.
J. Hickman (NRC) letter to O. Kingsley (Comed) dated May 22,1998, Request for Additional Information for Braidwood Station Units I and 2; and Byron Station Units 1 and 2.
I In reference 1, the Commonwealth Edison (Comed) Company submitted the requested 90-day ll l
response to Generic Letter (GL) 97-04, " Assurance of Suflicient Net Positive Suction Head for l
Emergency Core Cooling and Containment Heat Removal Pumps." In reference 2, the Nuclear j
Regulatory Commission (NRC) transmitted a Request for Additional Information (RAI) concerning j
the Byron and Braidwood response. This response, which was requested to be submitted within 30 days of May 22,1998, i.e., by June 22,1998, is provided in Attachment 1.
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n 9906290059 99062244y DR ADOCK 0
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' U.S. Nuclear R:gul: tory Commission June 22,1998 l
Additionally, Comed is updating the 90-day response to Question 2 of Generic Letter 97-04, as l
provided in Reference 1. This revised response is provided in Attachment 2 to this letter. The i
responses to Questions 1,4, and 5 of the 90-day response remain unchanged. The response to l
Question 3 is modified by the response to RAI Question 2 stated above.
If you need any additional information concerning this response, please contact Ms. Denise Saccomando (630)663-7283.
Sincerely, R. M.
rich Vice President - Regulatory Services
- Attachments cc:
Regional Administrator-RIII Byron Senicr Resident Inspector Braidwood Senior Resident Inspector Office ofNuclear Safety-IDNS l
o:genenc gl:LIC-98451 Aac (x
l ATTACHMENT 1 Response to Request for Information for I
Braidwood Nuclear Power Station, Un'.ts 1 and 2 and j
Byron Nuclear Power Station, Units 1 and 2 1
1 The following information is provided in response to the NRC Request for Additional Information dated May 22,1998.
1.
What is the maximum sump temperature assumed in the net positive suction head
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(NPSH) analyses?
J Response: The maximum sump temperature assumed in the current analysis of record for Byron,and Braidwood is 290 F.
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2.
In response to Question 3 it is stated that, "neither the methodology for calculating NPSH nor the acceptability of the analyses as documented in the above table have j
changed from that which was previously reviewed and approved by the NRC." This
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could imply that the inputs to the analyses have changed since it was last reviewed and '
I approved by the staff. If this is correct, specify the differences between the reviewed
'l and approved analyses and the current analyses, and whether the change was done I
under 10 CFR 50.59.
Response: The analysis performed for Updated Final Safety Analyses Report (UFSAR)
Sections 6.5.2.2 and 6.1.2 are the current analyses of record and was incorporated into the UFSAR under 10 CFR 50.59. The inputs and assumptions regarding NPSH have not changed from that previously reviewed other than the assumed percent sump blockage (i.e.
50% to 81%).
A new analysis which utilizes a conservatively assumed minimum Containment flood level to determine debris transport and effects on the Containment Sump is currently under engineering review. The analysis will be incorporated into the Byron and Braidwood UFSARs in a future revision, after being evaluated in accordance with 10 CFR 50.59.
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f TTACHMENT 2 Supplemental Response to Request for Information for Braidwood Nuclear Power Station, Units 1 and 2 and Byron Nuclear Power Station, Units I and 2 The following is supplemental information to the 90-day response letter dated January 5,1998, to Generic Letter 97-04, " Assurance of SuHicient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pumps."
Containment Sorav (CS) and Residual Heat Removal (RHR) Pumps The recirculation mode of operation gives the limiting NPSH requirement for these pumps. In the original analysis, the available NPSH was determined from the containment floor level relative to the pump elevations and the pressure drop in the suction piping from the Emergency Core Cooling System (ECCS) sump to the pumps. Consistent with the guidance provided in NRC Regulatory Guide 1.82, Revision 0, " Sumps for Emergency Core Cooling and Containment Spray Systems,"
this calculation assumed the ECCS sump screen to be 50 percent blocked and determined available NPSH for the CS and RHR pumps to be 26.8 ft and 25.1 ft, respectively. The required NPSH for the CS and RHR pumps is 22.5 ft and 19 ft, respectively.
A subsequent analysis was performed to evaluate the impact of undocumented / unqualified coatings on available NPSH for the CS and RHR pumps. This calculation determined that the resulting ECCS sump screen blockage of approximately 81 percent had negligible impact on NPSH. The available NPSH for the CS and RHR pumps was determined to be 26.6 ft versus 22.5 fl required and 24.9 fl versus 19 ft required, respectively.
Pump NPSII Required (ft)
NPSH Available(ft)
CS 22.5 26.6 RH 19 24.9 Chemical Volume (CV) and Safety Iniection (SI) Pumos The end of the injection mode of operation gives the limiting NPSH requirement for these pumps.
The available NPSH was determined from the elevation head and vapor pressure of the water in the Refueling Water Storage Tank (RWST) and the pressure drop in the suction piping from the RWST to the pumps. From Updated Final Safety Report (UFS AR) Table 6.3-1, " Emergency Core Cooling System Component Parameter," the required and available NPSH for the CV and SI pumps are unchanged from the January 5,1998 response, and are as follows.
o generegl.UC-98-051. doc
P7mp NPSH Required (ft)
NPSH Avcilable(ft)
CV 21 35 SI 25 37 e
j Recent Engineering reviews have identified inconsistencies between the UFSAR values and the various supporting calculations. Adequate NPSH has been confirmed for all cases; however, revision of the calculations is ongoing. Upon completion of this effort, the UFSAR will be revised under the provisions of 10 CFR 50.59, as appropriate, to be consistent with the design calculations.
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