ML20209G390
| ML20209G390 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 07/09/1999 |
| From: | Jamie Benjamin COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-96-06, GL-96-6, NUDOCS 9907190151 | |
| Download: ML20209G390 (6) | |
Text
Commonw calth li!Mn Comp.tn}
j Im.ille Generating station
+
j l
261Ii North 2Ist Ro.nl M.nrweilles,11. 613 i!-97s7
=
Tri sisesveni l
July 9,1999 1
United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 j
1 LaSalle County Station, Units 1 and 2 Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374
Subject:
Commonwealth Edison Company (Comed) Response to Nuclear Regulatory Commission (NRC) Generic Letter (GL)96-06," ASSURANCE OF EQUIPMENT OPERABILITY AND CONTAINMENT INTEGRITY DURING DESIGN-BASIS j
ACCIDENT CONDITIONS," dated September 30,1996.
J
References:
1.
NRC Generic Letter (GL) 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions", dated September 30, 196.
2.
F.R. Dacimo (Comed) Letter to the NRC Document Control Desk, " Supplemental Response to NRC Generic Letter (GL) 96-06" dated July 29,1998.
3.
W. T. Subalusky Letter to U. S. NRC, Response to NRC Generic Letter 96-06 dated, June 4,1997.
4.
NRC Letter,"Information Pertaining to Comed 1
Implementation of Modifications Associated with Generic Letter 96-06," dated February 6,1998.
In Response to the United States Nuclear Regulatory Commission (NRC)
Generic Letter 96-06 (Reference 1), Commonwealth Edison Company (Comed) committed, to perform corrective actions prior to restart of LaSalle Unit 2 from refuel outage L2R07. The purpose of this letter is to inform the NRC of the status of these commitments, and to request NRC concurrence for the use of ASME Section Ill Appendix F acceptance criteria to i
pe.manently qualify Unit 1 and 2 Penetrations M-49 and M-50.
'}i I 9907190151 990709 PDR ADOCK 05000373 P
PDR A l'nkom Compan>
7 July 9,1999 '
U.S. Nucle:r R:gulatory Commission O
Page 2 3
(Comed Letter (Reference 2), dated July 29,1998, provided the status of the Comed commitments regarding the same penetrations on LaSalle Unit 1. We have implemented the replicate modifications for LaSalle Unit 2, as described in Attachment A.
Also, for Penetrations M-49 and M-50, our initial response to the GL 96-06 (Reference 3) indicated that these penetrations were susceptible to GL 96-06 overpressurization conditions. After further evaluation and in accordance with the guidance provided by the NRC (Reference 4), we have performed detailed analysis of the piping and components and have verified that these penetrations meet the requirements using ASME Appendix F acceptance criteria. These evaluations are further described in Attachment B.
In conclusion, LaSalle County Station has complied with the requirements of NRC Generic Letter 96-06. We are requesting the NRC review our use of ASME-Section lli Appendix F acceptance criteria to permanently qualify Unit 1 and 2 Penetrations M-49 and M-50 for GL 96-06 overpressurization conditions. We would like approval to occur by October 1,1999 to suppod our Unit 1 L1 R08 Refuel Outage. Upon receipt of NRC approval, we will update our UFSAR to
)
include the use of ASME Section ill Appendix F for this application.
Should you have any questions conceming this letter, please contact Frank A. Spangenberg, Ill, Regulatory Assurance Manager, at
. (815) 357-6761, extension 2383.
Respectfully,
-R-Jeffrey A. Benjamin Site VI.:e President LaSalle County Station Enclosure cc:
Regional Administrator-NRC Region ill NRC Senior Resident inspector - LaSalle County Station 1
)
4
July 9,1999 '
U.S. Nuclear Regulatory Commission I
Page 3 Attachment A Summary of the status of the implementation of Design Changes for GL 96-06.
Penetration M-36: Reactor Recirculation Sample Line Pioina.
For LaSalle Unit 1, Design Change Package (DCP) 9700343 was implemented to add a bypass line with a new check valve around the inboard containment isolation valve. The new check valve is positioned such that the trapped fluid between the containment isolation valves would be relieved to the larger volume of the Reactor Recirculation B Loop, which can not be isolated from the reactor vessel.
DCP 9700344 implemented a similar design change for penetration M-36 in Unit 2.
Penetration M 30: Suction Pipina to the Reactor Water Cleanuo (RT) Recirculation Pumo.
For LaSalle Unit 1 Design Change Package (DCP) 9700518 was implemented to drill i
a vent hole in the inboard disc of the inboard containment isolation valve. Based on a calculation performed by the original valve manufacturer, the outboard disc would unseat at a pressure less than the piping maximum stress code allowable and provide a pressure relief path to the reactor vessel. The vent hole size selected was compatible with the relief capacity of the isolated water volume following a LOCA.
The valve function as a containment isolation valve will still be met because the 1
leakage boundary of the outboard disc is maintained and will be ensured by the leak rate testing requirements of 10CFR50 Appendix J.
DCP 9700525 implemented a similar design change to the inboard containment isolation valve for the Unit 2 Reactor Water Cleanup System to protect the Unit 2 M-30 penetration.
Penetrations M-16 and M-17: Supolv & Retum to the Reactor Buildina Closed Coolina Water Pipina for the Seals of the Reactor Recirculation Pumo.
l For LaSalle Unit 1, Design Change Package (DCP) 9700339 was implemented to add a relief valve between each penetration and its corresponding inboard containment isolation valve. The new relief valve discharges the relief capacity directly to the containment atmosphere away from any safety related components.
l l
DCP 9700340 was implemented to provide similar relief valves for penetrations M-16 and M-17 in Unit 2.
July 9,1999 U.S. Nuclear Regulatory Commission Page 4 i
Penetrations M-25. M 26; M-27. and M-28: Supolv & Retum to the Chilled Water Pioina for the Primary Containment Ventilation Heat Exchanaer Coils.
For LaSalle Unit 1, Design Change Packages (DCP) 9700335 and 9700336 were implemented to add a relief valve between each penetration and its corresponding inboard containment isolation valve. The new relief valve discharges the relief capacity directly to the containment atmosphere away from any safety related component.
DCP 9700337 was implemented to provide similar relief valves for penetrations M-26 and M-28 in Unit 2. DCP 9700338 was implemented to provide similar relief valves for penetrations M-25 and M-27 in Unit 2.
4 Penetration M-7: Residual Heat Removal (RHR) Pipino for Shutdown Coolina Mode RHR Pumo Suction Pioina from Reactor Recirculation.
For LaSalle Unit 1, Design Change Package (DCP) 9700341 was implemented to add a relief valve between the penetration and the inbcard containment isolation valve. The new relief valve discharges the relief capacity directly to the containment atmosphere away from any. safety related components.
DCP 9700342 has been implemented to add a similar relief valve for penetration M-7 in Unit 2.
l' l-
p L
~
July 9,1999 l-
_ U.S. Nuclear Regulatory Commission L
Page 5 Attachment B
. GL 96-06 Penetration M-49 and M-50 Qualification using ASME Section Ill,
- Appendix F; Reactor Recirculation System Flow Control Valve 1(2)B33-F060A/B Hydraulic Piping.
As outlined in NRC Generic Letter 96-06, thermally induced overpressurization of isolated water-filled piping sections in containment could jeopardize the ability of accident-mitigating systems to perform their safety functions and could also lead to a breach of containment integrity via bypass leakage. This thermally induced overpressurization would occur during conditions where the containment is isolated and subsequently heats up (e.g., LOCA). An overpresure condition can occur in isolated penetrations. Calculations were performed to qualify Unit 1 and 2 Primary Containment Penetrations M-49 & M-50, Reactor Recirculation (RR)
Flow Control Valve 1(2)B33-F060A/B Hydraulic Piping, for overpressurization concems resulting from NRC Generic Letter 96-06.
Penetrations M-49 and M-50 assemblies were qualified using ASME Section lli Appendix F,1974 Edition. The hydraulic lines' piping, isolation valves, and pipe supports for the piping located between the inboard isolation valves and the primary containment wall were qualified using the 1989 version of Appendix F.
i LaSalle County Station is committed to the 1974 Edition of ASME B&PV Section lli Appendix F. However, based on the results of the comparison of the 1974 and 1989 Editions that follow, it was decided that the more conservative 1989 Edition of Appendix F should be used for the Penetrations M-49 and Ms50 piping and valve calculations i
L A comparison of the 1989 and 1974 Editions of ASME B&PV Section lli Appendix F shows that the 1989 Edition of Appendix F contains more information on terminology and methods of analysis. The acceptance criteria of the 1989 Edition of Appendix F has additional requirements for the analysis methodologies used in the piping and valve qualifications, making it more restrictive than the acceptance criteria of the 1974 Edition of Appendix F. For valve qualification, Aiticle F-1341.2 of the 1989 Edition of Appendix F has additional limits for the maximum primary stress intensity and the average primary shear stress which
.are not in the 1974 Edition of Appendix F. For piping, Article F-1430 (b) of the l~
1989 Edition of Appendix F requires the primary stress to be the lesser of 3.0 Sm (stress intensity value) or 2 Sy (yield strength) which is more restrictive than just less than 3.0 Sm in the 1974 Edition of Appendix F.
It was determined that an ana!ytical approach to address the overpressurization issue associated with these penetrations would be 'a more practical solution than L
a physical design change. These penetrations contain Fyrquel Hydraulic Fluid.
L i
inadvertent actuation of a relieving device would relieve Fyrquel Hydraulic fluid I
1 July 9,1999 U.S. Nuclear Regulatory Commission Page 6 Inside the drywell. Additionally, modifying the Reactor Recirculation (RR) Flow Control Valve System to include relief valves either inside or outside the containment could adversely affect the normal operation of the Reactor Recirculation Flow Control Valve and introduce a RR Flow Control Valve I
transient if they actuate.'
In conclusion the safety related portions of the RR FCV Lines were analyzed to
]
demonstrate that the piping and components remain within acceptable stress limits for the thermally induced pressures that would result from a postulated LOCA.
i i
4
.