ML18102B567
| ML18102B567 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 08/29/1997 |
| From: | Eric Simpson Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LR-N97457, NUDOCS 9709090279 | |
| Download: ML18102B567 (10) | |
Text
'1,',
Public Service Electric and Gas Company E. C. Simpson Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-1700 Senior Vice President - Nuclear Engineering AUG 2 91997 LR-N97457 United States Nuclear Regulatory Conunission Document Control Desk Washington, DC 20555 REQUEST FOR ADJ;)ITIONAL INFORMATION MARGIN RECOVERY LCR S94-41 SALEM GENERATING STATION NOS. 1 AND 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 Gentlemen:
Public Service Electric & Gas Company (PSE&G) has received the NRC Staff's request for additional information (RAI) sent by verbal request on May 12, 1997, electronic transmittal on June 5, 1997, and electronic transmittal on July 23, 1997, pertaining to PSE&G's submitted amendment request for the Margin Recovery Program for Salem Generating Station dated May 10, 1996.
PSE&G's responses to the request are attached.
The order in which they are provided are:
Questions 1 -
7 from 6/5/97, Question 8 from 5/12/97, and Questions 9 -
10 from 7/23/97.
Should you have any questions regarding this additional information, we will be pleased to discuss them with you.
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Document Control Desk l10R-N9,7 4 5 7 Attachment 2
C Mr. H. J. Miller, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. L. Olshan, Licensing Project Manager -
Salem U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 14E21 Rockville, MD 20852 Ms. M. Evans (X24)
USNRC Senior Resident Inspector Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering 33 Arctic Parkway CN 415 Trenton, NJ 08625 AUG 2 91997
REF: LR-N97457 STATE OF NEW JERSEY SS.
COUNTY OF SALEM E. C. Simpson, being duly sworn according to law deposes and says:
I am Senior Vice President - Nuclear Engineering of Public Service Electric and Gas Company, and as such, I find the matters set forth in the above referenced letter, concerning Salem Generating Station, Units 1 and 2, are true to the best of my knowledge, information and belief.
Subscribed and Sworn to before me this My Commission expires on
' 1997 ELIZABETH J. KIDD NOTARY PUBLIC OF NEW JERSEY My Commission Expires Al>ril 25, 2000
Document Control.esk Attachment LR-N97457 LCR S94-41 RAI SALEM GENERATING STATION UNIT NOS. 1 AND 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 REQUEST FOR ADDITIONAL INFORMATION MARGIN RECOVERY LCR S94-41 PSE&G provides the following responses to NRC requests.
- 1)
The margin recovery program (MRP) containment pressure analysis from WCAP-13839 does not take credit for the zirconium water reaction as an energy sourcer whereas the analysis from the Final Safety Analysis report (FSAR) does include this energy.
- a.
State the rationale for not including this energy in the latest analysis.
As stated in WCAP-13839 with respect to mass and energy release calculations, Zirc-water reaction heat was not considered because the clad temperature did not rise high enough for the rate of the Zirc-water reaction heat to be of any significance.
- b.
Characterize the energy associated with the zirc-water reaction as a percentage of the total energy released into containment.
Based on discussions with Westinghouse, an explicit treatment of this energy component for a similar NSSS resulted in a contribution of approximately 6 BTU.
Since total energy released into containment is on the order of lxl0 9 BTUs, energy from Zirc-water reaction represents approximately 0.0000006% of total.
This contribution to the total energy released is effectively zero.
- c.
What would be expected impact on the containment pressure and temperature response if the zirc-water reaction energy were included?
Based on the energy release described above, in response to l)b., there would be no significant impact on containment pressure and temperature response by including the zirc-water reaction.
Page 1 of 7
Document Control.sk Attachment LR-N97457 LCR S94-41 RAI
- 2)
There was no one-to-one comparison made between the MRP
- analysis and the pre-MRP (i.e. current licensing basis) analysis.
- a.
Please indicate and explain any differences in the aforementioned analyses.
Consider differences in input assumptions, initial conditions, computer codes, physical models, and final pressure and temperature results.
It is particularly important that the staff have an understanding of the modeling and input differences between the MRP and pre-MRP analysis and the impact of these differences on the pressure and temperature results.
In particular, specify the differences in increased safeguard delay times, reduced fan cooler heat removal, and reduced SI flow between the pre-MRP and MRP analyses.
Containment analyses are characterized by the methodologies used, the input assumptions and the results.
LOCA mass and energy release analyses provided in the submittal via WCAP-13839, are based on the NRC approved methodology of WCAP-10325-P-A.
Current analyses of record are based on the earlier NRC approved methodology (WCAP-8264-P-A).
WCAP-10325-P-A provides a description of the changes applicable to the methodology relative to WCAP-8264-P-A.
As stated in WCAP-10325-P-A, the newer mass and energy release methods produce lower containment response results, while retaining significant conservatism.
The containment response analyses provided in this submittal, as do the current analyses of record, utilize the "Containment Pressure Analysis Code",
COCO code (WCAP-8327).
Page 2 of 7
Document Control~esk Attachment LR-N97457 LCR S94-41 RAI Following is a summary of key inputs used in the submitted analysis and changes relative to the current licensing basis:
Input Assumption Break Size & Location-Initial Power Level-Initial RCS Avg Temp-Initial Cont. Temp-Initial Cont. Pressure-Safety Injection Flow Rate-Number of CFCUs-CFCU Heat Removal-CFCU Delay Time-Number of Cont. Spray Pumps-Containment Spray Flow-Containment Spray Delay Time-CFCU/Spray Actuation Setpoints-Containment Net Free Volume-Page 3 of 7 Change, LCR S94-41 Relative to Current Analyses No change Reduced, LCR uses licensed power level (3411 MWth) + 2% uncert Uncertainty allowance increased by 1 °F No change No change Min SI flow rate with 5 percent reduction in SI Pump performance No change Reduced by approximately 20%
No change No change No change Increased by 5 seconds No change No change
Document Control.esk Attachment LR-N97457 LCR 894-41 RAI Individual sensitivity studies were not performed to characterize the impact of each change in input assumptions.
However, the following is a comparison of the net results for the submitted analyses relative to the current analyses:
Current LCR 894-41 Analysis Peak Contmt Press (psig) 45.53 41.2 Peak Contmt Temp (°F) 269.2 263.1
- b.
The MRP peak containment pressure is 41.2 psig, whereas the pre-MRP peak pressure is 45.53 psig.
Specify/discuss the source(s) of this erosion in conservatism.
The reduction in containment pressure result in the MRP analysis is due to the application of the NRC approved mass and energy release methodology of WCAP-10325-P-A. Overall, the effect on containment response of the changes described above, is an increase in margin for containment temperature and pressure design limits.
- 3)
[Question retracted by NRC]
- 4)
Page 4-1 of WCAP-13839 implies that the double-ended pump suction break is the most limiting in terms of containment temperature.
Has the main steamline break been reanalyzed to determine the peak temperature? If this case has not been considered, provide justification.
If it has been considered, state the results, and discuss how environmental qualifications of equipment would be affected.
Containment temperature response to steam line rupture is analyzed and presented in the submittal under Section 4.1.18.
Results provided in that section for peak containment temperature response from a steam line break have been included within the current equipment qualification program. [Note: It is correct that, with respect to the LOCA containment response analyses, the double-ended pump suction is bounding for containment temperature. WCAP-13839 is included with the license change request in order to incorporate the improved Page 4 of 7
L Document Control.esk Attachment LR-N97457 LCR S94-41 RAI LOCA mass and energy release model into the Salem
.licensing basis; but only for LOCA containment response considerations.]
- 5)
Have any significant changes been made to the COCO code used for the MRP analysis relative to the version used for the pre-MRP analysis?
There have not been any significant modifications to the COCO code with respect to solution techniques.
The primary focus has been on code maintenance and operating platform changes.
- 6)
It is the staffs understanding that the most recent version of WCAP-10325 is May, 1983.
If a more recent version has been used, specify the differences in the two versions, and provide the rationale for not using the more updated version.
The submitted WCAP-13893 uses the LOCA mass and energy release methodology of WCAP-10325-P-A, dated May 1983.
- 7)
The staff recalls that, a few years ago, the modeling of the energy released from the steam generators was being considered by some utilities/Westinghouse.
Has the modeling of the steam generator energy released changed between the pre-MRP and MRP analyses? If so, state how and discuss its impact.
Yes, modeling of steam generator energy release has changed.
The differences between the MRP and pre-MRP steam generator energy release modeling are described in WCAP-10325-P-A sections 2.2.3, Steam Generator Fluid Exit Condition and 3.2, Post Blowdown Sensitivities.
The MRP methodology as described in WCAP-10325-P-A uses saturated conditions at the steam generator outlet, while the pre-MRP (WCAP-8264-P-A) used superheated steam.
As described in WCAP-10325-P-A, use of saturated steam at the steam generator outlet has been determined to be conservative" Page 5 of 7
Document Control.esk Attachment LR-N97457 LCR S94-41 RAI
- 8.
Does Margin Recovery affect LTOP/POPS?
And, if so,
- what is the nature and extent of the effect?
Current Salem LTOP design basis RCS peak pressure results from the mass input case of an inadvertent injection of an intermediate head SI pump.
The heat input case is not limiting.
The RCS peak pressure resulting from the mass input case was previously calculated to be 452 psig with no reactor coolant pumps in operation.
The effects of static head and two operating reactor coolant pumps (RCPs) are calculated to add 39 psig to the peak pressure, resulting in a total peak pressure of 491 psig.
Administrative controls prohibit operating more than two RCPs in MODE 5, Cold Shutdown (RCS temperature ~ 200°F)
The MRP revises Salem TS 5.4.2 to include the lower RCS volume of 12,446 +/- 426 CFT. The lower bound RCS volume is more limiting than the current RCS volume used in the LTOP analysis.
The MRP's effect on LTOP is not significant. The revised LTOP analysis, considering the effect of reduced RCS volume as a result of MRP, has resulted in a new calculated RCS peak pressure of 455 psig with no RCPs operating, and 494 psig including the effects static head and two RCPs operating.
This slightly higher RCS peak pressure remains within the limits of the current Unit 1 TS 3/4.4.9 and Unit 2 TS 3/4.4.10, when ASME Code Case N514 is applied.
NRC approval to apply ASME Code Case N514 at Salem Generating Station was granted by NRC letter dated February 13, 1995.
- 9.
PSE&G should provide a statement indicating that all components of the Salem 1/2 reactor coolant loop piping and supports meet all licensing basis ASME Section III design requirements.
The licensing basis design requirements for the Salem Units 1 and 2 Reactor Coolant Loop Piping was ASA B31.1.-1955 for Unit 1 and UASA B31.l.0-1967 for Unit
- 2.
The only piping in these units with ASME Section III as the licensing basis is the Pressurizer Surge lines.
The piping evaluation performed during the Margin Recovery Program confirmed that the existing analyses results continue to be applicable for both the Reactor Coolant Loop Piping and the Pressurizer Surge Line.
Page 6 of 7
Document Control.esk Attachment LR-N97457 LCR S94-41 RAI
- 10.
PSE&G should also provide a statement indicating that
'~
- Salem 1/2 meets the ASME Section III requirement which states that the design temperature shall not be less than the actual maximum metal operating temperature.
As provided in Section 5.2.1 of the Salem UFSAR, the reactor coolant piping design temperature is 650°F.
The operating conditions proposed as part of the margin recovery program are less than the piping design temperature.
Page 7 of 7