ML18102A930
| ML18102A930 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 03/19/1997 |
| From: | Storz L Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LR-N97040, NUDOCS 9703250132 | |
| Download: ML18102A930 (11) | |
Text
Public Service Electric and Gas Company Louis F. Storz Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-5700 Senior Vice President - Nuclear Operations MAR 191997 LR-N97040 Un~ted States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 REQUEST FOR ADDITIONAL 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) dated December 18, 1996 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 submitted request are attached.
Should you have any questions regarding this additional information, we will be pleased to discuss them with you.
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9703250132 970319 PDR ADOCK 05000272 P
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Printed on W
Recycled Paper Sincerely,
Document Control Desk LR-N97040 Attachment 2
C Mr. H. J. Miller, Administrator - Region I U. S. Nuclear Regulatory Commissioh, 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 Mr. C. Marschall (X24)
USNRC Senior Resident Inspector Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering 33 Arctic Parkway CN 415 Trenton, NJ 08625 MAR 191997 95-4933
REF: LR-N97040 STATE OF NEW JERSEY SS.
COUNTY OF SALEM L. F. Storz, being duly sworn according to law deposes and says:
I am Senior Vice President -
Nuclear Operations 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 : /9th day of 'fr;~, 1997
~~~qt, NotaryPU1}~f New Jersey KIMBERLY JO BROWN NOTARY PUBLIC OF NEW JERSEY My Commission expires on ___
M_v_c_o1_m_11is_si_on_E_*xp_ire_s_A_pr_il 2_1_
- . _19_9_8 ___
MAR 191997 Document Control Desk LR-N97040 3
SRM/mrh BC Senior Vice President - Nuclear Engineering (N19)
General Manager -
Salem Operations (SOS)
Director -
QA/NSR (XOl)
Manager -
Nuclear Business Relations (N28)
Manager -
Salem Operations (SOl)
Manager -
System Engineering -
Salem (S02)
Manager -
Nuclear Safety Review (N38)
Salem Mechanical Engineering Manager (N51)
Onsite Safety Review Engineer ~ Salem (X15)
P. Cowan J. O'Connor T. Ross M. Van Noy General Solicitor, E. Selover (Newark, 5G)
Mark J. Wetterhahn, Esq.
Records Management (N21)
Microfilm Copy Files Nos. 1.2.1 (Salem) 2.3 (LCR S94-41)
Document Control Desk LR-N97040 LCR S94-41 RAI Attachment 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 the NRC requested items.
- 1)
Section 4. 0, "Accident Analysis" - Please provide discussion on any computer code used in the transient and accident analyses which are not approved by NRC staff.
Response
The computer codes used in the Margin Recovery Program safety analyses are part of the current licensing basis for Salem.
Therefore, no new codes were used that require NRC approval.
- 2)
Section 4.1.1, "Uncontrolled RCCA Bank Withdrawal From a Subcri ti cal Condi ti on" - Please provide transient DNBR curve to demonstrate that the criterion of the MDNBR is met during this transient.
Response
A transient DNBR curve was not generated for the Rod Withdrawal from Subcritical transient.
Bounding statepoint data determined from transient analysis is used to calculate the limiting DNBR. The curves provided in the submittal are consistent with the information presented in the current UFSAR licensing basis, and illustrate the transient analysis results for key parameters during this event.
Thus, PSE&G currently has no plans to create a transient DNBR curve.
(See UFSAR -
Fig 15.2-1, 2, and 3)
- 3)
Section 4.1.3, "Rod Cluster Control Assembly Misalignment" -
Please provide transient DNBR curve to demonstrate that the criterion of the MDNBR is met during this transient.
Response
Transient DNBR curves are not generated for the events listed in section 4.1.3.
The curves presented for the dropped rod event are consistent with those presented in the Salem UFSAR.
For the static rod misalignment, the case is reviewed to ensure that the DNBR limit is not violated based on the F~H values.
This is also consistent with the methodology used in the current licensing basis for this event.
For the transient rod misalignment (i.e., dropped Page 1 of 7
Document Control Desk Attachment LR-N97040
, LCR S94-41 RAI rod), bounding statepoint data determined from transient analysis is used to calculate the limiting DNBR.
The curves provided in the submittal are consistent with the information presented in the current UFSAR licensing basis analyses, and illustrate the transient analysis results for key parameters during this event.
Thus, PSE&G currently has no plan to create a transient DNBR curve.
(See UFSAR - page 15.2-15 and Fig 15.2-11 sheets 1 & 2, for the dropped rod event only)
- 4)
Section 4.1.4, "Uncontrolled Boron Dilution" - Please provide the results of an analysis to demonstrate sufficient times are available between the time of the alarm and the time of lost shutdown margin for all modes of plant operation per the SRP 15.4.6.
Response:.
To cover all phases of plant operation, boron dilution during refueling, startup, and power operation (Operational Modes 1, 2 and 6) are considered in this analysis.
The criterion used is the time from the start of dilution until the time of lost shutdown margin. Salem is not a Standard Review Plan plant, and does not explicitly analyze modes 3, 4, and 5.
This is consistent with the current UFSAR licensing basis analyses.
(See UFSAR - page 15.2-19)
- 5)
Section 4.1.5.3, "Single Reactor Coolant Pump Locked Rotor and Reactor Coolant Pump Shaft Break" - It is indicated that less than 5% of the total fuel rods experience DNB during a locked rotor event.
Please confirm that in your evaluation, all fuel rods with a transient DNBR less than 1.34 are assumed experiencing DNB and fuel failure.
Using the amount of fuel failure determined above, provide the results of an analysis to demonstrate that the radiological consequences are within 10 CFR Part 100 guidelines.
Response
In the Locked Rotor analysis, all fuel rods below the DNBR limit are assumed to experience DNB.
The 1.2% of fuel rods shown to be below the DNBR limit in our analysis is less than the 5% failed fuel rods assumed in the 10 CFR Part 100 dose calculation.
The 5% rods-in-DNB input to the dose calculation is included in our current radiological dose assumptions.
This can be confirmed by review of the information submitted to the Staff on October 17, 1996, ref.
letter LR-N96318, pertaining to this event.
Page 2 of 7
Document Control Desk Attachment LR-N97040 LCR S94-41 RAI
- 6)
Section 4.1.8, "Loss of Normal Feedwater" and Section 4.1.9, "Loss of Offsi te Power" -
The PORV 's were assumed operable during these transients.
However, the technical specification allows power operation with PORVs isolated.
Please provide the results of analyses considering PORVs inoperable.
Response
The PORVs are assumed to function only if modeling them creates more severe transient results.
Since this analysis is not a limiting pressurization transient, the PORVs are modeled.
Modeling the PORVs gives a more limiting pressurizer water volume.
This is consistent with the current UFSAR licensing basis analyses for Loss of Normal Feedwater and Loss of Offsite Power. (See UFSAR -
Fig 15.2-28, 4th plot of pressure shows pressure limited to 2350 which corresponds to a PORV set pressure, not a pressurizer safety valve set pressure.)
- 7)
Section 4.1.10, "Excessive Heat Removal Due to Feedwater System Malfunctions" - In the assessment of this section, it is indicated that this transient is less limiting than the excessive load increase evaluated in Section 4.1.11.
However, the results of an excessive load increase is not presented in Section 4.1.11.
Please provide the needed analysis results.
Response
Two Feedwater Malfunction events are analyzed; a feedwater heater bypass event and an increase in feedwater flow event.
Only the feedwater heater bypass event is bounded by the Excessive Load Increase response.
For the increase in feedwater flow events, the results are presented in the license change request submittal.
This is consistent with the current licensing basis in the UFSAR for Feedwater Malfunctions.
The results of the Excessive Load Increase are not presented in the submittal due to the non-limiting nature of the transient.
This transient does not require the generation of a reactor trip for mitigation.
This is also consistent with the current UFSAR.
(See UFSAR - page 15.2-45, ELI Fig 15.2-30, 32, 35, and 37 demonstrate nuclear power which shows that no reactor trip occurs.)
Page 3 of 7
Document Control Desk Attachment LR-N97040 LCR S94-41 RAI
- 8)
Section 4.1.13, "Main Steam System Failures" - Please provide transient DNBR curves for the accidental depressurization of main steam system and main steam line break events to demonstrate that the acceptance criteria of these events are met.
Response
DNBR curves were not generated for the Main Steam System failure events. Bounding statepoint data determined from transient analysis is used to calculate the limiting DNBR.
The curves provided in the submittal are consistent with the information presented in the current UFSAR licensing basis analyses and illustrate the transient analysis results for key parameters during this event.
Thus, PSE&G currently has no plan to create a transient DNBR curve.
(See UFSAR -
Fig 15.2-41, 42, 43a, 43b, and 43c Fig 15.4-49 through 53f)
- 9)
Section 4.1.14, "Spurious Operation of the SIS at Power" -
Please address the effect of this event regarding potential solid pressurizer. (concern raised in Westinghouse NSAL 013).
Response
The Inadvertent Safety Injection analysis and the issue raised in NSAL-93-013 are not impacted by the requested Technical Specification changes described in the submitted license change request.
[The issue raised in NSAL-93-013 involves non-conservative input assumptions that affect the results of pressurizer filling due to an Inadvertent SI transient.
Pressurizer filling is not affected by the requested Tech Spec Changes. Inadvertent SI is mentioned because it is also analyzed for DNBR (even though it is a non-limiting DNBR event).
Since it is analyzed for DNBR, it employs the RTDP methodology which is pending approval with this submittal.]
- 10) Section 4.1.15, "Single Rod Cluster Control Assembly Withdrawal at Full Power" - It is indicated that the results of this transient may cause fuel failure.
However, this is a condition II event (SRP 15.4.2) and no DNB is allowed for this transient.
Please discuss the acceptability of this analysis.
Response
For PWRs, Section 15.4.2 of the SRP applies to Bank Withdrawal at Full Power which is analyzed as a Condition II event in the UFSAR and in the submittal.
For Page 4 of 7
Document Control Desk Attachment LR-N97040 L~S94-41 RAI a Single RCCA Withdrawal at Full Power, Salem's UFSAR presently reflects this event as a Condition II event in light of Rod Control System deficiencies that were identified in 1993 (ref. Generic Letter 93-04 and Salem Amendments 144/122).
Salem has implemented corrective actions that enable the Single RCCA Withdrawal at Full Power to be returned to the original licensing basis, i.e., a Condition III event.
Revision to the Salem UFSAR is to be completed prior to restart of the Salem Units.
The acceptance criteria for the Single RCCA Withdrawal at Full Power used in the submittal is consistent with that of Condition III events, thus, it is acceptable.
- 11) Section 4.1.16, "Major Rupture of a Main Feedwater Line" -
Please provide the results of an analysis assuming PORVs inoperable.
This is because the technical specification allows power operation with PORVs isolated.
Response
Similar to item 6, the PORVs are assumed to function only if modeling them creates more severe transient results.
Since the Feedline Break analysis is not a limiting pressurization transient, the PORVs are modeled.
Modeling the PORVs results in less margin to the acceptance criteria.
The analysis presented in the submittal is the current UFSAR licensing basis Feedline Break analysis. (See UFSAR -
Fig 15.4-60b)
- 12) Section 4.1.17, "RCCA Ejection" -
The acceptance criteria of this event are specified in SRP 15.4.B.
Specifically, the transient peak system pressure is below 110% of design pressure and the radiological consequences are well within the 10 CFR Part 100 guidelines.
Please discuss the results of the analysis with respect to these acceptance criteria.
Response
Pressure during the Rod Ejection event is generically addressed in WCAP 7588, Rev lA, "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Method," dated December 1971 and as reviewed by the staff, ref. letter dated August 28, 1973 from D. B. Vassallo to R. Salvatori.
The results of the current dose assessments performed for the Rod Ejection Accident remain within the 10 CFR 100 guidelines for radiological consequences.
Results presented in the submittal are consistent with those presented in the current UFSAR licensing basis.
Page 5 of 7
Document Control Desk Attachment LR-N97040 LCR S94-41 RAI
- 13) Section 4.3, "Steam Generator Tube Rupture" -
Please describe the limiting single failure assumed in this analysis.
Response
The Steam Generator Tube Rupture event does not assume a single failure.
The methodology used is consistent with that presented in the current UFSAR licensing basis and with other Westinghouse plants that are of Salem's generation. (UFSAR presentation of SGTR page 15.4-49 through
- 56)
- 14)
Page 7, 8 lines from the top.
A reference is made to one T-hat RTD.
Should the correct number be three or two depending on methodology for a failed T-hot RTD?
One RTD would appear to be using the bypass manifolds not RTD bypass.
If only one RTD is used then the CSA for the electronics may be ambitious.
Table 2 states RTD used as three.
Response: Line eight of page seven is not correct.
The calculation utilizes all three T-hot RTDs as defined in Table 2.
- 15)
Page 8, Table 2.
RMTE is assumed to be 0.
Do plant procedures and available test equipment support this assumption?
Response
Yes.
Plant specific procedures along with the maintenance and test equipment specified in the procedures were reviewed and the resulting impact for RMTE is insignificant, (i.e., less than one tenth of the calibration tolerance).
- 16)
Page 8, Table 2.
Is hot leg streaming included in the uncertainty for T-average?
Is it included in hot leg enthalpy, Table 5, page 18?
Response
Yes.
Table 2 includes a term "PMA", Process Measurement Accuracy, which is comprised of the T-hot random and systematic streaming identified on Table 5, page 18, Hot Leg Enthalpy section.
Page 6 of 7
Document Control Desk Attachment LR-N97040 LCR S94-41 RA!
- 17)
Page 15, second paragraph.
Under what conditions is a systematic temperature error allowance included as a cross calibration systematic error.
Response
When the RdF RTDs were investigated for calibration anomalies, there was a concern that some of the RTDs exhibited a systematic error, acquired through the vendor calibration process.
This type of error is limited in magnitude and one directional, i.e., it is a bias.
The Salem Units utilize Weed RTDs and have not exhibited any systematic error during calibration.
Page 7 of 7