NL-04-0392, Response to RAI on the Proposed Technical Specifications Revision to Primary Containment Leakage Rate Testing Program
| ML040930186 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 03/30/2004 |
| From: | Sumner H Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-04-0392 | |
| Download: ML040930186 (20) | |
Text
H. L Sumner, Jr.
Southern Nuclear Vice President Operating Company, Inc.
Hatch Project Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.7279 SOUTHERNAh March 30, 2004 COMPANY Energy to Serve You r Wo rld9' Docket Nos.:
50-321 NL-04-0392 50-366 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin 1. Hatch Nuclear Plant Request for Additional Information on the Proposed Technical Specifications Revision to Primary Containment Leakage Rate Testing Program Ladies and Gentlemen:
By letter dated December 1, 2003 Southern Nuclear Operating Company (SNC) submitted to the NRC a proposed change to the Unit 1 and Unit 2 Technical Specifications (TS) for the Edwin I. Hatch Nuclear Plant. The amendment request proposes a change in the post-accident peak primary containment pressure (Pa) listed in TS section 5.5.12, "Primary Containment Leakage Rate Testing Program." This proposed change supports other efforts by SNC to increase the reactor nominal operating pressure for the Hatch Units. The containment evaluation performed for the pressure increase effort resulted in slightly higher post-accident peak calculated containment pressure values. Since the peak calculated containment pressures are explicitly listed in the Administrative section of the TS, a TS change is required.
Through a teleconference conversation with the NRC/NRR Hatch Project Manager, a request was made for SNC to provide a correspondence describing the additional effort associated with the pressure increase project. Also, an electronic communication (E-mail) was received by SNC requesting responses to three questions from a staff reviewer pertaining to the license amendment request (LAR). By letter dated March 10, 2004 SNC provided a response to question I of the E-mail, as well as a more detailed description of the pressure increase project scope as requested. The answers to questions 2 and 3 of the E-mail request are provided as a Request for Additional Information enclosure to this letter.
(Affirmation and signature are on the following page).
U. S. Nuclear Regulatory Commission NL-04-0392 Page 2 Mr. H. L. Sumner, Jr. states he is a Vice President of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and to the best of his knowledge and belief, the facts set forth in this letter are true.
This letter contains no NRC commitments. If you have any questions, please advise.
Respectfully submitted, so >
ERN NUCLEAR OPERATING COMPANY Sumner, Jr.
Sworn to and subscribed before me this 0 day of _6ACh
, 2004.
I.(5IIJI4IItIC..IJ/--.
A, :
,Notary Public
-M ycpmitnission expires: '1z 29-D
- J-IL Svhc/daj
Enclosure:
Request for Additional Information cc:
Southern Nuclear Operating Companv Mr. J. B. Beasley, Jr., Executive Vice President Mr. G. R. Frederick, General Manager - Plant Hatch RType: CHAO2.004 U. S. Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Mr. C. Gratton, NRR Project Manager - Hatch Mr. D. S. Simpkins, Senior Resident Inspector - Hatch State of Georgia Mr. L. C. Barrett, Commissioner - Department of Natural Resources
Enclosure Request for Additional Information on the Proposed Technical Specifications Revision to Primary Containment Leakage Rate Testing Program NRC Ouestion (2) Provide a comparison listing of the inputs for the new reactor steam dome nominal pressure to each input listed in the FSARfor the current licensing basis containment analyses (FSAR Table 6.2-6, "DBA-LOCA Initial Conditions, Assumptions, and Calculated Pressure Results'). Provide a comparison of the mass and energy releases and the pressure and temperature responses (current to revised pressure case), in graphicalform, and include the suppression pool temperature response. (a) Will the FSAR be revised to reflect these new analyses? (b) Ifnot, how wvill these analyses be maintained as "the current licensing basis containment analyses?"
SNC Response Comparison Listing of Inputs FSAR Table 6.2-6 provides the inputs and results of various containment analyses performed for Plant Hatch Units 1 and 2. These analyses include the Short-Term Containment Analysis (Peak Pressure), Short-Term Containment Analysis for Dynamic Loads Evaluation, Long-Term Containment Analysis (peak temperature) and Long Term Containment Analysis for ECCS NPSH considerations. All the inputs and results of FSAR Table 6.2-6 are not applicable to the Short-Term Containment Evaluations performed for the pressure increase that result in the slight increase in peak calculated containment pressures, Pa provided in the respective Technical Specifications. Table I of this enclosure provides a comparison of the relevant existing licensing basis and pressure increase Short-Term Containment Analyses key inputs and results.
Comparison of mass/energy/temperature/pressure The existing and revised time histories for various parameters (Pressure, Temperature, Mass and Enthalpy) are provided as a comparison in the following enclosed figures:
- Figures 1 and 8
- Figures 2 and 9
- Figures 3 and 10
- Figures 4 and 11
- Figures 5 and 12
- Figures 6 and 13
- Figures 7 and 14 Existing Unit 1 and Unit 2 Short-Term Containment Response Drywell/Wetwell Pressure vs. Time Revised Unit I and Unit 2 Short-Term Containment Response Drywell/Wetwell Pressure vs. Time Existing Unit I and Unit 2 Short-Tern Containment Response Drywell/Wetwell Temperature vs. Time Revised Unit I and Unit 2 Short-Term Containment Response-Drywell/Wetwell Temperature vs. Time Existing Unit 1 and Unit 2 Short-Term Containment Response Break Flow vs. Time Revised Unit 1 and Unit 2 Short-Term Containment Response Break Flow vs. Time Unit 1 and Unit 2 Short-Tern Containment Response - Energy Addition vs. Time Figures 7 and 14 represent the energy releases for the pressure increase analyses. The energy time history files were not provided in the current licensing basis engineering reports and are therefore not provided for comparison in this enclosure.
Page E-1
Enclosure Request for Additional Information on the Proposed Technical Specifications Revision to Primary Containment Leakage Rate Testing Program FSAR Revision As stated in SNC's letter NL-04-037 1, following completion of site implementation, the affected FSAR sections will be revised under the provisions of 10 CFR 50.71(e) to reflect the changes due to the 10 psi nominal operating pressure increase thereby maintaining current, the licensing basis for the containment analyses.
NRC Ouestion (3) It would appear that the increase in the reactor steam dome pressure would result in a higher suppression pool temperature at the end of the short-term, blowdown, period.
The submittal indicates that the suppression poolpeakshort-term temperatures are 1 F to 2 F higher as a result of the reactor steam dome pressure change (compare Table 6.2-6 of the FSAR to Table 1 of LAR). During the long-term, the temperature (andpressure) response might also be affected In reviewing the FSAR, it would appear that these reported temperatures are actually the long-term values (see for example FSAR Figure 6.2-23 or 6.2-11), the short-term suppression pool temperature response is not shown.
(a) Are the reported temperatures the long-term peakvalues? (b) On page 6.2-26 (Containment Long-Tern Response) of the FSAR, it is stated that after RPV depressurization is complete the suppression pool is the only heat sink in the containment.
Provide a discussion of the long-term analysis during the depressurization period which supports your statement that the long-term suppression pool temperature is not affected by the revised reactor steam dome pressure.
SNC Response The peak pool temperature provided in the December 1, 2003 submittal is the bounding long-term suppression pool temperature resulting from the Extended Power Uprate Long-Tern Containment (Peak Temperature) analysis. This analysis used the most conservative inputs (smaller pool volume, lower final feedwater temperature, most limiting core flow...etc.) from either Unit I or Unit 2 which produced a peak pool temperature of 208 'F. FSAR Table 6.2-6 lists the peak suppression pool temperatures resulting from the individual units' Long-Tern Containment Analysis for ECCS NPSH consideration. The impact of the pressure increase on the peak pool temperature was determined by evaluating the single bounding analysis. This evaluation concluded that conservatively assessing the effect of the increased sensible energy, the effects on the long-term containment DBA-LOCA response of the reactor dome pressure increase are insignificant (less than 0.3 IF increase in peak suppression pool temperature), and therefore the results of the single bound Long-Term (Peak Temperature) analysis remain valid.
On page 6.2-26 (Containment Long-Term Response) of the FSAR states: "During the long-term containment response (after RPV depressurization is complete), the suppression pool is assumed to be the only heat sink in the containment system. The effects of decay energy and stored energy on the suppression pool temperature are considered". The long term analysis is directed primarily at the suppression pool temperature response, considering the decay heat addition to the suppression pool. The long-term heat-up of the suppression pool following a LOCA is governed by the capability of the residual heat system to remove decay heat which is transferred from the Page E-2
Enclosure Request for Additional Information on the Proposed Technical Specifications Revision to Primary Containment Leakage Rate Testing Program reactor pressure vessel to the suppression pool. The existing Hatch long-term analysis takes no credit for passive structural heat sinks in the dry vell, and suppression chamber (airspace and pool). Therefore, after RPV depressurization is complete, the suppression pool is the only heat sink in the containment. Since the decay heat is dependent on the initial power level, which is unchanged with the pressure increase, the long-term containment response will not be significantly impacted by the 10 psi nominal operating pressure increase. The main effect of the pressure increase is on the increase in initial stored sensible energy in the fluid and the solid components within the reactor vessel.
The additional sensible energy is an insignificant contribution to the overall peak suppression pool temperature response compared to other more conservative input assumptions for the long-term containment analysis.
Reference:
"Hatch Pressure Increase Project - Time Histories for Short Term LOCA Containment Response using M3CPT blowdown model" Dated March 15, 2004.
"Hatch Pressure Increase Project - Time Histories for Short Term LOCA Containment Mass and Energy Release using M3CPT blowdown model" Dated March 19, 2004.
- 3. NEDC-32749P Extended Power Uprate Safety Analysis Report for Edwin 1.
Hatch Nuclear Plants Units 1 and 2, July 1997.
Page E-3
Enclosure Request for Additional Information on the Proposed Technical Specifications Revision to Primary Containment Leakage Rate Testing Program Table - 1 Short-Term Containment Analysis Key Inputs HNP - 1 HNP-1 HNP-2 HNP-2 Initial Conditions Existing 10 PSI Existing 10 PSI INCREASE INCREASE Containment Containment Evaluation Evaluation Reactor power level 2818.3 MWt 2818.3 MWt 2818.3 MWt 2818.3 MWt Initial Core Flow 78.5 Mlb/hr 78.5 Mlb/hr 77.0 Mlb/hr 77.0 Mlb/hr Steam Dome Temperature 551.0 °F 552.0°F 551.0°F 552.0°F Feedwater Temperature at 399.5 OF 399.5 OF 427.3 OF 427.3 OF Vessel Inlet Break Enthalpy 525.6 BTU/lb 527.0 BTUAb 528.8 BTUAb 530.2 BTU/lb RPV Dome Pressure 1053 psia 1063 psia 1053 psia 1063 psia Drywell Free Airspace 146,010 f 3 146,010 f 3 146,266 ft3 146,266 ft3 Volume (Including Vent System)
Drywell Pressure 1.75 psig 1.75 psig 1.75 psig 1.75 psig Drywell Temperature (drywell 150 OF 150 °F 150 °F 150 OF air)
Drywell Relative Humidity 20%
20%
20%
20%
Wetwell Free Airspace 112,900 f 3 112,900 ft3 109,800 ft3 109,800 ft3 Volume Wetwell Pool Volume 88,192 ft 88,190 fl3 89,670 ft3 89,670 ft3 Wetwell Pressure 1.75 psig 1.75 psig 1.75 psig 1.75 psig Wetwell Pool Temperature 100 OF 100 OF 100 OF 100 OF Wetwell Airspace 100 OF 100 OF 100 °F 100 °F Temperature Wetwell Airspace Relative 100%
100%
100%
100%
Humidity Operating Drywell to Wetwell 0 psid 0 psid 0 psid 0 psid Differential Pressure Wetwell Pool Surface Area in 9,500 ft2 9,500 ft2 9,500 ft2 9,500 ft2 contact with Wetwell Airspace I
I I
I Page E-4
Enclosure NL-04-0392 Figure 1 Existing HIt EX PW UP CASE IA (102EPU/100 )
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Enclosure NL-04-0392 Figure 12 Existing HT2 EX PW UP CASE 1A (102EPU/100 DBA BREAK FLOW
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Enclosure NL-04-0392 6.E+04 M 4.E+04 I-J 0-J U.
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Enclosure.
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