ML051010036

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NRC Audit Question 6
ML051010036
Person / Time
Site: Hatch  Southern Nuclear icon.png
Issue date: 07/08/2003
From: Sullivan K
- No Known Affiliation
To:
Office of Nuclear Reactor Regulation
References
FOIA/PA-2004-0277
Download: ML051010036 (9)
v • d • e


Text

A -Q o cr Control Number:

Audit Question Processing Form (For Staff use only)

AQ-6 Question Initiated By: Sullivan, Ken Date: 7/8/2003 Regulatory Basis for Question:

Question:

Is the plant designed to withstand all 11 SRV's opening from full power?

Response

The plant is designed to withstand all 11 SRVs opening from full power.

Specifically, the primary containment loads that result from operation of all 11 SRVs are detailed in the HNP-I and 1HNP-2 Plant Unique Analysis Reports (PUARs) as referenced in HNP-1 FSAR KA and HNP-2 FSAR 3.8B. As part of extended power uprate, additional containment loads analysis was performed that demonstrated that adequate design margins continued to exist for operation at current rated power, 2763 MWt. It is noted that this SRV containment dynamic loads analysis reflects the single SRV opening setpoint of 1150 psig for all HNP-I and HNP-2 SRVs.

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(For Staff use only)

Response b Reviewedk- Staff init: \..

NRC Status: In Progress__ _ _ __

-HNPR1-FSAR-KA~

SUPPLEMENT KA PLANT UNIQUE ANALYSIS OF THE MARK I CONTAINMENT SYSTEM KA.1 INTRODUCTION The Hatch Nuclear Plant-Unit 1 (HNP-1) containment system is one of the first-generation General Electric (GE) boiling water reactor (BWR) nuclear steam supply systems housed in a containment structure designated as the Mark I Containment System. The original design of the Mark I Containment System considered postulated accident loads previously associated with containment design, which included pressure and temperature loads associated with a loss-of-coolant accident (LOCA), seismic loads, dead loads, jet-impingement loads, hydrostatic loads due to water in the suppression chamber, overload pressure test loads, and construction loads. However, since the establishment of the original design criteria, additional loading conditions have been identified that arise in the functioning of the pressure-suppression concept used in the Mark I Containment System. These additional loads result from the dynamic effects of drywell air and steam being rapidly forced into the suppression pool (torus) during a postulated LOCA and from suppression pool response to safety relieyalye (SRV operation generally associated wit pint transient operati conditions. Because these were not considereditheoinesign oMEEark I Containment hydrodynamcloads System, the Nuclear Regulatory Commission (NRC) determined that a detailed reevaluation of the Mark I Containment System was required.

A two-phase program was identified to the NRC in May 1975. The first-phase effort, called the Short-Term Program (STP), provided a rapid assessment of the adequacy of the containment to maintain its integrity under the most probable course of the postulated LOCA. The first phase demonstrated the acceptability of continued operation during the performance of the second phase, called the Long-Term Program (LTP). In the LTP, detailed testing and analytical work was performed to define the specific design loads against which the containment was assessed to conform to established acceptance criteria.

The STP was completed in 1977, following the docketed submittal by Georgia Power Company (GPC) to the NRC of the HNP-1 plant unique analysis report (PUAR).(') Reevaluation of the Mark I Containment System (STP and LTP) was completed in December 1983, following the docketed submittal by GPC to the NRC of the HNP-1 PUAR.'-

KA.2 PLANT UNIQUE ANALYSIS REPORT The Mark I Containment System reevaluation results are detailed in the PUAR submitted to the NRC in December 1983 and revised in December 1989 (reference 3). Many of the analyses presented in reference 3 assumed a 95 0F initial suppression pool temperature. Subsequent analysis (reference 7) documents the acceptability of the reference 3 analyses at higher initial pool temperature (<1OO0 F). The PUAR demonstrates that the configuration of the plant, including structural modifications and load mitigation devices, meets the NRC requirements for KA-1 REV 20 7/02

HNP-1 -FSAR-KA the Mark I LTP as documented in the Mark I Containment Long-Term Program Safety Evaluation Report, NUREG-0661.12 ' -

The reference 7 analysis also documents the containment loads analysis performed to justify operation at 2763 MWt. Additional structural modifications were not required and NRC requirements for the Mark I LTP continue to be satisfied.

In summary, the PUAR report provides the following:

  • A review of the event sequences involving the Mark I Containment System related phenomena for the postulated LOCA and SRV actuation conditions.
  • A description of the major structural components of the HNP-1 containment system that were evaluated. The description includes both before and after status structural modifications.
  • A review of the design criteria used, which includes both the design specification covering the fabrication and erection of the modifications and the structural acceptance criteria applying to the design analysis.
  • A discussion of the system changes/additions made to the containment system to mitigate loads.
  • A description of the loads and load combinations as applied in the HNP-1 analysis.
  • A review of the computer programs used in the analysis.
  • A summary of the analytical methods and models employed In evaluating each of the structural components.
  • A summary of the analytical results for each structural component and a comparison with allowables, based on the structural acceptance criteria which demonstrate that the upgraded design-safety margins have been achieved.

KA.3 DESCRIPTION OF LTP MODIFICATIONS The components significantly affected by the postulated LOCA and SRV actuation events are the suppression chamber, vent system, torus internal structures, SRV piping and supports, and the torus-attached piping and supports. Detailed analysis of the components determined that structural modifications and system changes were required to establish the NRC design-safety margins specified for the Mark I LTP. Table KA-1 presents a summary of the LTP modifications to the HNP-1 containment system. The modifications are in addition to the STP changes summarized in Appendix A of the PUAR.( 3 )

KA-2 REV 20 7/02

HNP-1-FSAR-KA KA.4 EXPANDED OPERATING DOMAIN OPERATION A containment loads analysis was performed to demonstrate that ample margins for containment integrity still remain for plant operation in the expanded operating domain (EOD) at the maximum core inlet subcooling condition which is 100% power and 87% flow with reduced feedwater temperature. This analysis,( 5 ) which evaluated the containment pressure and temperature response and the containment hydrodynamic loads for a postulated design basis LOCA, was based on the methodology developed for the Mark I Long-Term Containment Program which is documented in the Mark I Containment Program Load Definition Report.(6)

The results of this analysis showed that the peak containment pressure in the EOD with reduced feedwater temperature is 51.6 psig, which is higher than the value reported in NEDO-24570 of 47.9 psig, but below the design value of 56 psig and within the design margins shown in the PUAR. The containment hydrodynamic loads with EOD conditions are also within the design margins shown in the PUAR.

KA.5 OPERATION DURING PERIOD OF EXTENDED OPERATION An analysis of the cumulative fatigue usage factor (CFUF) for the torus shell was performed to account for the period of extended operation. (See HNP-2-FSAR subsection 18.1.1 for a definition of the term "period of extended operation.") This analysis demonstrated the need to track actual thermal and dynamic loading events -to ensure the torus shell maintains an actual CFUF s 1.0 through the period of extended operation. The most limiting event for the torus is the steam blowdown resulting from the lifting of one or more main steam safety relief valves.

The component cyclic or transient limit program (HNP-2-FSAR subsection 18.2.12) performs tracking of operational events. The CFUF analysis is a time-limited aging analysis and is described in HNP-2-FSAR section 18.5.

KA-3 REV 20 7/02

HNP-1 -FSAR-KA REFERENCES

1. "Torus Support System and Attached Piping Evaluation for E.l. Hatch Nuclear Plant Unit 1, Mark I Containment," NRC Docket No. 50-321, Bechtel Power Corporation, August 1976.
2. "Safety Evaluation Report Mark I Containment Long-Term Program," NUREG-0661, U. S. Nuclear Regulatory Commission, July 1980.
3. "Plant Unique Analysis Report for E.l. Hatch Nuclear Plant-Unit 1, Mark I Containment Long-Term Program," Revision 3, NRC Docket No. 50-321, Bechtel Power Corporation, December 1989.
4. (Deleted)
5. "Limiting Reload Licensing Events for E.l. Hatch Nuclear Plant Unit 1 and Unit 2,"

EAS 65-1088, General Electric Company, October 1988.

6. "Mark I Containment Program Load Definition Report," Revision 2, NEDO-2188, General Electric Company, November 1981.
7. "Extended Power Uprate Safety Analysis Report for Edwin I. Hatch Nuclear Plant Units I and 2," NEDC-32749P, General Electric Company, July 1997.

KA-4 REV 20 7/02

HNP-2-FSAR-3 SUPPLEMENT 3.8B PLANT-UNIQUE ANALYSIS OF MARK I CONTAINMENT SYSTEM 3.8B.1 INTRODUCTION The Hatch Nuclear Plant Unit 2 (HNP-2) containment system is one of the first-generation General Electric (GE) boiling water reactor (BWR) nuclear steam supply systems (NSSSs) housed in a containment structure designated as the Mark I containment system. The original design of the Mark I containment system considered postulated accident loads previously associated with containment design, which included pressure and temperature loads associated with a loss-of-coolant accident (LOCA), seismic loads, dead loads, jet-impingement loads, hydrostatic loads due to water in the suppression chamber, overload pressure test loads, and construction loads. However, since the establishment of the original design criteria, additional, loading conditions have been identified that arise in the functioning of the pressure-suppression concept utilized in the Mark I containment system. These additional loads result from the dynamic effects of drywell air and steam being rapidly forced into the suppression pool (torus) during a postulated LOCA and from suppression pool response to safety relief valve (SRV)-

operation generally associated with plant transient operating conditions. Because these hydrodynamic loads were not considered in the original design of the Mark I containment system, the Nuclear Regulatory Commission (NRC) determined that a detailed reevaluation of the Mark I containment system was required.

A two-phase program was identified in the NRC in May 1975. The first-phase effort, called the Short-Term Program (STP), provided a rapid assessment of the adequacy of the containment to maintain its integrity under the most probable course of the postulated LOCA. Thus, the first phase demonstrated the acceptability of continued operation during the performance of the second phase, called the Long-Term Program (LTP). In the LTP, detailed testing and analytical work was performed to define the specific design loads against which the containment was assessed to establish conformance to established acceptance criteria.

The STP was completed in July 1977 following the docketed submittal by Georgia Power Company (GPC) to the NRC of the HNP-2 plant unique analysis report." Reevaluation of the Mark I containment system (LTP) was completed in September 1983, following the docketed submittal by Georgia Power Company to the JRC6ofttiei NP-2 PlantfUinique AnlysisReport (PUAR).(2 )

3.8B.2 PLANT UNIQUE ANALYSIS REPORT The Mark I containment system reevaluation results are detailed in the PUAR submitted to the NRC in September 1983 and revised in December 1989. (See Reference 2). Many of the analyses presented in Reference 2 assumed'a 950 F initial suppression pool temperature.

Reference 3 documents the acceptability of the Reference 2 analyses at higher initial pool temperature (<110°1). The PUAR demonstrates that the configuration of the plant, including structural modifications and load mitigation devices, meets the NRC requirements for the 3.8B-1 REV 20 7/02

HNP-2-FSAR-3 Mark I LTP as documented in the Mark I containment Long-Term Program Safety Evaluation Report, NUREG-0661.14 1 In summary the report provides the following:

  • A review of the event sequences involving the Mark I containment system related phenomena for the postulated LOCA and safety relief valve (SRV) actuation conditions.
  • A description of the major structural components of the HNP-2 containment system that were evaluated. The description includes both before and after status structural modifications.
  • A review of the design criteria used, which includes both the design specification covering the fabrication and erection of the modifications and the structural acceptance criteria applying to the design analysis.
  • A discussion of the system changes/additions made to the containment system to mitigate loads.
  • A description of the loads and load combinations as applied in the HNP-2 analysis.
  • A review of the computer programs used in the analysis.
  • A summary of the analytical methods and models employed in evaluating each of the structural components.-
  • A summary of the analytical results for each structural component and a comparison with allowables, based on the structural acceptance criteria that demonstrate that the upgraded design-safety margins have been achieved.

3.8B.3 DESCRIPTION OF LTP MODIFICATIONS The components significantly affected by the .postulated LOCA and .SRV actuation events-are_

the suppression chamber, vent system, torus internal structures, SRV piping and supports, and the torus-attached piping and supports. Detailed analysis of the components determined that structural modifications and system changes were required to establish the NRC design safety margins specified for the Mark I LTP. Table 3.8B-1 presents a summary of the LTP modifications to the HNP-2 containment system. The modifications are in addition to the STP changes summarized in Appendix A of the PUAR.t2 )

3.8B.4 EXPANDED OPERATING DOMAIN OPERATION A containment loads analysis was performed to demonstrate that ample margins for containment integrity still remain for plant operation in the expanded operating domain (EOD) at 3.86-2 REV 20 7/02

HNP-2-FSAR-3 the maximum core inlet subcooling condition which is 100% of original rated power (2436 MWt) and 87% flow with reduced feedwater temperature. This analysis ) which evaluated the containment pressure and temperature response and the containment hydrodynamic loads for a postulated design basis LOCA, was based on the methodology developed for the Mark I Long-Term Containment Program and documented in the Mark I Containment Program Load Definition Report.A6) The results of this analysis showed that the peak containment pressure in the EOD with reduced feedwater temperature is 46.7 psig which is higher than the value reported in NEDO-24569(7) of 43.0 psig, but below the design value of 56 psig and within the design margins shown in the PUAR. The containment hydrodynamic loads with EOD conditions are also within the design margins shown in the PUAR.

3.8B.5 EXTENDED POWER UPRATE OPERATION A containment loads analysis was performed to assure adequate margins exist for operation at 2763 MWt. The results, summarized in reference 9, are acceptable. The containment system performance analysis included short- and long-term pressure and temperature responses, LOCA containment dynamic loads, and SRV containment dynamic loads.. The analyses included the EOD for a core power of 2763 MWt and final feedwater temperature operation.

The peak containment pressure is 46.9 psig.

3.8B.6 OPERATION DURING PERIOD OF EXTENDED OPERATION An analysis of the cumulative fatigue usage factor (CFUF) for the torus shell was performed to account for the period of extended operation. (See subsection 18.1.1 for a definition of the term "period of extended operation.") This analysis demonstrated the need to track actual thermal and dynamic loading events to ensure the torus shell maintains an actual CFUF < 1.0 through the period of extended operation. The most limiting event for the torus is the steam blowdown resulting from the lifting of one or more main steam safety relief valves. The component cyclic or transient limit program (subsection 18.2.12) performs tracking of operational events. The CFUF analysis is a time-limited aging analysis and is described in section 18.5 3.813-3 REV 20 7/02

HNP-2-FSAR-3 REFERENCES

1. "Structural Evaluation of Pressure Suppression System for E. l. Hatch Nuclear Plant-Unit 2, Mark I Containment," Bechtel Power Corporation, Docket No. 50-366, July 1977.
2. "Plant Unique Analysis Report for E. l. Hatch Nuclear Plant-Unit 2, Mark I Containment Long-Term Program," Bechtel Power Corporation, Revision 2, Docket No. 50-366, December 1989.
3. "Elimination of the Suppression Pool Temperature Limit for Plant Hatch Units 1 and 2,"

EAS-19-0388, General Electric Company, March 1988.

4. "Safety Evaluation Report Mark I Containment Long-Term Program," NUREG-0661, U. S. Nuclear Regulatory Commission, July 1980.
5. "Limiting Reload Licensing Events for E. l. Hatch Nuclear Plant Unit 1 and Unit 2,"

EAS-65-1088, General Electric Company, October 1988.

6. "Mark I Containment Program Load Definition Report," Revision 2, NEDO-21888, General Electric Company, November 1981.
7. "Mark I Containment Program Plant Unique Load Definition: Unit 2," NEDO-24569, General Electric Company, September 1981.
8. "Power-Uprate Safety Analysis Report for Edwin I. Hatch Plant Units 1 and 2,"

NEDC-32405P, General Electric Company, December 1994.

9. "Extended Power Uprate Safety Analysis Report for Edwin I. Hatch Plant Units 1 and 2,"

NEDC-32749P, General Electric Company, July 1997.

3.8B-4 REV 20 7/02

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