Text

Technical Specifications Change 06-02 - License Amendment Request Associated with NRC Generic Letter 2004-02
	ML061500426
Person / Time
Site:	Sequoyah
Issue date:	05/25/2006
From:	Pace P; Tennessee Valley Authority
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	GL-04-002, TVA-SQN-TS-06-02
	Download: ML061500426 (27)
	v • d • e

Tennessee Valley Authority, Post Office Box 2000, Soddy-Daisy, Tennessee 37384-2000 May 25, 2006 10 CFR 50.59(c)(2)

TVA-SQN-TS-06-02 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN:

Document Control Desk Washington, D. C. 20555-0001 Gentlemen:

In the Matter of

)

Docket Nos.

50-327 Tennessee Valley Authority

)

50-328 SEQUOYAH NUCLEAR PLANT (SQN)

UNITS 1 AND 2 -

TECHNICAL SPECIFICATIONS (TS)

CHANGE 06 LICENSE AMENDMENT REQUEST (LAR) ASSOCIATED WITH NRC GENERIC LETTER (GL) 2004-02

Reference:

TVA letter to NRC, September 1, 2005, "Sequoyah Nuclear Plant Units 1 and 2 - Nuclear Regulatory Commission (NRC)

Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors -

Second Response" Pursuant to 10 CFR 50.90, TVA is submitting a request for an Operating License change to Licenses DPR-77 and DPR-79 for SQN Units 1 and 2.

The proposed change will modify the SQN design and licensing basis for the containment sump debris transport analysis as described in the SQN Updated Final Safety Analysis Report (UFSAR).

The current transport analysis for SQN is a two-dimensional physical transport model.

As informed in the referenced letter to NRC GL 2004-02, "Potential Impact of Debris Blockage on

AIy, Printed om recycled paper

U.S. Nuclear Regulatory Commission Page 2 May 25, 2006 Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors," TVA is requesting to update the analysis to a three-dimensional transport model.

TVA has determined that implementation of the proposed TS changes be congruent with the modification of the containment sumps during the respective refueling outage.

Therefore, NRC approval of the proposed LAR is requested to support TVA's schedule for issuing design packages for plant modifications to the containment sump.

SQN Unit 2 modification occurs first and is scheduled during the Unit 2 Cycle 14 refueling outage (November 2006).

SQN Unit 1 modification is scheduled during the Unit 1 Cycle 15 refueling outage (September 2007).

The LAR revisions were reviewed under the requirements of 10 CFR 50.59, "Changes, Tests and Experiments."

Based on this review, it was concluded that a license amendment is required in accordance with 10 CFR 50.59(c) (2).

TVA has determined that there are no significant hazards considerations associated with the proposed change and that the license amendment qualifies for categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c) (9).

Additionally, in accordance with 10 CFR 50.91(b) (1),

TVA is sending a copy of this letter and enclosures to the Tennessee State Department of Public Health.

There are no regulatory commitments associated with this submittal.

If you have any questions concerning this change, please contact James D. Smith at (423) 843-6672.

U.S. Nuclear Regulatory Commission Page 3 May 25, 2006 I declare under penalty of perjury that the foregoing is true and correct.

Executed on this 25th day of May, 2006.

Sincerely, P.

L.

Pace Manager, Site Licensing and Industry Affairs

Enclosures:

1. TVA Evaluation of the Proposed Changes

2. Changes to Updated Final Safety Analysis Report -

(mark-up)

3. Mode of Transport Logic Tree cc: See page 4

U.S. Nuclear Regulatory Commission Page 4 May 25, 2006 cc (Enclosures):

Framatome ANP, Inc.

P.

0. Box 10935 Lynchburg, Virginia 24506-0935 ATTN:

Mr. Frank Masseth Mr.

Edgar D. Hux 94 Ridgetree Lane Marietta, Georgia 30068 Mr.

Lawrence E.

Nanney, Director Division of Radiological Health Third Floor L&C Annex 401 Church.Street Nashville, Tennessee 37243-1532 Mr.

Douglas V. Pickett, Senior Project Manager U.S. Nuclear Regulatory Commission Mail Stop 08G-9a One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2739 Mr. William T. Russell 400 Plantation Lane Stevensville, Maryland 21666

ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT (SQN)

UNITS 1 AND 2 TVA Evaluation of the Proposed Changes

1.0 DESCRIPTION

This letter provides a license amendment request (LAR) that includes a proposed revision to the SQN Updated Final Safety Analysis Report (UFSAR).

The proposed UFSAR revision is provided in response to resolution of Generic Safety Issue No.

191 (GSI-191),

"Assessment of Debris Accumulation on PWR Sump Performance" and is in accordance with TVA letter to NRC dated September 1, 2005, "Sequoyah Nuclear Plant (SQN) Units 1 and 2 - Nuclear Regulatory Commission (NRC) Generic Letter 2004-02,

'Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized Water Reactors'"-

Second Response."

A change in methodology is proposed for SQN's containment sump debris transport analysis.

The change in methodology affects SQN's current design and licensing basis as described in Section 6.2.1.6 of the SQN UFSAR.

The proposed UFSAR revision was reviewed under the requirements of 10 CFR 50.59, "Changes, Tests and Experiments."

Based on this review, it was concluded that a LAR is required in accordance with 10 CFR 50.59©(2).

2.0 PROPOSED CHANGE

A description of the proposed UFSAR changes is provided below:

Proposed Change to UFSAR, Section 6.2.1.6, "Protective Coatings":

Information no longer applicable for SQN is removed from the UFSAR (Section 6.2.1.6, entitled "Protective Coatings") and is replaced with the following:

"As a result of the above information, TVA reevaluated the licensing basis for the containment sump screen blockage.

While designed and constructed before the issuance of NRC Regulatory Guide 1.82, Revision 0, the original licensing E1-I

basis for the containment recirculation sump included operability with an assumed 50 percent intake flow area blockage consistent with RG 1.82 recommendations.

Scale model testing confirmed that the design of the original sump intake structure (i.e.,

a 6 inch trash curb around the base of the sump intake structure and 0.25 inch mesh intake screens sloping upward and outward from the sump opening) was sufficient to meet the 50 percent blockage criteria.

To address the potential increased failed coating debris load, an evaluation was performed using a two-dimensional physical transport model (Reference 70) to confirm the ability of the containment sump to support containment spray and emergency core cooling system pump operation.

The evaluation methodology focused on a near-sump region of influence resulting from post-accident flow fields where debris transport to the sump intake was possible.

The evaluation quantified the amount of accident generated debris which could potentially be transported to the sump intake structure, established the head loss from the resulting intake screen debris blockage and confirmed that the minimum containment spray and emergency core cooling system pump suction head requirements would be met for the expected blockage.

A subsequent evaluation established a maximum limit on failed coatings which could be transported to the containment recirculation sump without degrading the capabilities of the required accident mitigation systems (Reference 71).

This evaluation established the revised licensing basis for containment sump intake blockage.

It was summarized and submitted for NRC review in Reference 85 and was accepted in Section 3.7 of Reference 86.

To address the additional concerns contained in NRC Generic Safety Issue No. 191 (GSI-191), "Assessment of Debris Accumulation on PWR Sump Performance," the containment sump was subsequently reanalyzed to address the susceptibility of the emergency core cooling and containment spray recirculation functions to the adverse effects of post-accident debris blockage and operation with debris laden fluids.

As summarized in Reference 87, the comprehensive reanalysis used the evaluation methodology described in Reference 88.

The revised analysis methodology included development of a three-dimensional computational fluid dynamics model to E1-2

establish debris transport characteristics (i.e., flow directions, velocities and turbulence) in the entire sump pool during post-accident sump recirculation operation.

The results of the reanalysis were used to size the flow area of the advanced design containment sump strainers which replaced the original sump intake structure.

Blockage testing of the advanced containment sump strainer design with a conservative debris load (which included the assumed failure of all qualified and unqualified coatings in containment) confirmed that the containment sump will support operation of the emergency core cooling system and the containment spray system under all anticipated debris loading conditions. This includes the failure of all coatings installed inside containment."

Proposed Change to UFSAR Section 6.2, "Reference Section":

The following references are added to the UFSAR:

Reference 85.

Reference 86.

Reference 87.

Reference 88.

R. Gridley letter to NRC dated September 16, 1987, "Sequoyah Nuclear Plant

- Containment Coatings."

NUREG-1232, Volume 2, "Safety Evaluation Report on Tennessee Valley Authority:

Sequoyah Nuclear Performance Plan," May 1988.

Randy Douet letter to NRC dated September 1, 2005, "Sequoyah Nuclear Plant NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors -

Second Response."

Nuclear Energy Institute Guidance Report No. NEI-04-07, "Pressurizer Water Reactor Sump Performance Evaluation."

3.0 BACKGROUND

The SQN UFSAR describes the original licensing basis for SQN's containment sump screen blockage evaluations.

A two-dimensional physical transport model currently confirms the ability of the containment sump to support containment spray E1-3

and ECCS pump operation during containment sump recirculation operation.

The evaluation establishes the current licensing basis for SQN as approved by NUREG-1232, Volume 2, "Safety Evaluation Report on Tennessee Valley Authority:

Sequoyah Nuclear Performance Plan," dated May 1988.

To address additional concerns in response to GSI-191 and NRC Generic Letter (GL) 2004-02, the SQN containment sump was reanalyzed.

The revised analysis uses a new transport analysis methodology that includes a three-dimensional fluid dynamics model for debris transport characteristics.

The three-dimensional model sizes the flow area for an advanced sump strainer design, which will replace SQN's original sump intake structure.

TVA's change to SQN's debris transport analysis from a two-dimensional model to a three-dimensional model constitutes a change in methodology affecting SQN's licensing basis.

Accordingly, TVA is submitting a proposed change to the SQN UFSAR for NRC review and approval.

These changes support plant modifications to the containment sump that utilize an advanced sump strainer design.

The following provides the results of the 10 CFR 50.59 evaluation that found NRC review and approval necessary:

Criteria:

10 CFR 50.59(c) (2) (viii):

"Result in a departure from a method of evaluation described in the UFSAR (as updated) used in establishing the design bases or in the safety analyses."

Evaluation Results:

SQN's evaluation methodology for the transport of debris to the containment sump during design basis accidents is currently based on a two-dimensional physical transport model.

The methodology focuses on a near-sump region of influence resulting from post-accident flow fields where debris transport to the sump intake is considered to be possible.

The methodology was reviewed by NRC and approved as part of SQN's licensing basis under NUREG-1232, Volume 2, "Safety Evaluation Report on Tennessee Valley Authority:

Sequoyah Nuclear Performance Plan," Section 3.7, May 1988.

To address issues raised by NRC GSI-191, and in response to recommendations in NRC GL 2004-02, an alternate methodology E1-4

is applied using a three-dimensional computational fluid dynamics model to establish debris transport characteristics (i.e.,

flow directions, velocities and turbulence) in the sump pool during post-accident sump recirculation operation.

Based on the above discussion, TVA's alternative methodology constitutes a change in the evaluation methodology from that currently described in the SQN UFSAR and results in the above 10 CFR 50.59 criteria being met.

Accordingly, TVA is submitting the proposed LAR for NRC approval.

4.0 TECHNICAL ANALYSIS

TVA's proposed revision to the SQN UFSAR regarding the debris transport evaluation for SQN's containment sump is based on methodology consistent with Nuclear Energy Institute Guidance Report No. NEI-04-07, "Pressurized Water Reactor Sump Performance Evaluation" as supplemented by the "Safety Evaluation by The Office of Nuclear Reactor Regulation Related to NRC Generic Letter 2004-02, Nuclear Energy Institute Guidance Report (Proposed Document Number NEI 04-07),

'Pressurized Water Reactor Sump Performance Evaluation Methodology.'"

The evaluation includes a three-dimensional computational fluid dynamics model that establishes debris transport characteristics (i.e.,

flow directions, velocities and turbulence) in the sump pool during post-accident sump recirculation operation.

The results of the three-dimensional evaluation have been used to size the strainer flow area for an advanced design containment sump strainer, which will ensure appropriate safety margins exist for SQN's ECCS and containment spray pumps.

A general description of the methodology described in the proposed UFSAR change is provided below:

Debris Transport Methodology The methodology used for the SQN analysis is based on the NEI 04-07 guidance report for refined analyses as modified by the NRC's safety evaluation report (SER),

as well as the refined methodologies suggested by the SER in Appendices III, IV, and VI.

The methodology addresses the four major debris transport modes as follows.

Blowdown transport - the vertical and horizontal transport of debris to all areas of containment by the break jet.

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" Washdown spray transport - the vertical (downward) transport of debris by the containment sprays and break flow.

" Pool fill transport - the horizontal transport of debris by break and containment spray flows from the refueling water storage tank (RWST) to areas that may be active or inactive during recirculation.

" Recirculation transport - the horizontal transport of debris from the active portions of the recirculation pool to the sump screen by the flow through the ECCS.

The specific effect of each mode of transport is addressed for each type of debris generated in accordance with the logic tree shown in Enclosure 3.

The logic tree is somewhat different than the baseline logic tree provided in the NEI guidance report to account for certain non-conservative assumptions identified by the SER including the transport of large pieces, erosion of small and large pieces, the potential for washdown debris to enter the pool after inactive areas have been filled, and the direct transport of debris to the sump screens during pool fill up.

Additionally, the generic logic tree was expanded to account for a more refined debris size distribution.

The basic methodology used for the SQN transport analysis includes:

1. Development of a three-dimensional model of the containment building interior features, which are significant enough to affect sump recirculation flow paths using computer aided drafting (CAD) software.

2.

Review of the sump transport flow paths to identify potential blockage points including screens, fences, grating, drains, etc., that could lead to water holdup.

3.

Debris types and size distributions were obtained from a debris generation analysis for each postulated break location.

4.

The fraction of debris blown into the ice condenser was determined based on the flow of steam during the blowdown.

5.

The quantity of debris washed down by ice melt and spray flow was conservatively determined.

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6.

The quantity of debris transported to inactive'areas or directly to the sump intake was calculated based on the volume of the inactive areas and sump cavities proportional to the water volume at the time these cavities were filled.

7.

Using conservative assumptions, the locations of each type/size of debris at the beginning of recirculation was determined.

8.

A computational fluid dynamics (CFD) model was developed to simulate the flow patterns that would occur during recirculation.

a.

The mesh in the CFD model was nodalized to sufficiently resolve the features of the CAD model.

b.

The boundary conditions for the CFD model were set based on the plant configuration of SQN during the recirculation phase.

c.

The ice melt and containment spray flows were included in the CFD calculation with the appropriate flow rate and kinetic energy to accurately model the effects on the containment pool.

d.

At the postulated loss-of-coolant accident break location, a mass source was added to the model to introduce the appropriate flow rate and kinetic energy associated with the break flow.

e.

A negative mass source was added at the sump location with a total flow rate equal to the sum of the spray flow and break flow.

f.

An appropriate turbulence model was selected for the CFD calculations.

g.

After running the CFD calculations, the mean kinetic energy was checked to verify that the model had been run long enough to reach steady state conditions.

h.

Transport metrics were determined based on relevant tests and calculations for each significant debris type present in the SQN containment building.

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9.

A graphical determination of the transport fraction of each type of debris was made using the velocity and telekinetic energy profiles from the CFD model output, along with the determined initial distribution of debris.

10. The recirculation transport fractions from the CFD analysis were input into the logic tree.

11. The quantity of debris that could experience erosion due to the break flow, spray flow, or ice melt drainage was determined.

12. The overall transport fraction for each type of debris was determined by combining each of the previous steps in the logic tree.

The CFD calculation for the recirculation flow was performed using the Flow-3D Version 8.2 computer code.

Flow-3D is a

commercially available general purpose computer code for modeling the dynamic behavior of liquids and gasses influenced by a wide variety of physical processes.

The program is based on the fundamental laws of mass, momentum, and energy conservation.

It has been constructed for the treatment of time-dependent, multi-dimensional problems.

The Flow-3D code used for SQN is configuration controlled under TVA's 10 CFR 50, Appendix B Quality Assurance Program.

Version 8.2 of the computer code has been validated and verified in accordance with established quality assurance requirements.

Justification for the Proposed UFSAR Changes Justification for TVA's proposed UFSAR change is based principally on the regulatory acceptance of the analytical methods described in Nuclear Energy Institute Guidance Report No. NEI-04-07 as supplemented by the NRC SER, as well as the refined methodologies suggested by the SER in Appendices III, IV, and VI.

The proposed analysis methodology represents a significant improvement and refinement of the present licensing basis sump transport methodology.

It is better suited to address the sump transport and blockage concerns raised by GSI-191 than the present SQN licensing basis methodology.

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5.0 REGULATORY SAFETY ANALYSIS This letter is a request to amend Operating Licenses DPR-77 and DPR-79 for SQN Units 1 and 2.

The proposed change will modify the SQN design and licensing basis for the containment sump debris transport analysis as described in the SQN Updated Final Safety Analysis Report (UFSAR).

The current transport analysis for SQN is a two-dimensional physical transport model.

TVA is requesting to update SQN's analysis to a three-dimensional transport model in response to NRC Generic Letter (GL) 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors."

5.1 No Significant Hazards Consideration TVA has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response

No.

The design function of the sump during accident conditions is to support emergency core cooling systems (ECCS) and containment spray system operation for recirculation.

The sump is a passive feature that does not act as an accident initiator, (i.e., failure of the sump would not initiate a design basis accident).

The proposed change to the UFSAR regarding debris transport analysis provides an overall improvement in the analysis for recirculation operation and does not change the consequences of accidents previously evaluated.

The change in methodology is neutral with regard to probability.

Consequently, the changes associated with the enclosed license amendment do not affect the frequency of occurrence for accidents previously evaluated in the UFSAR.

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Accident dose as previously evaluated in the UFSAR is unaffected by the proposed license amendment.

Based on the above discussion, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response

No.

The sump is a passive component and is not an accident initiator; i.e., failure of the sump will not initiate a design basis accident.

The sump'transport methodology is used to confirm the ability of the sump to perform all safety functions during normal and accident conditions.

Consequently, this activity does not create a possibility of a new or different type of accident than any previously evaluated.

3.

Does the proposed change involve a significant reduction in a margin of safety?

Response

No.

The changes addressed in TVA's proposed amendment are associated with methodology for debris transport to the containment sump.

The change does not affect specific safety limits, design limits, set points, or other critical parameters.

The transport methodology is used to confirm that the ECCS and containment spray systems will perform their safety functions for all accident conditions within existing equipment performance capability margins.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, TVA concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c); and, accordingly, a El-10

finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements/Criteria The applicable regulatory requirement/criteria for the proposed license amendment request (LAR) for Sequoyah Nuclear Plant (SQN) is the Nuclear Energy Institute Guidance Report No. NEI-04-07, "Pressurized Water Reactor Sump Performance Evaluation," as supplemented by the "Safety Evaluation by The Office of Nuclear Reactor Regulation Related to NRC Generic Letter 2004-02, Nuclear Energy Institute Guidance Report (Proposed Document Number NEI 04-07),

'Pressurized Water Reactor Sump Performance Evaluation Methodology,'" dated December 2004.

The current licensing basis for SQN is a two-dimensional model that was approved in NUREG-1232, Volume 2, "Safety Evaluation Report on Tennessee Valley Authority: Sequoyah Nuclear Performance Plan,"

dated May 1988.

Improved computational techniques update the debris transport analysis to a three-dimensional model.

The results of the three-dimensional model are proposed for incorporation into SQN's current licensing basis.

Accordingly, the enclosed UFSAR change is proposed to update SQN's design and licensing basis.

The General Design Criteria (GDC) which requires sump recirculation operability are as follows:

GDC 16, "Containment Design" - This GDC prescribes a reactor containment and associated systems be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.

TVA's proposed methodology change does not result in an adverse change to the post-accident sump recirculation operation.

The proposed methodology change addresses concerns of Generic Safety Issue No.

191 (GSI-191),

"Assessment of Debris Accumulation on PWR Sump Performance."

Revision of the debris transport methodology, in conjunction with plant modifications to the containment sump, will assure appropriate safety margins are maintained.

El-II

GDC 35, "Emergency Core Cooling" -

By this GDC, a

system is to be provided for abundant emergency core cooling with capacity to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented, and (2) clad metal-water reaction is limited to negligible amounts.

The recent analysis of the containment sump to address concerns of susceptibility of the ECCS and containment spray recirculation functions against post-accident debris blockage and operation with debris laden fluids results in a new sump strainer design to ensure adequate safety margin is maintained for emergency core cooling.

GDC 38, "Containment Heat Removal," mandates that a system be provided to remove heat from the reactor containment to reduce and maintain the containment pressure and temperature following any loss-of-coolant accident to acceptably low levels.

The proposed methodology change will continue to support heat removal of the reactor containment for long-term operations following an accident condition that includes sump recirculation operation.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20; however, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types of or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed El-12

change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment needbe prepared in connection with the proposed amendment.

7.0 REFERENCES

1.

TVA letter to NRC dated September 16, 1987, "Sequoyah Nuclear Plant -

Containment Coatings."

2.

NUREG-1232, Volume 2, "Safety Evaluation Report on Tennessee Valley Authority: Sequoyah Nuclear Performance Plan," dated May 1988.

3.

TVA letter to NRC dated September 1, 2005, "Sequoyah Nuclear Plant -

NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During-Design Basis Accidents at Pressurized Water Reactors -

Second Response."

4.

Nuclear Energy Institute Guidance Report No. NEI-04-07, "Pressurizer Water Reactor Sump Performance Evaluation."

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ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT (SQN)

UNITS 1 AND 2 Changes to'Updated Final Safety Analysis Report (Mark-up)

E2-1

Insert A As a result of the above information, TVA reevaluated the licensing basis for the containment sump screen blockage. While designed and constructed before the issuance of NRC Regulatory Guide 1.82, Revision 0, the original licensing basis for the containment recirculation sump included operability with an assumed 50 percent intake flow area blockage consistent with RG 1.82 recommendations. Scale model testing confirmed that the design of the original sump intake structure (i.e., a 6 inch trash curb around the base of the sump intake structure and 0.25 inch mesh intake screens sloping upward and outward from the sump opening) was sufficient to meet the 50 percent blockage criteria.

To address the potential increased failed coating debris load, an evaluation was performed using a two-dimensional physical transport model (Reference 70) to confirm the ability of the containment sump to support containment spray and emergency core cooling system pump operation. The evaluation methodology focused on a near-sump region of influence resulting from post-accident flow fields where debris transport to the sump intake was possible. The evaluation quantified the amount of accident generated debris which could potentially be transported to the sump intake structure, established the head loss from the resulting intake screen debris blockage and confirmed that the minimum containment spray and emergency core cooling system pump suction head requirements would be met for the expected blockage. A subsequent evaluation established a maximum limit on failed coatings which could be transported to the containment recirculation sump without degrading the capabilities of the required accident mitigation systems (Reference 71). This evaluation established the revised licensing basis for containment sump intake blockage. It was summarized and submitted for NRC review in Reference 85 and was accepted in Section 3.7 of Reference 86.

To address the additional concerns contained in NRC Generic Safety Issue No. 191 (GSI-191),

"Assessment of Debris Accumulation on PWR Sump Performance," the containment sump was subsequently reanalyzed to address the susceptibility of the emergency core cooling and containment spray recirculation functions to the adverse effects of post-accident debris blockage and operation with debris laden fluids. As summarized in Reference 87, the comprehensive reanalysis used the evaluation methodology described in Reference 88. The revised analysis methodology included development of a three-dimensional computational fluid dynamics model to establish debris transport characteristics (i.e.,

flow directions, velocities and turbulence) in the entire sump pool during post-accident sump recirculation operation. The results of the reanalysis were used to size the flow area of the advanced design containment sump strainers which replaced the original sump intake structure. Blockage testing of the advanced containment sump strainer design with a conservative debris load (which included the assumed failure of all qualified and unqualified coatings in containment) confirmed that the containment sump will support operation of the emergency core cooling system and the containment spray system under all anticipated debris loading conditions. This includes the failure of all coatings installed inside containment.

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IPage Submitted for Information Only SQN Containment Temperature Monitoring Temperature sensors are distributed throughout the ice bed of the ice condenser. Selected channels are displayed on a recorder in the main control room and provide actuation signals for the annunciation at preset deviations from the prescribed limits of the ice bed equilibrium temperatures.

Temperature sensors are strategically located throughout the containment. A group of these sensors is monitored in the main control room to ensure that the air temperature requirements for the proper operation of equipment is maintained.

Temperature monitoring and actuation signals associated with the heat removal system are described in Subsection 6.2.2.

Sump Level Control For a complete description of the Reactor Building sumps, see Subsection 5.2.7.

Leakage Detection Equipment For a complete description of the leakage detection methods and equipment, see Subsection 5.2.7.

Secondary Containment Instrumentation is provided in the main control room to give the operator information concerning the status of the active isolation valves, isolation dampers, and airlock doors in the secondary containment barrier.

The isolation valves and isolation dampers are equipped with limit switches. These limit switches show the full open or full closed position by indicating light in the main control room. For all the airlocks, each side of airlock is instrumented. For airlocks between secondary containment and primary containment, the instrumentation indicates the position of each airlock door and alarms in the main control room if both sides of an airlock are open simultaneously.

6.2.1.6 Protective Coatings Approximately 48,000 square feet of concrete surface is coated within the primary containment of each unit. These areas are coated with a catalyzed epoxy coating, which was tested in accordance with ANSI N101.2, "Protective Coatings for Light-Water Nuclear Reactor Containment Facilities," to demonstrate that the coating will remain intact on the surface to which it was applied during postulated LOCA conditions.

Major carbon steel components, such as the containment liner plates and domes, structural and miscellaneous steel, a large portion of the polar crane, etc., are protected with an inorganic zinc primer only, with no organic topcoat (approximately 79,000 square feet). In addition, approximately 46,000 square feet of inorganic zinc-primed steel is topcoated with a LOCA tested and approved catalyzed epoxy coating.

These coatings were tested in accordance with ANSI N101.2.

$6.2.doc 6.2-52 E2-3

SQN TVA is committed to adhere to Appendix B of 10 CFR 50 and ANSI N45.2 as required to produce a quality end product. Basically, TVA believes that the Quality Assurance Program (QA) for protective coatings inside the containment should control four activties in the coating program. The four major areas to be controlled are:

1.

The coating material itself, by extending requirements on the manufacturing progress and qualification of coating systems through the use of applicable portions of ANSI Standards N101.2 and N5.9 or its revision N512.

2.

The preparation of the surface to which coatings are to be applied.

3.

The inspection process.

4.

The application of the coating systems.

requirements.

All four of these controlled activities must have appropriate documentation and records to meet Appendix B TVA's protective coating program within the containment is in conformance with NRC Regulatory Guide 1.54. In addition, applicable provisions found in ANSI N101.4 have been incorporated into WVA surface preparation, coating application/inspection specifications, and coating QA procedures. In addition, all maintenance work on Coating Service Level I coatings shall use the same coating material as the existing system or a different coating that has been DBA qualified and approved for use with the existing system.

The amount of uncontrolled coatings allowed inside containment is limited to ensure that in a post-LOCA or MSLB environment, the uncontrolled coatings transported to the containment sump will not degrade the recirculation flow to the engineered safety systems via blockage of the containment sump screen or be ingested into the engineered safety systems and result in component degradation.

The original basis for qualification of coatings was the accident conditions resulting from a design basis LOCA. However, the containment temperature profile for the LOCA does not bound the temperature profile expected from an MSLB. Approximately 12,000 square feet of topcoated steel and 7,500 square feet of coated concrete inside containment, which were previously qualified, would not be qualified for the MSLB conditions------ - -

Replace with Insert A As a result of the above information, TVA reevaluated the licensing basis for the containment sum n

blockage with the help of a physical transport study performed by Westinghouse (Refere

). The purpose of the evaluation was to determine if the containment spray and emergenc cooling systems could be operated safely if debris were present from coating failures. The m ology focused on a near-sump region of influence based on post-accident flow fields and addr potential effects caused by the return of containment spray flow through the refueling canal d and by flow from the LOCA short-term blowdown. The Westinghouse study concluded that a mum of 84 ft2 of coatings could potentially fail in a DBA and be transported to the containmen culation sump without degrading the capabilities of the required accident mitigation systems (R ce 71).

I In addition, WVA 's N r i t r v l a e h

r p n i y f r v r e i g a o e t e s m o h B

Iconditionsanam oscenbokgpotltdithWetnhuesuyadcnlddhtte sumy Isump pef ce wsacpal.TeWtigoestud rvdsfracpaiiyorsn

~ cont "

ent coatings and was used as a basis to reestablish limits for the uncontrolled coatings log.

r-S6.2.oc

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SQN Removal of mirror insulation due to jet impingement or pipe whip due to a LOCA (and subsequent exposure of non-qualified coatings under the mirror insulation) on the Main Reactor Coolant Piping or Steam Generator is not considered credible based on the discussions in Section 3.6.1.1. Per Section 3.6.5.1, the dynamic effects of ruptures in the primary coolant loop have been eliminated.

6.2.2 Containment Heat Removal Systems 6.2.2.1 Design Bases Adequate containment heat removal capability for the Ice Condenser Reactor Containment is provided by the Ice Condenser (Section 6.5), the Air Return Fan System (Section 6.6), and the Containment Spray Subsystems whose components operate in the sequential modes described in Paragraph 6.2.2.2. One Containment Spray Subsystem consists of a Containment Spray train and a Residual Heat Removal Spray train (which is a portion of the Residual Heat Removal System Section 6.3).

The Containment Spray Subsystems consist of two trains of redundant equipment per reactor unit. There are four spray headers per unit. Two headers are supplied from separate Containment Spray trains; the other two are supplied by separate RHR Spray trains (see Table 6.2.1-1). Each individual train consists of a pump, a heat exchanger, appropriate control valves, required piping, and a header with nozzles located in the upper compartment of the containment with flow directed to obtain full coverage of the containment upper volume during an emergency. The systems use borated water supplied from the refueling water storage tank and/or the recirculation sump, as shown in Figure 6.2.2-1.

Minimum Engineered Safety Feature performance of the Containment Heat Removal Systems is achieved with the following:

1.

Ice Condenser (Section 6.5)

2.

One train of the Air Return Fan System (Section 6.6)

3.

One Containment Spray Train

4.

One Residual Heat Removal Spray Train (needed only after all the ice has melted)

The primary design basis for the Containment Spray Subsystems is to spray cool water into the containment atmosphere when appropriate in the event of a loss-of-coolant accident and thereby ensure that the containment pressure cannot exceed the containment shell design pressure as defined in Section 3.8.2.2. This protection is afforded for all pipe break sizes up to and including the hypothetical instantaneous circumferential rupture of the reactor coolant loop resulting in unobstructed flow from both pipe ends. The Containment Spray trains supplement the ice condenser until all the ice is melted approximately 3600 seconds after the LOCA at which time it and the Residual Heat Removal trains become the sole systems for removing energy directly from the containment. The Containment Heat Removal Systems are designed to provide a means of removing containment heat without loss of functional performance in the postaccident containment environment and to operate without benefit of maintenance for the duration of time necessary to restore and maintain acceptable containment conditions.

Although the water in the core after a loss-of-coolant accident is quickly subcooled by the Emergency Core Cooling System (Section 6.3), the design of heat removal capability of each Containment Heat Removal System is based on the conservative assumption that the core residual heat is released to the containment as steam which eventually melts all ice in the ice condenser.

$6.2.doc 6.2-54 E2-5

Insert B 85 R. Gridley to NRC, September-16,1987, "Sequoyah Nuclear Plant - Containment Coatings."

86.

NUREG-1232, volume 2, "Safety Evaluation Report on Tennessee Valley Authority: Sequoyah Nucrear Performance Plan," May 1988.

87.

Randy Douet to NRC, September 1.2005, "Sequoyah Nuclear Plant - NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors - Second Response."

88.

Nuclear Energy Institute Guidance Report No. NEI-04-07, "Pressurizer Water Reactor Sump Performance Evaluation."

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50.

Reference'deleted per Amendment 6.

51.