ML22195A159
| ML22195A159 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley, Millstone, Browns Ferry, Salem, Indian Point, Kewaunee, Harris, Wolf Creek, Point Beach, Watts Bar, Sequoyah, Byron, Braidwood, Summer, Prairie Island, Seabrook, Surry, North Anna, Turkey Point, Ginna, Diablo Canyon, Callaway, Vogtle, Farley, Robinson, South Texas, Cook, Comanche Peak  |
| Issue date: | 03/02/2007 |
| From: | Alsup R Westinghouse |
| To: | Bhatt S NRC/NRR/DSS/SNSB |
| Santosh Bhatt NRR/DSS 301-415-6066 | |
| References | |
| NSAL-06-6 | |
| Download: ML22195A159 (8) | |
Text
TO:
Those indicated L33 070302 004 FROM:
R. E. Alsup, Industry Affairs, BR 4X-C DATE:
March 2, 2007
SUBJECT:
TRANSMITTAL OF OE ITEM FOR ACTION OR INFORMATION - NSAL-06-06 Attached is an OE item being forwarded to each of the indicated organizations for action or information as required. Please provide your response, if action is required for your organization, by N/A. If action is taken on an OE item that was sent to you for information, please notify Corporate OE to correct the assignment.
ACTION INFORMATION BFN - E. D. Charlton
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SQN - P. W. Wilson
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[ x ]
WBN - S. D. Hudson
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[ x ]
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Engineering & Technical Svs:
(ATTN: Chuck Whitehead) [ ]
[ x ]
Rad Health/Chemistry/EP W. A. Nurnberger, BR 4T-C
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Nuclear Fuels J. F. Lemons, BR 3F-C
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Plant Operations Reliability A. J. Scales, BR 4T-C
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Materials Inventory C. L. Martin, LP 4W-C
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Nuclear Operation:
Nuc Assurance & Licensing NSRB Support, BR 4X-C
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Process Methods/Security H. R. Rogers, BR 4T-C
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Human Resources Industrial Safety, LP 3A-C
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Other Andrew Neal, HDC 1A-DST
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OE EVALUATION/RESPONSE (A)
APPLICABILITY EVALUATION NER NO.:
06-0892 TROI ID:
N/A (1)
Title:
Westinghouse NSAL-06-6, LOCA Mass and Energy Release Analysis (2)
Keyword(s):
Safety Analysis, LOCA, Loss of Coolant Accident, Containment, Containment Integrity (3)
Actions:
TVAN Corp Mech/Nuc Engineering (I) - Information (AI) - Additional Info (A) - Action (PA) - Previously addressed (NA) - Not Applicable (4)
Action Priority:
20 working days (5)
Initial Screening Reviewer:
Steve Hudson Date:
7/6/06 (6)
Screening Mtg. Date:
9/21/06 (7)
Applicability Explanation:
This NSAL states that Westinghouse identified eight procedural issues with past guidance provided for the performance of loss-of-coolant accident (LOCA) mass and energy (M&E) release analyses by Westinghouse which could potentially affect the M&E results used to evaluate containment integrity. Most plants are only affected by a subset of the issues and some are not affected by any issues. The impact of each issue is determined by assessing an appropriate pressure increase for each plant as appropriate. If a pressure increase is assessed, the limiting case may be unaffected because the increase occurs in the part of the transient that is well below the limiting result for a specific plant. For all affected plants, it has been determined that the maximum increase resulting from all of the issues combined would not cause the containment pressure to exceed the design limit. The purpose of this notification is to inform utilities of the M&E analysis issues and their potential impacts on peak containment pressures and temperatures. These issues should also be reviewed with respect to containment conditions assumed for other components and containment conditions for equipment qualification.
This item is N/A to BFN. WBN and SQN initiated PERs 105564 and 111026, respectively, to address this NSAL.
Additionally, Westinghouse issued letters WAT-D-11445 & WAT-D-11446 and TVA-06-30 in support of the issues of this NSAL not affecting WBN & SQN. This item will be sent for Information to TVAN Corporate Mechanical/Nuclear Engineering.
(8)
Related Documents:
PERs 105564 & 111026; WAT-D-11445, WAT-D-11446, & TVA-06-30 (9)
OE Preparer:
Steve Hudson Ext 365-3796 Date 3/02/07 (10 OE Manager:
Bob Alsup Ext 751-8251 Date 3/02/07 (B)
OE ITEM RESPONSE
SUMMARY
(1)
OE Item is Applicable:
Yes No (2)
PER Initiated:
Yes No (3)
Implementation Date (4)
Item Needs Further Review Yes No (If yes, explain below.)
(5)
Response Summary:
(6)
Item Closed:
Yes No (7)
Responsible Manager:
Ext.
Date TVA 40550 [09-97]
Page 1 of 1 NADP-3-1-1 [09-15-97]
Electronically approved records are authenticated in the Electronic Document Management System Nuclear Safety Advisory Letter Westinghouse Electric Company This is a notification of a recently identified potential safety issue pertaining to basic components supplied by Westinghouse.
This information is being provided so that you can conduct a review of this issue to determine if any action is required.
P.O. Box 355, Pittsburgh, PA 15230
Subject:
LOCA Mass and Energy Release Analysis Number: NSAL-06-6 Basic Component: Containment Analysis Date:
06/06/2006 Affected Plants: See page 6 for Plants with Westinghouse Mass & Energy Analysis Substantial Safety Hazard or Failure to Comply Pursuant to 10 CFR 21.21(a)
Transfer of Information Pursuant to 10 CFR 21.21(b)
Advisory Information Pursuant to 10 CFR 21.21(d)(2)
Yes No Yes Yes
References:
See page 6.
SUMMARY
Westinghouse has identified several procedural issues with past guidance provided for the performance of loss-of-coolant accident (LOCA) mass and energy (M&E) release analyses (References 1 & 2) by Westinghouse.
Eight issues have been identified that could potentially affect the M&E results which are used to evaluate containment integrity. Most plants are only affected by a subset of the issues and some are not affected by any issues. The impact of each issue is determined by assessing an appropriate pressure increase for each plant as appropriate. If a pressure increase is assessed, the limiting case may be unaffected because the increase occurs in the part of the transient that is well below the limiting result for a specific plant.
For all affected plants, it has been determined that the maximum increase resulting from all of the issues combined would not cause the containment pressure to exceed the design limit. The existence of these issues does not create a substantial safety hazard and is not reportable under 10 CFR 21.
The purpose of this notification is to inform utilities of the M&E analysis issues and their potential impacts on peak containment pressures and temperatures. These issues should also be reviewed with respect to containment conditions assumed for other components and containment conditions for equipment qualification. While this NSAL provides a generic discussion of the issues, plant specific information has been provided by separate letter for most plants at this time. Please contact your Customer Project Manager to obtain each individual plant notification of these issues and the impact on your plant. Most of these letters were issued in the Fall of 2005. Even though the plants impacted by the M&E analysis issues have been notified independently, it is prudent to document these issues in this NSAL.
Additional information, if required, may be obtained from the originators. Telephone (412) 374-5750 and 4079 Originators:
Approved:
J. T. Crane Regulatory Compliance and Plant Licensing B. F. Maurer, Acting Manager Regulatory Compliance and Plant Licensing L. C. Smith Containment and Radiological Analysis
NSAL-06-6 Page 2 of 6 ISSUE DESCRIPTION The LOCA results in M&E releases inside containment which cause an increase in the containment pressure during the accident. These M&E releases are used to determine the maximum calculated containment pressure which is compared to the containment design limit. Westinghouse has identified procedural issues with past guidance provided for performance of M&E release analyses. One issue concerns the analytical methodology, while the remaining issues are concerned with the selection and application of input parameters.
TECHNICAL EVALUATION Two approved M&E methods (References 1 & 2) have been used for the initial safety analysis and/or for evaluating plant changes such as upratings and steam generator (SG) replacement. The earlier M&E analysis method, WCAP-8264-P-A, is considered a mass driven model since it does not allow for steam condensation in the reactor coolant system (RCS) loops. Therefore, issues that would affect primary mass releases are most important to the WCAP-8264-P-A model and issues that affect SG secondary side energy transfer have a limited effect. The WCAP-10325-P-A model is more energy driven as a result of RCS loop condensation. Therefore, the issues that affect the SG secondary side have more effect. The estimated affect on the peak pressure for the following generic issues are bounding for both methods.
- 1)
Area of the Downcomer in the REFLOOD Code Westinghouse-designed reactors can be divided into downflow and upflow barrel baffle designs. The original guidance for calculating the downcomer area for downflow plants was incorrect. When the calculation was corrected, a larger downcomer area and thus a larger volume for downflow plants was calculated. This resulted in a longer time required for the emergency core cooling system (ECCS) to completely fill the downcomer. Sensitivity studies showed that the effect on LOCA M&E release, as determined by the effect on containment pressure, was about 1.2 psi.
- 2)
Area of the Upper Plenum in the FROTH Code The FROTH computer program is run in conjunction with the REFLOOD computer program and calculates the LOCA M&E releases for the post-reflood period until the SG secondary side pressure(s) is calculated to equilibrate at the containment design pressure. During this time period, the two-phase mixture levels in the core, upper plenum, hot leg and SG inlet plenum are the principle parameters of interest. Due to a misinterpretation of a database parameter, the cross-sectional area of the upper plenum (AUPP) was being over predicted, which resulted in a reduction in entrainment to the SGs and thus less steam production. Correction of the upper plenum area results in increased M&E releases. For plants for which Westinghouse calculated the containment pressure transient, the increase in peak pressure is approximately 0.8 psi. This penalty would only apply to plants that calculated the peak pressure to occur during the post-reflood period. However, if the double-ended hot leg (DEHL) break is currently the limiting break, the potential also exists to switch the most limiting break from the blowdown limited DEHL break to the double-ended pump suction (DEPS) break if the pressure difference between these breaks is less than 1 psi.
NSAL-06-6 Page 3 of 6
- 3)
Review of Other FROTH Inputs A review of the FROTH code user controlled input variables showed that ASGP, the SG inlet plenum area, which is used to calculate the void fraction in the SG inlet plenum, was based on a value that was generally too small. A value for ASGP that is smaller than can be justified based on the geometry of the SG is conservative with respect to calculation of the peak pressure. This is because the void fraction sets the mixture height in the SG plenum which has an effect on the entrainment of liquid into the SG tubes.
A lower void fraction is a benefit and occurs when the flow area is increased. A review of SG geometry for the inlet plena determined a more appropriate method for calculating ASGP. The result was a larger flow area and a reduction in entrained liquid. The reduction in entrained liquid reduces the mass and energy released post-reflood and is a benefit for the calculated containment pressure.
- 4)
WCAP-10325-P-A Model Features The Westinghouse LOCA M&E release model (Reference 2) was approved in February 1987.
Westinghouse identified the need to clarify two model features; the assumptions placed on the SG exit steam enthalpy during the post-reflood period and the assumed power level used in the LOCA M&E analysis. This issue has been resolved through discussions with the NRC. Their acceptance is documented in Reference 3. The NRC staff found that the assumption used for the SG exit enthalpy during the post-reflood period is conservative. The NRC staff also found that Westinghouses understanding of performing analyses at the licensed core power, regardless of power level, is an acceptable method, as long as the plant-specific calorimetric uncertainty is considered.
- 5)
Main Feedwater Addition Following a Reactor Trip The Westinghouse methods account for the addition of main feedwater (MFW) to the SG secondaries following a LOCA in the time frame from reactor trip until MFW isolation is calculated to occur. Full MFW flow is assumed prior to the reactor trip. The recent review called into question the current modeling of the isolation of the SG secondary side on a reactor trip signal. The continued addition of MFW after reactor trip is adding energy to the secondary side above 212ºF and therefore in the long-term this additional energy will be released to the containment. DEHL breaks are not affected and depending upon the time at which peak containment pressure is calculated, a penalty to peak pressure may occur for the DEPS minimum ECCS case. Sensitivity studies have shown either a penalty or a benefit when MFW addition after reactor trip is modeled. Generally, plants with a peak pressure result late in time (i.e., post-FROTH) may show a small penalty.
- 6)
AFW System Purge and Unisolatable Volumes After isolation of the MFW, a volume of hot MFW will reside in the main feed lines between the auxiliary feedwater (AFW) injection point and the SG secondary side. Once AFW flow is initiated, the hot MFW water will be pushed into the SG secondary side. As the SGs are calculated to depressurize, there may be additional volume trapped between the AFW injection point and the MFW isolation valve that will flash and be pushed into the SG secondary. These two concerns were not considered in the LOCA mass and energy release models. Addition of these effects to the LOCA mass and energy release calculation has been shown to increase the SG secondary side energy above 212ºF by several million BTUs. Thus, depending upon various plant design features and the time at which the containment peak pressure was calculated to occur, an effect on containment peak pressure of 1.3 psi is possible. If the steam generator design for a particular plant has separate MFW and AFW piping, this issue would not apply.
NSAL-06-6 Page 4 of 6
- 7)
AFW Flow for FROTH Code During work to support new LOCA M&E release analyses, information was requested on minimum AFW flow per SG, post-LOCA. In some cases it was found that the actual flow used in the analysis was a pump flow and not flow per a SG. This resulted in SG AFW flows that were high. AFW flow is only modeled in the FROTH code which calculates the transient from end of reflood until the SG secondary side pressure has been calculated to have depressurized to the containment design pressure. This is a time range of about 200 to 1000 seconds and varies depending upon plant type (i.e., 2, 3, or 4 loop), power level and containment design pressure. The effect on analysis results for reductions in AFW flow during this short period is an approximate penalty of 0.1 psi in peak containment pressure.
For plants that currently have a peak during the post-reflood phase and do credit some AFW flow, the overall peak pressure would be higher if the amount of AFW flow is reduced. For plants that currently have the overall peak pressure during the blowdown period or the reflood period, and some amount of AFW flow credited, it is possible that the overall peak pressure could shift to the post-reflood period if the amount of AFW flow was reduced. For plants that currently do not credit any AFW flow during the post-reflood period, there is no impact on the analysis results.
- 8)
Possibility of Asymmetric AFW Flow LOCA analyses are performed assuming that off-site power is lost coincident with the event, and with the limiting single failure of one diesel generator to start. If the plant design does not start the turbine-driven AFW pump on the loss-of-offsite power or a safety injection (SI) signal, the typical design will have one motor-driven AFW pump in operation which generally will not feed all SGs. Thus, one or more SGs may not receive any AFW flow. If the broken loop SG is the SG with no AFW flow, there will be an effect on the calculated LOCA M&E release. The current LOCA M&E release models do not contain a provision to model asymmetric AFW flow. Instead, this effect is bounded by the assumption of no AFW delivery.
Modeling no AFW flow can result in a peak containment pressure penalty of up to 1.0 psi. The magnitude of this effect on peak pressure and temperature is dependent upon the transient characteristic for any given plant. Plants that are limited by a blowdown peak are generally unaffected. Plants that calculate a peak pressure after reflood is over will calculate a penalty depending upon the AFW assumptions in the current analysis.
Combined Effect of All Eight Issues In reviewing the plant specific M&E analyses performed by Westinghouse, no plant was affected by all the issues described above. Additionally, some of the affected plants have no impact on the limiting case of the current analysis of record because the maximum calculated containment pressure is unaffected.
For the affected plants, the time of the peak pressure from the current analysis of record was determined and the effects that occur during the time of the peak pressure were combined using the square root of the sum of the squares (SRSS) technique. That is, each effect for specific analysis should be combined using the square root of the SRSS method to determine the overall increase in the DEPS minimum ECCS case.
Once the new DEPS-limiting ECCS case peak pressure has been determined, this value is compared against the DEHL case to assure that the limiting break location has not changed.
NSAL-06-6 Page 5 of 6 SAFETY SIGNIFICANCE The M&E release models contain known conservatisms which could be used for operability determinations to offset any increase in containment pressure due to these issues. WCAP-10325-P-A documents the conservatisms inherent in the methodology in Section 5.1, Model Conservatisms, which are on the order of 6 or more psi. Conservatisms discussed are:
- 1) Appendix K models used in reflood.
- 2) Assumptions related to SG secondary mass (i.e., nominal level plus uncertainty).
- 4) Uncertainty applied to decay heat and conservatism applied to the 1979 decay heat model.
If a plant was analyzed with the full scope of changes as previously discussed, the sum of all effects is
~5.8 psi increase in the calculated peak pressure. The SRSS of 2.6 psi is closer to the total effect seen for plants already analyzed with all effects. Further, the model benefits and conservatism associated with the WCAP-10325-P-A LOCA mass and energy release model can be applied to analyses performed with the WCAP-8264-P-A LOCA mass and energy release model.
The issues were evaluated for ice condenser plants where the containment pressure is controlled by the melting of the ice inside containment. Instead of applying the impact in a pressure increase, the penalty was converted into an energy value. Benefits were found in the calculation of the SG secondary mass, and other analysis inputs, that were greater than the additional energy.
Since the potential increase in the maximum calculated containment pressure does not exceed known model conservatisms, all containments would remain below their design pressure. Therefore, reasonable assurance exists that containment integrity would be maintained and that a substantial safety hazard would not be created.
NRC AWARENESS The NRC was not notified under the Part 21 process.
There have been communications with the NRC on topics dealing with WCAP-10325-P-A model features (SG exit enthalpy, NSSS power level) described in issue number 4 of the Technical Evaluation section above. The resolution of issue number 4 was found to be acceptable by the NRC (Reference 3).
RECOMMENDED ACTIONS If available, each plant should review the plant specific information provided via Westinghouse project letter. If Westinghouse performs the containment analysis, the maximum containment pressure may need to be revised. In this case, it is recommended that the utility review its documentation which refers to the value of maximum containment pressure. If Westinghouse provided the LOCA M&E results but did not perform the containment analysis, the utility should review their containment evaluation with respect to the issues identified in the project letter or alternatively this NSAL. Furthermore, these issues should be reviewed with respect to containment conditions assumed for other components and containment conditions for equipment qualification.
NSAL-06-6 Page 6 of 6 REFERENCES
- 1. WCAP-8264-P-A, Revision 1, Topical Report Westinghouse Mass and Energy Release Data for Containment Design, 8/31/1975.
- 2. WCAP-10325-P-A, Westinghouse LOCA Mass and Energy Release Model for Containment Design March 1979 Version, 5/1/1983.
- 3. Letter from Herbert N. Berkow (NRC) to Mr. James A. Gresham (Westinghouse): Acceptance of Clarifications of Topical Report WCAP-10325-P-A, Westinghouse LOCA Mass and Energy Release Model for Containment Design March 1979 Version (TAC No. MC7980);
October 18, 2005.
Plants with Westinghouse Mass & Energy Analyses Almaraz Angra Unit 1 Ascó Beaver Valley Units 1 & 2 Beznau Byron Units 1 & 2 Braidwood Units 1 & 2 Callaway Comanche Peak Units 1 & 2 D. C. Cook Units 1 & 2 Diablo Canyon Units 1 & 2 Farley Units 1 & 2 Ginna Indian Point Unit 2 Indian Point Unit 3 Kewaunee Kori Units 1 & 2 Kori Units 3 & 4 Krsko Maanshan Millstone Unit 3 North Anna Units 1 & 2 Point Beach Units 1 & 2 Prairie Island Units 1 & 2 Ringhals Units 3 & 4 Robinson Unit 2 Salem Units 1 & 2 Seabrook Unit 1 Sequoyah Units 1 & 2 Shearon Harris Unit 1 South Texas Units 1 & 2 Surry Units 1 & 2 Turkey Point Units 3 & 4 V.C. Summer Vandellós Vogtle Units 1 & 2 Watts Bar Wolf Creek Yonggwang Units 1 & 2 This document is available via the Internet at www.rle.westinghousenuclear.com. This site is a free service of Westinghouse Electric Co. but requires specific access through a firewall. Requests for access should be made to kleinwd@westinghouse.com.