ML120680062
| ML120680062 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 02/22/2012 |
| From: | Steelman W Entergy Operations |
| To: | Benton L, Kalyanam K Office of Nuclear Reactor Regulation |
| Kalyanam N, NRR/DORL/LPL4, 415-1480 | |
| References | |
| TAC ME6049 | |
| Download: ML120680062 (9) | |
Text
Page 1 NRC Question:
AADB staff would like to review the 10 CFR 50.59 evaluation that the licensee conducted that supports the indicated change to Waterfords current AOR for an postulated FHA. The reason for our request is because AADB cannot conduct a confirmatory analysis based on the currently docketed information available. Whereas if the AST was still the current AOR, I could have simply compared the licensees proposed changes against the docketed values provided in their AST amendment. However, being that their current AOR is no longer the AST analyses, but is now based on the AOR resulting from their change supported by their indicated 10 CFR 50.59 evaluation, I dont actually have any reference values to compare the proposed change against. Therefore, I cant make a reasonable assurance finding on this LAR without reviewing the 10 CFR 50.59 evaluation to verify what the reference values for the current FHA AOR actually are. Please let me know if you have any questions, comments, and/or concerns.
Waterford 3 Response:
W3F1-2010-0009 Attachment 1 Section 4 (page 3) describes the analysis change for this submittal is that number of fuel pin failures has increased from 60 pins to 472 pins.
W3F1-2010-0009 Attachment 1 Section 4 (page 3) also states that the revised dose calculation was performed using the same methodology that was approved in the NRC Waterford 3 Alternate Source Term (AST) Safety Evaluation Report (SER). Entergy submittal W3F1-2010-0009 Attachment 1 Section 4 (page 4) contains the following table:
Analysis of New Analysis Regulatory 10CFR50.67 Record Guide 1.183 Limit AOR Limit EAB 0.58 rem TEDE 4.56 rem TEDE <6.3 rem TEDE <25 rem TEDE LPZ 0.089 rem TEDE 0.70 rem TEDE <6.3 rem TEDE <25 rem TEDE MCR 0.105 rem TEDE 0.824 rem TEDE <5 rem TEDE <5 rem TEDE This table provides the current analysis of record results. The AST methodology used RADTRAD to calculate the results from the corresponding source terms. Since the source term is directly proportional to the dose results, a confirmatory analysis (hand calculation) could be performed by multiplying the previous dose results by the source term ratio (number of failed fuel pins 472/60) to validate the dose consequences remain within the regulatory limits. The table below does this calculation for both the current analysis of record (AOR) and the AST SER results. The table demonstrates that using either one still provides acceptable results.
Page 2 Analysis of SER Analysis Multiply by Regulatory Record 472/60 Guide 1.183 AOR RESULT Limit EAB 0.58 rem TEDE 4.56 rem TEDE <6.3 rem TEDE EAB 0.550 rem TEDE 4.33 rem TEDE <6.3 rem TEDE LPZ 0.089 rem TEDE 0.70 rem TEDE <6.3 rem TEDE LPZ 0.090 rem TEDE 0.71 rem TEDE <6.3 rem TEDE MCR 0.105 rem TEDE 0.83 rem TEDE <5 rem TEDE MCR 0.190 rem TEDE 1.49 rem TEDE <5 rem TEDE Waterford 3 believes all the information necessary to perform a confirmatory analysis is contained in the W3F1-2010-0009 submittal.
If the NRC is questioning the difference between the AOR and the AST SER results, the dose results were changed under the 10CFR50.59 program. The 10CFR50.59 report and associated Updated Final Safety Analysis Report (UFSAR) were both submitted to the NRC [references below - I could not find ADAMS numbers]. The AOR control room results were impacted by two changes. The control room volume was reduced and the control room unfiltered inleakage was changed from 200 cfm to 100 cfm. The AST SER Table 14 (page 42) gives a summary of the control room unfiltered inleakages used in the dose analyses. The unfiltered inleakage ranged from 100 cfm to 200 cfm primarily based upon the assumption used when generating the calculations. The Tracer gas testing validated that the lower value of 100 cfm remained bounding. During the AOR revision to lower the control room volume, the unfiltered leakage value was also lowered. The unfiltered inleakage of 100 CFM was already approved in the AST SER for multiple events and remains bounded by the Tracer gas testing requirements. This resulted in the previous AST SER result dropping to the reported AOR value. Attached is the 10CFR50.59 evaluation and associated UFSAR change. The 10CFR50.59 analysis and UFSAR pertinent sections (attached) have been highlighted to provide easier identification of information.
References
- 1. Letter W3F1-2008-0026, Report of Facility Changes, Tests and Experiments and Commitment Changes, May 1, 2008.
- 2. Letter W3F1-2007 -0054, Final Safety Analysis Report - Revision 301, December 7, 2007.
WSES-FSAR-UNIT-3 TABLE 15.7-6 (Sheet 1 of 2) Revision 301 (09/07)
PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT (DRN 04-704, R14)
Core Power Level: 3735 MWt Core Inventory: (Ci/MWt)
Kr-85 4.520E+02 Kr-85m 1.378E-01 Kr-87 5.849E-14 Kr-88 2.071E-04 I-131 9.867E+01 I-132 3.985E+01 I-133 9.764E+00 I-134 9.488E-23 I-135 4.991E-02 Xe-131m 1.551E+02 Xe-133 1.667E+04 Xe-133m 3.713E+02 Xe-135 2.223E+02 Xe-135m 1.625E+00 Fission Product Gap Fractions:
I-131 12%
Kr-85 14%
Other Noble Gases 5%
Other Halogens 5%
Alkali Metals 12%
Fuel Rods Failing (maximum): 60 rods Iodine Chemical Form *:
Elemental 4.85%
Organic 0.15%
Particulate 95%
- The releases from the pool are conservatively modeled as 99.85% elemental and 0.15% organic iodine.
Control Room Parameters:
(EC-5000081470, R301)
Volume 168,500 ft3 (EC-5000081470, R301)
Recirculation Flow Rate 3800 CFM Iodine Filter Efficiency 99% (elemental/particulate/organic)
Pressurization Flow 225 CFM (EC-5000081470, R301)
Unfiltered Inleakage 100 CFM (EC-5000081470, R301)
Breathing Rate 3.47E-04 m3/sec.
Control Room Occupancy Factors 0-24 hours 1.0 (DRN 04-704, R14)
1 I. OVERVIEW I SIGNATURES Facility: Waterford 3 Evaluation # I Rev. #: -!o::._ _ __
Proposed Change I Document: ER-W3-2005-0344-000 Description of Change:
CR-WF3-2004-2077 identified that the control room HVAC equipment room is effectively isolated from the rest of the control room envelope since there is no ventilation into or out of the room.
Calculation ECM97-013, "Control Room Envelope Volume", was updated via DRN 05-1203 to account for areas such as HVAC room that have minimal air exchange with the rest of the control room envelope. Therefore, the control room design bases, including HVAC, radiological dose and toxic gas analyses, needed to be reevaluated to incorporate the lower control room envelope volume.
The impact on control room toxic gas analysis has been addressed in ER-W3-2005-0476-000. An EC to address HVAC and other Mechanical Systems impacts will be performed per CR-WF3-2005-2606.
ER-W3-2005-0344 was initiated to address the radiological dose impact of the reduced control room volume and incorporates other model changes as discussed below:
Radiological Calculations:
ECS04 series calculations for offsite and control room personnel dose due to FSAR Chapter 15 events were revised to assess the impact of a reduced main control room volume assumption, in response to CR-WF3-2004-2077 and CR-WF3-2005-2606. Where appropriate, a minimum control room volume of 168,500 ft3 was assumed; where appropriate, the previously assumed control room volume of 220,000 ft3 was retained as an assumed maximum volume. Other changes to the offsite and Main Control Room personnel dose calculations include:
- The Main Control Room ventilation model was revised to allow operations to either leave the control room in the Recirculation mode or, at any point in the event, to place the control room in the Pressurized mode. That is, the calculations will provide results that bound operation in either mode for the duration of the event. Operators are assumed to select the preferred intake when placing the control room in the Pressurized mode.
- The Shield Building Ventilation System (SBVS) Charcoal filtration is not credited for iodine removal from the containment annulus region.
- A very conservative post-accident initial 6 minute Positive Pressure Period is assumed for the containment annulus for the reactor building release model to allow future Technical Specification relaxations on Annulus Negative Pressure.
- The SBVS leakage pathway portion of the Reactor Building airborne release model is revised to allow for Maintenance Hatch leakage at 2 days into the event, vice the previous 7 days.
Note that calculations for filter charcoal loadings use assumptions consistent with those of the offsite and main control room intake radiological dose calculations. However, EQ dose calculations have not incorporated the SBVS related relaxations which have been incorporated into the offsite and main control room intake dose calculations. Thus, those relaxations (e.g., 6 minute Positive Pressure Period and not crediting charcoal filtration) should be viewed as discretionary conservatisms in the RADTRAD-code based offsite and main control room intake dose calculations. Revisions to the EQ calculations will be required to take full advantage of those relaxations. As discussed in memo 1 Signatures may be obtained via electronic processes (e.g., PCRS, ER processes), manual methods (e.g., ink signature), e-mail, or telecommunication. If using an e-mail or telecommunication, attach it to this form.
10 CFR 50.59 EVALUATION FORM Sheet 2 of 6 W3Cl-2007-0006, the EQ dose calculations are not impacted by allowing Maintenance Hatch leakage at 2 days vice the previous 7 days.
Is the validity of this Evaluation dependent on any other change? DYes (8) No If "Yes," list the required changes/submittals. The changes covered by this 50.59 Evaluation cannot be implemented without approval of the other identified changes (e.g., license amendment request).
Establish an appropriate notification mechanism to ensure this action is completed.
Based on the results of this 50.59 Evaluation, does the proposed change DYes (8) No require prior NRC approval?
Preparer:
Paul Sicard /
Reviewer:
OSRC:~.~~
Chairman's N 01- () 1'6 OSRC Meeting #
II. 50.59 EVALUAliON Does the proposed Change being evaluated represent a change to a method of evaluation ONLY? If "Yes," Questions 1 - 7 are not applicable; answer only Question 8. If "No," answer DYes all questions below. rgJ No Does the proposed Change:
- 1. Result in more than a minimal increase in the frequency of occurrence of an accident DYes previously evaluated in the UFSAR? rgJ No BASIS:
The proposed activity involves changes to licensing and design basis calculations and documents to account for a reduction in the assumed main control room volume and incorporation of other modeling enhancements in the post accident radiological dose calculations. The proposed activity does not involve any physical changes to the plant equipment or the manner in which they operate. There is nothing associated with this change which would impact system performance or reliability, affect any system interface in a way that could lead to an accident, or increase the probability of operator error to cause any event previously evaluated in the FSAR. Thus, the proposed activity has no impact on the frequency of occurrence of an event previously evaluated in the FSAR associated with this change.
- 2. Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a DYes structure, system, or component important to safety previously evaluated in the UFSAR? rgJ No BASIS:
The proposed activity involves changes to licensing and design basis documents only, that document acceptable consequences for various radiological events analyzed in FSAR Chapter 15 (e.g., LOCA, Main Steam Line Break (MSLB), etc.).
The Small Break LOCA main control room dose calculation had previously credited the 65 CFM unfiltered in leakage corresponding to the Pressurized Mode at beyond 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> into the event. Now, consistent with other analyses, it credits a larger 100 CFM unfiltered inleakage, corresponding to the Recirculation Mode, throughout the event.
The calculations demonstrate that filter loadings continue to be acceptable. There is 8_ negligible impact on EQ dose due to the assumption of a reduced main control room volume. There is no impact on EQ zone maps (Le., on post-accident radiation levels that must be assumed for EQ) associated with the calculation changes for ER-W3-2005-0344, thus no associated increase in the likelihood of equipment malfunction. Because of the extensive workscope that would be involved in removing credit for fission product scrubbing of elemental iodine from EQ calculations, the relaxations in SBVS modeling assumptions have not been incorporated into EQ calculations. As discussed in memo W3Cl-2007-0006, the EQ dose calculations are not impacted by allowing Maintenance Hatch leakage at 2 days vice the previous 7 days. ER-W3-2005-0344 does not involve any changes to any plant equipment or the manner in which they operate. Therefore, the proposed activity does not impact the likelihood of a malfunction of an SSC important to safety previously evaluated in the UFSAR.
- 3. Result in more than a minimal increase in the consequences of an accident previously DYes evaluated in the UFSAR? rgJ No BASIS:
The calculations performed in support of ER-W3-2005-0344 continue to demonstrate that acceptance criteria for offsite and control room doses. No changes to the licensing basis results for offsite and main control room dose is required in FSAR Chapter IS, which reports the results of the radiological dose consequences as being less than or equal to the applicable RG 1.183 guidelines and GDC 19 control room dose limits for all reported radiological events.
The calculations assume a maximum 100 CFM unfiltered inleakage value, corresponding to the licensing basis for the Control Room in the Recirculation mode. The analysis for the Inside Containment Main Steam Line Break relaxes a previous excess conservatism, and now assumes a source term based on the 10%
DNBR fuel failure limit established by the Chapter 15 NSSS analysis, vice the previous and more conservative 2% fuel melt. Excess conservatism (i.e., excess conservatism in release fractions) in the noble gas source term for Small Break LOCA has also been removed. Therefore, the proposed activity does not result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR.
- 4. Result in more than a minimal increase in the consequences of a malfunction of a structure, DYes system, or component important to safety previously evaluated in the UFSAR? rgJ No BASIS:
The calculations performed in support of ER-W3-2005-0344 demonstrate that (1) the offsite and control room doses are within the acceptable limits and (2) the SBVS, CVAS and control room charcoal heat and mass loadings remains acceptable, and (3) the post-accident radiation levels specified in the EQ Zone Maps are unchanged by this ER. Therefore, this activity does not result in more than a minimal increase in the consequences of a malfunction of a structure, system, or component important to safety previously evaluated in the UFSAR.
- 5. Create a possibility for an accident of a different type than any previously evaluated in the DYes UFSAR? rgJ No BASIS:
Because there are no new interactions or relationships created by ER-W3-2005-0344, there is no possibility of creating an accident of a different type than previously evaluated. The proposed activity involves changes to licensing and design basis documents only. It does not involve any changes to any plant equipment or the manner in which they operate. There are no physical changes to the plant or changes to plant procedures required as a result of this change. No new accident initiators are introduced by this activity. Thus, since this ER does not impact any equipment, it does not create a possibility for an accident of a different type than any previously evaluated in the FSAR.
- 6. Create a possibility for a malfunction of a structure, system, or component important to safety DYes with a different result than any previously evaluated in the UFSAR? ~ No BASIS:
The proposed activity does not involve any physical changes to the plant. There are no required changes to plant procedures associated with the change. The proposed change provides increased flexibility to plant operators to switch the control room from the Recirculation mode to the Pressurized Mode any time following an accident, thus eliminating the sensitivity of the results of licensing basis radiological analyses to operator actions. Operators are assumed to select the preferred intake when taking the action to put the control room in the Pressurized mode; independent, redundant monitors are located at each intake to provide operators with the necessary information to select the preferred intake. Therefore, the proposed activity does not create a possibility for a malfunction of a SSC important to safety with a different result than any previously evaluated in the FSAR.
- 7. Result in a design basis limit for a fission product barrier as described in the UFSAR being DYes exceeded or altered? ~ No BASIS:
This ER is concerned with radiological analyses, which determine the radiological consequences (addressed in Questions 3 and 4) associated with FSAR Chapter 15 events, including those that involve postulated fuel failure. The control room and offsite dose consequences for all events were found to be within acceptable regulatory limits and no changes are required to the consequences documented in FSAR Chapter 15. The analysis for the Inside Containment Main Steam Line Break does relax a previous excess conservatism, and now assumes a source term based on the 10% DNBR fuel failure limit established by the Chapter 15 NSSS analysis, vice the previous and more conservative 2% fuel melt. The iodine and heat loading on all safety related HVAC systems were found to be acceptable. This ER does not result in changes to any of the fission product barriers (fuel cladding, Reactor Coolant System boundary and containment). Thus, there is no impact on any fission product barriers which are associated with calculations for offsite and control room dose, but rather these calculations determine the consequences which can result due to fuel failure or other failures of fission product barriers.
There is no impact on the design basis limits for the RCS (e.g., primary pressure or stresses) associated with how the release is determined. Thus, ER-W3-2005-0344 does not result in exceeding or altering any design basis limit for fission product barriers.
- 8. Result in a departure from a method of evaluation described in the UFSAR used in establishing DYes the design bases or in the safety analyses? ~ No BASIS:
There have been no changes in methodology associated with the proposed ER. All changes in the calculations are changes in input parameters and assumptions, such as the modeling of control room ventilation and of the shield building release pathways. The RADTRAD code, as with its methodology described in FSAR Section 15.B, continues to be used for offsite and control room intake dose calculations.
The radiological analyses performed to support ER-W3-2005-0344 continue to follow Regulatory Guide 1.183 methodology for Alternative Source Term
evaluations. Regulatory Guide 1.183 provides guidance for implementing the AST dose methodology into the plant design and licensing basis.
Control room personnel shine dose calculations and charcoal filter loading calculations continue to use the same methodology, and have been revised based upon conservative scaling to use modeling assumptions consistent with the offsite and control room intake dose calculations. The changes to EQ calculations associated with ER-W3-2005-0344 clarify the bases of those calculations and the EQ calculations have not been revised to incorporate the relaxations (e.g., 6 minute Positive Pressure period, not crediting charcoal filtration) incorporated as discretionary conservatisms in the offsite and main control room intake calculations; thus, there is no change in methodology.
Thus, there is no departure from a method of evaluation described in the FSAR associated with ER-W3-2005-0344.
If any of the above questions is checked "Yes," obtain NRC approval prior to implementing the change by initiating a change to the Operating License in accordance with NMM Procedure EN-U-103.