ML20199H003
| ML20199H003 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 01/21/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20199G993 | List: |
| References | |
| GL-96-03, GL-96-3, NUDOCS 9802040391 | |
| Download: ML20199H003 (13) | |
Text
_ - _ _ _ - _ _ _ _ _ _ _ - _ _ _ - - _
, peaasog ge +4 UNITE] STATES
- j. j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30006 4001
\ . . . . . /*
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REVIEW OF TNE PRESSURE TEMPERATURE LIMITS REPORT (PTLR) AND METHODOLOGY FOR THE RELOCATION OF THE REACTOR COOLANT SYSTEM (RCS)
PRESSURE-TEMPERATURE LIMIT CURVES AND LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP) SYSTEM LIMITS RELATED TO FACILITY OPERATING LICENSES NPF 37. NPF-66. NPF 72 AND NPF 77 COMMONWEALTH EDISON COMPANY BYRON STATION. UNITS 1 AND 2. AND BRAIDWOOD STATION. UNITS 1 AND 2 DOCKET NUMBERS 50-454. 50-455. 50-456. AND 50-457
1.0 INTRODUCTION
l By letter dated May 21,1997 (Reference 5), and supplemented by letters dated November 18, 1997 (Reference 6), December 3,1997 (Reference 7), January 8,1998 (Reference 9) and January 13,1998 (Reference 10), Commonwealth Edison Company (Comed), requested changes to the technical specifications (TS) for Bpon Units 1 and 2, ana Braidwood Units 1 and 2. The requested changes included (1) developing new reactor coolant system (RCS) pressure temperature (P/T) limit curves and low temperature overpressure protection (LTOP) system limits, (2) relocating the P/T limit curves and LTOP system limits from the TS to a licensee controlled document identified as a Pressure Temperature Limits Report (PTLR), and (3) changing the affected limiting conditions for operation and bases accordingly. These changes are made in accordance with Generic Letter (GL) 96 03, " Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits," dated January 31,1996, in addition, Comed also requested a change to relocate the surveillance capsule withdrawal i
schedule from the TS to the PTLR. Relocation of the reactor vessel surveillance capsule withdrawal schedule is made in accordance with GL 91-01, " Removal of the Schedule for the Withdrawal of Reactor Vessel Material Specimens from Technical Specifications," except that the schedule is relocated to the PTLR rather than the UFSAR. Changes to this schedule continue to be controlled by the requirements of Appendix H to 10 CFR Part 50.
The licensee's request to relocate the P/T limit curfes and LTOP system limits to the PTLR was submitted consistent with the guidance provided in GL 96-03 and WCAP 14040 NP-A (Reference 2), with three exceptions. These exceptions, discussed in Attachment E of the licensee's May 21,1997, submittal, involved (1) the computer program used for determination of the LTOP supoints, (2) the neutron transport cross section librar/ and dosimeter reaction cross sections used in determining the fluences, and (3) the version of the ASME Code used in determining the P/T limit curves and LTOP system limits . Exceptions 1 and 2 are addressed elsewhere in this evaluation. Exception 3 was addressed, in part, in the evaluation authorizing the use of the methodology in the ASME Section XI, Appendix G,1996 Addenda (Reference 11).
9002040391 900121 PDR ADOCK 05000454 P PDR
e 2
The Commission granted Byron and Braldwood exemptions (References 1,3, and 8) from the requirements of 10 CFR 50.60,' Acceptance Criteria for Fracture Prevention Measures for Lightwater Nuclear Power Reactors for Normal Operation," that allowed the plants to use the methodology in ASME Code Case N 514. Byron and Braidwood were also granted exemptions (Reference 11), permitting the use of the methodology described in toe 1996 ASME Code,Section XI, Appendix G, Article G 2000, in addition, the staff approved (Reference 12), the l Integration ef the Byron 1 and 2 and Braidwood 1 and 2 weld metal surveillance programs.
With regard to the LTOP system limits, the Code Case and the 1996 Addenda provide es:sntially the same guidance, in that they both recommend that LTOP systems limit the pressure in the vessel to 110% of the P/T limits and allow the use of enable temperatures less than 200'F or w coolant temperature corresponding to a reactor vessel metal temperature less than RT,a + 50'F, whichever is greater.
2.0 BACKGROUND
2.1 Neutron Fluence The fluence evaluation which is the basis for the current P/T curves was performed when the first fcur surveillance capsules were removed and evaluated at the end of the first cycle for each plant. However, at the time these evaluations were performed, ENDF/B IV based cross sections were used. Since that time the staff has identified a number of nonconservative values in the Iron inelastic cross sections. The cross sections now recommended by the staff are based on the ENDF/B VI c.oss section file.
The licensee proposed to move the P/T limit curves to the PTLR before an integr.:ted fluence reevaluation is performed based on the ENDF/B VI cross sections. This is scheduled to occur after the removal of the next set of surveillance capsules. This reevaluation will change the manner in whicia the materials are utilized, and, therefore will change the PTLR methodology.
Becaus? of the changes, the methodology will then have to be submitted to the NRC for prior review and approval. The licensee stated that the maximum operating time used to generate the P/T limit curves will be at most 85.8% of the current licensing basis time. Therefore, the revised operating time used to generate the curves will be no greater than 85.8% of the currently approved maximum value. This evaluation establishes a conservative fluence estimation. The f onconservatism of the ENDF/8 IV file is about 20%. The conservatisms are 14.2%, due to the shorter operating time; 6%, due to the initial adjustments to the measured data; and 5%, due to I
the effect of the low leakage loadings which have been practiced in all units since the second
{ cycle.
2.2 Pressure Temperature Limits The methodologies for assessing PTT limits and reactor pressure vessel (RPV) surveillance programs are discuased, in part, in the following documents: (1) 10 CFR Part 50, " Appendix G -
Fracture Toughness Requirements"; (2) 10 CFR Part 50, ' Appendix H - Reactor Vessel Material Surveillance Pregram Requirements"; (3) 10 CFR 50.60 " Acceptance Criteria for Fracture Prevention Measures for Lightwater Nuclear Power Reactors for Normal Operation"; (4) 10 CFR 50.61 " Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events"; and (5) Regulatory Guide 1.99, Revision 2 "Radiat;on Embrittlement of Reactor Vessel Materialc." The terms and methods used throughout this evaluation are discussed in detailin these sources.
3 in 1997, Comed submitted several requests to modify its methodology for determining the P/T limit curves for Byron 1 and 2 and Braidwood 1 and 2. One of the changes involved using a later version of the ASME Boiler and Pressure Vessel Code than that listed in 10 CFR 50.55a for determination of the P/T limit curves (i.e.,the licensee requested to use the methodology specified in the 1996 ASME Codt.,Section XI, Appendix G, Article G 2000 rather than the 1989 ASME Code). This change primarily altered the methodology used in determining the applied stress intensity due to thermal stresses. The other chanele involved integrating the weld metal surys.llance programs for Byron Units 1 and 2 and for Braldwood Units 1 and 2.
2.3 Low Temperatur.g Overoressure Protection System The LTOP system mitigates overpressure transients at low temperatures so that the integrity of the reactor coolant pressure boundary is not compromised by violating 10 CFR Part 50, Appendix G. The LTOP systems at Byron and Braidwood use combinations of pressurizer power operated relief valves (PORVs) and residual heat removal (RHR) suction relief valves to accomplish this function. The PORV portion of the system is manually enabled. When enabled, the system continuously monitors RCS temperature and pressure conditions. An auctioneered 2ystem temperature is continuously converted to an allowable pressure and then compared to the actual RCS pressure. The system logic first annunciates a main control board alarm whenever the measured pressure reaches a predetermined setpoint, thereby indicating a pressure transient is occurring. On a further increase in mecsured pressure, an actuation signal opens the PORVs in order to prevent pressure temperature conditions from exceeding allowable limits. The RHR suction relief valves have a constant setpoint and are available for !ow temperature overpressure protection whenever the ccrresponding RHR train is placed in service.
The design basis of the LTOP system cons lders both mass addition and heat addition translents I during water solid RCS conditions. The mass addition s:ncityses account for the injection from one charging pump. The heat addition analyses account for heat input from the secondary sides of all steam generators (SGs)into the RCS, upon starting a single reactor coolant pump (RCP).
The heat addition transient analyses assume the secondary side tempera'ures of the SGs are 50'F higher than the RCS temperature. The Byron and Braidwood proposed LTOP enable temperatures and actuation setpoints were established using the methodology presented in WCAP 14040-NP A,in combination with ASME Code Case N 514 and ASME Section XI, i
Appendix G,1996 Addenda.
3.0 EVALUATIONS 3.1 Neutron Fluence Evaluation The current capsule withdrawal schedule is as follows:
Byron Unit 1: Capsule W. Cycle B1R08, November, 1997 Byron Unit 2: Capsule X, Cycle B2R07, April, 1998 Braidwood Unit 1: Capsula W. Cycle '1R07, September, 1997 Braidwood Unit 2: Capsule W. Cyt F '.07, May, 1999 For the totalIron thickness of about 5.2 inches (2.5 inches for the core barrel and 2.7 inches for the neutron pads) the staff conservatively estimates that the underprediction of the vessel fluence due to the Iron cross sections in ENDF/B-IV is about 20%.
, U ~ - -
4 The projected operating time for Byron Unit 1 is 10.3 effective full power years (EFPYs) to the end of refueling cycle 9 at Byron Unit 1 (B1R09). Given that the currently approved license value is 12 EFPYs, operation would continue for 10.3/12 = 85.8% of the approved time. This is the longest operating time at all four units and is conservatively assumed to be app'; cable for all units. Byron 2 is projected to operate for 9.9 EIPYs until B2R08. It is currently approved for operation for 1'! EFPYs, so this modification to the P/T curves will apply for 9.9/12 = 82.5% of the approved time.
Braldwood Unit 1 is projected to operate for 9.56 EFPYs until A1R08. It is currently approved for 16 EFPYs, so this modification to the P/T c9tves will apply for 9.56/16 = 59.8% of the approved time. Braldwood Unit 2 is projected to operate for 10.18 EFPYs until A2R08. It is currently approved for 12 EFPYs, so this modification to the P/T curves will apply for 10.18/12 = 84.8%.
In addition, the licensee stated that, at the time of the current fluence calculation, the final values were increased by about 6% of the measured value from the capsule dosimetry. Therefore, this is a conservatism with respect to the calculated value. There is an estimated 5% conservatism due to the low leakage loadings practiced from cycle 2 to the present.
Given that the fluence is increasing linearly with respect to the number of EFPYs, the staff has determined that for Byron Unit 1 the nonconservatism amounts to about 20%. However, the conservatisms in the determination of tne fluence amount to about 14.2 + 5 + 6 = 25.2%. These l conservatisrns are, therefore, larger that the nonconservatism due to the use of the ENDF/B IV l cross section file. The staff, therefore, believes that the fluence values the licensee proposes to l use to generate the P/T curves are acceptable. The remaining units are even more conservative than Byron Unit 1, thus, they are also acceptable.
3.2 Prcssure Temperature Umits Evaluation Based on the information provided by the licensee, the methods used by the licensee in determining the vessel material data conform, in general, to the methodology approved by the NRC in WCAP-14040 NP A which endorses Regulatory Guide 1.99, Revision 2,
- Radiation Embrittlement of Reector Vessel Materials." Although the methodology used by the licensee is acceptable and consistent with that previously approved (WCAP 14040 NP A), there are some details of the methodology (not specifically addressed in WCAP i4040-NP-A) that require additional comment. These details include (1) the method used in determining the best estimate chemistry for welds, (2) the appropriate unirradiated reference temperature (RTm) value used in determining the adjusted referenco temperature (ART), and (3) the method for assessing integrated surveillance data. These areas are discussed below. In addition, the NRC's evaluation regarding the adequacy of the ARTS and the P/T limit curves for each unit is provided below, as are the details regarding implementation of the 1996 methodology that was addressed in reviewing the acceptability of the P/T limits.
3.2.1 Weld Best Estimate Chemistry The best estimate chemical composition (copper and nickel) for a heat of weld metal can be determined by soveral methods. Three of the more frequently used methods are the simple average, the mean-of the-means, and the coil weighted average method (copper only). In the simple aveoge method, all of the chemical composition data for that heat of material are averaged regardless of the source. (Sources of chemical composition data include weld metal qualificatioa tests, a plant's surveillance program, and nozzle dropout analyses.) in the mean-of-the-means approach, the mean value for each source of data for that particular heat of material is averrged to determine the best estimate chemical composition for the heat in the coil
\ .
S weighted average method, the mean value for each source of data is weighted by the number of coils of wire used in the fabrication of the weld. The wolghteu average ( sean) for each group is then average 1 to determine the best estimate chemical composition for the heat of material.
Selection of the appropriats method to use requires significant technicaljudgment. For example, the simple average method may be adversely influa9ed by numerous chemical analyses from one source of data. The mean-of the means approach, however, avoids this by placing equal welght on each source (which can also be a disadvantage since it gives a source of data with 1 chemistry sample the same weight as a source of data with 20 chemistry samples).
In addition, when a mean-of the means or coll-weighted average approach is used in determining the best estimate chemistry the way in which the chemistry data are grouped (in particular, those from weld qualification tests) can have a significant impact on the results. That is, the I resultant best estimate value can vary depending on whether the chemistry data are determined to be from "one weld" or from multiple welds, if weld qualification specimens were fabricated by the same manufadurer, within a short time span, usint similar welding input parameters, and using the same coll of ' veld consumables, the staff's r9 commendation is that all chemistry samples from that weld should be considered as "one weld" for the purposes of best estimate chemistry determ' nation. If information is not available to confirm the aforementioned details, but sufficient evidence exists to reasonably assume the details are the same, the best estimate chemistry should be evaluated both by assumin(, the data came from "one weld" and by assuming that the data came from an appropriate number of " multiple welds."
The licensee concluded that the mean of the means approach is the m<.,st appropriate method sloce it eliminates the inappropriate weighting effect which results from numerous analyses from i a particular weld block (i.e., source). The licensee further concluded that a coil weighted average approach is not a fundamentally sound basis for evaluating weld chemistry because of the lack of documentation of coil changes or intra coil splices that may have occurred or been present during production of the welds. i Given that the licensee concluded that the mean of the-means approach was the most :
appropriate method, the staff evaluated the method used by the licensee to group the weld qualification test data. The licensee chose to group the weld qualification data as coming from muitiple sources since information explicitly linking the data to other data was not available. The two heats of weld metal potentially affected by this issue are heats 442002 and 442011. With l this approach, for heat 442002, the licensee determined that the mean-of the-means value is 0.053% copper and 0.621% nickel. If the weld qualification test for a given woid is assumed to come from the same block, the mean-of the-means value would be 0.059% copper and 0.628%
nickel. The staff used 0.059% copper and 0.628% nickel as the best estimate chemical composition for heat 442002 and determined that the surveillance data were credible. Using the ratio procedure and the surveillance data, the chemistry factor was determined by the staff to be less (which is less conservative) than the value used by the licensee. For heat 442011, the licensee determined that the mean-of the means value is 0.032% copper and 0.666% nickel. If the weld qualification iest for a given weld is assumed to come from the same block, the mean-if the means value would be 0.033% copper and 0.667% nickel. The staff concluded that there
. vere no appreciable differences in the ART using e!ther of these chemical compositions for heat 442002.
Only heats 442002 and 442011 were impacted by the method (simple average, mean of-the-means, coil weighted average) chosen to determine the best estimate chemical composition.
For heat 44N02, the mean-of the-means approach, which the licensee used, provides ths most
6 conservrtive estimate of the chemical composition for the heat. (Copper and nickel are 0.029%
and 0.680%, respectively, for the simple average method. Copper and nickel are 0.059% and 0.628%, respectively, for the meart of the-means approach). On the other hand, for heat 442011, the simple average method yields a slightly more consersative chemical composition.
(Copper and nickel are 0.033% and 0.688%, respectively, for the simple average method.
Copper and nickel are 0.032% and 0.667%, respectively, for the mean-of the-means approach).
Given the differences in the estimates by the three methods and the number of samples for each source of data, the staff concludes that the values used by the licensee are acceptable.
However, as additional chemical composition data become available, the licensee should re-evaluate the appropriate method for determining the best estimate chemical composition and the !
appropriate method for grouping the data if the mean-of the means or coilweiqhted average approach are used.
3.2.3 Unirradiated Reference Temperature The unirradiated reference temperature, RTwo3g,is used in the determination of the ART. As discussed during a public meeting on November 12,1997 (see rueeting summary dated November 19,1997, ' Meeting Summary for November 12,1997 Meeting with Owners Group Repre;,entatives and NEl Regarding Review of Responses to Generic Letter 92 01, Revision 4, Supplement 1 Responses"), the staff recognized that in Jome instances there are significant differences in the RTuorg values used by licensees for the same heat of material. The NRC also indicated that this was a potentiallong term issue.
For Byron and Braldwood Units 1 and 2, the RTuotu values used in the evaluation differ for several heats of material as shown in Table 1. The licensee believes that the differing RTwovy values may be explained on the basis of the different flux lots used in the fabrication of the weld.
TABLE 1: DIFFERENCES IN RTuorg FOR MATERIALS IN BYRON 1 AND 2 RT ow Heat of Material Byron i Byron 2 Braiowood i Braidwood 2 31401 10 0 40.0 40 0 N/A 442002 30.0 10.0 N/A N/A 442011 10.0 40.0 40 0 40.0 As shown in Table 1, heat 31401 has it a greatest variability. The staff evaluated this variability and concluded that even if the most limiting RTuotg value (i.e.,40.0 'F) was useo in the Pyron 1 and Braidwood 1 evaluation of this heat cf material, the limiting material woulci not change and the RTp3, screening criteria would not be exceeded. This calculation was performed using the end-of license fluence reported in the NRC Reactor VesselIntegrity Database. Similarly, for heat 442011, the limiting material would not change and the RTpts screening criteria would not be exceeded for Byron 1 if an RTuota value of 40.0'F was used, in this evaluation, the staff assumed that the surveillance data from Braidwood 1 and 2 heat 442011 could be used in the evaluation of the Byron 1 vessel based on the similarity of the irradiation environments.
7 For heat 442002, if the limiting RTau value (i.e.,10*F)is used in the Byron 1 evaluation of this heat, this would become the limiting material by approximately 10'F but the RTp3 screening criteria would not be exceeded. For Byron 1, the weld in question is identified as WF-336 and was fabricated from weld wire heat 442002 and flux lot 8873. The similar weld at Byron 2 is identified as WF-447 and was fabricated from weld wire heat 442002 and flux lot 8064. The licensee believes that the differing initial RTag values may be explained on the basis of the different flux lots used. The staff finds this acceptable; nonetheless, the staff still considers this an issue that may lead to nJte cha nges as noted in the November 12,1997, meeting, in future revisions to the PTLR, the licenst e should assess the impact of its assumption that the vessel weld has the same RTag value as determined from the surveillance weld.
3.2.4 Credibility Evaluation for Intearated sprveillance Proaram To assess the credibility of surveillance data from welds when the data are from more than one source, the data should be ,1ormalized to account for differences in chemical composition and irradiation environment consistent with NRC regulations. The NRC previded guidance on performing these adjustments in the November 12,1997, public meeting as discussed above, in assessing the credibility of the data, the licensee did not adjust for chemical composition differences (i.e., apply the ratio procedure). The licensee did consider differences in the 7rradiation temperature between the specimens at Byron 1 and Byron 2 and concluded that the differences were small.
l The staff, on the other hand, assessed the credibility of the integrated wald metal surveillanca I program by normalizing the surveillance data to the averaDe chemical comrmsition and irradiation environment of the surveillance specimens as discussed below. The irrad i an environment (i.e., temperature) to which the surveillance weld metal at Byron 1 and 2 were exposed were considered identical to those to which the vessels were exposed (consistent with the licensee's conclusion that the effect is small); therefore, no adjustments to the surveillance data to account for temperature were made in assessing the credibility of the data. To account for chemical composition differences, the staff used the ratio procedure and normalized the data to the I
average chemical composition of the surveillance welds (i.e., copper = 0.0225; nickel = 0.701).
Even though the details between the staff and the licensee's evaluation differed, the net result was the same. That is, the data were determined to be credible. However,in future revisions to the PTLR and/or when additional surveillance data become available, the licensee should address the method for assessing the credibility of the data including the method for accounting for irradiation environment and chemical composibon differences. This is consistent with the licensee's statement that " temperature differences and chemistry factor ratios will be re-evaluated at all future scheduled capsule evaluations."
3.2.5 Byron Unit 1 Based on the material provided by the licensee, the staff confirmed that (1) the limiting material is intcrmediate shell forging SP 5933 with a 1/4T ART of 70'F and a 3/4T ART of 60'F and (2) the P/T limit curves are approprLite.
3.2.6 Byron Unit 2 Based on the insterial provided by the licensee, the staff confirmed that (1) the limiting material is circumferential weld, WF-447 (heat 442002) and (2) the P/T !imit curves are appropriate. The ARTS calculated by the staff (1/4T ART of 81.9'F and 3/4T ART of 67.6'F) were slightly lower
I 8
(i.e., less conservative) than that reported by the licensee (1/4T ART of 87.6'F and 3/4T ART of 71.5'F) as a result of minor differences in chemical composition and in applying the ratio procedure.
I 3.2.7 Draidwood Unit 1 Based on the material provided by the licensee, the staff confirmed that (1) the limiting materialis circumferential weld, WF 562 (heat 442011) and (2) the P/T curves are appropriate. The ARTS calculated by the staff (1/4T ART of 69.7'F and 3/4T ART of 60.6'F) were lower (i.e.,less conservative) than that reported by the licensee (1/4T ART of 76.6'F and 3/4T ART of 65.4'F) as a result of using a slightly different chemical composition for the weld. Furthermore, the P/T limit curves are more conservative than would be required using the methodology in WCAP.
14040 NP A. Since the licensee's results are more conservative, the staff concludes that they are appropriate.
3.2.8 Braidwood Unit 2 Based on the material provided by the licensee, the Staff confirmed that (1) the limiting material is circumferentialweld, WF 562 (heat 442011)with a 1/4T ART of 66.8'F and a 3/ T ART of 58.1'F and (2) fhe P/T curves are appropriate.
3.' Low Pressure Overoressure Protection System Evaluation 3.3.1 LTOP Enable Temperature i
The LTOP enable temperature is the temperature below which the LTOP system is required to be operable. The licensee's proposed enable temperature (1) accounts for instrumentation uncertainties associated with the instrumentation used to enable the LTOP system arid (2) Implement the ASME Code Case /1996 Addenda methodology of using an enable RCS Ivd temperature corresponding to the reactor vessel %-T metal temperature of RTuoi + 50 or 200'F, whichever is greater. Therefore, the minimuni allowed enable temperature was calculated as the larger of either RTwo3 + 50'F + E,n, + delta Tw, or 200'F. In this calculation, E,n, refers to instrument error while delta Tw, refers to the temperature difference between the reactor coolant and the metal at a distance one fourth of the vessel wall thickness from the inside surface in the beltline region.
The use of 50'F in the above methodology is consistent with ASME Code Case N 514 and the 1996 Addenda. Accounting for the delta Ty,.3 is consistent with Branch Technical Position (BTP)
RSB 5 2, which states that the enable temperature is defined La "the water temperature corresponding to the metal tempera'ure...at the beltline location (1/4T or 3/4T) that is controlling in the Appendix G limit calcult';on." This approach is also consistent with the ASME Code Case and the 1996 Addenda. Accounting for instrument uncertainty ensures that the LTOP system is not enabled at temperatures less conservative then are required by the aforementioned documents, Based on the above discussion the staff finds acceptable the licensee's implementation of the LTOP enable temperature methodology.
The licensee proposed an LTOP enable temperature of a 350'F for all four units. Based on vessel material data, instrumentation uncertainties, and delta Ty,.7, the minimum allowed ensble temperature for each of the units is as follows:
I
. t 9
TABLE 2: ENABLE TEMPERATURES Minimum A# owed Enable Calculated Temperature Fnable (Greater of Plant RT m E.,,, Delta-T., Temperatute Ca'eulated or 200*F)
Draidwood1 76 6*F 15.0*F 29.254*F 170 9'F 200'F Braidwood 2 66 9'F _, 15 0'F 29.254*F 161.2*F 200*F Byron 1 700'F 150*F 20 254'F 164 3*F 200'F Byron 2 87.6*F 15.0*F 29.254*F 181.9'F 200*F For the above listed RTwot values, the proposed LTOP enable temperature of a 350'F is conservative with respect to the minimum LTOP enable temperature allowed by WCAP-14040 NP A, ASME Code Case N 514 and the ASME Section XI, Appendix G,1996 Addenda .
Therefore, the staff finds the licensee's proposed enable temperature acceptable.
3.3.2 LTOP Actuation Setpoint The ASME Code Case and 1996 Addenda both allow LTOP systems to be designed to limit the peak pressure at the controlling location in the reactor to 110% of the P/T limits. Additionally, since overpressure events most likely occur during isothermal conditions in the RCS, the NRC has approved the use of the steady stata P/T limits for the design of LTOP. WCAP 14040-NP-A provides a methodology for calculating setpoints for LTOP systems that use pressurizer PORVs with variable setpoints. Section 3.2, "COMS Setpoint Determination," of WCAP- 14040-NP A provides discussions of parameters that need to be considered in deteimining the LTOP actuation setpoint. A summary of the licensee's LTOP analyses were submitted in the January 8 and 13,1998, letters. The licensee's analyses and proposed PORV lift setroints were based on WCAP 14040-NP-A and the P/T limits with 10% relaxation in accordance with ASME Code Case N-514 and the 1996 Addenda. The resulting PORVlift setpoints were provided as Figure 2.3 in each units PTLR.
Westinghouse performed mass addition and heat addition LTOP analyses for Byron and !
Braidwood using the LOFTRAN computer code. Use of the LOFTRAN computer code is I consistent with WCAP 14040-NP A. For the mass addition transients, the RCS was assumed to I be water solid. The analyses accounted for injection from a single charging pump and calculated !
the amount of overshoot that would occur during such a transient. To ensure consistency bstween the analysis assumptions and th9 TS, Surveillance Requirement (SR) 4.5.3.2 ensures that all other charging and safety injection (SI) pumps ars made incapable of injecting into the ,
HCS while in the LTOP region. This SR, however, uses a temperature of 330*F to make the i pumps inoperable instead of the licensee's proposed LTOP enable temperature of 350'F. Using l 330*F is acceptable because the minimum required enable temperature (Section 3.1 of this !
safety evaluation (SE))is 200*F. A configuration allowing SI pumps to be available when the l pressurizer levelis less than or equal to 5% is evaluated in Section 3.3.4 of this SE.
. t 10 Heat addition transients were also run with the RCS in a water solid condition. For these analyses, the secondary splem was assumed to be 50'F higher than the RCS. One RCP was assumed to start and consequent heat addition from all SGs was accounted for. Expansion of the reactor coolant resulted in pressurization of the RCS and actuation of the LTOP system. The resulting overshoot values were determined by these analyses. To ensure consistency between the analyses assumptions and the TS, a note is included in TS 3.4.1.3 to ensure that RCrs are not started unless the secondary side of any SG is less than 50*F nigher than the RCS.
For all cases analyzed, the licensee conservatively assumed one PORV failed and, therefore, credited only one PORV for pressure relief. Additionally, the licensee did not credit the RHR suction relief valves for mitigating the pressure transient in the analyses. An evaluation presented in Section 3.3.3 of this SE addresses the substitution of RHR suction relicivalves for PORV as currently allowed by TS 3.4.9.3. The licensee evaluated the results of the heat addition i and mass addition cases as a function of temperature and used the more conservative value of overshoot for LTOP setpoint calculations. The licensee accounted for static and dynamic l nad i effects as well as instrumentation uncertainties in the final determination of the LTOP setpoints.
The dynamic head effect wds divided into two regions. For RCS temperatures above 120'F, the licensee used a pressure drop corresponding to all four RCPs and both RHR pumps running.
For RCS temperatures less than or equal to 120'F, the licensee used a pressure drop corresponding to one RCP and both RHR pumps running. Plant administrative procedures prohibit running more than one RCP below 120'F.
The above analyses were performed using the LC'RAN computer code assuming the original i
steam generators. For Byron Unit 1 and Braidwood Unit 1, Framatome and Comet Nuclear Fur -
Services performed analyses that verified the Westinghouse calculated setpoints remain valid for the replacement steam generators. These enalyses were performed using RELAPS/ MOD 2 B&W l computer code, which is documented in Topical Repcrt BAW-10164P-A. Th;s code was l
i approved by the staff for both loss-of-coolant accidents (LOCA) and non-LOCA applications and is therefore acceptable for use in LTOP analyses.
Based on the above discussion, the staff finds the licensee's implementation of the WCAP-14040-NP A methodology and the proposed LTOP actuation setpoints as presented in Figures 2.3 of each unit's PTLR acceptable. In addition, the staff finds acceptable the licensee's request to perform LTOP analyses using the NRC approved RELAPS/ MOD 2 B&W computer code.
3.3.3 RHR Suction Relief Valves TS 3.4.9.3 requires at least two overpressure protection devices, each consisting of either an RHR suction relief valve or a PORV. This requirement would allow for RHR suction relief valves to be used in place of PORVs for LTOP. WCAP-14040-NP A does not address the use of the RHR suction relief valves. The staff's SE for the WCAP, dated October 16,1995, states that licensees who use the WCAP should address this issue in their PTLR submittal. The licensee addressed the use of the RHR suction relief valves in its letter dated January 8,1998. The licensee's evaluation justified the use of the RHR suction relief valves at its current lift setpoint of 450 psig and appropriately accounted for a setpoint drift of 3% and an accumulation of 10% as recommended by Article NC-7000 of the ASME Boiler and Pressure Vessel Code. The staff reviewed the licensee's evaluation and found it acceptable.
The licensee also included a requirement in the PTLR to evaluate 'he RHR suction relief valves in a similar manner whenever the P/T limits are revised. The staff finds this consistent with the recomrnendation in the SE for WCAP-14040 NP-A and, therefore, acceptable.
. t 11 3.3.4 jyallLbility of 81 Pumns with Pressurizer Level s 5%
To mitigate a loss of decay heat removal event, TS 3.5.4.2 requires at least one 81 pump and flow path to be available in either Mode 5 or Mode 6 with the pressurizer level less than or equal to 5%. The licensee evaluated this from an LTOP perspective. The licensee's evaluation ,
concluded that should an 81 pump be inadvertently started, operator.a would have longer than 10 minutes (licensee calculations show 18.4 minutes) before the pressurizer would become water solid. This would provide suffielent time to credit manual operator action to mitigate this event beforg overpressurization would occur.
~ The licensee further stated that typical operating practices at Byron and Braidwood ensure that only one Si pump is available. However, for circumstances where both Si pumps are available, administrative controls ensure that operators would have to take at least three independent manual actions to start more than one pump. These manual actions include moving the har d switch for each 81 pump from the Full to Lock position to the Run position and opening the SI ;
pump coid leg discharge isoladon valves. Therefore from an overpressure protection standooint, '
the potential for an inadvertent start of both GI pumps is not considered a credible event.
The staff finds the licensee's evaluation acceptable based on (1) the time available for operators to take manual action to mitigate the event of a single SI pump start, and (2) the administrative ,
controls for ensuring that the inadvertent start of both Si pumps event is not credible. ;
3.3.5 Bolt up Temoetsture The licensee did not include Instrumentation uncertainties in the bolt up temperature. However,- '
the licensee proposed to account for instrumentation uncertainties in the minimum pressurizailon ,
temperature (Section 2.5 of each unit's PTLR) to ensure that the RCS is not capable of being : ,
pressurized (i.e., the RCS remains vented) until the RCS temperature is greater than or equal to the minimum allowable bolt up temperature plus instrument uncertainty as detem,ined using a -
technique consistent with ISA S67.04 1994.' This was determined to be acceptable by the staff.
4.0 f,QNfcLUSIONS
- Based upon the staff evaluations, as discussed in Section 3.0 above, the i4RC concludes that it I is acceptable for the licensee to relocate the P/T limit curves and LTOP system limits from the )
Byron, Units 1 and 2, and Braidwood, Units 1 and 2. TS to a licensee controlled PTLR. The staff -
also concludes that it is acceptable to relocate the surveillance capsule withdrawal schedule to the licensee controlled PTLR since changes to this schedule are controlled by the requirements of Appendix H to 10 CFR Part 50. 1
- The staff has reviewed the proposed fluence values for the four Byron and Braidwood units for the revision of the P/T curves and finds that the existing approved values are conservative and, i therefore, acceptable. In addition, the P/T limits meet the requirements of 10 CFR Par 150, Appendix G and none of the RTm values exceed the screening criteria specified in 10 CFR
- 50.61. .
The 14RC notes, however, that the licensee should (1) re-esaluate the appropriate method for determining the best estimate chemical composition as additional chemical composition data become available, (2) assess the impact of its assumption that the vessel weld has the same 1
4
)
12 RT m value as determined from the surveillance weld in future revisions to the PTLR and/or when additional surveillance data become available, and (3) address the method for assessing the credibility of the data including the method for accounting for irradiation environment and chemical composition differences in future revisions to the PTLR.
The staff has also reviewed the licensee's implementation of WCAP-14040 with regard to LTOP ,
and the licensee's proposed LTOP system enable temperature and actuation setpoints. The i staff finds the licensee's proposal consistent witn the staff's SE approving the WCAP and also consistent with BTP RSB 5 2, ASME Code Case N 514 and ASME Section XI, Appsndix G,1996 Addenda. Based on the above discussion and the eva!ustion provided in Section 3 of this SE, the staff finds the licensas's proposed LTOP enable temperature and actuation setpoints acceptable. The staff further finds acceptable the licensee's request to use the RELAPS/ MOD 2-B&W computer code for LTOP arfyses.
5.0 REFERENCES
- 1. Letter from R. R. Assa, NRC, to D. L. Farrar, Commonwealth Edison Company,
" Exemption from Requirements of 10 CFR 50.60 Braidwood Station, Unit 1," July 13, ,
199S.
- 2. Letter from C. l. Grimes, NRC, to R. A. Newton, Westinghouse Electric Corporation,
" Acceptance for Referencing of Topical Report WCAP 14040, Revision 1, ' Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves,'" October 16,1995. (Also known as WCAP-14040-NP A).
- 3. Letter from G. F. Dick, NRC, to I. M. Johnson, Commonwealth Edison Company,
" Exemption from Requirements of 10 CFR 50.60, ' Acceptance Criteria for Fracture Prevention Measures for Lightwater Nuclear Power Reactors for Normal Operation'-
Byron Station, Units 1 and 2," November 29,1996.
- 4. Letter from J. B. Hosmer, Commonwealth Edison Company, to NRC Document Control Desk. " Reactor VesselIntegrated Surveillance Program 10 CFR 50, Appendix H, Section Ill.C,* May 6,1997. (WCAP 14824, Revision 1 is Attached).
- 5. Letter from J. B. Hosmer, Commonwealth Edison Company, to NRC Document Control Desk, " Application for Amendment to Appendix A, Technical Specifications, for Facility Operating Licenses Relocation of Pressure and Temperature Limas,' May 21,1997.
- 6. Letter from J. B. Hosmer, Commonwealth Edison Company, to NRC Document Control Dt,sk,
- Supplement to the Application for Amendment to Appendix A, Technical Specifications, for Facility Operating Licenses Relocation of Pressure and Temperature Limits," November 18,1997. (WCAP 14940 and Errata to WCAP 14940 and WCAP-14970 are Attached).
- 7. Letter from J. B. Hosmer, Commonwealth Edisor' Company, to NRC Document Control Desk, *Supplementalinformation Pertaining to Byron & Braidwood's Reactor Vessel integrated Surveillance Program," December 3,1997. (WCAP 14824 Revision 2, and Erratum to WCAP 14824, Revision 2 are Attached).
y _
13
" 8. Letter from G. F. Dick. NRC to I. M. Johnson, Comrnonwealth Edison Company, "Exemptica from Requirements of 10 CFR 50.60 - Braidwood Nuclear Station, Unit 2,*
December 12,1907.
E 9
Letter from H. G. Stanley, Commonwealth Edison Company, to NitC Decument Control Desk, "SupplementalInformation Pertaining to Technical Specification Amendment Rug;rding Pressure Temperature Curves Gyron and Braidwood Nuclear Power gr Sthiions," January 8,1998. (Errata to WCAP-14824 Revision 2, WCAP 14940 and WCAP 14970 are Attached).
- 10. Letter from H. G. Stanley, Commonwealth Edison Con )cny,io NRC Document Control
[
Desk," Supplemental!nformation Pertaining to Techi,ical Specification Amendment rtegarding Pressure Temperature Curves - Byron and Braidwood Nuclear Power Stations," Januai/13,1998,
- 11. Letter from G. F, Dick, NRC, to O. D. Kingsley, Commonwealth Edison Company,
" Exemption from Requirenants of 10 CFR 50.60- Byron, Ur..e 1 and 2, and Braidwood, Units 1 and 2,* January 16,1998,
- 12. Letter from R. A. Capra, NRC, to O. D. Kingsley, Commonwealth Edison Company, Mntegration of Reactor Pressure Vessel Surveillance Program '.:r Byron and Braidwood, Units 1 and 2," Jnneary 16,1998.
Principal Contributors: K. Karwoski L. Lois M. Ghuaibi M. W. Weston n
Date: January 20,1998 r
a- >.
w g
~
Q % r ~~