ML20199J037

Safety Evaluation Supporting Amends 98,98,89 & 89 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively

ML20199J037
Person / Time
Site:	Byron, Braidwood
Issue date:	01/23/1998
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20199J034	List: ML20199J028 ML20199J037 ML20199K468
References
NUDOCS 9802050202
Download: ML20199J037 (13)

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UNffSD STATES pj NUCLEAR RECULATCRY COMMISSI N wasameron, o.o.samen

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RAFETY EVALUATION SY THE OFFICE OF NUN 8AR REACTOR REGULATION 4

REVIEW OF THE PRESSURE TEMPE RATURE LIMITS REPORT (PYL t AND WlETHODOLOGY FOR THE RELOCATION OF "HE REAC' TOR COOLANT SYS'TMfRC1 PRESSURE TEMPERATURE LIMIT CURVES AND LOW "EMPERATURE OVEltPRESSU tE PROTECTION (LTOP) SYSTEM LIMITS Ef7ED TO FACILITY OPERATING LICENSES NPF.37. NPF et. NPF-72 AND NPF.77 QOMMONWEALTH EDISON COMPANY SYRON STATION. UNITS 1 AND 2. AND BRAIDWOOD STATION. UNITS 1 AND 2 DOCKET NUMSERS 50 454. 50 4$5. 50 456. AND 50 457 i

1.0 lb(IBpDUCTION 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 Byron Units 1 and 2, and 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 onntrolled document identified as a Pressure Temperature Limits Report (P(LR), and (3) changing the affected limiting conditions for operation and bases accordingly. These changes a,4 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 schedule from the TS to the PTLR. Relocation of the reactor vessel surveillance capsule withdrawal schedule is mode in accordance with OL 9101, " Removal of the Schedule for the Withdrawal of Reactor Vessel Material Specimens from Technical Specifications," except that the 1

schedule is relocated to the PTLR rather than the UFSAR. Changes to this schedule continue to be coYrolled by the requirements of Appendix H to 10 CFR Part 60.

The licensee's request to relocate the P/T limit curves and LTOP system limits to the PTLR was submitted consistent with the guidance provided in OL 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 setpoints, (2) the neutron transport cross-section library and dosimeter reaction cross sections used in determining the fluences, and (3) the version of th* 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).

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The Commission granted Byron and Braidwood exemptions (References 1,3, and 8) from tile requirements of 10 CFR 50.60, 'Accep%ce Criteria for Fracture Prevention Measures for Ughtwater Nuclear Power Reacton. P tmal Operation,' that allowed the plants to use the methodologyin ASME Code Cese '

. Byron and Braidwood were also granted exemptions (Reference 11), permitting the use 6.

w methodology described in the 1996 A8ME Code, j

Section XI, Appendix G, Article G 200s 'n addition, the staff appros

• Dforence 12), the integration of the Byron 1 and 2 and Sr_Jwood 1 and 2 weld metal surveillance programs.

With regard to ti,e LTOP system limits, the Code Case and the 1996 Addenda provide essentially 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 a coolant temperature conosponding to a reactor vessel metal temperature less then RT,m + 50*F. whichever is greater.

2.0 BACKGROUND

t 2.1 Neutron Fluence The fluence evaluation which is the basis for the current P/T curves was performed when the first four surveillance capsules were removed and evaluated at the end of the Arst cycle for each plant. However, at the time these evaluations were performed, ENDF/5 IV based cross settions were used. Since that time th6 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 i

ENDF/5 VI cross section file.

The licensee proposed to move the P/T limit curves to the PTLR before an integrated 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 i

manner in which the materials are utilized, and, therefore will change the PTLR methodology.

Because of the changes, the methodology will then have to be submitted to the NRC for prior i

review and approval. The licenses stated that the maximum operating time used to generate the P/T limit curves will be at most 85.8% of the cunent licensing basis time. - Therefore, the revised operating time used to generate the corvos will be no greater than 85.8% of the currently approved maximum value. This evaluation establishes a conservative fluence estimation. The nonconservatism of the ENDF/B IV file is about 20%. The conservatisms are 14.2%, due to the shorter cperating time; 6%, due to the initial adjustments to the measured data; and 5%, due to the effect of the low leakage loadings which have been practiced in all units since the second cycae.

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2.2 Pressure Temperature Limits The methodologies for assessing P/T limits and reactor pressure vessel (RPV) surveillance programs are discussed, 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 Matertal Surveillance Program 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 Pressurtzed Thermal Shock Events"; and (5) Regulatory Guide 1.99, Revision 2 " Radiation Embrittlement of Reactor Vessel Materials " The terms and methods used throughout this evaluation are discussed in detail in these sources.

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3 in 19g7. Comed submitted several requests,o modify its methodology for determining the P/T llmit curves for Byron 1 and 2 and Braidwood 1 and 2. One of the changes involved using a later version of the ASME Boller and Pressue:, Vos el Code than that listed in 10 CFR 50.55a for determination of the P/T limit curver g.e.,the f ansee requested to use the methodology specifit d in the 1996 ASME Code, Se. tion XI ApperW 1x 0, Article G-2000 rather than the 198g -

8 ASME Code). This change primarily 7.terod '.w methodology used in determining the applied stress intensity due to thermal stresses, i.,e other change involved integrating the weld metal -

surveillance programs for Byron Units 1 and 2 and for Braidwood Units 1 and 2.

2.3 Low Temnerature Overnra=_=e PraaW evatem 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 0. The LTOP systems at Byron and Braidwood use combinations of pressurtser power l

operated relief valves (PORVs) and residual heat removal (RHR) suction relief valves to accumplish this function. The PORV portion of the system is manually enabled. When enabled, the system continuously monitors RCS temperature and pressure conditions. An auctioneered system 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 measured 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 low temperature overpressure protection whenever the corresponding RHR train is placed in service.

The design basis of the LTOP system considers both mass. addition and heat addition transients during water solid RCS conditions. The mass addition analyses 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 temperatures 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, Appendix G,1996 Addenda.

3.0 EVALUATIONS 3.1 Neutron Fluence Evaluation The current capsule withdrawat schedule is as follows:

Byrca Unit 1: Capsule W, Cycle BIR08, November, 1997 Byron Unit 2: Capsule X, cycle B2R07, April, 1998 Braidwood Unit 1: Capsule W, Cyda A1R07, September,1997 Broldwood Unit 2: Capsule W, Cycle A2R07, May,

_199g

_ For the total tron _ 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 flusnce due to the Iron cross sections in ENDF/B IV is about 201

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 (81Rog). Given that the currenty approved license value is 12 EFPYs, operation would continue for 10.3/12 = 85.8% of the approved time. This is the at operating time at all four unha and is conservatively assumed to be applicable for s'l units.

2 is projected to operate for 9.9 EFPYs until B2R08. It is currently 5pd for operation for 12 EFPYs, so this modification to the PR ourves will apply for 9.9/12 = 32.6% of the approved time.

Brandwood Unit 1 is projected to operate for 9.88 EFPYs until A1R00, it is currenty approved for 18 EFPYs, so this modification to the PN ourves will appy for 9.86/16 = Sg.0% of the approved time.- Braldwood UnN 2 is projected to operate for 10.18 EFPYs until A2R08. H is currenty approved for 12 EFPYs, so this modification to the P6 ourves wB apply for 10.18/12 = 84.8%.

In addnion, the licensee stated that, at the time of the curtsnt 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 linearty 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 the fluence amount to about 14.2 + 5 + 6 = 25.2%. These conservatisms are, therefore, larger that the nonconservatism due to the use of the ENDF/B IV cross section file. The staff, therefore, believes that the fluence values the licensee proposes to use to generate the PR curves are acceptable. The remaining units are even more conservative than Byron Unit 1, thus, they are also acceptable.

3.2 Pressure Temperature Limits 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 Reactor Vessel Materials." Although the methodology used by the licensee is l

acceptable and consistent with that proylously approved (WCAP 14040 NP-A), there are some i

details of the methodology (not specificaly addressed in WCAP 14040 NP A) that require additional comment. These details include (1) the method used in determining the best-estimate chemistry for welds, (2) the appropriate unitradiated reference temperature (RTau) value used in determining the adjusted reference 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 PR limit curves for each unk is provided below, as are the details regarding implementation of the 1996 methodology that was addressed in reviewing the acceptability of the Pn limits.

3.2.1 Wald Best Estimate Ch> Jatty The best-estimate chemical composition (copper and nickel) for a heat of wold metal can be determined by several methods. Three of the more frequently used methods are the simple average, the mean-of-the means, and the coil weighted average method (copper ony), in the simple average 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 qualification tests, a plant's surveillance program, and noule dropout analyses.) in the mean-of-the-means approach, the mean value for each source of data for that particular heat of material is averaged to determine the best estimate chemical composition for the heat, in the coil i

5 weigided everage method, the mean value for each source of data is weighted by the number of coils of wire used in the fabrication of the wold. The weighted average ~(mean) for each group is then averaged to determine the best estimate W :nical composition for the heat of material.

Selection of the appropriate method to use requires signl6 card technicaljudgment. For example, the simple average method may be adversely influenced by numerous chemical analyses from one sou se of data. The mean-of the means approach, however, avoids this by placing equal weight 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 collweighted average approach is used in determining the best estimate chemistry, the way in which the chemistry data are grouped (in particular, i

those from weld qualification tests) can have a signl6 card impact on the results. That is, the resultant best estimate value can vary depending on whether the chemistry data are determined to be from *one weld" or from multiple wolds. If weld qualification specimens were fabricated by the same manufacturer, within a short time span, using similar welding input parameters, and using the same coil of weld consumables, the staff's recommtadation is that all chemistry samples from that wwld should be considered as *one weld' for the purposes cf best-estimate chemistry determination, 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 assuming the data came from *one weld" ar.d 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 most appropriate method since it eliminates the inappropriate weighting effect which results*from numerous analyses from 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 poduction of the welds.

Given that the licensee concluded that the mean-of the means approach was the most appropriate method, the staff evaluated the method ut.ed by the licensee to group the weld qualification test data. The licensee chose to group the weld qualification data as coming from multiple 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 hosts M2002 and M2011, With this approach, for heat M2002, the licensee determined that the mean-of the-means value is 0.053% copper and 0.621% nickel N the weld qualification test for a given weld is. assumed to come from the same block, the mean-of the means value would be 0.05g% copper and 0.628%

nickel. The staff used 0.05g% copper and 0.628% nickel as the best estimate chemical composition for heat M2002 and determined that the survelliance data were credible. Using the ratio procedure and the surveillance dats, the chemistry factor was determined by the staff to be less (which is less conservative) than the value used by the licensee. For heat M2011, the licensee determined that the mean-of-the means value is 0.032% copper and 0.666% nickel, if the weld quellfication test for a given weld is assumed to come from the same block, the mean-

- of the-means value would be 0.033% copper and 0.667% nickel. The staff cor:cluded that there were no appreciable differences in the ART using either of these chemical composhions for heat M2002.

Only heats M2002 and M2011 were impacted by the method (simple average, mean-of-the-means, coil weighted everage) chosen to determine the best-estimate chemical composition.

For heat M2002, the mean of-the-means approach, which the licensee used, provides the most

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8 conservative 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 0.0Sg% and a

0.82g%, respedively, for the mean-of the means approach). On the other hand, for heat 442011, the simple average method yloids a slightly more conservative 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 condudes 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 kr determining the best-estimate chemical ot,mposition and the appropriate method for grouping the dets if the mean-of-the means or ooil weighted everage approach are used.

3.2.3 Unirradiated Reference Temnerature The unitradiated reference temperature, RTam, is used in tne determination of the ART. As discussed during a public meeting on November 12,19g7 (sea meeting summary dated November ig, igg 7, " Meeting Summary for November 12,19g7 Vesting with Owners Group Representatives and NEl Regarding Review of Responsas to Generic Letter g2 01, Revision 1, Supplement 1 Responses"), the staff recognized that in some instances there are signifmant differences in the RTag values used by licensees for the same heat of material. The NRC also indmated that this was a potential long term issue.

For Byron and Braidwood Units 1 and 2, the RT values used in the evaluation differ for several heats of material as shown in Table 1. Yh$ licensee believes that the differing RTau values may be explained on the basis of the different flux lots used in the fabrication of the weld.

TABLE 1: DIFFERENCES IN RTem FOR MATERIALS IN BYRON 1 AND 2 RT,.,,ui Heat of Meteriel Bron i tron 2 tredwood 1 Bradwood 2 31401 10.0 40.0 40.0 N/A

1. 42002 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 the greatest variability. The staff evaluated this variability and concluded that even if the most limiting RTag value (i.e.,40.0 'F) was used in the Byron 1 and Braldwood i evaluation of this heat of material, the limiting material would not change and the RTn, screening celteria would not be exceeded. This calculation was performed using the end of license fluence reported in the NRC Reactor Vessel integrity Database. Simi! arty, for heat 442011, the limiting material would not change and the RTm screening criteria would not be exceeded for Byron 1 if an RTau value of 40.0'F was used. in this evaluation, the staff assumed that the surveillance data from Braldwood 1 and 2 heat 442011 could be used in the evaluation of the Byron 1 vessel based on the similarity of the irradiation environments.

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For heat 442002,if the limiting RTau value (i.e.,10'F)is used in the Byron 1 evaluation of tnis heat, this would become the limiting material by approximately 10'F but the RTm sorooning l

criteria would not be exceeded. For Byron 1, the weld in question is identified as WF 336 and was fabricated frorn weld wire heat 442002 and flux lot 8873. The similar weld at Byron 2 is Identified as WF-447 and was fabricated from wold wire heat 442002 and flux lot 8064. The iloonsee believes that thw differing initial RT. values may be explained on the basis of the i

different flux lots used. The staff finds this acc,eptable; nonetheless, the staff still considers this an issue that may lead to rule chances as noted in the November 12,1997, tweting. In future fevisions to the PTLR, the licensee 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 Credibilltv Evaluation for intearated Sure:M r.se Ta,s,sm To assess the credibility of surveillance data from welds when the data are from more than one source, the data should be normalized to account for diNorences in chemical composition and irradiation environment consistent with NRC regulations. The NRC provided guidance on performing these adjustments in the November 12,1997, public meedng as discussed above, in t

i assessing the orodibility of the data, the licensee did not adjust for cf.dmical composition j

differences (i.e., apply the retic procedure). The licensee did consider differences in the irradiation temperature between the specimens at Byron 1 and Byron 2 and concluded that the differences were small.

3 The staff, on the other hand, assessed the credibility of the integrated weld metal surveillance program by normalizing the surveillance data to the average chemical composition and irradiation j

environment of the surveillance specimens as discussed below. The irradiation environment j

(i.e., temperature) to whlch the surveillance weld metal at Byron 1 and 2 were exposed were i

considered identical to those to which the vessels were exposed (consistent with the licensee's j

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 j

average chemical crposition 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 not result i

was the same. That is, the date were determined to be credible. However, in future revisions to -

the PTLR and/or when additional surveillance data become availaide, the licensee should address the method for assessing the credibility of the data including the method for accounting for irradiation environment and chemical composition 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 materialis intermediate shell forging 5P 5g33 with a 1/4T ART of 70'F and a 3/4T ART of 60'F and (2) the P/T limit curves are appropriate.

3.2.6 Svron Unit 2 Based on the material provided by the licensee, the staff confirmed that (1) the limiting materialis circumferential weld, WF 447 (heat 442002) and (2) the P/T limit curves are appropriate.' The ARTS calculated by the staff (1/4T ART of 81.g'F and 3/4T ART of 67.6'F) were slightlylower

• 6 (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 resuM of minor differences in chemical composition and in applying the ratio procedure.

3.2.7 Braidwood Unit 1 Based on the material provided by the licensee, the staff confirmed that (1) the limitintI material is circumferential wold, WF 562 (heat 442011) and (2) the P/T curves are appropriate. "he ARTS calculated by the staff (il4T ART of 69.7'F and 3/4T ART of 60.6'F) were lower (i.s., 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 wold. Furthermore, the P/T limit curves are more conservative than would be required using the methodology in WCAP.

14040 NPA. Since the lloonsee'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 circumferential wold, WF.562 (host 442011) with a il4T ART of 66.6'F and a 3/4T ART of 68.1'F and (2) the P/T curves are appropriate.

3.3 Low Pressure Overnressure Protection System Evaluation 3.3.1 LTOP Enable Temoeraturs The LTOP enable temperaturs is the temperature below which the LTOP system is required to i

be operable. The licensee's proposed enable temperature (1) accounts for instrumentation uncertainties associated with the instrumentation used to enable the LTOP system and (2) implement the ASME Code Case /1996 Addenda methodology of using an enable RCS liquid temperature corresponding to the reactor vessel WT metal temperature of RT et + 50 or 200'F, whichever is Greater. Therefore, the minimum allowed enable temperature was calculated as the larger of either RT,,+ 50'F + E,. + delta Tw, or 200'F. In w.is calculation, E refers to inst:ument error while delta Twt refers to the temperature differer:ce between the reactor coolant and the metal at a distance one fourth of the vessel well thickness from the inside surface in the beltline region.

The use of 50'F in the above methodologyis consistent with ASME Code Case N 514 and the 1996 Addenda. Accounting for the delta Tw, is consistent with Branch Technical Position (BTP)

RSS 6 2, which states that the enable temperature is defined as "the water temperature corresponding to the metal temperature...at the beltline location (1/4T or 3/4T) that is controlling In the Appendix G limit calculation? 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 than are required by the aforementioned documents. Basta on the above discussion the staff finds acceptabla 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 uncertelnties, and delta Twg, the minimum allowed enable temperature for each of the unNs is as followw

e 9

TABLE 2: ENABLE TEMPERATURES Minimum Atowed EnsWe Colodeled Temperstwo EneWe teresterer Plant RT,

E.,

DetoTu Temperstwo Coloulated er 20FF) traiewood 1 78 FF 16FF 20.254*F 170.rF 20FF Steieweed 2 et9*F ilVF 20.264*F 101.2*F 20FF Dyren 1 700*F il FF 29.254*F 164.3*F 200*F Dyren 2 87.8*F 16.rF 20.264*F 181.9'F 20FF For the above listed RT,cn 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 Sution XI, Appendix G,1996 Addenda.

Therefore, the staff finds the liwresee's proposed enable temperature acceptable.

3.3.2 LTOP Actuation Setoolnt The ASME Code Case and 1996 Addenda ooth allow LTOP systems to be designed to limit the peak pressure at the contiolling 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 l

has approved the use of the steady state P/T limits for the design of LTOP. WCAP 14040 NP-A provides a methcdology for calculating setpoints for LTOP systems that use pressurtzer PORVs t

with variable setpoints. Section 3.2, "COMS Setpclnt Determination," of WCAP-14040 NP A provides discussions of parameters that nood to be considered in determining the LTOP octuation 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 setpoints 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 PORV lift 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 LOFTPAN computer code. Use of the LOFTRAN compV.sr code is consistent with WCAP 14040 NP A. For the mass addition transients, the RCS was assumed to be water eclid. The analyses accounted for iriection from a single charging pump and calculated the amount of overshoot that would occur during such a transient. To ensure consistency between the analysis assumptions and the TS, Surveillance Requirement (SR) 4.5.3.2 ensures thrt all other charging and safety ir(oction (SI) pumps are made incapable of irQecting into the RC4 while in the LTOP region. This SR, however, uses a temperature of 330*F to make the pmps inoperable Irnteed of the licensee's proposed LTOP enable temperature of 350*F. Using 330*F is acceptable because the minimum required enable temperature (Section 3.1 of th;s asfety evaluation (SE))is 200*F. A configuration allowing 81 pumps to be evallable when the pressurizer levells less than or equal to 5% is evalus,ted in Section 3.3.4 of this SE.

10 Heat-addition transients were also run with the RCS in a water solid condition. For these analyses, the secondary system was assumed to be 50'F higher than the RCS Or.e RCP was assumed to start and consequent heat addhion from all SGs was accounted for. Expansion of the reactor coolant resulted in pressurization of the RCS and actuation of the LTOP system. The resuhing overshoot values were determined by these analyses. To ensure consistency between Sie arealyses assumptons and the TS, a note is included in TS 3.4.1.3 to or.sure that RCPs are not started unless the secondary side of any SG is less than 50*F higher than the RCS.

For all cases analyzed, tne licensee conservatively assumed one PORV failed and, therefore, credited only one PORV for pressure relief. Additionalry, 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 sucan relief valves for PORV as currently allowed by TS 3.4.9.3. The licensee evaluated the results of the heat addition and mess-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 dynam;c head effects as well as instrumentatiors uncertainties in the final determination of the LTOP setpoints.

6 The dynamic head effect was divided into two regions. For RCS temperatures above 120*F, the Econses used a pressure drop corresponding to all four RCPs and both RHR pumps running.

l For RCS tortperatures less than or equal to 120'F, the licensee used a pressure drop corresponding to one RCP and both RHR pumps running. Plant administ ative procedures prohibit running more than one RCP below 120*F.

The above analyses were performed using the LOFTRAN computer code assuming the original steam generators.- For Byron Unit 1 and Braldwood Unit 1, Framatome and Comed Nuclear Fuel Services performed arulyses that verified the Westinghouse calculated setpoints remain valid for the replacement steam generators. These analyses were performed using RELAP5/ MOD 2-B&W computer code, which is riocumented in Topical Report BAW-10164P-A. This code was 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 prewnted in Figures 2.3 of each unit's PTLR acceptable. In addition, the staff finds acceptable the licensee's request to perform LTOP unalyses using the NRC approved RELAP5/ MOD 2 B&W computer code.

3.3.3 RHR Suction Relief y.glyg.g 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 med in place of PORVs for LTOP. WCAP-14040 NP-A does not address the use of the

~ RHR suction relief valves. The staffs SE for the WCAP, dated October 16,1995, states that Econsees who use the WCAP should address this issue in their PTLR submittal. The licensed addree ad the use of the RHR suction relief valves in its letter dated January 8,1998. The 5cetss's evaluation justiDd the use of the RHR suction relief valves at its current lift setpoint of 450 pi,g and appro?riately accounted for a setpoint drift of 3% and an accumulation of 10% as recommended by Anicle NC-7000 of the ASME Boiler and Pressure Vessel Code. The staff reviewed the licensee's evaluation and found it acceptable.

The licensee also incli.ded a requirement in the PTLR to evaluate the RHR suction relief valves in a similar manner whenever the P/T limits are revised. The staff finds this consistent with the recornmendation in the SE for WCAP-14040-NP-A and, therefore, acceptable.

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3.3.4 Avaliabilltv of 81 Pumos with Pressurizer Level s 5%

i To mitigate a loss of decay heat removal event, TS 3.5.4.2 requires at least one 81 pump and Sow 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, operators would have longer than 10 minutes (licensee calculations show 18.4 minutes) before the pressurtzer would become water i

solid. This would provide sufficient time to credit manual operstor action to mitigate this event i

before overpressurization would occur.

The licensee further stated that typical operating practices at Byron and Bra dwood ensure that i

_ only one 81 pump is available. However, for circumstances where both 81 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 hand switch for each 81 pump from the Pull to Lock position to the Run position and opening the 81 l

pump o@f leg discharge isolation valves. Therefore from an overpressure protection standpoint, the potential for an inadvertent start of both Si pumps is not cont,ldered a credible event.

i 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 81 pump start, and (2) the administrative controls for ensuring that the inadvertent stcrt of both Si pumps event is not credible.

' 3.3.5 Bolt-up Temocrature I

The licensee did not include instrumentation uncertainties in the bolt-up temp 6rature. However, the licensee proposed to account for instrumentation uncertainties in the minimum pressurization i

temperature (Section 2.5 of each unit's PTLR) to ensure that the RCS is not capable of being l

pressurized (i.e., the RCS rom.alns vented) until the RCS temperature is greater than or equal to j

the minimum allowable bolt-up temperature plus instrument uncertainty as determined using a technique consistent with ISA 867.04 - 1994. This was determined to be acceptable by the staff.

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4.0 CONCLUSION

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Based upon the staff evaluations, as discussed in Section 3.0 above, the NRC concludes that it is acceptable for the licensee to relocate the P/T limit curves and LTOP system limits from the t

Byron, Units 1 and 2l and Braidwood, Units 1 and 2, TS to a licensse controlled PTLR. The staff also concludes that it is acceptable to relocate the surveillance capsule withdrawal schedule to

.P the licensee controlled PTLR since changes to this schedule are controlled by the requirements of Appendix H S 10 CFR Part 50. -

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i The staff has reviewed the proposed fluence vaWes for the four Byron and Braidwood units for i

the revision of the P/T curves and finds that the existing approved values are conservative and, j

therefore, acceptable. In addition, the P/T limits meet the requirements of 10 CFR Part 50, Appendix G and none of the RTm values exceed the screening criteria specified in 10 CFR 50.61.

The NMC notes, however, that the licensee should /b re-evaluate the appropriate method for determining the best-estimsle chemical composition as additional chemical composition data E

become available, p) assess the impact of its assumption that the vessel weld has the same i

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i 12 RTecnu value as determined from the surveillance weld in future revisions te the PTLR and/or when additional surveillance data become available, and (3) address the method for assassing i

the credibility of the dats, 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 Snds the licensee's proposal consistent with the staff's SE approving the WCAP and also l

consisteat with BTP RSB 5-2, ASME Code Case N 514 and ASME Coction XI, Appendix G,1996 i

Addenda. Based on ti e above discussion and the evaluation provided in Section 3 of this SE, j

the staff finds the licensee's proposed LTOP enable temperature and actuation setpoints acceptable. The staff further Snds acceptable the licensee's request to use the REl.AP5/ MOD 2-B&W computer code for LTOP analyses.

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5.0 REFERENCER

1. Letter from R. R. Assa, NRC, to D. L Farrar, Commonwealth Edison Company, "Exemrtion from Requirements of 10 CFR 50.60 - Braidwood Station, Unit 1," July 13, j

1995.

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2. Letter from C. I. Grimes, NRC, to R. A. Newton, Westinghouse Electric Corporation,

" Acceptance for Referencing of Topical Repori WCAP-14040, Revision 1, ' Methodology Used to Develop Cold overpressure Mitigating System Setpoints and RCS Heatup and 4

Cooldown Limit Curves,"' October 16,1995. (Also known as WCAP-14040-NP-A).

O. Letter from G. F. Dick, NRC, to 1. 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'-

u Byron Station, Units 1 and 2," November 29,1996.

4. Letter from J. B. Hosraer, Commonwealth Edison Company, to NRC Document Control l

Desk, " Reactor Vessel in'egrated Surveillance Program 10 CFR 50, Appendix H, Section lll.C," May 6,1997. (WCAP-14824, Revision 1 is Attached).

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5. Letter from J. B. Hosmer, Commonwealth Edison Company, to NRC Document Control Desk, " Application for Amendment to Appendix A, Technical Specifications, for Facility i

Operating Licenses Relocation of Pressure and Temperature Limits," May 21,199/.

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6. Letter from J. B. Hosmer, Commonwealth Edisors Company, to NRC Document Control Desk,

• Supplement to the Application for Amendmer.l 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 Edison C,empany, to NRC Document Control Desk, " Supplemental Information 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).

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1 13 8.- Letter from G. F. Dick, NRC to I. M. Johnson, Cc,T T,cr.nalth Edison Company,

" Exemption nom Requirements of 10 CFR 50.60 - Braidwood Nuclear Station, Unit 2,*

December 12,1997.

s G. Letter from H. G. Stanley, CommonweaMh Edison Company, to NRC Document Control Desk,

• Supplemental Information Portaining to Technical Specification Amendment j

Regarding Pressure Temperature Curves - Symn and Braidwood Nuclear Power Stations,' January 8,1398. (Errata to WCAP 14824 Revision 2, WCAP 14940 and i

WCAP 14970 are Attadwd).

10. Letter from H. G. Stanley, Commonwealth Edison Company, to NRC Document Control 1

j Desk, " Supplemental information Pertaining to Technical Specification Amendment i

Regarding Pressure Temperatuie Curves - Byron and BrsiWood Nuclear Power j

Stations,' January 13,1998,

11. Letter from G. F. Dick, NRC, to O. D. Kingsley, Commonwealth Edison Company,

• Exemption from Requirements of 10 CFR 50.60 - Byron, Units 1 and 2, and Braidwood, Units 1 and 2,* January 16,1998.

12. Letter from R. A. Capra, NRC, to O. D. Kingsley, Commorwealth Edison Company,

" integration of Reactor Pressure Vessel Surveillance Program for Byron and Braidwood, Units 1 and 2,* January 16,1998.

Principal Contributors: K. Karwoski L Lois M. Shualbi M W. Weston t

Date: January 20,1998

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