ML20203A282

Safety Evaluation Supporting Amends 101,101,92 & 92 to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,respectively

ML20203A282
Person / Time
Site:	Byron, Braidwood
Issue date:	02/03/1998
From:	NRC (Affiliation Not Assigned)
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ML20203A279	List: ML20203A275 ML20203A282 ML20203B723
References
NUDOCS 9802230329
Download: ML20203A282 (23)

Text

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UNITE 3 STATES j

j NUCLEAR REGULATORY COMMISSION 2

WASHINGTON, D.C. 30606 4001

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.101 TO FACILITY OPERATING LICENSE NO. NPF-37.

AMENDMENT NO.101 TO FACILITY OPERATING LICENSE NO. NPF 66.

AMENDMENT No 92 TO FACILITY OPERATING LICENSE NO. NPF-72.

AND AMENDMENT NO. 92 TO FACILITY OPERATING LICENSE NO. NPF-77 l

COMMONWEALTH EDISON COMPANY BYRON STATION. UNIT NOS.1 AND 2 BRAIDWOOD STATION. UNIT NOS.1 AND 2 DOCKET NOS. STN 50-454. STN 50-455. STN 50-456 AND STN 5N57

1.0 INTRODUCTION

By letter dated February 28,1997, as r,upplemented by letters dated November 13,1996, and March 20, June 24, August 19 and November 3,1997, the Commonwealth Edison Company (Comed, the licensee) requested license amendments to revise the technical specifications (TS) for the Byron Station, Units 1 and 2, and for the BraiNod Station, Units 1 and 2. The intent of these requests was to eliminate certain steam generamt (SG) tube repair criteria as well as certain inspection and reporting requirements for the SG tubes. These license amendment requests were necessitated by the planned replacement of the original steam generators (OSG) in Byron, Unit 1, and Braidwood, Unit 1, which are Westinghouse Model D4 SG, with the replacement steam generators (RSG) which are Babcock & Wilcox, Intematbnal (BWI) SG. The proposed TS revisions reflect the significant design differences between the OSG and the RSG.

The SG tube repair criteria and the associated surveillance and reporting requirements are identified as the interim plugging criteria (IPC). These repair criteria were required to address a form of SG tube degradation in the OSG known as outer diameter stress corrosion cracking (ODSCC). The November 13,1996, and March 20, June 24, August 19 and November 3,1997, submittal provided additional information that did not change the initial proposed no significant hazards consideration determination.

Additionally, the P attemative repair criteria (ARC) for SG tube flaws within the OSG tubesheet as well as two separate OSG tube repair sleeving meihodologies are no longer needed in the Byron, Unit 1, and Braidwood, Unit 1, TS due to the design differences between the OSG and the R5G.

The IPC were added to the Byron, Unit 1, and Braidwood, Unit 1, TS in license ameridments issued on November 9,1995. These criteria and associated surveillance and reporting requirements were not added to the Byron, Unit 2, and Braidwood, Unit 2, TS since both these units have Westinghouse Model D5 SG which have not experienced any significant ODSCC SG tube degradation. While the pending license amendment requests only affect the Byron, Unit 1, i

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4 and Braidwood, Unit 1, TS, the requested license amendments apply to Unit 2 of both stations so as to maintain the continuity of the license amendment numbering system for both the Byron and Braidwood Stations.

Finally, the licensee, in the subject license amendment requests, proposed to restore the limiting l

TS value for the long-term dose equivalent lodine-131 (del) concentration in the primary coolant to its original licensing basis of 1.0 microcurie por gram ( Cilgm) from the present TS limiting values of 0.35 pCilgm for both Byron, Unit 1, and Braidwood, Unit 1. Similarty, the licensee has j

proposed to restore the short-term (i.e., less than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />) TS limit to its original licensing basis 4

of 60 Cilgm from the present TS vak e of 20 uCi/gm. Both the forthcoming installation of the l

RSG and the proposed restoration of the DEI to their originallicensing bases required the staff to j

reevaluate the radiation dose exposures w5ich could result from postulated design basis 3

accidents (DBA). For example, the RSG have a larger volume of water and metal than the OSG which affect the evaluation of radiation doses attributable to DBAs due to the difference in the l

release of steam during transient or accident conditions, i

Both the Byron, Unit 1, and the Braidwood, Unit 1, TS revisions are to be implemented in the first and subsequent operating cycles following installation of the RSG. For Byron, Unit 1, the RSG l

installation started in the refueling outage initiated in November igg 7. Accordingly, the revisions to the Byron, Unit 1, TS are now effective and will be implemented in the forthcoming fuel cycle i

(Cycle 9) starting in eady 1998. In a similar fashion, the avisions to the Braidwood, Unit 1, TS are effective now and will be implemented after the RSG installation in Braidwood, Unit 1, which is scheduled for the fall 1998 refueling outage (Cycle 8).

i 2.0 EVALUATION l

l 2.1 Discussion

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The NRC staff riviewed four separate DBA whose onsite and offsite radiological consequences l

would be altered as a result of the installation of the RSG in both Byron, Unit 1, and Braidwood, l

Unit 1. Though both stations are nearly identical, there are several differences between the two sites such as the local meteorology which affects the values of the atmospheric dispersion i

factors. Further, the unfiltered infiltration rates for the control rooms are also different between i

each station. Accordingly, though such features as the RSG, the reactor core design and containment are identical for each station, the staff analyzed each site separately in recognition

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of these site specific differences.

The results of these station specific analyses are preser:ted in separate tables in this safety l,

evaluation. Most of the attached tables, however, are applicable to both stations due to the nearly identical physical plant characteristics. The staff's conclusions in this Safety Evaluation (SE) are also presented separately for each station since the conclusions are slightly different for i-each station.

The removal of the IPC for ODSCC flaws from the Byron, Unit 1, and the Braidwood, Unit 1, TS required no independent staff analysis since the RSG do not have the same type of SG tube 3

support structures as the OSG and, therefore, the SG tube repair criteria for the ODSCC flaws which occurred in the OSG, are not applicable to the RSG. Accordingly, the RSG will not be

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subject to the relatively large end of cycle SG tube leakage which could occur under postulated u

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3-accident condi' ions. Similarly, the removal of the SG tube ARC for flaws occurring within the OSG tubesheets (i.e., the F* repair criteria) also required no independent staff analysis for the same reason as cited above. Tnis is also true for the proposed removal of the two separate sleeving methodologies presently in the Byron, Unit 1, and the Braidwood, Unit 1, TS. The removal of these various SG tube repair criteria and their associated surveillance and reporting -

requirements, are considered to be administrative in nature since they are not required to ensure the safe operation of the RSG due to the significant design differences between the OSG and the RSG.

2.2 Evaluation of the Redie:Osical Consecuences 2.2.1 Ryton. Unit 1. and Braidwood. Unit 1. Radie:ccice: Analysis of Postulated DBAs The licensee performed various reanalyses of the Byron, Unit 1, and Braidwood, Unit 1, for the postulated accident conditions analyzed in Chapter 15 of the Final Safety Analysis Report l

J (FSAR). These reanalyses were required due to the chance in the design of the Unit 1 SGs of l

both stations from the Wes'inghouse Model D4 steam generatorc to the Babcock & Wilcox steam l

generators and because the licensee has proposed an amendment to restore both the maximum instantaneous and the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> TS values for DEI to the original licensing basis levels.

The change in SG design also affects the manner in which the reactor responds in the event of certain accidents. This change in response can impact the radiological releases, thereby affecting offsite and onsite radiological doses.

On November 13,1996, the licensee submitted its revised steam generator tube rupture (SGTR) analysis, in letters dated February 11, May 20, July 18, and October 3,1997, the staff transmitted to the licensee requests for additional information (RAI). In letters dated March 20, June 24, August ig, and November 3,1997, the licensee provided responses to these RAls. In these responses, the licensee identified four postulated accidents whose consequences would be altered as a result of the change in SG design from the OSG to the RSG. These are:

1. Main Steamline Break

2. Steam GeneratorTube Rupture

3. Locked Rotor

4. Rod Ejection The licensee indicated that all other postulated accidents would be unaffected by the change in the SG design.

The licensee calculated the offsite doses for the postulated main steamline break (MSLB) and the SGTR accidents. The licensee calculated doses which met the acceptance criteria of Standard Review Plan (SRP) Sections 15.1.5 and 15.6.3. The licensee indicated that the control room operator radiation exposure doses from a postulated loss of coolant accident (LOCA) would bound the operator's doses from the MSLB and SGTR accidents. The licensee did not calculate any doses for the postulated locked rotor or rod ejection accidents even though the quantity of steam releascd during the course of these accidents increased due to the change in SG design and the quantity of melted fuelincreased for the rod ejection accident. The licensee also provided the change in steaming rates as a result of the change in steam generator design.

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The following sections provide the results of the staff's independent assessmert of the licensee's -

reanalysis of the FSAR Chapter 15 accidents affected by both the SG design change and the increase in the maximum permissible TS values of the primary coolant del.

2.2.2 Postulated Main Steamline Break Accident for Byron. Unit 1. and Braidwood. Unit i The staff performed a confirmatory evaluation cf the consequences of an MSLB accident outside l

containment. Two cases were analyzed. In the first case, a pre-existing spike was assumed to have occurred prior to the event. During the event, primary-to-secondary leakage was assumed 1

to c: cur et the TS maximum allowable rate of 150 gallons per day (gpd) per SG. For this case, the reactor coolant iodine specific activities were assumed to be at the TS F8gure 3.4-1 full power DEI limit of 60 pCi/gm. The secondary coolant lodme specific activity was assumed to be at the secondary coolant specific activity equilibrium value of 0.1 pCilgm.

Tne second case assumed that the postulated accident itself initiates a concurrent iodine spike.

The reactor coolant activity level was assumed to be at the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> TS value of 1.0 pCilgm del.

Secondary coolant activity was assumed to be at the TS limit of 0.1 uCilgm del. It is assumed

. that when the iodine spike occurs, it results in a release of iodine from the fuel gap to the reactor coolant at a rate in Cilut.it time wtilch is 500 times the normal iodine release rate necessary to maintain the reactor coolant at 1.0 uCi/gm (i.e., a spiking factor of 500). The licensee indicated.

that the MSLB evtnt did not result in failed fuel.

For both analyses, it was assumed that a 150 gpd primary-to secondary tube leak occurred in the faulted SG until it was laolated. For the intact SG, it was assumed that primary-to secondary l

leakage occurred at a rate of 150 gpd/SG for the duration of the accident. Concurrent with the l.

MSLB, it was also assumed that there would be a loss of offsite power. Consequently, the main condenser was unavailable for steam dump. Haat removal from the reactor core would have to occur, therefors, by discharge through the code safety valves. This was anticipated to occur for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> after which no further steam or radioactivity release would occur because the reactor heat removal (RHR) system would bo sufficient to maintain core cooling.

Table 1 presents the details of the staff's assumptions for both the Byron and Braidwood Stations. The results of the staff's assassment are presented in Tables 7, g and 11 for Byron, Urdt 1, and Tables 8,10 and 12 for Braidwood, Unit 1. Tables 7 and g resent the offsite doses.

p and Table 11 presents the control room doses for Byron, Unit 1. Similarly, the offsite doses for Braidwood, Unit 1, are in Tables 8 and 10 while the control room doses are in Table 12. For this costulated accident, the staff did not perform an assessment of the whole body dose associated with the release of noble gases because the thpoid dose is so limiting with respect to compliance with General Design Criterion (GDC) ig of Appendix A to 10 CFR Part 50 and the guideline values in 10 CFR Part 100. The doses were found to be within a small fraction of Part 100 guidelines for the accident-initiated spike case, within Part 100 for the pre-existing spike case, and less than GDC 19 requirements for the control room in both cases for both Byron, Unit 1, and Braidwood, Unit 1.

t 5-t 2.2.3 ' Postulated Steam Generator Tube Ruoture Accident for Byron. Unit 1. and Braidwood. Unit 1 l

The staff performed a confirmatory evaluation of the consequences of a postulated SGTR accident. As with the MSLB accident, two cases were analyzed. One case involved a pro-existing lodme spike and the other case, an accident-l.%isted spike. For the pre-existing spike case, the reactor coolant iodine activity level of the del was assumed to be at the full power level of 60 uCi/gm of the DEI in TS Figure 3.4-1. The secondaiy coolant iodine specific activity was based the TS normal operation limit of 0.1 pCilgm del.

I The accident initiated spike case assumed the SGTR event itself initiated a concurrent iodine i

spike. The reactor coolant was assumed to Me at the TS 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reactor coolant activity level of the DEI of 1.0 uCi/gm. The secondary system activity was assumed to be at the TS limit of 0.1

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uCilgm del. The SGTR is assumed to initiate an iodine spike which resuhs in a release of iodine i

from the fuel gap to the reactor coolant at a rate in Ci/ unit time which is 500 times the normal l.

iodine release rate necessary to maintain primary coolant at 1.0 uCilgm. The licensee indicated l

that the postulated SGTR accident did not result in any failed fuel.

For both analyses, it was assumed that a primary to-secondary leak occurred in the intact SGs at a rate of 150 gpd for the duration of the accident. It was also assumed for these two analyses l.

that offsite power was lost and the main condenser was unavailable as a heat sink to remove

- decay heat from the reactor. After 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> following this event, the analysis indicated that i.

reactor core heat had been sufficiently removed so that the unit could be placed on the RHR, and L

no further steam release or activity release was assumed to occur. For completeness, the staff

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incorporated the contribution from the intact SG for this period even though the dose contribution would be small from this source. The licensee had excluded this contribution from its i

assessment.

l Table 4 presents the assumptions used by the staff in its assessment of the Unit 1 RSG for both -

l the Byron and Braidwood Stations. Some of the assumptions presented for the MSLB are also relevant to the SGTR. Therefore, Table 1 should be referred to as a source of information for the l

, postulated SGTR accident also.

I The potential consequences of a postulated SGTR accident are presented in Tables 7,9 and 11

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for Byron, Unit 1, and in Table 18,10 and 12 for Braidwood, Unit 1. For this accident, the staff F

did not perform an assessment of the whole body dose associated with the release of noble

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gases because the thyroid dose is so limiting with respect to compliance with GDC ig and

10 CFR Part 100. The doses were found to be within a small fraction of Part 100 guidelines for

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the accident-initiated spike case, within Part 100 for the pre-existing spike case for offsite i

exposures, and less than the requirements in GDC ig for the control room in both cases for both l

Byron, Unit 1, and Braidwood, Unit 1.

2.2.4 Postulated Locked Rotor Accident for Byron. Unit 1. and Braidwood. Unit 1 4

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The staff performed a confirmatory assessment of the consequences of a postulated reactor i

coolant pump kscked rotor event and subsequent leakage of steam from the secondary system due to the leakace of primary coolant to the secondary system. Leakage from the primary side to the secondary side was assumed to exist prior to the accident. citial conditions assumed a i

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4 6-contribution to releases from the initial reactor coolant activity levels at the proposed TS values and contribution of at,tivity from five percent of the gap inventory being raleased to the primary coolant.

Table 5 presents the assumptions used by the staff in their assessment of the consequences of 4

7 a locked rotor accident. The staffs assessment of the potential consequences of a postulated locked rotor accident are presented in Tables 7,9 and 11 for Byron, Unit 1, and in Tables 8,10 and 12 for Braidwood, Unit 1. The doses were found to be less than the requirements in GDC 19 and within a small fraction of the guideline values in 10 CFR Part 100 fu offsite exposures for L

both Byron, Unit 1, and Braidwood, Unit 1.

i 2.2.5 Postulated Rod Election Accident for Byron. Unit 1. and Braidwood. Unit 1 i

The staff performed a confirmatory assessment of the consequences of a postulated rod ejection accident. It was assumed that the reactor was operating with equilibrium activity levels in the primary and secondary systems based upon a failed fuel rate of one percent and a primary-to-secondary leak rate of 150 gpd/SG. The staffs analyses assumed that the release pathway to j

the environment would be either via containri ont leakage or via primary-to-secondary leakage.

The first pathway assumes that the postulated rod ejection results in the release of activity from the fuel gap and that a certain amount of fuel melting occurs. This activity is released from the fuel to the primary coolant. The primary coolant is then released to containment where the j

activity 8eaks from the containment at a rate consistent with the TS maximum allowable contaironent leak rate for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at half that leak rate after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The second i

pathway assumes that the activity from the fuel gap and the melted fuel is transported to the primary coolant. The primary coolant is then assumed to leak to the secondary side at the TS limit for the SG (i.e., a leak rate of 150 gpd per SG). Activity is released from the secondary side i

via the relief valves since it is assumed that there is a loss of offsite power concurrent with the rod ejection. The licensee indicated that in the event of a rod ejection, gap activity would be released from ten percent of the fuel and that five percent of the fuelwould undergo melting.

L Table 6 presents the assumptions used by the staffin its assessment. The results of the staffs assessment are presented in Tables 7,9 and 11 for Byron, Unit 1, and in Tables 8,10 and 12 for Braidwood, Unit 1. The doses were found to be less than the requirements of GDC 19 and well within Part 100 guidelines for offsite exposures for Byron, Unit 1.

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The radiation exposure doses for Braidwood, Unit 1, were found to be less than the requirements of GDC ig and slightly gmater than "well within the guideline values in 10 CFR Part 100" as stated in Appendix A of SRP Section 15.4.8 for offsite exposures resulting from the containment j

leakage pathway. Although the doses exceeded the guidelines of Appendix A, the staff j

concluded that this departure from the criteria in SRP Section 15.4.8 is acceptable because:

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1. The doses are below the guideline values of 10 CFR Part 100; l '

2. The consequences were calculated on the basis of fuel cladding failures and fuel melting that are predicted by conservative models; and 2

3. The SRP acceptance criterion itself is conservative.

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7 2.2.6 Conclusions for Bvron. Unit 1 The staff has assessed those accidents for which the change from the Westinghouse Model D4

. (i.e., the OSG) to the BWI SG (i.e., the RSG) would impact the offsite and control room operator doses. As a result of this assessment, the staff has concluded that, for those accidents which

' are affected by the installation of the RSG and the restoration of the TS values of the short-term

- and steady state DEI to their original licensing bases, the doses would not exceed the dose guidelines presently contained in the Standard Review Plans for the postulated rod ejechon, MSLB, SGTR and locked rotor accidents and that in no case would the appropriate fraction of 10 CFR Part 100 be exceeded offsite nor would the GDC ig of 10 CFR Part 50, Appendix A, requirement limiting operator dose exposures, be exceeded. - Therefore, the staff finds that the proposed replacement of the Westinghouse Model D4 SG with the B&W SG as well as the change to the maximum permissible reactor coolant activity levels of the del in TS Section 3.4.8.a and TS Figure 3.4-1, are acceptatie from an evaluation of the radiological consequences.

Specifically, these changes restore the long-term (i.e., greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />) DEI to 1.0 uCi/gm and the short-term DEI to 60 pCi/gm.

2.2.7 Conclusions for Braidwood. Unit 1 The staff has assessed those accidents for which the change from the Westinghouse Model D4 SG to the BW1 SG would affect the offsite and control room operator doses. As a result of this assessment, the staff has concluded that, for those accidents which are affected by the change j

to the BWI SG, the doses would not exceed the dose guidelines presently contained in the Standard Review Plans for the MSLB, SGTR and locked rotor accidents and that in no case would the appropriate fraction of the radiation exposure guidelines in 10 CFR Part 100 be exceeded nor would the requirements for limiting radiation exposure to control room operators in GDC ig of 10 CFR Part 50, Appendix A, be exceeded. For the rod ejection accident, it was determined that the offsite doses exceeded the guidelines of SRP Section 15.4.8, However, this was considered acceptable because the assumptions with respect to fuel melting and fuel cladding failure were considered conservative and the offsite consequences were below the guideline values of 10 CFR Part 100, The control room operator doses from a rod ejection are within the requirements of GDC ig. Therefore, the staff finds that the proposed replacement of

- the Wsstinghouse Model D4 SG with the BWI SG as well as the change to the maximum permissible reactor coolant activity levels of the DEI in TS Section 3.4.8.a and TS Figure 3.4-1, are acceptable from a radiological standpoint.

2.2.8 Chanaes to the Technical Soecifications The changes to the TS of the Byron and Braidwood Stations are identical because the SG tube repair criteria being removed from the TS were identicalin Unit 1, in both stations. Some of these changes are primarily administrative in nature while others were reviewed for their effect on public health and safety (such as onsite and offsite radiation exposure doses which could occur under postulated accident conditions).

The Table of Contents for both stations are revised to reflect the restoration of the primary coolarit's maximum permissible short-term DEI to its originallicensing basis of 60 uCi/gm. This is reflected in the addition of a revised TS Figure 3.4-1 which increases the short-term DEI to

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60 Ci/gm for Unit 1 of both stations which are soplicable in the operating cycles after installation of the BWI RSG. This change is primarily administrative.

Sections 4.4.5.2 in the TS of both stations are being revised to indicate that the voltage-based SG tube repair criteria added in Joense Amendmert Nos. 77 for the Byron Station and License Amendment Nos. 6g for the Braidwood Station, are only applicable for the OSG which are being replaced by the BWI RSG. The voltage-based repair criteria are not applicable to the RSG since.

the RSG do not have the same type of SG tube support structures as the OSG and, therefore, the SG tube repair criteria for the ODSCC flaws which occurred in the OSG, are not applicable to the RSG.

Sections 4.4.5.3 " Inspection Frequencies," in the TS of both stations are being revised to reflect the installation of the RSG. Specifically, the proposed revision requires as an added feature, an inspection no later than 24 calendar months after initial operation following a SG replacement.

This revision is acceptable in that it applies a requirement on the inspection frequency of the first inservice inspection of the RSG which is identical to that for the OSG.

Section 4.4.5.4, " Acceptance Criteria," in the TS of both stations is being revised to twdefine the plugging or repair limit and the sleeving SG tube repair process itself, when implemerting either i

laser welded or tungsten inert gas (TIG) welded sleeves, as applicable only to Westinghouse Model D4 SG (i.e., Unit 1 of the Byron and Braidwood Stations) and Westinghouse Model D5 SG (i.e., Unit 2 of the Byron and Braidwood Stations). This revision does not affect in any way the existing definition of the pl"gging or repair limit for Model D4 or Model DS SG. Accordingly, the not affect of this revision is to make this definition and the sleeve rcpair process not applicable to the RSG. This is acceptable in that the RSG do not require SG tube sleeve repairs.

These TS Sections are also being revised to indicate that the repair criteria for SG tube flaws within the OSG tubesheet (i.e., the F* SG tubes) are only applicable to the OSG and not the RSG. This is also acceptable for the same reason cited above.

Section 4.4.5.5, " Reports," in the TS of both stations is being revised to indicate that the results of the F* tube inspections need be submitted to the Commission only for the OSG. This is,

acceptable in that the revision does not eliminate any existing reporting requirements for the OSG and this revision will not require the licensee to report F* indications in the RSG since the RSG are not subject to the same forms of SG tube degradation in the tubesheet area which have occurred in the OSG.

Section 3.4.8 and 4.4.8 in the TS for both stations are being revised to state that the present maximum permissible primary coolant TS long-term (i.e., greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />) value of the DEI concentration of 0.35 pCilgm is applicable to the OSG in both the Byron and Braidwood Stations only until their replacement by the RSG. After the installation of the RSG in these stations, the long-term DEI TS limits for the RSG are restored to 1.0 uCi/gm for Unit 1 of noth stations.

Similarly, the short-term maximum permissible DEI in TS Figure 3.4-1 in the TS applicable to the RSG of both stations are revised to restore these values to their original licensing bases. These revisions are acceptable as discussed in Sections 2.2.6 and 2.2.7 of this SE. In summary, the staff concluded in these sections that the behavior of the RSG under transient and postulated accident conditions with the short-term and long-term maximum permissible DEI limits restored to their original licensing bases, did not adversely affect public health and safety.

.g.

Finally, the bases sections in the TS of both stations are revised to reflect the forthcoming installation of the RSG and the restoration of the TS limits for the DEI to their original licensing bases.

3.0 STATE CONSULTATION

in accordance with the Commission's regulations, the lilinois State official was notified of the proposed issuance of the amendments. The Stat' '3*ficial had no comments.

4.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to the installation or use of a facility component located within the restricted area at defined in 10 CFR Part 20 and change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative l

occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (62 FR 66134). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b),

l no environmental impact statement or environmental assessment need be prepared in l

connection with the issuance of the amendments.

5.0 CONCLUSION

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

Attachmentt. As Stated Principal Contributors: J. Hayes M. D. Lynch Date: February 3, 1998 1j

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TABLE 1 Assumotions for the Postulated Main Steamline Break Accident By/l Replacement Steam Generators Byron and Braidwood Stations. Unit 1 lodine Partition Factor Faulted SG 1.0 intact SGs 0.1 Steam and H O Releases frem 2

Faulted SG 0120 minutes 1.18E5 S2eam Release from Intact SGs (Ibs) 0-2 hours 4.10E5 2-8 hours 9.49ES B 1B hours 6.77ES 16-24 hours 5.81ES 24-32 hours 5.22E5 32-40 hours 4.82E5 Primary to Secondary Leak Rate 150 (gpd/SG)

Time to isolate Faulted SG (min) 120 Flashing Fraction Variable with respect to time. Provided in Comed letter dated 11/3/97.

Scrubbing Fraction 0

Primary Bypass Fraction for intact 0

SGs Duration of Plant Cooldown (hrs) 40

. TABLE 1 (continued)

Assumptions for the Postulated Main Steamline Break Accident BWI Replacement Steam Generators Byron and Braidwood Stations. Unit 1 Primary coolant concentration of 60 uCi/gm of dose eeuivalent *l.

Pre-existina Soike Value fuCi/am)

• l = 46.2 321=

51.7

• 1 = 73.9

• l = 11.1 S i = 40.6 Volume and Mass of primary coo! ant and secondary coolant.

Primary Coolant Volume (ft')

12,062 @S86.2 "F Primary Coolant Temperature ( F) 586.2 Mass of Primary Coolant (Ibs) 538,361 Primary Coolant Pressure (psla) 2,293 Pressurizer Temperature ( F)

'657 Pressurizer Pressure (p)sla) 2,293 Pressurizer Volume (ft 1,150 Secondary Coolant Steam Volume (ft )

2,780 Secondary Coolant Liquid Volume (ft )

2,423 Secondary Coolant Steam Mass /SG (Ibs) 5,571 Secondary Coolant Liquid Mass /SG (lbs) 105,224 4

Secondary Coolant Steam Temperature ( F) 523 Secondary Coolant Feedwater Temperature ('F) 440 4

TS limits for DE *l in the primary and secondary coolant.

Primary Coolant DE *l concentration ( Cilgm)

Maximum Instantaneous Value 60 48 Hour Value 1.0 Secondary Coolant DE *l concentration (gC!/gm) 0.1 4

'l

. TABLE 1 fcontinued)

Assumotions for the Postulated Main Steamline Break Accident BWI Replacement Steam Generators Byron and Braidwood Stations. Unit 1 TS value for the primary to secondary leak rate.

Primary to secondary leak rate, any SG (gpd) 150 Primary to secondary leak rate, total (gpd) 600 Maximum primary to secondary leak rate to the faulted and intact SGs.

Faulted SG (gpm) 150 Intact SGs (gpm/SG) 150 Letdown Flow Rate (gpm) 75 Equilibrium Release Rate from Fuel for 1 uCilgm of Dose Equivalent '8'l Cl/ day

• l = 2,040 1321= 5,300 5 1= 5,330

• l = 7,370

• l = 5,300 Control Room Free Volume (ft )

4.05ES Filtered Recirculation Flow 4.45E4 (cfm)

Recirculation Efficiency for all 90 forms of lodine (%)

Makeup Filter Efficiency for 99 all forms of lodine (%)

Makeup Air Filtration Rate 5400 (cfm)

4

. TABLE 2 Assumptions for the Postulated Main Steamline Break Aeddent BWI Rectacement Steam Generators Byron Station. Unit 1 Control Room Unfiltered Airinfiltration Rate 89 (cfm)

Occupancy Factors 0-1 day 1.0 1-4 days 0.6 l

Atmospheric Dispersion Factors l

(sec/m )

8 Control Room 0-8 hours 4.05E-3 8-24 hours 1.9E-3 1-4 days 5.7E-4 4-30 days 3.8E-4 EAB 6.8E-4 LPZ 0-8 hours 2.3E-5 8-24 hours 1.5E-5 1-4 days 6.4E-6 4-30 days 1.4E-6 Spiking Factor for Accident Initiated 500 Spike

TABLE 3 Assumptions for the Postulated Main Steamline Break Aeddent BWI Roolacement Steam Generators Braidwood Station. Unit 1 Control Room Unfiltered Air Infiltration Rate 25 (cfm) l Occupancy Factors 0-1 day 1.0 1-4 days 0.6

(

Atmospheric Dispersion Factors (sec/m )

Control Room 0-8 hours 6.2E-3 8 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> -

3.2E-3 1-4 days 8.4E-4 4-30 days 1.4E-4 EAB 7.7E-4 LPZ 0-8 hours 7.9E-5 8-24 hours 5.2E-5 1-4 days 2.1E-5 4-30 days 5.6E-6 Spiking Factor for Accident-Initiated 500 Spike

4 '

TABLE 4 Assumotions for Postulated Steam Generator Tube Ruoture (SGTR) Accident BW1 Replacement Steam Generators Byron and Braidwood Stations. Unit 1 lodine Partition Factor 0.01 Steam Release from Defective SG O 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (ibs) 9.55E4

>2 hours (Ibs)

O Steam Release from Intact SGs (Ibs) 0-2 hours 1.61ES 2-40 hours assumed to be at the same rate as for a postulated main steamline break i

Reactor Coolant Released to 1.41ES Faulted SG (Ibs)

^

Primary to Secondary Leak Rate (gpd/SG) 150 Time to isolate Faulted SG (sec) 3300

~_

. TABLE 5 Assumptions for the Postulated Locked Rotor Accident BWI Replacement Steam Generators Byron and Braidwood Stations. Unit 1 Core Thermal Power Level (MWt) 3565 Duration of Plant Cooldown by Secondary System (br) 40 Gap Fraction:

lodines 0.12 "Kr 0.30 All others 0.10 Cladding Failure (%)

5 Primary to Secondary Leak Rate (gpd/SG) 150 1

lodine Partition Factorin SG 0.01 Steam,*'cleased from 4 SGs (Ibs) 0-2 hours 5.65ES 2-8 hours 1.09E6 8-16 hours 8.03E5 16-24 hours 7.07E5 24-32 hours 6.48E5 32-40 hours 6.08E5 Primary Coolant Dose Equivalent *l Le.al (gCi/gm) 60 I

l l

l

. TABLE 6 Assumptions for the Postulated Rod Election Accident BWI Replacement Steam Generators Byron and Braidwood Stations. Unit 1 Core Thermal Power (MWt) 3565 Fuel Cladding Defects (%)

10 Primary to Secondary Leak Rate (gpd/SG)

.150 Metted Fuel (% of core) 5 Activity released to reactor coolant from melted fuel and available for release (%)

Primary to Secondary Release 100 for noble gases Pathway 50 forlodides Containment Pathway 100 for noble gases 25 for lodides lodine Paitition Factor From Relief Valves 1.0 During Steaming 0.01 Containment Volume (ft')

2.76E6 Leak Rate (%/ day) 0.1 for t = 0-1 day 0.05 fori > 1 day Steam Dump fnam Relief Valves (Ibs) 1.16E5 Duration of Steam Dump from Relief Valves 500 (sec)

Time between Accident snd Equalization of 3250 Primary to Secondary System Pressure (sec)

Steaming Rate Refer to Table 1 for values

.~.

, -i t

TABLE 7 Thyroid Doses from Postulated Accidents (Rem)

BWI Reolacement Steam Generators Byron Station. Unit 1 Accident E6B LP2 1.

Main Steamline Break Coincident Spike

• 2.4 3.3 Pre-existing Spiks" 3.4 0.55 l

t 2.

Steam Generator Tube Rupture Coincident Spike

• 4.9 0.23 Pre-existing Spike **

24.7 0.85 i

3.

Locked Rotor

• 0.60 0.54 4.

Rod Ejection *"

i Primary to Secondary Pathway 6.1 0.50 Containment Pathway 67 26 i

• Acceptance Criterion is 30 rem

" Acceptance Criterion is 300 rem l

• " Acceptance Criterion is 75 rem i

to-me m

~

~. = _.

. i TABLE 8 1

Thyroid Doses from Postulated A;t idents (Rem)

BW1 Replacement Steam Generators Braidwood Station. Unit 1 Accident EAR LPl 4

1.

Main Steamline Break Coincident Spike

• 2.7 11 Pre-existing Spike" 3.8 1.9 2.

Steam Generator Tube Rupture Coincident Spike

• 5.5 0.79 Pre-existing Spike" 28 2.9 3.

Locked Rotor

• 0.70 2.1 4.

Rod Ejection *"

Primary to Secondary Pathway 6.9 1.7 Containment Pathway 76 104

• Acceptance Criterion is 30 rem

" Acceptance Criterion is 300 rem

"* Acceptance Criterion is 75 rem

4 TABLE 9 Whole Body Doses from Postulated Accidents (Rem)

BWI Replacement Steam Generators Byron Station. Unit 1 Accident E6B LPl 1.

Main Steamline Break Coincident Spike *

<1

<1 Pre-existing Spike **

<1

<1 2.

Steam Generator Tube Rupture Coincident Spike *

<1

<1 Pre-existing Spike"

<1

<1 3.

Locked Rotor

• 0.18 0.069 4.

Rod Ejection

• Primary to Secondary Pathway 0.27 0.009 Containment Pathway 0.39 0.039

• Acceptance Criterion is 2.5 rem

" Acceptance Criterion is 25 rem

• Acceptance Criterion is 6.25 rem

. TABLE 10 Whole Body Doses from Postulated Accidents (Rem) 4 BWI Replacement Steam Generators Braidwood Station. Unit 1 l

Accident E6B (P_Z 1.

Main Steamline Break Coincident Spike *

<1

<1 Pre-existing Spike"

<1

<1 2.

Steam Generator Tube Rupture Coincident Spike *

<1

<1 Pre-existing Spike"

<1

<1 3.

Locked Rotor

• 0.20

.0.27 4.

Rod Ejection *"

Primary to Secondary Pathway 0.30 0.031 Containment Pathway 0,44 0.14

• Acceptance Criterion is 2.5 rem

• • Acceptance Criterion is 25 rem

• • • Acceptance Criterion is 6.25 rem

,.a.

22 -

TABLE 11 h

Control Room Deses from Postulated Accidents (Rem)

BWI Replacement Steam Generators Byron Station. Unit 1 Accident Thyroid Whole Body i

1.

Main Steamline Bresk A

Coincident Spike 2.3

<1 Pre-existing Spike 0.68

<1 2.

Steam Generator Tube Rupture.

Coincident Spike 0.15

<1 Pre-existing Spike 0.55

<1 3.

Locked Rotor 0.43 0.76 4.

Rod Ejection Primary to Secondary Pathway 0.31 0.11 Containment Pathway 17 0.45 i

• Acceptance criteria are 5 rem whole body and 30 rem thyroid.

l 4

a f

e

.I k'

J r

=,-y,

=

e

4, l

$ l TABLE 12

.Qsntrol Room Doses from Postulated Accidents (Rem)

DWI Replacement St9am Generators Braidwood Station. Unit 1 4

Accident Thyroid Whole Body 1.

Main Steamline Break Coincident Spike 1.8

<1 Pre-existing Spike 0.29

<1 2.

Steam Generator Tube Rupture Coincident Spike 0.11

<1 Pre-existing Spike 0.41

<1 1

3.

Locked Rotor 0.34 1.2 4.

Rod Ejection Primary to Secondary Pathway 0.24 0.16 Containment Pathway 10 0.69

• Acceptance criteria are 5 rem whole body and 30 rem thyroid.

i

.