ML20138J757
| ML20138J757 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 05/05/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20138J755 | List: |
| References | |
| NUDOCS 9705080328 | |
| Download: ML20138J757 (24) | |
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1 UNITED STATES NUCLEAR REGULATO;RY COMMISSION WASHINGTON, D.C. 30006 4001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIQN RELATED TO AMENDMENT NO. 175 TO FACILITY OPERATING LICENSE NPF-9 AND AMENDMENT NO. 157 TO FACILITY OPERATING LICENSE NPF-17 l
DUKE POWER CCli28My, MCGUIRE NUCLEAR STATION. UNITS 1 AND 2 DOCKET NOS. 50-369 MQ_E9 M
1.0 INTRODUCTION
By letter dated September 30, 1994, as supplemented by letters dated September 18, 1995, and March 15, April 29, Mty 16, September 23, and October 28, 1996, and January 16, April 22, and May 2, 1997. Duke Power j
Company (DPC or the licensee) submitted a request for changes to the McGuire i
Nuclear Station, Units 1 and 2, Technical Fpecifications (TS). The requested changes are related to the replacement of se current Westinghouse Model "0" type preheat steam generators (SGs) with.eedring steam generators designed by Babcock & Wilcox International. The March 15, April 29, May 16, September 23, and October 28, 1996, and January 16, April 22, and May 2, 1997, letters provided clarifying information that did not change the scope of the September 30, 1994, application and the initial proposed no significant hazards consideration determination.
I The major design differences in the feedring steam generator with respect to l
the preheat design include the following:
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There are approximately 2,000 more tubes of a slightly smaller diameter.
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The tube bundle is about 8 feet taller.
The SG liquid mass at full power is approximately 20,000 lbm greater.
The above steam generator design differences result in the following thermal-i hydraulic changes:
The total p'rimary system volume is increased by about 10%.
The effective. tube bundle heat transfer area is increased by approximately 60%.
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- The full power programmed T., for McGuire is reduced by about 3 'F.
2.0 EVALUATION The licensee stated that all the analyses were perfonned consistent with NRC-approved methodology. The following are a list of the applicable Duke Power I
Company analysis methodology topical reports:
b 9705000328 970505 ADOCK0500g9 DR
. DPC-NE-3000-P, Thermal-Hydraulic Transient Analysis Methodology, Duke Power Company, Oconee, McGuire and Catawba Nuclear Stations, Revision 1, SER dated December 1995.
(Approved by letter, R. Martin to M. Tuckman, December 27,1995)
DPC-NE-3001-PA, Multidimensional Reactor Transients and Safety Analysis Physics' Parameters Methodology, Duke Power Company, McGuire and Catawba Nuclear Stations, November 30, 1991.
DPC-NE-3002, FSAR Chapte'r 15 System Transient Analysis Methodology, Duke Power Company, McGuire and Catawba Nuclear Station.
(Revision 1 approved by letter, R. Martin to M. Tuckman, December 28, 1995; Ravision 2-approved by letter, R. Martin to M. Tuckman April 26,1996)
DPC-NE-3004-P, Mass and Energy Release and Containment Response Methodology, Duke Power Company, McGuire and Catawba Nuclear Stations.
(Approved by letter, R. Martin to M. Tuckman, September 6, 1995)
In addition, the steam generator tube rupture and loss-of-coolant accident (LOCA) reanalyses relied upon the following NRC-approved vendor topical l
reports:
WCAP-10698-PA, "SGTR Analysis Methodology to Determine the Margin to l
Steam Generator Overfill," Westinghouse Electric Corporation.
BAW-10168, "BWNT Loss-of-Coolant Accident Evaluation Model for Recirculat'ing Steam Generator Plants," Revision 2, Babcock & Wilcox.
(Approved by letter R. Jones to J. M. Taylor, August 22,1996)
BAW-10168, "BWNT Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants," Revision 3, Babcock & Wilcox.
l (Approved by letter, M. Virgilio to J. H. Taylor, June 15,1994) l BAW-10174A, " Mark-BW Reload LOCA Analysis for the McGuire and Catawba l
Units," Revision 1, Babcock & Wilcox.
In its application, the licensee concluded that the SG replacement would not result in radiological consequences which would exceed the guideline values presented in Standard Review Plans (SRPs) for the main streamline failure, steam generator tube rupture (SGTR), locked rotor accident (LRA) and the rod i
ejection accident.: The basis for this. conclusion was that radiological.
consequences of design basis accidents (DBAs) are bounded by the current licensing basis.
In performing dose analyses for the rod ejection and locked rotor events, a primary-to-secondary leak rate is calculated.
T.'e licensee initially calculated the primary-to-secondary leak rate by the staff-approved methodology. The licensee then deviated from this methodology when it adjusted the leak rate used by the dose analysis by scaling the technical specification initial leak rate vs. using the leak rate assumed in the initial
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calculation.
In its submittal of May 2, 1997, the licensee provided a justification of the method of performing this scaling for these specific i
analyses. The staff reviewed the licensee's justification for this scaling l
method and finds it is reasonable and provides acceptable leak flow values for these analyses only. The licensee did not calculate control room doses for i
these DBAs. The staff reviewed the licensee's analyses, as noted below, and i
. compared the potential radiological consequences to the current licensing basis and the acceptance criteria presented in NUREG-800 and General Design Criterion 19 (GDC-19) of Appendix A to 10 CFR Part 50.
The licensee reevaluated each Final Safety Analysis Report (FSAR) transient and accident analysis to determine the effects of the steam generator replacement, and to ensure that thermal-hydraulic performance will not be i
degraded. The staff's review of the licensee's reanalysis is delineated in sections 2.1 through 2.3.
2.1 Design Basis Transients and Accidents The following thermal-hydraulic system transients were reanalyzed in order to ensure that the acceptance criteria continued to be met with the feedring steam generators. The licensee stated that all analyses were performed consistent with NRC-approved methodologies.
A. Mass and energy release for postulated loss-of-coolant accidents inside containment (FSAR Section 6.2.1.3)
This event.was reanalyzed to ensure that the peak containment pressure limit is not exceeded. Since the Reactor Coolant System (RCS) volume will be greater, the total mass released into containment will be greater.
In addition, during the depressurization of the RCS, the steam generators actually function as heat sources.
Since the feedring steam generator full-power liquid mass is greater than that of the Model D steam generators, the total energy available for removal by the RCS is increased. Both of these effects have the potential to yield more severe mass and energy release results. The analysis was performed in accordance with the analytical model and methodology described in topical report DPC-NE-3004. Although the containment pressure would be increased using the feedring steam generator, the peak lower containinent pressure calculated for Units 1 and 2 was 12.48 psig, which meets the 15 psig acceptance criterion. Therefore, the reanalysis results for this transient are acceptable.
B. Mass and energy release for postulated secondary system pipe ruptures inside containment (FSAR Section 6.2.1.4)
This event was reanalyzed to ensure that the peak containment temperature limit is not exceeded. Steam generator tube bundle uncovery is important since this initiates the release of superheated steam into containment.
Since the feedring steam generator design has tubes that are significantly taller than those in the Model D steam generators, the potential exists for earlier bundle uncovery. The analysis was performed in accordance with the analytical model and methodology described in topical report DPC-NE-3004. The peak temperature in the break compartment is calculated to be 313 *F, which does not exceed the i
environmental qualification (EQ) limit of 340 *F and is therefore l
acceptable.
1
. j C. Feedwater system malfunction causing an increase in feedwater flow (FSAR Section 15.1.2)
This event was reanalyzed to ensure that departure from nucleate boiling I
(DNB) does not occur. The transient involves an increase in core power resulting from an overcooling by the secondary system. The impact of the increased heat transfer area of the feedring steam generator was evaluated in this reanalysis. The reanalysis was performed in accordance with the analytical model and methodology described in topical reports DPC-NE-3000 and DPC-NE-3002. The minimum DNB ratio is calculated to be 1.86, which is well above the 1.55 design limit and is therefore acceptable.
D. Excessive increase in secondary steam flow (FSAR Section 15.1.3)
This event was reanalyzed to ensure that DNB does not occur. The i
transient involves an increase in core power resulting from an l
overcooling by the secondary system. T1e impact of the increased heat transfer area of the feedring steam generator was evaluated in this reanalysis.
The reanalysis was performed in accordance with the analytical model and methodology described in topical reports DPC-NE-3000 and DPC-3002. The transient analysis results were found to be acceptable as the licensee found that the reactor reached an equilibrium condition that did not challenge the overpressure delta temperature or overtemperature delta temperature reactor trip functions, which are designed to protect the core against DNB. Therefore, an explicit DNB ratio calculation was not performed.
E. Inadvertent opening of a steam generator relief or safety valve (FSAR 15.1.4)
This event was reanalyzed to confirm that the transient response is bounded by the DNB calculation performed for the steam system piping failure event. The analysis was performed in accordance with the analytical model and methodology described in topical reports DPC-NE-3000, DPC-NE-3001 and DPC-NE-3002. The peak thermal power level reached was less than half of that resulting from the steam system piping failure
. event and thus the DNB ratio will be higher than that calculated for a steam system piping failure event, as further discussed in item F.
An explicit DNB ratio calculation was not performed.
F. Steam system piping failure (FSAR Section 15.1.5)
This transient was reanalyzed to show that DNB does not occur and the radiological consequences are withi,n the limits established by SRP Section 15.1.5.
l This transient involves an increase in core power resulting from 1
i overcooling by the secondary system. The impact of the increased heat i
transfer area of the feedring steam generator is evaluated in this reanalysis. The analysis was performed in accordance with the analytical model and methodology described in topical reports DPC-NE-3000 and i
i DPC-NE-3001 The minimum DN8 ratio was calculated to be 1.78, which is
(
i
. well above the 1.45 design limit for this transient, and therefore it is acceptable.,
Both the staff and the licensee have evaluated the radiological consequences of a steam system piping (also referred to as main steam-line) failure using SRP (NUREG-0800) Section 15.1.5, Appendix A,
" Radiological Consequences of Main Steam Line Failure Outside Containment (PWR)." The licensee has submitted analyses of this accident that show no additional fuel failures attributed to the accident. The staff has independently calculated the doses from this type of accident. The results of the staff's calculations confirm the licensee's conclusion that the doses would be within the limits established by SRP 15.1.5,-
Appendix A.
The assumptions' used in calculating the doses are listed in Table 1 (attached), and the resulting calculated dose values are in Table 2 (attached).
The staff concludes that the radiological consequences of this accident are a small fraction of 10 CFR Part 100 for the~ accident initiated spike case, within Part 100 for the pre-existing spike case, and less than GDC-19 for both spike cases.
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G. Turbine trip (FSAR Section 15.2.3) l This event was reanalyzed to show that peak primary and secondary system pressures do not exceed the applicable limits. The increased heat transfer area of the feedring steam generator improves the ability of the secondary system to remove primary system heat and, therefore, potentially results in a more severe secondary side pressurization. The reanalysis was performed in accordance with the analytical model and methodology described in topical reports DPC-NE-3000 and DPC-NE-3002.
The peak primary and secondary pressures were calculated to be 2,674.3 and 1,285.7 psig, respectively.
This is acceptable as the corresponding acceptance criteria are 2,733.5 psig (110% of 2,485 psig, design pressure of the reactor coolant side) and 1,303.5 psig (110% of 1,185 psig, design pressure of the steam side).
H. Loss of nonemergency AC power to the station auxiliaries (FSAR Section 15.2.6)
This event was reanalyzed to demonstrate the adequacy of the natural circulation cooling in the modified reactor coolant loop configuration.
i The reanalysis was performed in accordance with the analytical model and methodology described in topical reports DPC-NE-3000 and DPC-NE-3002.
The results of this reanalysis were determined acceptable as they show that, within 10 minutes of the loss of offsite power, a stable natural circulation flow rate of approximately 5 percent of the full-power value is established with a core temperature difference (AT) of less that 30 *F.
The results of the reanalysis indicate that this event is bounded by turbine trip and is therefore acceptable.
l o I. Loss of normal feedwater flow (FSAR Section 15.2.7)
This event was reanalyzed to confirm that the transient response is bounded by the turbine trip event.
The reanalysis was performed in accordance with the analytical model and methodology described in topical I
reports DPC-NE-3000 and DPC-NE-3002. The position presented in DPC-NE-3002, that this transient is bounded by turbine trip, remains valid.
J. Feedwater system pipe break (FSAR Section 15.2.8) l This event was reanalyzed to demonstrate the capability of the degraded secondary system to effectively cool the reactor core. While the increase in heat transfer area would tend to improve the transient l
results, other factors, such as the restrictor, necessitate the i
reanalysis. The reanalysis was performed in accordance with the analytical model and methodology described in topical reports DPC-NE-3000 and DPC-NE-3002. The adequacy of the long-term cooling capability was I
verified by the prevention of hot leg boiling. This was found to be accestable as a minimum subcooling of 17.5 'F was reached during the over1 eating phase of the feedline break transient. The RCS peak pressure was within the 110 percent of RCS design pressure which meets the acceptance criteria.
K. Reactor coolant pump shaft seizure-locked rotor (FSAR Section 15.3.3)
This event was reanalyzed to show that the peak primary system pressure does not exceed the applicable limit and to determine the percentage of fuel rods that experience DNB. The results of the transient reanalysis were determined to be insensitive to the secondary system. The analysis l
was performed in accordance with the analytical model and methodology described in topical reports DPC-NE-3000 and DPC-NE-3002. Therefore, because the secondary system does not impact the transient, the analysis results are acceptable for this transient.
The licensee evaluated the offsite doses resulting from a postulated reactor coolant pump locked rotor accident (LRA) for the replacement steam generators.
In accordance with the SRP, the reactor is initially assumed to be operating at 102 percent of the licensed power level. A turbine trip and coincident loss of offsite power (LOOP) are incorporated l
into the analysis, which is consistent with the conditions given in Section 15.3.3-15.3.4 of the SRP.
l With the assumed loss of offsite power, releases are through the steam generator relief valves (PORVs and safety valves), i.e., no credit is taken for iodine partitioning in the condenser. During periods of SG tube bundle uncovery, the licensee assumed that all primary-to-secondary leakage will escape the SG with no partitioning, plateout or scrubbing of the iodine source term. As noted in Section 2.0 above, the TS value for primary-to-secondary leakage was not assumed in the licensee's calculation. The leakage was based upon the results the licensee obtained from a RETRAN analysis on the new SGs. The staff utilized these same RETRAN generated time-dependent primary-to-secondary leak rates in our dose analysis.
l l
. In calculating the consequences of this postulated event, the staff assumed that the LRA with the LOOP causes 15% of the fuel pins to enter departure from nuclear boiling (DNB) and breach, thereby releasing the fuel pin activity. Additional assumptions are listed in Table 3 (attached) based upon the Appendix to Section 15.4.9 of the SRP. The resulting doses are shown in Table 4 (attached).
The staff confirmed that the doses resulting from fuel failure would be less than Part 100 guidelines for offsite locations and within GDC 19 guidelines for the control room operators.
For releases associated with an accident initiated spike with no fuel failure the staff confirmed that the offsite doses would be a small fraction of Part 100.
L. Steam generator tube rupture (SGTR) (FSAR Section 15.6.3)
The SGTR event was reanalyzed to show that (a) DNB does not occur, (b) the calculated offsite doses do not exceed the acceptance criteria, and (c) steam generator overfill is avoided. When the ruptured area of the tube is no longer covered with liquid, a significant reduction in the iodine partition factor occurs; therefore, tube bundle recovery is an important phenomenon in the offsite analysis. Since the feedring steam I
generator design has tubes that are significantly taller than those in the Model D steam generators, the potential exists for a longer period of tube bundle uncovery, which has a negative impact on the offsite doses.
l Other factors associated with the feedring steam generator, which l
potentially impact the transient results, are the reduced tube diameter and the revised SG 1evel setpoints. The reanalysis was performed in accordance with the analytical model and methodology described in topical reports DPC-NE-3000, DPC-NE-3002, and WCAP-10698-PA.
I The minimum DNB ratio was calculated to be 1.93, which is well above the 1.55 design limit. Also, the time to overfill for the replacement generators was determined to be bounded by that for the Model D steam generators, which has been previously found to be acceptable.
i For the SGTR dose calculation, the staff used information provided by the licensee in several submittals listed above.
In accordance with SRP 15.6.3, two assessments were performed for the most limiting scenario, which is an SGTR with a loss of offsite power and fully stuck open atmospheric dump valve. The assessments included an accident-initiated iodine spike and a preexisting iodine spike.
For the accident-initiated iodine spike case, the doses at the EAB and LPZ are 10 percent of the guideline values of 10 CFR Part 100 and less than GDC-19. For the preexisting iodine spike case, the staff's calculations indicate that thyroid doses are within the acceptance criteria presented in SRP 6.4 and 15.6.3.
The assumptions used in calculating the doses are listed in Table 5 (attached), and the resulting calculated dose values are in i
Table 6 (attached). The staff concludes that the radiological consequences of an SGTR accident with a loss of offsite power and fully stuck open atmospheric dump valve are acceptable.
1 e
y
. M. Loss-of-coolant accidents (LOCA) (FSAR Section 15.6.5)
The licensee stated that a LOCA analysis, applicable to McGuire Units 1 and 2, was performed by B&W Nuclear Technologies (BWNT).
The analysis supports operation of the Duke Power units with the feedring steam generators. - The methodology employed in the analysis is in accordance with 10 CFR 50.46 and 10 CFR Part 50 Appendix K and is documented in l
topical reports BAW-10174, Revision 1 (erroneously stated as Revision 2 in the September 30, 1994, submittal but corrected in the March 15, 1996, submittal), and BAW-101068P, Revision 3.
The LOCA evaluation considered both large and small breaks.
l The maximum cladding temperature for a large break was calculated to be l
2,075 "F, which is less than the acceptance criteria limit of 2,200 *F of 10 CFR 50.46. The maximum local metal-water reaction was found to be 5.9 percent, which is well below the embrittlement limit of 17 percent as specified by 10 CFR 50.46. The total core metal-water reaction was calculated to be 0.61 percent as compared with the 1 percent criterion of 10 CFR 50.46, and the cladding temperature transient was terminated at a l
time when the core geometry is still amenable to cooling. As a result, the core temperature will continue to drop and the ability to remove decay heat generated in the fuel for an extended period of time will be l
provided. These results were found to be acceptable.
l I
The maximum calculated peak cladding temperature for the small breaks l
analyzed was found to be 1,200 *F.
The licensee uses the methodology l
delineated in the topical report BAW-10168, Revision 2, "BWNT Loss-of-l Coolant Accident Evaluation Model for Recirculation Steam Generator Plants." This result is acceptable as it is well below the acceptance criteria limits of 10 CFR 50.46 and there is no case that is limiting when compared to the results presented for the large breaks.
N. Affects to the " doghouse" from a postulated secondary system pipe rupture i
outside containment This event was reanalyzed to ensure that the " doghouse" (a structure that l
contains the safety valves for the main r. team and main feedwater lines) equipment qualification temperature limit is not exceeded. A key phenomenon in this reanalysis is tube bundle uncovery, since this initiates the release of superheated steam. Since the feedring steam generator design has tubes that are significantly taller than those in the Model D steam generators, the potential exists for earlier bundle i
uncovery. The reanalysis was performed ir accordance with the approved I
analytical model and methodology described in topical report DPC-NE-3004.
A comparison of the mass and energy release rates showed that the reanalyzed transient results were bounded by the existing vendor-supplied analysis and we,re acceptable. Therefore, an explicit " doghouse" temperature response calculation was not performed by the licensee.
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i 2.2 Other Transients Reanalyzed For Offsite Dose Only The radiological consequences were reanalyzed by the' licensee for the following events due to the differences in the thermal-hydraulic parameter of the feedring steam generator. Specifically, since the feedring generator design has tubes that are significantly taller than those in the Model D steam generators, the potential exists for a longer period of tube bundle uncovery, which negatively impacts the offsite dose calculations. The two events listed are FSAR Chapter 15 transients for which fuel failure are postulated to occur.
A. Single rod cluster control assembly (RCCA) withdrawal (FSAR Section 15.4.3.d)
The offsite dose results at Exclusion Area Boundary (EAB) were found not to exceed a small fraction (10 percent) of the 10 CFR Part 100 limits and are therefore acceptable.
B. Spectrum of RCCA ejection accidents (FSAR Section 15.4.8)
The staff evaluated the consequences of a rod ejection accident.
Ejection of a rod results in rapid reactivity insertion. A single case was evaluated, including both failed fuel' source term and release from the secondary side, and failed fuel source term and release from the containment. The licensee has calculated and assumed that 50 percent of the fuel elements will experience cladding failure, releasing all of its gap radioactivity. The released radioactivity is assumed to be mixed immediately with the primary coolant. When the tube bundles are covered, all primary-to-secondary leakage is assumed to be mixed with the liquid in each steam generator. For releases via the secondary path during periods of tube bundle uncovery, primary-to-secondary leakage is assumed l
to be released directly to the environment. Activity release to the l
environment may occur via each of two pathways. The activity from I
cladding failure in the first pathway involves a release of activity to the primary coolant which is then assumed to leak to the atmosphere as in the DBA LOCA (Table 7 attached).
In the second pathway, activity is transferred to the secondary coolant via primary-to-secondary leakage rate (Table 8 attached). With loss of offsite power and subsequent steam i
venting, some of the iodine transferred to the shell side is available l
for leakage to the environment.
In calculating the consequences of this postulated event, the staff calculated the doses as if all activity was released via each of the above pathways. The assumptions used by the staff for the radiological consequences following a postulated control rod ejection accident are listed in Tables 7 and 8 (attached) and the calculated doses are listed in Tables g and 10 (attached). The staff concludes that the radiological consequences of rod ejection accidents are well within the acceptance criteria of 10 CFR Part 100 and less than GDC-19.
2.3 Transients Not Reanalyzed i
j For the fo11 ming transients the licensee indicated and the staff agreed that reanalysis we. not required, since either (a) the analysis is unaffected by l
the steam generator replacement,-(b) any changes will not adversely impact the
. analysis result, or (c) the transient is bounded by a more limiting transient, which is being reanalyzed.
A. Peak Reverse Differential Pressure, Containment Subcompartment & Minimum Containment Pressure Analyses (FSAR Sections 6.2.1.1, 6.2.1.2, and 6.2.1.5)
B. Feedwater system malfunction causing a reduction in feedwater temperature (FSAR Section 15.1.1)
C. Loss of external load (FSAR Section 15.2.2)
D. Inadvertent closure of main steam isolation valves (FSAR Section 15.2.4)
E. Partial loss of forced reactor coolant flow (FSAR Section 15.3.1)
F. Complete loss of forced reactor coolant flow (FSAR Section 15.3.2)
G. Reactor coolant pump shaft break (FSAR Section 15.3.4)
H. Uncontrolled RCCA bank withdrawal from a subcritical or low power startup condition (FSAR Section 15.4.1) l
- 1. Uncontrolled RCCA bank withdrawal at power (FSAR Section 15.4.2)
J. RCCA misoperation (FSAR Section 15.4.3) l K. Startup of an inactive reactor coolant pump at an incorrect temperature (FSARSection15.4.4)
L. Chemical volume control system (CVCS) malfunction that results in a decrease in boron concentration in the reactor coolant (FSAR Section 15.4.6)
M. Spectrum of RCCA ejection accidents (FSAR Section 15.4.8)
N. Inadvertent operation of emergency core cooling system (ECCS) during power operation (FSAR Section 15.5.1)
O. Inadvertent opening of a pressurizer safety or relief valve (FSAR Section 15.6.1) 2.4 Technical Specification Changes The major Technical Specification changes proposed by the licensee, and the staff's evaluation of these changes, are delineated below. All of the changes are found by the staff to be consistent with all the reanalyzed licensing basis safety analyses-(see previous discussions), and are therefore acceptable. Other TS changes pertain to requirements that are no longer applicable after the replacement of the steam generators, or are administrative in nature, reflecting the use of the most current revision of the topical reports.
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i l Proposed Revision to TS Section 3.4.5 and Associated Bases, Table 4.4-1
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The licensee proposed to change the surveillance requirements to delete repair methods that are no longer applicable after the replacement of the steam generators. Reference to F*, sleeving, and interim plugging criteria, or alternate plugging criteria are deleted. The F* criterion relates to the determination of the minimum acceptable length of defect-free SG tubes within the tube sheet to establish structural integrity.
In addition, clarification is added to the surveillance requirements on performing initial inspections after replacement of the steam generators and when they will be performed.
Evaluation This proposed change to the TS deletes repair criteria that will no longer be applicable after the replacement of the SGs.
References to F*, sleeving, and interim plugging criteria, or alternate plugging criteria are removed because these methods of repair were approved specifically for use on the original SGs. These. changes will not alter the way surveillances are performed, and will entinue to meet the current intent of the requirements.
Since the repair methods and alternate criteria are no longer necessary, then the language in the TS and the TS Bases are obviously no longer necessary or need to be changed. The staff finds these changes acceptable.
i Proposed Revision to Tables 2.2-1 and 3.3-4
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The licensee proposed to change the low-low SG water level reactor trip setpoint from a variable setpoint that is proportional to nuclear power to a constant level setpoint of 16.7 percent of narrow range span.
The licensee also proposed to change the T, value from 588.2 'F to 585.1 'F.
Evaluation The variable setpoints were chosen to maximize the plant operating range while still ensuring that reactor trip on low-low level would occur following a feedline break inside containment. The new low-low level setpoints are consistent with reanalyzed licensing basis safety analyses. All of the reanalyzed transients that take credit for this trip function meet the applicable acceptance criteria, as previously stated.
The feedwater system pipe break was analyzed to demonstrate the capability of the degraded secondary system to effectively cool the core. The analysis was performed in accordance with NRC-approved methodology as described in topical reports DPC-NE-3000 and DPC-NE-3002. The staff has reviewed the impact of the design differences between the original preheat SGs and the replacement feedring SGs on structural modeling and analysis results. The feedwater l
system pipe break transient is significantly impacted by implementation of the feedring SGs, and requires several major assumption changes as a direct result of the design and location of the main feedwater nozzles. The licensee's l
discussion of assumption changes and the impact of changes in transient 1
results was reviewed and found to be reasonable and acceptable.
l For this accident, the loss of offsite power coincident with reactor trip is assumed, resulting in reactor coolant pump trip and delay in the startup of l
l
. the diesel generators for safety injection. Early main steam isolation valve (MSIV) closure was determined to be conservative in terms of earlier faulted SG dryout. Thus, in the revised assumptions, MSIV closure occurs coincident with turbine trip, which occurs on loss of offsite power. Therefore, the staff concludes that the licensee's approach to the analysis of this event is acceptable.
The licensee also proposed to change T,Ticensee stated that this new T value from 588.2 *F to 585.1 *F.
In its September 18, 1995, submittal, the was chosen based on returning the secondary side steam pressure to the original value after replacement of the steam generators, and that the new T w
assumed value for nominal full power in all applicable safety analy,s7s.as the These analyses were evaluated in Section 2.1 of this safety evaluation. Therefore, 1
the staff finds the change acceptable.
Proposed Revision to Table 3.3-4
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The licensee proposed to change the high-high SG water level setpoint for turbine trip and feedwater isolation. This is changed to 83.9 percent of narrow-range span.
l Evaluation The high-high setpoint was chosen to maximize the plant operating region while still ensuring that feedwater isolation on high-high level would occur before the actual water level in the SG reached the flood point.
The.new high-high level setpoint is consistent with all reanalyzed licensing basis safety analyses. The results of the increase in feedwater flow analysis, which is the only FSAR Chapter 15 transient that relies on this trip function, show that all applicable acceptance criteria are met. DPC Topical Report DPC-NE-3002, Revision 1, describes the DPC transient analysis methodology. The impact of the design differences between the feedring type replacement SGs and the original preheat SG on the feedwater flow analyses has been reviewed and approved by the staff as discussed earlier.
Proposed Revision to 3/4.4.6.2
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The licensee proposed to reduce the allowable primary-to-secondary system leakage to a total of 0.27 gpm through all steam generators and 135 gallons per day through any one generator.
Evaluation The leakage limit of 135 gallons per day is more stringent than the limit of 500 gallons per day, as required by current McGuire TS, and, therefore, is acceptable. The proposed change to a lower primary-to-secondary leakage limit for operation with the replacement steam generators will require that corrective action be taken more quickly in the event of steam generator i
leakage.
Proposed Revision to Section 5.4.2
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The licensee proposed to change the volume of the reactor coolant system (RCS) j from 12,040 i 100 cubic feet to 13,050 i 100 cubic feet.
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j i The mass and energy release for postulated loss-of-coolant accidents (LOCAs) l inside containment is analyzed to ensure that the peak containment pressure limit is not exceeded. Since the RCS volume is greater, the total mass released into containment will be greater.
In addition, during the depressurization of the RCS, the SGs actually function as heat sources.
Since the feedring SG full-power liquid mass is greater than that of the Model D E
preheat SGs, the total energy available for removal by the RCS is increased.
Both of these effects have the potential to yield more severe mass and energy release results. This event has been reanalyzed by the licensee, and shown to meet current acceptable limits relative to pressure, temperature, and stress allowables for the affected components (see also Section 2.1.A of this safety i
evaluation).
t Topical Report DPC-NE-3004-P describes DPC's methodology for simulating the mass and energy release and containment response to LOCAs and main steamline break accidents for the McGuire facility. The staff has previously reviewed and approved the analytical methodology including the thermal-hydraulic l
computer codes for predicting McGuire containment pressure and temperature l
responses to design-basis accidents. The staff also reviewed the impact of the design differences between the existing preheat SGs and the replacement feedring SGs on the analysis results. The peak lower containment pressures calculated for McGuire Units 1 and 2 were 10.29 and 12.48 psig, which meet the 15 psig acceptance criterion. Various break locations were analyzed and it was determined that the cold leg RCP discharge leg break case generates the l
highest energy release and the limiting large-break LOCA peak pressure. With l
feedwater flow and unaffected SGs promptly isolated, steamline break mass and energy releases are considerably less than those of the limiting LOCA.
However, since the blowdown fluid is superheated, the licensee has examined l
the lower containment temperature response to ensure that environmental qualification limits. for safety-related equipment are not exceeded.
The steamline break inside containment is limiting from considerations of environmental qualification of equipment structures and components in the lower containment. The peak temperature in the break compartment is calculated to be 313 'F with the original SGs. This is considered to be the bounding temperature and encompasses the design changes in the replacement feedring SGs. This calculated value is well below the environmental qualification (EQ) limit of 340 'F and therefore is acceptable (see also Section 2.1.B of this safety evaluation).
Proposed Revision to Section 6.9.1.9 The licensee proposed to change the references to topical reports DPC-NE-3000 i
and DPC-NE-3002 to reflect the use of the most current approved revisions.
Evaluation l
This change is administrative in nature in that the references are being updated to reflect the use of the most current approved topical reports, and is therefore, acceptable.
~
l.
3.0 STAFF CONCLUSION The staff has reviewed the licensee's submittal to support changes to the McGuire Technical Specifications affected by the replacement of the. steam i
generators and has found them to be acceptable as discussed in Section 2.0,
4.0 STATE CONSULTATION
1 In accordance with the Commission's regulations, the North Carolina State official was notified of the proposed issuance of the amendments. The State official had no comments.
5.0 MyIRONMENTAL CONSIDERATION The amendments change requirements with respect to installation or use of a 1
facility component located within the restricted area as 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 occupational radiation exposure. The' Comission has previously issued a proposed finding that the amendments involve no significant hazards i
consideration, and there has been no public coment on such finding (60 FR 56366 dated November 8, 1995). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR I
51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
6.0 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 Comission's regulations, and (3) the issuance of the amendments will not be inimical to the comon defense and security' or to the health and safety of the public.
Attachment:
Tables I through 10 Principal Contributors:
H. Balukjian H. Conrad J. Rajan J. Minns J. Hayes V. Nerses Date:
May 5, 1997
i TABLE 1 INPUT PARAMETERS FOR MCGUIRE UNITS 1 AND 2 EVALUATION OF A MAIN STEAM LINE FAILURE
- Power (MWt) 3479 1
Accident initiated spike case, DEI in primary coolant 1.0 pCi/g (case 1)
Preexisting spike case, DEI in primary coolant 60.0 pCi/g (case 2)
Primary-to-Secondary leak rate, as limited by TS 0.27 gpm Source spike factor after accident 500 Iodine released from affected steam generator within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Initial steam release from defective S/G (lbs) 199,600 Primary coolant steam mass (lbs) 582,456 i
Secondary coolant steam mass (lbs) 144,500
]
Range of flashing fractions s; 1 l
Atmospheric Dispersion Factors EAB (0-2 hours) 6.6 x 10
LPZ (0-8 hours) 2.5 x 10'5 j
Control Room (0-8 hours) 2.0 x 10'3 Control room parameters are listed in Table 3.
]
l
- All releases throuah secondary system / Dose Eauivalent lodine (DEI)
A-1
TABLE 2 CALCULATED DOSES FOR MCGUIRE UNITS 1 AND 2 MAIN STEAM LINE FAILURE Thyroid Thyraid LOCATION Pre-accident Accident Spike Initiated Spike Whole Body Exclusion Area Boundary 3.2 rem
- 1.6 rem ***
<1 rem **/****
Low Population Zone 0.5 rem
- 1.0 rem ***
<1 rem **/****
Control Room 0.6 rem *****
1.1 rem *****
<1 rem ******
- 10 CFR 100 Acceptance Criteria - 30 rem
- 10 CFR 100 Acceptance Criteria Whole Body - 2.5 rem
- 10 CFR 100 Acceptance Criteria = 300 rem
- 10 CFR 100 Acceptance Criteria - 25 rem
A-2
]
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l-TABLE 3 INPUT PARAMETERS FOR MCGUIRE UNITS 1 AND 2 EVALUATION OF A LOCKED ROTOR ACCIDENT l
Power Level (MWt) 3479 Percent of rods perforated 15 Percent of rods melted 0
Number of rods in core 50595 Peaking factor 1.65 Reactor Coolant volume (gal) 83,059 Control Room Parameters (see Table 3)
Atmospheric Dispersion EAB (0-2 hours) 6.6 x 10
LPZ (0-8 hours) 2.6 x 10-5 Control room 2.0 x 10'3 r
Steaming rates 180,000 lbs (0-1800 seconds) 115,000 lbs (1800-3600 seconds) l 110,000 lbs (3600-5400 seconds) 125,000 lbs (5400-7200 seconds) 170,000 lbs (7200-9792 seconds)
Tube Bundle uncovery 1 s/g (278-1143 seconds) 2 s/g (420-1030 seconds) 1 4
I A-3 I
1 j
TABLE 4 l,
CALCULATED DOSES FOR MCGUIRE UNIT 1 AND 2 LOCKED ROTOR ACCIDENT LOCATION Thyroid Whole Body l
Exclusion Area Boundary 41.7 rem
- 0.3 rem **
Low Population Zone 2.2 rem
- 0.017 rem **
Control Room 2.6 rem ***
0.025 rem ****
- 10 CFR 100 Acceptance Criteria - 300 rem for fuel failures
- 10 CFR 100 Acceptance Criteria Whole Body - 25 rem for fuel failures
l I
A-4
l a
TABLE 5 INPUT PARAMETERS FOR MCGUIRE UNITS 1 AND 2 EVALUATION OF A STEAM GENERATOR TUBE RUPTURE ACCIDENT 1.
Power level (MWt) 3479 2.
Primary coolant concentration of dose equivalent '3'I l
Pre-existina Soike Value Cifuci) 960nci/a i
- I = 46.4 132 I - 16.7
'33
- I - 74.0 I - 11.10 i
- I - 40.7 Accident Initiated Soike value (uCi/a) Oluci/a
- I
.077 132 1 =.028
'33*1
.123 l
- I
.018 I
.068 3
3.
Primary Coolant Volume (ft )
13,050 Primary Coolant Mass (lbs) 582,456 Steam released intact S/Gs (1bs) (2-8 hr) 959,000 Steam Temperature ( F) "F) 525 Feedwater Temperature (
436 4.
Letdown flow rate (gpm) 75 l
- 5. TS limits for DE *{in the primary and secondary coolant Primary coolant DE concentration (pCi/g) 1.0 SecondarycoolantDE'3{I(pCi/g) 0.1 6.
Steam releases (lbs)
Faulted SG:
(0-2 hr) 210,415 Intact SG:
(0-2 hr) 201,000 7.Releap31e rate for 1.0 pCi/g of do.ie equivalent *I (Ci/hr) 1-9.96 152*1-25.93
- I=
26.74 i
- I-41.81 I-27.72 i
3 j
- 8. Atmospheric dispersion factor (s/m )
Exclusion Area Boundary (0-2 hrs) 6.6 x 103 l
Low Population Zone (0-8 hrs 2.6 x 10' 1.0x10)3 Control Room 9.
Control room parameters 3
Volume (ft )
116,000 Pressurization rate (cfm) 1800 Makeup efficiency (%)
99 Unfiltered inleakage (cfm) 10 Occupancy factor 1
Iodine Protection Factor 65 l
A-5 I
l
i l
l TABLE 6 CALCULATED DOSES FOR NCGUIRE UNITS 1 AND 2 STEAM GENERATOR TUBE RUPTURE ACCIDENT l
Thyroid Thyroid LOCATION Pre-accident Accident Spike Initiated Spike Whole Body Exclusion Area Boundary 41 rem
- 29 ***
0.02 rem **
Low Population Zone 1.6 rem
- 1.2 rem ***
0.02 rem ****
Control Room 1.9 rem *****
1.4 rem *****
<0.1 rem ******
- 10 CFR 100 Acceptance Criteria - 300 rem
- 10 CFR 100 Acceptance Criteria Whole Body - 25 rem
- 10 CFR 100 Acceptance Criteria - 30 rem
- 10 CFR 100 Acceptance Criteria - 2.5 rem
A-6
1 TABLE 7 INPUT PARANETER FOR NCGUIRE UNIT 1 AND 2 EVALUATION OF ROD EJECTION ACCIDENT PRIMARY RELEASES Power level (MWt) 3479 Peaking factor 1.65 Containment leak rate (%/ day) 3 Containment bypass leakage (%)
7 Filtered leakage fraction (%)
93 Iodine split Elemental (%)
75 Organic (%)
25 Particulate (%)
0 Filter efficiencies Elemental, organic and particulate (%)
95 Percent of iodine and noble gas activity 10 in gap of failed fuel Fraction of rods perforated 0.5 Atmospheric Dispersion factors CAB (0-2 hours)
- 9. 5 x 10
LPZ (0-8 hours) 2.6 x 10'5 Control room 2.0 x 10'3 Tube Bundle Uncovery -
I s/g (69-1426 Seconds)
I s/g (69-1030 Seconds) 2 s/g (69-1008 Seconds)
A-7
i TABLE 8 INPUT PARANETER FOR MCGUIRE UNIT 1 AND 2 EVALUATION OF R0D EJECTION ACCIDENT SECONDARY RELEASES Power level (MWt) 3479 Peaking factor 1.65 Fraction of Rods perforated 0.5 Primary Coolant Volume (gal) 83,059 Percent of iodine and noble gas activity 10 in gap of failed fuel Atmospheric Dispersion factors (see Table 3)
Tube Bundle Uncovery 1426 Seconds (1 s/g) 69-1030 Seconds (1 s/g) 69-1008 Seconds (2 s/g)
Steaming Rates 180,000 lbs (0-1800 seconds) 85,000 lbs (1800-3600 seconds) 70,000 lbs (3600-5400 seconds) 40,000 lbs (7200-8100 seconds)
A-8
^
TABLE 9 CALCULATED DOSES FOR MCGUIRE UNITS 1 AND 2 ROD EJECTION ACCIDENT PRIMARY RELEASES LOCATION Thyroid Whole Body 4
Exclusion Area Boundary 38.7 rem
- 0.18 rem **
Low Population Zone 13.2 rem
- 0.01 rem **
Control Room 21.3 rem ***
0.02 rem ****
- 10 CFR 100 Acceptance Criteria = 25% of Part 100 - 75 rem
- 10 CFR 100 Acceptance Criteria Whole Body - 25% Part 100 - 6.25 rem
l
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TABLE 10 i
CALCULATED DOSES FOR MCGUIRE UNITS 1 AND 2 ROD EJECTION ACCIDENT SECONDARY RELEASES LOCATION Thyroid Whole Body
]
Exclusion Area Boundary 23.1 rem
- 0.17 rem **
Low Population Zone 1.0 rem **
0.001 rem **
Control Room 1.2 rem ***
0.024 rem ****
1 l
- 10 CFR 100 Acceptance Criteria - 25% of Part 100 - 75 rem
- 10 CFR 100 Acceptance Criteria Whole Body - 25% Part 100 - 6.25 rem
i i
i 4
- 4 A-10