ML20137Y891
| ML20137Y891 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 04/21/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20137Y890 | List: |
| References | |
| NUDOCS 9704230271 | |
| Download: ML20137Y891 (16) | |
Text
'
g UNITED STATES
- s j
NUCLEAR REGULATORY COMMISSION
' WASHINGTON, D.C. 20006 4001
'49*****
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 223 TO FACILITY OPERATING LICENSE NO. DPR-77 AND AMENDMENT NO. 214 TO. FACILITY OPERATING LICENSE NO. DPR-79 TENNESSEE VALLEY AUTHORITY SE000YAH NUCLEAR PLANT. UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328
1.0 INTRODUCTION
By letter dated April 4,1996 (Reference 1), as. supplemented by letters dated -
January 10., February 7, February 13, March 17, March 19, March 20, March 25 April 1, April 6, April 10, April 11, and April 18, 1997, the Tennessee Valley Authority (TVA, the licensee), requested amendments pursuant to 10 CFR 50.90 to Facility Operating License No. DPR-77 and No.' DPR-79 at Sequoyah Nuclear Plant (SQN), Units 1 and 2, respectively. The proposed amendments would change the SQN Technical Specifications (TSs) to allow for the conversion from Westinghouse fuel to Framatome Cogema Fuel-(FCF, Framatome), designated Mark-BW. The planned fuel conversion would begin with fuel cycle 9 for each unit.
The amendments would revise the SQN TS and their Bases to reflect the fuel design and vendor change. The licensee's evaluation was contained in a Topical Report BAW-10220P, " Mark-BW Fuel Assembly Application for Sequoyah Nuclear Units 1 and 2," (Reference 25).
The January 10, February 7, February 13, March 17, March 19, March 20, March 25, April 1,. April 6, April 10, April 11, and April 18, 1997, letters provided clarifying information that did not change the scope of the April 4,1996, application and the initial proposed no significant hazards consideration determination.
2.0 BACKGROUND
The. proposed new fuel vendor, FCF, previously Babco'ck and Wilcox (B&W), has provided fuel for pressurized-water reactors for many years, and B&W has specifically provided fuel for Westinghouse designed nuclear steam supply systems (NSSS).
The fuel design that is going to be used in the SQN cores is the FCF MARK-BW fuel design. The Trojan, Catawba and McGuire reactors have all had fuel supplied by B&W/Framatome with NRC staff approval.
The staff has previously completed a review effort specifically intended to allow both fuel design and methodology conversion for utilities with Westinghouse NSSSs. All of the methodologies and topical reports that are used to support the fuel and 9704230271 970421 PDR ADOCK 05000327 P
PDR n
e t
methodology change have been reviewed and approved by the staff under different generic applications.
Framatome fuel design and transient analysis have been approved by the staff for application to a Westinghouse NSSSs. The plant-specific applications of these methodologies have been reviewed by the staff for this license application and found to be acceptable.
In addition to the fuel change, the minimum required reactor coolant system (RCS) flow is also being reduced by approximately 5% with this license amendment request.
Because of the fuel design and methodology changes, the plant TS must undergo extensive revisions.
Framatome previously submitted, and the staff approved, a methodology to convert the plant TS commensurate with the fuel and methodology changes. The revised Sequoyah TSs that have been proposed generally follow this methodology. Certain specific changes are included as t'
discussed in Section 3.0 of the safety evaluation (SE). The use of a Core i-Operating Limits Report (COLR), that has already been approved for use at SQN
)
l allows a number of the cycle-specific TS limits to be included in the COLR l
h rather than the TS. The staff has reviewed the approved methodology and the proposed TS with the conclusion that the proposed TS are adequate.
The l
Framatome methodology is contained in Topical Report BAW-10163P-A, " Core i
Operating Limit Methodology for Westinghouse-Designed PWRs" (Reference 13).
Each of the TS changes is individually described in Section 3.0, TECHNICAL SPECIFICATION CHANGES, of this SE.
The new Framatome fuel wil1' be inserted into a core that is made up of mostly Westinghouse fuel.
The majority of the Westinghouse fuel is the Vantage-5H (V-5H) design; however, there will be limited use of Westinghouse Standard fuel.
In the first transition core for Unit 1, only 10 of the 193 assemblies in the core will be the Westinghouse Standard design. All three fuel designs and all of the analysis methodologies have previously been approved by the staff, however, the transition ' cores (cores with a mix of Westinghouse and Framatome fuel) have to be evaluated individually from a thermal hydraulic, stru.ctural, and plant response to the loss-of-coolant accident (LOCA) standpoint. The three fuels in the same reactor core may behave differently than a full core of any of the designs. As a result, Framatome has performed an extensive analysis to. justify the mixed core application *. The resident V,5H design was tested by Framatome to evaluate the specific flow characteristics to establish a core design with appropriate limits to provide acceptable margin.
The thermal-hydraulic analysis for the transition to Framatome fuel is described in detail in Section 2.3 THERMAL HYDRAULIC EVALUATION of this SE.
The original submittal and the subsequent submittals have undergone considerable revision. The original submittal was updated by three revisions.
Additionally, in answering the staff questions, aspects of the original submittal and the January, 10, 1997, discussion of mixed core thermal i
hydraulic analysis were updated and changed. Some of the important information has been withdrawn by the licensee' because it contained incorrect statements.
Revisions and corrections have been submitted for the TS changes, the TS bases changes, the LOCA analysis methodology and results, the non-LOCA l
transient analysis and methodology, and the thermal hydraulic analysis methodology.
In a letter dated February 7,1997, (Reference 6), the licensee i
i provided supplemental information in response to the staff's requests for l
1 additional information (RAI) on January 8 and January 14, 1997. On I
February 27, 1997, representatives of the licensee and its consultants met the NRC staff at Rockville, Maryland. Subsequently, conference calls were held on March 3, 4, 12, and 13, 1997.
In a letter dated March 17, 1997 (Reference 4),
i the licensee provided a revised response to the staff's RAI to include i
additional information requested by the staff during the February 27, 1997, meeting, and during the telephone conferences mentioned above. The April 6, 1997, submittal provides a final and complete set of answers to all the staff i
questions. As a result, the entire docketed correspondence associated with this license amendment must be consulted to establish the basis for this i
action.
j 3.0 EVALUATION l
3.1 LOCA Analysis l
The licensee has performed a reanalysis of the LOCA for the new fuel and the j
lower RCS flowrate. The analysis was. performed using the approved evaluation model (EM) described in Topical Report BAW-10168 Revision 2 and Revision 3, i'
"RSG LOCA, BWNT Loss-of-Coolant Acci< lent Evaluation Model for Recirculating-Steam Generator Plants" (References 14 and 15). Both revisions of the EM were previously reviewed and approved by the staff. The analysis was performed
)
i using the~ computer codes contained in Topical Report BAW-10164P-A,
{
"RELAPS/ MOD 2-B&W"'(Reference 17), which has also been reviewed and approved by i
the staff. The staff reviewed and approved these methodologies for use in
)
i NSSSs designs that utilize recirculating loop. steam generators like the SQN t
design and concludes that use of these methodologies for the SQN plant is j
acceptable.
l The calculations were performed using a conservative set of initial-conditions. The power was assumed to be 102% of rated thermal power with 15%
i of the steam generator tubes assumed to be plugged. The lower RCS flowrate j
was assumed and the beginning of life fuel and moderator parameters were used.
L The single failure chosen for tho analysis was the loss of one entire train of i
emergency core cooling system (FCCS). equipment and, when appropriate, flow is i
.modeled to be lost through the break rather than injected into the core. A maximum total peaking of 2.5 is assumed. The staff has reviewed these assumed initial conditions and found them to be conservative and, therefore, finds l
. them acceptable.
-3.1.1 Larae-Break LOCA 4
A number of sensitivity studies were performed to determine the set of conditions that will produce the limiting results. A discussion is provided to show that generic parameters like the RELAP time-step and the nodding configurations chosen will yield acceptable results. An evaluation was also performed for important LOCA parameters. A time-in-life, ECCS flow, break j-location, discharge coefficient or break size, break type and containment pressure study was presented. The analysis results indicate that a beginning of life, minimum ECCS flow, cold leg break at the discharge of the reactor coolant pump yields limiting results. The break size or discharge coefficient
)
4 it
4 study revealed that a double-ended cold leg break with a discharge coefficient of 1.0 will also yield limiting results. The study of pipe break type, (i.e., rupture vs. pipe split) indicated that a pipe rupture resultstin a i
higher peak clad temperature (PCT). The containment backpressure study r
indicsted that the SQN Updated Final Safety Analysis Report (UFSAR) base case 1
shown in the original submittal figure 5.4-4 yields the most conservative results.
Based on the results of the sensitivity studies presented and
+
j reviewed by the staff, the parameters that will yield the most limiting LOCA l
i results have been identified and are appropriate for use in verifying that the peaking limits are acceptable and establishing compliance with 10 CFR 50.46.
i Five separate LOCA calculations were performed for five different axial power shapes, where the axial peak locations in the core are at 2.9, 4.6, 6.3, 8.0, i
and 9.7 feet.
The peaking, F, chosen for the analysis are presented. in Figure 5.5-1 of the submittal,. The peak F values occurring at each of the above locations are 2.5, 2.5, 2.415, 2.3,,and 2.146. These calculations were performed to assess the adequacy of the peak linear heat rates chosen. The peaking limits are assumed at each core location and the LOCA results are evaluated to determine if the limits have been exceeded. The results verified that these peaks produce acceptable results. The PCT occurs when the peak.
linear heat rate occurs at the 8.0. foot elevation and the maximum localized and core-wide oxidation occur at the 2.9 foot elevation with all the results meeting the ECCS acceptance criteria.
(
The maximum PCT is calculated to be 2115 *F which is less than the 10 CFR 50.46(b) limit of 2200 *F.
The maximum calculated localized oxidation is 5.5%
which is less than the limit of 17% and the maximum calculated core-wide oxidation is 0.77% which is less than the 1% acceptance criterion. As a result, acceptable results are obtained for the new MARK-BW fuel and the peaking limits that were assumed yielded acceptable results.
3.1.2 Small-Break LOCA The original set of calculations performed and submitted in Reference 1 to support the license amendment were performed using a version of the EM that had not been approved by the NRC. During the staff review of the evaluation nodel, the model was changed to conform to the requirements of 10 CFR 50.46 and Appendix K and subsequently received staff approval (Reference 14 and 15).
As a result of staff questions, a new set of calculations was performed (Reference 2).
During the staff review and while responding to additional i
staff questions,'Framatome discovered that this new set of calculations did j
not model the break flow correctly and unrealistic break' flows were being calculated.
The unrealistic break flows made the analysis results unacceptable.
Framatome discovered that two input errors caused the problems and performed a new set of calculations with acceptable results (Reference 8).
The input error is described in detail in supplemental responses to the initial staff questions in References 7, 8, and 10. After the error was corrected, Framatome independently verified that the break flow was being correctly calculated in accordance with Appendix K of 10 CFR Part 50, using the Moody break flow correlation. The staff has reviewed the material and concluded that the small-break LOCA analysis presented in References 8. and 10 were performed using the approved evaluation model and that the break flow is i
acceptably modeled for SQN.
?
4
, A spectrum of bra k sizes was chosen to bound a range of break sizes.
Break size!, of 1.5, 2.5, 2.75, 3.0, 3.25, and 5.0 inches in diameter were analyzed.
The 2.34 inch diameter break of the centrifugal charging pump injection line was aiso performed.
This break assumes that all centrifugal charging pump flow is either lost through the break with the other train assumed to be lost due to the assumed single failure.
From the analysis results, the limiting break was the 2.75 inch diameter break. The PCT was calculated to be 1162 F,
which is well under the acceptance criterion of 2200 *F of 10 CFR 50.46(b).
The other 10 CFR 50.46(b) acceptance criteria are also met. The highest
)
localized oxidation was calculated to be 0.004% which is well below the 17%
1 acceptance criteria. The calculated core wide oxidation is well below the 1%
l limit because the highest calculated localized oxidation was well below 1%
(0.004%).
No fuel clad ruptures are predicted and a coolable core geometry is maintained.
With regard to long-term cooling, this amendment does not affect j
the long-term cooling and the existing evaluations remain applicable. As a j
result, the staff has concluded that the ECCS acceptance criteria for the small-break LOCA were met and the results are acceptable.
3.1.3 Mixed Core LOCA Analysis An evaluation of the mixed core effects on the LOCA analysis has been performed on the large-break LOCA because these results are much more limiting than the small-break LOCA.. Because of the small differences in the different fuel assembly pressure drops, flow can cross between adjacent assemblies causing higher or lower flow in particular assemblies. This effect has been evaluated to determine the overall effect on the limiting PCT.
Analysis were performed for the resident Westinghouse Standard and V-5H fuel.
The PCT associated with this fuel is calculated to be 1911 *F for each of these fuel types in a mixed core. The assumed flow for these calculations was at a higher RCS flow than is allowed with this license amendment; however, West.inghouse has evaluated the effect of the RCS flow with no changes in the core average temperature to be very small. The calculated PCT for the MARK-BW fuel is 2115 F, assuming the lower core flow.
These calculations and.results were performed for a full core of each fuel type.
The' pressure drop kssociated with the V-5H fuel is approximately 1 psi higher than with the MARK-BW fuel and slightly lower for the Westinghouse Standard fuel. As a result, it is expected that the MARK-BW and Westinghouse Standard fuels would benefit or receive raore flow in a mixed core configuration and the Westinghouse V-5H would be adversely affected and receive less flow.
This is a situation very similar to the core reloads at Catawba and McGuire where the Mark-BW fuel was inserted into a core of Westinghouse OFA fuel. An analysis was performed for the Catawba and McGuire reload to determine the effect of the mixed core on the PCT. The analysis performed indicated that the difference in PCT would be less than 20 F.
As a result, the licensee will i
carry a 20 *F PCT penalty on the Westinghouse Vantage-5H in the PCT analysis.
This penalty will be carried for all V-5H fuel that has not been burned for two cycles.
Fuel burned more than two cycles will not produce enough heat to be limiting with regard to PCT.
The staff finds this approach acceptable.
l
t
3.2 Non-LOCA Transient Analysis The non-LOCA transient analysis have been performed using the methodology outlined in the approved Topical Report BAW-10169P-A, "RSG Plant Safety Analysis, B&W Safety Analysis Methodology for Recirculating Steam Generator Plants." The methodology uses a RELAP5/M002-BW with a subchannel evaluation -
model-LYNXT to evaluate the transient response. The non-LOCA evaluation model approved by the NRC uses a " full power model" and a " low power model." To simplify the reload analysis process, Framatome has modified the low power model 'to model all full power and low power events. The main difference between the two models is the use of a hot and an average channel in RELAP in the full power model and the elimination of the modeling of the hot channel in i
the low power model.
Framatome stated that the hot channel modeled in RELAP i
is not important because a subchannel model, LYNXT, is used to make the detailed hot channel calculations. A description of the change and a set of comparative results for one transient were presented in Reference 7 and reviewed by the staff. The results of the comparative calculations indicate that modeling of the hot channel in RELAP does not significantly affect the analysis results. As a result, the staff finds the new modeling approach acceptable only for this application.
The licensee reanalyzed six non-LOCA transients. All of the other UFSAR Chapter 15 accident analyses have been evaluated to determine if existing analyses are adequate. The discussion provided by the licensee is contained in References 7 and 10. The staff has reviewed each of the UFSAR Chapter 15 accident analyses to determine the effect of the fuel change and the flow reduction.
For each transient that has not been specifically reanalyzed in this submittal, the staff has found that the existing UFSAR analysis remains acceptable, the transient has been bounded by another more limiting transient 1
that has been analyzed, or the reload-specific analysis is adequate. As a result, the staff has determined that the approach is acceptable for this' license amendment request.
The six limiting transients that have been reanalyzed using approved methods are the rod cluster control assembly (RCCA) withdraw at full power, the loss of electrical load,.both with and without pressurizer spray and the power-operated relief valve (PORV), the four pump (reactor coolant pump) coastdown, the locked reactor coolant pump rotor, and the main steam line break, both with and without rod withdrawal. The transients chosen were intended to be the most limiting of the overcooling, heatup, loss of RCS flow, and reactivity anomaly transients.
A conservative set of initial conditions was used to perform the analysis.
The reactor power was assumed to be 102%, and no increased pressure credit was taken when the transient pressure was modeled to increase in the departure from nucleate boiling ratio (DNBR) calculations. A limiting single failure is always assumed along with a the most reactive control stuck rod in the fully withdrawn position. A statistical core design (SCD) methodology was utilized for the DNBR calculations (discussed in greater detail in Section 2.3 THERMAL HYDRAULIC EVALUATION), with the exception of the main steam line rupture because the RCS pressure drops below the point where the SCD methodology yields acceptable results.
For the main steam line break, the calculations are performed using a deterministic core design approach. The subchannel code j
i i,.
-7 LYNXT, the critical heat. flux correlation BWCMV, and the statistical core design methodology have all been approved by the staff.
The staff reviewed these inputs for the non-LOCA transients and found them acceptable.
- The uncontrolled RCCA bank withdrawal at power transient. analysis was performed using the Framatome methodology with the new fuel at the lower flowrate. A conservative set of initial conditions was choser..
These assumptions will yield limiting results. At the initiation of tha transient, the reactivity inserted because of the rod withdrawal causes reactor power and i
pressure to increase.. The reactor is modeled to trip on high neutron flux.
The results indicate that the minimum DNBR is above the limit and no safety limits are reached.
The loss of external load or turbine trip transient analysis was modeled to achieve the most limiting pressure transient for both the primary and secondary systems. One of the three code safety relief valves is modeled not to open, a positive moderator temperature coefficient is chosen and the limiting -fuel conditions are chosen as analysis inputs. The primary pressure control systems, the PORVs and pressurizer sprays are. assumed.to not operate to control primary system pressure, thus achieving the peak primary system.
pressure. The primary pressure control systems are modeled to calculate the peak secondary system pressure because the high pressure reactor trip will not occur if the pressure' control systems operate and the reactor trip is delayed a short period of time. The delay in the reactor trip causes more heat to be transferred to the secondary system and a higher secondary peak pressure.
For the peak RCS pressure calculation, with no PORVs or sprays modeled, the reactor trips on high RCS pressure and the calculated peak pressure is 2740 psia which is below the limit of 2748 psia.
For the peak main steam line pressure, with the pressurizer spray and PORV modeled, the first reactor trip 4
setpoint reached is on OTAT and the calculated peak pressure is 1201 psia which is below the limit of 1208 psia. The calculated DNBR reriained well above the limit.
t The complete loss of forced reactor coolant flow and the single reactor coolant pump locked rotor were modeled to minimize the DNBR.
For the loss of l
forced flow, the reactor is mode. led to trip on reactor coolant pump underfrequency and,he minimum DNBR remains above the safety limit.
For the locked reactor coolant pump rotor, the reactor trip on low RCS flow and the minimum DNBR is below the limit for a short period of time. The DNBR is calculated to occur with less than 5% of the fuel pins experiencing DNBR.
This is acceptable and remains bounded by the existing analysis for the locked RCP rotor.
The peak cladding temperature is calculated to be 1104 F and no significant oxidation is expected to occur.
The peak pressures for these transient also do not challenge the safety limits and remain below 2600 psi.
The licensee calculated the effects for the main steam line break both with a coincident rod withdraw and without a coincident rod withdraw. A conservative set of assumptions are modeled. The only pump that is assumed to inject borated water following the tr'ansient is the one safety injection pump. The other safety injection pump is assumed to be out of service and the high head safety injection pumps or charging pumps are also not modeled. As a result, there is no injection of borated water prior to the RCS pressure dropping below approximately 1500 psig. The water modeled is assumed to be at 1950 ppm
i
. 4 boron even though the TS required boron concentration in the borated water storage tank is 2500 ppm.
The one assumption that is not modeled conservatively is the boron concentration in the ECCS piping. The water in the ECCS piping up to the second check valve from the RCS is all assumed to be at 1950 ppm boron.
This is not conservative because check valve leakage from the unborated RCS could dilute the ECCS piping. The licensee justified this assumption with the following reasoning. There can be some leakage and dilution of the boration in the piping because of tha difference in the modeled boron concentration and the required boron concentration, 1950 vs. 2500 ppm. Any check valve leakage 1
that may occur will be detected by the pressure monitors in the discharge of the safety injection piping.
If any leakage is observed, the piping will be flushed to assure the piping is full of borated water. Any RCS leakage will also cause the piping to pressurize. Once the piping is pressurized to RCS pressure leakage will stop until the next valve or check valve begins to leak.
Two check valves would have to leak to cause significant boron dilution in the ECCS piping. Additionally, there is a monthly TS requirement to vent this particular injecti6n path. That TS is intended to prevent voids from forming in the ECCS piping, however, the vent piping (3/4-inch) is opened in each of four locations for 10 minutes. The volume of water that is displaced by the venting will be replaced by borated water from the BWST at 2500 pmm boron causing the piping to remain full of borated water.
In addition, the licensee has also performed a sensitivity study to show the limiting break would remain the limiting break if no boron is assumed in the ECCS piping up to the safety i
injection pump. To dilute this stretch of piping it would require the leakage of three check valves. As a result, this particular assumption is acceptable for tnis particular plant and this evaluation.
For the main steam line rupture a conservative set of assumptions and initial
- conditions was assumed. The analysis assumes that the reactor is at no-load, end of life, zero boron concentration with a limiting set of fuel and moderator conditions.
The steam generators are assumed to have no plugging and the reactor coolant pumps are modeled to continue running during the transient for some of the cases to obtain the most severe reactor cooldown.
All four generators are modeled to blowdown through the break prior to main steam isolation, and the ruptured generator is modeled to completely blow down. The return to criticality and power is modeled as a result of the cooldown and the limiting moderator and fuel temperature feedback are enhanced by the modeling of the most reactive rod stuck in the fully withdrawn position.
Four separate main steam line rupture cases were analyzed. The double ended rupture, upstream and downstream of the steam measurement device, with and without offsite power available. The limiting case was the double ended steam lineruptureoftheupstreampipingtothemainsteammeasuringdevicethat results in a 5.326 ft flow area, with offsite power available. The results indicate that the power excursion peaks at something under 20% power with no DNB calculated.
Because the RCS pressure goes so low, the W-3 CHF correlation is used to calculate DNBR rather than the statistical core design utilizing the BWCMV correlation.
The results are acceptable.
The final main steam line rupture considered was the at-power case assuming the coincident withdraw of the control rods. The control rods' are assumed to
4 1
I P
_g_
1 withdraw because the main steam cooldown causes the rods to respond by
~
i withdrawing to keep RCS temperature within the programmed band.
For this 3
event, the minimum DNBR also remains above the limit and the results are acceptable.
Although the main steam line rupture is a limiting transient and fuel damage j
is not prohibited, the calculated results indicated that no safety limits are i
challenged, DNBR is not achieved, and no fuel damage is predicted.
The main l
steam line rupture is the most limiting main steam line transient and the consequences are more severe than the other more frequent transients. Because a
no fuel damage is predicted for the limiting transient, the other less limiting main steam transients need not be analyzed. The staff finds this 4
conclusion acceptable.
I The licensee has reanalyzed a number of transients and provided discussion of all the Chapter 15 transient analysis to show that the proposed TS changes will not cause adverse consequences. The staff has reviewed the analysis and i
evaluations and has concluded that the transients have been adequately evaluated and the results are acceptable.
i(
I 3.3 Thermal Hydraulic Evaluation 1
A thermal hydraulic evaluation has been performed for the new fuel at the i
lower flow condition at SQN. A set of calculations was performed to assure i
that the fuel performance is acceptable. Another detailed analysis was i
performed to evaluate the effect of having three different types of fuel j
designs in the same core. This analysis resulted in a mixed core DNBR j
penalty.
The thermal hydraulic evaluation was performed using an approved methodology contained in Topical Report BAW-10156-A, "LYNXT-Core Transient Thermal-
{
Hydraulic Program," a thermal-hydraulic analysis code. The DNBR calculations are performed using the staff approved methodology in BAW-10159P-A, "BWCMV Correlation of Critical Heat Flux in Mixing Vane Grid Fuel Assemblies." The core design methodology is contained in BAW-10170P-A, " Statistical Core Design for Mixing Vane Cores." The analysis methodologies have been approved for 4
each of the fuel designs utilized at SQN.
The limiting peaking factors associated with the reference power distribution used for the Sequoyah thermal hydraulic analysis are Fj=1.64, which corresponds to the maximum allowable minus 4% of measurement uncertainty, F,-1.55, and the bundle average peak of 1.557.
l The analysis methodology and the critical heat flux correlation have been i
approved for full core applications.
The evaluations performed for Sequoyah l
resulted in acceptable thermal hydraulic performance.
Because there will be 1;
three'different types of fuel in the Sequoyah core for a number of cycles, a mixed core analysis becomes an_important area for this reload application.
1 The licensee has proposed to use the full core methodology to evaluate all l
bounding mixed core configurations and use the most limiting result to establish the transition core penalty. The transition core penalty is applied to the results of the calculations performed on the full core of MARK-BW fuel.
The statistical core design approach has established a statistical design 4
limit, or' safety limit. A value higher is chosen for the thermal design s
4
,o.
-e
= -, - - + -
- w-
l l l limit. The percent difference between the thermal design and the statistical design-limit is referred to as the retained margin. The retained margin i
allows penalties to be taken against the thermal design limit without j
exceeding the statistical design or safety limit. The DNBR penalty associated with the mixed core flow effects is assessed against the retained margin associated with the statistical-safety limit..
l l
The penalty determined was calculated in a manner consistent with the manner i
that the penalty was chosen for Trojan plant in a similar mixed core i
configuration.
The penalty was calculated using different combinations of I
both fuel and limiting peaking factors. The calculations, use actual assembly flow data obtained by running flow and pressure drop tests for both the'V-5H t
, and the MARK-BW fuel. Data was already available for the Standard fitel. The j
penalty calculated was intended to bound the worst possible cases.of fuel combinations and peaking factors. The actual core configuration and expected peaking factors were then used to verify that the penalty sufficiently bounds the expected conditions.. The' staff finds this approach and the proposed i
penalty acceptable for the following reasons..Both worst case and cycle specific calculations were performed, using flow tests for the individual fuel types, and the results presented show the penalty will adequately bound the actual cycle specific plant configuration. The calculational methods.have been reviewed and approved by the staff for this cycle-specific reload application. The penalty presented will apply to the initial Sequoyah transition core.
The necessary mixed core DNBR penalties associated with i.
future.SQN cores are expected.to go down as the number of V-5H assemblies also go down. As stated in Reference 24, the licensee's procedures require that prior to taking credit for any such reductions in the penalty factor for i
future fuel cycles, the. licensee shall submit the matter of reducing the
~
i pena 1ty factor to the NRC staff for review and approval.
License Conditions y i
' for. Unit I and Unit 2 (2.C.(25) and 2.C(18), respect'ively) have been included l
with the amendment. Therefore, this is acceptable.
~
3.4 Mechanical Desian Analysis
)
i The new fuel, MARK-BW fuel, is described in BAW-10172P-A, " MARK-BW Mechanical l
Design Report," and has been approved by the staff. The fuel design has undergone minor modifications to the bottom nozzle and the supporting welds and rivets. The staff has reviewed the description of these changes according to the acceptance criteria in the approved topical report and the staff concluded that these changes are minor and acceptable. Additionally, for the mixed core analysis a series of calculatjens using an approved methodology were presented to evaluate'the effect of combined seismic and LOCA loads. The analyses were performed considering different combinations of MARK-BW, V-5H, and Standard fuel types. The results showed that the combined loads of seismic and LOCA are conservatively within the crushing load limits.
The licensee evaluated the structural integrity of the FCF Mark-BW fuel assembly for the normal operating, upset and faulted conditions. The components evaluated are hold down spring, clamp screw, guide thimbles, spacer grids, top and bottom nozzles. The methodology of the licensee's evaluation follows the Topical Report BAW-10172P (Reference 22) which was reviewed and approved by the.NRC staff and is considered applicable to SQN.
The stress limits were determined in accordance with the ASME Code Section III,
4 i
Subsection NG.
The licensee indicated that this is consistent with the SQN UFSAR.
2 The evaluation consisted of performing structural analyses of the FCF including guide thimble buckling evaluation and a detailed finite element analysis of the top and bottom nozzles. The maximum calculated stresses and applied loads at critical components were compared to the design basis allowable limits in Tables 8-1 and 8-5 of Reference 2 for the normal operating and faulted conditions, respectively. The maximum vertical loads for the faulted condition are provided in Table 8-4.
These loads are less than the allowable loads provided in Reference 7.
The maximum cumulative usage factor (CUF) was determined to be less than 1.0 for the hold down spring for the 4
design fuel service life under the loading cycles described in Table 8-2 of Reference 25.
Based on its review, the staff finds all calculated maximum stresses and CUF provided by the licensee are below the allowable limits in' the Subsection NG of ASME Section 3 and, therefore, acceptable.
1The licensee assessed the effect of flow induced vibration for the critical components of the instrument sheath and guide thimbles.
In response to the staff's request for additional information, the licensee indicated that the spacer grids provide sufficient support to limit the rod lateral motion and the calculated maximum crossflow velocity is small in comparison to the design acceptance crossflow velocity for the proposed FCF (References 6 and 7).
Based on its review, the staff concludes that the use of FCF fuel will not increase the potential for the flow induced vibration.
The comparison of design data provided in Table 3-1 of Reference 25 indicates that the FCF fuel (Mark-BW) is similar in design to tSe Westinghouse fuel (V-5H) currently licensed and operated at SQN Units 1 and 2.
Test results provided in Table 3-3 of Reference 25 show that these two fuel assemblies are
. dynamically similar in stiffness, natural frequencies, and damping.
Based on its review, the staff concludes that the use of FCF will not have significant impact on the structural responses to static or dynamic loads that were considered in the design basis analyses and, therefore, will not have adverse effects on the core supporting internal components.
Based on its review of the information provided by the licensee, the staff 4
concludes that the licensee's evaluation is found to be acceptable and the analysis adequate for use at SQN. The staff concludes that the maximum stress and fatigue usage factors at critical locations of the FCF components are within the ASME Section III allowable limits. Therefore, the use of the FCF MARK-BW fuel at SQN Units 1 and 2 is acceptable with respect to the structural integrity and mechanical design of the fuel assemblies and the supporting reactor internals.
4.0 TECHNICAL SPECIFICATION CHANGES AND LICENSE CONDITIONS TS Table 2.2-1 j
The Design RCS flow is changed to 90,045 (87,000 X 1.035) rather than 91,400. The new lower design flow has been justified and found to be acceptable and supported by appropriate and acceptable analysis.
Using
1 ea i !
the lower core flow, a reanalysis of the limiting DNBR transients has been performed with acceptable results.
The f AI) methodology has been amended to conform to the Framatome i
method,(ology in accordance with the. approved topical report.
i The ranges l
for which f (AI)=0 for: measured q - q between -29 percent and +5 l
3 percent has been changed to the v,alues, in the COLR identified by QTNL*
l and QTPL*. These values are also used for the AT trip reduction setpoints. The amount that the AT trips are reduced is also being 4
changed from the fixed values of 1.5 percent and 0.86 percent to COLR
~
{
values that change from cycle to cycle, QTNS* and QTPS*. These changes are acceptable because the methodology used to generate these values has j
been approved by the staff and is applicable to SQN.
It is also appropriate for these values to be contained in the COLR and the e
methodology that must be used to. generate the values has been approved by the staff and is referenced in the TSs.
j The f (AI) methodology has also been amended. The ranges for which 2
f (AI)=0 for measured q - q has been amended from all values to the 3valuesintheCOLRidenkifie,dbyQPNL*andQPPL*. These values are also used for the AT trip reduction setpoints. The amount that the AT trips are reduced is also being specified as QPNS* and QPPS*.
These changes are acceptable because the methodology used to generate these values has been approved by the staff.
It is also appropriate for these values to be contained in the COLR and the methodology that must be used to generate the' values has been approved by the staff and is referenced in i
the TSs.
The TS bases are also being amended to support the license amendment.
TS Bases section 2.1.1 REACTOR CORE has been modified to describe the new approaches in the OTAT and OPAT methodology. The description in the TS bases including the 4% uncertainty has already been applied to the F " values of-1.70 for the MARK-BW fuel and 1.62 for the Westinghouse f$el.
j TS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR-F (Z)
A number of changes have been made to this TS in order to facilitate the change in methodology. The F,(Z) spec is now being referred to as F,(X, Y, Z).
The limits are now maintained in the COLR. This change results from the new methodology. New action items are added and are consistent with the approved methodology. The staff finds the changes acceptable because the changes are consistent with the approved methodology.
The SR 4.2.2.2 along with the discussion in the Bases section is being changed to comply with the new Framatome methodology.
TS 3/4.2.3 NUCLEAR ENTHALPY HOT CHANNEL FACTOR The title of this TS is being editorially changed to NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - F (X,Y).
Consistent with the heat flux hot channelfactor,theaccepta$lelimitsarebeingmaintainedintheCOLR
1 i
- i i and the actions and surveillance requirements contained in TS 4.2.3.2 l'
are changed to be consistent with the approved methodology. The bases section is also being amended for censistency with the new approach.
TS 3/4.2.4 00ADRANT POWER TILT RATIO
- The action statements.that require a reduction of thermal power when the excore power detectors indicate a measured quadrant power tilt ratio (QPTR) in excess of 1.02 are being changed to require that power be reduced by 3% for every 1% the QPTR exceeds the measured value of 1.02 rather than the measured value of 1.0.
Basing the power reduction on-
- the QPTR limit value of 1.02 rather than from the base value of 1.0 is acceptable because the Framatome methodology applies an allowance for peaking up to the QPTR limit.
The bases section is also being changed to be consistent with the new approach.
Table 3.2-I DNB PARAMETERS The Reactor Coolant System Total Flow limit is being changed from a
' value of 2 378,400 to values specified in Figure 3.2-1.
Figure 3.2-1 has a flowrate vs thermal power fraction (% RTP) plot that varies the flow ibit between 90% and 100% power. The change is acceptable because the margia to DNBR has been maintained for the lower flows by derating the power.
SECTION 6.0 REPORTING REQUIREMENTS Section 6.9.1.14, CORE OPERATING LIMITS REPORT, is being amended to include the f (AI)' and f (AI) limits referenced from TS 2.2.1.
The W(z) 3 2
factor associated with specification'3/4.2.2 is being eliminated and other editorial changes are being made. These changes are consistent with the approved methodology and the discussion of the other TS changes described above.
Section 6.9.1.14a is being amended to include the new approved reference topical reports.
References to WCAPs 9272-P-A, 101216-P-A, and WCAP-13631-P-A are being replaced by WCAP 10054-P-A, and BAWs 10180P-A, 10169P-A, 10163P-A, 10168P-A Rev 2, and 10168P-A Revision 3.
These changes are consistent with the new methodologies.
Plant operation continues to be limited in accordance with the values of cycle-specific parameter limits that are established using NRC-approved methodologies to assure that all applicable limits (e. g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident limits) of the safety analysis are met. Therefore, the NRC staff concludes that the proposed changes are acceptable.
Consistent with the licensee's submittal dated April 18, 1997, License Conditions 2.C.(25) and 2.C.(18) have been added to the SQN licenses for Units 1 and 2, respectively, as discussed in Section 3.3 above.
In addition, a number of editoria1' changes have also been proposed such as the retitling of TS sections, insertion of.the term " Power", capitalization of " Trip" and correcting the spelling of " Ratio" in TS 3.2.4 and changes to the phraseology
1
, throughout to be consistent with the FCF approach. The staff has reviewed all of these changes and finds them to be acceptable.
5.0 STATE CONSULTATIQH In accordance with the Commission's regulations, the Tennessee State official was notified of the proposed issuance of the amendments. The State official j
had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a i
facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves 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 Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding j
1 (61 FR 20856 dated May 8, 1996). The amendment also involves changes to recordkeeping, reporting or administrative procedures and requirements.
~;
Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and (c)(10).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be 1
prepared in connection with the issuance of the amendments.
7.0 CONCLUSION
i The Commission has concluded, based on the considerations discussed above, i
that:
(1) there is reasonable assurance that the health and safety of the l
public will not be endangered by operation in the proposed manner, (2) such 4
activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common 4
defense and security or to the health and safety of the public.
Principal Contributors:
C. Jackson, T. Huang, S. Wu, C. Wu, Date: April 21,1997
1 l
l !
REFERENCES i
i 1.
Tennessee Valley Authority to NRC, letter TVA-SQN-TS-96-01, " Conversion from Westinghouse Fuel to Framatome Cogema Fuel at Sequoyah Nuclear i
Plant Units 1 and 2," April 4, 1996.
t t
2.
Request for additional, information dated January 8,1997.
i l
3.
Request for additional information correction dated January 14, 1997.
t j
4.
TVA submittal dated February 7, 1997.
f t
5.
TVA submittal dated February 13, 1997.
6.
TVA response to staff questions dated, February 7, 1997.
l 7.
TVA response to staff questions dated March 17, 1997.
8.
TVA response to staff quest' ions dated March 20, 1997.
' 9.
TVA response to staff questions dated March 25, 1997.
I 10.
TVA response to staff questions and supplemental Technical Specifications changes, dated April 1, 1997.
11.
TVA response to staff questions dated, April 6, 1997.
12.
TVA submittal addressing mixed core thermal hydraulics dated i
January 10, 1997.
13.
BAW-10163P-A, Core Operating Limit Methodology for Westinghouse-Designed PWRs.
j 14.
BAW-10168P-A Rev. 2, RSG LOCA, BWNT Loss-of-Coolant Accident Evaluation i
Model for Recirculating Steam Generator Plants.
I 15.
BAW-10168P-A Rev. 3, RSG LOCA, BWNT Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants.
16.
BAW-10169P-A, RSG Plant Safety Analysis, B&W Safety Analysis Methodology for Recirculating Steam Generator Plants.
17.
BAW-10164P-A, RELAP5/ MOD 2-B&W, 18.
BAW-10156P-A LYNXT - Core Transient Thermal-Hydraulic Program.
19.
BAW-10170P-A, Statistical Design Limit Methodology.
20.
BAW-10159P-A, BWCMV Correlation of Critical Heat Flux in Mixing Vane Grid Fuel Assemblies, 1
0
e o 21.
Hernan, Ronald W., letter to James H. Taylor, EXTENSION OF APPLICABILITY OF BWCMV TO VANTAGE-5H FUEL WITHOUT INTERMEDIATE FLOW MIXERS FOR SEQUOYAH UNITS 1 AND 2 CYCLE 9 RELOAD (TAC NO.' M93557), dated September 16, 1996.
l 22.
BAW-10172P-A, MARK-BW Mechanical Design Report.
3 23.
TVA submittal dated April 10, 1997.
24.
TVA letter dated April 11,.1997.
25.
BAW-10220P, Mark-BW Mechanical Design Report, Babcock & Wilcox (B&W)-
Topical Report, March 1996. (proprietary) i 26.
TVA letter to staff - Revision 2 to TS 96-01 dated March 19, 1997.
l 27.
Telefax from TVA dated April 18, 1997, accepting proposed license
{
condition.
1 I
n h
b t
t
,