ML20205N385
| ML20205N385 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 10/25/1988 |
| From: | Schnell D UNION ELECTRIC CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| Shared Package | |
| ML20205N389 | List: |
| References | |
| ULNRC-1850, NUDOCS 8811040124 | |
| Download: ML20205N385 (31) | |
Text
1901 Gratet Street Post Offxe Bas 143 St. ltut Mnoud 63165 314.S51 2650
(.bVION
$LECTRIC ^
U U""""
$$$ October 25, 1988 <
U.S. Nuclear Regulatory Commissjon I ATTN: Document Control Desk Mail Station Pl-137 Washington, D.C. 20555 i Gentlemen: ULNRC- 18 50 DOCKET NUMBER 50-483 i CALLAWAY PLANT CYCLE 4 RELOAD LICENSE AMENDMENT Referenco: ULNRC-1807 dated 7-15-88, Cycle 4 Transition Core DNBR Penalty -
Union Electric herewith transmits an application for amendment to Facility Operating License No. NPF-30 for the Callaway Plant. This ar ndment request includes changes in support of the ca laway Cycle 4 Roload.
Cycle 4 will be the second and final transition core to a full VANTAGE 5 core for Cycle 5. Changes in Cycle 4 which require revision of the Technical Specifications i include: increased F (z) and F-delta-H peaking factors; incorporation of a po@itive moderator temperature coefficient; increased refueling water storage i tank / accumulator boron concentrations; and increased ,
sodium hydroxide concentration in the containment spray additive tank. j l
- Included as part of this amendment request are
- a l description of the Cycle 4 core; a description of l
! onhancements made under 10CFR 50.59 to callaway Plant l which are related to Cycle 4 and improvo plant '
! performance; Safety Evaluations and a Significant Hazard "
l Evaluation in support of the required Tcchnical [
l Specification changes; and marked-up Technical i t
Specification and FSAR pages.
- Included in the safety evaluation associated with the positive moderater temperature coefficient, as !
reflected in the draft FSAR changes, is the reanalysis of i boron dilution transients. This reanalysis incorporates i a revised methodology which includes the offects of i
density compensation on the dilution flow ratos. The reanalysis continues to demonstrato acceptable results [
for boron dilution transients. 9
{ bt 8811040124 881025 h ADOCK 05000483 t
t t
i.
In addition, Union Electric notes that the 12.5% transition core DNBR penalty discussed in Technical Specification Bases
, 3/4.2.2 and 3/4.2.3 is greatly reduced (less than 2%) when l consideration is given to WCAP-il837, "Extension of Methodology-for Calculating Transition Core DNBR Penalties," as discussed in the referenced letter.
This request has been reviewed and approved by the callaway Onsite Review Committee and the Nuclear Safety Review Board. It has been determined that this request does not involve any unroviewed safety questions as defined in 10CFR 50.59 nor a significant hazard consideration as determined by the three j factor test por 10CFR 50.92.
Union Electric currently plans a Refuel 3 carly start date i of March 17, 1989 and a Cycle 4 carly re-start date of May 8, 1989. We therefore request that NRC approve this amendment request by April 21, 1989.
3 Enclosed is a check for the $150 application foo required by 10CFR 170.21.
Very truly yours, i
Donald F. Schnell DS/plh Attachments ULNRC-1850 dated 10/25/88 The $150 application fee was mailed separately from this application.
Reference:
Union Electric Co. check #360244 dated l 10/25/88 on Boatman's Bank of Troy, Missouri. If there are any questions contact Dave Shafer at 314-554-3104.
I i
l i
l l
l -
STATE OF MISSOURI )
) S S-CITY OF ST. LOUIS )
Alan C. Passwater, of lawful age, being first duly sworn upon 03th says that he is Manager, Licensing and Filels (Nuclear) for {
Union Electric Company; that he has read the foregoing document and I known the content thereof; that he has executed the same for and on behalf of said company with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.
/14 /) l By i
Alan C. Passwater
! Manager, Licensing and Fuels l Nuclear SUBSCRIBED and sworn to before me this 475 b day of ,1987 i
BARGAR 0 h{Fy
.PFA NOTARY WBUC, STATE OF MISSCURI Mf COV!,tt$il0N EXPtRts AP2il 22,1983 ST. LOUIS COUNTY
Y
!k-e cc Gerald Charnoff, Esq.
Shaw, Pittman, Potts & Trowbridge i 2300 N. Street, N.W.
Washington, D.C. 20037 Dr. J. O. Cermak CFA, Inc.
4 Professional Drive (Suite 110)
Gaithersburg, MD 20879 R. C. Knop "
Chief, Reactor Project- Branch 1 U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137 Bruce Little ;
Callaway Resident Office U.S. Nuclear Regulatory Commission RR#1~
Steedman, Missouri 65077 Tom Alexion (2)
Office of _ Nuclear Reactor. Regulation U.S. Nuclear Regulatory' Commission
- Mail Stop 316 7920 Norfolk Avenue Bethesda, MD 20014 Ron Kucera, Deputy Director Department of Natural Resources P.O. Box 176 Jefferson City, MO 65102 Manager, Electric Department Missouri Public Service Commission P.O. Box 360 Jefferson City, MO 65102 I
1 i
._-----m. . . _ . . . . _ . _ _ . - _ _ , . . _ , . - . . _ _ ,,
l Attechment 1 ULNRC- 18 50 CYCLE 4 RELOAD DESCRIPTION i
I l
l J
CYCLE 4 RELOAD DESCRIPTION, Plans for Cycle 4 consist of 88 new 17 x 17 Westinghouse VANTAGE 5 (V-5) fuel assemblics (64 V-5 assemblics in Region 6A at 3.6 w/o U-235 and 24 V-5 assemblics in Region 6B at 4.0 w/o U-235),
together with 9 Optimized Fuel Assemblics (OFAs) and 96 V-5 assemblics carried over from Cycle 3. The 88 assemblics being added for Cycle 4 will have 10,112 Integral Fuel Burnablo Absorbers (IFBAs). WCAP-10444-P-A contains descriptions of the V-5 design features. In addition, the following changes will bc implemented in Cycle 4:
o Increased F-delta-H and Fg peaking factors o Increased RWST and accumulator boron concentrations associated with a positive moderator temperature coefficient (PMTC) o Increased sodium hydroxide concentration in the spray additivo tank (SAT) as a result of changes in the boron concentrations in the RWST and accumulators o Axial Blankots o Debris-Filter Bottom Nozzlos (DFBN) o Enhanced Performanco Rod Cluster Control Assemblics (SP-RCCA) o Reallocation of Goncric DNBR Margin INCREASED PEAKING FACTORS Increases in the measured F-delta-H peaking factor from 1.49 to 1.59 (corresponding to an analytical F-delta-H change from 1.55 to 1.65) and in the F peaking factor from 2.32 to 2.50 allow for more flexibility in(z)fuel management schemes, longer fuel cyclos, and improvement of fuel economy and neutron utilization.
Those increased peaking factors are consistent with values used for V-5 fuel in the Cycle 3 licensing submittal, ULNRC-1470 dated 3/31/87. Attachment 2 provides a safety evaluation which addresses those peaking factor increases.
POSITIVE MODERATOR TEMPERATURE COEFFICIENT AND INCREASED
{ RWST/ACCUMULA'ICR BORON CONCENTRATIONS i Technica] Specifications for Cycle 4 will allow a +5pcm/*F l modcrator temperature coefficient (MTC) below 70% of rated l thermal power, decreasing linearly from +5 to O pcm/'F at 100% of l
l 1 i
l
rated thermal power. The implementation of PMTC supports reductions in fuel cycle costs by reducing the required burnable absorber inventory, particularly for longer cycles which need a larger number of burnable absorbers to control MTC at the beginning of the cycle. However, to support longer cycles and i the implementation cf PMTC while retaining the required shutdown margin, the Technical Spec!4 sations for RWST and accumulator baron concentration rangen mast be increased to 2350-2500 ppm and 2300-2500 ppm, respectively.
l Safety evalum .ans have been performed for those changes and demonstrate that the safety criteria are mot for the proposed PMTC, with a revised boron dilution accident analysis, and that there is no adverse offect on accident consequences as a result of the increased RWST and accumulator boron concentrations. I Attachment 3 documents those safety evaluations.
INCREASED SAT SODIUM HYDROXIDE CONCENTRATION As a result of the proposed implementation of PMTC and increased RWST/ accumulator boron concentrations, it has also been determined that a higher sodium hydroxide (NaOH) concentration will be needed in the Spray Additive Tank such that the containment spray system will be able to achieve the minimum long i term post-LOCA sump pH of 8.5. The higher NaOH concentration l (31-34 weight porcent) is due to the increased baron I concentrations discussed above. Attachment 3 provides a safety l cvaluation which addresses the increased NaOH concentration in the SAT.
1 i The following changes are being implemented under the provisions of 10CFR 50.59 and require no license changes.
AXIAL BLANKETS The primary role of axial blankets in the reactor core is to improve fuel utilization (thus Icducing fucl costs) by reducing axial neutron leakage. Noutrons born near the top or bottom of the assembly have a high probability of migrating into the surrounding wator/ core support structure without causing fission.
By implementing axial blankets, a larger proportion of enriched fuel is located in higher flux regions of the core, thus reducing the probability of neutron leakage. Although the use of axial blankets necessitates an increase in U-235 onrichment in the enriched portion of the assembly to maintain the same reactivity, the overall enrichment, averaged over the entire assembly length, is lower in blankoted fuel than in non-blanketcc Eic1.
The safety implications of loading fuel with axial blankets are well documented, and are of the same nature as those arising from the insertion of fresh non-blanketed fuel in a reload core. The structural and thermal-hydraulic characteristics of blanketed fuel are virtually the same as non-blankoted fuel, while the neutronic differences arc well within the ranges encountered in normal reload cores. The safety-related neutronic impact of axial blankets is that of altering the axial burnup and isotopic distributions of the fuel assembly. Howcycr, all reload corcs feature widely varying axial burnup and isotopic distributions, regardless of the use of axial blankets. In addition, since the blankets are relatively small and are located in regions of low flux, their impact on safety is negligible. Axial blankets are further described in WCAP-10444-P-A.
DEBRIS-FILTER BOITOM NOZZLES The DFBN is a revised version of the current 17 x 17 nozzle design. The revised design includes an improved pattern of flow holes which:
- 1. Reduces the passage of debris into the fuel assembly,
- 2. Maintains the structural integrity of the current nozzlo design,
- 3. Maintains the hydraulic performance of the current design.
The revised diamotor flow hole pattern includes an increased number of smaller holes (190 mils vs. 360 mils) that reduces the possibility of debris entering the active region of the fuel assembly. Based on extensive testing, the DFBN can reduce the passage of debris into the fuel region by 90% or greater. This reduction of debris is expected to decrease the incidence of debris-related fuel rod failures, and ultimately reduce RCS coolant activity. Since the DFBN is designed to maintain the hydraulic proporties of the current nozzle, its use has no adverso impact on safety.
ENHANCED PERFORMANCE ROD CLUSTER CONTROL ASSEMBLIES Two W9stinghouse EP-RCCAs have boon procured for use in Cycle 4.
These EP-RCCAs retain the basic features of the original RCCAs while adding onhancements to extend service life (i.e. high purity stainless stool tubing and wear resistant coating). Thoso RCCAs meet the functional and design critoria of the original RCCAs. The major improvement to the EP-RCCAs is a thin chromo electroplate applied over the length of t':tc rodlet cladding.
This coating is more resistant to cladding wear against reactor internals guides and should increase Callaway RCCA service life.
The small increase in outer diamotor of the rodlots due to plating is not expected to increase rod drop timos significantly.
In fact, Prairic Island Unit 2 showed no change in drop times after having installed Enhanced Performance RCCAs with chromo plating.
)
I I- REALLOCATION OF GENERIC DNBR MARGIN The Cycle 4 Technical Specification Bases will'also discuss a reallocation of generic DNBR margin. This is being donc to increase the Callaway ITDP core power uncertainty from 1.1% to l
2.0%. Sufficient gencric DNBR margin currently exists to compensate for this increase in-the core power uncertainty. This
. reallocation will permit the use of an optional steam flow l
calorimetric methodology at the callaway Plant.
The Cycle 4 licensing submittal consists of the following attachments:
Attachment 1 - Cycle.4 Reload Description
. , Attachment 2 - Safety Evaluation for Increased Peaking Factors Attachment 3 - Safety Evaluation for Positive Moderator Temperature coefficient, Increased RWST/ Accumulator Boron Concentrations, and Increased SAT Sodium Hydroxide Concentration Attachment 4 - Cycle 4 Technical Specification Changes Attachment 5 - Significant Hazards Evaluation for Cycle 4 Attachment 6 - Draft FSAR Changes for Cycle 4 i
l l
l 1
1 I
a
Attachment 2 ULNRC- 1850 SAFETY EVALUATION FOR INCREASED PEAKING FACTORS
TABLE OF CONTENTS Section lltli E191
1.0 INTRODUCTION
AND
SUMMARY
1 2.0 NUCLEAR DESIGN EVALUATION 3 3.0 THERMAL-HYDRAULIC DESIGN EVALUATION 6 4.0 ACCIDENT EVALUATION 8 4.1 Non-LOCA Accidents 8 4.2 LOCA Accidents 17
5.0 REFERENCES
19 1
4317F:6-881021
. l
1.0 INTRODUCTION
AND
SUMMARY
This report presents the safety evaluation performed for the Callaway Plant-operating with increased F-delta-H and Fn (z) peaking factors. Increases in the measured F-delta-H peaking factor from 1.49 to 1.59 and Fn (z) from 2.32 to 2.50 allow for more fleFibility in the fuel management schemes i.e., longer fuel cycles, improvement of fuel economy and neutron utt11zation. Both the 1.59 F-delta-H (1.65 with measurement uncertainty) and 2.50 Fg(z) va1ues are consistent with the values used in the Cycle 3 Licensing Submittal (ULNRC-1470, 3/31/87) for the Callaway Plant, but were not licensed for the Cycle 3 reload.
The incidents analy;:ed and reported in the Callaway FSAR which could be potentially affected by the increr.ses in F-delta-H and F (z) 9 have been evaluated and are discussed in this report.
The results of the evaluations / analyses described herein lead to the following conclusions:
- 1. The changes in the Technical Specification measured F-delta-H limit from 1.49 to 1.59 is supported by current DNB analyses. All DNBR margins and analyses are as described in the current Callaway Plant FSAR (Reference 1) and are unaffected by this change in F-delta-H.
- 2. The changes in the Technical Specification 9F (z) limit from 2.32 to 2.50 has been evaluated to ensure that the cladding integrity and fuel meiting at the "hot spot" are maintained within the applicable safety analysts limits. All safety analyses as described in the current Callaway Plant F3AR (Reference 1) are unaffected by this change in F (z).
n 4317F:6-881021 1
- 3. The proposed changes to the Technical Specifications do not affect the VANTAGE 5 design / safety bases used in the VANTAGE 5 Reference Core Report (NCAP-10444-P-A) or those contained in subsequent reload safety evaluations.
4317F:6-881021 2
2.0 NUCLEAR DESIGN EVALUATION 2.1 F_-Delta-H Increase The limit on the nuclear enthalpy rise hot channel factor, F-delta-H, will take the following form in the Technical Specifications:
(1)
F$H I 1.59 (1 + 0.3(1-P))
where P - THERHAL POWER / RATED THERMAL POWER, and Ffy - Measured values of F-delta-H obtained by using the movable incore detector to obtain a power distribution map with appropriate uncertainties.
The increase in the measured radial peaking factor limit (1.49 to 1.59) will allow additional flexibility for fuel management and for determining core loading patterns. The new limit is applicable to VANTAGE 5 fuel and made possible by the additional thermal margins provided by the Intermediate Flow Hlxing (IFH) grids.
A few 0FA assemblies (approximately nine) will remain in the core in the upcoming cycle, Cycle 4. The DNB analyses for these assemblies are based on the current measured F-delta-H Technical Specification limit of 1.49. These assemblies will reside in the core for their third cycle and are of sufficiantly low reactivity that they will always operate at powers substantially below the current F-delta-H limit. The Callaway Plant Cycle 4 loading pattern has been modeled and the OFA assemblies will operate below an F-delta-H value of 1.20. Rather than implement more complex dual Technical Specification limits, with one limit applicable to VANTAGE 5 fuel and the other applicable to 0FA/LOPAR fuel, a single limit appropriate to VANTAGE 5 fuel will be provided. This recognizes the fact that the remaining 0FA assemblies, or any other OFA/LOPAR assemblies previously discharged from the core, are of sufficiently low reactivity that they cannot physically operate at powers near the current 1.49 measured radial peaking factor Ilmit. This fact is verified analytically during the normal reload design evaluation.
4317F:6-881021 3
4 As always, cycle specific reload core analysis performed in accordance with the methodology described in Reference 2 demonstrates that the new radial peaking factor Ilmit is met. No changes to the current methodology are required as a result of this Technical Specification change.
The increased radial peaking factor limit may, depending on the cycle specific fuel loading pattern, result in increases in the stuck rod F-delta-H value for steamline break accidents. The effect of this increased stuck rod peaking factor has been evaluated in the DN8R analysis for this event. No other key safety parameters used as input to the FSAR Chapter 15 accident analyses are impacted by the increased radial peaking factor limit.-
2.2 [g(z) Increase The limit on the heat flux hot channel factor, F n (z), will take the following form in the Technical Specifications:
F (z) 1 (2.50/P) * (K(z)) for P > 0.5, and n
Fg (z) 1 (5.00) * (K(z)) for P 1 0.5 where P = THERHAL POWER / RATED THERHAL P0HER, and K(z) - the function obtained from Figure 3.2-2 in the Callaway Technical Specifications for a given core height location. A revised K(z) function applicable to Callaway with the increased F (z) ilmit is provided in Attachment 4 of this 9
submittal.
The increased total peaking factor limit will allow additional flexibility in fuel management and core operation as well as accommodate the increased radial peaking factor Ilmit and introduction of fuel incorporating axial blankets.
The increased radial peaking factor discussed in Section 2.1 will result in increases in the total peaking factor, F (z), experienced in the Callaway n
core. The introduction of fuel using natural urantum axial blankets may also result in slight increases in the axial component of the total peaking
)
4317F:6-881021 4
_ _ _ - - _ _ _ _ _ _ _ _ )
factor. Only minimal increases are expected from this, since reduced length Integral Fuel Burnable Absorbers (IFBAs) will also be incorporated into the fuel design to reduce axial peaking factors.
Three cycles of representative core calculations for the Callaway VANTAGE 5 licensing submittal (Reference 3) indicate that the new increased F9(z) limit will be met for Callaway reload cores operating with axial blankets and the increased radial peaking factor F-delta-H limit. Actual Callaway reload cores will employ the usual methods of enrichment variation and burnable absorber usage to ensure compilance with the new Technical Specification peaking factor limits. As always, cycle specific reload core analysis performed in accordance with the methodology described in Reference 2 will demonstrate that the new total peaking factor Ilmit will be met. No changes to the current methodology will be required. The increased total peaking factor limit will have no impact on other key safety parameters used as input to the FSAR Chapter 15 accident analyses.
An additional minor change impacts the Axial Flux 01fference Technical Specification. The current Specification requires that the target axial flux difference be updated at least once per 31 Effective Full Power Days. This ,
can be done by re-measuring the target flux difference of each operable excore channel (which must be done at least once per 92 Effective Full Power Days) or by linear interpolation between the most recently measured value and 0% at the end of cycle life. Since reload cores typically have a target axial flux difference slightly more negative than 0%, this spec?fication is modified to allow remeasuring or linear interpolation between the most recently measured value and the calculated value at the end of cycle life. The calculated value is determined during the nuclear design evaluation of the reload cycle and is contained in the Callaway Nuclear Design Report.
4317F:6-881021 5
3.0 THERMAL-HYDRAULIC DESIGN EVALUATION 3.1 Introduction The proposed Callaway Technical Specification basis change which impacts the DNBR calculations is the value of F-delta-H determined from equation (1) given in Section 2.1. The measured radial peaking factor limit increase from 1.49 to 1.59 has a direct impact on DNBR calculations.
The core limits of the Tecnnical Specification Figure 2.1-1 include a restriction that the average enthalpy at the vessel exit must be less than the enthalpy of saturated liquid to assure the proportionality between AT and core power. The exit enthalpy restriction is more limiting than DNBR at low heat fluxes and is independent of the radial peaking factor as shown in the following relation:
h - h + <h sat (2) out in where h out - average coolant Cnthalpy at vessel exit (BTU /lb,)
h gy - vessel inlet coolant enthalpy (BTU /lb,)
Q - total power (BTU /hr) l G - total core coolant flow (Ib,/hr) 3.2 Justification for Increastna F-delta-H The Callaway Cycle 4 core will consist of over 907. of the Hestinghouse VANTAGE 5 type fuel assemblies, and the remainder (approximately 9 fuel assemblies) will be of the Optimized Fuel Assembly design. DNB analyses performed for the VANTAGE 5 fuel assemblies as part of the Callaway VANTAGE 5 Itcensing submittal were based on a peak measured F-delta-H limit of 1.59, therefore, the change in the Technical Specification limit on measured F-delta-H from 1.49 to 1,59 is supported by the current DNB analyses, and all DNBR margint [
and analyses are unaffected by this change.
4317F:6-881021 6
DNB safety analyses for the remaining fuel assemblies are based on the current Technical Specification limit of 1.49. As described in Section 2.0, the optimized fuel assemblies available for possible insertion into future Callaway cores (in the event of unforeseen fuel loading problems) have achieved burn-up levels which restrict the peak power levels they can achieve. As part of the reload process, verification of 0FA design limits is performed. Based on expected peaking factors for the Ca.!!nway optimized fuel assemblies, no violation of the current DNB design ' oasis for the optimized fuel will occur.
The core limit curves (Technical Specification Figure 2.1-1) remain valid for both VANTAGE 5 fuel assemblies at a measured 1.59 F-delta-H and optimized fuel assemL11es at the design measured peaking factor of 1.49.
3.3 Conclusions In summary, the effect of increasing the 100% tated power measured F-delta-H from 1.49 to 1.59 has already been accommodated in the DNB safety analyses of the VANTAGE 5 fuel assemblies and does not result in any change in DNBR wargins for Callaway. There is no impact on the few remaining optimized fuel assemblies, since the burn-up levels these assemblies have achieved precludes their reaching peaking factors beyond the current safety analysis limit of 1,55 (corresponding to a measured limit of 1.49).
4317F:6-881021 7
4.0 ACCIDENT EVALUATION This section summarizes the effect of the increase in the F-delta-H and F (z) Technical Specification limits for the Callaway Plant on the non-LOCA 9
and LOCA design-basis safety analyses.
4.1 Non-LOCA Accidents 4.1.1 F-delta-H Increase An increase in full power F-delta-H limit does not directly affect the system transient response of the non-LOCA events presented in the Callaway Plent FSAR. Rather, the normal operation F-delts-H limit is used in the determination of the DNBR for those events for which DNB is a safety acceptance criterton. An increase in F-delta-H results in a decrease in the DNBR value for a given set of thermal-hydraulic conditions. However, as noted in Section 3.2, this effect has already been accounted for in the Callaway FSAR non-LOCA analyses performed to support the introduction of VANTAGE 5 fuel in Cycle 3. These analyses are documented in the current Callaway Plant FSAR (Reference 1).
This section summartzes the applicability of the current Callaway Plant FSAR Chapters 15 and 6 (Reference 1) in support of an increase in the Technical Specification 11mit for measured F-delta-H from 1.49 to 1.59. Also addressed is the impact of this Technical Specification change on the Category I results of WCAP-10961-P, (Reference 4).
Overtemoeratutt_And overoortLSelta-T Protectjon (FSAR Section 15.0)
As noted in Section 3.2, the current Callaway Plant Technical Specification (Figure 2.1-1) core Ilmit curves are valid for VANTAGE 5 fuel assemblies at a 1.59 measured F-delta-H and optimized fuel assemblies at the design measured peaking factor of 1.49. Furthermore, it is noted in Sections 2 and 3 that the nine optimized fuel assemblies to be loaded into Cycle 4 are of sufficiently low reactivity that they cannot operate above the current Technical Specification limit for F-delta-H. On this basis, the current Technical 4317F:6-88102) B
Specification Overtemperature and Overpower Delta-T (OTDT/0PDT) setpoint equation constants continue to protect the core safety ilmits as shown in Figure 15.0.1 of the Callaway Plant FSAR. The system transient responses for the FSAR non-LOCA events that rely on OTDT/0PDT for protection are not affected. Therefore, the current Callaway Plant Technical Specification values for the OTDT/0PDT setpoints remain valid for the increase in the F-delta-H Technical Specification limit.
Feedwater System Halfunctions That Result in a Decrease in Feedwater Temoerature (FSAR Section 15.1.1)
This ANS Condition 11 event is bounded by "Fecdwater System Halfunctions that Result in an Increase in feedwater Flow" (15.1.2). The safety analysis DNB design basis is met and the conclusions of the FSAR remain valid for the increased F-delta-H Technical Specification.
Feedwater System Halfunctions That Result in an Increase in Faedwater Flow (FSAR Section 15.1.2)
For this ANS Condition II event, cases are analyzed for both full power and zero power conditions. The zero pcwer cue, as discussed in the FSAR, is bounded by "Uncontrolled RCCA Bank Hithdrawal from a Subcritical or Low Power Startup Condition" (15.4.1). For the full power case, the transient is effectively terminated by a turbine trip and feedwater isolation on high-high steam generator level. The core Ilmits applicable for the OFA fuel at the 1.49 measured F-delta-H design limit and the VANTAGE 5 fuel at the increased Technical Specification limit are not exceeded at the most Ilmiting point in the transient. Therefore, the safety analysis DNBR limits (i.e., for OFA and
! VANTAGE 5) are met and the conclusions of the FSAR remain valid.
l EActisive IncitA1e.in Se n dary Steam Flow (FSAR Section 15.1.3) i l
For this ANS Condition II event, cases are analyzed at beginning and end of l life condittoas both with and without automatic rod con'rol. In all cases,
( the transient approaches an equilibrium condition and a ceactor trip does not l result. The core limits applicable for the OFA fuel at tie 1.49 measured l
4317F:6-881021 9 t
4 F-delta-H design Iln.it and the VANTAGE 5 fuel at the increased Technical Specification Ilmit are not exceeded at the most Ilmiting point in the transient. Therefore, the safety analysis DNBR limits are met and the conclusions of the FSAR remain valid.
Inadvertent Ooenino of a Steam Generator Relief or Safety Valve (FSAR Section 15.1.4)
This ANS Condition II event is bounded by "Steam System Piping Failure" (15.1.5). The safety analysis DNB design basis is met and the conclusions of thc FSAR remain valid.
ligam System Pioina Failure (FSAR Section 15.1.5)
For this ANS Condition IV ever t, the ANS Conditiof II minimum DNBR limit criterion is applied. The analyses are performed assuming zero power initial conditions and peaking factors consistent with the most reactive RCCA stuck out of the core. An increase of the full power F-delta-H limit results in an increase of the stuck RCCA peaking factor. The effects of the increased stuck rod F-delta-H value have been evaluated. The safety analysis DNBR limits are met and the conclusions of the FSAR remain valid.
Loss of External Electrical Load (FSAR Section 15.2.2)
This ANS Condition II event is bounded by "Turbine Trip" (15.2.3). The safety analysis DNB design basis is met and the concluslotas of the FSAR remain valid.
Isrbine Trio (FSAR Section 15.2.3)
For this ANS Condition II event, cases are analyzed at beginning and end of life conditions both with and without pressurizer control. For these cases, the transient is terminatad by either reactor trip on high pressurizer pressure or overtemperature delta-T. The core limits applicable for the OFA fuel at the 1.49 measured F-delta-H design limit and the VANTAGE 5 fuel at the increased Technical Specification Ilmit are not exceeded at the most limiting point in the transient. Therefore, the safety analysis DNBR limits are met and the conclusions of the FSAR remain valid.
4317F:6-881021 10
Inadvertent Closure of MSIVs (FSAR Section 15.2.4)
This ANS Condition II event is bounded by the "Turbine Trip" event (15.2.3).
The safety analysis DNB design basis is met and the conclusions of the FSAR remain valid.
Loss of Non-Emeroency AC Power to the Station Auxiliaries (FSAR Section 15.2.6)
This ANS Condition II event is analyzed to show that adequate heat removal capability exists to remove core decay heat and stored energy following reactor trip. This criterion is not affected by the F-delta-H limit. Alth respect to the ONB criterion, this event is bounded by "Complete Loss of Forced Reactor Coolant Flow' (15.3.2). The safety analysis ONB design basis is met and the conclusions of the FSAR remain valid. This statement is equally applicable for the Callaway loss of AC Power results reported in NCAP-11884 (Reference 5).
Loss of Normal Feedwater Flow (FSAR Section 15.2.7)
This ANS Condition II event is analyzed to show that adequate heat removal capability exists to remove core decay heat and stored energy following reactor trip. This criterion is not affected by the F-delta-H limit. With respect to the DNB criterton, this event is bounded by the "Turbine Trip" event (15.2.3). The safety analysis DNB design basis is met and the FSAR conclusions remain valid. This statement is equally applicable for the Callaway Loss of Normal Feedwater results reported in NCAP-11884 (Reference 5).
Etthittr__Sy_stafrLP_ int _BLtAh (FSAR Section 15.2.8)
This is an ANS Condition IV event, thus the minimum DNBR limit acceptance criterton is not applicable. Therefore, the increased F-delta-H limit does not impact the analyses and the conclusions of the FSAR remain valid. This statement is equally applicable for the Callaway feedline Break transient results reported in NCAP-11884 (Referenct 5).
4317F:6-881021 11
Partial loss of Forced Reactor Coolant Flow (FSAR Section 15.3.1)
For this ANS Condition II event, the FACTRAN code is used to calculate the core hot channel heat flux transient based upon nucitar power and flow from LOFTRAN. The FSAR analysis, applicable to 0FA and VANTAGE 5 fuel, assumed a full power F-delta-H safety analysis limit of 1.65 for the transient heat flux calculation. This safety analysis assumption is documented in Table 15.0-2 of the FSAR.. The transient is terminated by a low RCS loop flow reactor trip and the core limits applicable for the OFA fuel at the 1.49 meacured F-delta-H design limit and the VANTAGE 5 fuel at the increased Technical Specification limit are not exceeded at the mest limiting point in the transient.
Therefore, the safety analysis DNBR Ilmits are met and the conclusions of the FSAR remain valid.
Comolete Loss of Forced Rgaetor Coolant Flow (FSAR Section 15.3.2)
For this ANS Condition III event, the FACTRAN code is used to calculate the core hot channel heat flux transient based upon nuclear power and flow from LOFTRAN. The FSAR analysis, applicable to 0FA and VANTAGE 5 fuel, assumed a full power F-delta-H safety analysis limit of 1.65 for the transient heat flux calculation. This safety analysis assumption is documented in Table 15.0-2 of the FSAR. The transient is terminated by reactor trip on reactor coolant pump undervoltage and the core Ilmits applicable for the OFA fuel at the 1.49 measured F-delta-H design limit and the VANTAGE 5 fuel at the increased Technical Specification limit are not exceeded at the most limiting point in the transient. Therefore, the safety analysis ONBR limits are met and the conclusions of the FSAR remain valid.
Et1Clor Coolant Pumo Shaft Selzure (Locked Rotor) (FSAR Section 15.3.3)
This is an ANS Condition IV event analyzed for determination of peak RCS pressure and peak fuel clad temperature assuming DNB to occur in the core.
The FACTRAN code is used to calculate the core hot spot heat flux transient based upon nuclear power and flow from LOFTRAN. The FSAR analysis, applicable to 0FA and VANTAGE 5 fuel, assumed a full power F-delta-H safety analysis limit of 1.65 for the transtent heat flux calculation used to evaluate the i percentage of fuel rods in DNB. This safety analysis assumption is documented 4317F:6-881021 12
in Table 15.0-2 of the FSAR. The calculated percentage of fuel rods in DNB remains within the current safety analysis limit of 51., Therefore, the FSAR Locked Rotor safety analysis supports the Technical Specification F-delta-H limit increase.
Reactor Coolant Pumo Shaft Break (FSAR Section 15.3.4)
This ANS Condition IV event is bounded by "Reactor Coclant Pump Shaft Seizure" (15.3.3). Therefore, the conclusions of the FSAR remain valid.
Uncontrolled RCCA Bank Withdrawal from a Subtritical or Low Power Startuo Condition (FSAR Section 15.4.1)
This ANS Condition II event is performed at zero power conditions and is terminated by reactor trip on the power range high neutron flux low setpoint.
The core limits applicable for the OFA fuel at the 1.49 measured F-delta-H design limit and the VANTAGE 5 fuel at the increased Technical Specification limit are not exceeded at the most 11miting point in the transient.
Therefore, the safety analysis DNBR limits are met and the conclusions of the FSAR remain valid.
Uncontrolled RCCA Bank Withdrawal at Power (FSAR Section 15.4.2)
For this ANS Condition II event, various power levels and reactivity insertion rates for both minimum and maximum reactivity feedback are analyzed. The transients are terminated by an overtemperature delta-T or high neutron flux l
reactor trip. The core limits applicable for the OFA fuel at the 1.49 l measured F-delta-H design Ilmit and the VANTAGE 5 fuel at the increased Technical Specification limit are not exceeded at the most limiting point in the transient. Therefore, the safety analysis DNBR limits are met and the
, conclusions of the FSAR remain valid.
l l
l RCCA Misoperation (FSAR Section 15.4.3) l RCCA Hisoperation is categorized into four types of events. Three of these are classified as ANS Condition II events: dropped RCCA, dropped hCCA bank, and statice.11y misaligned RCCA. The fourth, single RCCA withdrawal, is 4317F:6-881021 13 ,
~
classified as an ANS Condition III event. The effects of the increased F-delta-H peaking factor have b)en evaluated for thw DNB acceptance criteria.
The conclusions in the FSAR are verified. Specifically, the DNS acceptance criterion is met for the Conditi n9 II events and the calculated number of fuel rods experiencing DNB is within the current safety analysis limit of 5% for the Condition III event.
Startuo of an Inactive Reactor Coolant Pumo at an Incorrect Temoerature (FSAR Section 15.4.4)
For this ANS Condition II event, the FACTRAN code is used to calculate the heat flux transient based upon nuclear power and flow from LOFTRAN. The FSAR !
analysis, applicable to 0FA and VANTAGE 5 fuel, assumed an F-delta-4 consistent with a full power safety analysis limit of 1.65 for the transient heat flux calculation. This safety analysis assumption is documented in Table 15.0-2 of the FSAR. The transtent is terminated by reactor trip on the high neutron flux P-8 trip setpoint. The core limits applicable for the OFA fuel i at the 1.49 measured F-delta-H design limit and the VANTAGE 5 fuel at the increased Technical Specification limit are not exceed:d at the most limiting point in the transient. Therefore, the safety analysis DNBR limits are met and the conclusions of the FSAR remain valid.
CVCS Halfunction That Results in a Decrease in the Boron Concentration in the Reactor Coolant (FSAR Section 15.4.6)
' This ANS Condition II event is analyzed to show that adequate time exists for
' automatic or operator action to terminate an inadvertent dilution prior to a loss of shutdown margin. The transtent calculation does not include an explicit evaluation for the Condition II DNB acceptance criterion. The i
analysis assumptions do not depend on F-delta-H. Therefore, the FSAR analysis
! results and FSAR conclusions are unaffected by the increase in the Technical
! Specification limit for F-delta-H.
i 4317F:6-881021 14 I
Inadvertent Loadina and Ooeratina with a Fuel Assembly in Imoroner Position (FSAR Section 15.4.7) 1he increased F-delta-H has no impact on this transient. Therefore, the FSAR conclusions remain valid.
Spectrum of RCCA Eiection Accidents (FSAR Section 15.4.8)
This is a Condition IV event, thus the minimum DNBR limit acceptance criterton does not apply. The increased F-delta-H limit does not affect the FSAR analyses and the FSAR conclusions remain valid.
Inadvertent Ooeration of Er _raency Core Coolina System (ECCS) Durina Foytt Ooeration (FSAR Section 15.5.1)
For this ANS Condition II event, the transient is initiated by a spurious i
safety injection signal. The injection of borated water drives nuclear power and RCS temperature down. Hence, the most limiting thermal-hydraulic condition occurs at the initiation of the transient. Therefore, the safety analysis DNBR limits are met and the conclusions of the FSAR remain valid.
CVCS Halfunction,that Increases Reactor Coolant Inventory (FSAR Section 15.5.2) l For this ANS Condition II event, the transient is initiated by operator error or control signal malr' unction. The resulting increase in reactor coolant inventory is characterized by increasing pressurizer level, increasing l pressurizer pressure, and constant boron concentration. The transient is analyzed to demonstrate that there is adequate time for operator mitigation of pressurizer filling. The transient assumptions are chosen to minimize available operator action time. The resulting DNBR transient would be bounded by CVCS ms1 functions which result in the inadvertent decrease in ACS boron concentration (addressed in Section 15.4.6). The safety analysis assumption -
for F-delta-H would not affect the transtent pressurizer water volume; therefore, the FSAR conclusions remain valid for the proposed Technical Specification revision.
4317F:f-881021 15
~
Inadvertent Ocenina of a Pressurizer Safety or Relief Valve (FSAR Section 15.6.1)
For this ANS Condition II event, the transient is terminated by an overtemperature delta-T reactor trip. The core limits applicable for the OFA fuel at the measured 1.49 F-delta-H design limit and the VANT%iE 5 fuel at the increased Technical Specification limit are not exceeded ;c the most limiting point in the transient. Therefore, the safety siidlysis CNBR limits are met and the conclusions of the FSAR remain valid.
Steam Generator Tube Ruotutt (FSAR Section 15.6.3)
This is a Condition IV event, thus the minimum DNBR limit acceptance criterion ,
does not apply. The increased F-delta-H limit does not affect the SGTR analyses and the conclusions of References 6 and 7 remain valid.
Mass _and Enerav Release Analysis for Postulated Secondary Ploe Ruotures Inside Costainment (FSAR Section 6.2.1.4)
The limiting Steamline Break transient for DNB evaluation is found in FSAR Section 19 1.5. The transient analyzed for Section 6.2.1.4 of the FSAR is des'gned to max' size break mass and energy releases ins de containment. The analysis assumptions are not dependent on F-delta-H. Therefore, the conclusions in the FSAR remain valid.
S.itamline BrtALB13s/Enerav Releasts for Egulpment Environmeat&LQualification Outside containmtat (HCAP-10961-P)
The limiting Steamline Break transient for DNB evaluation is found in FSAR .
Section 15.1.5. The transient analyzed for NCAP-10961 is designed to maximize
! break mass and energy releases for EQ outside containment. The analysis assumptions are not dependent on F-delta-H. Therefore, the NCAP results l applicable to the Callaway Plant (Category I) remain valid for the increase in j
the Technical Specification F-delot-H limit.
F 4317F:6-881021 16 l __
j
! 4.1.2 Fg (z) Increase :
This section summarizes the applicability of the current Callaway Plant j non-LOCA FSAR safety analyses to suppo-t a revised Technical Specification j 11mit for peak Fg (z) as a function of core height. Certain non-LOCA transients are sensitive to the maximum nF (z) limit value. The revised j
, Technical Specification incresses the maximum value from 2.32 to 2.5. This l
- effect has already been accountre for in the Callaway Plant non-LOCA analyses (
performed to support the intrc' duction of VANTAGE 5 f!iel in Cycle 3. These l
analyses are documented in the current Callaway Plant FSAR (Reference 1). {
) Two non-LOCA FSAR Chapter 15 accidents incorporate the F (r) limit into the n
, safety analysis assumptions. These are the Section 15.3.5 Reactor Coolant ;
I Pump Sheft Seizure and Section 15.4.8 Spectrum of RCCA Efection Accidents l j transients. These Condition IV events are analyzed to ensure that clad l j integrity and fuel melting acceptance criteria are met. For both of these events, Reference 1 documents that all appitcable safety analysis acceptance ,
criteria are met incorporating assumptions which support a maximum Technical :
Specification F n (z) limit of 2.5. All other non-LOCA Jes*gn basis ;
calculations are unaffected by this change. j l
4.2 LQCA ACCIDENTS t
The Small and large Break LOCA analyses performed as part of the VANTAGE 5 !
Itcensing submittal to support the VANTAGE 5 reloads were based on a peak F (z) of 2.50 at the core mid-planc (6 feet) for the t.arge Break LOCA n
analysis and 2.31 at the top of the core (12 feet) based on an Fg of 2.375 at 10 feet for the limiting Small Break LOCA power shape (See FSAR Figure l 15.6-45). In the Large Break LOCA analysis to suppo't the VANTAGE 5 reloads, [
F-delta-H was based on a peak of 1.f.5 with a choppoc'. cosine power shape. The Small Break LOCA analysis was also performed with a? F-delta-H of 1.65. The LOCA Ilmit of 1.65 for F-delta-H corresponds to the measured To:hnical i Specification Ilmit of 1.59 for the measured Enthalpy Rise Hot Channel -
f Factor. On this basis, the changes in the Technical Specification limits for l the Heat Flux Hot Chennel Factor (F (z) and the Nuclear Enthalpy Rtse Hot ;
9 Channel Factor (F-delta-H) are supported by the current VANTAGE 5 1.0CA !
analyses (Reference 3).
t 4317F:6-881021 17 i
j
The Callaway Cycle 4 core will consist of approximately 184 VANTAGE 5 fuel assemblies and 9 0FA fuel assemblies. The Small and Large Break LOCA analyses for the remaining 0FA fuel are based on the current safety analysis limits of F (z) of 2.32 and F-delta-H of 1.55. The OFA fuel planned for insertion n
into the Cycle 4 core has experienced burnup, reducing the stored energy. It has been determined that the peak power levels of these assemblies are restricted due to the secrued burnup. Based on the expected lower power levels of the OFA fuel, no violations of the large and Small Break LOCA analyses will occur.
In summary, the effect of increasing F g(2) and F-delta-H has already been accommodated in the Large and Small Break LOCA Analyses for the VANTAGE 5 fuel. Also, there is no adverse effect on the remaining 0FA fuel, since the burn-r3 levels preclude reaching peaking factors beyond current LOCA limits.
(
I i
i i
I 4317F:6-881021 18
~
5.0 REFERENCES
- 1. Union Electric Company, Callaway Plant, Final Safety Analysis Report, Docket Number 50-483 Revision OL-2,
- 2. Davidson, S. L. (Ed.), et. al., "Westinghouse Reload Safety Evaluation Methodology," WCAP-9272-P-A, July 1986.
- 3. ULNRC-1470, dated 3/31/87, Cycle 3 Licensing Submittal.
- 4. Butler, J. C., Love, D. S., et al., "Steamline Break Mass / Energy Releases for Equipment Environmental Qualtftcation outside ContairN nt,"
NCAP-10961-P, Revision 1, October 1985.
- 5. Leach, C. , Gongaware, B., Tuley, R. , "Implementation of the Steam i
Generator Low-Low Level Reactor Trip Time Delay and Environmental Ailowance Modifier in the Callaway Plant," WCAP-11884, August 1988, submitted by ULNRC-1822 dated 8/30/88.
I
- 6. SLNRC 86-01 dated 1/8/86.
- 7. ULNRC-1518 dated 5/27/87.
i 4317F:6-881021 19
^
Attachment 3 ULNRC-1850 SAFETY EVALUATION FOR POSITIVE MODERATOR TEMPERATURE COEFFICIENT, INCREASED RWST/ ACCUMULATOR BORON CONCENTRATIONS, AND INCREASED SAT SODIUM HYDROXIDE CONCENTRATION
_ . _ . .