ML22285A120

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Attachment 3 to Enclosure 1: Marked-Up Technical Specifications Bases (for Information Only)
ML22285A120
Person / Time
Site: Callaway Ameren icon.png
Issue date: 10/12/2022
From:
Ameren Missouri, Union Electric Co
To:
Office of Nuclear Reactor Regulation
Shared Package
ML22285A115 List:
References
ULNRC-06768
Download: ML22285A120 (6)
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Text

Attachment 3 to to ULNRC-06768 Page 1 of 6

ATTACHMENT 3

MARKED-UP TECHNICAL SPECIFICATIONS BASES (FOR INFORMATION ONLY)

The following pages depict the changes proposed to the existing Technical Specification Bases. These pages are provided for information only with the final changes processed in accordance with the provisions of TS 5.5.14, "Technical Specifications (TS) Bases Control Progr am."

5 pages follow this cover sheet

Reactor Core SLs B 2.1.1

B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs

BASES

BACKGROUND GDC 10 (Ref. 1) requires that specified a cceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (A OOs). This is accomplished by having a departure from nucleate boili ng (DNB) design basis, which requires that the minimum departure from nucleate boiling ratio (DNBR) of the limiting rod during Condition I and II events is greater than or equal to the DNBR correlation limits.

To meet this correlation limit design basis while acc ounting for uncertainties for Westinghouse fuel, for Revised The rmal Design Procedure (RTDP) analyses, uncertainties in plant ope rating parameters, nuclear and thermal parameters, fuel fabrication par ameters, computer codes, and DNB correlation (WRB-2) predictions are comb ined statistically to obtain the overall DNBR uncertainty factor. This DNBR uncertainty factor is used to define the design limit DNBR, which corresponds to a 95% probability with 95% confidence that DNB will not occur on the limiting fuel rods during Condition I a nd II events. Since the parameter uncertainties are considered in determini ng the RTDP design limit DNBR values, the plant safety analyses are per formed using input parameters at their normal values. The design limit DN BR values are 1.21 and 1.22 for thimble and typical cells, respectively, fo r VANTAGE 5 fuel. In addition, margin has been maintained by meeting safe ty analysis DNBR limits above the design limit DNBR to offset known DN BR penalties and to provide DNBR margin for operating and design flexibi lity. Reference 3 discusses non-RTDP transients. These transients are anal yzed using the WRB-2, W-3, ABB-NV, or WLOP DNB correlation, as app licable for the specific transient. The correlation limits for WRB-2, W-3, ABB-NV, and WLOP are 1.17, 1.30, 1.13, and 1.18, respectively.

For Framatome GAIA fuel, uncertainties are statistical ly applied to the boundary conditions analyzed rather than to the DNBR de sign limit. The DNBR safety limit for GAIA fuel assemblies is 1.12 for the ORFEO-GAIA Critical Heat Flux (CHF) correlation and 1.15 for the ORFEO-NMGRID CHF correlation with COBRA-FLX using the P-SCHEME Solver. The ORFEO-NMGRID CHF correlation DNBR safety limit is 1.18 in CO BRA-FLX with the PV Solver. The DNBR safety limit is 1.12 for the ORFEO-GAIA CHF correlation and 1.15 for the ORFEO-NMGRID CHF correl ation with XCOBRA-IIIC. Reference 5, 6, 7

The restrictions of this SL prevent overheating of th e fuel and cladding, as well as possible cladding perforation, that would re sult in the release of fission products to the reactor coolant. Overheating of the fuel is (continued)

CALLAWAY PLANT B 2.1.1-1 Revision 13 Reactor Core SLs B 2.1.1

BASES

BACKGROUND prevented by maintaining the steady state peak linear heat rate (LHR)

(CONTINUED) below the level at which fuel centerline melting occu rs. Overheating of the fuel cladding is prevented by restricting fuel op eration to within the nucleate boiling regime, where the heat transfer coef ficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enou gh to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontr olled release of activity to the reactor coolant. Reference 4 further discuss es the fuel centerline temperature design basis.

Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onse t of DNB and the resultant sharp reduction in heat transfer coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally we aker form. This weaker form may lose its integrity, resulting in an u ncontrolled release of activity to the reactor coolant.

The proper functioning of the Reactor Trip System ( RTS) and steam generator safety valves prevents violation of the r eactor core SLs.

APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY operation and AOOs. The reactor core SLs are establi shed to preclude ANALYSES violation of the following fuel design criteria:

a. There must be at least 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the limiting hot fue l rod in the core does not experience DNB; and
b. The hot fuel pellet in the core must not experien ce centerline fuel melting.

The Reactor Trip System Allowable Values in Table 3.3.1-1, in combination with all the LCOs, are designed to preven t any anticipated combination of transient conditions for Reactor Cool ant System (RCS) temperature, pressure, RCS flow, DI, and THERMAL POWER level that would result in a departure from nucleate boiling ra tio (DNBR) of less than the DNBR limit and preclude the existence of flow insta bilities.

(continued)

CALLAWAY PLANT B 2.1.1-2 Revision 13 Reactor Core SLs B 2.1.1

BASES APPLICABLE Protection for these reactor core SLs is provided by th e proper operation SAFETY of the steam generator safety valves and the followi ng automatic reactor ANALYSES trip functions:

(CONTINUED) a. High pressurizer pressure trip;

b. Low pressurizer pressure trip;
c. Low reactor coolant system flow;
d. Overtemperature DT trip;
e. Overpower DT trip; and
f. Power Range Neutron Flux trip.

The SLs represent a design requirement for establishi ng the RTS Allowable Values identified previously. LCO 3.4.1, "R CS Pressure, Temperature, and Flow Departure from Nucleate Boilin g (DNB) Limits,"

and the assumed initial conditions of the safety analy ses (as indicated in the FSAR, Ref. 2) provide more restrictive limits to e nsure that the SLs are not exceeded.

SAFETY LIMITS The reactor core safety limits figure pr ovided in the COLR shows the loci of points of THERMAL POWER, pressurizer pressure, and average temperature below which the calculated DNBR is not l ess than the design limit DNBR values, the average enthalpy in the hot le g is less than or equal to the enthalpy of saturated liquid, or the e xit quality is within the limits defined by the DNBR correlation.

The reactor core SLs are established to preclude the viol ation of the following fuel design criteria:

a. There must be at least a 95% probability at a 95 %

confidence level (the 95/95 DNB criterion) that the h ot fuel rod in the core does not experience DNB; and

b. There must be at least a 95% probability at a 95 %

confidence level that the hot fuel pellet in the core does not experience centerline fuel melting.

The reactor core SLs are used to define the various R TS functions that the above criteria are satisfied during steady state operatio n, normal operating transients, and anticipated operational occurrences (A OOs). To ensure

(continued)

CALLAWAY PLANT B 2.1.1-3 Revision 13 Reactor Core SLs B 2.1.1

BASES SAFETY LIMITS that the RTS precludes the violation of the above cri teria, additional criteria (CONTINUED) are applied to the Overtemperature DT and Overpower DT reactor trip functions. That is, it must be demonstrated that th e average enthalpy in the hot leg is less than or equal to the saturation e nthalpy and that the core exit quality is within the limits defined by the DNBR correlation.

Appropriate functioning of the RTS ensures that for va riations in the THERMAL POWER, RCS pressure, RCS average temperature, RCS flow rate, and DDI that the reactor core SLs will be satisfied during steady state operation, normal operational transients, and AOOs.

Reference 4 discusses the fuel temperature design basis. Figure 15.0-1 of Reference 2 depicts the protection provided by the O verpower DDT reactor trip function against fuel centerline melting.

APPLICABILITY SL 2.1.1 only applies in MODES 1 and 2 because these are the only MODES in which the reactor is critical. Automatic protect ion functions are required to be OPERABLE during MODES 1 and 2 to ensu re operation within the reactor core SLs. The steam generator saf ety valves or automatic protection actions serve to prevent RCS hea tup to the reactor core SL conditions or to initiate a reactor tr ip function, which forces the unit into MODE 3. Allowable Values for th e reactor trip functions are specified in LCO 3.3.1, "Reactor Trip S ystem (RTS)

Instrumentation." In MODES 3, 4, 5, and 6, Applicab ility is not required since the reactor is not generating significant THERMAL POWER.

SAFETY LIMIT The following SL violation responses are applicable to the reactor core VIOLATIONS SLs. If SL 2.1.1 is violated, the requirement to go to MODE 3 places the unit in a MODE in which this SL is not applicable.

The allowed Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> recognizes the importance of bringing the unit to a MODE of operation where this SL is not applicable, and reduces the probability of fuel damage.

CALLAWAY PLANT B 2.1.1-4 Revision 13 RCS Pressure SL

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.

2. FSAR, Chapter 15. Reviewer Note: FSAR Section 4.7.4 is a new FSAR Section to capture
3. FSAR Section 4.4.1.1. Framatome GAIA fuel-specific content.
4. FSAR Section 4.4.1.2 and 4.7.4.1.2.
5. ANP-10341(P)(A), Revision 0, The ORFEO -GAIA and ORFEO-NMGRID Critical Heat Flux Correlations.
6. FS1- 0050690, Revision 3.0, ORFEO -GAIA and ORFEO-NMGRID Design Limit Validation for XCOBRA-IIIC.
7. FS1- 0058371, Revision 1.0, Validation of the ORFEO -NMGRID CHF Correlation with the PV Solver.

CALLAWAY PLANT B 2.1. 1 -5 Revision 0

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