ML053550032

Handout of Draft Proposed Change #1 Technical Specifications Bases Changes from 12/12/05 Meeting on Safety Limit 2.1.1.1, Part 21

ML053550032
Person / Time
Site:	Hatch
Issue date:	12/12/2005
From:	BWR Owners Group, Southern Nuclear Operating Co
To:	Office of Nuclear Reactor Regulation
References
NUREG-1433, NUREG-1434
Download: ML053550032 (8)

Text

DRAFT PROPOSED CHANGE #1 TECHNICAL SPECIFICATIONS BASES CHANGES (Handout for December 12,2005 Meeting on Safety Limit 2.1.1.1, Part 21)

Reactor Core SLs B 2.1.1 BASES APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY operation and AOOs. The reactor core SLs are established to preclude ANALYSES violation of the fuel design criterion that an MCPR limit is to be established, such that at least 99.9% of the fuel rods in the core would not be expected to experience the onset of transition boiling.

Reactor Protection System setpoints (LC0 3.3.1 . l , "Reactor Protection System (RPS) Instrumentation"), in combination with the other LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System water level, pressure, and THERMAL POWER level that would result in reaching the MCPR limit.

2.1.1 .la Fuel Claddinq lnteqritv [General Electric Comanv (GE) Fuel1 GE critical power correlations are applicable for all critical power calculations at pressures 2 785 psig and core flows 2 10% of rated flow.

For operation at low pressures or low flows, another basis is used, as follows:

Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be > 4.5 psi. Analyses (Ref. 2) show that with a bundle flow of 28 x lo3 Iblhr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be > 28 x lo3 Iblhr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt.

With the design peaking factors, thiscorresponds to a THERMAL POWER > 50 % RTP. Thus, a THERMAL POWER limit of 25% RTP for reactor pressure < 785 psig is conservative.

2.1 .I.Ib Fuel clad din^ lntearitv [Advanced Nuclear Fuel Cor~oration

/ANF) Fuel]

The use of the XN-3 correlation is valid for critical power calculations at pressures > 580 psig and bundle mass fluxes > 0.25 x lo6 lblhr-ft2 (Ref. 3). For operation at low pressures or low flows, the fuel cladding integrity SL is established by a limiting condition on core THERMAL POWER, with the following basis:

Provided that the water level in the vessel downcomer is maintained above the top of the active fuel, natural circulation is sufficient to ensure a minimum bundle flow for all fuel assemblies that have a relatively high power and potentially can approach a critical heat flux condition. For the ANF 9x9 fuel design, the minimum bundle flow is > 30 x 10' Iblhr. For the ANF 8x8 fuel design, the minimum bundle flow is > 28 x lo3 Iblhr. For all designs, the coolant minimum bundle flow and maximum flow area are BWW4 STS B 2.1.1 -2 Rev. 3.0,03131104

There is an exception to the applicability of SL 2.1.1.1 for depressurization AOOs, for example the pressure regulator failure -open (PRFO) event in Chapter [ I 51 of the FSAR.

In depressurization transients, critical bundle power increases and actual bundle power decreases, which tends to increase the critical power ratio, (Reference 5). This provices more margin to the conditions that would produce the onset of transition boiling.

Consequently, SL 2.1.1.1 does not apply to AOOs which result in reactor depressurization.

Reactor Core SLs B 2.1 .I BASES SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to the release of radioactive materials to the environs.

SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations.

APPLICABILITY SLs 2.1.1 .I, 2.1 .I.2, and 2.1 .I.3 are applicable in all MOD SAFETY LIMIT Exceeding an SL may cause fuel damage and create a potential for VIOLATIONS radioactive releases in excess of 10 CFR 100, "Reactor Site Criteria,"

limits (Ref. 4). Therefore, it is required to insert all insertable control rods and restore compliance with the SLs within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.

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

2. NEDE-24011-P-A (latest approved revision).

3. XN-NF524(A), Revision 1, November 1983.

-4. 10 CFR 100.

[

BWR14 STS Rev. 3.0, 03131104

INSERT 2 SL 2.1.1.1 is not applicable during reactor depressurization transients as discussed in the APPLICABLE SAFETY ANALYSIS section.

INSERT 3 GE Energy-Nuclear, 10 CFR Part 21 Communication, SCO5-03, "Potential to Exceed Low Pressure Technical Specification Safety Limit, March 29,2005.

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.

Main Steam Line lsolatio~

1.a. Reactor Vessel Water Level Low Low Low. Level 1 Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, isolation of the MSlVs and other interfaces with the reactor vessel occurs to prevent offsite dose limits from being exceeded. The Reactor Vessel Water Level

- Low Low Low, Level 1 Function is one of the many Functions assumed to be OPERABLE and capable of providing isolation signals. The Reactor Vessel Water Level Low Low Low, Level 1 Function associated with isolation is assumed in the analysis of the recirculation line break (Ref. 1). The isolation of the MSLs on Level 1 supporls actions to ensure that offsite dose limits are not exceeded for a DBA.

Reactor vessel water level signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level

- Low Low Low, Level 1 Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Reactor Vessel Water Level - Low Low Low, Level 1 Allowable Value is chosen to be the same as the ECCS Level 1 Allowable Value (LC0 3.3.5.1) to ensure that the MSLs isolate on a potential loss of coolant accident (LOCA) to prevent offsite doses from exceeding 10 CFR 100 limits.

This Function isolates the Group 1 valves.

i.b.Main Steam Line Pressure - Low Low MSL pressure indicates that there may be a problem with the turbine pressure regulation, which could result in a low reactor vessel water level condition and the RPV cooling down more than 100°F/hr if the pressure loss is allowed to continue. The Main Steam Line Pressure - Low Function is directly assumed in the analysis of the pressure regulator BWRl4 STS B 3.3.6.1-6 Rev. 3.0,03/31K)4

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) failure (Ref. 2). For this event, the closure of the MSlVs ensures. that

. the RPV temperature change limit (10O0Flhr)is not reached.

The MSL low pressure signals are initiated from four transmitters that are connected to the MSL header. The transmitters are arranged such that, even though physically separated from each other, each transmitter is able to detect low MSL pressure. Four channels of Main Steam Line Pressure Low Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Allowable Value was selected to be high enough to prevent excessive RPV depressurization.

The Main Steam Line Pressure - Low Function is only required to be OPERABLE in MODE 1 since this is when the assumed transient can occur (Ref. 2).

This Function isolates the Group 1 valves.

1.c. Main Steam Line Flow - Hiah Main Steam Line Flow High is provided to detect a break of the MSL and to initiate closure of the MSIVs. If the steam were allowed to continue flowing out of the break, the reactor would depressurize and the core could uncover. If the RPV water level decreases too far, fuel damage could occur. Therefore, the isolation is initiated on high flow to prevent or minimize core damage. The Main Steam Line Flow High -

Function is directly assumed in the analysis of the main steam line break (MSLB) (Ref. 1). The isolation action, along with the scram function of the Reactor Protection System (RPS), ensures that the fuel peak claddirlg temperature remains below the limits of 10 CFR 50.46 and offsite doses do not exceed the 10 CFR 100 limits.

BWRl4 STS B 3.3.6.1-7 Rev. 3.0,03/31 104