ML053250329

Technical Specifications Bases, Revision 1

ML053250329
Person / Time
Site:	Susquehanna
Issue date:	11/08/2005
From:	Susquehanna
To:	Office of Nuclear Reactor Regulation
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-PPL Rev. 0 SDM B3.1.1 BASES REFERENCES (continued)

5. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).

6. FSAR, Section 4.3.

7. PL-NF-90-001-A, Application of Reactor Analysis Methods for BWR Design and Analysis," Section 2.4, July 1992, Supplement 1-A, August'1995, Supplement 2-A, July 1996, and Supplement 3-A, March 2001.

I TS B 3.1-7 SUSQUEHANNA - UNIT 1 i Revision 1

PPL Rev. 1 SDV Vent and Drain Valves

- B 3.1.8 B3.1 B 3.1.8 REACTIVITY CONTROL SYSTEMS I

Scram D c r VV

.on V v Scram Discharge Volume (SDV) Vent and Drain Valves BASES BACKGROUND The SDV vent and drain valves are normally open and discharge any accumulated water in the SDV to ensure that sufficient volume is available at all times to allow a complete scram. During a scram, the SDV vent and drain valves close to contain reactor water. The SDV is a volume of header piping that connects to each hydraulic control unit (HCU) and' drains into an instrument volume. There are two SDVs (headers) and two instrument volumes, each receiving approximately'one half of the control rod drive (CRD) discharges. The two instrument volumes are connected to a common drain line with two valves in series. Each header is connected to a common vent line with two valves in series. The header piping is sized to receive and contain all the water discharged by the CRDs during a scram. The design and functions of the SDV are described in Reference 1.

APPLICABLE SAFETY ANALYSES The Design Basis Accident and transient analyses assume all of the control rods are capable of scramming. The acceptance criteria for the SDV vent and drain valves are that they operate automatically to:

a.

Close during scram to limit the amount of reactor coolaint discharged so that adequate core cooling is maintained and offsite doses remain within the limits of 10 CFR 100 (Ref. 2); and

b.

Open on scram reset to maintain the SDV vent and drain path open so that there is sufficient volume to accept the reactor coolant discharged during a scram.

Isolation of the SDV can also be accomplished by manual closure of the SDV valves. Additionally, the discharge of reactor coolant to the SDV can be terminated by scram reset or closure of the HCU manual isolation valves. For a bounding leakage case, the offsite doses are well within the limits of 10 CFR 100 (Ref. 2), and adequate core cooling is maintained (Ref. 3). The SDV vent and drain valves allow continuous drainage of the SDV during normal plant operation (continued)

SUSQUEHANNA - UNIT 1

' - B 3.147 Revision 0

-1PPL Rev. 1 SDV Vent and Drain Valves

- B 3.1.8 BASES APPLICABLE SAFETY ANALYSES (continued) to ensure that the SDV has sufficient capacity to contain the reactor coolant discharge during a full core scram. To automatically ensure this capacity, a reactor scram (LCO 3.3.1.1, "Reactor Protection System (RPS)

Instrumentation") is initiated if the SDV water level in the instrument volume exceeds a specified setpoint. The setpoint is chosen so that all control rods are inserted before the SDV has insufficient volume to accept a full scram.

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1

SDV vent and drain valves satisfy Criterion 3 of the NRC Policy Statement (Ref. 4),

LCO The OPERABILITY of all SDV vent and drain valves ensures that the SDV vent and drain valves will close during a scram to contain reactor water discharged to the SDV piping. The SDV vent and drain valves are required to be open to ensure the SDV is drained. Since the vent and drain lines are provided with two valves in series, the single failure of one valve in the open position will not impair the isolation function of the system. Additionally, the valves are required to open on scram reset to ensure that a path is available for the SDV piping to drain.freely at other times.

'APPLICABILITY In MODES 1 and 2, scram may be required; therefore, the SDV vent and drain valves must be OPERABLE. In MODES 3 and 4, control rods are not able to be withdrawn (except as permitted by LtO 3.10.3 and LCO 3.10.4) since the reactor mode switch is in shutdown and a control rod block is applied. "This provides adequate controls to ensure that only a single control rod can be withdrawn. Also, during MODE 5, only a single control rod can be withdrawn from a core cell containing fuel assemblies.

Therefore, the SDV vent and drain valves are not required to be OPERABLE in these MODES since the reactor is subcritical and only one rod may be withdrawn and subject to scram.

ACTIONS The ACTIONS table is modified by Note 1 indicating that a separate Condition entry is allowed for the SDV vent line and the SDV drain line. This is acceptable, since the I

' (continued)

SUSQUEHANNA - UNIT 1

. TS / B 3.1-48 Revision I

PPL Rev. 1 SDV Vent and Drain Valves B 3.1.8 BASES ACTIONS Required Actions for each Condition provide appropriate compensatory (continued) actions for each inoperable SDV line. Complying with the Required Actions may allow for continued operation, and subsequent inoperable SDV lines are governed by subsequent Condition-entry and application of associated Required Actions.

The ACTIONS table is modified by a second note stating that a isolated line may be unisolated under administrative control to allow draining and venting of the SDV. When a line is isolated, the potential for an inadvertent scram due to high SDV level is increased. During these periods, the line may be unisolated under administrative control. This allows any accumulated water in the line to be drained, to preclude a reactor scram on high SDV level. This is acceptable since administrative controls ensure the valve can be closed quickly, if a scram occurs with the valve open.

A.1 When one SDV vent or drain valve is inoperable in one or more lines, the associated line must be isolated to contain the reactor coolant during a scram. The 7 day Completion Time is reasonable, given the level of redundancy in the lines and the low probability of a scram occurring while the valve(s) are inoperable and the line is not isolated.- The SDV is still isolable since the redundant valve in the affected line is-OPERABLE.

During these periods, the single failure criterion is not preserved, and a

'higher risk exists to allow reactor water out of the primary system during a scram.

B.1 If both valves in a line are inoperable, the line must be isolated to contain the reactor coolant during a scram.

The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time to isolate the line is based on'the low probability of a scram occurring while the line is not isolated and unlikelihood of significant CRD seal leakage.

If any Required Action and associated Completion Time is not met, the plant must be brought to a MODE in which the LCO (continued)

SUSQUEHANNA - UNIT I TS / B 3.1-49 Revision 1

PPL Rev. 1 SDV Vent and Drain Valves 8 3.1.8 BASES ACTIONS

- C.1 (continued) does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.8.1 REQUIREMENTS During normal operation, the SDV vent and drain valves should be in the open position (except when performing SR 3.1.8.2) to allow for drainage of the SDV piping: Verifying that each valve is in the open position ensures that the SDV vent and drain valves will perform their intended functions during normal operation. This SR does not require any testing or valve manipulation; rather, it involves verification that the valves are in the correct position.

The 31 day Frequency is based on engineering judgment and is consistent with the procedural controls governing valve operation, which ensure correct valve positions.

SR 3.1.8.2 During a scram, the SDV vent and drain valves should close to contain the reactor water discharged to the SDV piping. Cycling each valve through its complete range of motion (closed and open) ensures that the valve will function properly during a scram. The 92 day Frequency is based on operating experience and takes into account the level of redundancy in the system design.

SR 3.1.8.3 SR 3.1.8.3 is an integrated test of the SDV vent and drain valves to verify total system performance. After receipt of a simulated or actual scram signal, the closure of the SDV vent and drain valves is verified. The closure time of (continued)

SUSQUEHANNA-UNIT 1

.. :..B 3.1-50 Revision 0

PPL Rev.l1 SDV Vent and Drain Valves B 3.1.8 BASES SURVILLNCE SR 3.1.8.3 (continiued)

REQUIREMENTS 30 seconds after receipt of a scram signal is based on the bounding leakage case evaluated in the accident analysis based on the requirements of Reference 2. Similarly, after receipt of a simulated or actual scram reset signal, the opening of the SDV vent and drain valves is verified. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.1.1 and the scram time testing of control rods in LCO 3.1.3 overlap this Surveillance to provide complete testing of the assumed safety function.

The 24 month Frequency is-based on the need to perform portions of this Surveillance under th'e conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at the

• 24 month Frequency; therefore, the Frequency was concluded to be acceptable from a reliability standpoint REFERENCES I1.

FSAR, Section 4.6.

2.

IO CFR 100.

3.

NUREG-0803, 'Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping," August1I981.

4.

Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).

5.

TSTF-404-A, Rev. 0.

SUSQUEHANNA -UNITI1 TS/B 3.1 -51 Revision I

B. 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)

PPL Rev. 0 APLHGR B 3.2.1 BASES BACKGROUND The APLHGR is a measure of the average LHGR of all the fuel rods in a fuel assembly at any axial location. Limits on the APLHGR are specified to ensure that limits specified in 10 CFR 50.46 are not exceeded during the postulated design basis loss of coolant accident (LOCA).

-APPLICABLE SAFETY ANALYSES SPC performed LOCA calculations for the SPC ATRIUM'M-10 fuel design. The analytical methods and assumptions used in evaluating the fuel design limits from 10 CFR 50.46 are presented in References 3, 4, 5, and 6 for the SPC analysis. The analytical methods and assumptions used in evaluating Design Basis Accidents (DBAs) that determine the APLHGR Limits are presented in References 3 through 9.

I I

I LOCA analyses are performed to ensure that the APLHGR limits are adequate to meet the Peak Cladding Temperature (PCT), 'maximum cladding oxidation, and maximum hydrogen generation limits of 10 CFR 50.46. The analyses are performed using calculational models that are consistent with the requirements of 10 CFR 50, Appendix K. A complete discussion of the analysis codes are provided in References 3, 4, 5, and 6 for the SPC analysis. The PCT following a postulated LOCA is a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is not strongly influenced by the rod to rod power distribution within the assembly.

APLHGR limits are developed as a function of fuel type and exposure.

The SPC LOCA analyses also consider several alternate operating modes in the development of the APLHGR limits (e.g., Extended Load Line Limit Analysis (ELLA), Suppression Pool Co6ling Mode, and Single Loop Operation (SLO)). LOCA analyses were performed for the regions of the power/flow map bounded by the 100% rod line and the APRM rod block line (i.e., the ELLA region). The ELLA region is analyzed to determine whether an APLHGR multiplier as a function of core flow is required. The results of the analysis demonstrate the PCTs are within '

the 10 CFR 50.46 limit, and that APLHGR multipliers as a function of core flow are not required.

I

-(continued)

SUSQUEHANNA-UNIT1 -

TS / B 3.2-1 Revision 1.

PPL Rev. 0 APLHGR B 3.2.1 BASES APPLICABLE -

SAFETY ANALYSES (continued)

The SPC LOCA analyses consider the delay in Low Pressure Coolant Injection (LPCI) availability when the unit is operating in the Suppression Pool Cooling Mode. The delay, in LPCI availability is due to the time required to realign valves from the Suppression Pool Cooling Mode to the LPCI mode. The results of the analyses demonstrate that the PCTs are within the 10 CFR 50.46 limit.

Finally, the SPC LOCA analyses were performed for Singe-Loop Operation. The results of the SPC analysis forATRIUM

-10 fuel shows that an APLHGR limit which is 0.8 times the two-loop APLHGR limit meets the 10 CFR 50.46 acceptance criteria, and that the PCT is less than the limiting two-loop PCT.

The APLHGR satisfies Criterion 2 of the NRC Policy Statement (Ref.

10).

I I

I LCO The APLHGR limits specified in the COLR are the result of the DBA analyses.

APPLICABILITY The APLHGR limits are primarily derived from LOCA analyses that are assumed to occur at high power levels. Design calculations and operating experience have shown that as power is reduced, the margin to the required APLHGR limits increases.-At THERMAL.POWER levels

<.25% RTP, the reactor is operating with substantial margin to the APLHGR limits; thus, this LCO is not required.

ACTIONS A.1 If any APLHGR exceeds the required limits, an assumption regarding an initial condition of the DBA may not be met Therefore, prompt action should be taken to restore the APLHGR(s) to within the required limits such that the plant operates within analyzed conditions. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is sufficient to restore the APLHGR(s) to within its limits and is acceptable based on the low probability of a DBA occurring simultaneously with the APLHGR out of specification.

(continued)

SUSQUEHANNA-UNIT 1

.TS/B3.2-2 Revision 2

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PPL Rev. 0 APLHGR B 3.2.1 BASES ACTIONS B.1 (continued)

If the APLHGR cannot be restored to within its required limits within the associated Completion Time, the'plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 25% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 25% RTP in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.1.1 REQUIREMENTS APLHGRs are required to be initially calculated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is 2 25% RTP and then every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.

Additionally, APLHGRs must be calculated prior to exceeding 50% RTP unless performed in the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. APLHGRs are compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on both engineering judgment and recognition of the slowness of changes in power dist ibution during normal operation. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowance after THERMAL POWER 2 25% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels and because the APLHGRs must be calculated prior to exceeding 50% RTP.

REFERENCES

1. Not used.

2.

Not used.

3.

ANF-91-048(P)(A), "Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model," January 1993.

4.- ANF-CC-33(P)(A) Supplement 2, "HUXY: A Generalized Multirod Heatup Code with 10CFR50 Appendix K Heatup Option,"

January 1991.;

5.

XN-CC-33(P)(A) Revision 1, "HUXY: A Generalized Multirod Heatup Code with 10CFR50 Appendix K Heatup Option Users Manual," November 1975.

(continued)

SUQUHAN UNT1T.

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I iI II.1 II SUSQUEHANNA -UNIT 1 -

TS / B 3.2-3 '.

Revision 1 p

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PPL Rev. 0 APLHGR B 3.2.1 BASES References (continued)

6. XN-NF-80-19(P)(A), Volumes 2, 2A, 2B, and 2C "Exxon Nuclear Methodology-for Boiling Water Reactors: EXEM BWR ECCS Evaluation Model," September 1982.

7.

FSAR, Chapter 4.

8.

FSAR, Chapter 6.

9.

FSAR, Chapter 15.

10. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).

. N ~

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i SUSQUEHANNA - UNIT 1 TS B 3.2-4 Revision 2

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