Text

Proposed License Amendments 285 & 253, Extended Power Uprate Application Regarding Withdrawal of Change to Technical Specification SR 3.3.1.1.8 - Calibration Frequency for Lprms
	ML072280247
Person / Time
Site:	Susquehanna
Issue date:	08/03/2007
From:	Mckinney B; Susquehanna
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	PLA-6254
	Download: ML072280247 (47)
	v • d • e

!rltt T. MalcKnney PPL SusquelMnna, LLC Sr. Vice President & Chief Nuclear Officer 769 Salem Boulevard Berwick, PA 18603 Tel. 570.542.3149 Fax 570.542.1504 AUG 0 3 2007 pp btmckinney@pplweb.com U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop OP 1-17 Washington, DC 20555 SUSQUEHANNA STEAM ELECTRIC STATION PROPOSED LICENSE AMENDMENT NO. 285 FOR UNIT 1 OPERATING LICENSE NO. NPF-14 AND PROPOSED LICENSE AMENDMENT NO. 253 FOR UNIT 2 OPERATING LICENSE NO. NPF-22 EXTENDED POWER UPRATE APPLICATION REGARDING WITHDRAWAL OF CHANGE TO TECHNICAL SPECIFICATION SR 3.3.1.1.8 -

CALIBRATION FREQUENCY FOR LPRMS Docket Nos. 50-387 PLA-6254 and 50-388 Reference. PPLLetter PLA-6076, B. T. McKinney (PPL) to USNRC, "ProposedLicense Amendment Numbers 285for Unit I OperatingLicense No. NPF-14 and 253for Unit 2 OperatingLicense No. NPF-22 Constant Pressure Power Uprate," dated October 11, 2006.

Pursuant to 10 CFR 50.90, PPL Susquehanna LLC (PPL) requested in the above Reference approval of amendments to the Susquehanna Steam Electric Station (SSES)

Unit 1 and Unit 2 Operating Licenses (OLs) and Technical Specifications (TS) to increase the maximum power level authorized from 3489 megawatts thermal (MWt) to 3952 MWt, an approximate 13% increase in thermal power. The proposed Constant Pressure Power Uprate (CPPU) represents an increase of approximately 20% above the Original Licensed Thermal Power (OLTP).

The purpose of this letter is to withdraw the proposed change to Technical Specification 3.3.1.1 Surveillance Requirement SR 3.3.1.1.8 which requested a longer frequency between calibrations of the Local Power Range Monitors (LPRMs). The proposed change would have changed the calibration frequency from 1000 megawatt days per metric ton (MWD/MT) to 2000 MWD/MT. In order to support this change, NRC requested additional analysis be performed. The analysis will take several months to develop. Therefore, so as not to extend the approval of the CPPU submittal, PPL is withdrawing the proposed change to SR 3.3.1.1.8.

-A0-0 f

Document Control Desk PLA-6254 contains the marked-up Technical Specification pages for Section 3.3.1.1 for Units 1 & 2, which supersede those pages that were transmitted in the above reference. Attachment 2 contains the revised marked-up Technical Specification Bases pages for Section B3.3. 1.1 for Units 1 & 2 for information only.

There are no new regulatory commitments associated with this submittal.

PPL has reviewed the "No Significant Hazards Consideration" and the "Environmental Consideration" submitted with the Reference relative to the Enclosure. We have determined that there are no changes required to the "Environmental Consideration."

The original "No Significant Hazards Consideration" contained statements on the LPRM calibration frequency. The "No Significant Hazards Consideration" in Attachment 3 to this letter has been revised to delete any reference to the LPRM calibration frequency.

The removal of the reference to the LPRM calibration frequency does not invalidate the original conclusion.

If you have any questions or require additional information, please contact Mr. Michael H. Crowthers at (610) 774-7766.

I declare under perjury that the foregoing is true and correct.

Executed on: '

B. T. McKinney Attachment I - Revised Technical Specification Pages for Section 3.3.1.1 Units 1 & 2.

(Mark-ups)

Attachment 2 - Revised Technical Specification Bases Pages for Section B3.3. 1.1 Units I

& 2. (Mark-ups - For Information Only)

Attachment 3 - Revised No Significant Hazards Consideration.

Copy: NRC Region I Mr. R. V. Guzman, NRC Sr. Project Manager Mr. R. R. Janati, DEP/BRP Mr. F. W. Jaxheimer, NRC Sr. Resident Inspector

Attachment 1 to PLA-6254 Revised Technical Specification Pages for Section 3.3.1.1 Units I & 2 (Mark-ups)

PPL Rev. 0 RPS Instrumentation 3.3.1.1 ACTIONS (continued)*

CONDITION REQUIRED ACTION COMPLETION TIME C. One or more C.1 Restore RPS trip capability. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Functions with RPS trip capability not maintained.

D. Required Action and D.1 Enter the Condition referenced Immediately

.associated* in Table 3.3.1.1-1 for the Completion Time of channels.

Condition A, B, or C not met.

E. As required by E.1 Reduce THERMAL POWER to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Required Action D.1 and referenced in Table 3.3.1.1*-1.

F. As required by F.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Required Action D.1 and referenced in Table 3.3.1.1-1.

G. As required by G.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Required Action D.1 and referenced in Table 3.3.1.1-1.

H. As required by H.1 Initiate action to fully insert all Immediately Required Action D.1 insertable control rods in core and referenced in cells containing one or more fuel Table 3.3.1.1-1. assemblies.

(continued)

SUSQUEHANNA - UNIT 1 3.3-2 pe to Tflt PR and ARTSRME-LLLA

PPL Rev. 0 RPS Instrumentation 3.3.1.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME As required by 1.1 Initiate alternate method to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Required Action D.1 detect and suppress thermal and referenced in hydraulic instability oscillations.

Table 3.3.1.1-1 AND 1.2 Restore require channels to 120 days OPERABLE.

J. Required Action and J.1 Reduce THERMAL POWER to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated o RTP.

Completion Time of Condition I not met.

SUSQUEHANNA - UNIT 1 3.3-3 Retype tor P S and ARTS LLLA

PPL Rev. 0 RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS

1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains RPS trip capability.

SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.1.1.2 Perform CHANNEL CHECK. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months SR 3.3.1.1.3--- ---------------------- --------NOTE .---------------...------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER ! RTP.

Verify the absolut -rnce between the 6verage 7 days power ran onitor (APRM) channels and the alcul*TP.power is *_2% RTP while operating at SR 3.3.1.1.4 ----------------------------------- NOTE ---------------------------

Not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

Perform CHANNEL FUNCTIONAL TEST. 7 days (continued)

SUSQUEHANNA - UNIT 1 3.3-4 Retype to yreUeol P S and ARTSýAE~LLLA

PPL Rev. 0 RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.11 ---------------- NOTES ----------------------------..

1. Neutron detectors are excluded.

2. For Function 1.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

Perform CHANNEL CALIBRATION. 184 days SR 3.3.1.1.12 - -------- NOTES ------.-.-....-------- -

1. For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

2. For Functions 2.b and 2.f, the CHANNEL FUNCTIONAL TEST includes the recirculation flow input processing, excluding the flow transmitters.

Perform CHANNEL FUNCTIONAL TEST. 184 days SR 3.3.1.1.13 Perform CHANNEL CALIBRATION. 24 months SR 3.3.1.1.14 Perform CHANNEL FUNCTIONAL TEST. 24 months SR 3.3.1.1.15 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months SR 3.3.1.1.16 Verify Turbine Stop Valve-Closure and Turbine 24 months Control Valve Fast Closure, Trip Oil Pressure-Low Functions are not bypassed when THERMAL POWER is >:W/o RTP.

(continued)

SUSQUEHANNA - UNIT 1 3.3-6 andLL

PPL Rev. 0 RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.17

1. Neutron detectors are excluded.

2. For Function 5 "n" equals 4 channels for the purpose of determining the STAGGERED TEST BASIS Frequency.

3. For Function 2.e, "n" equals 8 channels for the purpose of determining the STAGGERED TEST BASIS Frequency. Testing of APRM and OPRM outputs shall alternate.

Verify the RPS RESPONSE TIME is within limits. 24 months on a STAGGERED TEST BASIS SR 3.3.1.1.18 ------------ ----- NOTES------------

1. Neutron detectors are excluded.

2. For Functions 2.b and 2.f, the recirculation flow transmitters that feed the APRMs are included.

Perform CHANNEL CALIBRATION. 24 months SR 3.3.1.1.19 Verify OPRM is not b passed when APRM Simulated 24 months Thermal Power is >: , and recirculation drive flow is < value equivalent-the core flow value defined in the COLR.

SR 3.3.1.1.20 Adjust recirculation drive flow to conform to reactor 24 months core flow.

SUSQUEHANNA - UNIT 1 3.3-7 etype to ref PR and ARTS/ LLLA

PPL Rev. 0 RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 1 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

1. Intermediate Range Monitors

a. Neutron 2 3 G SR 3.3.1.1.1 5 122/125 divisions Flux-High

SR 3.3.1.1.4 of full scale SR 3.3.1.1.6 SR 3.3.1.1.7 SR 3.3.1.1.11 SR 3.3.1.1.15 3 H SR 3.3.1.1.1 S 122/125 divisions SR 3.3.1.1.5 of full scale SR 3.3.1.1.11 SR 3.3.1.1.15

b. Inop 2 3 G SR 3.3.1.1.4 NA SR 3.3.1.1.15

.5(a) 3 H SR 3.3.1.1.5 NA SR 3.3.2.2.15

2. Average Power Range Monitors

a. Neutron 2 3 (c) G SR 3.3.1.1.2 S<20% RTP Flux-High SR 3.3.1.1.7 (Setdown) SR 3.3.1.1.8 SR 3.3.1.1.12 SR 3.3.1.1.18

b. Simulated 3 (c) F SR 3.3.1.1.2 Thermal SR 3.3.1.1.3 Power-High SR 3.3.1.1.8 and:_ 115.5% RTP SR 3.3.1.1.12 SR 3.3.1.1.18 SR 3.3.1.1.20 (continued)

(a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(b) - + 64.. / RTP when reset for single loop operation per LCO 3.4.1, "Recirculation Loops Operating."

(c) ach APRM channel provides inputs to both trip systems 0 .S57 W LAI3) 4+~~

SUSQUEHANNA - UNIT 1 3.3-8 e o r0laet PR and A ELLLA 6;;i -

PPL Rev. 0 RPS Instrumentation 3.3.1.1

. Table 3.3.1.1-1 (page 2 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

2. Average Power Range Monitors (continued)

c. Neutron 3(c) F SR 3.3.1.1.2 _<120% RTP Flux-High

SR 3.3.1.1.3 SR 3.3.1.1.8 SR 3.3.1.1.12 SR 3.3.1.1.18

d. Inop 1,2 3 (c) G SR 3.3.1.1.12 NA

e. 2-Out-Of-4 1,2 2 G SR 3.3.1.2 NA Voter SR 3.3.1.12 SR 3.3.1.15 SR 3.3.1.17

f. OPRM Trip 3(*C) I SR 3.3.1.2 (d)

.0 %FIT SR SR 3.3.1.8 3.3.1.12 SR 3.3.1.18 SR 3.3.1.19 SR 3.3.1.20

3. Reactor Vessel 1,2 2 G SR 3.3.1.1.9 < 1093 psig Steam Dome SR 3.3.1.1.10 Pressure-High SR 3.3.1.1.15

4. Reactor Vessel 1,2 2 G SR 3.3.1.1.1 'a 11.5 inches Water Level-Low, SR 3.3.1.1.9 Level 3 SR 3.3.1.1.10 SR 3.3.1.1.15

5. Min Steam 8 F SR 3.3.1.1.9 !5 11% closed Isolation Valve-- SR 3.3.1.1.13 Closure SR 3.3.1.1.15 SR 3.3.1.1.17

.6. Drywell Pressure-- 1,2 2 G SR 3.3.1.1.9 < 1.88 psig High SR 3.3.1.1.10 SR 3.3.1.1.15 (continued)

(c) Each APRM channel provides inputs to both trip systems.

(d) See COLR for OPRM period based detection alg6rithm (PBDA) setpoint limits.

SUSQUEHANNA - UNIT 1 TS / 3.3-9 Retype to ect IVMS and ART ELLLA

PPL Rev. 0 RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 3 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

7. Scram Discharge Volume Water Level--High

a. Level 1,2 2 G SR 3.3.1.1.9 < 66 gallons Transmitter SR 3.3.1.1.13 SR 3.3.1.1.15 5(a) H SR 3.3.1.1.9 2 _<66 gallons SR 3.3.1.1.13 SR 3.3.1.1.15

b. Float Switch 1,2 2 G SR 3.3.1.1.9 < 62 gallons SR 3.3.1.1.13 SR 3.3.1.1.15 5 (a) 2 H SR 3.3.1.1.9 <_62 gallons SR 3.3.1.1;13 SR 3.3.1.1.15

8. Turbine Stop E SR 3.3.1.1.9 < 7% closed Valve-Closure SR 3.3.1.1.13 SR 3.3.1.1.15 joRTP 4 SR 3.3.1.1.16 SR 3.3.1.1.17

9. Turbine Control ŽRTP '22 E SR 3.3.1.1.9 > 460 psig Valve Fast Closure, SR 3.3.1.1.13 Trip Oil Pressure- SR 3.3.1.1.15 Low SR 3.3.1.1.16 SR 3.3.1.1.17 10 Reactor Mode 1,2 2 G SR 3.3.1.1.14 NA Switch- Shutdown SR 3.3.1.1.15 Position 5(a) H SR 3.3.1.1.14 NA 2

SR 3.3.1.1.15

11. Manual Scram 1,2 2 G SR 3.3.1.1.5 NA SR 3.3.1.1.15 5 (a) 2 H SR 3.3.1.1.5 NA SR 3.3.1.1.15 (a) With any control rod withdrawn from a core cell containing one or more tuel assemblies.

SUSQUEHANNA - UNIT 1 3.3-10

PPL Rev. 1 RPS Instrumentation 3.3.1.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. One or more C.1 Restore RPS trip capability. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Functions with RPS trip capability not maintained.

D. Required Action and D.1 Enter the Condition referenced Immediately associated in Table 3.3.1.1-1 for the Completion Time of channels.

Condition A, B, or C not met.

E. As required by E.1 Reduce THERMAL POWER to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Required Action D.1 < &Y.RTP.

and referenced in Table 3.3.1.1-1.

F. As required by F;1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Required Action D.1 and referenced in Table 3.3.1.1-1.

G. As required by G.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Required Action D.1 and referenced in Table 3.3.1.1-1.

H. As required by H.1 Initiate action to fully insert all Immediately Required Action D.1 insertable control rods in core and referenced in cells containing one or more fuel Table 3.3.1.1-1. assemblies.

(continued)

SUSQUEHANNA - UNIT 2 3.3-2 and A 'ELLLAI

PPL Rev. 1 RPS Instrumentation 3.3.1.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME As required by 1.1 Initiate alternate method to detect 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Required Action D.1 and suppress thermal hydraulic and referenced in instability oscillations.

Table 3.3.1.1-1 AND 1.2 Restore require channels to 120 days OPERABLE.

J. Required Action and J.1 Reduce THERMAL POWER to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated < oTP.

Completion Time of Condition I not met.

SUSQUEHANNA-UNIT2 3.3-3 PRReLr

PPL Rev. 1 RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS

NOTES --------------------..... . .... -- ----------

1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains RPS trip capability.

SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.1.1.2 Perform CHANNEL CHECK. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months SR 3.3.1.1.3 - NOTE -------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER > ocRTP.

- 3--

Verify the absolute differ een the average 7days power range APRM) channels and the calcul ower is: 22%RTP while operating at SR 3.3.1.1.4 ---------------------- NOTE------------------

Not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

Perform CHANNEL FUNCTIONAL TEST. 7 days (continued)

SUSQUEHANNA - UNIT 2 3.3-4

'-RetpTe to~jwelect PR ad /RIMELL

PPL Rev. 1 RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY

NOTES--------

SR 3.3.1.1.11

1. Neutron detectors are excluded.

2. For Function 1.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

Perform CHANNEL CALIBRATION. 184 days SR 3.3.1.1.12 -------------------------------- NOTES---------------

1. For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2..

2. For Functions 2.b and 2.f, the CHANNEL FUNCTIONAL TEST includes the recirculation flow input processing, excluding the flow transmitters.

Perform CHANNEL FUNCTIONAL TEST. 184 days SR 3.3.1.1.13 Perform CHANNEL CALIBRATION. 24 months SR 3.3.1.1.14 Perform CHANNEL FUNCTIONAL TEST. 24 months SR 3.3.1.1.15 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months SR 3.3.1.1.16 Verify Turbine Stop Valve-Closure and Turbine Control 24 months Valve Fast Closure, Trip Oil Pressure-Low Functions are not bypassed when THERMAL POWER is RTP.

(continued)

SUSQUEHANNA - UNIT 2 3.3-6 PRWMt t

PPL Rev. 1 RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.17 ---------------- NOTES----------------

1. Neutron detectors are excluded.

2. For Function 5 "n" equals 4 channels for the purpose of determining the STAGGERED TEST BASIS Frequency.

3. For Function 2.e, "n" equals 8 channels for the purpose of determining the STAGGERED TEST BASIS Frequency. Testing of APRM and OPRM outputs shall alternate.

Verify the RPS RESPONSE TIME is within limits. 24 months on a STAGGERED TEST BASIS SR 3.3.1.1.18 ---------------------- NOTES ---------------

1. Neutron detectors are excluded.

2. For Functions 2.b and 2.f, the recirculation flow transmitters that feed the APRMs are included.

Perform CHANNEL CALIBRATION. 24 months SR 3.3.1.1.19 Verify OPRM is not bypassed when APRM 24 months Simulated Thermal Power is - and recirculation drive flow is <- value equivalent the core flow value defined in the COLR.

SR 3.3.1.1.20 Adjust recirculation drive flow to conform to reactor 24 months core flow.

SUSQUEHANNA - UNIT 2 3.3-7 PR a nd TS/MEL

PPL Rev. 1 3.3.1.1 Table 3.3.1.1-1 (page 1 of 3)

Reactor Protection System Instrumentation APPLICABLE MODES OR CONDITIONS OTHER REQUIRED REFERENCED SPECIFIED CHANNELS PER FROM REOUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS TRIP SYSTEM ACTION D.1 REQUIREMENTS VALUE

1. Intermediate Range Monitors

a. Neutron . 2 G SR 3.3.1.1.1 S 122/125 divisions Flux-High of full scale SR 3.3.1.1.4 SR 3.3.1.1.6 SR 3.3.1.1.7 SR 3.3.1.1.11 SR 3.3.1.1.15 5(a) 3 H SR 3.3.1.1.1 <_122/125 divisions SR 3.3.1.1.5 of full scale SR 3.3.1.1.11 SR 3.3.1.1.15

b. Inop 2 3 G SR 3.3.1.1.4 NA SR 3.3.1.1.15 3 H SR 3.3.1.1.5 NA SR 3.3.1.1.15

2. Average Power Range Monitors

a. Neutron 2 G SR 3.3.1.1.2 :520% RTP Flux--High SR 3.3.1.1.7 (Setdown) SR 3.3.1.1.8 SR 3.3.1.1.12 -- 60La 0.

SR 3.3.1.1.18 6;

b. Simulated 1 F SR 3.3.1.1.2 *5 .62 W 6FTP(" n Thermal SR 3.3.1.1.3 +644. ARTPb) and Power-High SR 3.3.1.1.8 :5 115.5% RTP SR 3.3.1.1.12 SIR 3.3.1.1.18 SR 3.3.1.1.20 (continued)

(a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(b) Co RTP when reset for single loop operation per LCO 3.4.1, "Recirculation Loops Operating."

(c) Ea*/hAPRM channelprovides i.tt-4e-t rip serns SUSQUEHANNA - UNIT 2 TS / 3.3-8 (AR and AR tL9LA

PPL Rev. 1 RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 2 of 3)

Reactor Protection System Instrumentation APPLICABLE MODES OR CONDITIONS OTHER REQUIRED REFERENCED SPECIFIED CHANNELS PER FROM REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS TRIP SYSTEM ACTION D.1 REQUIREMENTS VALUE

2. Average Power Range Monitors (continued)

c. Neutron 1 F SR 3.3.1.1.2 _ 120% RTP Flux-High SR 3.3.1.1.3 SR 3.3.1.1.8 3ýO 3 (c) SR 3.3.1.1.12 SR 3.3.1.1.18

d. Inop 1,2 G SR 3.3.1.1.12 NA

e. 2-Out-Of-4 1.2 2 G SR 3.3.1.2 NA Voter SR 3.3.1.12 SR 3.3.1,15 SR 3.3.1.17 3'c) I SR 3.3.1.2 (d)

f. OPRM Trip SR 3.3.1.8 SR 3.3.1.12 SR 3.3.1.18 SR 3.3.1.19 SR 3.3.1.20

3. Reactor Vessel 1,2 2 G SR 3.3.1.1.9 :5 U)93 psig Steam Dome SR 3.3.1.1.10 Pressure-High SR 3.3.1.1.15

4. Reactor Vessel 1,2 2 G SR 3.3.1.1.1 __11.5 inches Water Level- SR 3.3.1.1.9 Low, Level 3 SR 3.3.1.1.10 SR 3.3.1.1.15

5. Main Steam 1 8 F SR 3.31.1.9 !5 11% closed Isolation SR 3.3A1.1.13 Valve--Closure SR 3.3.1.1.15 SR 3.3.1.1.17

6. Drywell 1,2 2 G SR 3.3.1.1.9 51 .88 psig Pressure-High SR 3.3.1.1.10 SR 3.3.1.1.15 (continued)

(c) Each APRM channel provides inputs to both trip systems.

(d) See COLR for OPRM period based detection algorithm (PBDA) setpoint limits.

SUSQUEHANNA - UNIT 2 TS/3.3-9 e, to re lect ELLLA

ýan dand

PPL Rev. 1 PPL Rev. ORPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 3 of 3)

Reactor Protection System Instrumentation APPLICABLE MODES OR CONDITIONS OTHER REQUIRED REFERENCED SPECIFIED CHANNELS PER FROM REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS TRIP SYSTEM ACTION D.1 REQUIREMENTS VALUE

7. Scram Discharge Volume Water Level--High

a. Level 1,2 2 G SR 3.3.1.1.9 f 66 gallons Transmitter SR 3.3.1.1.13 SR 3.3.1.1.15 5(s) 2 H SR 3.3.1.1.9 _566 gallons SR 3.3.1.1.13 SR 3.3.1.1.15

b. Float Switch 1,2 2 G SR 3.3.1.1.9 5 62 gallons SR 3.3.1.1.13 SR 3.3.1.1.15 51a1 2 H SR 3.3.1.1.9 !5 62 gallons SR 3.3.1.1.13 SR 3.3.1.1.15

8. Turbine Stop > RTP 4 E SR 3.3.1.1.9 5 7% dosed Valve--Closure SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.16 SR 3.3.1.1.17

9. Turbine Control TRRTP 2 E SR 3.3.1.1.9 >?460 psig Valve Fast SR 3.3.1.1.13 Closure, Trip SR 3.3.1.1.15 Oil Pressure-- SR 3.3.1.1.16 Low SR 3.3.1.1.17

10. Reactor Mode 1.2 2 G SR 3.3.1.1.14 NA Switch-- SR 3.3.1.1.15 Shutdown Position 5(a) H SR 3.3.1.1.14 2 NA SR 3.3.1.1.15

11. Manual Scram 1,2 2 G SR 3.3.1.1.5 NA SR 3.3.1.1.15 51a) H SR 3.3.1.1.5 NA 2

SR 3.3.1.1.15 (a) With any control rod wilhdrawn from a core cell containing one or more fuel assemblies.

SUSQUEHANNA - UNIT 2 3.3-10 Retype to reflect RNMoand

ý"PRýNMS a deRTectL ARTS/MELLLA_.,

Attachment 2 to PLA-6254 Revised Technical Specification Bases Pages for Section B3.3.1.1 Units 1 & 2 (Mark-ups - For Information Only)

PPL Rev. 2 RPS. Instrumentation B 3.3.1.1 BASES APPLICABLE 2.a. Average Power Range Monitor Neutron Flux - High (Setdown)

SAFETY ANALYSES, For operation at low power (i.e., MODE 2), the Average Power Range LCO, and Monitor Neutron Flux - High (Setdown) Function is capable of generating a APPLICABILITY trip signal that prevents fuel damage resulting from abnormal operating (continued) transients in this power range. For most operation at low power levels, the Average Power Range Monitor Neutron Flux - High (Set.down) Function will provide a secondary scram to the Intermediate Range Monitor Neutron.

Flux - High Function because of the relative setpoints. With the IRMs at Range 9 or 10, it is possible that the Average Power Range Monitor Neutron Flux - High (Setdown) Function will provide the primary trip signal for a corewide increase in power. I-<

No specific safety analyses take direct credit fg-t 9

Monitor Neutron Flux - High (Setdown) Fun ion. I indirectly ensures that before the reactor je swil position, reactor power does not exceed 6oRTP operating at low reactor pressure and low core floP prevents fud'amage during significant reactivi in POWER < T.

The Allowable Value is based o r ve.

when THERMAL POWER is 4 P The Average Power Range Monitor Neutron Flux - High (Setdown) Function must be OPERABLE during MODE 2 when control rods may be withdrawn since the potential for criticality exists. In MODE 1, the.Average Power Range Monitor Neutron Flux -High Function provides protection against reactivity transients and the RWM protects against control rod withdrawal error events.

(continued)

SUSQUEHANNA - UNIT 1 TS /

PPL Rev. 2 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2J. Oscillation Power Ranqe Monitor (OPRM) Trip SAFETY ANALYSES, The OPRM Trip Function provides compliance with GDC 10, "Reactor LCO, and Design,". and GDC 12, "Suppression of Reactor Power Oscillations" thereby APPLICABILITY providing protection from exceeding the fuel MCPR safety limit (SL) due to (continued) anticipated thermal-hydraulic power oscillations.

References 17, 18 and 19 describe three algorithms for detecting thermal-hydraulic instability related neutron flux oscillations: the period based detection algorithm (confirmation count and cell amplitude), the amplitude based algorithm, and the growth rate algorithm. All three are implemented in the OPRM Trip Function, but the safety analysis takes credit only for the period based detection algorithm. The remaining algorithms provide defense in depth and additional protection against unanticipated oscillations. OPRM Trip Function OPERABILITY for Technical Specification purposes is based only on the period based detection algorthm.

The OPRM Trip Function receives input signals from the local power range monitors (LPRMs) within the reactor core, which are combined into "cells".

for evaluation by the OPRM algorithms. Each channel is capable of detecting thermal-hydraulic instabilities, by detecting the related neutron flux oscillations, and issuing a trip signal before the MCPR SL is exceeded.

Three of the four channels are requird4,,e OPERABLE.

The OPRM Trip is automat ly ena pass removed) when THERMAL POWER is 2 RTP, as indicated by the APRM Simulated Thermal Power, and reactor core flow is : the value defined in the COLR, as indicated by APRM measured recirculation drive flow. This is the operating region where actual thermal-hydraulic instability and related neutron flux oscillations are expected to occur. Reference 21 includes additional discussion of OPRM Trip enable region limits.

These setpoints, which are sometimes referred to as the "auto-bypass" setpoints, establish the boundaries of the OPRM Trip enabled region. The APRM Simulated Thermal Power auto-enable setpoint has 1% deadband while the drive flow setpoint has a 2% deadband. The deadband for these setpoints is established so that it increases the enabled region once the region is entered.

(continued)

SUSQUEHANNA - UNIT 1 TS/ B 3.3-13 .ed to nclud S a S/MELLL

PPL Rev. 2 RPS Instrumentation B 3.3.1.1 APPLICABLE 2".f. Oscilli n oer Ra Monitor (OPRM) Trip ntiued SAFETY e R ~ n(ERB e e ANALYSES,- The R Trip n is required to be RALE wh e lant is at LCO, and .>!iao RTP. Theu 10RTP level is selec to provide rgin in the APPLICABILITY unlikely event that a reactor power inc ase transien curring without (continued) operator action while the plant is 0 "ng belo o RTP causes a

-powerincrease-to-or-beyondthe -APRM-Simulated Thermal Power OPRM Trip auto-enable setpoint. This OPERABILITY requirement assures that the OPRM Trip auto-enable function will be OPERABLE when required.

An APRM channel is also required to have a minimum number of OPRM cells OPERABLE for the Upscale Function 2.f to be OPERABLE. The OPRM cell operability requirements are documented in the Technical Requirements Manual, TRO 3.3.9, and are established as necessary to support the trip setpoint calculations performed in accordance with methodologies in Reference 19.

An OPRM Trip is issued from an APRM channel when the period based detection algorithm in that channel detects oscillatory changes in the neutron flux, indicated by the combined signals of the LPRM detectors in a cell, with period confirmations and relative cell amplitude exceeding specified setpoints. One or more cells in a channel exceeding the trip conditions will result in a channel OPRM Trip from that channel. An OPRM Trip is also issued from the channel if either the growth rate or amplitude based algorithms detect oscillatory changes in the neutron flux for one or more cells in that channel. (Note: To facilitate placing the OPRM Trip Function 2.f in one APRM channel in a "tripped" state, if necessary to satisfy a Required Action, the APRM equipment is conservatively designed to force an OPRM Trip output from the APRM channel if an APRM Inop condition occurs, such as when the APRM chassis keylock switch is placed in the Inop position.)

There are three "sets" of OPRM related setpoints or adjustment parameters: a) OPRM Trip auto-enable region setpoints for STP and drive flow; b) period based detection algorithm (PBDA) confirmation count and amplitude setpoints; and c) period based detection algorithm tuning parameters.

(continued)

SUSQUEHANNA - UNIT 1 TS /B 3.3-14 ARTS/,FELLLA

PPL Rev. 2 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 7.a, 7.b. Scram Discharge Volume Water Level - High (continued)

SAFETY ANALYSES, Four channels of each type of Scram Discharge Volume Water Level - High LCO, and Function, with two channels of each type in each trip system, are required to APPLICABILITY be OPERABLE to ensure that no single instrument failure will preclude a (continued) scram from these Functions on a valid signal. These Functions are required in MODES 1 and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn. At all other times, this Function may be bypassed.

8. Turbine Stop Valve - Closure Closure of the TSVs results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited.

Therefore, a reactor scram is initiated at the start of TSV closure in anticipation of.the transients that would result from the closure of these valves. The Turbine Stop Valve - Closure Function is the primary scram signal.for the-turbine trip event analyzed in Reference 5.. For this event,. the reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the End of Cycle Recirculation Pump Trip (EOC-RPT) System, ensures that the MCPR SL is not exceeded. Turbine Stop*Valve - Closure signals are initiated from position switches located on each of the four TSVs. Two independent position switches are associated with each stop valve. One of the two switches provides input to RPS trip system A; the other, to RPS trip system B. Thus, each RPS trip system receives an input from four Turbine Stop Valve - Closure channels, each consisting of one position switch. The logic for the Turbine Stop Valve - Closure Function is such that three or more TSVs must be closed to produce a scram. This Function must be enabled at THERMAL POWER

_2 RTP. This is accomplished automatically by pressure instruments "sensing turbine first stage pressure. Because an increase in the main turbine bypass flow can affect this function non-conservatively, THERMAL P*POWER is derived from first stage pressure. The main turbine bypass val *must not cause the trip Function to be bypassed when THERMAL POWERi _ oRTP.

The Turbine Stop Valve - Closure Allowable Value is selected to be high enough to detect imminent TSV closure, thereby reducing the severity of the subsequent pressure transient.

(continued)

SUSQUEHANNA - UNIT 1 TSI/B e3PR MIS a4ee nd7 EELLL

PPL Rev. 2 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 8. Turbine Stop Valve - Closure (continued)

SAFETY ANALYSES, Eight channels (arranged in pairs) of Turbine Stop Valve - Closure Function, LCO, and with four.channels in each trip system, are required to be OPERABLE to APPLICABILITY ensure that no single instrument failure will preclude a scram from this

.(continued) Function if any three TSVs should close. This Function is required, consistent with analysis assumptions, whenever THERMAL POWER is YcoRTP. This Function is not required when THERMAL POWER is

/. r*eo. RTP. since the Reactor Vessel Steam Dome Pressure-High and the verage Power Range Monitor Neutron Flux-High Functions are adequate to maintain the necessary safety margins.

9... Turbine Control Valve Fast Closure, Trip Oil Pressure -. Low Fast closure of the TCVs results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited.

Therefore, a reactor scram is initiated on TCV fast closure in anticipation of the transients that would result from the closure of these valves. The Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Function is the primary scram signal for the generator load rejection event analyzed in Reference 5. For this event, the reactor scram.reduces the amount of energyrequired to be absorbed and, along with the actions of the EOC-RPT System, ensures that the MCPR SL is not exceeded.

Turbine Control Valve Fast Closure, Trip Oil Pressure - Low signals are initiated by the electrohydraulic control (EHC) fluid pressure at each control valve. One pressure instrument is associated with each control valve, and the signal from each transmitter is assigned to a separate PS logic channel.

This Function must be enabled at THERMAL POWER > TP. This is accomplished automatically by pressure instruments sensing ture first stage pressure. Because an increase in the main turbine bypass fl w can affect this function non-conservatively, THERMAL POWER is deriv from first stage pressure. The main turbine bypass valves must not caus the trip Function to be bypassed when THERMAL POWER is ý I RTP The Turbine Control Valve Fast Closure, Trip Oil Pressure - Low A e

Value is selected high enough to detect imminent TCV fast closure.

(continued)

SUSQUEHANNA - UNIT 1 TS/ B

PPL Rev. 2 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 9. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low (continued)

SAFETY ANALYSES, Four channels of Turbine Control Valve Fast Closure, Trip Oil Pressure-LCO, and Low Function with two channels in each trip system arranged in a APPLICABILITY one-out-of-two logic are required to be OPERABLE to ensure that no single (continued) instrument failure will preclude a scram from this Function on a valid signal.

_ThS ,ninP r r st Wn~th the analysis assumptions,.

whenever THERMAL POWER is YoRTP. This Function is not required when THERMAL POWER is < o RTP, since the Reactor Vessel Steam Dome Pr - . Average Power Range Monitor Neutron Flux -

igh Functions are adequate to maintain the necessary safety margins.

10. Reactor Mode Switch - Shutdown Position The Reactor Mode Switch - Shutdown Position Function provides signals, via the manual scram logic channels, to each of the four RPS logic channels, which are redundant to the automatic protective instrumentation channels and provide manual reactor trip capability. This Function was not specifically credited in.the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

The reactor mode switch is a single switch with four channels, each of which provides input into one of the RPS logic channels.

There is no Allowable Value for this Function; since the channels are mechanically actuated based solely on reactor mode switch position.

Four channels of Reactor Mode Switch .- Shutdown Position. Function, with two channels in each trip system, are available and required to-be OPERABLE. The Reactor Mode Switch - Shutdown Position Function is required to be OPERABLE in MODES 1 and 2, and MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn.

(continued)

SUSQUEHANNA - UNIT 1 T7S 1 B 3.3-22 Retyped to include PRNMS and ARTSIMELLLA

PPL Rev. 2 RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.3 REQUIREMENTS (continued) To ensure that the APRMs are accurately indicating the true core average power, the APRMs are calibrated to the reactor power calculated from a heat balance. The Frequency of once per 7 days is based o, orcha LPRM sensitivity, which could affect the APRM reading

-ewen-performances -- :;:;.-

ing. i SR whe RTP is provided that requires the SR to be met only a - RTP because it is difficult to accurately

--- 1r64in APRM indication of core THERMAL POWER consistent with a heat z3balance unnc w

r bc RTP. At low power levels, a high degree of accuracy is he Iarge, inherent margin to thermal limits (MCPR, LHGR and APIL o RTP, the Surveillance is required to have been, satisfatorilyperfored--within the last 7 days, in accordance with*

R3ic allows an increaRMALPOWER abov if the 7 day Frequency is not met per SR 3.0.2. In i vent, the SR must be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reaching or exceedin RTP. Twelve hours is based on operating experience and inconsideration of providing a reasonable time in which to complete the SR.

SR 3.3.1.1.4 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function.

As noted, SR 3.3.1.1.4 is not required to be performed when entering MODE 2 from MODE 1, since testing of the MODE 2 required IRM Functions cannotbe performed in MODE 1 without utilizing jumpers, lifted leads, or movable links. This allows entry into MODE 2 if the 7 day Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2 from MODE 1. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.

A Frequency of 7 days provides an acceptable level of system average unavailability over the Frequency interval and is based on reliability analysis (Ref. 9).

(continued)

SUSQUEHANNA - UNIT 1 TS/ B

PPL Rev. 2 RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.15 REQUIREMENTS (continued) The LOGIC. SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The functional testing of control rods (LCO 3.1.3), and SDV vent and drain valves (LCO 3.1.8),

overlaps this Surveillance to provide complete testing of theassumed safety

function_ -.... .

The. LOGIC SYSTEM FUNCTIONAL TEST for APRM Function 2.e simulates APRM and OPRM trip conditions at the 2-out-of-4 Voter channel inputs to check all combinations of two tripped inputs to the 2-out-of-4 logic in the voter channels and APRM related redundant RPS relays.

The 24 month Frequency is based on the need to perform portions of this Surveillance under the 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 that these components usually pass the Surveillance when performed at the 24 month Frequency.

SR 3.3.1.1.16 This SIR ens s that scrams initiated from the T~urbine Stop lye--Closure and Tur *e Control Valve Fast Closure, Trip Oil Pressure-Lo *unctions l be inadvertently bypassed when THERMAL POWER is /o RTP.

is is performed b a Functional check that ensures the scram feature is not by 0 .F use main turbine bypass flow can affect t unction no servatively (THER s derived from turbine irst stage ssure), the opening of the main turbine bys must not ca e tri Function to be bypassed when Thermal Power is.- * / R'TP.

1Iany bypass chan I's trip function is nonconservative (i.e., the Functions are bypassed at> /6 RTP, either due to open main turbine bypass valve(s) or other reasons), en the affected Turbine Stop Valve-Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure-Low Functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition (nonbypass). If placed in the nonbypass condition, this SR is met and the channel is considered OPERABLE.

The Frequency of 24 months is based on engineering judgment and reliability of the components.

(continued)

SUSQUEHANNA - UNIT 1 TS / B 3.3-38 Re to ic RNMS and AR ELLLA_ -

PPL Rev. 2 RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.17 (continued)

REQUIREMENTS (continued) After 8 cycles, the sequence repeats.

Each test of an OPRM or APRM output tests,each of the redundant outputs from the 2-Out-Of-4 Voter channel for that Function and each of the corresponding relays in the RPS. Consequently, eaoh of the RPS relays is tested every fourth cycle. The RPS relay testing frequency is twice the frequency justified by References 15 and 16.

SR 3.3.1.1.19 This surveillance involves confirming the OPRM Trip auto-enable setpoints.

The auto-enable setpoint values are considered to be nominal values as discussed in Reference 21. This surveillance ensures that the OPRM Trip is enabled (not bypassed) for the correct values of APRM Simulated Thermal Power and recirculation drive flow. Other surveillances ensure that the APRM Simulated Thermal Power and recirculation drive flow properly correlate with THERMAL POWER (SR 3.3.1.1.2) and core flow (SR 3.3.1.1.20), respectively.

If any auto-enable setpoint is nonconservative (i.e., e OP rip is bypassed when APRM Simulated Thermal Power Ž o and recirculation drive flow < value equivalent to the core flow value defined in the COLR, then the affected channel is considered inoperable for the OPRM Trip Function. Alternatively, the OPRM Trip auto-enable setpoint(s) may be adjusted to place the channel in a conservative condition (not bypassed).

If the OPRM Trip is placed in the not-bypassed condition, this SR is met and the channel is considered OPERABLE.

For purposes of this surveillance, consistent with Reference 21, the conversion from core flow values defined in the COLR to drive flow values used for this SR can be conservatively determined by a linear scaling assuming that 100% drive flow corresponds to 100 Mlb/hr core flow, with no adjustment made for expected deviations between core flow and drive flow below 100%.

The Frequency of 24 months is based on engineering judgment and reliability of the components.

(continued)

SUSQUEHANNA - UNIT 1 TS /B 3.3-41

PPL Rev. 2 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE Average Power Range Monitor (APRM) (continued)

SAFETY ANALYSES, LCO, and Three of the four APRM channels and all four of the voter channels are APPLICABILITY required to be OPERABLE to ensure that no signal failure will preclude (contin'ued) a scram on a valid signal. In addition, to provide adequate coverage of the entire core consistent with the design bases for the APRM Functions 2.a, 2.b, and 2.c, at least (20) LPRM inputs with at least three LPRM inputs from each of the four axial levels 6t which the LPRMs are located must be OPERABLE for each APRM channel, with no more than (9), LPRM detectors declared inoperable since the most recent APRM gain calibration. Per Reference 23, the minimum input requirement for an APRM channel with 43 LPRM inputs is determined given that the total number of LPRM outputs used as inputs to an APRM channel that may be bypassed shall not exceed twenty-three (23). Hence, (20) LPRM inputs needed to be operable. For the OPRM Trip Function 2.f, each LPRM in an APRM channel is further associated in a pattern of OPRM "cells," as described in References 17 and 18.

Each OPRM cell is capable of producing a channel trip signal.

2.a. Average Power Range Monitor Neutron Flux-High (Setdown)

For operation at low power (i.e., MODE 2), the Average Power Range Monitor Neutron Flux-High (Setdown) Function is capable of generating a trip signal that prevents fuel damage resulting from abnormal operating transients in this power range. For most operation at low power levels, the Average Power Range Monitor Neutron Flux-High (Setdown)

Function will provide a secondary scram to the Intermediate Range Monitor Neutron Flux-High Function because of the relative setpoints.

With the IRMs at Range 9 or 10, it is possible that the Average Power Range Monitor Neutron Flux-High (Setdown) Function w) pr*Jde the primary trip signal for a corewide increase in power. (I -z -'

No specific safety analyses take direct credit f eA Range Monitor Neutron Flux-High (Set n) Functti Function indirectly ensures that befor e reactor.

in the run position, reactor power es not exceedi when operating at low reactor essure and low core, Therefore, it indirectly prevents fuel da during significant increases with THERMAL POWER < RTP_ . 7 The Allowable Value is based on prey g significant increases in power when THERMAL POWER is <4 RTP.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-8 S O A-N~~A-U.2 Retyped to

PPL Rev. 2 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.f. Oscillation Power Range Monitor (OPRM) Trip (continued)

SAFETY ANALYSES, LCO, and References 17, 18 and 19 describe three algorithms for detecting APPLICABILITY thermal-hydraulic instability related neutron flux oscillations: the period (continued) based detection algorithm (confirmation count and cell amplitude), the amplitude based algorithm, and the growth rate algorithm. All three are implemented in the OPRM-Trip Function, -but-the safety-analysis takes credit only for the period based detection algorithm. The remaining algorithms provide defense in depth and additional protection against unanticipated oscillations. OPRM Trip Function OPERABILITY for Technical Specification purposes is based only on the period based detection algorithm.

The OPRM Trip Function receives input signals from the local power range monitors (LPRMs) within the reactor core, which are combined into "cells" for evaluation by the OPRM algorithms. Each channel is capable of detecting thermal-hydraulic instabilities, by detecting the related neutron flux oscillations, and issuing a trip signal before the MCPR SL is exceeded. Three of the our channels are required to be OPERABLE.

The OPRM Trip is automioa' Ayenaed (bypass removed) when THERMAL POWER is _, RTP, as indicated by the APRM Simulated Thermal Power, and reactor core flow is - the value defined in the COLR, as indicated by APRM measured recirculation drive flow.

This is the operating region where actual thermal-hydraulic instability and related neutron flux oscillations are expected to occur. Reference 21 includes additional discussion of OPRM Trip enable region limits.

These setpoints, which are sometimes referred to as the "auto-bypass" setpoints, establish the boundaries of the OPRM Trip enabled region.

The APRM Simulated Thermal Power auto-enable setpoint has 1%

deadband while the drive flow setpoint has a 2% deadband. The deadband for these setpoints is established so that it i eases the enabled region once the region is entered. s '" ' , e e Trip Function is required to be OPERA when t e plant is at RTP. Th RTP level is selecte o provide argin in the un ikely even* areactor power increas t nsient o rring withou or action while the plant is op ing below&/o RTP causes a power increase to or beyond the APRM Simulated

.Thermal Power OPRM Trip auto-enable setpoint. This OPERABILITY requirement assures that the OPRM Trip auto-enable function will be OPERABLE when required.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-14 Retyped4er-Include P S and AR S ELLA

PPL Rev. 2 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 8. Turbine Stop Valve-Closure (continued)

SAFETY ANALYSES, LCO, and valves. The Turbine Stop Valve-Closure Function is the primary scram APPLICABILITY signal for the turbine trip event analyzed in Reference 5. For this event, (continued) the reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the End of Cycle Recirculation Pump Trip

-(EOC RPT)-Sy-sttr, e t SL is not exceeded.

Turbine Stop Valve-Closure signals are initiated from' position switches located on each of the four TSVs. Two independent position switches are associated with each, stop valve. One of the two switches provides input to RPS trip system A; the other, to RPS trip system B. Thus, each RPS trip system receives an input from four Turbine Stop Valve-Closure channels, each consisting of one position switch. The logic for the Turbine Stop Valve:--Closure Function is such that three or more TSVs must be closed to produce a scram. This Function must be M W RTP. This is accomplished automatically by pressure instruments sensing turbine first stage pressure. Because an increase in the main turbine bypass flow can_

ffect this function non-conservatively, THERMAL POWER is derived fTm first stage pressure. The main turbine bypass valves must not c-*se the trip Function to be bypassed when THERMAL POWER is

_f/t RTP.

The Turbine Stop Valve--Closure Allowable Value is selected to be high

enough to detect imminent TSV closure, thereby reducing the severity of the subsequent pressure transient.

Eight channels (arranged in pairs) of Turbine Stop Valve-Closure Function, with four channels in each trip system,. are required to be OPERABLE to ensure that no single instrument failure will precludea scram from this Function if any three TSVs should close. This Function is co nt with analysis assumptions, whenever THERMAL POWERs I /I RTP. This Function is not required when THERMAL RTP since the Reactor Vessel Steam Dome

P:*-*1igh and the Average Power Range Monitor Neutron Flux-High Functions are adequate to maintain the necessary safety margins.

(continued)

Incnined SUSQUEHANNA - UNIT 2 TS / B 3.3-21

PPL Rev. 2 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 9: Turbine Control Valve Fast Closure. TriD Oil Pressure-Low SAFETY ANALYSES, LCO, and Fast closure of the TCVs results in the loss of a heat sink that produces APPLICABILITY reactor pressure, neutron flux, and heat flux transients that must be (continued) limited. Therefore, a reactor scram is initiated on TCV fast closure in anticipation of the transients that would result from the closure of these

-7vaIves--The-Turbine-Cont rDIVaIve-Fa-st-Closire--Tfi-Oil PPs-ure-ý-Low Function is the primary scram signal for the generator load rejection event analyzed in Reference 5. For this event, the reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the EOC-RPT System, ensures that the MCPR SL is not exceeded.

Turbine Control Valve Fast Closure, Trip Oil Pressure-Low signals are initiated by the electrohydrauliccontrol (EHC) fluid pressure at each control valve. One pressure instrument is associated with each control valve, and the signal from each transmitter is assigned to a separate RPS logic channel. This Function must be enabled at THERMAL POwER A RTP. This is accomplished automatically by pressure I-en st-sensing turbine first stage pressure. Because an increase in the main turbine bypass flow can affect this function non-conservatively, SHERMAL POWERis derived from first stage pressure. The main ne b pass valves must not cause the trip Function to be bypassed

-- whenTH E*/ RTP.

The Turbine Control Valve Fast Closure, Trip Oil Pressure-Low Allowable Value is selected high enough to detect imminent TCV fast closure.

Four channels of Turbine Control Valve Fast Closure, Trip Oil Pressure-Low Function with two channels in each trip system arranged in a one-out-of-two logic are required to be OPERABLE to ensure that no

single instrument failure will preclude a scram from this Function on a valid .si I ... ........ . consistent with the analysis sumptions, whenever THERMAL POWE Ao RTP. This Function is not required when THERMAL POWER is o RTP, since

?,i..Cthe Reactor Vessel Steam rna- wFe-_,_ntheaDn Average Power Range Monitor*Neutron Flux-High Functions are adequate to maintain~the necessary safety margins.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-22

PlPL Rev. 2 RPS Instrumentation B 3.3.1.1

-BASES SURVEILLANCE SR 3.3.1.1.1 and SR 3.3.1.1.2 (continued)

REQUIREMENTS (continued) Agreement criteria which are determined by the plant staff based on an investigation of a combination of the channel instrument uncertainties, may be used to support this parameter comparison and include indication and readability. If a channel is outside the criteria, it may be an indication

-that-the-instr--ment-has-drifted-outside-its-limit-and does not necessarily indicate the channel is Inoperable.

The Frequency of once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for SR 3.3.1.1.1 is based upon operating experience that demonstrates that channel failure is rare.

The Frequency of once every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for SR 3.3.11.1.2 is based upon operating experience that demonstrates that channel failure is rare and the evaluation in References 15 and 16. The CHANNEL CHECK supplements less formal checks of channels during normal operational use of the displays associated with the channels required by the LCO.

SR 3.3.1.1.3 To ensure that the APRMs are accurately indicating the true core average power, the APRMs are calibrated to the reactor power calculated from a heat balance. The Frequency of once per.7 days is based on minor changes in LPRM sensitivity, which could affect the APRM reading between performances of SR 3.3.1.1.8.

A restriction to satis RTP is provided that require to be met ,o RTP because it is difficult to urately majM indication of core THERMAL POWER consi with a h /,YoRTP. At low power levels, a 4y-gh sfee7-,6' accuracy is unnecessary because of the large, inherent "fargntoterrnal limits (M PR- I'Hll 'n ARLHG.*). ,A4------

./ 6 RTP, urvei ance is required to have been satisfactorily pararmed within thp. l.st 7 dvn inaccordance with SR 3.0.2. A Note is provided which allows an increase in THERMAL W-ER bak. if the 7 day

.In this evVii, the Srmrust be 21 performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reaching or exceen o RTP..

Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-31 NMS and A L

PPL Rev. 2 RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.15 (continued)

REQUIREMENTS (continued) The LOGIC SYSTEM FUNCTIONAL TEST for APRM Function 2.e simulates APRM and OPRM trip conditions at the 2-out-of-4 Voter channel inputs to check all combinations of two tripped inputs to the 2-out-of-4 logic in the voter channels and APRM related redundant RPS

--relays.

The 24 month Frequency is based on the need to perform portions of this Surveillance under the 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 that these components usually pass the Surveillance when performed at the 24 month Frequency.

SR 3.3.1.1.16 This SR ensures that scrams iniliated from the Turbine Stop Valve-Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure-Low Functions will not be inadvertently bypassed when THERMAL POWER is 0o RTP. This is performedly a Functional check that ensures the cram featurge i se /. RTP. Because main turbine bbypa can affect this function nonconservatively (THERMAL WeER is derived from turbine first stage pressure), the opening of the main turbine bypass valves must not cause the trip Function to be bypassed when Thermal Powe 7a RTP.

any bypass channel's trip function is nonconservative (i.e., the

. RTP, either due to open main turbine

bypass valve(s) orother rea ons), then the affected Turbine Stop

Valve-Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure-Low Functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition (nonbypass).

If placed in the nonbypass condition, this SR is met and the channel is considered OPERABLE.

The Frequency of 24 months is based on engineering judgment and reliability of the components.

(continued)

SUSQUEHANNA - UNIT 2 TS /B 3.3-38 Inl RNM T/EL

PPL Rev. 2 RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.19 REQUIREMENTS (continued) This surveillance involves confirming the OPRM Trip auto-enable setpoints. The auto-enable setpoint values are considered to be nominal values as discussed in Reference 21. This surveillance ensures that the OPRM Trip is enabled (not bypassed) for the correct výlues of AP-RM-Simulated Thermal Power and recirculation drive flow.

Other surveillances ensure that the APRM Simulated Thermal Power and recirculation drive flow properly correlate with THERMAL POWER (SR.3.3.1.1.2) and core flow (SR 3.3.1.1.20), respectively.

If any auto-enable setpoint is nonconserva the OPRM Trip is bypassed when APRM Simulated Thermal Powe- and recirculation drive flow 5 value equivalent to the core flow value defined in the COLR, then the affected channel is considered inoperable for the OPRM Trip Function. Alternatively, the OPRM Trip auto-enable setpoint(s) may be adjusted to place the channel in a conservative condition (not bypassed). If the OPRM Trip is placed in the not-bypassed condition, this SR is met and the channel is considered OPERABLE.

For purposes of this surveillance, consistent with Reference 21, the conversion from core flow values defined in the COLR to drive flow values used for this SR can be conservatively determined by a linear scaling assuming that 100% drive flow corresponds to 100 MIb/hr core flow, with no adjustment made for expected deviations between core flow and drive flow below 100%.

The Frequency of 24 months is based on engineering judgment and reliability of the components.

SR 3.3.1.1.20 The APRM Simulated Thermal Power-High Function (Function 2.b) uses drive flow to vary the trip setpoint. The OPRM Trip Function (Function 2.f) uses drive flow to automatically enable or bypass the OPRM Trip output to RPS. Both of these Functions use drive flow as a representation of reactor core flow. SR 3.3.1.1.18 ensures that the drive flow transmitters and processing electronics are calibrated. This SR adjusts the recirculation drive flow scaling factors in each APRM channel to provide the appropriate drive flow/core flow alignment..

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-41 ~~~-R~tyed to I clud -~fMS and ART-S~tvtEtt

Attachment 3 to PLA-6254 Revised No Significant Hazards Consideration

Attachment 3 to PLA-6254 Page 1 of 10 5.1 No Significant Hazards Consideration PPL Susquehanna has evaluated whether or not a significant hazards consideration is involved with the proposed change, by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Extended Power Uprate Response: No.

The probability (frequency of occurrence) of a Design Basis Accident occurring is not affected by the increased power level, because Susquehanna continues to comply with the regulatory and design basis criteria established for plant equipment. A probabilistic risk assessment demonstrates that the calculated core damage frequencies do not significantly change due to Constant Pressure Power Uprate (CPPU). Scram setpoints (equipment settings that initiate automatic plant shutdowns) are established such that there is no significant increase in scram frequency due to CPPU. No new challenges to safety-related equipment result from CPPU.

The changes in consequences of postulated accidents, which would occur from 102% of the CPPU rated thermal power (RTP) compared to those previously evaluated, are acceptable. The results of CPPU accident evaluations do not exceed the NRC-approved acceptance limits. The spectrum of postulated accidents and transients has been investigated, and are shown to meet the plant's currently licensed regulatory criteria. In the area of fuel and core design, for example, the Safety Limit Minimum Critical Power Ratio (SLMCPR) and other applicable Specified Acceptable Fuel Design Limits (SAFDLs) are still met.

Continued compliance with the SLMCPR and other SAFDLs will be confirmed on a cycle-specific basis consistent with the criteria accepted by the NRC.

Challenges to the Reactor Coolant Pressure Boundary were evaluated at CPPU conditions (pressure, temperature, flow, and radiation) were found to meet their acceptance criteria for allowable stresses and overpressure margin.

Challenges to the containment have been evaluated, and the containment and its associated cooling systems continue to meet 10 CFR 50, Appendix A, Criterion 16, Containment Design; Criterion 38, Containment Heat Removal; and Criterion 50, Containment Design Basis. The increase in the calculated post-LOCA suppression pool temperature above the currently assumed peak temperature was evaluated and determined to be acceptable.

Attachment 3 to PLA-6254 Page 2 of 10 Radiological release events (accidents) have been evaluated, and meet the guidelines of 10 CFR 50.67.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

RHR Service Water System and Ultimate Heat Sink Technical Specification and Methods Change Response: No.

The proposed changes do not involve any new initiators for any accidents nor do they increase the likelihood of a malfunction of any Structures, Systems or Components (SSCs). Implementation of the subject changes reduces the probability of adverse consequences of accidents previously evaluated, because inclusion of the manual spray array bypass isolation valves and the small spray array isolation valves in the Technical Specifications (TS) increases their reliability to function for safe shutdown.

The use of the ANS/ANSI-5.1-1979 decay heat model in the UHS performance analysis is not relevant to accident initiation, but rather, pertains to the method used to evaluate currently postulated accidents. Its use does not, in any way, alter existing fission product boundaries, and provides a conservative prediction of decay heat. Therefore, the change in decay heat calculational method does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Therefore, the proposed fehange does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Containment Analysis Methods Change Response: No.

The use of passive heat sinks and the ANS/ANSI-5.1-1979 decay heat model are not relevant to accident initiation, but rather, pertain to the method used to evaluate postulated accidents. The use of these elements does not, in any way, alter existing fission product boundaries, and provides a conservative prediction of the containment response to DBA-LOCAs. Therefore, the Containment Analysis Method Change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Attachment 3 to PLA-6254 Page 3 of 10 Feedwater Pump / Condensate Pump Trip Change Response: No Feedwater pump trips and condensate pump trips rarely occur. A low water level SCRAM on loss of one feedwater pump or one condensate pump is bounded by the loss of all feedwater transient in Final Safety Analysis Report (FSAR)

Appendix 15E. A trip of one feedwater pump or a trip of one condensate pump does not result in the loss of all feedwater. The Feedwater Pump / Condensate Pump Trip Change is included in the CPPU Probabilistic Risk Assessment (PRA).

The best estimate for the Susquehanna Steam Electric Station (SSES) Core Damage Frequency (CDF) risk increase due to the CPPU is 6E-08 for Unit 1 and 7E-08 for Unit 2 which are in the lower left comer of Region III of Regulatory Guide Regulatory (Reference 15) (i.e., very small risk changes). The best estimate for the Large Early Release Frequency (LERF) increase is 1.OE-09/yr for both units, which is also in the lower left comer of the Region III range of Regulatory Guide 1.174. Therefore, the Feedwater Pump / Condensate Pump Trip Change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Main Turbine Pressure Regulation System Response: No.

Technical Specification 3.7.8 does not directly or indirectly affect any plant system, equipment, component, or change the process used to operate the plant.

Technical Specification 3.7.8 would ensure acceptable performance, since it would establish requirements for adhering to the appropriate thermal limits, depending on the operability of the main turbine pressure regulation system. Use of the appropriate limits assures that the appropriate safety limits will not be exceeded during normal or anticipated operational occurrences. Thus, Technical Specification 3.7.8 does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Extended Power Uprate Response: No.

Equipment that could be affected by EPU has been evaluated. No new operating mode, safety-related equipment lineup, accident scenario, or equipment failure mode was identified. The full spectrum of accident considerations has been

Attachment 3 to PLA-6254 Page 4 of 10 evaluated and no new or different kind of accident has been identified. CPPU uses developed technology and applies it within capabilities of existing or modified plant safety related equipment in accordance with the regulatory criteria (including NRC approved codes, standards and methods). No new accidents or event precursors have been identified.

The SSES TS require revision to implement EPU. The revisions have been assessed and it was determined that the proposed change will not introduce a different accident than that previously evaluated. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

RHR Service Water System and Ultimate Heat Sink Technical Specification and Methods Change Response: No.

The subject changes apply Technical Specification controls to new UHS manual bypass isolation valves and the existing small spray array isolation valves. The design functions of the systems are not affected.

The addition of manually operated valves in the system, operational changes and the Technical Specification changes do not create the possibility of a new or different kind of accident from any previously evaluated.

The use of the ANS/ANSI-5.1-1979 decay heat model is not relevant to accident initiation, but rather pertains to the method used to evaluate currently postulated accidents. The use of this analytical tool does not involve any physical changes to plant structures or systems, and does not create a new initiating event for the spectrum of events currently postulated in the FSAR. Further, it does not result in the need to postulate any new accident scenarios. Therefore, the decay heat calculational method change does not create the possibility of a new or different kind of accident from any accident previously evaluated Containment Analysis Methods Change Response: No.

The use of passive heat sinks and the ANS/ANSI-5.1-1979 decay heat model are not relevant to accident initiation, but pertain to the method used to evaluate currently postulated accidents. The use of these analytical tools does not involve any physical changes to plant structures or systems, and does not create a new initiating event for the spectrum of events currently postulated in the FSAR.

Further, they do not result in the need to postulate any new accident scenarios.

Attachment 3 to PLA-6254 Page 5 of 10 Therefore, the Containment Analysis Method Change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Feedwater Pump / Condensate Pump Trip Change Response: No The occurrence of a reactor SCRAM is already considered in the current licensing basis and is not an accident. A SCRAM resulting from the trip of a feedwater pump or a condensate pump is bounded by a loss of all feedwater event. The loss of all feedwater transient is already considered in the plant licensing basis. The SCRAM due to the feedwater or condensate pump trip does not change the results of the loss of all feedwater transient in any way. Therefore, the Feedwater Pump /

Condensate Pump Trip Change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Main Turbine Pressure Regulation System Response: No.

Technical Specification 3.7.8 w:ll not directly or indirectly affect any plant system, equipment, or component and therefore does not affect the failure modes of any of these items. Thus, Technical Specification 3.7.8 does not create the possibility of a previously unevaluated operator error or a new single failure.

Therefore, Technical Specification 3.7.8 does riot create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Extended Power Uprate Response: No.

The CPPU affects only design and operational margins. Challenges to the fuel, reactor coolant pressure boundary, and containment were evaluated for CPPU conditions. Fuel integrity is maintained by meeting existing design and regulatory limits. The calculated loads on affected structures, systems and components, including the reactor coolant pressure boundary, will remain within their design allowables for design basis event categories. No NRC acceptance criterion is exceeded. Because the SSES configuration and responses to transients and postulated accidents do not result in exceeding the presently approved NRC

Attachment 3 to PLA-6254 Page 6 of 10 acceptance limits, the proposed changes do not involve a significant reduction in a margin of safety.

RHR Service Water System and Ultimate Heat Sink Technical Specification and Methods Change Response: No.

Implementation of the subject changes does not significantly reduce the margin of safety since these changes add components and Technical Specification controls for the components not currently addressed in the Technical Specifications. These changes increase the reliability of the affected components/systems to function for safe shutdown.

Therefore, these changes do not involve a significant reduction in margin of safety.

The ANS/ANSI-5.1-1979 model provides a conservative prediction of decay heat.

The use of this element is consistent with current industry standards, and has been previously accepted by the staff for use in containment analysis by other licensees, as described in GE Nuclear Energy. "Constant Pressure Power Uprate," Licensing Topical Report NEDC-33004P-A, Revision 4, dated July 2003; and the letter to Gary L. Sozzi (GE) from Ashok Thandani (NRC) on the Use of the SHEX Computer Program and ANSI/ANS 5.1-1979, "Decay Heat Source Term for Containment Long-Term Pressure and Temperature Analysis," July 13, 1993.

Therefore, the decay heat calculational method change does not involve a significant reduction in the margin of safety.

Containment Analysis Methods Change Response: No.

The use of passive heat sinks and the ANS/ANSI-5.1-1979 decay heat model are realistic phenomena, and provide a conservative prediction of the plant response to DBA-LOCAs. The use of these elements is consistent with current industry standards, and has been previously accepted by the staff for other licensees, as described in GE Nuclear Energy: "Constant Pressure Power Uprate," Licensing Topical Report NEDC-33004P-A, Revision 4, dated July 2003; the letter to Gary L. Sozzi (GE) from Ashok Thandani (NRC) on the Use of the SHEX Computer Program; and ANSI/ANS 5.1-1979, "Decay Heat Source Term for Containment Long-Term Pressure and Temperature Analysis," July 13, 1993.

Therefore, the Containment Analysis Method Change does not involve a significant reduction in the margin of safety.

Attachment 3 to PLA-6254 Page 7 of 10 Feedwater Pump / Condensate Pump Trip Change Response: No A low water level SCRAM on loss of one feedwater pump or one condensate pump is bounded by the loss of all feedwater transient in FSAR Appendix 15E.

The loss of all feedwater transient is a non-limiting event that does not contribute to the setting of the fuel safety limits. Consequently, a SCRAM resulting from a feedwater pump or condensate pump trip does not reduce the margin to fuel safety limits. Therefore, the potential for a SCRAM resulting from a feedwater pump trip or a condensate pump trip does not involve a significant reduction in the margin of safety.

Main Turbine Pressure Regulation System Since Technical Specification 3.7.8 does not alter any plant system, equipment, component, or processes used to operate the plant, the proposed change will not jeopardize or degrade the function or operation of any plant system or component governed by Technical Specifications. Technical Specification 3.7.8 preserves the margin of safety by establishing requirements for adhering to the appropriate thermal limits.

Conclusion for All Changes Based upon the above, PPL Susquehanna concludes that the proposed amendment presents no significant hazards consideration, under the standards set forth in 10 CFR 50.92(c) and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements / Criteria 5.2.1 Analysis Extended Power Uprate 10 CFR 50.36 (c)(2)(ii) Criterion 2, requires that TS LCOs include process variables, design features, and operating restrictions that are initial conditions of design basis accident analysis. The Technical Specifications ensure that the SSES system performance parameters are maintained within the values assumed in the safety analyses. The Technical Specification changes are supported by the safety analyses that were performed consistent with NRC approved methodology approved for SSES and continue to provide a comparable level of protection as the current Technical Specifications. Applicable regulatory requirements and

Attachment 3 to PLA-6254 Page 8 of 10 significant safety evaluations performed in support of the proposed changes are described in Attachment 4.

RHR Service Water System and Ultimate Heat Sink Technical Specification and Methods Change GDC-5 requires SSCs important to safety not to be shared among nuclear power units unless it can be shown that sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units. The proposed changes do not affect compliance with GDC-5 as described in FSAR Section 3.1. The RHRSW system and UHS continue to be designed such that no single active failure will prevent their safety function from being achieved.

GDC-44 requires that a system to transfer heat from structures, systems, and components important to safety to an ultimate heat sink be provided. The system safety function shall be to transfer the combined heat load of these structures, systems, and components under normal operating and accident conditions. The proposed change does not affect compliance with GDC-44 as described in FSAR Section 3.1. The RHRSW system and the UHS are designed to Seismic Category I requirements. Redundant safety related components served by RHRSW and the UHS are supplied through redundant supply headers and returned through redundant discharge or return lines. Electric power for operation of redundant safety related components of RHRSW is supplied from separate independent offsite and redundant onsite standby power sources. No single active failure renders RHRSW or the UHS incapable of performing its safety function.

Regulatory Guide 1.27 Revision 2 applies to nuclear power plants that use water as the ultimate heat sink. The proposed change does not affect compliance with Regulatory Guide 1.27 Revision 2 as described in Sections 3.13 and 9.2.7 of the FSAR. The UHS continues to be capable of providing sufficient cooling for 30 days to permit simultaneous safe shutdown and cooldown of both SSES units and maintain them in a safe shutdown condition.

10 CFR 50.36(c)(3), requires that Technical Specification LCOs include surveillance requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. The Technical Specification changes are supported by the safety analyses that were performed consistent with NRC approved methodology and continue to provide a comparable level of protection as current Technical Specifications.

Attachment 3 to PLA-6254 Page 9 of 10 Containment Analysis Methods Change 10 CFR 50, Appendix A, GDC-1 6 requires that a reactor containment be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity, and that it be assured that containment design parameters important to safety not be exceeded for as long as postulated accident conditions require.

The evaluations described in Attachment 4, Section 4.1 demonstrate that containment parameters stay within their design limits.

10 CFR 50, Appendix A, GDC-50 requires that the reactor containment structure be designed so that the structure and its internal compartments can accommodate the calculated pressure and temperature conditions resulting from any loss of coolant accident. The evaluations described in Attachment 4, Section 4.1 demonstrate that containment parameters stay within their design limits.

Feedwater Pump / Condensate Pump Trip Change General Design Criterion 10 (GDC 10), "Reactor Design," in Appendix A, "General Design Criteria for Nuclear Power Plants," 10 CFR Part 50 states, in part, that the reactor core and associated -coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded. General Design Criterion 20 (GDC 20), Protection System Functions, in Appendix A, "General Design Criteria for Nuclear Power Plants," 10 CFR Part 50 states, in part that the protection system shall be designed to initiate automatically the operation of appropriate systems including reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a results of anticipated operational occurrences.

During the feedwater pump and condensate pump trip transients for CPPU, the water level may reduce to the level at which a SCRAM is initiated. Fuel design limits are not exceeded during this transient.

The Feedwater Pump / Condensate Pump Trip Change will not cause the MCPR safety limit to be violated nor the fuel cladding strain to exceed 1%. Therefore, the requirements of GDC- 10 and GDC-20 regarding acceptable fuel design limits is satisfied.

Main Turbine Pressure Regulation System Title 10 of the Code of Federal Regulations (10 CFR) establishes the fundamental regulatory requirements with respect to reactivity control systems. Specifically, General Design Criterion 10 (GDC 10), "Reactor design," in Appendix A, "General Design Criteria for Nuclear Power Plants," 10 CFR Part 50 states, in part, that the reactor core and associated coolant, control, and protection systems

Attachment 3 to PLA-6254 Page 10 of 10 shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded.

Technical Specification 3.7.8 will ensure that the MCPR Safety Limit will not be violated and that fuel cladding strain will not exceed 1%. This satisfies the requirement of GDC-10 regarding acceptable fuel design limits.

5.2.2 Conclusion Based on the analyses provided in Section 4, Technical Analysis, the proposed change is consistent with applicable regulatory requirements and criteria. In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such 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 defense and security or to the health and safety of the public.

v t e Letter Sequence Withdrawal
TAC:MD3309, (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance, Acceptance, Acceptance, Acceptance Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement Administration Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance Questions 19 March 2007 15 March 2007 20 March 2007 29 March 2007 12 April 2007 10 April 2007 12 April 2007 16 April 2007 12 April 2007 12 April 2007 18 April 2007 30 April 2007 27 April 2007 27 April 2007 27 April 2007 30 April 2007 14 May 2007 12 July 2007 1 August 2007 Answers Results Approval Other: ML071620334, ML073320383, ML073390289, ML073390306, ML081000038, PLA-6187, Response for Additional Information Requested in Regard to Proposed License Amendment No. 285, Constant Pressure Power Uprate Environmental, PLA-6255, Proposed License Amendments 285 and 253, Constant Pressure Power Uprate Application Regarding Steam Dryer Flow Effects Verbal and Confirmatory Requests for Additional Information Responses, PLA-6266, Proposed License Amendment No. 285 for Unit 1, License No. NPF-14 & Proposed License Amendment No. 253 for Unit 2, License No. NPF-22, Extended Power Uprate Application Regarding Change to Technical Specification SR 3.7.1 - RHRSW and UHS, PLA-6370, License Renewal Application, Amendment to Section 4.2.2, PLA-6633, Providing Notice That the License Conditions 2.C.(36)(a)3, 2.C.(36(b)7 & 2.C.(36)(f) Have Been Met, PLA-6876, Th Rio Steam Dryer Visual Inspection Report, PLA-7047, Th Rio Steam Dryer Visual Inspection Report
MONTHYEAR2007-04-12 [Table View]

PLA-6254, Proposed License Amendments 285 & 253, Extended Power Uprate Application Regarding Withdrawal of Change to Technical Specification SR 3.3.1.1.8 - Calibration Frequency for Lprms

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PLA-6254, Proposed License Amendments 285 & 253, Extended Power Uprate Application Regarding Withdrawal of Change to Technical Specification SR 3.3.1.1.8 - Calibration Frequency for Lprms

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