View source for PLA-6977, Proposed Amendment No. 313 to License NPF-14 and Proposed Amendment No. 285 to License NPF-22: Change to Technical Specification Surveillance Requirement (SR) 3.5.1.12

{{Adams
| number = ML13158A096
| issue date = 06/06/2013
| title = Proposed Amendment No. 313 to License NPF-14 and Proposed Amendment No. 285 to License NPF-22: Change to Technical Specification Surveillance Requirement (SR) 3.5.1.12
| author name = Helsel J
| author affiliation = PPL Susquehanna, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000387, 05000388
| license number = NPF-014, NPF-022
| contact person = 
| case reference number = PLA-6977
| document type = Letter type:PLA, License-Application for Facility Operating License (Amend/Renewal) DKT 50, Technical Specification, Bases Change
| page count = 16
}}

=Text=
{{#Wiki_filter:Jeffrey M. Helsel           PPL Susquehanna, LLC Nuclear Plant Manager                 769 Salem Boulevard Berwick. PA 18603 Tel. 570.542.3510 Fax 570.542.1504 jmhelsel@pplweb.com U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 SUSQUEHANNA STEAM ELECTRIC STATION PROPOSED AMENDMENT NO. 313 TO LICENSE NPF-14 AND PROPOSED AMENDMENT NO. 285 TO LICENSE NPF-22:
CHANGE TO TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENT (SR) 3.5.1.12                                                     Docket Nos. 50-387 PLA-6977                                                                             and 50-388 Pursuant to 10 CFR 50.90, PPL Susquehanna, LLC (PPL), hereby requests approval of the following proposed amendments to the Susquehanna Steam Electric Station (SSES)
Unit 1 and Unit 2 Technical Specifications (TS), as described in the Enclosure. The proposed amendments would change Surveillance Requirements (SR) 3.5.1.12 in Technical Specification 3.5.1 "ECCS -Operating." Specifically, the proposed amendments would eliminate the TS requirement for the Automatic Depressurization System (ADS) valves to open during manual actuation of the ADS circuitry.
Additionally, the proposed amendments change the surveillance frequency from "24 months on a staggered test basis for each valve" to "24 months".
The proposed change will reduce the potential for valve seat misalignment and leakage during operation that can be caused by opening the valve to perform SR 3.5.1.12.
Further, the proposed amendment does not involve a significant hazard consideration as discussed in the evaluation.
PPL requests approval of the proposed change to the Unit 1 and Unit 2 Technical Specifications by May 1, 2014. PPL further requests that the approved amendment be issued to be effective immediately upon approval with the implementation to be completed within 60 days. contains a mark up of the proposed Technical Specification change. contains a markup of the Technical Specification Bases for info1mation.
The change has been reviewed by the SSES Plant Operations Review Committee and by the Susquehanna Review Committee. In accordance with 10 CFR 50.91(b), PPL

Document Control Desk PLA-6977 Susquehanna, LLC is providing the Commonwealth of Pennsylvania with a copy of this proposed License Amendment request.
There are no regulatory commitments associated with the proposed changes.
If you have any questions or require additional infotmation, please contact Mr. Duane L. Filchner at 610-774-7819.
I declare under penalty of petjury that the foregoing is true and con*ect.

==Enclosure:==

PPL Susquehanna Units 1 and 2 Evaluation of Proposed Change toTS SR 3.5.1.12 and the Con*esponding TS Bases Attachments:  Proposed Unit 1 and Unit 2 Technical Specification Changes (Mark-ups)  Proposed Unit 1 and Unit 2 Technical Specification Bases Changes (Mark-ups)
Copy:
NRC Region I Mr. P. W. Finney, NRC Sr. Resident Inspector Mr. J. Whited, NRC Project Manager Mr. L. J. Winker, PA DEP/BRP

Enclosure to PLA-6977 PPL Susquehanna Units 1 and 2 Evaluation of Proposed Change to TS SR 3.5.1.12 and the Corresponding TS Bases
: 1. DESCRIPTION
: 2. PROPOSED CHANGE
: 3. BACKGROUND
: 4. TECHNICAL ANALYSIS
: 5. REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements/Criteria 5.3 Precedents
: 6. ENVIRONMENTAL CONSIDERATIONS
: 7. REFERENCES

Enclosure to PLA-6977 Page 1 of7 PPL EVALUATION

==Subject:==
PPL Susquehanna, LLC Evaluation of Proposed Change to the Unit 1 and Unit 2 Technical Specifications (TS) Surveillance Requirements (SR) 3.5.1.12 in Section 3.5.1 "ECCS - Operating" and the associated TS Bases
: 1. DESCRIPTION PPL Susquehanna, LLC (PPL) proposes to change SR 3.5.1.12 to allow stroking of an Automatic Depressu1ization System (ADS) valve actuator while uncoupled from an ADS valve stem to meet the requirements in SR 3.5.1.12. This change would allow for determination that an ADS valve is OPERABLE without requiring the ADS valve to be opened, thereby reducing the potential for valve seat misalignment and leakage.
The ADS valves that are removed during each refueling outage will continue to be manually actuated during bench-testing at an offsite facility to meet the requirements of the set point testing program. Uncoupling of the ADS valve stem from the plant-installed remote manual actuation equipment will allow increased testing of the manual actuation valve solenoids without cycling the valve.
If approved, PPL plans to implement the proposed Unit 1 and Unit 2 TS amendments within 60 days.
Mark-ups of the proposed change to the Unit I and Unit 2 Technic.al Specifications (TS) and Bases are included in Attachments 1 and 2.
: 2. PROPOSED CHANGE The proposed revision to the Unit 1 and Unit 2 Technical Specification SR 3.5.1.12 changes the cun*ent requirement that "each ADS valve opens when manually actuated" to the requirement that "each ADS valve actuator strokes when manually actuated,.
Additionally, the surveillance frequency is changed from "24 months on a STAGGERED TEST BASIS for each valve solenoid" to "24 months".
Corresponding changes are also required to the TS Bases Sections B SR 3.5.1.12.

Enclosure to PLA-6977 Page 2 of7
: 3. BACKGROUND The SSES Unit 1 and Unit 2 ADS valves are Crosby Dual Function Safety/Relief Valves, Model HB-65-BP. These valves are designed to perform either as a safety valve or as a relief valve. The safety mode of operation (ADS) is independent and separate from the relief mode.
Each Susquehanna unit utilizes six of the reactor safety/relief valves (S/RVs) as ADS valves to reduce reactor pressure during small breaks in the event of High Pressure Coolant Injection (HPCI) failure. After the reactor vessel pressure is reduced to the capability of the low-pressure systems, Residual Heat Removal (RHR) low pressure coolant injection (LPCI) mode and Core Spray (CS), these low pressure systems provide inventory makeup so that acceptable post-accident temperatures are maintained. The ADS functions to depressurize the Reactor Coolant System (RCS) to allow the combination of the LPCI and CS operation to inject into the RCS. The ADS valves can be opened automatically or manually.
SR 3.5.1.12 currently verifies that the ADS valves can be manually opened. The Susquehanna set point testing program includes manual actuation of the ADS valves during a bench-test in accordance with the valve control program. After the ADS valves have been tested, reworked and reinstalled in the plant, SR 3.5.1.12 is performed on those ADS valves to manually actuate them (i.e., opens and closes the valves) utilizing the valve actuator.
: 4. TECHNICALANALYSIS The ADS (safety mode of operation) is initiated when the increasing static inlet steam pressure overcomes the restraining spring and frictional forces acting against the inlet steam pressure to move the disc in the opening direction.
The relief mode of operation is initiated when an electrical signal is received at any or all of the solenoid valves located on the pneumatic relief-mode actuator assembly. The manual actuation of the ADS valves is initiated from the control room. The solenoid and air control valve open to allow an air source to pressurize the lower side of the piston in the pneumatic cylinder to push it upwards. This action is transmitted through a lever arm and dog pivot mechanism, which in turn pulls the valve lifting nut upwards thereby opening the valve to allow steam to discharge through the valve. Upon de-energization of the solenoid, the air valve repositions to allow the pressurized air in the cylinder to vent to atmosphere and thus closing the valve.
Experience at other nuclear sites has indicated that repeated manual actuation of the ADS valves can lead to undesirable seat leakage during operation. ADS valve leakage is

Enclosure to PLA -6977 Page 3 of7 directed to the primary containment suppression pool, which in tutn causes increased suppression pool water temperatures and the need for increased cooling.
Cun*ently, after the S/RVs are removed from the plant they are sent to an offsite test facility where as-found testing, valve rework, and a cettified bench test are performed to demonstrate operation, prior to shipment back toSSES. After the valves are reinstalled in the plant, the ADS valves are tested to meet SR 3.5.1.12 surveillance requirements by manually actuating them from the control room in accordance with plant procedures.
The design of the SSES valves was reviewed by both the vendor and PPL. The design appears to be adequate and is not considered to be a cause for valve seat leakage as a result of manual operation of the ADS valves.
This proposed change to SR 3.5.1.12 will allow the valve stems to be uncoupled from their actuators to allow testing of the manual actuation electrical circuitry, manual actuation of the solenoid and air control valve, and the operation of the pneumatic cylinder without causing the ADS valve to change position (open).
The ADS valves will continue to be manually actuated during bench-testing at an offsite test facility as a part of certification testing in which the set point for each S/RV is verified prior to installation in the plant. The cunent practice of replacing a portion of the S/RVs each operating cycle (24 months) will be maintained. A main disk exercise test, performed on all SIRVs at least once every 6 years and when the entire valve assembly is shipped to the certified test facility, will ensure that the main disks can freely open.
* The only potential for a failure would be a human performance issue related to incotTect reassembly of the valve and actuator. This is a relatively simple procedure that requires removal of two cotter pins and reassembly in accordance with a plant procedure.
Independent verification of the valve reassembly is required by the procedure to ensure that the valve is properly reassembled.
The proposed change to SR 3.5.1.12 also requires testing of all manual valve solenoids on a 24 month frequency rather than a staggered test frequency basis. This change is beneficial since the operation of the all the actuators will be demonstrated every 24 months.
Therefore, all the components of the ADS will continue to be tested at the current frequency in accordance with the setpoint testing program requirements and this proposed change to SR 3.5.1.12.

Enclosure to PLA-6977 Page4 of7
: 5. REGULATORY SAFETY ANALYSIS 5.1      No Significant Hazards Consideration
: 1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The proposed change does not modify the method of demonstrating the operability of the Safety/ Relief Valves (S/RVs) in both the safety and relief modes of operation. The proposed change does modify the method for demonstrating the proper mechanical functioning of the SIRVs. The SIRVs are required to function in the safety mode to prevent ovetpressurization of the reactor vessel and reactor coolant system pressure boundary during various analyzed transients, including Main Steam Isolation Valve closure. SIRVs associated with the Automatic Depressurization System are also required to function in the relief mode to reduce reactor pressure to permit injection by low pressure Emergency Core Cooling System (ECCS) pumps during certain reactor coolant pipe break accidents. The cun*ent testing method demonstrates the proper mechanical functioning of the SIRVs in both modes through manual actuation of the S/RVs. The proposed testing method results in acceptable demonstration of the SIRV functions in both the safety and relief modes, and therefore provides assurance that the probability of S/RV failure will not increase. None of the accident safety analyses are affected by the requested TS changes and the consequences of accidents mitigated by the SIRVs will not increase.
Therefore, the proposed amendment does not result in a significant increase in the probability or consequences of any previously evaluated accident.
. 2. Do the proposed changes create the possibility of a nelv or different kind of accident from any accident previously evaluated?
Response: No The proposed change modifies the method of testing of the SIRVs, but does not alter the functions or functional capabilities of the SIRVs. Testing under the proposed method is perfotmed in offsite test facilities and in the plant during outage periods when the SIRV functions are not required. Existing analyses address events involving an S/RV inadvertently opening or failing to reclose. Analyses also address the failure of one or more S/R.Vs to open. The proposed change does not introduce any new failure mode.

Enclosure to PLA-6977 Page 5 of7 Therefore, it does not create the possibility of a new or different kind of accident from any accident previously evaluated.
: 3. Do the proposed changes involve a significant reduction in a margin of safety?
Response: No The proposed amendment provides for a complete verification of the functional capability of the S/RVs by performing tests, inspections, and maintenance activities without opening the valves while installed in the plant. This alternative testing and associated programmatic controls will provide an overall level of assurance that the S/RVs are capable of performing their intended accident mitigation safety functions. The proposed amendment does not affect the valve setpoints or adversely affect any other operational criteria assumed for accident mitigation. No changes are proposed that alter the setpoints at which protective actions are initiated, and there is no change to the operability requirements for equipment assumed to operate for accident mitigation.
Moreover, it is expected that the alte1native testing methodology will increase the margin of safety by reducing the potential for S/RV leakage resulting from testing.
Additionally, the increased testing frequency of the manual actuation circuitry is beneficial since the valves will no longer be tested on a staggered test frequency.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, PPL concludes that the proposed change does not involve a 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 10CFR50.36 requires in part that the operating license of a nuclear production facility include technical specifications. Paragraph (c)(2)(ii) of that part requires that a limiting condition for operation (LCO) of a nuclear reactor must be established for each item meeting one or more of four criteria. The S(RV functions identified in LCO 3.5.1 meet Criterion 3, "A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier." Paragraph (c)(3) fut1her requires the establishment of surveillance requirements, "relating to test, calibration, or inspection to assure ... that the limiting conditions for operation will be met." As discussed above, the proposed change in the surveillance requirements for the S/RVs are sufficient to demonstrate the safety and relief modes of operation for the S/RVs, and therefore, are sufficient to ensure that the limiting conditions for operation are met.

Enclosure to PLAM6977 Page 6 of7 In conclusion, based on the considerations discussed above, (I) 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 approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.
5.2.1 Precedent The NRC has approved similar license amendment requests for LaSalle Units 1 and 2, Fitzpatrick, and Hope Creek.
: 6. ENVIRONMENTAL CONSIDERATION 10 CFR 51.22(c)(9) identifies certain licensing and regulatory actions, which are eligible for categorical exclusion from the requirement to perform an environmental assessment. A proposed amendment to an operating license for a facility does not require an environmental assessment if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant hazards consideration; (2) result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite; or (3) result in a significant increase in individual or cumulative occupational radiation exposure. PPL Susquehanna, LLC has evaluated the proposed change and has determined that the proposed change meets the eligibility criteria for categorical exclusion set fo11h in 10 CFR 51.22(c)(9).
Accordingly, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with issuance of the amendment. The basis for this determination, using the above criteria, follows:
As demonstrated in the "No Significant Hazards Consideration'' evaluation, the proposed amendment does not involve a significant hazards consideration.
There is no significant change in the operational transients. The proposed change does not involve any physical alteration of the plant (no new or different type of equipment will be installed) or change in methods governing normal plant operation.
There is no significant increase in individual or cumulative occupational radiation exposure.

Enclosure to PLA-6977 Page 7 of7
: 7. REFERENCES 7.1  SSES Final Safety Analysis Report, Chapter 6.
7.2  INPO Event Report, Level 3, 11-24, "Maintenance Error results in Uncontrolled Reactor Vessel Depressurization," dated July 26, 2011.
7.3  NRC Information Notice 2003-01, "Failure of a Boiling Water reactor Target Rock Main Steam Safety/ReliefValve,'' dated January 15,2003.
7.4  General Electric Nuclear Energy Service Information Letter, "Target Rock Safety Relief Valve Failure to Open," dated December 20,2002.

Attachment 1 to PLA-6977 Technical Specification Markups (Unit 1 and 2)

PPL Rev. 4 ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE                                              FREQUENCY SR 3.5.1.9   --------------------*******NOTE---------*****************-....
Not required to be performed until12 hours after reactor steam pressure and flow are adequate to perform the test Verify~  with reactor pressure~ 165 psig. the HPCI             24 months pump can develop a flow rate ~ 5000 gpm against a system head corresponding to reactor pressure.
SR 3.5.1.1 0 -------------***----------------NOTE-----*---------------*****
Vessel Injection/spray may be excluded.
Verify each ECCS Injection/spray subsystem                     24 months actuates on an actuaJ or simulated automatic initiation signal.
SR 3.5.1. 11 --------------------*--***-----NOTE----*-****-**------------
Valve actuation may be excluded.
Verify the ADS actuates on an actual or simulated              24 months automatic initiation srgnal.
SR 3.5.1.12 ---*-******-*****-*--*******--*NOTE----***--********-*-*******
Not required to be performed until12 hours after reactor steam pressure and flow are adequate to perform the test.
Verify each ADS valve ep&R&:actuator strokes                   24 months oo-a when manually actuated.                                        S-lAGGeReD TeST BJ',SIS fe~
(continued)
I SUSQUEHANNA - UNIT 1                                                                   Amendmentm

PPL Rev.1 ECCS - Operating 3.5.1 SURVEILLANCE REQUJRE=M=E=N.:....:.T~S__.(c.;;...;o;.;...:;nt=ln=ue.;;.;.~d)"------.....--------
SURVEILLANCE                                             FREQUENCY SR 3.5.1.9  *****-*******--*************NOTE~****--**********************
Not required to be performed until12 hours after reactor steam pressure and flow are adequate to perform the test.
Verify. with reactor pressures 165 pslg, the HPCI              24 months pump can develop a fJow rate~ 5000 gpm against a system head corresponding to reactor pressure.
SR 3.5.1.10 ***************************-***NOTE*-*********************--
Vessel injection/spray may be excluded.
Verify each ECCS Injection/spray subsystem                     24months actuates on an actual or simulated automaUc initiation signal.
SR 3.5.1.11 **********-*************'"***-NOTE--*****--*******-********
Valve actuatlon may be excluded.
Verify the ADS actuates on an actual or simulated             24 months automatic Initiation signal.
SR 3.5.1.12 -*********-**-****************-NOTE*************-----****-*--
Not required to be performed untu 12 hours after reactor steam pressure and flow are adequate to perform the test.
Verify each ADS valve e~actuator strokes                      24 months eR-a when manually actuated.                                       s:rAGGI!:RED TEST BASIS fer eaeh *.ralve-seleneid (continued)
SUSQUEHANNA* UNIT 2                          3.5*6                                      Amendment 1'61

Attachment 2 to PLA-6977 Technical Specification Bases Markups (Unit 1 and 2)

PPLRev.~
ECCS-Operating 83.5.1 BASES SURVEILLANCE    SR 3.5.1.12 {continued)
REQUIREMENTS overpressure protection are verified, per ASME requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours after reactor steam pressure and flow are adequate to perform the test. The 12 hours allowed for manual actuation after the required pressure is reached is sufficient to achieve stable conditions and provides adequate time to complete the Surveillance.
Another method is by manual actuation of the ADS valve at atmospheric temperature and pressure during coJd shutdown. When using this method, proper functioning of the valve actuator and solenoid is demonstrated by visual observation of actuator movement. Actual diss travel is measured dHring vahm refurbishment ans testing f)er AS Me rec:twirements. Lifting the va1*1e at atmosJ>heric pressure is tho preferreeJ method because lifting the valves 'Nfth steam flov.* increases the likelihood that the vai*Je 'Nill leak. The ADS actuator will be disconnected from the valve to ensure no damage is done to the valve seat or to the valve internals. Each valve shall be bench-tested prior to reinstallation. The bench- test along with the test on the ADS actuator establishes the OPERABILITY of the valves. The Note that modifies this SR is not needed when this method is used because the SR is performed during cold shutdown.
SR 3.5.1.11 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function. The Frequency of 24 months on a STAGGeReD TeST BASIS ensures that both solenoids for each ADS valve are alternately tested. The Frequency is based on the need to perform the Surveillance under the conditions that apply just prior to or during a startup from a plant outage. Operating experience has shown that these components usually pass the SR when performed at the 24 month Frequency. which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
SR 3.5.1.13 This SR ensures that the ECCS RESPONSE TIME for each ECCS injection/spray subsystem is less than or equal to the maximum value assumed in the accident analysis. Response Time testing acceptance criteria are included in Reference 13. This SR is modified by a Note that allows the instrumentation portion of the response time to be assumed to be based on historical response time data and therefore, is excluded from the ECCS RESPONSE TIME testing. This is allowed since the instrumentation (continued)
I SUSQUEHANNA- UNIT 1                   TS I 8 3.5-17                                    Revision4

PPL Rev. 3 ECCS-Operating 83.5.1 BASES SURVEILLANCE REQUIREMENTS   SR 3.5.1.12 (continued) overpressure protection are verified, per ASME requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until12 hours after reactor steam pressure and flow are adequate to perform the test. The 12 hours allowed for manual actuation after the required pressure is reached is sufficient to achieve stable conditions and provides adequate time to complete the Surveillance.
Another method is by manual actuation of the ADS valve at atmospheric temperature and pressure during cold shutdown. When using this method, proper functioning of the valve actuator and solenoid is demonstrated by visual observation of actuator movement. Aoti:Jal disc tFa'lel is measured during l,<<alve r=eft:lrbishment and testing per ASM~ Fe(ii:Jirements. Lifting tl:le
               .valve at atmospheric pressur:e requires eentroJiing the aotYator to set the valve diso softly on its seat to prevent valve damage. Lifting of the valves at atmospheric pressure is tho prefoFFed method beoat:tse lifting the valves with steam flew inereases the likelihooEI that the valve ,...411 Jeak. The ADS actuator will be disconnected from the valve to ensure no damage is done to the valve seat or to the valve internals. Each valve shall be bench tested prior to reinstallation. The bench-test along with the test on the ADS actuator establishes the OPERABILITY of the valves. The Note that modified this SR is not needed when this method is used because the SR is performed during cold shutdown.
SR 3.5.1.11 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function. The Frequency of 24 months on a ST/\GG~RED TeST BASIS ensures that both solenoids for each ADS valve are alternately tested. The Frequency is based on the need to perform the Surveillance under the conditions that apply just prior to or during a startup from a plant outage. Operating experience has shown that these components usually pass the SR when performed at the 24 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
SR 3.5.1.13 This SR ensures that the ECCS RESPONSE TIME for each ECCS injection/spray subsystem is less than or equal to the maximum value assumed in the accident analysis. Response Time testing acceptance criteria are included in Reference 13. This SR is modified by a Note that allows the instrumentation portion of the response time to be assumed to be based on historical response time data and therefore, is excluded from the ECCS RESPONSE TIME testing. This is allowed since the instrumentation (continued)
SUSQUEHANNA- UNIT 2                TS I B 3.5-16                                      Revision~*}}