ML20217C580
| ML20217C580 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 09/26/1997 |
| From: | Sorensen J NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20217C583 | List: |
| References | |
| NUDOCS 9710010436 | |
| Download: ML20217C580 (10) | |
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Northern states Power company Prairio Island Nuclear Generating Plant 1717 Wakonada Dr. E.
Welch MN 55W9 September 26,1997 10 CFR Part 50 Section 50.90 U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING STATION Docket Nos. 50 282 License Nos. DPR-42 50-306 DPR-60 License Amendment Request Dated September 26,1997 Am9ndment of Auxiliary Feedwater System and instrumentation System Attached is a request for a change to the Technical Specifications, Appendix A of the Operating Licenses, for the Prairie Island Nuclear Generating Plant. This request is submitted in accordance with the provisions of 10 CFR Part 50 Section 50.90.
This amendment request proposes changes to; (1) Technical Specification 3.4 B to i
change the minimum conditions for the Auxiliary Feedwater System (AFW). The proposed change would permit Startup Operations to continue with an inoperable Turbine Priven AFW Pump and/or associated system valves due to an inability to complete Inservice Testing requirements and flow verification following the completion of maintenance, replacement, or repairs performed below 350 F or Cold Shutdown, and (2) Technical Specification Table 3.5.2.B, Engineered Safety Feature Actuation System Instrumentation, to clarify the operability of the AFW system in Mode 2 when the Feedwater Pumps are not required to be operated due to low feed demand.
Exhibit A contains a description of the proposed change, the reason for requesting the change, the supporting safety evaluation and significant hazard determination. Exhibit Y_ '"
B contains current Prairie Island Technical Specification pages marked up to show the proposed changes. Exhibit C contains the revised Technical Specification pages.
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9710010436 970926 PDR ADOCK 0500o282
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USNRC NORTHERN STATF.S POWER COMPANY Sept' ember 26,1997 Page 2 of 2 If you have any questions concerning this License Amendment Request, please contact Gene Eckholt at 612-388-1121 ext. 4063.
Joel P. Sorensen Plant Manager Prairie Island Nuclear Generating Plant l
c:
Regional Administrator-Ill: NRC NRR Project Manager, NRC Senior Resident inspector, NRC State of Minnesota, Attn: Kris Sanda J. E Silberg Attachments:
Affidavit Exhibit A - Evaluation of Pronosed Changes to the Technical Specificatiens I
Exhibit B - Proposed Char
'4arked Up on Existing Technical Specification Pages Exhibit C - Revised Techt...
- m cification Pages
UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING STATION DOCKET NO 50-282 50-306 REQUEST FOR AMENDMENT TO OPERATING LICENSES DPR-42 AND DPR-60 LICENSE AMENDMENT REQUEST DATED September 26,1997 Northern States Power Company, a Minnesota corporation, requests authorization for changes to Appendix A of the Prairie Island Operating License as shown on Attachments labeled as Exhibit A, B, and C. Exhibit A describes the proposed changes, reasons for the changes, safety evaluation and a significant hazards evaluation. Exhibits B and C are copies of the Prairie Island Technical Specifications I
incorporating the proposed changes.
This letter contains no restricted or other defense information.
NORTHERN STATES POWER COMPANY l
By
//Joel P So' rensen' Plant Manager Prairie Island Nuclear Generating Plant On this the b ofb@
/99'hefore me a notary public in and for said county, personally appeared Joel P/Sorenseri, Plant Manager, Prairie Island Nuclear Generating Plant, and being first duly sworn acknowledged that he is authorized to execute this document on behalf of the Northern States Power Company, that he knows the contents thereof, and that to the best of his knowledge, information, and belief the statements made in it are true and that it is no interposed for delay.
O_- he w b
@ NOTARYPUBUCMINNESOTA MARCIA K. LaCORE j
HENNEPIN COUNTY MyConmselon Expiras Jan.31,2000 w::: ::::::::::::::::::::::::::::::a
-t Exhibit A L
Peg 31
.i Exhibit A Prairie Island Nuclear Generating Plant License Amendment Request September 26,1997 Evaluation of Proposed Changes to the Technical Specifications Appendix A of Operating Licenses DPR-42 and DPR-60 Pursuant to 10 CFR Part 50, Sections 50.59 and 50.90, the holders of Operating Licenses DPR-42 and DPR-60 hereby propose the following changes to the Technical Specifications contained in Appendix A of the Facility Operating Licenses:
BACKGROUND This license amendment request proposes changes to Technical Specification (TS) 3.4.B. " Auxiliary Feedwater System" and Table TS.3.5.28 " Engineered Safety Feature Actuation System Instrumentation." The proposed change to TS 3.4.B would provide specific guidance for conducting post m sintenance operational testing of the Turbine Driven Auxiliary Feedwater Pump (TDA :W Pump) and associated system valves in meeting operability and Limiting Conditions for Oreration (LCOs) for startup with an inoperable TDAFW Pump and associated system valves. The proposed change to Table TS.3.5.2B would permit ine bypass of the auto start feature of the AFW pumps for the trip of both Main Feedwater Pumps !n Mode 2 when the Feedwater Pumps are not required to be operated due to low feed demand.
As currently written, TS 3.4.B.1.a requires that the TDAFW Pump and associated system valves for a specific unit be operable to make or maintain the reactor critical or exceed a reactor ccolant system average temperature of 350 F. The LCOs for the Auxiliary Feedwater System do not provide relief in a startup situation in that they require the startup operation be discontinued until an inoperable TDAFW Pump and associated system valves are restored to operable. Operability for ASME CODE Class 1,2, and 3 pumps is determined using the definition of operability and the requirements of TS 4.2.A.2. Since the TDAFW Pump is an ASME Code Class 3 pump, it must meet the additional requirements of TS 4.2.A.2.
In question is the ability to declare the TDAFW Pump and/or associated system valves operable following replacement, repair, or maintenance performed below 350 F or Cold l
Shutdown. TS 4.2.A.2 commits inservice testing to meet the requirements of ASME l
Section XI Boiler and Pressure Vessel Code. The testing of pumps is accomplished in L
accordance with ASME OM-1988 Part 6, This states the following:
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Exhibit A Pcg32 Section 4,4, "Effect of Pump Replacement, Repair, and Maintenance on Reference Values - When a reference value or set of values have been affected by repair, replacement, or routine servicing of a pump, a new reference value or set of values shall be determined or the previous values reconfirmed by an inservice test run before declaring the pump operable."
For pump replacement, repair, or maintenance, the inservice testing must be completed prior to declaring the pump operable. In addition, to test AFW flow for syster.1 valves associated with the TDAFW Pump when maintenance has been performed, requires the use of the TDAFW pump. However, for work conducted during cold shutdown, this testing can not be done until the unit is in startup operations and sufficient steam is available to complete the icsts using the TDAFW Pump.
Engineered Safety Feature Actuation System Instrumentation requirements of Table TS.3.5.2B item 7.d, requires that both channels of the actuation system be operable in Mode 2 for the AFW System. During Startup and Shutdown operations when there is a low feed demand, the normal method for providing Steam Generator feedwater is by using the Motor Driven AFW Pump. This is required at low power levels, because extended operation of the Main Feedwater Pumps in the recirculation mode would cause damage to the Main Feedwater Pumps due to heating and vibration of the l
pumps. With the Main Feedwater Pumps secured during startup and shutdown operations, the auto start feature for trip of both Main Feedwater pumps is bypassed using the Shutdown Auto position of the AFW Pump control switches to prevent the inadvertent auto start of the AFW Pumps. This proposed change to the Instrumentation Actuation Table clarifies that the auto start feature due to loss of both Main Feedwater Pumps is not a required function in Mode 2 and that the AFW system still is operable because the required functions are capable of being performed. The other auto start features of the AFW pumps (lo-lo steam generator water level, safety injection signal, and loss of both 4.16 kV normal buses (turbine driven pump only)) are still available and the AFW system is capable of performing its required function.
The fo! lowing proposed Technical Specification changes are required to permit startup operation following a shutdown where repair, replacement, or maintenance has been conducted on the TDAFW Pump and/or associated system valves, and clarify that the bypassing of the auto start signal for the AFW pumps in Mode 2 on loss of both Main Feedwater Pumps still maintains operability of the AFW System.
Proposed Chanaes A brief description of the proposed changes is provided below. The specific wording changes to the Technical Specifications are shown in Exhibits B and C.
Exhibit A
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Page 3 Technical Specification 3.4.B.2 Auxiliary Feedwater System Limitina Condition of
' Operation: Changed the requirement to discontinue startup operations if the TDAFW Pump and/or associated system valves are not operable based on the completion of inservice testing requirements and flow verification.
Reason for the Chance: During startup operation, if a TDAFW Pump is inoperable based on not having completed inservice testing or associated system valves are inoperable based on verifying flow, discontinuing startup operations will prevent the pump and associated system valves from being restored to an operable status. Startup operation needs to continue to a condition where sufficient steam is available to complete the inservice testing and flow testing to return the pump and associated system valves to an operable status. Justification is provided below in the Safety Evaluation Section.
Technical Specification 3.4.B.2.a Auxiliary Feedwater System Limitina Condition of Operation: Added requirements that for the startup to continue, the inoperability status of the TDAFW Pump must be based solely on the failure to perform inservice testing requirements. Inoperability of the system valves associated with the TDAFW Pump is limited to conducting flow verification testing. The testing must be completed prior to exceeding 10% reactor power or prior to exceeding 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after RCS temperature exceeds 350 F.
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Reason for the Chanae: Limiting the inoperability status of the TDAFW Pump specifically to that caused by an inability to complete inservice testing, requires that the pump must meet the requirements for operability (with the exception of inservice testing) cs stated in the definition section of Technical Specifications for startup to continue. Associated system valves require only flow verification to establish operability. The time requirement is consistent with that permitted for inoperability of the TDAFW Pump and system valves at power and the reactor power limitation is consistent with the existing Surveillance Requirement that verifies AFW flow lollowing cold shutdown. Justification is provided below in the Safety Evaluation Section.
Table TS.3.5-2B Enaineered Safety Feature Actuation System Instrumentation: Added a note to Mode 2 item 7.d for the Auxiliary Feedwater auto start initiated by the trip of both Main Feedwater Pumps. This requirement may be bypassed during Startup and Shutdown Operations when feedwater is being supplied by the AFW pumps. This is required to prevent the inadvertent starting of the AFW pumps at low power levels in Mode 2 with the reactor critcal and before starting or securing of the Main Feedwater System Pumps.
Exhibit A Pcg3 4 Resson for the Chance: The AFW system is still operable in this configuration in Mode
- 2. The only auto start feature of the AFW Pumps provided by the Instrumentation System not in effect, is the Loss of Both Main Feedwater Pumps. At low feed rates when feed is being supplied by an AFW pump, this is not a required function of the AFW System. Justification is provided below in the Safety Evaluation Section.
Safety Evaluation The Auxiliary Feedwater (AFW) System has one turbine driven pump and one motor driven pump each desl nad to supply 200 GPM with the Steam Generator (SG) 0 pressure at the SG safety valve setpoint. Each pump is capable of supplying 100% of the required AFW System flow based on analysis supporting the AFW System design basis. The AFW System design is based on an abnormal operational transient event.
Abnormal operational transients are certain anticipated plant transients which are anticipated to occur in the design life of the plant requiring special attention. The AFW System event is a Loss of Normal Feedwater.
A Loss of Normal Feedwater results in a reduction in capability of the secondary l
system to remove the heat generated in the reactor core. As described in the USAR Chapter 14 " Safety and Accident Analysis", an analysis performed using computer simulation of the plant determined that following a Loss of Normal Feedwater, the AFW System is adequate to remove stored and residual heat. The analysis assumed that only the motor driven AFW Pump would be available at one minute following the initiaCon of the event.
Results of the analysis show that a single motor driven AFW Pump deliverin; 200 GPM starting at one minute does not result in any adverse condition in the core. The Reactor Coolant System (RCS) pressure transient does not result in water release from the pressurizer relief or safety valves, nor does it result in uncovering the tube sheets of the Steam Gene ator being supplied with water.
Currently, Technical Specifications prevent the continuation of startup unless the TDAFW Pump and system valves are operable. The proposed amendments to technical specifications would allow for startup to continue with the Turbine Driven AFW Pump and associated system valves inoperable until a point where sufficient steam is available to complete operational testing per Inservice testing requirements and flow verification. The determination of inoperability for the TDAFW Pump during startup and shutdcan operations must be based only on not meeting the requirements for performing inservice testing. Material deficiencies, non-conforming conditions, required support systems, and periodic testing in accordance with TS section 4.8.A must support operability so that there is reasonable assurance that the TDAFW pump is capable of performing its design function. This will allow a determination of operability for the TDAFW pump as soon as the inservice testing is complete. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time allowed to complete the Inservice testing is consistent with tt. current permitted allowed outage time for the TDAFW Pump during Power Operations. The
Exhibit A Pag @5 10% reactor power limitation is consistent with the current surveillance testing to verify the normal flow path of the AFW system to the steam generators following Cold Shutdown.
The addition of the note in Technical Specification Table TS.S.3.2B item 7.d is not a functional change and only provides clarification for the operability of the AFW System.
In Mode 2, the AFW system is still operable with the auto start feature of the AFW Pumps bypassed for loss of both main feedwater purnps. At low feed rates when feed is being supplied by the Motor Driven AFW pump, this is not a required function of the AFW System.
Based on the above discussion, Northern States Power Company believes there is reasonable assurance that the health and safety of the public will not be adversely affected by these proposed changes to Technical Specifications.
Determination of Sianificant Hazards Considerations The proposed changes to the Operating License have been evaluated to determine whether they constitute a significant hazards consideration F.s required by 10 CFR Part 50, cection 50.91 using the standards provided in section 50.92. Th.s analysis is provided below:
- 1. The proposed amendment will not involve a sianificant increase in the probability or consecuences of an accident previously eva!uated.
Since none of the proposed changes involve a physical change to the plant, the mechanisms that could cause a Loss of Normal Feedwater have not changed. The probability that a Loss of Normal Feedwater will occur is not altered.
This change still requires that the motor driven AFW Pump and assoc;ated system valves are operable during Startup Operations. Analysis of the Loss of Normal Feedwater transient shows that a single AFW Pump provides sufficient AFW flow to prevent any adverse conditions in the core. The condition of an inoperable TDAFW Pump is already permitted during power operations where the consequences of the event would be more severe than during startup. Since there are no consequences from the Loss of Normal Feedwater event at power, the consequences during startup would still be none, but the margins would be larger because; (1) the amount of residual heat generated is less because reactor power at the start of the event is less and (2) the power history is lower resulting in less decay heat.
Thus these changes do not involve an increase in the probability or consequences of an accident previously analyzed.
Exhibit A PQga 6
- 2. The proposed emendment will not create the possibility of a new or different kind of accident from.any accident previously analyzed.
The proposed changm do not create the possibility of a new or different kind of accident previously evaluated because the proposed changes de not introduce a new mode of op3 ration or testing, or make physical changes to the plant The proposed changes do not alter the design, function, operation, or testing of any plant component, therefore the possibility of a new or different kind of accident from those previously analyzed woula not be created by these changes to Technical Specifications.
- 3. Tho procosed amendment will not involve a slanificant reduction in the maroin of safety.
Margins previously established for the Loss of Normal Feedwater event, were analyzed for different initial conditions. The Loss of Normal Feedwater event l
was analyzed for Power Operations. This analysis determined that no adverse conditions would occur in the core. Since there are no consequences from the Loss of Normal Feedwater event at power, the consequences during s'artup would still be none but the margins would be grea'er because; (1) the amount of residual heat generated is less because reactor power at the start of the event is less and (2) the power history is lower causing less decay heat.
Therefore the proposed change does not result in a significant reduction in the margin of safety currently ostablished.
Based on the evaluation described above, and pursuant to 10 CFR Part 50, section 50.91, Northern Statss Power Company has determined that operating the Prairie Island Nuclear Generating Plant in accordance with the propcsed license amendment request does not involve any significant hazards considorations as defined Nuclear Regulatory Commission regulations in 10 CFR. Part 50 section 50.92.
Environmental Assessment Northern States Power Company has evaluated the proposed changes and determined that:
- 1. The changes do not involve a significant hazards consideration, or
- 2. The changes do not involve a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or
Exhibit A Pcga 7
- 3. The changes do not involve a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed changes meet the eligibility criterion for categorical exclusion set forth in 10 CFR part 51 section 51.22(c)(9). Therefore, pursuant to 10 CFR Part 51 section 51.22(b), an environmental assesament of the proposed changes is not required.
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