ML051520473
| ML051520473 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 05/24/2005 |
| From: | Balduzzi M Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 2.05.005 | |
| Download: ML051520473 (13) | |
Text
-a
'~En terg'y Entergy Nuclear Operations, Inc.
Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 Michael A. Balduzzi Site Vice President May 24, 2005 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001
SUBJECT:
Entergy Nuclear Operations, Inc.
Pilgrim Nuclear Power Station Docket No: 50-293 License No: DPR-35 Request for Amendment to the Technical Specifications (TS) - Deletion of Requirement Related to NRC Approval of Engineering Evaluation for Elevated Safety Relief Valve Discharge Pipe Temperature LETTER NUMBER:
2.05.005
Dear Sir or Madam:
Pursuant to 10 CFR 50.90, Entergy Nuclear Operations Inc. (Entergy) proposes to amend the Pilgrim Station Facility Operating License, DPR-35. This proposed license amendment deletes the requirement for requesting NRC approval of an engineering evaluation when a Safety Relief Valve discharge pipe temperature exceeds the limit specified in Technical Specifications.
Entergy has reviewed the proposed amendment in accordance with 10 CFR 50.92 and concludes it does not involve a significant hazards consideration. The attachments provide a description of the proposed change and mark-up of the Technical Specification and Bases pages.
Entergy requests approval of the proposed amendment by June 1, 2006. Once approved, Entergy will implement the amendment within 60 days.
The commitments made in this letter are contained in Attachment 3 of this letter.
If you have any questions or require additional information, please contact Bryan Ford at (508) 830-8403.
UDOo 205005
Entergy Nuclear Operations, 1nc.
Pilgrim Nuclear Power Station Letter Number: 2.05.005 Page 2 I declare under penalty of perjury that the foregoing is true and correct.
Executed on the 24th of May 2005.
WGUdm Attachments
- 1. Evaluation of the Proposed Change (5 pages)
- 2. Mark-up of Technical Specification pages (3 pages)
- 3. Summary of Commitments (1 page) cc:
Mr. John P. Boska, Project Manager Office of Nuclear Reactor Regulation Mail Stop: 0-8B-1 U.S. Nuclear Regulatory Commission 1 White Flint North 11555 Rockville Pike Rockville, MD 20852 U.S.Nuclear Regulatory Commission Region 1 475 Allendale Road King of Prussia, PA 19406 Senior Resident Inspector Pilgrim Nuclear Power Station Ms. Cristine McCombs, Director Mass. Emergency Management Agency 400 Worcester Road P.O. Box 1496 Framingham, MA 01702 Mr. Robert Walker, Director Massachusetts Department of Public Health Radiation Control Program 90 Washington Street Dorchester, MA 02121 205005
ATTACHMENT 1 Evaluation of the Proposed Chanae
Subject:
Deletion of Requirement Related to Request for NRC Approval of Engineering Evaluation for Elevated Safety Relief Valve Discharge Pipe Temperature
- 1.
DESCRIPTION
- 2.
PROPOSED CHANGE
- 3.
BACKGROUND
- 4.
TECHNICAL ANALYSIS
- 5.
REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration 5.2 Environmental Consideration
- 6.
COORDINATION WITH OTHER PENDING TS CHANGES
- 7.
REFERENCES
Letter 2.05.005 Page 1 of 5 Deletion of Requirement Related to Request for NRC Approval of Engineering Evaluation for Elevated Safety Relief Valve Discharge Pine Temperature
- 1.
DESCRIPTION Pursuant to 10 CFR 50.90, Entergy proposes to amend the Technical Specifications (TS) for Pilgrim Nuclear Power Station. This proposed change deletes the TS 3.6.D.4 requirement for NRC approval of an engineering evaluation when a Safety Relief Valve (SRV) discharge pipe temperature exceeds the limit specified in TS 3.6.D.3.
Entergy will continue to comply with the SRV operability requirements and monitor the SRV discharge pipe temperatures as specified in TS 3.6.D.1 and.2 and TS 4.6.D.1 through.4, respectively.
- 2.
PROPOSED CHANGE Technical Specification 3.6.D.2 Note is deleted. TS 3.6.D.3 and.4 are changed to indicate "Deleted". These changes are made on TS pages 3/4.6-6 and 3/4.6-7.
The last sentence in the last paragraph on TS Bases page B3/4.6-8 is deleted. This page is included for information only. provides marked-up TS and Bases pages. A marked-up Bases page is provided for information only.
- 3.
BACKGROUND Pilgrim has four installed two-stage Target Rock SRVs on the main steam lines inside the primary containment, as listed below:
RV-203-3A on Main Steam Line A RV-203-3B on Main Steam Line D RV-203-3C on Main Steam Line D RV-203-3D on Main Steam Line B The SRVs are provided to relieve primary steam to the suppression pool by self-actuation, and each SRV has an ASME rated 898,000 Ibs/hour discharge capacity at 1126 psig as part of the reactor vessel overpressure protection design. The SRVs also fulfill the automatic depressurization function of the core standby cooling systems (emergency core cooling system) under design basis accident conditions via an automatic or manual actuation of the electro-pneumatic portion of the automatic depressurization system (ADS). FSAR Section 4.4 describes the design and operational requirements of the SRVs and FSAR Section 14 describes the related safety analysis.
The current Pilgrim TS 3.6.D.1 and.2 provide SRV operability requirements, and TS 3.6.D.3 and.4 require reporting by way of requesting NRC approval of an engineering evaluation for continued reactor operation beyond 90 days with a SRV discharge pipe temperature greater than 212 F for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Letter 2.05.005 Page 2 of 5 The Note in TS 3.6.D.2 specifically states that Technical Specifications 3.6.D.2 - 3.6.D.5 apply to the installed two-stage Target Rock SRVs.
TS 3.6.D.3 requires an engineering evaluation justifying continued reactor operation if elevated SRV discharge pipe (tail pipe) temperatures in excess of 212' F occur for a period greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. TS 3.6.D.4 requires NRC approval of the evaluation to continue operation beyond 90 days, and corrective actions (removal, testing, repair, recalibration, and reinstallation of the SRV) at the next cold shutdown of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more duration.
TS 3.6.D.1 and.2 (excluding the Note) stipulate SRV operability requirements consistent with those contained in Boiling Water Reactor Standard Technical Specifications (NUREG-1433, Rev. 3) Section 3.4.3 (Reference 1). The requirement for NRC approval of an engineering evaluation is not included in the STS and is adequately addressed by NRC Generic Letter (GL) 91-18, Rev. 1. Accordingly, Entergy proposes to delete TS 3.6.D.3 and.4, with a corresponding revision to the Bases and deletes the Note in TS 3.6.D.2.
- 4.
TECHNICAL ANALYSIS The SRVs are part of the reactor coolant pressure boundary and operate by power actuation (i.e., ADS) or self-actuation by process high pressure. The SRVs limit peak vessel pressure during overpressure transients to satisfy ASME code requirements. The postulated transients for which safety/relief valve actuation is required are described in Chapters 4 and 14 and in Appendices Q and R of the Updated Final Safety Analysis Report. The ADS provides a means to rapidly depressurize the primary system down to a pressure at which low-pressure cooling systems can provide makeup. In the event of a small or medium break loss of coolant accident (LOCA), the ADS function would be required if the high pressure coolant injection system is unable to maintain vessel water level.
TS 3.6.D.1 and.2 govern the operability requirements for the SRVs. No changes are made to these operability requirements.
TS 3.6.D.1 and.2 require that the SRVs be operable during reactor power operations and prior to reactor startup from cold condition or whenever reactor coolant pressure is greater than 104 psig and temperature greater than 3400 F. The surveillance requirements 4.6.D.1 and.2 ensure the SRVs are inspected and tested. The surveillance requirement 4.6.D.3 requires that the SRV discharge pipe temperatures be monitored daily. This surveillance assures SRV reliability to perform the intended safety functions and remain operable as required by TS 3.6.D.1 and.2 to perform the safety functions.
The Note in TS 3.6.D.2 and TS 3.6.D.3 and.4 were introduced in License Amendment No. 56, dated March 20, 1982. At that time, NRC GL 91-18, "Information to Licensees Regarding NRC Inspection Manual Section on Resolution of Degraded and Nonconforming Conditions," guidance was not available. The TS required engineering evaluation and the corrective actions represent evaluation and resolution of a
Letter 2.05.005 Page 3 of 5 non-conforming condition consistent with the guidance later provided by GL 91-18, Rev.
- 1. Entergy will continue to perform engineering evaluations of non-conforming or degraded conditions of the SRVs. Entergy will revise the SRV surveillance procedure to trigger the performance of an engineering evaluation if the temperature of any SRV discharge pipe exceeds 2120F during reactor operation for a period greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (Attachment 3).
Deleting the requirements for NRC approval of the engineering evaluation does not impact the safety function, reliability, and operability of the SRVs, and does not impact the safety analysis.
- 5.
REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration Entergy Nuclear Operations, Inc. (Entergy) proposes to delete the Pilgrim Station Technical Specifications (TS) 3.6.D.3 and.4, and delete a Note in TS 3.6.D.2.
The proposed change deletes the requirement for NRC approval of an engineering evaluation for continued reactor operation beyond 90 days and corrective actions related to the safety relief valve discharge pipe temperature greater than 21 20F for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Entergy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment 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?
Response: No.
This proposed change deletes an administrative requirement for NRC approval of an engineering evaluation to resolve a non-conforming and degraded condition that is required by NRC Generic Letter 91-18 (GL), Rev.
1, information to Licensees Regarding NRC Inspection Manual Section on Resolution of Degraded and Nonconforming Conditions". The SRVs will be maintained operable, inspected, and tested to perform their safety function as required by the current Specifications and any SRV non-conforming or degraded condition will be addressed in accordance with GL 91-18. The proposed change also deletes a Note regarding installed two-stage Target Rock SRVs. The deletion of an administrative requirement and the Note does not change the plant response to the design basis accident and does not increase the probability of inadvertent SRV operation. Therefore, the proposed change does not significantly increase the probability or consequences of any previously evaluated accidents.
Letter 2.05.005 Page 4 of 5
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The safety function of the SRVs is to provide over-pressure protection of the primary coolant pressure boundary and also for the automatic functions to rapidly depressurize the primary system to a pressure at which low-pressure cooling systems can provide makeup. The proposed change deletes an administrative requirement and a Note related to installed two-stage Target Rock SRVs, and does not introduce any new modes of equipment operation or failure. Therefore, the proposed change does not 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?
Response: No.
The ability of the SRVs to perform their safety function is maintained during operation and will continue to be tested as required in accordance with TS 3/4.13, Inservice Code Testing. The proposed change deletes an administrative requirement that is adequately addressed by following GL 91-18, Rev. 1. Deletion of an administrative requirement does not reduce the margin of safety. Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Accordingly, Entergy 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 Environmental Consideration The proposed change deletes an administrative requirement and a Note applicable to a system/component located within the restricted area, as defined in 10 CFR 20. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, (iii) a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.
Letter 2.05.005 Page 5 of 5
- 6.
COORDINATION WITH OTHER PENDING TS CHANGES The TS page 3/4.6-7 (Attachment 2) is affected by the TS Change request for Single Recirculation Loop Operation submitted by Entergy Letter No. 2.04.074, dated September 2, 2004. Pilgrim will reconcile this TS page when Single Recirculation Loop Operation TS change is approved.
- 7.
REFERENCES
- 1.
NUREG-1 433, "Standard Technical Specifications for General Electric Plants, BWR/4," Section 3.4.3.
ATTACHMENT 2 MARKED-UP TECHNICAL SPECIFICATION AND BASES PAGES (3 pages)
TS pages:
TS Bases page:
3/4.6-6 and 3/4.6-7 B3/4.6-8
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6 PRIMARY SYSTEM BOUNDARY (Cont)
- c. With no required leakage detection systems Operable, be in Cold Shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
D.
Safety and Relief Valves
- 1. During reactor power operating conditions and prior to reactor startup from a Cold Condition, or whenever reactor coolant pressure is greater than 104 psig and temperature greater than 340F, both safety valves and the safety modes of all relief valves shall be operable. The nominal setpoint for the relief/safety valves shall be selected between 1095 and 1115 psig.
All relief/safety valves shall be set at this nominal setpoint +
11 psi.
The safety valves shall be set at 1240 psig +/- 13 psi.
4.6 PRIMARY SYSTEM BOUNDARY (Cont)
D.
Safety and Relief Valves
- 1. Testing of safety and relief/safety valves shall be in accordance with 3.13.
- 2. At least one of the relief/safety valves shall be disassembled and inspected each refueling outage.
- 3. Whenever the safety relief valves are required to be operable, the discharge pipe temperature of each safety relief valve shall be logged daily.
- 4. Instrumentation shall be calibrated and checked as indicated in Table 4.2.F.
l
, 2. If Specification 3.6.D.1 is not met, an orderly shutdown shall be initiated and the reactor coolant pressure shall be below 104 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
C-Note: Technical Specifi 3.6.D.2
-only when tTarget Rock SRVs
- 3. f the temperature of any afety relief discharge pi
- /exceeds 212'F during no 1
//reactor power opeajt fr a
//period of greaterda 24
//hoursa enierg evaluaric all be performed justi ng continued operation the corresponding
\\temperature increases.
Amendment No. 42X-9 6,-88 T-1337-139T-149
- r.
3/4.6-6
UMMNG CONDIONS FOR OPERATON SURVEJLLAN6 RgaUIREMENTS.
3.6 PRIMARY SYSTEM BOUNDARY (Cont) 4.6 PRIMARY SYSTEM BOUNDARY (Cont)
- D.
Safetv Rerief Valves (Con't)
- 4. Any safety relief valve whose a
pipe temperature exceeds 2120F for hours or more shall be removed a e
next cold shutdown of 72 hou~
oe
/
1
,)
tested in the as~bund cnc~, and 7
A recaibrated as necasss nor to
- reinstallation. Fower,
eiatin shall not continue beond -days frm the initial discovery of dTarge pipe
-temperaturlin excess of 212F for more th 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without prior NRC appr al of the engineering evaluahon d
eared in3.6.0.3. fC<
- 5. The limiting conditions of operation for I
co4 o74 the instrumentation that monitors tailx§gg>
pipe temperature are given in Table 3.2-
/
-S;Ag~c Xp O
p E
Jet Pumos E Jet Pumos
- 1. Wheneverthe reactoris in the/tartup or NOTES 21eldsallepumps shall beNTS a.Not required to be performed until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> erale. If it is det ed t
after the associated recirculation loop is in pu is inoerabl n oce shutwn peration.
nal bnlae d
e fcosMiV a6 S
Co
- 2. Not required to be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after >25% Rated Thermal Power.
Ii W
enevertere heei iigg I
2,J.
,n l
ffireactor in the startup or run modes, jet pump T* 57T 1.h n
operability shal e checked daily by vernying l
{
IZ o'
S tat no two affhe following conditions occur/
l_Simultane ly./
- 1. T two recirculatic ops have a flow
<A,5
,imoalance of 10 r more when the pumps are ooe ed at the same sueed.
l
- 2. The indica d value of core flow rate varies frTe the value derived frot o
flow msasurements by more Q0%.
\\
/
- 3. The diffuser to Icwer plm aifferentia!
pressure reading on individual jet pump varies from estabr.ed jet purp delta F characteristics by more than 10%./
Amendment No. 4, EC, 6-,
^, t^-
7-
BASES:
3/4.6 PRIMARY SYSTEM BOUNDARY (Cont)
D.
Safetv and Relief Valves (Cont)
A main steam line isolation with flux scram has been selected to be used as the safety valve sizing transient since this transient results in the highest peak vessel pressure of any transient when analyzed with an indirect scram.
The original FSAR analysis concluded that the peak pressure transient with indirect scram would be caused by a loss of condenser vacuum (turbine trip with failure of the bypass valves to open).
However, later observations have shown that the long lengths of steam lines to the turbine buffer the faster stop valve closure isolation and thereby reduce the peak pressure caused by this transient to a value below that produced by a main steam line isolation with flux scram.
Item 3 above indicates that no credit be taken for the primary scram signal generated by closure of the main steam isolation valves.
Two other scram initiation signals would be generated, one due to high neutron flux and one due to high reactor pressure. Thus item 3 will be satisfied by assuming a scram due to high neutron flux.
Relieving capacity of 40% (4 relief/safety valves) results in a peak pressure during the transient conditions used in the safety valve sizing analysis which is well below the pressure safety limit.
The relief/safety valve settings satisfy the Code requirements that the lowest safety valve set point be at or below the vessel design pressure range to prevent unnecessary cycling caused by minor transients. The results of postulated transients where inherent relief/safety valve actuation is required are given in Appendices R and Q of the Final Safety Analysis Report.
Experience in safety valve operation shows that a testing of at least 50% of the safety valves per refueling outage is adequate to detect failures or deterioration.
The tolerance value of +/- 1% is in accordance with Section III of the ASME Boiler and Pressure Vessel Code.
An analysis has been performed which shows that with all safety valves set 1% higher, the reactor coolant pressure safety limit of 1375 psig is not exceeded.
The relief/safety valves have two functions; i.e., power relief or self-actuated by high pressure.
Power relief is a solenoid actuated function (Automatic Pressure Relief) in which external instrumentation signals of coincident high drywell pressure and low-low water level initiate the valves to open.
This function is discussed in Specification 3.5.D.
In addition, the valves can be operated manually.
Pilgrim's experience with 2 stage safety/relief valves has demonstrated that minimum leakage exists when the tailpipe temperature is 215 Fahrenheit.
Therefore, a reporting requirement Erb oe FTFV 2
f is conservative restimely reporting before leakage reaches significant Revision 1,,
B3/4.6-8 AmenmentNo.
-,-56T-3
ATTACHMENT 3 Summary of Commitments (1 page)
Commitment ID Description Due Date
- 1.
Revise SRV surveillance procedure to include the Prior to the performance of an engineering evaluation if the implementation of temperature of any SRV discharge pipe exceeds this license 2121F during normal operation for a period greater amendment.
than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.