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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217N8911999-10-15015 October 1999 Forwards Rept of Changes,Tests & Experiments at Pilgrim Nuclear Power Station for Period of 970422-990621,IAW 10CFR50.59(b).List of Changes Effecting Fsar,Encl ML20217D3951999-10-13013 October 1999 Forwards Request for Addl Info Re Util 990806 Submittal on USI A-46, Implementation Methodology Used at Pilgrim Nuclear Power Station, Per GL 87-02 ML20217E1581999-10-0808 October 1999 Forwards Insp Rept 50-293/99-05 on 990726-0905.Three Violations Noted & Being Treated as Ncvs.Violations Include Failure to Assure That Design Bases Correctly Translated Into Specifications ML20217C3151999-10-0606 October 1999 Forwards Scenario Package for Pilgrim Nuclear Power Station Nrc/Fema Evaluated Exercise Scheduled for 991207.Without Encl ML20217D5591999-10-0505 October 1999 Documents Pilgrim Nuclear Power Station Five Yr Survey of Main Breakwater.Survey Has Determined That Pilgrim Main Breakwater Is Intact & Remains Adequately Constructed to Perform Designed Safety Function ML20217C8051999-10-0505 October 1999 Forwards Proprietary Results of Audiologic Evaluations for Jp Giar,License SOP-10061-3.Attachment Clearly Shows Requirements for Operator Hearing Ability Are Met. Proprietary Info Withheld,Per 10CFR2.790(a)(6) ML20212J8301999-09-30030 September 1999 Informs of Completion of mid-cycle PPR of Pilgrim Nuclear Power Station.Staff Conducts Reviews for All Operating NPPs to Integrate Performance Info & to Plan Insp Activities at Facility Over Next Six Months ML20216J9961999-09-29029 September 1999 Forwards Resume of Person Identified as Acting RPM in Licensee to NRC Re Notification That Person Named in License Condition 11 of 20-07626-02,is No Longer Employed at Pilgrim Station.Resume Withheld,Per 10CFR2.790 ML20212F7871999-09-24024 September 1999 Advises That Util 990121 Application for Amend Being Treated as Withdrawn.Proposed Changes Would Have Modified Facility UFSAR Pertaining to Values for post-accident Containment Pressure Credited in Pilgrim Net Positive Head Analyses ML20212H1381999-09-23023 September 1999 Submits Info in Support of Request Filed on 990730 to Grant one-time Exemption from 10CFR50,App E,Authorizing Biennial Full Participation Emergency Preparedness Exercise to Be Conducted in 2002 Instead of 2001 ML20212H1441999-09-23023 September 1999 Withdraws 990121 Request for License Change Re Emergency Core Cooling Sys Net Positive Suction Head,Due to Incorrect Datum Preparation ML20212C2861999-09-16016 September 1999 Forwards SER Accepting Licensee 981123 Request for Relief RR-E1,RR-E5,RR-E6 Pursuant to 10CFR50.55a(a)(3)(i) & Request for Relief RR-E2,RR-E3 & RR-E4 Pursuant to 10CFR50.55a(a)(3)(ii) ML20216F3451999-09-16016 September 1999 Forwards Summary Rept Providing Results of ISI Conducted at PNPS on-line & Refueling Outage (RFO 12) ML20216E7111999-09-0909 September 1999 Forwards License Renewal Application Including Form NRC-398 & Form NRC-396 for Jp Giar,License SOP-10061-3.Without Encls ML20216E5891999-09-0707 September 1999 Forwards Copy of Pilgrim Station Organization Structure. Encl Refelcts Changes in Upper Mgt Level Structure.Changes Were Effective 990901 ML20211M4501999-09-0303 September 1999 Informs That Pilgrim Nuclear Power Station Plans to Conduct Full Participation Emergency Preparedness Exercise with Commonwealth of Ma on 991207,IAW 10CFR50,App E,Section IV.F.2 ML20211M9161999-08-31031 August 1999 Submits Review & Correction of Info in Reactor Vessel Integrity Database (Rvid),Version 2,re Pilgrim Station ML20211J8391999-08-30030 August 1999 Forwards Rev 1 to Provisional Decommissioning Trust Agreement for Plant,Changing Portions of Agreement to Permit Up to Two Distributions & Clarify Formula for Distribution ML20211H5701999-08-27027 August 1999 Forwards Insp Rept 50-293/99-04 on 990610-0725.Two Violations Identified Being Treated as non-cited Violations ML20211C3381999-08-19019 August 1999 Provides semi-annual LTP Update,Including Schedule, Commitment Descriptions,Progress Since Last Update & Summary of Changes.Rev Bars Indicate Changes in Status Since Last Submittal ML20210U7521999-08-18018 August 1999 Forwards from Massachusetts State Senator T Murray Opossing Merger Between Bec Energy & Commonwealth Energy Systems ML20210U6691999-08-18018 August 1999 Forwards from Massachusetts State Senator T Murray Opposing Merger Between Bec Energy & Commonwealth Energy Systems ML20210U5761999-08-18018 August 1999 Responds to Opposing Merger of Bec Energy & Commonwealth Energy Sys in Commonwealth of Massachusetts. Informs That for Sale,Nrc Responsible for Only Ensuring That Entergy Technically & Financially Qualified to Operate NPP ML20210U5151999-08-17017 August 1999 Forwards Notice of Withdrawal of Application for Approval of Indirect Transfer of FOL for Pilgrim in Response to .Approval No Longer Needed Since Beco Sold Interest in Pilgrim to EOI on 990713 ML20211B3841999-08-16016 August 1999 Forwards Response to NRC Second RAI Re Pressure Locking & Thermal Binding of SR power-operated Gate Valves ML20210U4831999-08-13013 August 1999 Forwards fitness-for-duty Program Performance Data Sheets for Period of 990101-0630,per 10CFR26.71(d) ML20210S0891999-08-0909 August 1999 Forwards Amend 11 to Indemnity Agreement B-48 Signed by Boston Edison Co & Entergy Nuclear Generation Co ML20210R6251999-08-0606 August 1999 Provides Supplementary Info on USI A-46 Implementation Methodology at Pilgrim Station,To Enable NRC to Perform Evaluation & Issuance of Plant Specific SER for Plant ML20210M9411999-08-0202 August 1999 Requests That NRC Treat Pending Actions Requested by Beco Prior to 990713,as Requests Made by Entergy.Ltr Requests That Minor Administrative Changes to License Amend 182 & Associated Ser, ,reflect 990713 Transfer ML20210H8761999-07-30030 July 1999 Requests That NRC Grant Exemption from Requirements of 10CFR50,App E,Section IV F,Which Would Authorize Rescheduling of 2001 Biennial Full Participation Emergency Preparedness Exercise for Pilgrim Station to 2002 ML20210H8661999-07-29029 July 1999 Provides Revised Response to GL 96-06 & Addresses NRC Insp Concern for Containment Penetration X-12.Info Submitted to Facilitate NRC Review & Closeout of Subject GL for Plant ML20216E2321999-07-26026 July 1999 Discusses GL 92-01,rev 1,suppl 1, Rv Structural Integrity. NRC Revised Info in Rvid & Releasing as Rvid Version 2 ML20216D4131999-07-22022 July 1999 Informs That J Conlon,License OP-11040-1,terminated Employment with Beco on 990703,per 10CFR50.74.Individual Will Not Participate in Util Licensed Operator Requalification Training Program ML20210E2231999-07-20020 July 1999 Discusses Arrangements Made by Dennis & M Santiago During 990615 Telephone Conversation for NRC to Inspect Licensed Operator Requalification Program at Pilgrim During Wk of 991004 ML20210C4151999-07-19019 July 1999 Informs That Util Intends to Submit Approx Eight Licensing Actions in FY00 & Eight in FY01,in Response to Administrative Ltr 99-02.Actions Are Not Expected to Generate Complex Reviews ML20210A9441999-07-14014 July 1999 Responds to Re Changes to Pilgrim Nuclear Power Station Physical Security Plan Identified as Issue 2,rev 14, Addendum 1,respectively.No NRC Approval Is Required IAW 10CFR54(p) ML20210F3711999-07-14014 July 1999 Informs NRC That Effective 990713,listed Pilgrim Station Security Plans Have Been Transferred from Boston Edison to Entergy & Are Still in Effect ML20209G2251999-07-0909 July 1999 Forwards Insp Rept 50-293/99-03 on 990419-0609.Five Severity Level IV Violations of NRC Requirements Identified & Being Treated as non-cited Violations,Consistent with App C. Several Individual Tagging Errors Occurred ML20209C4661999-07-0707 July 1999 Forwards SE Accepting Addendum on Proposed Change in Corporate Ownership Structure Involving Entergy Nuclear Generation Co ML20209C3851999-07-0606 July 1999 Forwards Redacted Draft of Decommissioning Trust Agreement Re Transfer of PNPS & NRC Operating License & Matls License from Boston Edison Co to Entergy Nuclear Generating Co ML20209C7761999-07-0606 July 1999 Submits Annual Summary Rept of Changes Made to QAP Description as Described in QA Manual,Vol Ii.Rept Covers Period of Jul 1998 Through June 1999.No Changes Made During Period ML20196J7251999-07-0101 July 1999 Informs of Completion of Licensing Action for GL 96-01, Testing of Safety-Related Logic Circuits, for Pilgrim Nuclear Power Station ML20209B9411999-06-30030 June 1999 Discusses Deferral of IGSCC Welds to RFO 13.Deferral of Welds to Refueling Outage 13 Does Not Impact Acceptable Level of Quality & Safety Per 10CFR50.55(a)(3)(i) Since Plant in Compliance W/Exam Percentage Requirements ML20209B9431999-06-30030 June 1999 Provides Formal Notification That Closing Date for Sale & Transfer of Pilgrim Station Scheduled to Occur on 990713. a Wang Will Be Verbally Notified of Time of Sale Closing ML20196H2381999-06-29029 June 1999 Forwards SER Denying Licensee 980820 Request for Alternative Under PRR-13,rev 2 for Use of Code Case N-522 During Pressure Testing of Containment Penetration Piping ML20209B9791999-06-29029 June 1999 Forwards Rev 13A to Pilgrims COLR for Cycle 13,IAW TS 5.6.5 Requirements.Rev 13A Provides cycle-specific Limits for Operating Pilgrim During Remainder of Cycle 13 ML20209A8761999-06-28028 June 1999 Forwards SER Authorizing Licensee 990317 Relief Request to Use ASME Code Case N-573 as Alternative to ASME Code Section XI Article IWA-4000 for Remainder of 10-year Interval Pursuant to 10CFR50.55a(a)(3)(i) ML20209A8701999-06-25025 June 1999 Responds to NRC Request for Info Re Y2K Readiness of Computer Sys at Nuclear Power Plants. Y2K Readiness Disclosure for Plant,Reporting Status of Facility Y2K Readiness Encl ML20210U5901999-06-25025 June 1999 Opposes Merger of Bec Energy & Commonwealth Energy Sys in Commonwealth of Massachusetts.Expresses Skepticism Re Claim by Companies That Consumers Will Benefit from Proposed Consolidation & four-year Freeze in Base Rates ML20209C3431999-06-22022 June 1999 Forwards Addendum 1,Rev 14 to Pilgrim Station Security Plan,Iaw 10CFR50.54(p)(2).Changes Proposed Have Been Implemented & Constitute Increase in Plant Defense Plan Commitments.Encl Withheld,Per 10CFR73.21 1999-09-09
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20217N8911999-10-15015 October 1999 Forwards Rept of Changes,Tests & Experiments at Pilgrim Nuclear Power Station for Period of 970422-990621,IAW 10CFR50.59(b).List of Changes Effecting Fsar,Encl ML20217C3151999-10-0606 October 1999 Forwards Scenario Package for Pilgrim Nuclear Power Station Nrc/Fema Evaluated Exercise Scheduled for 991207.Without Encl ML20217C8051999-10-0505 October 1999 Forwards Proprietary Results of Audiologic Evaluations for Jp Giar,License SOP-10061-3.Attachment Clearly Shows Requirements for Operator Hearing Ability Are Met. Proprietary Info Withheld,Per 10CFR2.790(a)(6) ML20217D5591999-10-0505 October 1999 Documents Pilgrim Nuclear Power Station Five Yr Survey of Main Breakwater.Survey Has Determined That Pilgrim Main Breakwater Is Intact & Remains Adequately Constructed to Perform Designed Safety Function ML20216J9961999-09-29029 September 1999 Forwards Resume of Person Identified as Acting RPM in Licensee to NRC Re Notification That Person Named in License Condition 11 of 20-07626-02,is No Longer Employed at Pilgrim Station.Resume Withheld,Per 10CFR2.790 ML20212H1441999-09-23023 September 1999 Withdraws 990121 Request for License Change Re Emergency Core Cooling Sys Net Positive Suction Head,Due to Incorrect Datum Preparation ML20212H1381999-09-23023 September 1999 Submits Info in Support of Request Filed on 990730 to Grant one-time Exemption from 10CFR50,App E,Authorizing Biennial Full Participation Emergency Preparedness Exercise to Be Conducted in 2002 Instead of 2001 ML20216F3451999-09-16016 September 1999 Forwards Summary Rept Providing Results of ISI Conducted at PNPS on-line & Refueling Outage (RFO 12) ML20216E7111999-09-0909 September 1999 Forwards License Renewal Application Including Form NRC-398 & Form NRC-396 for Jp Giar,License SOP-10061-3.Without Encls ML20216E5891999-09-0707 September 1999 Forwards Copy of Pilgrim Station Organization Structure. Encl Refelcts Changes in Upper Mgt Level Structure.Changes Were Effective 990901 ML20211M4501999-09-0303 September 1999 Informs That Pilgrim Nuclear Power Station Plans to Conduct Full Participation Emergency Preparedness Exercise with Commonwealth of Ma on 991207,IAW 10CFR50,App E,Section IV.F.2 ML20211M9161999-08-31031 August 1999 Submits Review & Correction of Info in Reactor Vessel Integrity Database (Rvid),Version 2,re Pilgrim Station ML20211J8391999-08-30030 August 1999 Forwards Rev 1 to Provisional Decommissioning Trust Agreement for Plant,Changing Portions of Agreement to Permit Up to Two Distributions & Clarify Formula for Distribution ML20211C3381999-08-19019 August 1999 Provides semi-annual LTP Update,Including Schedule, Commitment Descriptions,Progress Since Last Update & Summary of Changes.Rev Bars Indicate Changes in Status Since Last Submittal ML20211B3841999-08-16016 August 1999 Forwards Response to NRC Second RAI Re Pressure Locking & Thermal Binding of SR power-operated Gate Valves ML20210U4831999-08-13013 August 1999 Forwards fitness-for-duty Program Performance Data Sheets for Period of 990101-0630,per 10CFR26.71(d) ML20210S0891999-08-0909 August 1999 Forwards Amend 11 to Indemnity Agreement B-48 Signed by Boston Edison Co & Entergy Nuclear Generation Co ML20210R6251999-08-0606 August 1999 Provides Supplementary Info on USI A-46 Implementation Methodology at Pilgrim Station,To Enable NRC to Perform Evaluation & Issuance of Plant Specific SER for Plant ML20210M9411999-08-0202 August 1999 Requests That NRC Treat Pending Actions Requested by Beco Prior to 990713,as Requests Made by Entergy.Ltr Requests That Minor Administrative Changes to License Amend 182 & Associated Ser, ,reflect 990713 Transfer ML20210H8761999-07-30030 July 1999 Requests That NRC Grant Exemption from Requirements of 10CFR50,App E,Section IV F,Which Would Authorize Rescheduling of 2001 Biennial Full Participation Emergency Preparedness Exercise for Pilgrim Station to 2002 ML20210H8661999-07-29029 July 1999 Provides Revised Response to GL 96-06 & Addresses NRC Insp Concern for Containment Penetration X-12.Info Submitted to Facilitate NRC Review & Closeout of Subject GL for Plant ML20216D4131999-07-22022 July 1999 Informs That J Conlon,License OP-11040-1,terminated Employment with Beco on 990703,per 10CFR50.74.Individual Will Not Participate in Util Licensed Operator Requalification Training Program ML20210C4151999-07-19019 July 1999 Informs That Util Intends to Submit Approx Eight Licensing Actions in FY00 & Eight in FY01,in Response to Administrative Ltr 99-02.Actions Are Not Expected to Generate Complex Reviews ML20210F3711999-07-14014 July 1999 Informs NRC That Effective 990713,listed Pilgrim Station Security Plans Have Been Transferred from Boston Edison to Entergy & Are Still in Effect ML20209C7761999-07-0606 July 1999 Submits Annual Summary Rept of Changes Made to QAP Description as Described in QA Manual,Vol Ii.Rept Covers Period of Jul 1998 Through June 1999.No Changes Made During Period ML20209C3851999-07-0606 July 1999 Forwards Redacted Draft of Decommissioning Trust Agreement Re Transfer of PNPS & NRC Operating License & Matls License from Boston Edison Co to Entergy Nuclear Generating Co ML20209B9431999-06-30030 June 1999 Provides Formal Notification That Closing Date for Sale & Transfer of Pilgrim Station Scheduled to Occur on 990713. a Wang Will Be Verbally Notified of Time of Sale Closing ML20209B9411999-06-30030 June 1999 Discusses Deferral of IGSCC Welds to RFO 13.Deferral of Welds to Refueling Outage 13 Does Not Impact Acceptable Level of Quality & Safety Per 10CFR50.55(a)(3)(i) Since Plant in Compliance W/Exam Percentage Requirements ML20209B9791999-06-29029 June 1999 Forwards Rev 13A to Pilgrims COLR for Cycle 13,IAW TS 5.6.5 Requirements.Rev 13A Provides cycle-specific Limits for Operating Pilgrim During Remainder of Cycle 13 ML20209A8701999-06-25025 June 1999 Responds to NRC Request for Info Re Y2K Readiness of Computer Sys at Nuclear Power Plants. Y2K Readiness Disclosure for Plant,Reporting Status of Facility Y2K Readiness Encl ML20210U5901999-06-25025 June 1999 Opposes Merger of Bec Energy & Commonwealth Energy Sys in Commonwealth of Massachusetts.Expresses Skepticism Re Claim by Companies That Consumers Will Benefit from Proposed Consolidation & four-year Freeze in Base Rates ML20209C3431999-06-22022 June 1999 Forwards Addendum 1,Rev 14 to Pilgrim Station Security Plan,Iaw 10CFR50.54(p)(2).Changes Proposed Have Been Implemented & Constitute Increase in Plant Defense Plan Commitments.Encl Withheld,Per 10CFR73.21 ML20195G3721999-06-0707 June 1999 Informs That Proposed Indicators Failed QA Assessments for Digital Verification,Validation & Control of Software. Proposed Mod Can Be Completed on-line ML20195B5021999-05-27027 May 1999 Provides Suppl Info to 990203 Request of Beco That NRC Consent to Indirect Transfer of Control of Util Interest in License DPR-35.Request Described Proposed Merger of Bec Energy with Commonwealth Energy Sys ML20207D4681999-05-24024 May 1999 Provides Addl Info to That Included in Beco Ltr 98-123 Dtd 981001,addressing NRC Concerns Described in GL 96-06, Concerning Waterhammer in Reactor Bldg Closed Cooling Water Sys ML20195B9051999-05-20020 May 1999 Forwards Completed Renewal Applications for Listed Operators.Without Encls ML20206J4901999-05-0606 May 1999 Forwards Completed License Renewal Application,Including Forms NRC-398 & 396 for Sc Power,License OP-6328-3 ML20206P0711999-05-0606 May 1999 Forwards NRC Form 396, Certification of Medical Exam by Facility Licensee, for K Walz,License SOP-10886-1.Encl Withheld IAW 10CFR2.790(a)(6) ML20206D3621999-04-27027 April 1999 Informs NRC That Final Five Sys self-assessments Required to Fulfill Commitment Made in 980828 Response to Insp Rept 50-293/98-04 Were Completed on 990422.Completion Was Delayed by High Priority Refueling Outage 12 Preparatory Work ML20205R9871999-04-21021 April 1999 Forwards Affidavit of JW Yelverton of Entergy Nuclear Generation Co Supporting Request for Withholding Info from Rept on Audit of Financial Statements for Year Ended 971231. Pages 16 & 18 of Subj Rept Also Encl ML20207B0891999-04-20020 April 1999 Forwards e-mail Message from Constituent,J Riell Re Y2K Compliance of Nuclear Power Plant in Plymouth,Massachusetts. Copy of Article Entitled Nuke Plants May Not Be Y2K Ready Also Encl ML20206A2741999-04-16016 April 1999 Dockets Encl Ltr Which Was Sent to AL Vietti-Cook Re Condition of Approval of Transfer of License & License Condition for DPR-35.Encl Resolves Issues Between Attorney General of Commonwealth of Massachusetts & Applicants ML20205P9131999-04-16016 April 1999 Submits Applicant Consent to Listed Condition of Approval of Transfer of License & License Condition for License DPR-35 & Affirmatively Request That NRC Adopt Listed Language in Order ML20205P9271999-04-16016 April 1999 Withdraws Motion for Leave to Intervene & Petition for Summary Or,In Alternative,For Hearing.Requests That NRC Adopt Condition of Approval of Transfer of License & License Condition Agreed to Beco & Entergy Nuclear Generation Co ML20205Q9231999-04-15015 April 1999 Forwards Proprietary & non-proprietary Addl Info in Support of Request to Transfer of Plant FOL & Matls License to Entergy Nuclear Generation Co.Proprietary Info Withheld,Per 10CFR2.790 ML20205P9631999-04-15015 April 1999 Provides Attachments a & B in Support of Request for Transfer of Plant Operating License & NRC Matl License from Beco to Entergy Nuclear Generation Co as Submitted in Ref 1. Info Provided in Response to Request at 990413 Meeting ML20205H9281999-04-0707 April 1999 Requests Withdrawal of Uwua Locals 369 & 387 Unions Joint Intervention in Listed Matter ML20205F3731999-04-0202 April 1999 Submits Addl Info Provided in Support of Request for Transfer of Pilgrim Nuclear Power Station Operating License & Matls License.State of Ma Order Authorizing Divestiture & Copy of Financial Arrangement Encl ML20204H3771999-03-26026 March 1999 Informs That Local 387,Utility Workers Union of America,AFL- Cio Voted to Approve New Contract with Entergy Nuclear Generation Co & Voted to Accept Boston Edison Divestiture Agreement ML20205D4231999-03-24024 March 1999 Forwards Decommissioning Funding Rept for Pilgrim Nuclear Power Station,In Accordance with 10CFR50.75(f)(1) 1999-09-09
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o so.amsa a i Pilgrim Nuclear Power Station g , Rody Hill Road Plymouth, Massachusetts 02360 l LJ. Olivier Vice Presided Nuclear and Citation Director January 30 ,1998 BECo Ltr. 2.98.012 l U.S. Nuclear Regulatory Commission l Attention: Document Control Desk Washington, DC 20555 l
Docket No. 50-293 License No. DPR-35 Request for NRC approval of an Engineering Evaluation:
Elevated Tailpipe Temperature on Safety Relief Valve 203 38 in accordance with Pilgrim Nuclear Power Station (PNPS) Technical Specification 3.6.D.4, Boston Edison requests NRC review and approval of the attached engineering evaluation of elevated tailpipe temperatures associated with safety-relief valve (SRV) 203-3B. This evaluation 1 was reviewed by the Operational Review Committee on January 15,1998.
The PNPS Technical Specification 3.6.D.4 states:
Any safety relief valve whose discharge pipe temperature exceeds 212*F for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or more shall be removed at the next shutdown of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more, tested in the as found condition and recalibrated as necessary prior to reinstallation. Power operation shall not continue beyond 90 days from the initial discovery of discharge pipe temperatures in excess of 212*F for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without prior NRC approval of the engineering evaluation delineated in 3.6.D.3.
Technical Specification 3.6.D.3 states, in part, that an engineering eva uation shall be performed justifying continued operation for the corresponding temperature increases.
An elevated tailpipe temperature associated with safety / relief valve 203-3B (i.e.,132*F) was first observed on December 11,1997. The temperature slowly increased, and on December 24, 1997, the temperature reached and remained in excess of 212 F for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The tailpipe temperature is currently approximately 217'F where it has remained relatively stable.
SRV 203-3B is operable in its present condition. The leakage is minor in nature and it has tentatively been attributed to pilot staDe leakage.
if SRV203-3B tailpipe temperature exceeds 235'F 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 /> or exceeds 2ETF at any time, then an orderly shutdown of the reactor shall commence, as recommended in the attached engineering evaluation.
l.lil.l l.ll.lllll.ll.ll.l i bCC\ ,
.~vv"~ Page 1 of 2 9002120229 900130 PDR ADOCK 05000293 P PDR
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Pilgrim Nucle:r Power Station Your review cnd approv;l is requested pri:r 13 M:rch 24,1998 t3 preclud3 a shutdown cf Pilgrim Station. If addition:lInformation is required, pl2cs3 contact Mr. B:b C nnon tt (508) 830-8321.
i ,
hev L. J. Olivier RLC/deg id: reqeat/radmisc
Attachment:
Engineering Evaluation i
cc: Mr. Alan B. Wang, Project Manager Project Directorate 1-3 Mall Stop: OWFN: 14B2 Office of Nuclear Reactor Regulatic7 1 White Flint North 11555 Rockville Pike Rockville, MD 20852 U.S. Nuclear Regulatory Commission Region 1-475 Allendale Road Kiag of Prussia, PA 19406 Senior Resident inspector Pilgrim Nuclear Power Station Page 2 of 2
. EE98-004
- s.
BOSTON EDISON COMPANY ENGINEERING EVALUATION L Initiating Documenti PR97.3565
- 2. Affected (System. Subsystem. Train. ComDonent. or Device)
Target Rock Coriv .on Two-Stage Safety Relief Valve 203-3B
- 3. Specified Functions of the AtTected Item The safety relief valves are part of the reactor coolant pressure boundary and operate by power actuation (i.e., automatic depressurization system) or self-actuation by high process pressure. The safety relief valves limit peak vessel pressure during overpressure transients to satisfy ASME code requirements. The postulated transients for which safety / relief valve actuation is required are given in Chapter 14 and in Appendices R and Q of the FSAR.' The automatic depresrurization system 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 LOCA, this function would be required if high pressure ECCS was unable to maintain vessel water hvel.~
- 4. Enferences
- 1. Technical Specifications and associated bases 3.6.D.1,3.6.D.2,3.6.D.3,3.6.D.4 and 3.6.D.5.
- 2. General Electric Report NSE 13-0282, " Pilgrim Plant, SRV Tailpipe Steam Temperature Correlation for SRV Lee.kage Monitoring System," dated February 1982.
- 3. General Electric Report NEDE-30476, "Setpoint Drift Investigation of Target Rock Two-Stage Safety / Relief Valve (Final Report)," dated February 1984.
- 4. Operability Evaluation for Target Rock Corporation Two-Stage Safety Relief Valve 203-3D, dated 8/17/91.
- 5. NRC approval of operability evaluation for 203-3D, incoming NRC letter 1.91.288 dated 10/24/91.
- 6. TCH92-133," Root Cause/ Corrective Action Response for 203-3D" (PR92.0338/F&MR 91-373).
Page1
. EE98-004
- 7. Operability Evaluation for Target Rock Corporation Two-Stage Safety Relief
, , Valve 203-3 A dated 11/2/93.
- 8. Wyle Lab, Test Report No. 41211-0 dated,4/25/91,
- 9. Supplemental Reload I.icensing Report for PNPS Reload 11, Cycle 11, PDC 96-g 17," Reload 11/ Cycle 12 Core Design"
- 10. MR #19703052
- 11. Operability Evaluation for Target Rock Corporation Two-Stage Safety Relief Valve 203-3B, pilot serial n.imber 1025, dated 2/9/96.
E 12. Operability Evaluation for Target Rock Corporation Two Stage Safety Relief Valve 203-3D pilot serial number 1054, dated 11/4/97.
- 13. NEDO-22159 General Electric Boiling Water heactor Increased SRV S;mmer Margin Analysis for PNPS Unitl-June 1982.
- 5. Operability Concerl
{
Safety Relief Valve (SRV) 203-3B is leaking. This condition was detected by tailpip-temperature monitoring instrumentation on 12/11/97 and documented in Problem Report 97-3565. The SRV 203-3B tailpipe temperature trended to 132 'F on 12/11/97,145 *F on 12/16/97,212 'F on 2/24/97 and stabilized at 217 *F on 1/2D8 This temperatuie profile supports a condition indicative of pilot stage leakage.
Technical Specification 3.6.D.3 requires an engineering evaluation to support continued operation if the temperature of any safety relief valve discharge pipe exceeds 212'F 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 /> during normal reactor power operation (Ref.1). The Technical Specification Bases states that minimal les.kage exists when tht tailpipe temperature is 215 F, and therefore, a conservative temperature of 212 F was chosen.
- 6. Onera'oility Recommendation (Check one)
@ Operable -
O Inoperable Page 2
. EE98-004 7,. Basis for Recommendadon (Use additional sheets as necessary)
The most likely leakage paths through the Target Rock Corporation (TRC) two-stage safety relief valve are: (1) through the main stage, past the main disc and seat interface, or (2) through the pilot stage, past the disc and seat interface. General Electric (GE) and TRC (the valve's manufacturer) representatives have in the past indicated main stage leakage is typically substantial and increases faster than pilot stage leakage and pilot stage leakage is mon. on than main disc leakage.
SRV 203 3B pilot 1025 had high tailpipe temperatures between 1/24/96 and 2/6/96.
The operability evaluation determined the elevated temperature which eventually settled out at 217 'F was the result of minor pilot leakage. This was later confirmed at Wyle Labs where diagnostic "r .found" setpoint and leakage tests were performed.
SRV 203-3D pilot 1054 and main stage 008 were replaced on 12/3/97 due to high tailpipe temperature. Although diagnostic testing is incomplete, a leaking pilot valve is suspected.
Due to the similarity of the increase in tailpipe temperatures to SRV-3B pilot 1025 leakage, the most probable cause for the leakage presently experienced by SRV-3B is pilot leakage. This condition may clear with a lowering of reactor pressure; however, it will more likely remain at some low level (218-220 F ), for a period of months.
The consequences ofleakage across either the pilot or main stage boundary for SRV 203-3B must be addressed, since leakage increases may occur later and may occur at either location. Pilot stage leakage affects valve lift set point and response time while main stage leakage loes not.
Pilot Stage Leakage Pilot stage leakage can affect the performance of the two stage Target Rock SRV in the pressure actuated mode (i.e., safety mode). The effects of leakage on valve performance have been extensively studied and basically consist of the following: (1) setpoint drift, (2) response time changes (Ref. 3).
The leakage rates studied by GE and TRC range from 200 lbs/hr to 1000 lbs/hr. Test results indicate that set point prespire increased to approximately 1% at 225 lbs/hr and to 2% at a leakage rate of approximately 400 lbs/hr. The setpoint then decreased 2% per 100 lbs/hr of additio: alleakage. The eiTect ofleakage rate on setpoint is illustrated in Reference 3. Based on TRC test results, pilot stage leakage up to 1000 lbs/hr did not significantly affect the SRV setpoint (Ref. 3).
Response time is the interval from pilot actuation to main disc lift. The normal response time for a two stage TRC SRV is approximately 0.4 seconds. Response time varies with leakage rate A slower response time results in a higher peak reactor vessel pressure t . iring the safety mode, and a faster response time results in a lower peak reactor pressure. A slower response time also Page 3
a EE98-004 '
, results from a higher tailpipe temperature (i.e., higher steam leakage). The impact ofleakage on
~
- response time is presented in the plant analysis section of this report.
Main Stage Leakage
' Main stage leakage is an uncommon problem in the industry according to TRC. This view is -
substantiated by the volumes ofinformation available on relief valve ler.kage, all of which is a result of pilot stage leakage, Leakage across the main stage boundary is an economic concern -
because of the rotential for seat and/or disc dam:_ge. TRC and GE adviw that leakage across the main disc will nat affect the ability of the SRV to operate in either the pressure actuated or power-actuated modea. . Leakage across the main stage should not cause the SRV to inadverte :tly open and cause a rapid depressurization or fail to reclose aRer operating.
Plant Analyss General Electric has performed sensitivity analyses on PNPS showing that even if driR results in an opening pressure 10% above the nominal setpoint for all Pilgrim's installed SRVs, the peak -
pressure for the MSIV closure-flux scram event is less than the upset limit of 1375 psig. Based . i on these results, GE concludes that BWRs with TRC two-stage SRVs can tolerate driRs significantly above the 1% technical specification setpoint tolerance.
'Also, the peak vessel pNssure would increase by 5 psig if one SRV experienced the leakage L induced maximum response time delay of 0.9 second (Ref. 3). This is much less than the 77 psig margin between reactor vessel pressure for cycle 12 and the upset limit of 1375 psig (Ref. 9).
The impact of either a delay in SRV response time or an increase in SRV opening pressure on critical power thermal margin is minimal. This is due to the rapid insertion oflarge negative-control reactivity during transients before the higher pressure can contribute to any significant _-
Ladditional core power production due to core void collapse. This was demonstrated in NEDO-22159 (Ref.13 ), where a 30 psig increase in SRV opening setpoint resulted in only 0.1% _ j increase in peak fuel rod heat flux following a limiting pressurization event. This was specifically
- evaluated for PNPS for cycle 6f However, it would also apply to cycles 7-12 due to the -
insignificant contribution of SRV pressure relief to the mitigation of the core power excursion -
associated with limiting pressurization events. Reactivity shutdown via reactor scram renders the core essentially subcritical before SRV pressure relief can be effective in moderating the void
- collapse due to the pressurization event.
SRV Leakane Versus Tailoipe Temocrature and SRV Setocint
, L The maximum allowable SRV 203-3B tailpipe temperature of approximately 255'F can be correlated to a ceam leakage flow rate of approximately 225 lbs/hr, while steam leakage of 1000 lbs/hr corresponds to a tail pipe temperature of approximately 275 F. It is acceptable to continue operation with a tailpipe temperature ofless than or equal to 255"F since test data has demonstrated that the possible relief valve setpoint drin at this temperature is equivalent to +1%
(Ref. 3).
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_ , f .; EE98 004 Riant Parameter Effects on Tailpine Temocrature
- Drywell temperature: Sensitivity analysis predicts that the tailpipe temperature is relatively in-sensitive to drywell temperature variations over the entire range of steam leakage (Ref. 2).-
Reactor pressure: The temperature of the steam at the exit of the relief valve decreases as reactor pressure increases. Any effect on downstream tailpipe temperature may be offset by increased leakage s ates at higher rmi tor pressure.l The temperature limit of 255'F was based on normal reactor operating pressure for the exit steam (Ref. 2).
- Containment pressure: The safety relief valve tail pipe is equipped with vacuum breakers that prevent drawing a column of torus water into the tailpipe. The tailpipe will be at atmospheric .
pressure prior to inetiing and slightly above atmospheric pressure after inerting the containment.
The effects of containment pressure on tailpipe temperature are negligible because the differeace in tailpipe pressure due to inerting is only a few psig. Also the maximum leakage flow rate of up to 1000 lbs/hr will na be sufficient to pressurize the tailpipe, thereby affectiag temperature (Ref. 2). Therefore, containment pressure effects are judged to be negligible.
Eauioment ouamation Each safety relief valve has one solenoid valve which is attached to a manifold mounted on the air
. operator for the valve. The leakage flow through the safety relief valve will raise the temperature -
of the main valve body, base, pilot assembly and associated tailpipe. The solenoid ' valve is environmentally qualified, considering in-part the normal ambient temperature to which it is exposed. The solenoid valve is not in direct contact with any part of the safety relief valve which will experience appreciable elevated temperature because of the leakage through the valve..
Therefore, the solenoid valve will not be exposed to any significant amount of conducted heat but could be exposed to a slightly higher ambient temperature.- The solenoid valve is mounted _as an appendage off the safety relief valve in a configuration that maximizes air circulation around it and -
minimizes the ambient temperature to which the solenoid ' valve is exposed. Therefore, the effects of minor leakage through the SRV 203-3B safety relief valve are judged to hr.ve no appreciable affect on the enviranmental qualification of the safety relief valve solenoid.
CQElusi9a SRV 203-3B is operable in its present condition. The leakage that has occurred is of a minor i nature attributed to pilot stage leakage and is acceptable as discussed previously. Either -
_ _ intermittent or continuous leakage within the limits described below is acceptable for continued operation.. Tests and analyses have shown that ieakage rates of approximately 225 lbs/hr (equivalent to 255'F) should not impact the SRV setpoint by more than +1%.
Based on past experience with leeking pilot valv's, a slightly lower action limit has been selected for this SRV in order to assure reliable SRV operation and reduce damage to the pilot seat and Page 5
,. EE98-004 disc. If the tailpipe temperature exceeds 235'F 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 /> or exceeds 250*F at any time, an orderly shutdown of the reactor shall commence.
8.Comnensatory Measures / Conditions Required Justification for Continued Operation Technical specification surveillance 4.6.D.3 requires that SRV tailpipe temperature be logged
] daily. This surveillance shall be performed at an increased frequency of once per hour, to compensate for the reduced margin between the normal maximum tailpipe temperature of 212 F and 235'F.
Temporary Modification 97-66 has been installed to reconfigure the alarm circuitry on safety relief valve temperature recorder TR260-20. The alarm point of RV-203-3B which normally annunciates at 200*F has been reset to annunciate at 230 F.
If SRV 203-3B tailpipe temperature exceeds 235'F 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 /> or exceeds 250'F at any time, then an orderly shutdown of the reactor shall commence.
This relatively small leakage is not expected to cause torus water temperature or drywell temperature to change significantly; therefore, an increased surveillance interval for these parameters is not warranted.
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