ML100770450

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Proposed License Amendment to Technical Specifications: Revised Technical Specification for Setpoint and Setpoint Tolerance Increases for Safety Relief Valves (SRV) and Spring Safety Valves (Ssv), and Related Changes
ML100770450
Person / Time
Site: Pilgrim
Issue date: 03/15/2010
From: Bronson K
Entergy Nuclear Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
2.10.016, TAC M79265
Download: ML100770450 (35)
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Text

S' Entergy Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 Kevin H. Bronson Site Vice President March 15, 2010 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station Docket No. 50-293 License No. DPR-35 Proposed License Amendment to Technical Specifications: Revised Technical Specification for Setpoint and Setpoint Tolerance Increases for Safety Relief Valves (SRV) and Spring Safety Valves (SSV), and Related Changes

REFERENCES:

1. NEDC-31753P, "BWROG In-Service Pressure Relief Technical Specification Revision Licensing Topical Report," dated February 1990
2. Letter from A. C. Thadani (NRC) to C. L. Tully (BWR Owners' Group), "Acceptance for Referencing of Licensing Topical Report NEDC-31753P, 'BWROG In-Service Pressure Relief Technical Specification Revision Licensing Topical Report,' (TAC No.

M79265) dated March 8,1993

3. GE Hitachi Nuclear Energy Report, NEDC-33532P, "Pilgrim Nuclear Power Station Safety Valve Setpoint Increase", Rev. 0, March 2010 LETTER NUMBER: 2.10.016

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Nuclear Operations, Inc. (Entergy) hereby requests an amendment to the Pilgrim Operating License Technical Specifications (TS) to address plant modifications to revise the setpoint and setpoint tolerances for Safety Relief Valves (SRVs) and Spring Safety Valves (SSVs) and changes related to the replacement of (i) four (4) Target Rock two-stage SRVs with more reliable three-stage SRVs, and (ii) two existing Dresser 3.749 inch throat diameter SSVs with Dresser 4.956 inch throat diameter SSVs.

Entergy Nuclear Operations, Inc Letter Number 2.10.016 Pilgrim Nuclear Power Station Page 2 of 3 The proposed modifications will require TS revisions to address: (a) increased setpoints and setpoint tolerances for SRVs and SSVs; (b) increased Alternate Rod Insertion (ARI) and Recirculation Pump Trip (RPT) Setpoints; (c) increased Reactor Steam Dome Pressure Limit; (d) revised SSV and SRV surveillance requirements for consistency with Boiling Water Reactor (BWR) Standard Technical Specifications (STS); (e) revised High Pressure Coolant Injection (HPCI) System and Reactor Core Isolation Cooling (RCIC) System pump surveillances for consistency with BWR STS; and (f) relocation of SRV tailpipe temperature monitoring to the Updated Final Safety Analysis Report (FSAR).

The proposed changes are described in detail in the following attachments:

Attachment 1: Description and Evaluation of the.Proposed TS changes Attachment 2: Marked-up pages of the current TS and Bases Attachment 3: (Proprietary Version) General Electric Hitachi Nuclear Energy Report, NEDC-33532P, "Pilgrim Nuclear Power Station Safety Valve Setpoint Increase", Rev. 0, March 2010 Attachment 4: (Non-Proprietary Version) General Electric Hitachi Nuclear Energy Report, NEDC-33532, "Pilgrim Nuclear Power Station Safety Valve Setpoint Increase", Rev. 0, March 2010 The marked-up Technical Specification BASES are provided for information only. , NRC approval is not required. They will be implemented in accordance with the TS 5.5.6, "Bases Control Program."

Attachment 3 contains proprietary information as defined in 10 CFR 2.390, "Public inspections, exceptions, request for withholding." General Electric Hitachi (GEH), as the owner of the proprietary information, has executed an Affidavit provided within Attachment 3, which identifies that the information has been handled and classified as proprietary, is customarily held in confidence, and has been withheld from public disclosure. Accordingly, it is requested that the proprietary information (Attachment 3) be withheld from public disclosure in accordance with the provisions of 10 CFR 2.390 and 10 CFR 9.17, "Agency records exempt from public disclosure."

A non-proprietary version of the information is provided as Attachment 4.

Entergy requests NRC approval of the proposed Pilgrim TS amendment by March 15, 2011, to support the restart from Refueling Outage (RFO)-18. Based on the NRC approval, the amendment will be implemented prior to initiating startup for Cycle 19.

This letter contains no new regulatory commitments.

If you have any questions regarding the subject matter, please contact Joseph R. Lynch at 508 830 8403.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the tS*' day of iX,2,-c 2010.

Sincerely, I-tKevin Bronson, Pilgrim Site Vice President

Entergy Nuclear Operations, Inc Letter Number 2.10.016 Pilgrim Nuclear Power Station Page 3 of 3 : Description and Evaluation of the Proposed TS Changes (14 pages). : Marked-up pages of the Current TS and Bases (17 pages). : (Proprietary Version) GE Hitachi Nuclear Energy Report, NEDC-33532P, "Pilgrim Nuclear Power Station Safety Valve Setpoint Increase", Rev. 0, March 2010 (73 Pages) : (Non-Proprietary Version) GE Hitachi Nuclear Energy Report, NEDC-33532, "Pilgrim Nuclear Power Station Safety Valve Setpoint Increase", Rev. 0, March 2010 (70 Pages) cc: Mr. James S. Kim, Project Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission One White Flint North O-8C2 11555 Rockville Pike Rockville, MD 20852 Mr. Samuel J. Collins, Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 NRC Resident Inspector Pilgrim Nuclear Power Station Director, Mass Emergency Management Agency (MEMA) 400 Worcester Road Framingham, MA 01702 Director, Massachusetts Department of Public Health (MDPH)

Radiation Control Program Commonwealth of Massachusetts 529 Main Street, Suite 1 M2A Charlestown, MA 02129-1121

Attachment 1 to Enterqy Letter 2.10.016 Proposed License Amendment to Technical Specifications Revised Technical Specification for Setpoint and Setpoint Tolerance Increases for Safety Relief Valves (SRV) and Spring Safety Valves (SSV), and Related Changes

1. DESCRIPTION
2. PROPOSED CHANGES
3. BACKGROUND
4. TECHNICAL ANALYSIS
5. REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements/Criteria
6. ENVIRONMENTAL CONSIDERATION
7. COORDINATION WITH PENDING TS CHANGE REQUESTS
8. REFERENCES Page 1 of 14

Attachment 1 to Enterqy Letter 2.10.016 Proposed License Amendment to Technical Specifications

1.0 DESCRIPTION

Pursuant to 10 CFR 50.90, Entergy requests an amendment to Pilgrim Operating License Technical Specifications (TS). Entergy is preparing a modification package to upgrade the reliability and performance of the Pilgrim Safety Relief Valves (SRVs) and Spring Safety Valves (SSVs) by replacing the four (4) Target Rock two-stage SRVs with more reliable three-stage SRVs, and two existing Dresser 3.749 inch throat diameter SSVs with Dresser 4.956 inch throat diameter SSVs. The modification package will result in system changes which involve the following:

  • Replace the Target Rock 2-stage SRVs with Target Rock 3-stage SRVs to improve the set pressure and leakage performance of the SRVs.

Replace the SSVs with a similar larger capacity model to lower peak ATWS pressure R

sufficiently to accommodate the setpoint and tolerance increases.

  • Increase the as-found setpoint tolerance for the SRVs and SSVs from +/- 1% to +/- 3% to provide greater assurance that the valves will meet setpoint requirements.
  • Raise the nominal set pressure of the SRVs by 40 psig (1115 psig to 1155 psig) to increase the simmer margin to 120 psi to reduce the vulnerability to SRV pilot leakage.
  • Raise the nominal set pressure of the SSVs by 40 psig (1240 psig to 1280 psig) to maintain the current difference between the SRV and SSV setpoints to ensure that the SSVs will not open during any anticipated operational transients.
  • Raise the High Reactor Steam Dome Pressure Trip Level Setpoint for Alternate Rod Insertion (ARI) and Recirculation Pump Trip (RPT) by 40 psig (1175 +/- 5 psig to 1215 +/- 5 psig) to reduce likelihood of unnecessary ARI actuation and recirculation pump trips.
  • Lower the High Reactor Pressure Feedwater Pump Trip Setpoint from 1415 psig to 1315 psig to support required ATWS mitigating functions.

" Increase the Reactor Steam Dome Pressure Limit from 1325 psig to 1340 psig to provide sufficient margin to accommodate variation in reactor pressure for future cycle specific over-pressure protection analysis.

  • Increase Reactor Core.Isolation Cooling (RCIC) System turbine and pump speed, power, and steam flow requirements to account for elevated reactor steam pressure. In addition, an increase is required to the RCIC Steam Line Break Detection Instrument Setpoint.

The proposed modifications will require TS revisions to address: (a) increased setpoints and setpoint tolerances for SRVs and SSVs; (b) increased Alternate Rod Insertion (ARI) and Recirculation Pump Trip Set (RPT) Setpoints; (c)rincreased Reactor Steam Dome Pressure Limit; (d) revised SSV and SRV surveillance requirements for consistency with Boiling Water Reactor (BWR) Standard Technical Specifications (STS); (e) revised High Pressure Coolant Injection (HPCI) System and Reactor Core Isolation Cooling (RCIC) System pump surveillances for consistency with BWR STS; and (f) relocation of SRV tailpipe temperature monitoring to the Updated Final Safety Analysis Report (FSAR). A mark-up of the proposed TS changes is included in Attachment 2.

The proposed TS changes were evaluated and justified based on information contained in an industry topical report for increasing SRV setpoint tolerance (Reference 1), the NRC Safety Evaluation Report (SER) that approved the SRV Setpoint Topical Report (Reference 2), and plant specific evaluations which were identified in the NRC SER as additional analyses that must be provided in order to justify the proposed SRV tolerance change. BWR STS surveillance methodology was also relied on to identify and justify the component surveillance revisions.

Page 2 of 14

Attachment 1 to Enterqy Letter 2.10.016 Proposed License Amendment to Technical Specifications The plant specific evaluations that were performed are referenced in Attachment 3, GE Hitachi (GEH) Nuclear Energy Report, NEDC-33532P, "Pilgrim Nuclear Power Station Safety Valve Setpoint Increase", March 2010 (Reference 3). This report was prepared in response to the NRC Safety Evaluation Report approving the Licensing Topical Report, NEDC-31753P, "BWROG In-Service Pressure Relief Technical Specification Revision Licensing Topical Report," which was developed to support the use of a _ 3% lift setpoint tolerance for SRVs and SSVs.

The proposed Pilgrim TS change is similar to the following previously approved TS Changes:

  • Dresden Units 2 and 3, Amendment Nos. 223 and 215, (TAC Nos. MD2166 and MD2167) dated June 21, 2007
  • Grand Gulf Unit 1, Amendment No. 123, (TAC No. M94857), dated June 12, 1996
  • River Bend Unit 1, Amendment No. 109, (TAC No. MA6185), dated February.9, 2000
  • Browns Ferry Units 2 and 3, Amendment Nos. 251 and 210, (TAC Nos., M97413 and M97414), dated May 18,1998
  • Oyster Creek, Amendment No. 261 (TAC No. MC9149), dated September 13, 2006
  • Susquehanna Units 1 and 2, Amendment Nos. 201 and 175, (TAC Nos., MB3273 and MB3274), dated March 7, 2002
  • Quad Cities Units 1 and 2, Amendment Nos., 235 and 230, (TAC Nos., MD3689 and MD3690), dated November 1, 2007 Entergy is planning to replace all four SRVs and both SSVs during Refueling Outage (RFO)-18.

RFO-1 8 is planned to commence on or about April 17, 2011. License Amendment approval is required prior to restart from RFO-1 8.

2.0 PROPOSED CHANGE

identifies the proposed changes in the form of marked-up current Pilgrim Technical Specifications pages and marked-up TS BASES pages. The BASES pages are provided for NRC information only. TS BASES changes will be completed in accordance with the TS 5.5.6, BASES Control Program and incorporated into TS upon receipt of the NRC approved License Amendment.

The proposed TS changes are summarized in Table 1 below:

Table 1: Proposed TS Changes Item TS No. and/or TS Page No. or TS Proposed Changes No. BASES Page No. (bold face) 1 TS 2.1.4, page 2-1 Reactor steam dome pressure is revised from <51325 to *1340 psig at any time when irradiated fuel is present in the reactor vessel.

2 BASES: 2.0 SAFETY LIMITS (Cont). BASES for REACTOR STEAM DOME PRESSURE BASES Page B2-4 (2.1.4) are revised consistent with TS 2.1.4.

3 TS Table 3.2.F, on TS Page 3/4.2-26 Tailpipe Temperature Indication; asterisk in Note (5) is and NOTES FOR TABLE 3.2.F on TS removed; and Note (6) is revised and relocated to Page 3/4.2-28 FSAR.

4 TABLE 3.2-G on TS Page 3/4.2-29, High Reactor Dome Pressure Trip Level Setting "1175 High Reactor Dome Pressure Trip +/- 5 psig" is revised to "1215 +/- 5 psig".

Level Setting and BASES page 3B3/4.2-6 Page 3 of 14

Attachment 1 to Enterov Letter 2.10.016 Proposed License Amendment to Technical Specifications 5 SR 4.5.C on TS Page 3/4.5-7 HPCI system testing for HPCI pump operability surveillance is revised consistent with BWR Standard TS methodology.

6 SR 4.5.D.1 on TS Page 3/4.5-8 is RCIC system testing for RCIC pump operability revised surveillance is revised consistent with BWR Standard TS methodology.

7 TS BASES 3/4.5 BASES, Page BACKGROUND, 3ra paragraph is revised for ADS to B3/4.5-21 reflect new values for SRVs.

8 TS 3.6.D.1, 3.6.D.2 NOTE and TS 3.6.D.1, 3.6.D.2 NOTE and 3.6.D.3 are revised to 3.6.D.3 on TS Page 3/4.6-6, and TS reflect the new SRV and SSV setpoints and adoption of 3.6.D.4 and 5 on TS Page 3/4.6-7 BWR Standard TS surveillance methodology.

9 BASES: 3/4.6.D on BASES Page SRV and SSV BASES are revised to reflect the three B3/4.6-7 and B3/4.6-8 stage SRVs and new SSVs and new BWR Standard TS surveillance methodology.

3.0 BACKGROUND

The safety objective of the Pressure Relief System is to prevent over pressurization of the reactor vessel and the attached piping which forms the Reactor Coolant System (RCS). The Pressure Relief System includes four (4) SRVs and two (2) SSVs. The SSVs discharge directly into the interior space of the drywell. The SRVs discharge through their individual discharge piping, terminating below the minimum suppression pool (torus) water level.

The four SRVs and the two SSVs will be replaced. These valves are installed on the Main Steam lines located in primary containment between the reactor pressure vessel and the main steam line flow restrictors. The affected SRVs are RV-203-3A, -3B, -3C and -3D and the affected SSVs are RV-203-4A and RV-203-4B.

SRV RV-203-3A is located on steam line A SRV RV-203-3B is located on steam line D SRV RV-203-3C is located on steam line D SRV RV-203-3D is located on steam line B SSV RV-203-4A is located on steam line A SSV RV-203-4B is located on steam line C The SRVs require replacement because the current two-stage Target Rock SRVs have been unreliable performers with respect to leaking while in-service and the subject of setpoint drift.

SRV pilot valve leakage has led to multiple plant shutdowns and the setpoint drift problem resulted in exceeding current TS limits and numerous Licensee Event Reports (LERs). It has been determined that pilot valve leakage is due to low simmer margin and high as-found lift setpoints are due to corrosion bonding at the pilot valve disc/seat. To address current SRV performance problems, Entergy has performed extensive investigations and feasibility studies.

The preferred option for correcting these problems is to replace all SRVs and SSVs during the next refueling outage. RFO-1 8 is currently planned to start on or about April 17, 2011.

Entergy will be replacing the existing four (4) Target Rock Model 7567F two-stage SRVs with Target Rock Model 0867F three-stage SRVs. The setpoints for the SRVs will be increased from 1115 psig to 1155 psig, and the allowable as-found setpoint tolerance will be increased from the current +/-1% to +/-3%. The new SRVs have the same bore size as the existing two-stage SRVs.

Following as-found testing, the SRV setpoint will be restored to the current setpoint tolerence of

+/-_1%.

Entergy will also be replacing the existing two Dresser type 3777Q SSVs with new Dresser type 3707RR SSVs. The SSV setpoints will be increased from 1240 psig to 1280 psig, and the allowable as-found setpoint tolerance will be increased from the current +/-1% to +/-3%. The new Page 4 of 14

Attachment 1 to Enteray Letter 2.10.016 Proposed License Amendment to Technical Specifications SSV throat diameter will be increased from 3.749 inches to 4.956 inches for increased relief capacity. Following as-found testing, the SSV setpoint will be restored to the current setpoint tolerence of +/- 1%.

In addition, Pilgrim has elected to adopt BWR STS surveillance test methodology to demonstrate SSVs and SRVs operability. As a result, the existing TS surveillance for the two-stage Target Rock SRVs for tailpipe temperature monitoring, as specified in TS 3.6.D.3, 4, and 5, is not required and will be revised and relocated into the UFSAR.

The design Code for the SRVs is ASME Section III, 1968 through summer 1970 Addenda and for SSVs is ASME Section III, 1965 through summer 1966 addenda. ASME Code Section III, Article 9, N-911.2 (5) requires that, if bellows are used, means must be provided to detect bellows failure. The bellows integrity-monitoring capability will be identified and described in the Pilgrim UFSAR.

The SSVs and SRVs relieve reactor pressure through the main steam lines by direct mechanical actuation due to system overpressure. The mechanical overpressure setpoint of the SSVs and SRVs provides the Nuclear System Pressure Relief safety function that protects the reactor vessel and associated RCS pressure boundary. Each of the four SRVs can be individually opened by an operator using a control switch located in the main control room.

Each of the four SRVs opens automatically by the Automatic Depressurization System (ADS) system. SRV opening by either the control switch or ADS requires energization of a solenoid mounted on the SRV that admits pressurized nitrogen gas to an air operator mounted on the SRV. Automatic actuation of the solenoid valve is a function of the Automatic Depressurization System (ADS) and is relied on open the valve to reduce reactor pressure so that the Low Pressure Coolant Injection (LPCI) System and the Core Spray System (CS) can reflood the core to protect the fuel following certain postulated transients or accidents. The ADS function and logic features are not changed or affected by the proposed plant modifications.

Increased SSV and SRV setpoint and setpoint tolerances will result in higher reactor pressure during accident and transient conditions. Abnormal Operational Occurrences (AOO) and Anticipated Transient Without Scram (ATWS) events required re-evaluation. A review of high reactor pressure trip level settings established for the Alternate Rod Insertion (ARI),

Recirculation Pumps Trip (RPT) and Feedwater Pump Trip was performed. The analysis provided in Reference 3 is based on the proposed ARI and RPT setpoints that were increased from 1175 + 5 psig to 1215 + 5 psig and a Feedwater Pump Trip setpoint that was lowered from 1415 psig to 1315 psig. The high reactor pressure Feedwater Pump Trip instrumentation is not included in the Technical Specifications.

Increased reactor pressure during accidents and transients required an assessment of high pressure system operation and test requirements. Based on these reviews, the HPCI System and the Standby Liquid Control (SLC) System does not require modification. However, the RCIC System requires a turbine and pump speed increase to pump against the increased reactor pressure and the motor operated RCIC pump Injection Valve (MO-1301-49) requires modification. Applicable procedures will be updated to address revised RCIC System operating requirements. In addition, Pilgrim has elected to adopt BWR STS pump test methodology for High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) System pump testing.

During RFO-18, four new SRVs and two new SSVs will be installed. During subsequent refueling outages, Entergy will replace and test SRVs and SSVs in accordance with the current IST program. Since Pilgrim is on a 24-month operating cycle, the replacement and testing of SRVs differs from the NRC SER NEDC-31753P, but complies with the current ASME OM Code requirement.

Page 5 of 14

Attachment 1 to Enterpy Letter 2.10.016 Proposed License Amendment to Technical Specifications

4.0 TECHNICAL ANALYSIS

The proposed TS changes were evaluated and justified based on information contained in an industry topical report for increasing SRV setpoint tolerance (Reference 1), the NRC SER that' approved the SRV Setpoint Topical Report (Reference 2), and plant specific evaluations which were identified in the NRC SER as additional analyses that must be provided in order to justify the proposed SRV tolerance change. BWR STS surveillance methodology was also relied on to identify and justify the component surveillance revisions.

The analyses performed to address the SSV and SRV setpoint tolerance revision as identified in the Topical Report, the associated NRC SER, and the plant specific analyses performed to satisfy SER requirements, envelope the analyses that are required to address the SSV and SRV setpoint change and the revision to SSV blowdown capacity. The analyses performed identify maximum reactor pressure during accidents and transient conditions and assess impact on structures, systems and components (SSCs) and their ability to meet required safety functions. Where necessary, TS revisions and plant modifications were identified to ensure SSC operability.

Reference 1, BWROG Licensing Topical Report, NEDC-31753P was reviewed and approved by the NRC as documented in a safety evaluation issued by Reference 2. The NRC determined that it was acceptable for licensees to submit TS amendment requests to revise the safety function setpoint tolerance to +/- 3%, provided that the setpoints for those valves are restored to within +/- 1% prior to reinstallation. The NRC also indicated in its safety evaluation that licensees planning to implement TS changes to increase the setpoint tolerances should provide the following plant-specific analyses.

1. Transient analysis, using NRC approved methods, of abnormal operational occurrences as described in NEDC-31753P utilizing a +/- 3% lift setpoint tolerance for the SRVs and SSVs.
2. Analysis of the design basis overpressure event using the +/- 3% tolerance limit for the safety valve setpoints to confirm that the vessel pressure does not exceed ASME pressure vessel code transient limits.
3. Plant-specific analyses described in items 1 and 2 should assure that the number of SRVs and SSVs included in the analysis corresponds to the number of valves required to be operable in the TS.
4. Re-evaluation of the performance of high pressure systems (e.g., pump capacity, discharge pressure, etc.), motor-operated valves, and vessel instrumentation and associated piping considering the +/- 3% tolerance limit.
5. Evaluation of the +/- 3% tolerance on any plant-specific alternate operating modes (e.g.,

increased core flow, extended operating domain, etc.) should be completed.

6. Evaluation of the effects of the +/- 3% tolerance limit on the containment response during Loss of Coolant Accidents (LOCAs) and the hydrodynamic loads on the safety valve discharge lines and containment should be completed.

In support of the proposed TS changes, GEH and Entergy have performed specific analyses and evaluations, and the results are referenced in Attachment 3. The GEH analyses addresses impact of (a) increased setpoints for SRVs and SSVs, (b) increased SRV and SSV +/- 3%

tolerance limits and (c) increased SSV capacity. The results of these analyses have determined that the impact of the setpoint changes, increased tolerances, and increased SSV capacity are acceptable.

Page 6 of 14

Attachment 1 to Enteray Letter 2.10.016 Proposed License Amendment to Technical Specifications 4.1 Transient Analysis and Over-Pressure Analyses GEH Report Sections 2, 3, and 5.3 address NRC SER items 1, 2, and 3, as follows:

The ASME Boiler and Pressure Vessel Code requires the reactor pressure vessel to be protected from overpressure during transient conditions by self-actuated safety valves. For the purpose of the analyses, all four SRVs and two SSVs are assumed to operate in the safety mode. The Safety Limit for the reactor ,steam dome pressure has been selected such that it is at a pressure below which it can be shown that the integrity of the reactor coolant system is not endangered. The reactor pressure limit of 1340 psig as measured in the vessel steam dome was derived from the design pressures of the reactor vessel. The peak pressures for the piping systems connected to the reactor vessel have been recalculated based on a reactor steam dome peak pressure of 1340 psig. These peak pressures are below the lowest of the transient pressures permitted by the applicable design code: ASME Boiler and Pressure Vessel (B&PV)

Code (1965 Edition, including January 1966 Addendum) for the pressure vessel, USAS Piping Code Section B31.1 for the steam space piping, and ASME Section III for the reactor coolant system recirculation piping. The ASME B&PV Code permits pressure transients up to 10% over the vessel design pressure (110% x 1250 = 1375 psig). The USAS Piping Code and ASME Section III permit pressure transients and other occasional loads whose combined effect do not exceed stress levels based on the duration of the loads and the applicable service limit.

The critical characteristics of the SRVs include set pressure with tolerance and rated relief capacity at set pressure. The as-left setpoint tolerance (+/- 1% of nominal setpoint) will not be changed by this modification. The nominal set pressure is increased by 40 psig to 1155 psig and the analyzed as-found setpoint tolerance is increased to +/- 3%. The delay and stroke times will remain unchanged.

(

The critical characteristics of the SSVs include set pressure with tolerance and rated relief capacity at set pressure. The as-left setpoint tolerance (+/- 1% of nominal setpoint) will not be changed by this modification. The nominal set pressure is increased by 40 psig to 1280 psig and the analyzed setpoint tolerance is increased to +/- 3%. The opening response time is unchanged. The new SSVs will have an increased bore size and larger outlet than the existing SSVs. The increased throat diameter of the SSVs increases the valve relief capacity.

All SRVs and SSVs are required to be operable to satisfy the ASME over pressure analyses.

The setpoints are established to ensure that the applicable code limits for peak reactor and coolant piping pressure are satisfied.

To comply with the ATWS Rule, 10 CFR 50.62, Pilgrim previously modified the Standby Liquid Control (SLC) System, and implemented instrumentation to initiate Alternate Rod Insertion (ARI) and Recirculation Pump Trip (RPT). GEH has evaluated the increase in setpoints, setpoint tolerance, and SSV capacity for ATWS transients. The proposed increase in the SRV and SSV setpoints, setpoint tolerance, and SSV capacity does not impact the initiation or operation of SLC System. The setpoint for the initiation of ARI and RPT changes as a result of the increased SRV setpoint, as discussed in the Pilgrim-specific evaluation (Attachment 3). TS Table 3.2-G has been revised to reflect the increased trip level setting for ARI and RPT initiation functions. These setpoints have been increased consistent with SRV/SSV setpoint increases to avoid spurious activation during non-ATWS abnormal operating occurrences. The increased ARI/RPT setpoint is acceptable per the GEH analyses. Transient analyses results indicate that all acceptance criteria are met.

Page 7 of 14

Attachment 1 to Enteray Letter 2.10.016 Proposed License Amendment to Technical Specifications 4.2 Impact on High Pressure Systems GEH Report Sections 3 and 5 addresses NRC SER item 4, as follows:

The High Pressure Coolant Injection system ensures that the reactor core is adequately cooled to limit fuel clad temperature in the event of a small break in the nuclear system and loss of coolant which does not result in rapid depressurization of the reactor vessel. GEH analyses of the HPCI system due to the increased SRV setpoint of 1155 psig and +/- 3% setpoint tolerance concluded that HPCI turbine and pump design speed increase was not required and adequate system margin exists to accommodate the SRV/SSV setpoint changes. Since there are no system changes required, the current design conditions for the HPCI system are unchanged.

The Reactor Core Isolation Cooling (RCIC) system provides makeup water at reactor pressures up to the new SRV setpoint. Pilgrim-specific evaluation of the RCIC system due to' the increased SRV setpoint of 1155 psig and +/-3% drift concluded that a speed increase is required to 4628 RPM. Based on the evaluations in Attachment 3, a design speed increase of up to 4700 RPM is acceptable. This speed increase requires more steam flow which impacts the high steam flow isolation setpoints. The increase in pressure and flow impacts the steam supply isolation, steam supply admission and pump injection motor operated valves. The valve review process is discussed below.

/

Taking into account the increase in reactor pressure due to increased SRV/SSV setpoints, the, HPCI surveillance 4.5.C and RCIC surveillance 4.5.D.1 for pump operability are revised to be consistent with the surveillances format specified in BWR Standard Technical Specifications.

RCS boundary valves are under review to assure that they will perform their function at the increased design pressure resulting from the SRV/SSV set point and setpoint tolerance increase. All valves reviewed to date were successfully screened with the~exception of the RCIC Pump Injection Valve (MO1301-49). This motor operated valve (MOV) does not demonstrate sufficient margin based on a review of the weak link and torque/thrust analyses.

PNPS will modify valve components or replace MOVs as a part of the SRV/SSV modification package to assure that sufficient margin exists prior to implementing the full modification package. Interfacing piping is also under review as,a part of this project. To date, there are no interfacing piping issues identified that require modification. However, should any modifications be required they will be performed in accordance with the Pilgrim design and 10 CFR 50.59 processes.

The ATWS analysis in Section 3 of the attached GEH NEDC-33532P report is based on the' SLC system delivering 39 gpm of 8.42% sodium pentaborate solution with a minimum B10 enrichment of 54.5 atom-percentage. These SLC system equipment parameters are the minimum required by PNPS Technical Specifications and provide a hot shutdown capability equivalent to 10 CFR 50.62 requirements to inject 86 gpm of 13% sodium pentaborate solution with a B10 enrichment of 19.8 atom-percentage (natural enrichment). The SLC system includes two SLC pumps and each pump, discharge has a separate relief valve installed to prevent system over-pressurization. Only one out of the two SLC pumps is operated for ATWS mitigation. Entergy has reviewed both relief valve setpoints and verified that each of the relief valves will remain closed when the associated pump is in operation with the maximum lower plenum pressure of 1212 psig which was reported in Section 3 of the attached report NEDC-33532P.

Page 8 of 14

Attachment 1 to Enterqy Letter 2.10.016 Proposed License Amendment to Technical Specifications 4.3. Alternate Operating Modes GEH Report Section 2.4 addresses NRC SER item 5, as follows:

GEF has evaluated the impact of increased setpoints, setpoint tolerance, and SSV capacity on Pilgrim-specific alternate operating modes. The alternate operating modes, including the Maximum Extended Load Line Limit Allowed (MELLLA), Increased Core Flow (ICF), Feedwater Temperature Reductions; Turbine Bypass Out-Of-Service (OOS), Main Steam Isolation Valve (MSIV) OOS and Single Loop Operation (SLO) were considered in determining the most restrictive analytical conditions (i.e., the most limiting operating mode) for performing the analysis associated with the proposed TS change. Therefore, the impact of the proposed change on the Pilgrim-specific alternate operating modes has been explicitly addressed and determined to be acceptable.

4.4 Containment Response Due to Dynamic Loads GEH Report section 4 addresses NRC SER item 6, as follows:

Entergy evaluated the impact of SRV and SSV discharge loads on primary containment. The SRV discharge loads are defined by parameters that include:

" SRV discharge line (SRVDL) and containment geometry

  • Water leg length in the SRVDL at the time of SRV opening

" SRV flow capacity and SRV opening pressure Since an SRV setpoint increase and the setpoint tolerance will increase the SRV valve opening pressure, the SRV discharge dynamic loads will increase. Entergy has evaluated the SRV dynamic load increases for the associated piping and torus submerged structures and the evaluation concluded that all piping and structures were found to meet Code requirements.

The SSV discharge loads are defined by parameters that include the SSV flow capacity and SSV opening pressure.

Since an SSV setpoint increase and setpoint tolerance will increase the SSV safety valve opening pressure and an increase in the SSV throat size will increase the SSV flow capacity, the SSV dynamic loads are expected to increase. Entergy has evaluated the SSV dynamic loads for the associated piping and the evaluation concluded that all piping and structures were found to meet Code requirements.

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration Entergy requests an amendment to Pilgrim Operating License Technical Specifications (TS) to (a) increase the setpoints for Safety Relief Valves (SRVs) and Spring Safety Valves (SSVs) and (b) setpoint tolerances for SRVs and SSVs from +/- 1% to +/- 3% and (c) TS changes resulting from the increased SRV and SSV setpoints. The proposed change revises (i) TS 2.1.4, reactor steam dome pressure limit to <1340 psig, (ii) relocates SRV tailpipe temperature indication instrumentation included in TS Table 3.2.F and SRV tailpipe temperature monitoring included in TS 3.6.D.3, 4, and 5 into the Pilgrim Final Safety Analysis Report (UFSAR), (iii) revises TS 3.6.D to reflect the increased SRV and SSV setpoint tolerance drift limits of +/-,3%, (iv) revises trip level setting in TS Table 3.2-G for the initiation of Anticipated Transient Without Scram (ATWS) Alternate Rod Insertion and Recirculation Pump Trip instrumentation, (v) revises the surveillance Page 9 of 14

Attachment 1 to Enterqy Letter 2.10.016 Proposed License Amendment to Technical Specifications requirement (SR) 4.5.C and 4.5.D for reactor core isolation cooling (RCIC) and high pressure coolant injection (HPCI) pump operability to be consistent with the BWR Standard Technical Specification format, and (vi) revises the applicable TS Bases describing the above TS changes.

The proposed changes do not alter the TS requirements for the number of SRVs and SSVs required to be operable, the SRV testing frequency, or the manner in which the valves are operated. The current TS requirement to adjust the SRV and SSV as-left tolerance to within +/- 1 % of the nominal lift setpoint, prior to returning a valve to service, is not being changed.

According to 10 CFR 50.92 (c), "Issuance of amendment," a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

Entergy has evaluated the proposed Pilgrim TS changes using the criteria in 10 CFR 50.92, and has determined that the proposed changes do not involve a significant hazards consideration. The following information is provided to support a finding of no significant hazards consideration.

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

Response: No The proposed change increases the allowable as-found SRV and SSV setpoint tolerance, determined by test after the valves have been removed from service, from +/-

1% to +/- 3%. The proposed change also increases the SRV and SSV setpoints.

Analysis of these changes demonstrates that reactor pressure will be maintained below the applicable code overpressure limits. The proposed change increases the SSV discharge capacity due its increased throat diameter. The proposed change does not alter the TS requirements for the number of SRVs and SSVs required to be operable, the allowable as-left lift setpoint tolerance, the testing frequency, or the manner in which the valves are operated. Consistent with current TS requirements, the proposed change continues to require that the safety valves be adjusted to within

+/- 1 % of their nominal lift setpoints following testing. The proposed increase in the SRV and SSV setpoint complies with the ASME Boiler and Pressure Vessel (B&PV)

Code (1965 Edition, including January 1966 Addendum) for the pressure vessel, USAS Piping Code Section B31.1 for the steam space piping, and ASME Section III for the reactor coolant system recirculation piping. Since the proposed change does not alter the manner in which the valves are operated, there is no significant impact on the reactor operation.

The proposed change does not involve a change to the safety function of the valves.

The proposed TS revision involves no significant changes to the operation of any systems or components in normal or accident operating conditions. Therefore, these Page 10 of 14

Attachment 1 to Enteray Letter 2.10.016 Proposed License Amendment to Technical Specifications changes will not increase the probability of an accident previously evaluated.

Since an SSV setpoint increase and setpoint tolerance will increase the SSV safety valve opening pressure and an increase in the SSV throat size will increase the SSV flow capacity, the SSV dynamic loads are expected to increase. Entergy has evaluated the SSV dynamic loads for the associated piping. All piping and structures were found to meet Code requirements.

Since an SRV setpoint and the setpoint tolerance increase will increase the SRV valve opening pressure, the SRV discharge dynamic loads will increase. Entergy has evaluated the SRV dynamic load increases for the associated piping and torus submerged structures and the evaluation concluded that all piping and structures were found to meet Code requirements.

The proposed revision to the HPCI and RCIC pump operability determination surveillance follows the format of BWR Standard Technical Specification surveillance, and complies with in-service testing for pump operability determination in accordance with ASME OM Code requirement.

Generic considerations related to the change in setpoints and setpoint tolerance were addressed in NEDC-31753P, "BWROG In-Service Pressure Relief Technical Specification Revision Licensing Topical Report," and were reviewed and approved by the NRC in a safety evaluation dated March 8,, 1993. General Electric Hitachi Company (GEH) completed plant-specific analyses to assess the impact of increase in SRV and SSV setpoints and increase in the setpoint tolerance from +/- 1% to +/- 3%.

The impact of the increases in the SRV and SSV setpoints and increases in the setpoint tolerances, as addressed in this analyses, included vessel overpressure, Updated Final Safety Analysis Report (UFSAR) Chapter 14 events, ATWS, Loss of Coolant Accident (LOCA), containment response and dynamic loads, high pressure systems performance, operating mode and equipment out of service. The proposed change is supported by GEH analysis of events that credit the SRVs and SSVs.

The plant specific evaluations, required by the NRC's safety evaluation and performed to support this proposed change, demonstrate that there is no change to the design core thermal limits and adequate margin to the reactor coolant system pressure limits exists. These analyses also demonstrate that operation of Core Standby Cooling Systems (CSCS) is not adversely affected and the containment response following a LOCA is acceptable. The plant systems associated with these proposed changes are capable of meeting applicable design basis requirements and retain the capability to mitigate the consequences of accidents described in the UFSAR. Therefore, these changes do not involve an increase in the consequences of an accident previously evaluated.

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

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

Response: No The proposed change increases the allowable as-found lift setpoint tolerance for the Pilgrim SRV and SSV valves. The proposed change to increase the tolerance was developed in accordance with the provisions contained in the NRC safety evaluation for NEDC-31753P. SRVs and SSVs installed in the plant following testing will Page 11 of 14

Attachment 1 to Enterqy Letter 2.10.016 Proposed License Amendment to Technical Specifications continue to meet the current tolerance acceptance criteria of _ 1 % of the nominal setpoint. The proposed change does not affect the manner in which the overpressure protection system is operated; therefore, there are no new failure mechanisms for the overpressure protection system.

The proposed changes do not change the safety function of the SRVs and SSVs, or HPCI and RCIC systems. There is no alteration to the parameters within which the plant is normally operated. The increase in SRV and SSV setpoints, setpoint tolerance, and increased SSV discharge capacity are not precursors to new or different kind of accidents and do not initiate new or different kind of accidents. The impact of these changes have been analyzed and found to be acceptable within the design limits and plant operating procedures.

As a result, no new failure modes are being introduced. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No The margin of safety is established through the design of the plant structures, systems, and components, the parameters within which the plant is operated and the establishment of the setpoints for the actuation of equipment relied upon to respond to an event. The proposed change modifies the setpoints at which protective actions are initiated, and but does not change the requirements governing operation or availability of safety equipment assumed to operate to preserve the margin of safety.

Establishment of the +/- 3% SRV and SSV setpoint tolerance limit does not adversely affect the operation of any safety-related component or equipment. Evaluations performed in accordance with the NRC safety evaluation for NEDC-31753P have concluded that all design limits will continue to be met.

Therefore, the proposed change does not involve a significant reduction in the margin of safety.

Based upon the above, 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 Applicable Regulatory Requirements and Criteria The current +/- 1% tolerance band on the SRVs and SSVs setpoints stems from the original acceptance criterion defined by the Pilgrim Technical Specifications. Nuclear power plant licensees have experienced difficulty in meeting the typical' +/- 1 % lift setpoint tolerance. As a result, the BWROG developed NEDC-31753P to support the use of the +/-

3% lift setpoint tolerance.

NEDC-31753P was reviewed and approved by the NRC as documented in Reference 2.

The NRC determined that it is acceptable for licensees to submit TS amendment requests to revise the main steam safety relief valve lift setpoint tolerance to +/- 3%,

provided that the setpoints for those valves tested are restored to +/- 1% prior to reinstallation. The NRC also indicated in its safety evaluation that licensees planning to implement TS changes to increase the setpoint tolerances should provide a plant specific Page 12 of 14

Attachment 1 to Enterqy Letter 2.10.016 Proposed License Amendment to Technical Specifications analysis. The plant specific analysis for Pilgrim is provided in Attachment 3.

The existing main steam safety relief valves are tested in accordance with Pilgrim Technical Specifications and the ASME OM Code, "Code for Operation and Maintenance of Nuclear Power Plants." The Pilgrim fourth ten year inservice testing (IST) program implements the 1998 Edition through 2000 Addenda of the ASME OM Code. Appendix I, "Inservice Testing of Pressure Relief Devices in Light-Water Reactor Nuclear Power Plants," Section 1-1300, "Guiding Principles," of the ASME OM Code requires that a sample of valves from each valve group be periodically tested and that each value will be tested at least once every 5 years. The as-found acceptance criteria for those valves tested is the greater of either the t tolerance limit of the owner established set-pressure acceptance criteria (I.e., currently +/- 1%) or +/- 3% of the valve nameplate set-pressure.

Since the current ASME OM Code allows the +/- 3% limit to be used, no relief from the ASME OM Code is required with regard to the setpoint tolerance change. However, a change to the TS is required to revise the owner-established setpressure acceptance criteria to +/- 3%.

The proposed increase in the SRV and SSV setpoint complies with the ASME Boiler and Pressure Vessel (B&PV) Code (1965 Edition, including January 1966 Addendum) for the pressure vessel, USAS Piping Code Section B31.1 for the steam space piping, and ASME Section III for the reactor coolant system recirculation piping.

In conclusion, based on the considerations discussed above, (1) there is a 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 NRC's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

Entergy has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, -Standards for Protection Against Radiation." However, the proposed amendment 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, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22, "Criterion for Categorical Exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review", Paragraph (c)(9). Therefore, in accordance with 10 CFR 51.22, Paragraph (b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 COORDINATION WITH PENDING TS CHANGES There are no pending proposed TS changes that are being filed for license amendment that would impact the SRV and SSV TS setpoint changes.

Page 13 of 14

Attachment 1 to Enterqy Letter 2.10.016 Proposed License Amendment to Technical Specifications

8.0 REFERENCES

1. NEDC-31753P, "BWROG In-Service Pressure Relief Technical Specification Revision Licensing Topical Report, dated February 1990
2. Letter from A. C. Thadani (NRC) to C. L. Tully (BWR Owners' Group), "Acceptance for Referencing of Licensing Topical Report NEDC-31753P, 'BWROG In-Service Pressure Relief Technical Specification Revision Licensing Topical Report' (TAC No. M79265),"

dated March 8, 1993

3. GE Hitachi Nuclear Energy Report, NEDC-33532P, "Pilgrim Nuclear Power Station Safety Valve Setpoint Increase", Rev. 0, March 2010 Page 14 of 14 to Enterqy Letter 2.10.016 Marked-Up TS Pages and BASES Pages (17 Paqes)

2.0 SAFETY LIMITS 2.1 Safety Limits 2.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% of rated core flow:

THERMAL POWER shall be < 25% of RATED THERMAL POWER.

2.1.2 With the reactor steam dome pressure > 785 psig and core flow > 10% of rated core flow:

MINIMUM CRITICAL POWER RATIO shall be > 1.08 for two recirculation loop operation or > 1.1 or single recirculation loop operation.

2.1.3 Whenever the reactor is in the cold shutdown condition with irradiated fuel in the reactor vessel, the water level shall not be less than 12 inches above the top of the normal active fuel zone.

2.1.4 Reactor steam dome pressure shall be < a p_.sig at any time when irradiated fuel is present in the reactor vessel.

2.2 Safety Limit Violation With any Safety Limit not met within two hours the following actions shall be met:

2.2.1 Restore compliance with all Safety Limits, and 2.2.2 Insert all insertable control rods.

Amendment No. 15, 27, 2, 7-2,133, 146, 471, 101, 2,0, 22.,-.*2*2 2-1

A BASES: I'A stl'ý W 2.0 SAFETY LIMITS (Cont) / sa-,

REACTOR STEAM The Safety Limit for the reactor steam dome pressure has DOME PRESSURE been selecte such that it is at a pressure below which it can (2.1.4) be shown that Sue integrity of the system is ndangered.

The reactor pres re vessel is designed ection III of the ASME Boiler and ssure Vessel C (1965 Edition, including the Janua 966 Adden m), which permits a maximum pressure tran aent of.. 0%, 1375 psig, of design pressure 1250 psig. The ty Umit of 1325 psig, as measured by the reactor e dome pressure indicator, is equivalent to 1375 psi at the I est elevation of the reactor coolant system. T reactor cool t system is designed to ASME Section IT or the reactor reci lation piping, which permits a mr imum pressure transient 120% of design pressure:f 1148 psig at 5620F for suctio l~iping and 1241 ipsig at 562O fo.ichrepipin~g. The pres sre Safety Limit )

'*is selected to be the lowest transient overpressure allowed by e applicable codes.

REFERENCES 1) "General Electric Standard Application for Reactor Fuel,"

NEDE-24011 -P-A (through the latest approved amendment at the time the reload analyses are performed as specified in the CORE OPERATING LIMITS REPORT).

2) General Electric Thermal Analysis Basis (GETAB): Data, Correlation and Design Application, General Electric Co. BWR Systems Department, January 1977, NEDE-1 0958-PA and NEDO-1 0958-A. I/
3) "Methodology & Uncertainties for SLMCPR Evaluations,"

NEDC-32601-P-A (August 1999).

4) "Power Distribution Uncertainties for Safety Limit MCPR Evaluations," NEDC-32694-P-A (August 1999).
5) "GE 11 Compliance with Amendment 22 of GESTAR I1,"

NEDE-.31917P (April 1991).

6) "GE 14 Compliance with Amendment 22 of GESTAR II,"

NEDC-32868P (December 1998).

- Adj W

Revision 225 Amendment No. 15, 133, 116, 1:71, 191 B2-4

INSERT A The Safety Limit for the reactor steam dome pressure has been selected such that it is at a pressure below which it can be shown that the integrity of the reactor coolant system is not endangered. The reactor pressure ýlimit of 1340 psig as measured in the vessel steam dome was derived from the design pressures of the reactor vessel. The peak pressures for the piping systems connected to the reactor vessel have been recalculated based on a reactor steam dome peak pressure of 1340 psig. These peak pressures are below the lowest of the transient pressures permitted by the applicable design code: ASME Boiler and Pressure Vessel (B&PV) Code (1965 Edition, including January 1966 Addendum) for the pressure vessel, USAS Piping Code Section B3 1.1 for the steam space piping, and ASME Section III for the reactor coolant system recirculation piping. The ASME B&PV Code permits pressure transients up to 10% over the vessel design pressure (110% x 1250

= 1375 psig). The USAS Piping Code and ASME Section III permit pressure transients and other occasional loads whose combined effect do not exceed stress levels based on the duration of the loads and the applicable service limit.

INSERT B

7) "Pilgrim Nuclear Power Station Safety Valve Setpoint Increase," GE Hitachi Nuclear Energy Report, NEDC-33532P, Rev. 0 (March, 2010).

I it PNPS 4

,,-riLE 3.2. F (Cont)

Minimum # of SURVEILLANCE INSTRUMENTATION Operable Instrument 7

Parameter Type Indication Notes Channels Instrument # and Ranne 7 TI-5021-2A TRU-5021-IA Suppression Chamber Water Temperature Indicator/

Muitipoint Recorder (1) (2) (3) (4)

'K 30-230°F (Bulk)

TI-5022-2B Suppression Chamber Indicator/

TRU-5022-IB Water Temperature Multipoint Recorder (1) (2) (3) (4) 30-230°F (Bulk) 1 PID-5021 DrywelVTorus Diff. Indicator -.25 - +3.0 psig (1) (2) (3) (4)

Pressure 1 PID-5067A Drywell Pressure Indicator -.25 - +3.0 psig (1) (2) (3) (4)

PID-5067B Torus Pressure Indicator - 1.0 - +2.0 psig l/Valve (a)Primary Safety/Relief Valve (a) Acoustic monitor (5) or Position (b) Thermocouple (b) Backup I/Valve (a)Primary Safety Valve Position (a) Acoustic monitor (5) or Indicator (b) Thermocouple (b) Backup il~t~a"- See Note (6)--.TIrp Inditatiar T riap.ure Theff= Plr' (I

LI-1001-604A Torus Water Level Indicator/Multipoint (1) (2) (3) (4) 2 LR-1001-604A (Wide Range) Recorder 0-300" H2 0 LI-1001-6048 LR-1001-604B Torus Water Level Indicator/Multipoint (1) (2) (3) (4)

(Wide Range) Recorder 0-300" H20 Amendment No. 5 83-2-/ 3/4,2-26

NOTES FOR TABLE 3.2.F (1) With less than the minimum number of instrument channels, restore the inoperable channel(s) within 30 days.

(2) With the instrument channel(s) providing no indication to the control room, restore the indication to the control room within seven days.

(3) If the requirements of notes (1) or (2) cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(4) These surveillance instruments are c siderad to be redundant to each other.

(5) At a minimum, the primary or back- 4 .arameter indicators shall be operable for each valvejiben the valves are required to be -operable. With both primary and back n%)trument channels inoperable either return one (1) channel to operable s--ltus within 31 days or be in a shutdown mode within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The following instruments are associated with the safety/relief and safety valves:

Primary Valve Acoustic Monitor Tail Pipe Temperature Thermocouple 203-3A ZT-203-3A TE6271 1

  • 203-3B ZT-203-3B TE6272 203-3C ZT-203-3C TE6273 203-3D ZT-203-3D TE6276 203-4A ZT-203-4A TE6274-B 203-4B ZT-203-4B TE6275-B (6) At a minimum, for t eoouespoinSRtalietmerture one o the dual thtrmocouples will be operable for each SRV whe a ves are required to es inoperable, it shall be returned to an o er on within 31 day~s or hl b laced in a sutdown mode within 24 our (7) With less than the minimum number of operable instrument channels, restore the inoperable channels to operable status within 7 days or prepare and submit a special report to the Commission within 14 days of the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the channelsto operable status.

Amendment No. 487-893;-193, 3/4.2-28

PNPS IW TABLE 3.2-G INSTRUMENTATION THAT INITIATES RECIRCULATION PUMP TRIP AND ALTERNATE ROD INSERTION Minimum Number of Operable or Tripped Instrument Channels Per Trip System (1) Trin Function Trin Level Setting 2 High Reactor Dome Pressure Low-Low Reactor > -46.3" I

2 Water Level indicated level Actions (1) There shall be two (2) operable trip systems for each function.

(a) If the minimum number of operable or tripped instrument channels for one (1) trip system cannot be met, restore the trip system to "r operable status within 14 days or be in at least hot shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(b) If the minimum operability conditions (l.a) cannot be met for both (2) trip systems, be in at least hot shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

  • Vil.LL. 177 Amendment No. 42; -62:-4i 3/4.2-2 3/4.2-29

BASES:

3.2 PROTECTIVE INSTRUMENTATION (Cont)

The recirculation pump trip/alternate rod insertion systems are consistent with the "Monticello RPT/ARI" design described in NEDO-25016 (Reference 1) as referenced by the NRC as an acceptable design (Reference 2) for RPT.

Reference 1 provides both system descriptions and performance analyses. The pump trip is provided to minimize reactor pressure in the highly unlikely event of a plant transient coincident with the failure of all control rods to scram. The rapid flow reduction increases core voiding providing a negative reactivity feedback. High pressure sensors and low water level sensors initiate the trip. The recirculation pump trip is only required at high reactor power levels, where the safety/relief valves have insufficient capacity to relieve the steam which continues to be generated in this unlikely postulated event. Requiring the trip to be operable only when in the RUN mode is therefore conservative. The low water level trip function includes a time delay of nine (9) seconds +/- one (1) second to avoid increasing the consequences of a postulated LOCA. This delay has an insignificant effect on ATWS consequences.-,.

Alternate rod.insertion utlizes the same initiation logic and functions as RPT and provides a diverse means of initiating a reactor scram. ARI uses sensors diverse from the reactor protection system to depressurize the scram pilot air header, which *jln t_ýus e--as_ 01rs tol to be inserted.

References . . ' 'ct o ._..-_._ _.' ._ _ _ _

1. NEDO-25016, "Evaluation of Anticipated Without Scram for the transiens Monticello Nuclear Generating Plant," September 1976.
2. NUREG-0460, Volume 3, December 18.

The drywell temperature limitations of Specification 3.2.H.1 ensure. that safety related equipment will not be subjected to excess temperature.

Exposure to excessive temperatures may degrade equipment and can cause loss of its operability.

The temperature elements for monitoring drywell temperature specified in Table 3.2.H were chosen on the basis of their reliability, location, and their redundancy (dual - element RTD's). These temperature elements are the primary elements used for the PCILRT.

The "nominal instrument elevations" provided in. Tables 3.2.H and 4.2.H assist personnel in locating the instruments for surveillance and maintenance purposes and define the approximate containment region to be monitored. The "nominal instrument elevations" are not intended to provide a precise instrument location.

Revision _tTr/

- f---3nd en 4-No.- 42r-, 59 .-1 96 7-149 B3/4. 1 -6

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 CORE AND CONTAINMENT COOLING 4.5 CORE AND CONTAINMENT COOLING SYSTEMS_ SYSTEMS C. HPCI System C. HPCI System

1. The HPCI system shall be operable whenever there is irradiated fuel in the reactor vessel, A reactor pressure is greater than 150 psig., and reactor coolant temperature is greater than 365°F, except as specified in 3.5.C.2 below.
2. From.and after the date that the HPCI system is made or found to be inoperable for any reason, continued reactor operation is permissible only during the succeeding 14 days unless such system is sooner made operable, providing that during such 14 days all active components of the ADS system, the RCIC system, the.

LPCI system and both core spray systems are operable.

3. If the requirements of 3.5.C cannot be met, an orderly shutdown of the reactor shall be initiated and the reactor shall be in the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Amendment No. +/-7 3/4.5-7 I

INSERT C

-------------- Note --------------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform test.

b. Pump When tested as specified in 3.13, verify with reactor pressure < 1035 Operability and > 940 psig, the HPCI pump can develop a flow rate > 4250 gpm against a system head corresponding to reactor pressure.

INSERT D

--------------- Note --------------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform test.

d. Flow rate at Once / operating cycle, verify with reactor pressure < 150 psig, the 150 psig HPCI pump can develop a flow rate > 4250 gpm against a system head corresponding to reactor pressure.

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 CORE AND CONTAINMENT COOLING 4.5 CORE AND CONTAINMENT COOLING SYSTEMS SYSTEMS D. Reactor Core Isolation Cooling D. Reactor Core Isolation Cooling (RCIC)

(RCIC) System System 0 1. The RCIC system shall be 1. HPCI system testing shall be as operable whenever there is follows:

irradiated fuel in the reactor vessel, reactor pressure is greater than a. Simulated Once/

150 psig, and reactor coolant Automatic Operating temperature is greater than 365-F, Actuation Test Cycle except as specified in 3.5.D1 below. 6 , I'it._

2. From and after the date that the RCIC system is made or found to be inoperable for any reason, continued reactor operation is permissible only during the '

succeeding 14 days unless such system is sooner made operable, providing that during such 14 days the HPCIS is operable.

3. If the requirements of 3.5.D cannot be met, an orderly shutdown of the
c. Motor Operated As Specified in 3.13

{

reactor shall be initiated and the Valve reactor shall be in the Cold Operability Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4W Amendment No. "1134 3/4.5-8

INSERT E Note --------------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform test.

b. Pump When tested as specified in 3.13, verify with reactor pressure < 1035 Operability and > 940 psig, the RCIC pump can develop a flow rate > 400 gpm against a system head corresponding to reactor pressure.

INSERT F Note --------------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform test.

d. Flow rate at Once / operating cycle, verify with reactor pressure < 150 psig, the 150 psig RCIC pump can develop a flow rate > 400 gpm against a system head corresponding to reactor pressure.

ADS System B 3/4.5.E B 3/4.5 CORE AND CONTAINMENT COOLING SYSTEMS 314.5.E. Automatic Depressurization (ADS) System 0

BACKGROUND This specification ensures the operability of the ADS under all conditions for which the automatic or manual depressurzatlon of the nuclear system is an essential response to station abnormalities.

The nuclear system pressure relief system provides automatic nuclear system depressurization for small breaks in the nuclear system so that the low pressure coolant injection (LPCI) and the core spray systems can operate to protect the fuel barrier.

Because the Automatic Depressurization System does not provide makeup to the reactor primary vessel, no credit is taken for the steam cooling of the core caused by the system actuation to provide further conservatism to the CSCS. Performance analysis of the Automatic Depressurzation System is considered only with respect to its depressurizing effect in conjunction with LPCI or CoreSpx. There are four valves provid.@d and each has a capacity of 0 b/hr at a reactor pressure ote-44psig. .

APPLICABLE The limiting conditions for operating the ADS are derived from the SAFETY ANALYSIS Station Nuclear Safety Operational Analysis (FSAR Appendix G) and a detailed functional analysis of the ADS (FSAR Section 6).

ACTIONS The allowable out of service time for one ADS valve is determined as 14 days because of the redundancy and because of HPCI operability; therefore, redundant protection for the core with a small break In the

'nuclear system is still available.

SURVEILLANCES The testing interval for the core and containment cooling systems Is based on industry practice, quantitative reliability analysis, judgment and practicality. The core cooling systems have not been designed to be fully testable during operation. For example, complete ADS testing during power operation causes an undesirable loss-of-coolant inventory. When components are tested and found inoperable the impact on system operability is determined, and corrective action or Limiting Conditions of Operation are initiated. A simulated automatic actuation test once each cycle combined with code inservice testing of the pumps and valves is deemed to be adequate testing of these systems. The ADS test circuit permits continued surveillance on the operable relief valves to assure that they will be available if required.

The surveillance requirements provide adequate assurance that the core and containment cooling systems will be operable when required.

Revision 202,-269-W B3/4.5-21

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENtS 3.6 PRIMARY SYSTEM BOUNDARY (Cont) 4.6 PRIMARY SYSTEM BOUNDARY (Cont)

c. With no required leakage 6Y detection systems Operable, be in Hot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Cold Shutdown within the following 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 D. Safety and Relief Valves

1. During reactor power operating *Testing of sa ety and relief/safety conditions and prior to reactor startup valv shall be in accordance with from a Cold Condition, or whenever reactor coolant pressure is greater than 104 psig and temperature 2. At least o of the relie safety valves greater than 340 0 F, both safety shall be dis semble and inspected valves and the safety modes of 11 each refuelin outa e.
3. Whenever the saty relief valves are required to be per le, the discharge pi e temp ture of each valves shall set

\setpoint +/-

be t this nominal si Tl safety valves daily./ reli valve shall e logged safety shall be~set at 1240 psi 13 psi.

4. Instr entation shall be cali rated
2. If Specification 3.6.D.1 is not met, an and ecked as indicated in Table orderly shutdown shall be initiated 4.2.F.

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 />.

7' NOTE Tech"'mcl Specifications 3.W.3--

3.6.D.5 a)DIv to the two St6ae Taraet Amendment No. 12, 56, 88, 133, 139, 1149, 222 ,2 3/4.6-6

INSERT G

1. As specified in accordance with 3.13, verify the safety function lift setpoints of the safety and relief valves as follows:

Number of Safety and Relief Valves Setpoint (psig) 2 Safety 1280+38.4 4 Relief 1155 + 34.6 Following testing, lift settings shall be within + 1%.

INSERT H Note Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform test.

2. Once / operating cycle, verify each relief valve opens when manually actuated.

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6 PRIMARY SYSTEM BOUNDARY (Cont) 4.6 PRIMARY SYSTEM BOUNDARY (Cont) 5 pipe tem rature excee( rt12iF for 24 S hours or mnr shall b femoved at the nextucoldws,, f 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more, tested in tshe as- und condition, and eclirted f3nessary prior to I reinstallat .. -_

5. h tigcniions cfoperation for the nstrumentation that m itors tail pipe temperature are given in Tab e 3.2-F.

E. Jet Pumps E. Jet Pumps

1. Whenever the reactor is in the Startup or NOTES Run Modes, all jet pumps shall be 1. 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 /> Operable. If it is determined that a jet after the associated recirculation loop is in pump is inoperable, the reactor shall be operation.

in Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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.

Whenever there is recirculation flow with the reactor in the Startup or Run Modes, jet pump operability shall be checked daily by verifying at least one of the following criteria (1, 2, or 3) is satisfied for each operating recirculation loop:

1. Recirculation pump flow to speed ,ratio differs by s 5% from established patterns, and jet pump loop flow to recirculation pump speed ratio differs by < 5% from established patterns.
2. Each jet pump diffuser to lower plenum differential pressure differs by 5 20% from established patterns.
3. Each jet pump flow differs by < 10% from established patterns.

Amendment No. 45, 66, 71, 03, 133, 205, 219,-U, t=." 3/4.6-7

BASES:

3/4.6 PRIMARY SYSTEM BOUNDARY- (Cont)

C. Coolant Leakage (Cont)

The 2 gpm limit for unidentified coolant leakage rate increase within any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period is a limit specified by the NRC in Generic Letter 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping." This limit applies only during the RUN mode to accommodate the expected coolant leakage increase during pressurization.

The total leakage rate consists of all leakage, which flows to the drywell equipment drain sump (Identified leakage) and floor drain sump (Unidentified leakage).

In addition to the sump monitoring of coolant leakage, airborne radioactivity levels of the drywel atmosphere is monitored by the Reactor Pressure Boundary Leak Detection System. This system consists of two panels capable of monitoring the primary containment atmosphere for particulate and gaseous radioactivity as a result of coolant leaks.

D. Safety and Relief Valves ,r The valve sizing analysis con 'dered fourLj' elief/safety valves and two 8rooS

-- safety valves. Th pressures are established in accordance with the following three requirements of Section Ill of the ASME Code:

I

1. The lowest safety valve must be set to open at or below vessel design pressure and the highest safety valve be set at or belo 10 f design pressure.
2. The valves must limit the reactor$ressure to no more than 110% of design pressure. I
3. Protection systems directly related to the valve sizing transient must not be credited with action (i.e., an indirect scram must be assumed).

Revision 155, 177, 240, 269, 27*

j2 B3/4.6-7

BASES:

3/4.6 PRIMARY SYSTEM BOUNDARY (Cont)

D. Safety 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.

Relievin c'pacity of 6 relief/safety valvein c wsafety

'combination with aey.

valveKes'ults in a pea-pressure during the transient conditions used in tMe safety valve sizing analysis which is well below the pressure safety limit.

The relief/safety valve settings-satisfy the Code requirements that the lowest safe alve set point be at or below the vessel design pressure range to prevent unnecessa cycling caused by minor transients. The results of postulated transients wher .nher relief/safety valve actuation is required are th inal Safety Analysis Report. '

,'*7 Experience in safety valve operation shows that a testing of at least 50% of the safety

({,*/0 valves r refueling outage is adequate to detect failures or deterioration. The tolerance

-* uee of/+ti is in accordance with Section III of the ASME Boiler and Pressure Vesel,,")

Code. An a alysis has been performed which shows that with all safety valves se higher, the reactor coolantpressure safety limit of 1375 psig is not exceeded.

The relief/safety valves h'-eTw"l ions; 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 In addition, the valves can be operated manually.

Revision 41-46, 41-74, 269, 73, B3/4.6-8 I

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