Similar Documents at Hatch |
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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20217D3061999-10-13013 October 1999 SER Accepting Licensee Proposed Changes to Edwin I Hatch Nuclear Plant Emergency Classification Scheme to Add Emergency Action Levels Related to Operation of Independent Spent Fuel Storage Installation ML20212A6641999-09-13013 September 1999 Safety Evaluation Authorizing Relief Request RR-V-16 for Third 10 Yr Interval Inservice Testing Program ML20210J9631999-08-0202 August 1999 SER Finding That Licensee Established Acceptable Program to Verify Periodically design-basis Capability of safety-related MOVs at Edwin I Hatch Nuclear Plant,Units 1 & 2 ML20210J9271999-08-0202 August 1999 SER Finds That Licensee Performed Appropriate Evaluations of Operational Configurations of safety-related power-operated Gate Valves to Identify Valves at Plant,Susceptible to Pressure Locking or Thermal Binding ML20207E7631999-06-0303 June 1999 Safety Evaluation Concluding That Licensee Proposed Alternative to Use Code Case N-509 Contained in RR-4 Provides Acceptable Level of Quality & Safety.Considers Rev 2 to RR-4 & RR-6 Acceptable ML20206G1691999-05-0404 May 1999 SER Approving Requirements of Istb 4.6.2(b) Pursuant to 10CFR50.55a(a)(3)(ii) ML20207M1891999-03-11011 March 1999 SER Accepting Relief Request for Authorization of Alternative Reactor Pressure Vessel Exam for Circumferential Weld ML20196J4931998-12-0707 December 1998 Safety Evaluation Accepting Proposed Alternatives in Relief Requests RR-V-12,RR-V-15,RR-P-15,RR-V-7,RR-V-12,RR-V-14 & RR-V-15 ML20153G2481998-09-24024 September 1998 SE Concluding That Licensee Implementation Program to Resolve USI A-46 at Plant Adequately Addressed Purpose of 10CFR50.54(f) Request ML20239A2531998-09-0303 September 1998 SER Accepting Licensee Request for Relief Numbers RR-17 & RR-18 for Edwin I Hatch Nuclear Plant,Units 1 & 2.Technical Ltr Rept on Third 10-year Interval ISI Request for Reliefs for Plant,Units 1 & 2 Encl ML20236W3441998-07-30030 July 1998 Safety Evaluation Accepting Relief Requests for Second 10-yr ISI for Plant,Units 1 & 2 ML20236V5191998-07-28028 July 1998 Safety Evaluation Accepting Proposed License Amend Power Uprate Review ML20236L1821998-07-0707 July 1998 Safety Evaluation Accepting 980428 Proposed Alternative to ASME Boiler & Pressure Vessel Code,Section Xi,Repair & Replacement Requirements Under 10CFR50.55a(a)(3) ML20212A1981997-10-16016 October 1997 Safety Evaluation Denying Licensee Request for Relief from Implementation of 10CFR50.55a Requirements Re Use of 1992 Edition of ASME Code Section XI for ISI of Containments ML20216J8971997-09-12012 September 1997 SER Related to General Electric Nuclear Measurement Analysis & Control Power Range Neutron Monitoring Sys Upgrade Southern Nuclear Operating Co,Units 1 & 2 ML20216E9671997-09-0505 September 1997 Safety Evaluation Accepting ,As Suppl by 970902 Request for Relief to Request RR-V-11 Re IST & S/Rv ML20210S9141997-09-0303 September 1997 Safety Evaluation Accepting Licensee Request for one-time Relief from GL 88-01 for Insp of Category E Welds at Plant, Unit 1 & 2 ML20217N9381997-08-21021 August 1997 SE Re New & Revised Relief Requests Submitted by 970130,0307 & 25 Ltrs in Relation to Third 10-yr Pump & Valve IST Program ML20217N9811997-08-21021 August 1997 Safety Evaluation for Third 10-year Pump & Valve Inservice Testing Program,Southern Nuclear Operating Co,Inc,Hatch, Units 1 & 2 ML20148U6141997-07-0707 July 1997 Safety Evaluation Accepting Licensee Proposal for Third 10-yr Interval for Pump & Valve Inservice Testing Program ML20141A1981997-06-17017 June 1997 Safety Evaluation Accepting Licensee Design Criteria for Sizing ECCS Suction Strainers ML20141A1431997-06-16016 June 1997 Safety Evaluation Accepting Third 10-yr Inservice Insp Program Plan & Associated Requests for Relief.Relief Not Required for RR-08 ML20137N1811997-04-0404 April 1997 Safety Evaluation Supporting Amends 206 & 147 to Licenses DPR-57 & NPF-5,respectively ML20134P3661997-02-21021 February 1997 SER Accepting Test & Technical Evaluations Performed for Reactor Vessel Shell Welds,Per 10CFR50.55a(g)(6)(ii)(A)(5) ML20138J5211997-02-0505 February 1997 Safety Evaluation Accepting Temporary Request for Relief from ASME Code Repair Requirements for ASME Code Class 3 Valve ML20134B3301997-01-28028 January 1997 SE Accepting Revised QA Program for Plant ML20129F8211996-10-24024 October 1996 Safety Evaluation Accepting Licensee Actions IAW Current Industry Practice & BWRVIP Guidelines for Reinspection of BWR Core Shrouds ML20059E6961993-10-21021 October 1993 Safety Evaluation Supporting Amends 190 & 129 to Licenses DPR-57 & NPF-5,respectively ML20128C1931992-11-20020 November 1992 Safety Evaluation Accepting Licensee Response to Suppl 1 to GL 87-02 ML20127L4511992-11-18018 November 1992 Safety Evaluation Accepting Justification to Cancel Commitment on Seven Human Engineering Discrepancies ML20248F9791989-09-20020 September 1989 Safety Evaluation Accepting Okonite Taped Cable Splice as Electrical Connection to Replace Terminal Blocks in Selected Low Voltage Transmitter Measuring Loops ML20247H7261989-03-16016 March 1989 Safety Evaluation Re Use of Radioiodine Protection Factor for Sorbent Canisters ML20207M0431988-10-13013 October 1988 Safety Evaluation Denying Util 880711 Request for Relief from Hydrostatic Test Requirements of Section XI of ASME Code for Class 2 Portion of Main Steam Lines Between Outboard MSIVs & Turbine Stop Valves ML20153F9941988-05-0202 May 1988 Safety Evaluation Supporting Amend 153 to License DPR-57 ML20238A6801987-09-0404 September 1987 Safety Evaluation Re Insps & Repairs of Igscc.Plant Can Be Safely Operated for Another 18-month Fuel Cycle in Present Configuration ML20236F9831987-07-29029 July 1987 Safety Evaluation Supporting Util 831107,840229 & 860821 Responses to Generic Ltr 83-28,Items 3.1.1,3.1.2,3.2.1 & 3.2.2 ML20235X5271987-07-20020 July 1987 SER Supporting Util Response to Generic Ltr 83-28,Item 2.1, (Part 2) Re Vendor Interface Programs (Reactor Trip Sys Components) ML20235P8421987-07-14014 July 1987 Safety Evaluation Re Acceptance of Offsite Dose Calculation Manual as Updated & Corrected Through 861231 ML20215M3941987-06-22022 June 1987 Safety Evaluation Re Request for Relief from Inservice Insp Requirements ML20236F6151987-04-0101 April 1987 Safety Evaluation Re Analytical Method Used by Licensee to Evaluate Critical Stresses Re Mark I Containment Program Vacuum Breakers Adequate.Max Stress in Breakers Less than 30% of Code Allowable.Existing Design Structually Adequate ML20207U1441987-03-19019 March 1987 Undated Safety Evaluation Re Plant.Section 9, Radwaste Sys, of FSAR Also Encl ML20210S2731986-09-29029 September 1986 Safety Evaluation Re Inservice Insp Program & Requests for Relief ML20211F0241986-06-12012 June 1986 Safety Evaluation Supporting Util Listed Responses to Generic Ltr 83-28,Item 2.1 (Part 1) Re Identification & Classification of Reactor Trip Sys Components ML20211B3201986-05-30030 May 1986 SER Accepting Licensee 831107 & 840219 Responses to Generic Ltr 83-28, Items 3.1.3 & 3.2.3 Re post-maint Testing Requirements ML20205M9461986-04-24024 April 1986 Safety Evaluation Supporting Plant Operation in Present Configuration for 18-month Fuel Cycle.Plans for Insp &/Or Mod of Svc Sensitive Austenitic Stainless Steel Piping Sys Requested 3 Months Before Start of Next Refueling Outage ML20151Y4631986-01-29029 January 1986 Safety Evaluation Supporting Amend 122 to License DPR-57 ML20137M5311986-01-21021 January 1986 SER Supporting 850718 & 1127 Requests for Reconsideration of Relief from Requirements of Section XI of ASME Code Re Exam of Supports on ASME Piping ML20141F1261985-12-26026 December 1985 Safety Evaluation Supporting Amends 120 & 59 to Licenses DPR-57 & NPF-5,respectively ML20136A8461985-12-23023 December 1985 Safety Evaluation Re Responses to Generic Ltr 83-28,Items 3.1.1,3.1.2,3.2.1,3.2.2 & 4.5.1.Addl Info Requested on Items 3.1.1,3.1.2,3.2.1 & 3.2.2.Item 4.5.1 Acceptable ML20137E2191985-12-23023 December 1985 Safety Evaluation Re Util Response to Generic Ltr 83-28,Item 1.1, Post-Trip Review (Program Description & Procedure). Program & Procedures Acceptable 1999-09-13
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217D3061999-10-13013 October 1999 SER Accepting Licensee Proposed Changes to Edwin I Hatch Nuclear Plant Emergency Classification Scheme to Add Emergency Action Levels Related to Operation of Independent Spent Fuel Storage Installation HL-5845, Monthly Operating Repts for Sept 1999 for Ei Hatch Nuclear Plant.With1999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Ei Hatch Nuclear Plant.With ML20217A1691999-09-22022 September 1999 Part 21 Rept Re Engine Sys,Inc Controllers,Manufactured Between Dec 1997 & May 1999,that May Have Questionable Soldering Workmanship.Caused by Inadequate Personnel Training.Sent Rept to All Nuclear Customers ML20212A6641999-09-13013 September 1999 Safety Evaluation Authorizing Relief Request RR-V-16 for Third 10 Yr Interval Inservice Testing Program HL-5836, Monthly Operating Repts for Aug 1999 for Edwin I Hatch Nuclear Plant.With1999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Edwin I Hatch Nuclear Plant.With ML20210J9631999-08-0202 August 1999 SER Finding That Licensee Established Acceptable Program to Verify Periodically design-basis Capability of safety-related MOVs at Edwin I Hatch Nuclear Plant,Units 1 & 2 ML20210J9271999-08-0202 August 1999 SER Finds That Licensee Performed Appropriate Evaluations of Operational Configurations of safety-related power-operated Gate Valves to Identify Valves at Plant,Susceptible to Pressure Locking or Thermal Binding HL-5818, Monthly Operating Repts for July 1999 for Ei Hatch Nuclear Plant,Units 1 & 2.With1999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Ei Hatch Nuclear Plant,Units 1 & 2.With HL-5805, Monthly Operating Repts for June 1999 for Ei Hatch Nuclear Plant,Units 1 & 2.With1999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Ei Hatch Nuclear Plant,Units 1 & 2.With ML20207E7631999-06-0303 June 1999 Safety Evaluation Concluding That Licensee Proposed Alternative to Use Code Case N-509 Contained in RR-4 Provides Acceptable Level of Quality & Safety.Considers Rev 2 to RR-4 & RR-6 Acceptable HL-5795, Monthly Operating Repts for May 1999 for Ehnp Units 1 & 2. with1999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Ehnp Units 1 & 2. with ML20206G1691999-05-0404 May 1999 SER Approving Requirements of Istb 4.6.2(b) Pursuant to 10CFR50.55a(a)(3)(ii) HL-5784, Monthly Operating Repts for Apr 1999 for Ei Hatch Nuclear Plant,Units 1 & 2.With1999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Ei Hatch Nuclear Plant,Units 1 & 2.With HL-5766, Monthly Operating Repts for Mar 1999 for Ei Hatch Nuclear Plant,Units 1 & 2.With1999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Ei Hatch Nuclear Plant,Units 1 & 2.With ML20207M1891999-03-11011 March 1999 SER Accepting Relief Request for Authorization of Alternative Reactor Pressure Vessel Exam for Circumferential Weld HL-5755, Monthly Operating Repts for Feb 1999 for Ei Hatch Nuclear Plant,Units 1 & 2.With1999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Ei Hatch Nuclear Plant,Units 1 & 2.With ML20206P6981999-01-0707 January 1999 Ehnp Intake Structure Licensing Rept HL-5726, Monthly Operating Repts for Dec 1998 for Ei Hatch Nuclear Plant,Units 1 & 2.With1998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Ei Hatch Nuclear Plant,Units 1 & 2.With ML20196J4931998-12-0707 December 1998 Safety Evaluation Accepting Proposed Alternatives in Relief Requests RR-V-12,RR-V-15,RR-P-15,RR-V-7,RR-V-12,RR-V-14 & RR-V-15 HL-5714, Monthly Operating Repts for Nov 1998 for Ei Hatch Nuclear Plant,Units 1 & 2.With1998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Ei Hatch Nuclear Plant,Units 1 & 2.With HL-5706, Monthly Operating Repts for Oct 1998 for Hatch Nuclear Plant Units 1 & 2.With1998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Hatch Nuclear Plant Units 1 & 2.With ML20155B6121998-10-28028 October 1998 Safety Evaluation of TR SNCH-9501, BWR Steady State & Transient Analysis Methods Benchmarking Topical Rept. Rept Acceptable HL-5691, Monthly Operating Repts for Sept 1998 for Ei Hatch Nuclear Plant,Units 1 & 2.With1998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Ei Hatch Nuclear Plant,Units 1 & 2.With ML20153G2481998-09-24024 September 1998 SE Concluding That Licensee Implementation Program to Resolve USI A-46 at Plant Adequately Addressed Purpose of 10CFR50.54(f) Request ML20239A2531998-09-0303 September 1998 SER Accepting Licensee Request for Relief Numbers RR-17 & RR-18 for Edwin I Hatch Nuclear Plant,Units 1 & 2.Technical Ltr Rept on Third 10-year Interval ISI Request for Reliefs for Plant,Units 1 & 2 Encl HL-5675, Monthly Operating Repts for Aug 1998 for Ei Hatch Nuclear Plant,Units 1 & 21998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Ei Hatch Nuclear Plant,Units 1 & 2 ML20238F7131998-08-31031 August 1998 9,change 2 to QAP 1.0, Organization HL-5667, Monthly Operating Repts for July 1998 for Ei Hatch Nuclear Plant,Units 1 & 21998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Ei Hatch Nuclear Plant,Units 1 & 2 HL-5657, Ro:On 980626,noted That Pami Channels Had Been Inoperable for More than Thirty Days.Cause Indeterminate.Licensee Will Replace Automatic Function W/Five Other Qualified Pamis of Like Kind in Drywell & Revised Procedures1998-07-30030 July 1998 Ro:On 980626,noted That Pami Channels Had Been Inoperable for More than Thirty Days.Cause Indeterminate.Licensee Will Replace Automatic Function W/Five Other Qualified Pamis of Like Kind in Drywell & Revised Procedures ML20236W3441998-07-30030 July 1998 Safety Evaluation Accepting Relief Requests for Second 10-yr ISI for Plant,Units 1 & 2 ML20236V5191998-07-28028 July 1998 Safety Evaluation Accepting Proposed License Amend Power Uprate Review ML20236N6751998-07-0909 July 1998 Part 21 & Deficiency Rept Re Notification of Potential Safety Hazard from Breakage of Cast Iron Suction Heads in Apkd Type Pumps.Caused by Migration of Suction Head Journal Sleeve Along Lower End of Pump Shaft.Will Inspect Pumps ML20236L1821998-07-0707 July 1998 Safety Evaluation Accepting 980428 Proposed Alternative to ASME Boiler & Pressure Vessel Code,Section Xi,Repair & Replacement Requirements Under 10CFR50.55a(a)(3) HL-5653, Monthly Operating Repts for June 1998 for Ei Hatch Nuclear Plant,Units 1 & 21998-06-30030 June 1998 Monthly Operating Repts for June 1998 for Ei Hatch Nuclear Plant,Units 1 & 2 HL-5640, Monthly Operating Repts for May 1998 for Ei Hatch Nuclear Plant,Units 1 & 21998-05-31031 May 1998 Monthly Operating Repts for May 1998 for Ei Hatch Nuclear Plant,Units 1 & 2 ML20248B8651998-05-15015 May 1998 Quadrennial Simulator Certification Rept HL-5628, Monthly Operating Repts for Apr 1998 for Ei Hatch Nuclear Plant1998-04-30030 April 1998 Monthly Operating Repts for Apr 1998 for Ei Hatch Nuclear Plant HL-5604, Monthly Operating Repts for Mar 1998 for Edwin I Hatch Nuclear Plant,Units 1 & 21998-03-31031 March 1998 Monthly Operating Repts for Mar 1998 for Edwin I Hatch Nuclear Plant,Units 1 & 2 ML20216B2711998-02-28028 February 1998 Extended Power Uprate Safety Analysis Rept for Ei Hatch Plant,Units 1 & 2 HL-5585, Monthly Operating Repts for Feb 1998 for Ei Hatch Nuclear Plant,Units 1 & 21998-02-28028 February 1998 Monthly Operating Repts for Feb 1998 for Ei Hatch Nuclear Plant,Units 1 & 2 HL-5571, Monthly Operating Repts for Jan 1998 for Edwin I Hatch Nuclear Plant,Unit 11998-01-31031 January 1998 Monthly Operating Repts for Jan 1998 for Edwin I Hatch Nuclear Plant,Unit 1 HL-5551, Monthly Operating Repts for Dec 1997 for Ei Hatch Nuclear Plant,Units 1 & 21997-12-31031 December 1997 Monthly Operating Repts for Dec 1997 for Ei Hatch Nuclear Plant,Units 1 & 2 ML20199B0561997-12-31031 December 1997 Rev 0 GE-NE-B13-01869-122, Jet Pump Riser Weld Flaw Evaluation Handbook for Hatch Unit 1 HL-5581, Annual Operating Rept for 1997, for Ei Hatch Nuclear Plant Units 1 & 21997-12-31031 December 1997 Annual Operating Rept for 1997, for Ei Hatch Nuclear Plant Units 1 & 2 HL-5533, Monthly Operating Repts for Nov 1997 for Ei Hatch Nuclear Plant,Units 1 & 21997-11-30030 November 1997 Monthly Operating Repts for Nov 1997 for Ei Hatch Nuclear Plant,Units 1 & 2 HL-5514, Monthly Operating Repts for Oct 1997 for Edwin I Hatch Nuclear Plant,Units 1 & 21997-10-31031 October 1997 Monthly Operating Repts for Oct 1997 for Edwin I Hatch Nuclear Plant,Units 1 & 2 ML20212A1981997-10-16016 October 1997 Safety Evaluation Denying Licensee Request for Relief from Implementation of 10CFR50.55a Requirements Re Use of 1992 Edition of ASME Code Section XI for ISI of Containments ML20211M6491997-10-0808 October 1997 Addenda 1 to Part 21 Rept Re Weldments on Opposed Piston & Coltec-Pielstick Emergency stand-by Diesel gen-set lube-oil & Jacket Water Piping Sys.Revised List of Potentially Affected Utils to Include Asterisked Utils,Submitted ML20211H5311997-10-0101 October 1997 Rev 2 to Unit 1,Cycle 17 Colr ML20211H5251997-10-0101 October 1997 Rev 3 to Unit 1,Cycle 17 Colr 1999-09-30
[Table view] |
Text
g- ,
[j#"%g4-UNITED STATES s* j NUCLEAR REGULATORY COMMISSION ,
\...../ ,
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
- l. RELATED TO THE THIRD TEN-YEAR INSERVICE TESTING INTERVAL PROGRAM SOUTHERN NUCLEAR OPERATING COMPANY. INC.
EDWIN 1. H' ATCH NUCLEAR PLANT UNITS 1 AND 2 ,
DOCKET NOS. 50-321 AND 50-366 !
1.0 INTRODUCTION
Title 10 of the Code of Federal Reaulations (10 CFR) Section 50.55a, requires that inservice testing (IST) of certain American Society of Mechanical Engineers (ASME) Code Class 1,2, and 3 pumps and valves be performed in accordance with Section XI of the ASME Boiler and
- Pressure Vessel Code (the Code) and applicable addenda, except where alternatives have been authorized or relief has been requested by the licensee and granted by the Commission pursuant to paragraphs (a)(3)(i), (a)(3)(ii), or (f)(6)(i) of 10 CFR 50.55a. In proposing alternatives or requesting relief, tne licensee must demonstrate that: (1) the proposed alternatives provide an acceptable level of quality and safety; (2) compliance would result in hardship c. unusual difficulty without a compensating increase in the level of quality and safety; or (3) conformance is impractical for its facility. Section 50.55a authorizes the Commission to approve alternatives ar:d to grant relief from ASME Code requirements upon making the necessary findings.- NRC guidance contained in Generic Letter (GL) 60 04, " Guidance on Developing Acceptable Inservice Testing Programs," provides alternatives to the Code requirements determined ecceptable to the staff. Further guidance was given in GL 89-04, Supplement 1, and NUREG-1482, " Guidelines for Inservice Testing at Nuclear Power Plants."
The third 10-year interval program for IST of pumps and valves at Hatch, Units 1 and 2, was submitted in a letter dated September 15,1995. The third 10-year interval for both units began on January 1,1996, and will end on December 31,2006. ' In accordance with the proposed alternative of Relief Request (RR) RR-G-1, which was approved in a :etter dated August 29, 1995, the applicable Codes used in the Hatch IST program are the ASME Operation and Maintenance (OM) Code-1990 for pumps and valves, with the exception of relief valves. The applicable Code for relief valves is the ASME OM Code-1995.
RR P-14 was initially submitted in a letter dated March 25,1397. The staff's safety evaluation (SE) dated August 21,1997, authorized the use of the proposed alternative testing method for an interim period of 120 days. RR-P-14 was subsequently revised and resubmitted to the NRC in a letter dated March 16,1999. The NRC's findings with respect to authorizing alternatives and granting or denying the proposed revised relief request are gben below.
~
9905070108 990504 PDR ADOCK 05000321 P PDR
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2.0 RELIEF REQUEST RR-P-14 (2"d Submittal)
The licensee proposes an alternative to the differential pressure measurement requirements of ISTB 4.6.2(b) for the four residual heat removal pumps (RHR) and two core spray (CS) pumps in each unit. The licensee proposes to measure the discharge pressure and calculate the !
differential pressure by assuming a constant' suction pressure.
2.1 Licensee's Basis for Requesting Relief The licensee states:
The RHR and CS pumps are aligned to the suppression pool (torus) during all modes of normal plant operation which results in a virtually constant suction pressure. IST is performed utilizing a full flow test line which circulates water to and from the suppression pool. The Plant's Technical Specifications require that the suppression pool be maintained within a narrow range of level, temperature, and internal pressure during plant operation which results in a suction pressure l of approximately 5 psig. The Technical Specification operability limits for the l suppression pool are itemized below, j Unit 1/ Unit 2 l 1
Level 2 146" & s 150" l internal Pressure s 1.75 psig Water Temperature s100'F These Technical Specification operability limits for the suppression pool result in a maximum
- difference in calculated pump suction pressure of < 2 psig. This 2 psig maximum difference is insignificant when performing IST considering the normal discharge pressure of the RHR and CS pumps (see table below). This 2 psig variance is also insignificant in the calculation of differential pressure (AP = Po-P) when considering the Code acceptable operating range (i.e.,
95110% for vertical line shaft pumps and 90-110% for centrifugal pumps) from Table ISTB 5.2-V 2b and the allowable i2% instrument accuracy. Therefore, measurement of differential I
pressure provides no added benefit for determining pump operational readiness or for monitoring pump degradation.
Reference Discharge Maximum Pump Pressure Variance Unit 1 RHR 180 - 193 psig 1.11% max.
Unit 1 CS 305 - 310 psig 0.66% max.
Unit 2 RHR 172 - 190 psig 1.16% max.
Unit 2 CS 285 - 290 psig 0.70% max.
The following table summarizes several years worth of pump IST data. This summary confirms that the RHR and Core Spray pumps' suction pressures are consistent and are relatively insignificant in comparison with the pumps' discharge pressures. Applying an average suction
i
., . I i
I pressure of 5 psig, when calculating differential pressure, will provide data that is meaningful for l assessing operational readiness and for monitoring pump degradation.
PUMP MPL MIN. MAX. AVG. REMARKS No. PRES. PRES. PRES i 1E11-C002A 3.9 3.8 5.1 (52) Or=8000 gpm, Apr=116 psid 1E11-C002B 3.2 6.25 4.8 (47) Or=7700 gpm, Apr=185 psid 1E11-C002C 3.0 6.2 4.8(46) Or=7700 gpm, Apr=176 psid 1E11-C002D 3.4 6.0 4.6 (40) Or=7700 gpm, Apr=183 psid 1E21-C001 A 2.5 5.8 4.1 (68) Or=4625 gpm, Apr=289 psid 1E21-C001B 1.7
- 5.9 3.7 (47) Or=4625 gpm, Apr=282 psid 2E11-C002A 3.0 6.8 5.2 (50) Or=8000 gom, Apr=187 psid 2E11-C002B 4.3 7.1 5.3 (48) Or=7800 gpm, Apr=180 psid 2E11-C002C 3.0 6.9 5.3 (55) Or=7900 gpm, Apr=182 psid 2E11-C002D 3.8 6.2 4.9 (47) Or=7700 gpm, Apr=175 psid 2E21-C001 A 4.15 6.9 5.1 (43) Or=4750 gpm, Apr=302 psid 2E21-C001B 3.3 6.4 5.0 (53) Or=4750 gpm, Apr=303 psid AVERAGE 3.3 6.4 4.9 N/A Number in parenthesis "()" indicates the number of test valves averaged to get indicated value.
- One time occurrence only.
Additionally, a test gage is required to be installed to perform IST of each pump. The permanently installed pump suction pressure gages encompass a wider range of pressures than does IST and thus exceed the OM Code allowable range limit (3 times the reference value). The installed RHR pump gages must account for the pressure experienced with the RHR loop in the shutdown cooling mode of operation. The installed CS pump gages must account for the pressure experienced with the CS suction aligned to the Condensate Storage Tank. Therefore, a test gage, which satisfies the Code range limits, must be temporarily installed each time that IST is required.
Applying a constant pump suction pressure, when calculating differential pressure, will allow the IST to be performed with the installed pressure gages thus lessening the burden on operations personnel responsible for the testing. Since test gages are required to be calibrated both prior to and after usage, it also eliminates the possibility of invalidating test data due to a gage being damaged during transportation, installation or removal.
Mechanical degradation of centrifugal pumps, which experience significant differences in suction (inlet) pressure, would be indicated by changes in the differential pressure. However, for these pumps, the suction pressure variance is insignificant in comparison to the developed
4 head (pressure). Therefore, monitoring discharge pressure and calculating differential pressure assuming a constant 5 psig suction pressure provides an adequate method to determine operational readiness and detect potential degradation.
2.2 Alternative Testing The licensee proposes:
Pump suction pressure will be assumed to be 5 psig based on a review of several years of IST data which support suction pressure being virtually constant. During IST, pump differential pressure will be calculated by measuring pump discharge pressure and subtracting 5 psig. This value will then be compared to the corresponding reference value. The acceptance criteria of Table ISTB 5.2-2b will be applied for assessing pump operational readiness and for monitoring potential pump degradation. This testing method meets the intent .
of the Code for monitoring pump operationa! readiness and degradation, and j relieves the Licensee the burden associated with the use of temporary test l
gages. Therefore, the proposed alternative is requested pursuant to 10 CFR l 50.55a(a)(3)(ii).
2.3 Evaluation 1 l
The licensee requests relief from the requirements of ISTB 4.6.2(b) for the residual heat removal pumps,1E11-C002A, B, C & D and 2E11-C002A, B, C & D, and the core sprey pumps, 1E21-C001 A & B and 2E21-C001 A & B. If a direct indicating instrument is not provided, this section of the Code requires that differential pressure be determined by the difference between the pressure at a point in the inlet pipe and the pressure at a point in the discharge pipe. The licensee proposes to measure the discharge pressure and calculate the differential pressure by assuming a constant suction pressure of 5 psig.
The range of the permanently installed pressure gauges at the pumps' inlet exceed the OM Code allowable range limit (3 times the reference value), and so temporary gauges would need to be installed for each test. Accordingly, these temporarily installed gauges would need to be calibrated both prior to and after usage. These extra steps which are necessary for !
compliance with the requirements of ISTB 4.6.2(b) create a hardship for the licensee without a compensating increase in the level of quality and safety.
Discharge pressure can be used in lieu of differential pressure for evaluating pump hydraulic performance if variations in pump inlet pressure are small. NUREG/CR-6396, Section 3.3.2 provides items to consider for justifying the use of discharge pressure instead of differential pressure, it includes:
(1) The inlet pressure'is small in comparison with the discharge pressure (maximum deviation of 2%).
(2) The maximum expected variation in the inlet pressure from test to test is relatively small as determined by control procedures and technical specification limits and as verified by historical data.
P
.. . L (3) The Code required acceptance criteria are not relaxed.
(4) Everithough some uncertainty is introduced by this method, applying the Code -
' acceptance criteria for differential pressure for this application adds conservatism.
'(5)~. If a significant blockage occurs at the pump suction, this condition would affect the discharge pressure and/or flow measurement and would not go undetected.
The licensee's submittal meets all'the above criteria. The proposed alternative testing method provides an acceptable means of evaluating pump performance without causing a significant decrease in the ability to monitor operational readiness.
3.0 CONCLUSION
The proposed alternative to the requirements of ISTB 4.6.2(b) is authorized pursuant to 10 CFR 50.55a(a)(3)(ii). Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Principal Reviewer: M. Kotzalas Date: May 4, 1999 i
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