Safety Evaluation Supporting Demonstration of Containment Purge & Vent Valve Operability.Info Submitted Demonstrates Ability of 20-inch & 24-inch Purge & Vent Valves to Close Against Buildup of Containment Pressure During Dba/LocaML20129E141 |
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FitzPatrick |
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07/01/1985 |
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Office of Nuclear Reactor Regulation |
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Shared Package |
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ML20129E111 |
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NUDOCS 8507160742 |
Download: ML20129E141 (5) |
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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20216D9541999-07-28028 July 1999 Safety Evaluation Authorizing Proposed Alternatives for Second 10-year Interval Pursuant to 10CFR50.55a(a)(3)(ii) ML20202J0891999-02-0303 February 1999 Safety Evaluation Accepting Rev 2 of Third Interval Inservice Testing Program for Pumps & Valves for James a FitzPatrick Nuclear Power Plant ML20196J3501998-12-0404 December 1998 SER Accepting License Response to GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves ML20196F9251998-11-25025 November 1998 Safety Evaluation Re Thrid 10-year Interval Inservice Insp Program Relief Requests for Plant ML20195K4211998-11-17017 November 1998 Safety Evaluation Authorizing Proposed Alternative in Relief Request VRR-05 Per 10CFR50.55a(a)(3)(i) & PRR-01,PRR-02R1, PRR-03,PRR-04,VRR-02,VRR-03 & VRR-04 Per 10CFR50.55a(a)(3)(ii) ML20195E1051998-11-13013 November 1998 Safety Evaluation Accepting Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power Operated Gate Valves, Issued 950817 ML20155H5801998-11-0303 November 1998 Safety Evaluation Authorizing Postponement of Beginning of Augmented Exam Requirements of 10CFR50.55a(g)(6)(ii)(A)(2) Re ASME Code,Section XI ML20155H5321998-11-0303 November 1998 Safety Evaluation Authorizing Alternative to ASME Code Requirements for CRD Bolting ML20154L6591998-10-14014 October 1998 Safety Evaluation Accepting Licensee Proposed Alternative to Snubber Visual Inservice Exam Intervals & Sampling Rates Requirements Contained in ASME Code,Section Xi,Subsection Iwf,Article IWF-5000 ML20153D2591998-09-21021 September 1998 SER Accepting Proposed Alternative Testing of Containment Following ECCS Suction Strainer Replacement ML20236V8181998-07-29029 July 1998 Safety Evaluation Accepting Request for Relief from Implementation of Requirements of 10CFR50.55a Related to Containment Repair & Replacement Activities for James a FitzPatrick Nuclear Power Plant ML20236X5881998-07-29029 July 1998 Safety Evaluation Supporting Amend 245 to License DPR-59 ML20154L9201998-07-10010 July 1998 SER Accepting Rev to Reactor Vessel Surveillance Capsule Withdrawal Schedule for James a Fitzpatrick Nuclear Power Plant ML20217A4631998-03-23023 March 1998 Safety Evaluation Accepting Use of Three Heats/Lots of Hot Rolled XM-19 Matl in Core Shroud Repair Assemblies Re Licenses DPR-16 & DPR-59,respectively ML20198S3031998-01-14014 January 1998 Supplemental SE Accepting RG 1.97,rev 2 Recommendations for Containment Isolation Valve Position Indication Instrumentation at NPP ML20128G2441993-02-0909 February 1993 Correction to NRC SE Associated W/Ts Amend 184,dtd 921217. SE Restates Portion of Section 2.0 ML20126F4771992-12-23023 December 1992 Safety Evaluation Granting Licensee Relief from ASME Code Requirements for Repair of RWCU Equalizing Line Until Next Refueling Outage ML20059K4331990-09-13013 September 1990 Safety Evaluation Accepting Util 881110,890328 & 900129 Submittals Re IGSCC Insp & Repair for Facility Reload 8/ Cycle 9 Refuel Outage ML20056B4011990-08-20020 August 1990 Safety Evaluation Approving Licensee Relief Request R14 & Denying Requests R15 & R5A Re Hydrostatic Test Requirements ML20058P4331990-08-13013 August 1990 Safety Evaluation Accepting ATWS Recirculation Pump Trip Sys Design Mod ML20206F5701988-11-18018 November 1988 Safety Evaluation Re Compliance w/10CFR50.62 ATWS Rule Re Alternate Rod Injection & Recirculating Pump Trip Sys ML20206D5231988-11-10010 November 1988 Safety Evaluation Supporting 880309 Request for Relief from Hydrostatic Test Requirement for HPCI & Rcic,Provided That Alternative Testing Performed ML20148B1001988-03-14014 March 1988 Safety Evaluation Accepting Util Justification for Deviations from Reg Guide 1.97 for post-accident Monitoring Variables ML20236G5801987-10-27027 October 1987 Safety Evaluation Supporting Util 850930,860827 & 1208 Submittals of Second 10-yr Inservice Insp Program Plan & Associated Relief Requests from ASME Code Insp Requirements ML20238E1461987-09-0808 September 1987 Safety Evaluation of Util 870415 Proposed Design for Standby Liquid Control Sys.Design Meets Requirements of 10CFR50.62 Re ATWS ML20237L5241987-09-0101 September 1987 Safety Evaluation Supporting Util 831109,840629 & 850702 Responses to Generic Ltr 83-28,Items 2.1 & 4.5.2 Re Equipment Classification & Vendor Interface & Reactor Trip Sys Reliability, Respectively ML20236J8151987-07-30030 July 1987 Safety Evaluation Re Insps for & Repairs of IGSCC During Reload 7/Cycle 8 Refueling Outage.Facility Can Be Safely Operated for One 18-month Fuel Cycle in Present Configuration ML20212F9421986-12-31031 December 1986 Safety Evaluation Supporting Amend to License DPR-59, Changing Tech Specs Re Second Level of Undervoltage Protection ML20214R5871986-11-24024 November 1986 Safety Evaluation Accepting Util Actions to Ensure Structural Integrity of Vacuum Breakers in Mark I Containments ML20210T3101986-10-0202 October 1986 Safety Evaluation Accepting Util 860228 Submittal of Rev 2 to Offsite Dose Calculation Manual on Interim Basis ML20205E3601986-08-0606 August 1986 Safety Evaluation on Util 830806,1109 & 840330 Responses to Generic Ltr 83-28,Items 3.1.1,3.1.2,3.2.1,3.2.2 & 4.5.1. Licensee Complied W/All Items ML20210K5571986-04-18018 April 1986 Safety Evaluation Supporting Util Request for Relief from First 10-yr Inservice Insp Requirements for Class 1,2 & 3 Components ML20137S9901985-09-26026 September 1985 Safety Evaluation Accepting MSIV Leakage Control Sys,Per GDC 54, Piping Sys Penetrating Containment ML20134D2071985-08-0909 August 1985 Safety Evaluation of Util 831109 & 840629 Responses to Generic Ltr 83-28,Item 1.1, Post-Trip Review (Program Description & Procedure). Program & Procedures Acceptable ML20133F0011985-07-30030 July 1985 Safety Evaluation Accepting Util 831109 & 840629 Responses to Generic Ltr 83-28,Items 3.1.3 & 3.2.3 Re post-maint Testing ML20129E1721985-07-0101 July 1985 Safety Evaluation Re Radiological Consequences of Hypothetical LOCA While Purging Containment at Power. Radiological Consequences Acceptable ML20129E1411985-07-0101 July 1985 Safety Evaluation Supporting Demonstration of Containment Purge & Vent Valve Operability.Info Submitted Demonstrates Ability of 20-inch & 24-inch Purge & Vent Valves to Close Against Buildup of Containment Pressure During Dba/Loca ML20127E7221985-06-17017 June 1985 SER Supporting Util 840629 Response to Generic Ltr 84-09, Recombiner Capability Requirements of 10CFR50.44(c)(3)(ii) ML20127B3251985-06-10010 June 1985 Interim Safety Evaluation Approving Util 830630 Procedures Generation Package (PGP) for Emergency Operating Procedures Upon Resolution of Exceptions Noted in Section 2.PGP Submitted Per Generic Ltr 82-33 Re Suppl 1 to NUREG-0737 ML20127C7961985-06-0606 June 1985 Safety Evaluation Re Insp & Repair of RCS Piping.Plant Can Be Safely Returned to Operation in Present Configuration for Duration of Cycle 7 ML20140G5731975-07-15015 July 1975 Safety Evaluation Supporting Tech Spec Changes to License DPR-59 to Revise Suppression Pool Water Temp Limits 1999-07-28
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARJAFP-99-0277, Monthly Operating Rept for Sept 1999 Fpr Ja FitzPatrick Nuclear Power Plant.With1999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 Fpr Ja FitzPatrick Nuclear Power Plant.With ML20217A9931999-09-30030 September 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data JAFP-99-0261, Monthly Operating Rept for Aug 1999 for Jafnpp.With1999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Jafnpp.With JAFP-99-0236, Monthly Operating Rept for July 1999 for Ja FitzPatrick Npp. with1999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Ja FitzPatrick Npp. with ML20216D9541999-07-28028 July 1999 Safety Evaluation Authorizing Proposed Alternatives for Second 10-year Interval Pursuant to 10CFR50.55a(a)(3)(ii) ML20196H8621999-06-30030 June 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data, June 1999 Rept JAFP-99-0211, Monthly Operating Rept for June 1999 for Ja FitzPatrick Nuclear Power Plant.With1999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Ja FitzPatrick Nuclear Power Plant.With JAFP-99-0175, Annual Summary of Changes,Tests & Experiments for 1997/1998. with1999-06-0202 June 1999 Annual Summary of Changes,Tests & Experiments for 1997/1998. with JAFP-99-0181, Monthly Operating Rept for May 1999 for Jafnpp.With1999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Jafnpp.With JAFP-99-0166, Monthly Operating Rept for Apr 1999 for Ja FitzPatrick Nuclear Plant.With1999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Ja FitzPatrick Nuclear Plant.With JAFP-99-0142, Monthly Operating Rept for Mar 1999 for Ja FitzPatrick Nuclear Power Plant.With1999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Ja FitzPatrick Nuclear Power Plant.With JAFP-99-0092, Monthly Operating Rept for Feb 1999 for Ja FitzPatrick Nuclear Power Plant.With1999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Ja FitzPatrick Nuclear Power Plant.With ML20207E9311999-02-26026 February 1999 Part 21 Rept Re Sprague Model TE1302 Aluminum Electrolytic Capacitors with Date Code of 9322H.Caused by Aluminum Electrolytic Capacitors.Affected Capacitors Replaced ML20202J0891999-02-0303 February 1999 Safety Evaluation Accepting Rev 2 of Third Interval Inservice Testing Program for Pumps & Valves for James a FitzPatrick Nuclear Power Plant ML20199M0891999-01-22022 January 1999 Part 21 Rept Re Failure of Square Root Converters.Caused by Failed Aluminum Electrolytic Capacitory Spargue Electric Co (Model Number TE1302 with Mfg Date Code 9322H).Sent Square Root Converters Back to Mfg,Barker Microfarads,Inc JAFP-99-0011, Monthly Operating Rept for Dec 1998 for Ja FitzPatrick Nuclear Power Plant.With1998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Ja FitzPatrick Nuclear Power Plant.With ML20198F9991998-12-0404 December 1998 Assessment of Licensing Basis for Use of Containment Overpressure Credit for Net Positive Suction Head Analyses Power Authority of State of New York,James a Fitzpatrick Nuclear Power Plant ML20196J3501998-12-0404 December 1998 SER Accepting License Response to GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves JAFP-98-0396, Monthly Operating Rept for Nov 1998 for Ja FitzPatrick Npp. with1998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Ja FitzPatrick Npp. with ML20196F9251998-11-25025 November 1998 Safety Evaluation Re Thrid 10-year Interval Inservice Insp Program Relief Requests for Plant ML20195J7521998-11-18018 November 1998 Rev 7 to Jaf Colr ML20195K4211998-11-17017 November 1998 Safety Evaluation Authorizing Proposed Alternative in Relief Request VRR-05 Per 10CFR50.55a(a)(3)(i) & PRR-01,PRR-02R1, PRR-03,PRR-04,VRR-02,VRR-03 & VRR-04 Per 10CFR50.55a(a)(3)(ii) ML20195E1051998-11-13013 November 1998 Safety Evaluation Accepting Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power Operated Gate Valves, Issued 950817 ML20197G6221998-11-0606 November 1998 Non-proprietary Rev 7 to HI-971661, Licensing Rept for Reracking of Ja FitzPatrick Sfp ML20155H5321998-11-0303 November 1998 Safety Evaluation Authorizing Alternative to ASME Code Requirements for CRD Bolting ML20155H5801998-11-0303 November 1998 Safety Evaluation Authorizing Postponement of Beginning of Augmented Exam Requirements of 10CFR50.55a(g)(6)(ii)(A)(2) Re ASME Code,Section XI JAFP-98-0360, Monthly Operating Rept for Oct 1998 for Ja FitzPatrick Nuclear Power Plant.With1998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Ja FitzPatrick Nuclear Power Plant.With ML20155C2821998-10-30030 October 1998 Non-proprietary Rev 0 to GENE-187-30-1598 Np, CRD Bolting Flaw Evaluation for Ja FitzPatrick Nuclear Power Plant ML20154L6591998-10-14014 October 1998 Safety Evaluation Accepting Licensee Proposed Alternative to Snubber Visual Inservice Exam Intervals & Sampling Rates Requirements Contained in ASME Code,Section Xi,Subsection Iwf,Article IWF-5000 JAFP-98-0322, Monthly Operating Rept for Sept 1998 for Ja Fitzpatrick Nuclear Power Plant.With1998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Ja Fitzpatrick Nuclear Power Plant.With ML20153D2591998-09-21021 September 1998 SER Accepting Proposed Alternative Testing of Containment Following ECCS Suction Strainer Replacement ML20153B5611998-09-0101 September 1998 Rev 1 to JAF-SE-98-013, RHR & Core Spray Suppression Pool Suction Strainer Replacement ML20151X6891998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Ja FitzPatrick Nuclear Power Plant ML20237E8361998-08-25025 August 1998 Rev 6 to Colr ML20237E9471998-08-0808 August 1998 Rev 6 to Colr JAFP-98-0264, Monthly Operating Rept for July 1998 for Ja FitzPatrick Nuclear Plant1998-07-31031 July 1998 Monthly Operating Rept for July 1998 for Ja FitzPatrick Nuclear Plant ML20236X5881998-07-29029 July 1998 Safety Evaluation Supporting Amend 245 to License DPR-59 ML20236V8181998-07-29029 July 1998 Safety Evaluation Accepting Request for Relief from Implementation of Requirements of 10CFR50.55a Related to Containment Repair & Replacement Activities for James a FitzPatrick Nuclear Power Plant ML20153B5781998-07-28028 July 1998 Rev 0 to JAF-SE-98-025, HPCI & RCIC Suppression Pool Suction Strainer Replacement ML20236X3831998-07-14014 July 1998 Rev 2 to JAF-RPT-MULTI-02671, Summary of Detailed Evaluation for NRC Generic Ltr 96-06 ML20154L9201998-07-10010 July 1998 SER Accepting Rev to Reactor Vessel Surveillance Capsule Withdrawal Schedule for James a Fitzpatrick Nuclear Power Plant JAFP-98-0222, Monthly Operating Rept for June 1998 for Ja FitzPatrick Nuclear Power Plant1998-06-30030 June 1998 Monthly Operating Rept for June 1998 for Ja FitzPatrick Nuclear Power Plant JAFP-98-0193, Monthly Operating Rept for May 1998 for Ja FitzPatrick Nuclear Plant1998-05-31031 May 1998 Monthly Operating Rept for May 1998 for Ja FitzPatrick Nuclear Plant ML20248F3531998-05-21021 May 1998 Part 21 Rept Re Electronic Equipment Repaired or Reworked by Integrated Resources,Inc from Approx 930101-980501.Caused by 1 Capacitor in Each Unit Being Installed W/Reverse Polarity. Policy of Second Checking All Capacitors Is Being Adopted JAFP-98-0168, Monthly Operating Rept for Apr 1998 for Ja FitzPatrick Nuclear Power Plant1998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for Ja FitzPatrick Nuclear Power Plant JAFP-98-0128, Monthly Operating Rept for Mar 1998 for Ja FitzPatrick Nuclear Power Plant1998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for Ja FitzPatrick Nuclear Power Plant ML20217A4631998-03-23023 March 1998 Safety Evaluation Accepting Use of Three Heats/Lots of Hot Rolled XM-19 Matl in Core Shroud Repair Assemblies Re Licenses DPR-16 & DPR-59,respectively JAFP-98-0091, Monthly Operating Rept for Feb 1998 for JAFNPP1998-02-28028 February 1998 Monthly Operating Rept for Feb 1998 for JAFNPP ML20202G9081998-02-0606 February 1998 Safety Evaluation Re Amend to License DPR-59 to Revise TS Tables 3.2-2 & 4.2-2 JAFP-98-0058, Monthly Operating Rept for Jan 1998 for Ja FitzPatrick Nuclear Power Plant1998-01-31031 January 1998 Monthly Operating Rept for Jan 1998 for Ja FitzPatrick Nuclear Power Plant 1999-09-30
[Table view] |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ,
REGARDING DEMONSTRATION OF CONTAINMENT PURGE AND VENT VALVE OPERABILITY ,
i POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT .
DOCKET No. 50 333 1.0 Recui rement Demonstration of operability of the containment purge and vent valves, par-ticularly the ability of these valves to close during a design basis accident, is necessary to assure containment isolation. This demonstration of oper-ability is required by BTP CSB 6-4 and SRP 3.10 for containment purge and vent valves which are not sealed closed during operational conditions 1, 2, 3, and 4
2.0 Description of Puroe and Vent Valves Valve Size Number (Inches) location Use >
27 A0V-111 24 Outside containment - drywell Not given 27 A0V-112- 24 Outside containment - drywell Not given 27 A0V-113 24 Outside containment - drywell Not given 27 A0V-114 24 Outside containment - drywell Not given 27 A0V-115 20 Outside containment - wetwell Not given 27 A0V-116 20 Outside containment - wetwell Not given 27 A0V-117 20 Outside containment - wetwell Not given 27 A0V-118 20 Outside containment - wetwell Not given The 20-inch and 24-inch valves are butterfly-type Model 9222 manufactured by
, Fisher Controls Company and are equipped with G. H. Bettis Company model 733C-SR80 and 732C-SR80 operators, respectively. The valves are equipped with mechanical stops to limit the disc angle opening to 50' (90"= full . open).
PASNY proposes in Reference G to further limit the 24-inch valve numbers 27 A0V-111,112, and 113 to 40' open.
8507160742 850701 hDR ADOCK 0500 3
- 3 3.0 Demonstration of Operability 3.1 The Power Authority of the State of New York (PASNY) has provided oper-ability demone..ation information for the purge and vent system isolation valves at their James A. Fitzpatrick Nuclear Power Plant in the following submittals: .
A. PASNY letter dated May 12, 1981 from J. P. Bayne (PASNY) to T. A. Ippolito (NRC).
B. PASNY letter dated June 13, 1980 f rom J. P. Bayne (PASNY) to T. A. Ippolito (NRC).
C. PASNY letter dated May 6,1980 from P. J. Early (PASNY) to T. A. Ippolito (NRC).
D. NRC letter dated January 13, 1984 from D. G. Eisenhut (NRC) to l J. P. Bayne (PASNY).
e E.
PASNY letter dated February 24, 1984 from J. P. Bayne (PASNY) to D. B.
Vassallo (NRC).
F.
NRC letter dated October 9,1984 from D. G. Eisenhut (NRC) to J. P. Bayne (PASNY).
G.
PASNY letter dated November 26, 1984 fromJ.P.Bayne(PASth)to D. B. Vassallo (NRC).
H.
PASNY letter dated May 3,1985 from J. P. Bayne (PASNY) to D. B. Vassallo (NRC).
. 3.2 PASNY's approach to operability a'emonstration is based on the following assumptions:
- a. Single valve operation, i.e., redundant in-series valve is to have failed open,
- b. Pressure losses due to inlets, piping / ducting, filters, etc. are neglected.
- c. For valves with asymmetric discs, flow is assumed toward the hub side for purposes of predicting dynamic torques.
Dynamic torque (T ) predictions stem from coefficients developed by bench testsonmodelvakvesrepresentingthedesignofthein-servicevalves.
Analytical techniques involving scaling are used to determine T for the actual valve sizes. The I.S.A. paper entitled, "Effect of Flui0 Compressibility on Torque in Butterfly Valves," gives the basis for Fisher's approach to TD prediction.
Fisher's approach to evaluating critical valve parts in this valve design is to -
detennine maximum allowable APs across the valve at a given disc angle. This maximum allowable AP is based on the weakest operating part of the valve but does not include the operator and associated mounting hardware. The maximum allowable 4P for each disc angle (10 increments) is compared to the operating pressure condition.
From this, the maximum disc-opening angle is selected.
The Fisher developed computer program used to establish the maximum opening angle is described as follows:
- 1. For a given valve at some angle of opening, the program begins by calculating the loading. This includes a hydrostatic load on the disc, seating torque, bushing, and packing torque and dynamic torque.
e .
- 2. After the loading is detemined, the program calculates stresses in the shaft, key, pin, and bushing for a specific 4P. These calculated stresses are compared to the allowable stresses. For the ferritic steel shaft which is the weakest member, the design (allowable) stress intensity (Sm) from Section III of the ASME Boiler and Pressure Vessel Code is taken as 1/3 of the minimum ultimate tensile strength at room tempera ture. Fisher uses a value of 0.75 Sm for an allowable shear stress.
- 3. The program calculates stress and changes AP iteratively until the allowable stress matches the calculated stress. This determines the maximum allowable pressure drop for that angle of opening based on the stress at a single point. This process is done for cases 1, 2, 3, 4, and
. 5 (as defined below) for each angle of opening.
Case 1 - Stress in the shaft at the disc hub due to bending and torsion.
Case 2 - Stress in the shaft at the disc hub due to torsion and transverse shear.
Case 3 - Stress at the pinned disc-shaft connection.
Case 4 - Stress at the keyed actuator-shaft connection Case 5 - Stress at the shaft bushing.
- 4. The program output shows a AP which is calculated at each point for each angle of opening, including two4Ps for Case 1 (one based on maximum shear stress and one based on maximum tensile stress) for a total of six A Ps. The smallest AP of these six is then repeated as the allowable AP at the bottom of the column. The actuator torque for the lowest AP (allowable 4P) is also listed.
3.3 Reference G contains a re-analysis by PASNY based on a revised LOCA containment pressure response curve and a proposed further limitation to a maximum opening angle of 40* for 24-inch valves 27 A0V-111, -112, and -113.
These three 24-inch valves have 90* elbows upstream with valve shafts at 90* -
to the plane of the elbow.
The analysis by PASNY applies a factor of three to the closing torque of the 24-inch valves, with shafts out of plane with the upstream elbow, and is based on analytical methods used by Fisher Control.
Analysis results include the AP capability of the valve shaft considered to be l the most critical valve component, and the actuator torque margin during l closure.
4.0 Evaluation 4.1 In Reference C, PASNY provided information describing accident conditions used by Fisher Controls to assess the subject valves. Fisher makes the assumption that the total containment pressure is the AP across each of the
l isolation valves from the full open to the full closed position during closure. For valves in lines from the drywell, 39 psi was given as !
the design pressure. For valves in lines from the wetwell, 22 psi was t given as the design pressure. PASNY noted that 39 psi envelopes all drywell pressure values obtained from the most recent drywell respo_nse analysis and that 22 psi envelopes all wetwell pressure values obtained in the first 8 seconds based on the wetwell response analysis.
Given that 39 psi is the peak containment pressure predicted for the drywell, the staff finds that using a AP of 39 psig as the basis for operability assess-ment is acceptable. In the case of the wetwell valves, a AP of 22 psig is also acceptable.
Fisher's model valve bench test programs used to develop dynamic torque coefficients (Cr) were configured with straight pipe inlets. Testing did not include inlet c6nfigurations involving elbows and therefore the effects on C can not be quantified for those Fitzpatrick valves which have elbow u stream of the valve.
Information available from other valve manufacturers indicated that for a given valve design at the same conditions, the ratio of C7 (elbow-shaft in plane) to C (strai ht pipe) is greater than one and the fatio of C bow-shaft o t plane to CT (straight pipe) is greater than two in sdme(el-in-stances. Use of straight pipe developed C7 s for in service valves with an upstream elbow configuration would result in dynamic torque predictions that are not acceptable.
Based on information available, the staff believes that where bench tests did not include elbows in the piping configuration, a factor of 1.5 times the CT (straight pipe) for an elbow-shaft in plane valve installation con-
. figuration and a factor of 3 times the Cr (straight pipe) for an elbow-shaft out of plane valve installation configur& tion would yield conservative values -
of TD*
4.2 Valve loads for the 20-inch and 24-inch purge and vent valves are predicted by PASNY in the analysis included as part of Reference G. The analysis is based on Fisher Controls developed equations for their 9200 series butterfly valves. Since Fisher does not have data to quantify the effect on CT (dynamic torque coefficient) of upstream elbows, PASNY applies a torque multiplication factor of three for those valves with upstream elbows and shafts 90 out-of-plane. The staff finds the approach used by PASNY in the prediction of valve torques conservative.
The use of 0.75Sm as the allowable shear stress by Fisher (Section 3.2 of this evaluation) exceeds the generally accepted allowable shear stress of 0.6Sm (ASME Code -NB3227.2) or the AISC allowable shear stress of 0.4Sy (Sy= tensile yie16). However, the conservatism in the Fisher analysis results in the
1 l
0.6 Sm allowable shear stress not being exceeded. For example, for the valves that constitute the worst case, A0V-111, 112 & 113, that is, there is an elbow upstream and the valve shaft is out of the plane of the elbow, the shaft AP capability from Table 1 of Reference G at the 30' valve opening is 27 psi.
The AP developed at 30' is 19.23 psi, and the ratio of APs is 1.4. Thus, the calculated shear stress does not exceed 0.651 (0.75/1.4 = 0.54Sdi).
4.3 The capability of the Bettis 732C-SR 80 and 733C-SR 80 actuators to close the valves from the 50' and 40' limited opening positions is shown in curves provided with the analysis contained in Reference G. The curves compare required valve torque developed during closure against the containment pressure ramp with the available actuator spring torque. Adequate torque
. margins are shown for each valve.
4.4 The structural adequacy of the interfacing hardware (actuator / valve) is demonstrated in the Fisher Controls analysis provided as part of Reference H.
The calculated shear stresi is shown to be less than the allowable shear stress for the bolt material.
4.5 Seismic qualification for the 20-inch and 24-inch valve assemblies is addressed by the Fisher Controls Company report number CD72-234 dated July 19, %
1982.
5.0 Conclusion '
We have completed our review of infonnation submitted to date concerning operability of containment purge and vent valves for James A. Fitzpatrick Nuclear Power Plant. Sections 4.1, 4.2, 4.3 and 4.4 are the basis for the conclusion of the staff. We find the information submitted demonstrates the ability of the 20-inch and 24-inch purge and vent valves to close against the
, buildup of containment pressure in the event of a DBA/LOCA. The 50 opening angle limitation applies to valve numbers 27 A0V-114 , -115. -116, -117, and -
-118 and the 40' opening angle limitation applies to valve numbers 27 A0V-111,
-112, and -113. The Technical Specifications for valves 27 A0V-115, 27 A0V-116, 27 A0V-117 and 27 A0V-118 should reflect a closing time of less than or equal to 8 seconds and the limitation of opening angles as previously described for all valves.
Principal Contributor: R. Wright, EQB Dated: July 1,1985