ML20062J679
| ML20062J679 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/31/1980 |
| From: | Rosztoczy Z Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20062J670 | List: |
| References | |
| NUDOCS 8011040480 | |
| Download: ML20062J679 (4) | |
Text
s UNITID STATES'OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC _ SAFETY _ AND LICENSING BOUD dn the matter of 5'
METROPOLITAN EDISON COMPANY, Docket No. 50-2E9
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ETAL (Three Mile Island Nuclear Station Unit 1)
NRC STAFF TESTIMONY OF ZOLTAN R. USZTCCZY IN RESPONSE TO BOARD QUESTION 5 ON UCS CONTENTION 12
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Q.1 The TMI-2 accidcat demonstrated that some safe y-related equipment may have been exposed or was in inrninent danger of being exposed to environmental stresses beyond that for which it was qualified.
The board's concern is primarily with such equipment qualification.
In addition, envimnmental stress to safety-re;ated equipment will be of concern to the extent that such equipment is not included in existing staff requirements.
A.
The only known environmental stress to which certain safety related equipment at TMI-2 had been exposed which was beyond that for which it was qualified was submergence due to flooding.
TMI-1 initiated corrective measures which involve moving instruments affected to higher elevations thersby eliminating the concern. The corrective measures will be cor.:pleted, prior tc TMI-1 restart.
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2.
Which items df Regulatory Guide 1.89 have been e andfathered with respect to TMI-1? Explain any justification for allowing restart without compl$nce with the grandfathered items.
A.2 To date, no items of 8sgulatory Guide 1.89 have been grandfathered with respect to THI-1. The extent to which the environmental quali-fication of equipment important to safety satisfies the requirements of IEEE 323-1974 and Regulatory Guide 1.89 will be addressed in the staff's review of the licensee's response to IE Bu11etir.79-31B,~ dated Sept. 4 Mh and Nov.1,1980.
If any equipment is found deficient based on the abc7e review the staff will require requalification or replacement with sui. ably qualified equipment in accordance with the above standard and regulatory guide.
This review will provide reasonable assurance that safety related equipment is acceptably qualified to appropriate req 2irements. It is expected that the work by the staff, in the form of a safety evaluation report, will be completed by February 1,1981 per the Connission Order of May 23, 1980 (CLI-80-21). Any modifications required will be inplemented prior to June 30, 1982. During the review process the staff l('
may dec'ide that some items in Regulatory Guide 1.E9 :an be grandfathered with respect to TMI-1. Should this be the case, the staff's justi-fication for such a position will be articulated in cepth in the SER.
3.
What are the environmental qualification criterit which equipment inside of containment must meet with respect t.o radiation levels and length of time of exposure? (Address the Ir.:erim Staff Position ~
on Environmental Qualification of Electrical Equipment. (NUREG-0588),.
A.3 The environmental qualification P.equirements which equipment inside of containment must meet with respect to radiation levels and length of time
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of exposure are contained in the DDR Guidelines, K'JRIS-0538.
Both the DDR guidelines and NUREG-0588 are sic:ilar it that they provide the Jnethods for determining the radiation source tern wten considering 5 LOCA events inside containment. These methods corsi:er the radiation source ten: resulting from an event which completely depressurizes the primary system and releases the source tern inventory to the containment.
NURIG-0578 provides the radiation source tenn te be : sed for determining the qualification doses for equipment in close ;rcxicity to recirculating fluid systems inside of containment as a result of L*CA..
This method considers a LOCA event in which the primary sys ec nay net depressurize and the source term inventory remains in the co:la.nt.
NUREG-0588 also provides the radiation sour:e ten: t: be used for quali-fying equipment following non-LOCA events inside con:af ncent.
The length of time to which equipment must be q;aliffed is evaluated.
individually for each component considering its safe:y function. The quali-fied life is given in the licensees equipment 4:alification submittal, is subiect to staff review, and will be reportef in -he SER.
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PROFESS 10fiAL QUALIFICATI0fis 0F 10LTAfi R. ROSITOCIY m
Fy name is Zoltan R. Roszto:2y.
I am currently employed by the ll.S. Nuclear Regulatory Commission as Branch Chief cf the Equipment Qualification Branch, Divisicn of Engineering, Office of Nuclear Reactor Regulation (NRR). Prior to the formation of the Equipment Qualification Branch I served as Branch Chief cf the Analysis Branch, NRR.
My duties and responsibilities include the supervision of the review and evaluation of the capabilities of safety related systems and componenis to perfcrm their design functions under all normal, abnormal and accidental environmer. tai conditiers and in the event of seismic oc:urences. I participate in the develcpment of criteria, guides and regulations pertaining to equipment qualifi catt er..
I am a graduate of the University of Tecnnical Sciences, Hungary.
I have also re:eived a Master of Science degree in M=chanical Eagineering frcm the University of California and a Ph.D. in Nuclear Engineering from the University f
of Ari:ona.
- v Prior to my appointment with the NRC, I was employed by Combustion Engineering Inc.,
Wincsor, Connecticut (1964-1973) and by the Institute of Engineering Research.
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. University of California (1959-1952). During the last six years of my empicyment
. with Combustion Engineering (CE) I served as Manager, Safety Analysis.
In this capacity I was responsible for directing the safety evaluation of CE su;piied nucl ear power, systems. I also coordinated CE's participation in the rule making e
hearing on,* ergency Core Cooling Effectiveness.
}
I am a member of the American Nuclear Society and of Sigma Xi.
4
Table 3 (Cont.)
II.J.2.1 -
Reorient Const. Inspection Program II.J.2.2 Indep. Measurement in the Const. Inspect. Prog.
II.J.2.3 Assign Resident Inspectors to all Const. Sites II.J.3.1 Grgan. & Staffing to Oversee Design & Const.
II.J.3.2 Issue Reg. Guide on Management for Design & Const.
II.J.4.1 Revise Deficiency Reporting Requirements II.K.l.1 Review Chronology of TMI-2 II.K.1.2 Review Transients Similar to TMI-2 II.K.1.3 Review Procedures for Recognizing Void Formation II.K.1.4 Review Procedures for Proper ESF & HPI Operation II.K.1.5 Safety-Related Valve Position II.K.1.6 Review of Containment Isolation Design & Procedures II.K.1.7 Positive Control on Valves That Could Comp. AFW II.K.1.8 Assurance of two 100% Flow Paths for AFW II.K.1.9 Review of Procedures to Prevent Inad. Radio. Releases II.K.1.10 Procedures for remov/ restoring Safety-Related Systems II.K.1.11 Instructions fcr all Maint. & Ops. Personnel
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II.K.1.12 One Hour Notification II.K.1.13 Tech. Spec. Changes II.K.1.14 Review Procedures to Deal with Hydrogen II.K.1.15 Comunications for Non-automatic AFW Initiation II.K.1.16 Procedures for Control of PORY II.K.1.17 Eliminate SI activation on Pzr. Low Level II.K.1.18 Procedures & Training on Natural Circulation II.K.1.19 Modifications to Reduce PORY Actuation II.K.1.20 Procedures for Manual Reactor Trip II.K.1.21 Safety-Grade Anticipatory Reactor Trip II.K.1.22 Actions for HR Systems when Feedwater not Avail.
II.K.1.23 Reactor Vessel Level Indication II.K.1.24 LOCA Analysis for Determining When to Trip RCPs II.K.1.25 Develop Guidelines for Oper. Actions Based on LOCA Anal.
II.K.1.26 Revise Emerg. Proc. & Train Opers. Based Upon Guidelines II.K.1.27 Analysis, Guidelines, & Procedures for ICC II.K.1.28 Provide Design for Auto. Trip of RCPs I
II.K.2.1 Upgrade Timeliness & Reliability of AFW System l
II.K.2.2 Separate AFW and ICS II.K.2.3 Hard-Wired. Control-Grade Antic. Reactor Trip II.K.2.4 SBLOCA Analysis, Procedures, and Training II.K.2.5 Complete TMI-2 Simulator Training II.K.2.6 Reevaluate Analysis of Dual-Level OTSG Setpoint II.K.2.7 Reevaluate DB-1 Transient of 09/24/77 II.K.2.8 Continued Upgrading of AFW System II.K.2.9 FMEA of ICS II.K.2.10 Hard-Wired Safety-Grade Antic. Reactor Trip II.K.2.11 Operator Training & Drilling II.K.2.12 Transient Analysis and Proc. for Momt. SBLOCAs t
l II.K.2.13 Thermal-Mechanical Report (RV Integ. Feed-&-Bleed)
II.K.2.14 Demonstrate Lift Freq. of PORV is Acceptable II.K.2.15 Anal. of Effect of Slug Flow on OTSG Tubes II.K.2.16 Impact of RCP Seal Damage with LOOP & SBLOCA
.l II.K.2.17 Analysis of Potential Voiding During Transients II.K.2.18 Analysis of FW and Other Anticipated Transients II.K.2.19 Benchmark Anal. of Sequential AFW Flow to OTSG II.K.2.20 Anal. System Response to SBLOCA repressurized to PORY SP II.K.2.21 LOFT 3-1 Predictions 3-3
Table 3 (Cont.)
i II.K.3.1 Autoir.atic PORY Isolation System II.K.3.2 Report on Safety Effect PORY 1s01. System II.K.3.3 '
Report PORY & Safety Valve Fail. & Challenges II.K.3.4 Upgrade Rel, & Red, of Non-Safety Grade Equip.
II.K.3.5
' Continue to Study Need to Trip RCPs During LOCA II.K.3.6 Instrumentation to Verify Natural Circulation II.K.3.7 Eval. FGRV Opening Probability II.K.3.8 Consideration, of Need for Diverse DHR Method II.K.3.9 Modification to Westinghouse PID II.K.3.10 Proposal by W Licensees for Antic. Trip Setpoint II.K.3.ll Controlled Use of Control Components Type PORV II.K.3.12 Confinn Antic. Rx. Trip on W Plants II.K.3.13 Separation of HPCI & RCIC oii BWRs II.K.3.14 Isolation of Isolation Condensers on BWRs II.K.3.15 Modify Break Detection Logic on BWRs II.K.3.16 Reduce Challenges & Failures of Relief Valves II.K.3.17 Report on ECCS Outages II.K.3.18 Modifications to ADS Logic on BWRs II.K.3.19 Interlock on Recir. Pumps for BWRs II.K.3.20 Loss of Serv. Water at Big Rock Point II.K.3.21 Restart of Core Spray & LPCI for BWRs II.K.3.22 Auto. Switchover for RCIC Suction for BWRs II.K.3.23 Central Water Level Recording for BWRs II.K.3.24 Confonn Adeg. of Space Cooling HPCI & RCIC for BWRs II.K.3.25 Effect of Loss of AC on Pump Seals for B.WRs (W & CE)
II.K.3.26 Study on RHR Reliability for Fuel Pool Cooling for BWRs IT.K.3.27 Conanon Reference for Vessel Level Inst, for BWRs II.K.3.28 Qualification of ADS Accumulators for BWRs II.K.3.29 Study Performance of Isolation Condensers for BWRs II.K.3.30 Revise SBLOCA Model to Comply with 10 CFR 50.46 & App. K II.K.3.31 Plant Specific Calculations to show Compliance 10 CFR 50.46 II.K.3.32 Experimental Verification of Two-Phase Natural Circ.
II.K.3.33 Evaluate Elimination of PORY II.K.3.34 RELAP-4 Model Development II.K.3.35 Evaluation of Effects CFT Injection on SBLOCA II.K.3.36 Staff Audits of B&W SBLOCA II.K.3.37 Analysis of Plant Response to Isolated SBLOCA II.K.3.38 Analysis of Plant Response to SB in Pzr. Spray Line II.K.3.39 Eval. of Effects of Slugs Caused by HPI & CFT Flow II.K.3.40 Evaluation of RCP Seal Damage and Leaskage During SBLOCA II.K.3.41 Submit Predictions of LOFT Test L3-6 with RCPs Running II.K.3.42 Submit Information on Effects of Noncondensible Gases II.K.3.43 Eval. of Mech. Effects of Slug Flow on OTSG Tubes II.K.3.44
' Eval. Antic. Trans. with Single Failure for BWRs II.K.3.45 Eval. Depressurization with Other Than ADS for BWRs II.K.3.46 Response to Michelson Concerns for BWRs II.K.3.47 Test Program for SBLOCAiModel Verification for BWRs II.K.3.48 Eval, of Reliability as a Result of Implem. of B&OTF Recs.
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II.K.3.49 Review of Procedures by NRC II.K.3.50 Review of Procedures by NSSS Vendor i
II.K.3.51 Symptom-Based Emergency Procedures II.K.3.52 Operator Awareness of Revised Emergency Procedures II.K.3.53 Two Operators in the Control Room II.K.3.54 Simulator Upgrade for SBLOCAs II.K.3.55 Operator Monitoring of Control Board II.K.3.56 Simulater Training Requirements II.K.3.57 Ident. Water Sources Prior to Man. ADS for BWRs 3-4
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Table 3 (Cont.)
III.A.I.)
Improve Licer.see Emerg. Preparedness (Short-Term)
III.A.1.2 Upgrade Licensee Emerg. Support Facilities III.A.1.3' Maintain Supplies of Thyroid Blocking Agent III.A.2.1 Improve Licensee Emerg. Preparedness (Long-Ters)
III.A.2.2 Development of G.:idance & Criteria III.A.3.1 NRC Role in Responding to Nuclear Emergencies III.A.3.2 Improve Operations Center III.A.3.3 Comunications III.A.3.4 Nuclear Data Link III.A.3.5 Training. Drills, & Tests III.A.3.6
' Interaction of NRC with Other Agencies III.B.1 Transfer of Responsibilities to FEMA III.B.2 Implementation of NRC's 'and FEMA's Responsibilities III.C.1 Infonnation Available for News Medin & Public III.C.2 Develop Policy & Training for News Media Interface III.D.l.1 Primary Coolant Saurces Dutside Containment III.D.1.2 Radioactive Gas Management III.D.I.3 Vent. System and Radiciodine Adsorber Criteria III.D.1.4 Radwaste System Design Features
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III.D.2.1 Radiological Monitoring of Effluents III.D.2.2 Radiciodine, C-14, & Tritium Path. Dose Arial.
III.D.2.3 Liquid Fathway Radiological Control III.D.2.4 Offsite Dose Measurements III.D.2.5 Offsite Dose Calculation Manual III.D.2.6 Independent Radiological Measurements III.D.3.1 Radiation Protection Plans III.D.3.2 Health. Physics Improvements III.D.3.3 Inplant Radiation Monitoring III.D.3.4 Control Room Habitability III.D.3.5 Radiation Worker Exposure Data Base IV.A.1 Seek Legislative Auth to 5'.rengthen Enforcement IV.A.2 Revise Enforcement Policy IV.B.1 Revise Practices for Iss. Inst. & Info to Licensees IV.C.1 Extend Lessons Learned from TMI to Other NRC Programs IV.D.1 NRC Staff Training IV.E.1 Expand Research on Quantification of Safety Decision Making IV.E.2 Plan for Early Resolution of Safety Issues IV.E.3 Plan for Resolving Issues at CP Stage IV.E.4 Resolve Generic Issues by Rulemaking IV.E.5 Assess Currently Operating Reactors IV.F.1 Increased IE Scrutiny of Power Ascension Test Program IV.F.2 Evaluate Impacts of Financial Discentives of Safety IV.G.1 Develop a Public Agenda for Rulemaking IV.G.2 Periadic & Systematic Reevaluation of Existing Rults IV.G.3 Improve Rulemaking Procedures IV.G.4 Study Alternatives for Improved Rulemaking Process IV.H NRC Particiption in Rac. Policy Council V.)
Develop NRC Policy Statement on Safety V.2 Study Elimination of Nonstfaff Responsibilities V.3 Strengthen Role of ACRS V.4 Study Need for Additional Advisory Comittees V.5 Improve Public & Intervenor Participation in Hearings V.6 Study Construction-During-Adjudication Rules V.7 Study Need for TMI-Related Legislation V.S Study Need to Estab. Indep. Nuc. Safety Board V.9 Study the Reform of the Licensing Process 3-5
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Table 3 (Cont.)
V.10 Study NRC Top Management Structure V.11 Reexamine Organ. & Functions of NRC Offices V.12 R'evise Delegations of Authority to Staff V.13 Clarify & Strengthen Roles of Comm., Chr.irman, & EDO V.14 Auth. to Delegate Emerg. Response to Single Comm.
V.15 Achieve Single Location (Long-Term)'
V.16 Achieve Single Location (Interim)
V.17 Reexamine Commission Role in Adjudication 1
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Work assignments prior to this presen't,'.positiori1nclude:
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