ML19338F193
| ML19338F193 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/02/1980 |
| From: | Zudans J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19338F186 | List: |
| References | |
| ISSUANCES-SP, NUDOCS 8010070619 | |
| Download: ML19338F193 (10) | |
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i UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE M ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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METROPOLITAN EDISO:4 COMPA'iY,
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(Three Mile Island Nuclear
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Station Unit 1)
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NRC STAFF TESTIMONY OF JOHN J. ZUDANS RELATIVE TO REACTOR COOLANT PRESSURE BOUNDARY COMPLIANCE WITH GDC 1, 14, 15 & 30 (UCS CONTENTION 6) 0.1.
Please state your name and position with the NRC.
A.
My name is John J. Zudans.
I am an employee of the Nuclear Regulatory Commission assigned to the Equipment Qualification Branch, Division of Engineering, Office of Nuclear Reactor Regulation.
I am a senior Mechanical Engineer assigned to the Seismic and Dynar;'c Load Qualifi-cation Section.
i C.2.
Have you prepared a statement of professional qualifications?
4 A.
Yes.
A copy of this statement is attached to this testimony.
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'i. 3.
Please state the nature of the responsibilities you have had with respect to the Three Mile Island Nuclear Station, Unit 1.
A.
Soon af ter the accident at Three ile Island Unit 2 (TMI-2) on March 28, 1979 i was asked to evaluate the Residual Heat Removal Pumps at TMI-2 which are similar to those at TMI-l for possible use for long term decay j
heat removal.
8010 070 $ / 9
F Q.4.
Please state the purpose of this testimony.
A.
The purpose of this testimony is to address UCS Contention #6.
USC Contention 6 reads as follows:
" Reactor coolant system relief and safety valves form part of the reactor coolant system pressure boundary. Appropriate qualification testing has not been done to verify the capabi-lity of these valves to function during normal, transient, and accident conditions.
In the absence of such testing, verification compliance with GDC 1,14,15 and 30 cannot be found and the public health and safety is endangered."
Q.5.
What are the requirements of General Design Criteria (GDC) 1,14,15, and 30?
A.
General Design Criteria 1 (GDC 1) as stated in the Code of Federal Regulations Part 50 Appendix A requires that " structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards comensurate with the importance of the safety functions to be performed.
Where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency and shall be supplemented or modified as necessary to assure a quality product in keeping with the required safety function.
A quality assurance program shall be established and implemented in order to provide adequate assurance that these structures, systems,-and components will satisfactorily perform their safety functions.
Appropriate records of the design, fabrication, erection, and testing of structures, systems, and components important to safety shall be maintained
f by or under the contro,1 of the nuclear power unit licensee throughout l
l the life of the unit."
l GDC 14 requires that "the reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely
' low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture."
i GDC 15 requires that ' ne reactor coolant system and associated l
auxiliary, control, and protection systems shall be designed with t
l sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during.'iy condition of normal operation, including anticipated operational occurrences."
GDC 30 requires that " components which are part of the reactor coolant pressure boundary shall be designed, fabricated, erected, and tested to the highest quality standards practical and that means shall be provided for detecting and to the extent practical, identifying the location of the source of reactor coolant leakage."
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Q.6.
What are the requirements which the Reactor Coolant Pressure Boundary including safety and relief valves (SD) nust neet to coraply t: :, the reovirenaats cf 3;C 1,14,1:3, and 3J?
A.
The current staff position with respect to the requirements which must l
l be met to comply with GDC 1,14,15, and 30 require that applicarts 1
assessjtheir RCPB including safety and relief valves to the following standards:
- 1) Standard Review Plan (SRP) 3.9.2, "Dynanic Testing and Analyses of Systems, Components, and Equipment,"-
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- 2) SRP 3.9.3, "ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures,"
- 3) Regulatory Guide 1.67, " Installation of Overpressure Protection Devices,"
- 4) Regulatory Guide 1.68, "Preoperational and Initial Startup Test Programs for Water Cooled Power Reactors,"
- 5) Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems,"
- 6) The appropriate sections of Appendix B to 10 CFR 50.
Q.7.
In what respect is the Staff's position with respect to the requirements of GDC 1,14,15, and 30 regarding RCPB, including safety and relief valves, not met?
A.
The staff'e position requires that the safety and relief valves function as expected during design transient and accident conditions.
The extent to which the current staff interpretation of the requirements of GDC 1, 14, 15, and 30, relative to the reactor coolant system safety and relief valves, are not yet verified is that the tests performed to date did not cover loadings which result from two-phase flow or solid fluid flow.
The reactor coolant system safety valves were originally designed and tested for operation on saturated steam in accordance with the applicable edition and addenda of Section III of the ASME Boiler and Pressure Vessel Code.
Additionally, the safety valves have been designed to be functional afterjexposure to loads resulting from the maximum hypothetical earthquake.
for the TMI-l site.
As required by Article 9 of the Code, the safety valve relieving capacity has been provided so that the pressure limitation l
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specified in the Code.will be maintained under all of the system transients or accidents postulated to occur. The power operated relief valve (PROV) is a-pilot operated valve and does not replace a code required safety valve nor does it contribute to the Ccde required re-lieving capacity for the reactor coolant system.
However, the PORV was designed to the same ASME Code requirements as the safety valves as it relates to pressure boundary integrity.
Q.8.
What is being done to demonstrate that the safety and relief valves at TMI-l can withstand the loadings resulting from tnese flow conditions?
A.
A test program has been initiated by the Electric Power Research Institute (EPRI) which will address safety and relief valve operability.
Metropolitan Edison Company (hET-ED) in the TMI-l Restart Report has committed to participating in this test program and has as one of its objectives to satisfy the long-term requirement on SRV testing a3 set forth in Section 2.1.2 of NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations."
In the staff's SER on TMI-l restart (NUREG-0680), the staff requires that MET-ED justify that the EPRI test program is applicable to the TMI-l SRVs.
Should this program demonstrate that these valves are not qualified for the above stated loadings the staff will require the licensee to take corrective actions.
Q.9.
When will the test program be completed?
A.
Prese5t schedules indicate that this testing will be completed by July 1, 1981.
Q.10.
Would the health and safety of the public be endangered should TMI-l be allowed to restart prior to completion of the EPRI test program?
A.
An analysis of a stuck open PORV has been performed (see NRC testimony by W. Jensen in response to UCS Contention #6) and the rasults showed that no fuel damage is predicted to occur.
In addition, the following measures have or will be implemented by the licensee prior to restart to lessen the severity of a stuck open PORY:
(a) if the PORV should fail open, sensors which will be installed prior to restart at the PORY discharge will allow the operator to determine if the PORY is open or shut (see TMI-l Restart SER NUREG-0680 pages C8-ll to C8-13);
(b) TMI-1 Small Break LOCA Procedures requira the PORV block valve to be closed early in a LOCA; (c) the PORV and PORV block valve are all powered from emergency busses as part of the originally approved TMI-l design and therefore meet short term lessons learned Item 2.1.1 (see TMI-l Restart SER NUREG-0680, pages 8-8 to 8-9);
(d) small break LOCA emergency procedures have been upgraded at TMI-1 and have been approved by the NRC (as discussed on page 1-15 of NUREG-0680).
Furthermore, as stated on Page 2-1 of NUREG-0565, "with the increase in PORY lift setpoint, the reduction in the setpoint of the high pressure reactor trip and the addition of the anticipatory reactor trips, lifting of tWh PORV is not likely to occur for the loss of feedwater and turbine y trip tra:.n c..ts."
Thus these valves will be challenged considerably less.
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,. This has been verified by operating experience since there have been 20 transients (as of 6/80) which would have, with the old setpoints, opened the PORV and did not with the new setpoints. The lessening of challenges to the PORVs provides reasonable assurance that PORV failures will be greatly lessened.
With regard to the safety valves, there is presently no evidence that these valves will not operate properly during the anticipated transients which produce two phase flow and solid fluid flow.
In fact the transient which occurred at the Crystal River Nuclear Unit on February 26, 1980 l
provides evidence that the safety valves would perform their intended l
functions under these load conditions. The Crystal River facility has a B&W nuclear steam system and components similar to those at TMI-1.
Based on the above considerations, operation of TMI-l prior to completion of the EPR1 test program would not endanger the health and safety of the public.
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PROFESSIONAL QUALIFICATIONS OF JOHN J. ZUDANS My name is John J. Zudans.
I am currently employed by the U.S. Nuclear Regulatory Commission as a Senior Mechanical Engineer, Equipment Qualification Branch, Divis. ion of Engineering, Office of Nuclear Reactor Regulation, (NRR).
Prior to the NRR reorganization I was a member of the Engineering Branch, Division of Operating Reactors, NRR.
My duties and responsibilities include the review and evaluation of structural nachanical aspects as related to safety issues involving equipment qualification in nuclear reactor facilities being licensed or operating.
I am specifically involved with rechanical and environmental qualification of pumps and valves.
In this capacity I am responsible for evaluating purge and vent valve opera-bility for all operating reactors, deep draft pump operability, and I am also involved in reviews of various active safety related components such as relief valves, block valves and their associated equipment.
I am a graduate of Villanova University with a Bachelor of Science Degree (1970) in mechanical engineering.
I am also attending the University of Maryland towards a masters degree in mechanical engineering.
Prior to my appointment with tne NRC, I was employed by Stone & Webster Engineering Corp., Cherry Hill, N.J. (1974-1976) and Ingersoll-Rand Co.,
Pnillipsburg, N.J. (1972-1974)
My dutiek as a Principal Engineer at Stone & Webster included tne design and.~
analyses of containment structures and attachements thereto.
While employed at Ingersoll-Rand Co., my duties included the design, analyses and testing of pumas
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i used in the U.S.' Navy nuclear submarine program.
Operability and reliability j
of these components was a key requirement in performance of my duties.
Professional Societies American Society of Mechanical Engineer Member of the ASME Commt.ttee on Operation and Maintenance of Nuclear Power Plants-WG on Inservice Testing of Pumps and Valves.
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f OUTLINE This testimony of John J. Zudans contains the NRC Staff's response to UCS Contention 6.
The purpose of this testimony is to demonstrate that, contrary to the assertions made in the contention, additional qualification testing of reactor coolant system relief and safety valves is not required to provide reasonable assurance of no undue risk to public health and safety.
Conclusions to be drawn from this testimor.y:
-- Except for verification testing of their ability to withstand loadings result-ing from two-phase and solid-fluid flow, reactor coolant pressure boundary safety and relief valves meet the Staff's cut rent interpretation of the requirements of GDC 1, 14, 15 and 30.
-- Such verification testing is presently scheduled to be completed by July, 1981.
-- Analyses of the consegeunces of a stuck onen PORV predict that no fuel damage will occur.
-- Improvements in design and emergency procedures to be completed prior to restart will decrease the likelihood of PORV failure.
-- The recent transient at Crystal River provided evidence that the safety valves will perform properly under two-phase flow and solid-fluid flow conditions.
-- Operation of TMI-1 prior to completion of the verification testing will not endanger public health ano safety.
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