ML20081B187
| ML20081B187 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 10/24/1983 |
| From: | Lewis M LEWIS, M. |
| To: | Atomic Safety and Licensing Board Panel |
| Shared Package | |
| ML20081B075 | List: |
| References | |
| NUDOCS 8310270356 | |
| Download: ML20081B187 (9) | |
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WITED STATTS CF A: ERICA NUCIZAR REGUIATORY C')K4IS2 ION 00CKETED 3*f ore the Atonio i Safety and LicenDine n 3 card USNRC In the Patter of Docket nod 500M326ndA$2.2 Philadelphia Electric Company (IGS Units land 2
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Statement of Parvin I. Iewis In support of His Response to Applicant's
- iotion for Summary Disposition of Contention I-62.
Q.l.
State your name.
A. l.a.Marvin I. Iewis. A stat 6 ment of ny professional and other qualifications is attached and entitled " Resume' q.2.
You are familiar with Contention I-62. Answer yes or no.
A.2.a. Yes.
Q.3. '4 hat sections of the FSAR for IGS are pertinent to your Contention I-627 A.3. Although several sections of the FSAR areraated to the considerationc of thermal shock, there are presently very few parts of the FSAR which bear directly upon my belief that' the PTS in the IGS has not been properly analyzed.
Perhaps FSAR Section 4.3 2.8 comes closest to answering my concerns, but it is also deficient in several material and icportant charmeteristics.
Q.4 Please define FIS.
A.4. FIS is a condition that may' affect some P*as and 3'4Rs.
PTS results from the introduction of cold coolant onto, into, or close to a hot pressure vessel while pressure is or becomes high. Thermal stresses are go?.uced in the pressure tenperature boundary when cold coolant is introduced into, onto or proximate to the RPV. These thermal stresses, in conjunction with stresses which occuras a result of high vessel pressure, chugging load, vibration, seismic 1 cad, i
l and any other cumulative effect, have the potential to cause crack propagation in vessel materials and materials in the. pressure temperature boundary. The materials of which the reactor pressure vessel is made can become embrittled as a result of substantial neutron bombardment.
'4 hat constitutes substantial neutron bombardment is still a matter that requires further research. This embrittlement could adversely affect the ability of the RPV materials to withstand all the combined and additive or cumulative stresses which exist in a RP7.
Frs has been recognized as a probles in some - FJas because 1.) Very high tem;eratures have been observed in FJas during rapid cooldowns.
8310270356 831024 PDR ADOCM 05000352 G
Q,. 5 Describe why you belie"e that FIS is significant in 3'4Rs such as the IGS.
A.5 FIS can be a significant problem for FJEs since the necessary ingredients--
high reactor pressure combined with thermal and other stresses and :letermination of what constitutes significant neutron embrittlement -- have not been adequately analyzed or in some cases ignored. Specific reasons for these statements include
, but are not limited to, the following:
- 1) The pressure in a 3'4R foll1ws the water steam saturation curve. However, those instances and scenarios wherein the water steam saturation curve does not enter the temperature pressure determination are ignored. Two ; articular examples come to mind:
a)At Indian Point 2 on 10-17-30,the RTV was submerged in 9 feet of.. cold river water.(lThis event is especially significant to the IGS. A previous contention submitted to the Board by this intervenor was dismissed partially on the basis that the IGS site had a very large reserve of water. (2 b) The possibility Sf stratification within the RPV has not been fully analyzed.
This stratification could allow a steam bubble to form at the top of the reactor.
The water at the bottom of the reactor would stratify with the hottest water in contact with the steam at the top of the reactor. Any circumstance wherein water circulation would be adversely affected could give rise to this set of circumstances.
The insulating properties of the steam in comparison to wat er or water steam mixture could then provide an "unanalyzed Reactor Stress During Cooldown." (3.
There are many possibilities for a F4R to experience FIS that have not baen analyzed.
The above are only two.
Nonetheless, they demonstrate that the thermal arxi combined stressec in a F4R have not been adequately considered for FIS.
- 1) Nuclear Safety ?a6 azine Volume 24-4 Jul Aug 33 Fressure Vessel Thermt.1 Shock:
Experience at US F4Rs 19631981 " by D.L.Phung and '4m. 3.Cottrell.
- 2) Contention I-57 (Lewis) There is an insufficient inventory of water on site or in the F4ST to provide adequate assurance of cooling in the case of an SDV pipe break.
3/ "UNANALYZED REAcr0R VEESEL THERMAL STRESS DURING C00LDOWN(3N 83 h2)" April 12,1933, Eisenhut to Ccamissioners.
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2)The nsutron fluanca at the vassel wall in a 3XR should be low conpared with a FWR because of the presence of a large water filled annulus, a 2" chroud, and a substantially lower reactor core power density. However, several competing considera, ions were not included in the LG5 analysis of neutron fluence.
These t
competing considerations include, but are not limited to, a) fuel management considerations in a BWR consider hig burnup rates as a primary consideration. (1) F.las have been recently required to consider minimization of neutronfluxt3theRPVwall.(2)
These two points taken together give rise to a concern that the fluenca at a 3VR vessel wall may ba greaterthan a nominal flux obtained thru colculational techniques.
b)the water filled annulus in a SWR may not be completely filled with water. A steam water 2 phase, system could also exist in tre annulus. Over time a steam water attenuation can ba substantially less than a single phase water attenuation of theneutronflux.(3) c) the shroud may not be continuous. There has been a history of bolts com;ing off non-safety related ; arts in reactors and other debris breaking loose.
(4)
Whether of not an hiatus in the shroud could occur at a dangerous point as far as neutron embrittlement is involved and whether or not that hiatus can endanger the toughness of the RPV materials are matters that bear analysis.
(1) NUCIEAR REACTOR ENGINEERING 3rd Edition 5. Glasstone and A. Eesonske Para 3.I33 and 8.198.
(2 ) Enclosure A,NRC Staff Evaluation of FIS Nov 1982 Para 9.4 Page 9.4 and Appendix I Flux Redsetion Programs.
(3) N R E (see(1)above) Para.1.102.
(4) Regulatory Guide 1.133, Loose Part Detection Program for the Prinary System of INRs. This Reg Guide grew out of a history ond need caused by parts that broRe off of non safety related and occasiinally safety related equipment.
Intervenor respectfully brings the Board's attention to the fact that he submitted comments on this Reg Guide in 10-15-77, and is therefore very familiar with the Reg Guide history, l
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The dssign, construction, testing, peration and surveillance, togethsr with the phycical b2havior of ths 3'ia assures that E is not a problen for the E only if competing conoiderations that can make ITS a problen at 3MRs are ignored.
Q.6 and Q.7 Please describe the codes and standards to which the 103 apV are designe d and fabricated.
A.6 and A.7 The statement of Sapath Ranganath in support of Motion for Sums:7
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Disposition of Contention I-62 describes the codes sufficiently. There is n?
need to repeat it here.
Q.3.
In the " Statement of Sampath Ranganath In Suport of y.otion for Summary Disposition of Contention I-62" a statement ht
" thesecaterials were tested to the augmented requiremants specified by GE" appears. Do you know what these
" augmented requirements are, and why thy were " augmented"by GI?
A.3 No, but I would sure like to. This is definitely an area to explore in cross examinatinn.
Q.9 Again in the "Statene nt..." what specifielly does" operational experienca" refer to?
A.9 Cften " operational experience"will not really refer to experience at operating commercial nuclear power plants of the same or similar design. Sometine "operatioral experience may merely refer to calculational techniques using other computers. Unless " operational experience"lalearly defined as to reactor, placement in reactor, operatin6 history in reacter and many other variables, " operational experience "can mean almost anything.
"Operatonal experience " may or may not bear directly on this contention. Any answer that depends on an undefined " operational experience " should not be relied upon at this stage of litigation.
A.10 '4 hat specifically does the intervenor believe is inadequate or lackin6 in the Applicant's determination of fluences?
A.10 The Applicant admits that the fluences are just calculations. This is l
as it should be in light of the incompleteness of the IGS. However, the l
Applicant believes that "these calculations have been compared to field measurements and found to conservatively (sic) overpredict the neutron flux."
This overprediction is most probably for only a few points in some reactor where there are neutron dosimeters.
I as particulady troubled about points in the l
IGS that may not be adequately monitored by dosimeters. There is no reason to l
l believe that the calculations, which are admittedly only representative,will hold sway for all important welds and structures in the IG3,
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q,11% hat ara tha possible effects of those points which are not adequately monitored by dosimeters in the IG3?
A.11 As described in the Intervenor's answer to Q.5 above, there are considerations that allow a point to be bombarded by higher fluxes than calculated by the Applicant.
There has not been sufficient analysis to assure that hese points will not sustain significant neutron fluence which can cause crack propogation during a IT7.
Q.12 The Applicant has performed " confirmatory fracture analyM s." Do you have any apprehensions about fracture analysis?
please state them.
I'**
A.12 I have several apprehensions about fracture analysis:
- 1) Fracture analysis is a highly empirical calculational technique. It depends upon choosing the right parameters and equations. These parameters and equations change with size ard other variables. presently the fracture analysis has never been verified with a full scale test. The largest size of pressure vessel used to verify the fracture analysis calculations have been 1/6th scale models of WR vessels. There are many differences between WRs and F4Rs.
Inlightofthefactthatonly1/6thscalemodelsofIWEshavebeenusedto verify the fracture analysis and that there are many differences between F4R and Was, I at most apprehensive about accepting fracture analysis and fracture mechanics calculational techniques as adequate.
- 2) The assumption about crack depth, crack tip geometry, and crack propogation seem to jump out of the air and do not appear to have substantive justification. (1)
(1) NIP 2G 07 % Vol 1 Rev 1 Resolution of Task A 11 Reactor Vessel "aterials Toughness Safety Issue Appendix H discusses the 1/6th scale models used.See 1) above.
j Comment 12d "The most important omission in the criteria is the reference size to be l
used in the calculations. TheJ/Tequals50linesuggestsanassumedflawdepthof 0.25 inches."
l Please note how this consent is phrased,"most important", " suggests", dassumed."
i These are my problems in accepting the fracture analysis. Further, this intervenor did not send in these comments so these comments were not planted by this intervenrr, but are real concerns of other interested parties. See 2) above.
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'4 hat particular concerns do you.* ave about the neutron surveillance program et th3 IGC7 A.13 This is a very sore subject. Part of the original reason that FIC was brought forward as a problem is that tha predicted veseel fluence at several rer ctors was much less than the fluence actually experienced during operation.
The Applicantlas stated that his calculations " conservatively overpredict... com;ared to field measurements." Nonetheless, there have been several instances where experienced and calculational fluences have not compared favorably. Also ORNL states,"(Theestimate1 uncertainty)ofFluences at. the vessel wall locations may be as high as 505."
(1)
This is only part of my concern about fluences and the actual differences may be much higher as I pointed out in my answer to A.10.
Q.14. '4 hat are your overall conclusions concerning the effectof FT5 on th IG RPV?
A.14 The conditions necessary for the ocurrence of PTS on P4Rs can occur at IT. The only way that* an opposite conclusion can be reached is to ignore all pertinent and important facts and history.
- 1) Evaluation of the Threat To PRR Vessel Integrity Posed by FIS Events.Kryteret al ORNI/ TM 8072 INURm /CR 20B3 Oct. 71931 Page 6.6 Para 6.3 l
l Although I use a quote that suggests that the uncertainty is limited to D, I am not emphasizing Table 6.1." Uncertainties for calculational and uosimetry Measurement frocedures in I' Rs" wherein a simple addition of the known uncertainties leads i
to a figure for uncertainty well in excess of 100%.
United Etates if A 3rica Nuclear Seplatory Commission Sofore the Atomic Safety and Licensing Board 00CKETED In the Matter of
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AFFIDAVII' 0F MARVIN I. IF'4If,3.P.E.,INTERVENO3 AT THE ICS OL IEAREd$'N' L.
My name is Marvin I. Iewis. I am a Registered Professional Engineer adnd and intervenor at the IGS OL Hearings before the Nuclear Regulatory Commission Atomic Safety and Lincensing 3 card.
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I have written the enclosed response to tre Applicant's Motion for Sunnary Disposition of Contention I-62.
The statements are true and correct and complete to the best of my knowled e, information and belief.
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Date Parvin I. Iewis Subscribed and Sworn to 3efore me this date cI f N 'u.) > -71/
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Marvin I fewis 0
Tyg Address: 6504 3radford Terrace Phila PA 19149 Telephone :
(215) cU 9 5964 (Home)
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Brat;cy 3.A.
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Metallurgical Engineering Drexel Institute of Technology 1960 Graduate work Fngineering University of Panna 1961 Graduate work Chemistry St. Joseph's 1964 thru 1963 Professional License:
Registered Professional e gineer No. 011729-I 1960 to present n
Technician exterience:
U.S. Naval Materials :aboratory Ed 599 Phila Naval 3ase 1956 thru 1958 N.Y. Shipbuilding Corp.
Camder N.J.
1958 thru 1959 Professional exterience:
General Electric Company 1960 1963 Philadelphia Board of Education 1963 thru 1971 Enerry related experierice:
Environmental Coalition on Nuclear Power Action Director 1971 thru 1976 Board Member 1976 to present Citisen Action in the North East (CANE)
Energy Chairnan 1992 to present
11.
I have been active in energy-related concerns over a decade.
'>aring this time, I have provided many environmental and citizen-action groups my expertise and time on many energy related subjects.
I have also appeared on television and radio shows including several appearances on Pennsylvania perspective with Joe Funter and a regular talk show on WISP F.M.
I have also intervened either for myself or citizen action groups in proceedings before the Nuclear Re6ulatory Commission. Department Of Energy, and the Pennsylvania Public Utility Commission.
y.y formal consents to the Departnent of Energy, Nuclear Regulatory Commission and the Invironmental notection Agency have t*en accepted and incorporated into the respective regulations on the subjects of transportation of nuclear wastes and radioactive waste gas systems.
Previous to my involvement with energy, I was employed as a materials engineer with the General Electric Compny. l4y employment with GI was at several different locations and in several different capacities.
I started in a Thermophysical Properties Iaboratory. I organized the laboratory from the start and was in charge of it.
I was transferred to several locations within GE to write materials specificaticcs.
I wrote materials specifications on all manner of materials from astronaut drinking water to superinsulation.
As an undergraduate at Drexel, I worked as a technician. In the Naval.%teriAls Iaboratory, I worked -
in the. Metallurgical Section.
I investigated and performed tests on ferrous, non ferrous and ceramic materials.
I also worked as a welding inspector for the New York Shipbuilding Corporation. Aside from the magnaflux inspection of welds, I also ultMonically inspected the reactor vessel for the NSS Savannah.
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