ML18082A197
| ML18082A197 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 04/10/1980 |
| From: | Pasedag W NRC COMMISSION (OCM) |
| To: | |
| Shared Package | |
| ML18082A196 | List: |
| References | |
| NUDOCS 8004180416 | |
| Download: ML18082A197 (9) | |
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD tn the Matter of
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PUBLIC SERVICE ELECTRIC &
GAS COMPANY Docket No. 50-272 (Salem Nuclear Generating Station, Unit No. 1)
Proposed Issuance of Amendment to Facility Operating License No. DPR-70 Question:
Answer:
DIRECT TESTIMONY OF WALTER F. PASEDAG IN RESPONSE TO BOARD QUESTION NO. 5 In the event of a gross loss of water from the spent fuel storage pool at Salem 1, what would be the difference in consequences between those occasioned by the pool with th:
expanded storage proposed by the Licensee and those occasioned by the present pool?
The Staff has reviewed the potential for a gross loss of water from the present and expanded spent fuel pool at Salem.
Our review has identified no credible mechanism for a loss of water from* the pool which would result in any substantial off-site dose consequences.
The spent fuel pool consists of a reinforced concrete basin of wall thickness exceeding 8 feet on all sides and 24 feet at the bottom.
The entire pool is lined by a 1/4 inch steel liner.
The pool integrity under all postulated accident conditions was reviewed at the time of licensing.
The additional structural loading resulting from the pool expansion is well under 1% of the total lumped mass in the fuel handling building analytical model, and, therefore, does not appreciably change the structural response of the spent fuel pool.
The walls have been investigated for the seismic effect of the heavier racks and stored fuel.
The ~igh density racks have no appreciable effect on the structural stability and seismic response of the spent fuel handling building.
Therefore, the leaktightness of the expanded pool under all postulated accident conditions is assured, and no appreciable change in the margin of protection arises from the pool modification.
The pool design includes a weld channel leak collection system which is intended to collect any leakage of the liner welds.
After collection in the weld channels this leakage is piped to the radwaste system via ten (10) one inch diameter leak-off tubes which discharge to a radwaste drain.
The largest credible leakage from the spent fuel pool would occur if all 10 leak-off tubes were to discharge at their maximum capacity. This scenario requires multiple punctures of the spent fuel pool liner, and therefore, is considered highly unlikely.
A maximum leak rate of no more than 710 gpm could occur in this case, resulting in a rate of decrease in the pool water level of l.l inches per minute.
This leakage would be detected by the indication and alarm of the leak collection sump in the control room and result in the automatic operation of the sump pump.
In addition, prolonged leakage would result in a low spent fuel pool water level alarm.
Following detection of this leakage the tubes could readily be capped to withstand the maximum back pressure of 19 psig.
The potential radiological consequences from any accidental release of water from the spent fuel pool would be directly proportional to the fission and activation product concentrations in the water.
In our Environmental Impact Appraisal of the Salem spent fuel pool modification (Staff Exhibit 6C, Section 5.3.1) we concluded that the additional release of radioactive material to the spent fuel pool water resulting from the additional stored f~el is insignificant.
I Consequently, the difference in radiological consequences of a spill of thi~
water would also be insignificant.
- We also have evaluated the differences in the liquid pathway between the Salem site and the typical site evaluated in detail in the Staff 1 s Liquid Pathway Generic Study (NUREG-0440) in order to determine whether special r
site-specific factors might be present at the Salem site.
We examined the groundwater transport, surface water transport, and usage of the water bodies surrounding the Salem site and found that the Salem site compares favorably with the typical estuary site of the Generic Liquid Pathway Study.
Our evaluation indicates slower dispersion of postulated releases via the liquid pathway compared to the typical estuary site of NUREG-0440.
We conclude, therefore, that there are no site-specific peculiarities with respect to the Salem site which would invalidate our conclusions concerning liquid releases stated in the Environmental Impact Appraisal.
In our attempt to define the meaning of a 11gross loss of water 11 we have also considered a hypothetical, non-mechanistic, instantaneous loss of all cooling water in the present and expanded spent fuel pool combined with an inability, for unspecified reasons, of refilling the pool, or providing any other mode
- of cooling other than natural (convective) air cooling.
In view of the thorough review of the integrity of the spent fuel pool, even under design basis earthquake conditions, such an event is considered incredible, and clearly exceeds all design bases.
Accordingly, such an event should be classified as a 11class 9 accident 11
- For fresh spent fuel, continued denial of water cooling capability may eventually lead to oxidation and failure of the clad, and to overheating of the U02 fuel, with the potential for the release of the fission products in the U02 fuel in either the present or the expanded pool.
The doses at the site boundary resulting from this postulated release would depend heavily on the postulated scenarios for the mechanism of the water loss, subsequent cooling attempts, building integrity, etc.
In order to estimate the differences in the potential consequences of this hypothetical event arising from the pool modification, the onset of self-sustaining clad oxidation may be used as a conservative criterion for the release of the fission products from the fuel.
A detailed calculation of the heat-up of spent fuel in various configurations is given in a report by Sandia Laboratories (Spent Fuel Heat-Up Following Loss of Water During Storage, NUREG/CR-0649).
From this report it is apparent that PWR fuel in the configuration of the modified Salem storage racks cannot reach temperatures for self-sustaining clad oxidation if its age (since removal from the reactor) exceeds 280 days.
Since the additional fuel stored in the expanded pool would be at least four years old, as described in section 5.3.l of the Staff's Environmental Impact Appraisal, no additional clad failures, and hence no additional releases beyond those expected from newly discharged fuel would occur as a result of the SFP modification.
Based on the foregoing considerations we reach the following conclusions concerning the relative effects of the Salem spent fuel pool modification:
(1)
Th~ worst Credible loss of wa~er from the fuel pool would occur if the spent fuel pool liner were punctured simultaneously in ten locations such that all ten leak-off tubes would discharge water at their maximum capacity.
Because of the multiple failures which would have to occur to
- realize this scenario, this event is considered highly unlikely.
Our evaluation of this event indicates that there are no substantial differences in the radiological consequences arising from the modification of the pool.
(2)
A loss of all water from the pool is not considered credible, and would exceed all design requirements for the present and expanded spent fuel pool.
If no mitigation of this hypothesized event is assumed, the.
radiological consequences could be large, as a result of possible overheating and clad failure of any newly discharged fuel in the pool.
These consequences could occur either with the present, or with the expanded pool.
A detailed comparison would require specification of a scenario for the loss of water and make-up capacity.
However, we conclude that any additional fuel in the pool as a result of the pool modification would not contribute to the consequences of this event.
(3)
The expansion of the spent fuel pool at the Salem site does not constitute an exceptional case with respect to the liquid pathway or design features of the spent fuel pool resulting in risks substantially greater than for an average plant.
Therefore, we conclude that the environmental consequences of Class 9 accidents need not be evaluated.
Statem*nt of Qualification*
Dr. Allan S. Benjamin April 2, 1980 6uril\\9 th* paat ~e* year*, I have been employ9d as ~PhD Member of th* Teehnica1 Staff in th* Nuclear Fu*l Cycle Safety Reaearch Department at Sandia National Laboratori**
iri Al.buqu*rque, New Mexico.
My respona.lbilitie* include project aanaqement and lead technical accountability for two r***arch proj*eta of high priority regarding improved aafety in light water reactor*.
Th* f ir*t project examines th* desiqn and *Y*tem implication*
and phenomanolo9ieal aspects of filtered-vented containment systems ** a ~aa.n* for mitiqating th* cona*quance* of core meltinq accident*.
Durinq th* paat three months, I have alao been asai9ned t.h* technical lead in assessing the f*a*ibil~ty of retrofittinq auch systems into th* Zion and Indian Point reactors.
I have been the author of two Sandia reports on the *ubject, a principal author of articles published in the Journal of the Inetitute of Electrical and Electronic* Enqineers (IEEE Spectrum) and the Transaction* of th* American Nuclear Society, and pr***nter of an invited paper at th* Atomic Industrial Forum Workshop on Licensinq and Technical Issue*-
Poat TM.I.
The *econd proj*ct that I am currently involved with inv*sti-9a~es experimentally the chemical forms and transport behavior of fission product* in reactor primary *Y*t~* during core meltinq accidents, aa a means for determining the amounts and forms of radioactive material that e*c*p* to the contaiND*nt.
In conjunction with thi* activity, I recently **rved on a planning coamtittee of industry r*presentativea to recommend data qat.hering prioritie* for the Three Mil* Island reactor.
At Sandia, I have also been principal inveatiqator for a number of oth*r reactor-related safety studies. includinq (1) an *****mnent of *pent fuel h*atup following lo** of water during storage, (2) the modelin9 of heat t.ranster phenomena durinq core-concrete interactions that snay occur durinq core m*lbfovn accidents, and (3) a *urvay of contain-ment analy*i* procedure* for liquid metal fast breeder r*actor**
Hy analy*i* of the spent fuel problem *panned 1-l/2 years reaultinq in **veral publicati.on* (a Sandia report, a paper in Nuclear Technoloqy. and two paper* in th* ANS Transactions).
Th* work on core-concrete heat tranafer interaction* i* also beinq publi*h*d and presented at several forum* *pon*ored by the.A1!lerican Nuclear Society/European Nuclear society.
Prior ~d Sandia, % was employed for eleven yeara *~ ~RW Sy*~enm, Redondo Beach, Cali~ornia, wh*r* I performed analy*** related to re-entry vehicl* teehnoloqy in th* followinq areaes
... (1) aerodynamic h-t tran*fer to bodi** of r*volution. (2) non-i*othe:mal wall effects on heat transfer, (3) radiant h*at
~anafer, (4) boundary layer tran*ition. (5) conc!uation in on*, two, and thr** 41.Jllenaiona, including use of thermal analyser aethods, (6) ablation and ph*** chanqe ph*nomena
. nlatin9 to h.at *hi*ld*' (7) change of *hap* during ~*-entry, (8) rain and duat *ro*ion. (9) vi*coua and inviacid auper*onic flova, (10) drag determination, (11) trajectory an.alyais, (12) l***r effect* on heat shield*, (13) tranapiration cooling. (14) evaluation of exparim*ntal facilities, (15) analyaic ot experimental data, and (16) computer **thod* for l*rv*-*c*l*
data handling.
From 1972 to 1975, I va* project manager of
~* Erosion Data Analysis Proqram for the Air Fore* Material*
Laboratory and th* Space and Miaail* Syatmns Organizatian.
In 1977, culminating a six-year work-study proqram in which I Va* aided by a Coop*rative Fellowship ~rom TRW Systems, I wa* awarded a PhD degree from UCLA with a 111Ajor in heat and **** tran*t*r and minors in fluid mechanic* and probl.un
- olv1n9/d*ci*ion makincJ.
My ov*rall 9rad* point average was 3.86 ou~ of a possible 4.oo, the only B being in an elee~ive coura* in physioloqy.
Hy 4i***rtation con*i*t*d of a detailed atudy of numerical met.hod* for *olving incompre**ible, reoir-culatinq flow problems at high Reynolds numbers.
The re*ult*
h*v* b**n publi*h.d in th* Journal of Computational Phy*ic*
(December 1979).
Prior to UCLA, I obtained th* Bachelor of Science degr.. in 196<< and the Master of Scienc* deqre* in 1966 from Bratin\\
Oniverai't.y, Providence, Rhode Island.
At Brown, I snaintained grade point averages of 3.53 ** an underqraduate and 3.62 **
a 9raduate student, and I received ~y diploma with the distinction *ma~na cum laude.*
I vaa *l*cted to Sigma Xi and Tau Beta Pi National Honor Soc:ietiea, wa* awarded a aeholar*hip from Alcoa and a fellow*hip from NASA,, va*
designated a Francis Wayland Scholar, and wa* a recipient or the Brown Enqineerinq Amaociation Prize.
Ny ma*t*r* thesis Va* on th* *ubject of heat tran*fer aero** turbulent, ineom-pr***ibl.e boundary layer* over both amooth and rou9h aurfacea.
I am currently a Member cf thw.AIB*rican Society of Mechanical En9in**r* (ASHE).
PROFESSIONAL QUALIFICATIONS WALTER F. PASEDAG Environmental Evaluation Branch Division of Operating Reactors I am employed as a Section Leader in the Environmental Evaluation Branch*,
Division of Operating Reactors, U. S. N.uclear Regulatory Commission, Washington, D. C.
My duties are to provide technical supervision and review the work of personnel assigned to my section.
My responsibilities include planning, coord-inating, and reviewing the safety and environmental evaluations of modifications to reactor facilities in the areas of external hazards, radiological accident analyses, radiatton protection, and overall environmental assessments and analyses.
I am also responsible for providing guidance for and technical review of various documents including Safety Evaluations, Environmental Assessments, and Environ-mental Statements and for providing guidance for the development of technical positions for reactor and site standards, codes, and criteria associated with programs assigned to the* section.
I recetved a BS degree. in Engineering Science from the State University of New York at Stony Brook in 1966~ The sam~ year, I joined the Graduate Student Training Program at the Westinghouse Educational Center in Pittsburg, Pennsylvania.
My professional experience in the nuclear power industry i.ncludes six years of activity in the nuclear safety field at the Atomic Power and Nuclear Energy Systems Divisions of Westinghouse, and, under a technical exchange agreement, at the Reactor Development Department of Siemens A. G., Germany.
During this (period,
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- from 1966 to 1972, I have held positions of increasing technical responsibilities in the areas of radiatton shielding and personnel safety, emergency core cooling systems design and evaluation, including the development of thermal-hydraulic models of the primary system during the reflood phase of the loss-of-coolant accidentf development of design criteria for ventilation and filtration systems, and the design and analysis of post-accident fission product removal and control systems including containment spray and dual containment systems.
I joined the Atmoic Energy ColTTllission as a Nuclear Engineer in the.
Accident Analysis Branch in 1972, where I was responsible for the development of analytical models for the evaluation of engineered safety features relating to radiological safety and the review and evaluation of power reactor plants in the area of radiological and site safety.
I am the author of several reports and articles concerning the effective-ness of various engineered safety features for U/R and HTGR reactors which are published in the technical literature.
I am a member of the American Nuclear Society, and the Standards Committee of the ANS.
I am a licensed Professional Engineer registered in the Commonwealth of Pennsylvania.