ML20135E158
| ML20135E158 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 02/28/1997 |
| From: | Zwolinski J NRC (Affiliation Not Assigned) |
| To: | Linnell W S AFFILIATION NOT ASSIGNED |
| Shared Package | |
| ML20135E162 | List: |
| References | |
| NUDOCS 9703060300 | |
| Download: ML20135E158 (8) | |
Text
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n UNITED STATES g
j NUCLEAR REGULATORY COMMISSION o
WASHINGTON. D.C. 20066 0001 8
%,.....,o February 28, 1997 Mr. William S. Linnell, II Spokesperson Maine Safe Energy P.O. Box 4034 Portland, ME 04101 l
Dear Mr. Linnell:
I am responding to your letter to Chairman Jackson dated January 30, 1997, concerning your participation in the Commission meeting on Maine Yankee.
In that letter, you provided advance copies of slides to be prr.sented at the meeting and also included two questions to which I am respording.
In your first question you asked the approximate inventory cf radioactive material (RM) (in curies) produced annually by the Maine Yaniee plant, and the inventory of RM in the Maine Yankee spent fuel pool.
RM is produced in nuclear power plants as a result of the fission of nuclear full in the reactor. The inventory of curies produced is related to the amount of fuel burned.
Each operating nuclear reactor has been licensed by tie Commission to minimize the release of radiation to the public/ environment in accordance with applicable regulations, based on an analysis of the maximum amo1nt of RM generated at the end of a reactor core life.
For this conservative design basis accident analysis, worst case assumptions are made (including the number of curies of RM produced). The resulting calculated doses for a spectrum of hypothetical accidents must be below regulatory limits.
After a plant has been licensed, the Commission does not require an annual report of the total RM produced in the reactor or contained in the spent fuel pool. However, monitoring and reporting is required for the amount of radiation that is released to the environment, along with the resulting public doses.
An estimate of the amount of RM (in curies) that is produced by a 1000 MWe nuclear reactor was included in NUREG-75/014 (WASH-1400), dated October 1975.
In Table 3-1 of this document, it listed that for the reactor core abopt 1/2 hour after shutdown, the radioactive inventory would be about 8.1 x 10 curies, and that in the spent fuel pool storage pool with an inventory of 2/3 ofacore1caded,1/3corewith3-daydecayandIf3corewith150-daydecay the radioactive inventory would be about 1.3 x 10 curies.
Note that the RM in curies significantly decreases over time with natural decay rates for the various elements.
In addition, inventories of RM were estimated by Brookhaven National Laboratories for Millstone Unit 1 (a Boiling Water Reactor) and for R. E. Ginna (a Pressurized Water Reactor) plants for different decay times.
The results are presented in NilREG/CR-4982, " Severe Accidents in Spent Fuel Pools in Support of Generic Safety Issue 82," puolished in July 1987.
Copies of these referenced NUREGs will be forwarded to you ir,' separate enclosure.
l NRC FILE CENTER COPY OFCA i
9703060300 970220 PDR ADOCK 05000309 H
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s William S. Linnell, II You should be aware that licensed nuclear plant designs include numerous safety systems which provide defense in depth, redundancy, a series of i
physical barriers to radiation release, and fission control. These design features enable the plants to operate in a safe manner with fuel burnup rates that could generate the amounts of RM discussed above.
l In your second question, you asked if a controversy regarding Emergency Core Cooling System (ECCS) testing was resolved and what tests and studies exist to demonstrate resolution.
It was recognized, by the Atomic Energy Commission (AEC) staff in the mid 1960s, that there was a need to improve the design l
capabilities of ECCSs installed in nuclear plants. The " Interim Acceptance j
Criteria for Emergency Core Cooling Systems for Light Water Reactors" were developed by the AEC in the spring of 1971 by a regulatory staff task force, with support from AEC laboratory personnel engaged in related safety research and testing.
The interim criteria were based on the experience that had been gained in the case-by-case evaluation of the ECCSs for over 50 reactors by the AEC staff and the. accumulation of experimental data and information concerning the expected behavior of ECCSs.
One particular body of information considered was that from the semiscale tests performed at the National Reactor Testing Station by the Idaho Nuclear
' Corporation during the period of November 1970 through March 1971. These tests were designed to verify certain assumptions used in mathematical prediction models of some of the events which may occur during a loss-of-coolant accident relative to the performance of an ECCS.
The results of some semiscale water injection tests were not as predicted by the then-available AEC hydraulic code, and, in fact, the injected water failed to flow up through the electrically heated rods (simulating a reactor core) as expected.
On the basis of the uncertainty revealed in the predictive model, the task force recommended that some additional conservative assumptions be included in all evaluation models used in connection with the interim ECCS acceptance criteria. There were, however, reasons for believing that the sepiscale results were peculiar to the equipment used in this test and not completely
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representative of hydraulic behavior in large systems. A number of additional tests were conducted as discussed in NUREG CR-4945, " Summary of the Semiscale Program (1965-1986)." I have enclosed the executive summary from that report i
for your information. A copy of the referenced NUREG will be forwarded in a separate enclosure.
i The interim position (for ECCS systems) was brought into the rulemaking
. process in the early 1970s and was finally issued as 10 CFR 50.46 in its original version in early 1974. This process incorporated lessons learned from the tests that had been conducted.
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Wil'liep 5. Linnell, II.
February 28, 1997 I trust this information is responsive to your concerns.
4 Sincerely, 4
Original signed by:
John A. Zwolinski, Deputy Director i
Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation i
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Enclosure:
Summary of the Semiscale Program (1965-1986) t n
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NAME FMiraalia HThompson SJackson DATE 2/13/97 1/lf/97 A 46/97 0FFICIAL RECORD COPY DOCUMENT NAME:
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1 William S. Linnell, II I trust this information is responsive to your_ concerns.
Sincerely, k
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John A.\\ Zwolinski, Deputy Director Divisioh of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Enclosure:
Summary of the Semiscale Program (1965-1986) j l
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t Letter to William S. Linnell, II, dated February 28. 1997 from John A. Zwolinski.
DISTRIBUTION:
IDsket File 50-309.:(w/oHginalfincoming).'/
PUBLIC (w/ incoming)
OGC EDO #G970075 OPA H. Thompson OCA E. Jordan SECY # CRC-97-0116 P. Norry N. Olson J. Blaha C. Norsworthy H. Miller, RI C. Poslusny (w/ incoming) l C. Paperiello M. O'Brien F. Miraglia/A. Thadani M. Boyle (Email)
R. Zimmerman C. Conte i
B. Sheron S. Varga T. Martin J. Zwolinski W. Travers J. Stolz l
PD I-3 R/F (w/ incoming)
D. Dorman P. Milano NRR Mail Room (EDO #G970075 w/ incoming) (012/G/18)
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NUREGICR494F; EGG 2fiO9 Distribution Category: R2
SUMMARY
OF THE SEMISCALE PROGRAM (1965-1986)
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G. G. Loomis Published July 1987 EG&G Idaho, Inc.
Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20656 Under DOE Contract No. DE AC07-761D01570 FIN No. A6038 ENCLOSURE
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i EXECUTIVE
SUMMARY
i The Semiscale experimental program provided March 1979, emphasis was shifted to the investiga.
J both integral and separate effects non-nuclear tion of small breaks and other abnormal transients l
thermal-hydraulic data bases for light water reactor with the Mods using full height components.
safety issues spanning the time frame 1%5 to 1986.
The Semiscale data base was used by the water i
In over 20 years of testing the Semiscale program reactor research community primarily in two ways:
produced more than 250 experiments at prototypi-first, for exploratory research and second, for code cal pressures and temperatures concerning most development and assessment. Test series that were
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major issues associated with loss-of coolant acci-performed as exploratory research include: alter-dents (LOCA) and other abnormal transients for a nate ECC injection where lower plenum, upper pressurized water reactor (PWR). The data was head, and pump suction injection effectiveness used to stimulate thinking about phenomena asso-were noted; reduced initial pressure large break ciated with these PWR safety issues and as a basis IDCAs investigated the concept of spontaneous 2
for code / data comparison in code development nucleation and the effect of CHF/DNB response; and assessment efforts. This document first cata-steam generator tube rupture concurrent with large logs the various Semiscale facilities and data bases break IDCA (LBIDCA) investigated the effect of I
and then discusses the impact of Semiscale data on additional steam binding caused by the tube rup-licensing (10 CFR Part 50 - Appendix K) issues, ture on reflood phenomena and transient severity; r
and safety technology. Secondly, major thermal-steam generator tube rupture as the initiating event; hydraulic phenomena observed in Semiscale experi-general operator response effectiveness during a j
ments are summarized, followed by a discussion on steam generator tube rupture as the initiating event; i
theimpact of the Semiscale program on code devel-system recovery experiments where primary feed opment and assessment. Finally, this document and bleed effectiveness was analyzed, including sta-catalogs the Semiscale data base as to applicability tion blackout assumptions; effect on small break of the data base for code assessment and develop-IDCA (SBIDCA) transient severity due to various ment purposes.
allowed upper head to downcomer bypass flows; The Semiscale program was conducted for the secondary side breaks; and finally ultra-small Atomic Energy Commission (AEC), the Energy break IDCAs with degraded high presure injec-Research and Development Administration tion (HPI) and the required operator responses to (ERDA), the Department of Energy (DOE), and mitigate the consequences of such accidents.
the Nuclear Regulatory Commission (NRC). The Many of these exploratory research series used i
program was performed by a variety of contractors; commercial PWR Emergency Operating Proce-Phillips Petroleum, Idaho Nuclear Corporation, dures (EOPs) as a guide in performing the experi-Acrojet Nuclear Corporation, and lastly, EG&G ment. Semiscale was capable of performing most Idaho, Inc. The Semiscale program used a series of operator functions such as pressurizer auxiliary facilities called " Mods" to address water reactor spray, safety injection / pressurizer power operated safety concerns. All these Mods were designed to relief valve (SI/PORV) feed and bleed operations, simulate a commercial Westinghouse four loop Si operation, pump restart, steam generator feed l
PWR. Semiscale Mods contained most of the and steam operations, and pressurizer internal major components found in large PWRs; active heater operation. Several separate effects data loops with pumps, steam generators, and a vessel bases were also generated under exploratory with an electrically heated core. Because the system research, including natural circulation, reflood was non-nuclear, extensive and accurate measure-heat transfer, and tee critical flow. 'Ibst series per-ment systems were possible. The non-nuclear fea. - formed to provide data for development and assess-ture allowed fast experimental turnaround.
ment included: basic large, intermediate, and small Regardless of Mod, the scaling approach to model break IDCA signature response. Many of these a PWR was basically power to-volume scaling and areas under exploratory research were also used for the scaling factor was on the order of 1:1705 of a code assessment efforts.
commercial PWR. The early Mods were designed The Semiscale experimental data impacted water to investigate large break design basis accidents reactor safety research by providing data on relative involving double-ended offset shears of cold les conservatism of Appendix K to 10 CFR Part 50 piping. Following the TMI 2 accident in for licensing issues. Semiscale data showed an lii
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overall conservatism and thoroughness of Appen-ity of that concept during a SBIDCA. Techniques dix K assumptions most notably in the area of criti-cal heat flux (CHF) and post-CHF heat transfer, were developed for measuring primary and second.
l, reflood, pump modeling, two-phase flow model-ary side heat transfer coefficients using triplet ther-l ing, and break Dow modeling.
mocouples and steam generator riser gamma Semiscale provided scaled data to help assess the densitometers. Using this technique, existing corre-applicability of using scaled data to infer actual lations for secondary side heat transfer were evalu-PWR expected response, in addition, common ated and found to be inadequate. An important scaling distortions were identified in scaled facili-parameter for transient analysis is secondary mass inventory. Semiscale developed a technique, using ties. Semiscale data aided in the extrapolation of existing measurements, to measure the mass inven-scaled data response for reactor transients to the actual expected PWR response. This was accom-tory in a steam generator secondary during steam-plished by comparing Semiscale data to other ing conditions. However, nor mal measurements are larger scaled (but sometimes less well instru-perturbed by flow effects. Semiscale sponsored research for state-of-the-art measurement develop-mented) experimental data. Semiscale event timing ment throughout its charter. Semiscale instrumen-and scaled magnitudes of parameters compare tation techniques have been widely used by LOFT, favorably with LOFT (1/50), PKL (1/134 scale)
LOBI, ROSA, FIST, MIST, Flecht-Seaset, UPTF, and LOBI (1/500 scale), and actual plant experi-UMCP, CCTF, SCTF, PKL, and THTF programs.
ence (TMI-2 and GINNA transient data). Semi-Measurement development included drag devices,
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1 scale scaling studies in addition to the comparison turbine meters, gamma and x-ray densitometers, i
with other facilities show that Semiscale is in a class 4
of facilities that replicate most first order effects two-phase flow condensing systems, optical probes using STORZ lens and video cameras, and core rod thought important for selected reactor transients, clad temperature measurement devices.
i As a part of this scaling effort, Semiscale experi-The Semiscale program created a detailed data ments formed a basis of an international coopera-base with a great variety of high pressure /high tem-tive counterpart testing program including perature thermal hydraulic phenomena repre-ROSA-IV and LOBI.
Semiscale provided data in a timely manner, sented. Semiscale tests examined such phenomena addressing ongoing safety issues as they arose.
as counter current flow, signature responses to p
blowdown, reflood thermal and hydraulic Safety issues addressed by Semiscale testing response, core thermal and hydraulic response to included: steam generator tube rupture during full high pressure boiloff, condensation effects, two-power operation; preferred primary coolant pump phase flow regime mapping, critical flow in nozzles operation during SBLOCA; station blackout dur-and Tec's, and flooding behavior in pressurizer ing full power conditions; relative effectiveness of surge lines during PORV operation. The Semiscale operator actions during abnormal transients; effec.
data added to the world data base on these items tiveness of natural circulation heat removal during and can be used by thermal-hydraulic researchers in SBLOCA and operational transients; alternate fields other than water reactor safety.
forms of ECCS: transient severity during steam The primary charter of the hmi@ program was generamr tube rupture concurrent with large break to produce integral system effects data for use in code LOCAs; vessel void formation and removal tech-assessment and dewlopment efforts. Omink con-niques during secondary side breaks. Issues tributed to those efforts in the followmg areas: stand-addressed specific to small break LOCA safety included: the effect of " liquid holdup" on suction ard problems; double-blind pretest predactions; pretest scoping calculations; and most importantly in the area seal formation and resulting manometric core level of berdssking mdes for a giwn transient type.
depression; primary core liquid boiloff and depres.
Codes that were benefited by the Semiscale < lata surization following accumulator injection; ultra include RELAPS TRAC-BDI, TRAC-PFI, and a
small breaks with degraded emergency core ecoling RELAP4. On the order of 404 mink integral exper-3 4
(ECC); transient severity relative to break size and iments haw been used in the = ament of these codes.
location; and the effect of upper head to down-Semiscale impacted code dchent mainly because comer bypass flow on transient severity.
Serruscale was a mnstant user of the expenmental wr-
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&miscale provided experimental test beds to sions of the codes for pretest scoping calculat ons and improve safety technology. Semiscale successfully pretest predictions. Many unrtammanted minor cod-D tested an actual Westinghouse reactor vessel liquid ing errors and modeling and nnathion problems levelindicating system (RVLIS) and showed viabil-were solwd through this interaction. Specific examples
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of 4mi=k's importance to code development are the nomena. The data has been catalogued as to poor, fair,
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hmi=P pump model, transition boiling model, pres-good, or excellent retrienbility of configuration suruer vertkal stratification model, small break vapor reporting. The same raungs apply for data quality for pull through, and liquid entrainment of Tee junction nimcment purposes. An extensive data base connng a j
modehng, and dewlopmental assessment compansons wide range ofissues has a rating of exceBent configura.
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between code and data. The applicability of Senuscale tion availability and data quality. The hmiaA data exga,Ar,s for code development and assessment will continue to be used in the assessment efforts both
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are small. irnermediate, and large break issues and phe-universities, as well as foreign users.
efforts has been catalogued. Included in this catalog by NRC contractors, domesne wndors, unhtzs, and i
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EDO Principal Correspondence Control FROM DUE: 02/14/97 EDO CONTROL: G970075 DOC DT: 01/30/97 FINAL REPLY:
!Willicm S. Linnell II iMains Safe Energy TO:
Chairman Jackson FOR SIGNATURE OF :
- GRN CRC NO: 97-0116 Miraglia DESCt t
ROUTING:
QUESTIONS CONCERNING MAINE YANKEE Thompson
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Jordan Norry l
Blaha i
Miller, RI Paperiello,NMSS DATE: 02/03/97
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ASSIGNED TO:-
CONTACT:
NRR Miraglia SPECIAL INSTRUCTIONS OR REMARKS:
Put EDO and Chairman on for concurrence.
Chairman's Office to review response prior to dispatch.
NPR'PECEIVED:
FEBRUARY 4, 1997
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OFFICE OF THE SECRETARY CORRESPONDENCE CONTROL TICKET PAPER NUMBER:
CRC-97-0116 LOGGING > ATE: Feb 3 97 ACTION OFFICE:
EDO AUTHOR:
WILLIAM LINNELL AFFILIATION:
MAINE ADDRESSEE:
CHAIRMAN JACKSON LETTER DATE:
Jan 30 97 FILE CODE: IDR 5 MAINE YANKEE
SUBJECT:
VIEWGRAPHS FOR 2/4/97 COMMISSION MEETING ON MAINE YANKEE i
ACTION:
Direct Reply DISTRIBUTION:
CHAIRMAN, COMRS, SECY, OGC
-SPECIAL HANDLING: SECY.TO ACK CONSTITUENT:
l NOTES:
OCM # 7103 (CHAIRMAN SHOULD REVIEW RESPONSE PRIOR TO DISPATCH) 1 DATE DUE:
Feb 97 l'
SIGNATURE:
DATE SIGNED:
i AFFILIATION:
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EDO -- G970075