ML20112A573
| ML20112A573 | |
| Person / Time | |
|---|---|
| Site: | Pennsylvania State University |
| Issue date: | 03/01/1985 |
| From: | Richardson W PENNSYLVANIA STATE UNIV., UNIVERSITY PARK, PA |
| To: | NRC OFFICE OF ADMINISTRATION (ADM) |
| References | |
| NUDOCS 8503180372 | |
| Download: ML20112A573 (240) | |
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LICENSE RENEWAL APPLICATION Table of Contents
- 1. Letter of Transmittal ............................................
- 2. License Renewal Application ......................................
3 Financial Statements ............................................. Red
- 4. Safety Analysis Report ........................................... Yellow
'5.. Technical Specifications ......................................... Green
- 6. Environmental Impact Appraisal ................................... Blue
- 7. Operator and Senior Operator Requalification . . . . . . . . . . . . . . . . . . . . . Orange
- 8. Special Nuclear Material Requirements and Neutron Source.......... White Requirements v
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-THE PENNSYLVANI A STATE UNIVERSITY OFFICE OF THE PRESIDENT (j 201 OLD MAIN UNIVERSITY PARK. PENNSYLVANIA 16802 Executive Vice President 814-865-2505 tnd Provost of the University 1 March 1985
- Director of Nuclear Reactor Regulation U.S. Nuclear. Regulatory Commission
. Washington, DC 20555 Attn: Document Control Desk
. Ref: License Renewal Application for The Pennsylvania State University Penn State Breazeale Reactor, License R-2, Docket 50-5 I
Dear Sir:
I' The Pennsylvania State University respectfully submits application for renewal of facility License R-2 for continued operation of the Penn State Breazeale Reactor TRIGA pool type reactor. This renewal application is made pursuant-to the general requirements of Title 10 Code of Federal Regulations (10CFR) parts, 20, 50, 51, 55, 70, and 73.
The applications 'for license renewal contains' the following:
- l. Three (3) signed and notarized original and nineteen (19) i copies of the renewal application.
- 2. Three (3). signed and notarized original and six (6) additional copies of Financial Considerations.
l 3. Nineteen (19) copies of the following:
e I- . Safety Analysis. Report
- Technical Specifications 4
Environmental Impact Appraisal Operator Requalification Program SNM Requirements l
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i-l AN EQUAL OPPORTUNITY UNIVERSITY i
Director of Nuclear Reactor 1 March 1985
. Regulation .
.t The Emergency Plan for the Penn State Breazeale Reactor as well as the PSBR Physical Security Plan were approved previously. Therefore, as suggested in the July 26, 1984 letter from Cecil 0. Thomas to R. G. Cunningham, copies of these two plans are not included in this 4
submission.
It is requested that License R-2 be renewed utilizing the' supporting documents submitted as part of this license renewal application, and that these documents supersede previous corresponding submittals made to the NRC for License R-2.
The documents along with the excellent operating history of the PSBR over the past appr'eximate 30 years support the intent and desire of The Pennsylvania %te University to continue to operate the reactor in a safe and competent manner and to continue to promote ,
research and educational programs through utilization of the reactor facility. We believe that the renewal cf the R-2 license involves no significant hazard considerations.
Respectfully, bC h1 William C. Richardson 0~- Executive Vice President and Provost Enclosures cc: Pennsylvania Department of Environmental Resources American Nuclear Insurers t
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FACILITY LICENSE R-2 DOCKET NO. 50-5
. _ Name of Applicant: The Pennsylvania State University Address of Applicant: University Park, PA 16802 Name of Reactor: Penn State Breazeale Reactor (PSBR)
~ Description of Applicant's The Pennsylvania State University is a Land Business and Administration: Grant University funded in part by the State of Pennsyvlania to provide educational and
- research programs in engineering, the sciences, agriculture and liberal arts. The Breazeale Reactor provides nuclear services for
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researchers throughout the northeast. The Breazeale Reactor is operated by the Nuclear
, Engineering Department within the College of
- Engineering of The Pennsylvania State University System. '
k University Officials Associated Forrest J. Remick
[ With the Administration of the . Acting Vice President for
- PSBR: Research and Graduate Studies 2~
The Pennsylvania State University Wilbur L. Meier, Jr.
- . Dean College of Engineering The Pennsylvania State University George J. McMurtry
~ Associate Dean
. College of Engineering L The Pennsylvania State University Warren F. Witzig Professor and Department Head Nuclear Engineering Department The Pennsylvania State University Samuel H. Levine i
Director Breazeale Nuclear Reactor The Pennsylvania State University 4
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( Class of License: Class 104C (Research and Utilization Facility)
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Use of License: Renewal of Facility License R-2 required to continue operation of the 1 MW TRIGA Research Reactor of the Breazeale Nuclear Reactor for education, training, research and irradiation services.
Duration of License: 20 years Other NRC Licenses Associated SNM 95 With The Pennsylvania State Byproduct 37-185-4 University Breazeale Nuclear Byproduct 37-185-5 Reactor Byproduct 37-185-6 Finanical Qualifications: See Attachment I Supporting Documents: - Safety Analysis Report
- Technical.3pecifications
- Environmental Impact _Bppraisal
- Operator Requalif+ car. ion Program
- SNM Requirements
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I hereby certify the above statements to be true and correct to the best of my knowledge and belief.
C- - /y William C. Richardson Executive Vice President and Provost Subscribed to and before me on this date /4'c'./v T[f , Notary Public in and for Centre County, Pennsylvania hldf4 h- Ydffd$
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MARY A. NORTHAMER, Notary Pubi10 My commission expires univ, ray park. centre cc . Pa. .
My Comminion Lapues May 0. IN O
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e ' FINANCIAL STATEMENTS
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- j. Financial Qualifications of Applicant (PSBR) ......................... :
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, Audited Financial Statement of PSU ...................................
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'. (including Hershey Medical Center)
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1 FINANCIAL QUALIFICATIONS OF APPLICANT
- 1. GENERAL INFORMATION t
\Y~ ~ The Pennsylvania State University's funuamental responsibility is to provide programs of instruction, research and public service. It has ten colleges that offer undergraduate majors leading to baccalaureate and associate degrees. In addition to the University Park Campus in the municipality of State College, Behrend College in Erie, and Capitol Campus in Middletown, full-time instruction is available at seventeen. Commonwealth Campuses.
Graduate study is offered by the Graduate School in .17 distinct academic and professional degrees in 127 fields of study at several locations in Pennsylvania: the University Park Campus in State College, Behrend College in Erie, Capitol Campus in Middletown, the King of Prussia Graduate Center near Philadelphia, and The Milton S. Hershey Medical Center in Hershey.
Control of the University is vested in a Board of Trustees of thirty-two
. members. Members ex-officio are: . the Governor of the Commonwealth; the President.of the University; the State Secretary of Education; the State Secretary of Agriculture; and the State Secretary of Environmental Resources.
Terms of the other. trustees are three years. Six trustees are appointed by the Governor, nine are elected by the alumni, six by delegates from county agricultural societies, and six by delegates from county industrial societies.
The Pennsylvania State University la the land grant college of the Commonwealth of Pennsylvania that is supported by: -
- a. Student tuition and charges
/'N b. Commonwealth of Pennsylvania
(,) c. Auxiliary enterprises
- d. United State Government
- e. Medical center hospital
- f. Private grants and contracts ,
- g. Sales and services of educational departments An audited financial statement of the University for 1983-84 fiscal year is enclosed.;
The funding for -the Penn State Breazeale Reactor presently comes through the Department of Nuclear Engineering in the College of Engineering and was expended during 1983/8h as follows:
Salaries & Wages $338,778 Fringe Benefits 89,197 Department Allotment 76,195 Equipment 16,494 overhead 200,155 Building Depreciation 19,330 Tuition 6,060
$746,209 This level of funding is expected to ae maintained throughout the license renewal period.
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- 2. ESTIMATED CCSTS OF PERMANENTLY CLOSING DOWN THE FACILITY The estimated costs of permanently closing down the facility based on the In-Place Entombment option (described in Reg. Guide 1.86) are as followsi Shipping Fuel Assemblies Off Site h
($2000/ element) (164 elements) $328,000 Shipping Radioactive Fluids, Wastes and Highly Radioactive Components Off Site (15 tons) ($1000/ ton) 15,000 Entombment of Remaining Radioactive Material Control Rods, etc. In Concrete Within the Reactor Pool (360 yd3 concrete) ($50/yd3) 18,000 TOTAL $361,000 The In place entombment option of Reg. Guide 1.86 requires an appropriate and continuing surveillance program and a possession only license. We propose that the routine surveillance conducted by University Police Services of all University Buildings is an adequate surveillance program for such material entombed in the fashion described. The cost of such a surveillance program is negligible to the University since auch surveillance would be conducted even if the tomb weren't in the building.
The above should be considered as an option, probably the most conservative one. The ultimate option is properly lef t open until decommissioning is eminent. Entombment is probably the most conservative option (expensive) from a financial viewpoint and thus presented. It is expected that essentially all radioactive material could be removed form the site allowing h unrestricted future use.
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- Audited f,,inanClall
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Statements l r The Pennsylvania State University Fiscal Year Ended June 30,1984 g
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THE PENN5YLVANIA STATE UN lVERS1TY
' BOARD OF TRUSTEES as of October 12, 1984 Appointed by the Covernor Mumbers Ex of f icio Elected by Alumal JAY B. CLASTER RICHARD L. THORNBLRGH H. JESSE ARNELLE President Governor of the Connonwealth Attorney M. L. Claster & Sons, Inc.
PEIROSE HALLOWELL EDWARD R. BOOK CECILE M. SmlNGER Secretary Oiairman and Chlet Director. Contributions Department of Agriculture Executive of ficer and Community Affairs Hurshev Entertainment Westinghouse Elec. Corp. BRYCE JORDAN and Fe wt Company President of the University President MARGARET A. SMITH Owner, enti Cross United Mine Workers of America Acting Secretary Keys Inn, Inc.
Department of Education WILLIAM K. ULERICH MARIAN U. COPPERSMITH Chairman of the Board NICHOLAS DeBENEDICTIS President, Dbegan Signs, The Progressive Publishing Co., Inc. Secretary Inc., and Barash
$ Department of Environmental Advertising, Inc.
MARIE P. WALSH Resources Undergraduate Student LAWRENCE G. FOSTER Corporate Vice President EWARD P. ZEWRELLI for Public Relations State Senator Johnson and J0hnson KENNETH L. HOLDERMAN Elected by Delegates from Elected by Delegates from Retired A4ricultural Societies Industrial Societies NANCY V. KlDO D. EUGENE GAYMAN HWARD 0. BEAVER, JR. Licensed Psychologist Vice President Director and Retired Oialrman of Private Practice Pa. Formers Association the Board .
Car water Technology Corp. J0EL N. MYERS ROGER A. MADIGAN President H. THOMAS HALLOWELL, JR. Mcu-Weather, Inc.
State Representative Chairman of the Board SPS Technologies HELEN D. WISE 081E SNIDER Managing Partner, $1nging Executive Director Drook Farms BERNARD HANKIN Delaware State Educaticn President, Bernard Henkin Builders Associatlon RENO H. THOP*AS Co-Owner, Brooks End Farms J. LLOYD HUCK President and Oilef Operating CHARLES E. WISMER, JR. Of ficer, Merck & Co., Inc.
Master, Pennsylvania Emeritus Trustees State Grange STANLEY G. SCHAFFER Retired - Retired MILTON FRITSCHE BOYD E. WOLFF Owner, Wolfden Farm QUENTIN E. WOOD RALPH D. HETZEL, JR. - Retired Chairman of the Eberd and Chief Executive officer i
Quaker State Oil Refining Corp. G. ALBERT SHOEMAKER - Retired UNIVERSITY ADMINISTRATION as of October 12, 8984 RICHARD E. CRtBD STEVE A. GARBAN BRYCE JORDAN Senior Vice President President of the University Senior Vice President for Finance & Operations for Administration and Treasurer WILLIAM C. RICHAROSON HARRY PRYSTOWSKY inocutive Vice President and Senior Vice President for Hoelth
'rovost of the University Atfairs and Dean of the College of Medicine l
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CONTENTS The University Dollar 2 Letter of Transmittal 4 Auditors
- Opinion 5 FinancialStatements:
Balance Sheets 6 S*.atement of Current Funds Revenues, Experditures, and Other Chsnges 12 Statements of Changes in Fund Balances:
Unrestricted Current Funds 14 Restricted Current Funds 15 f Loan Funds 16
\ Endowment Funds 17 Funds Held in Trust for The Milton S. Hershey Medical Center 18 Unexpended Plant Funds 19 Funds for Renewals and Replacements 20 Retirement ofIndebtedness Funds 21 investment in Plant and Changes in Net investment in Plant 22 -
Notes to FinancialStatements 23 l'
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UNIVERSITY DOLLARh
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SOURCE PRIVATE GRANTS AND CONTRACTS 3.6c 1.6c SALES ANDSERVICESOF EDUCATIONALDEPARTMENTS OTHER SOURCES 4.0c
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^ STUDENTTUITION 24.0c ANDCHARGES
% l' MEDICAL CENTER HOSPITAL 10.6c D
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.Cf l w.p. Vf UNITED STATES oOvERNueNT 13.2c -
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- '" 1 23.9C OFPENNSYLVANIA AUXlOARY ENTERPRISES 19.1c 9?#
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A U 1983 - 1984 CURRENT FUNDS APPLICATION INSTRUCTION 26.7C 19.2C AUXILIARYENTERPRISES m., ,
i I f *, MN l f PUBUCSERVICE 4.2C l
i i INSTITUTIONALSUPPORT 5.8C 13.1C ORGANIZEDRESEARCH OPERATION AND MAINTENANCE OF PHYSICAL PLANT 6.4C 11.1C MEDiCALCENTERHOSP1TAL ACADEMICSUPPORT 6.7C STUDENT SERVICES 6.8C ANDSTUDENT AfD APPLICATION BY OBJECT D'-MISMNf*?'S ,. . ?dhESES 58.1 c SALAR S. WAGES, AND FRINGE BENEFITS T,*$T " ~ 4 4 ' ? ',7 ? ?pM _ OTHER CURRENT EXPENDITURES b i.-L.. 4 u.u.2 38.DC AND TRANSFERS r* EQUIPMENT AND N 3.4C MINOR IMPROVEMENTS 3
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TH E PEN NS Y LVA NI A STATE UNIVERSITY 408 OLD MAIN UNIVERSITY PARK, PENNSYLVANIA 16 02 October 12, 1984 Dr. kiryce Jordan, President The Pennsylvania State University
Dear Dr. Jordan:
The Audited Financial Statements of the University for the fiscal year ended June 30, 1984 are presented on the accompanying pages. These represent a complete and permanent record of the finances of the University for that year.
These financial statements have been examined by Deloitte Haskins & Sells, certified Public Accountants of Philadelphia, Pennsylvania and their report has been made a part of this record.
Respectfully submitted, Kenneth S. Babe Assistant Vice President -
Audits and Internal Control
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Senior Vice President for Finance & Operations and Treasurer
. kk AWl MARY A. NORTHAMIR, Notary Public Unieersity Park, Centre Co., Pa.
My Commission Expires May 6,1985 4
AN i 0041 OPIORTtrNITY Ifritvi HstTY u
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Haskins+ Sells 2500 t h.ee G.eard Plata Ph.laJsetpho. Pennsylvano 19102 (215) 569-3500 C:able oEHANDS AUDITORS' OPINION To the Board of Trustees of The Pennsylvania State University:
We have examined the balance sheets of The Pennsylvania State University as of June 30, 1984 and 1983 and the related state-ments of current funds revenues, expenditures and other changes, and of changes in fund balances for the year ended June 30, 1984 Our examinations were made in accordance with generally accepted auditing standards and, accordingly, included such
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tests of the accounting records and such other auditing pro-( cedures as we considered necessary in the circumstances. We previously examined and expressed our unqualified opinion dated October 10, 1983 on the statement of current funds revenues, expenditures and other changes (not presented herein) of the University for the year ended June 30, 1983. Fund totals for that year are included to provide a basis ror comparison with 1984 In our opinion, the aforementioned financial statements present fairly the financial position of the University at June 30, 1984 and 1983 and the current funds revenues, expenditures and other changes, and the changes in fund balances for the year ended June 30, 1984, in conformity with generally accepted ac-counting principles app).ied on a consistent basis during the two years ended June 30, 1984.
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October 12, 1984 A
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O Exhibit A THE PENN$YLVAN IA STATE UN I VERSITY BALANCE SHEETS, JUNE 30, I984 AND 1983 ASSETS June 30, 1934 June 30, 1983 Current funds:
Cash and temporary investments $40,546,107 5 42,628,414 investments - af scost (approximate market value, 52,134,000 and 57,707,000) 2,197,602 7,574,695 Accrued Interest raceIvable 1,649,313 1,483,071 Inventories I3,321,989 12,595,963 Prepaid expenses and deferred charges 2,422,423 2,878,368 Accounts receivable -
United States Government 7,822,372 10,568,940 Other, not of allowances of 18,655,563 and 56,655,228 27,387,849 23,583,873 Total current funds 595,347,655 O
5101,313,324 1
Loan funds:
Cash and temporary investments 5 645,142 5 132,671 Accounts receivable 39,529 32,823 Accrued interest receivable 606,803 618,352 Loans to students, not of allowance of 34,617,607 -
National Direct Student Loans 21,965,396 21,023,352 Health Professions Student Loans 1,138,665 1,163,749 Other 3,678,703 3,003,594 Deposit - United Student Aid Funds 23,000 23,000 investments - at cost (approximate market value, 54,178,000 and 53,840,000) 3,271,157 3,771,288 Due from endowment funds 755,499 755,499 Total loan funds $32,123,894 5 30,524,328 i
t See notes to financial statements.
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L 8ALAHCE SHEET 5 JUNE 30, 1984 ANO I983 LiA8iLITIES ANO FUND 0ALANCES l
1 June 30, 1984 June 30, 1983 I 1
Current funds:
Notes payable - due within one year --
1 10,000,000 Accounts payable 132,311,140 32,414,247 Accrued litterest payable --
70,000
. Accrued salaries and wages 9,520,425 9,009,703 Accrued payroll taxes 1,479,973 1,520,615 5tudents' deposits 2,297,122 2,333,440 )
Deferred revenues 18,547,468 17,142,153 I Fund balances -
Restricted fExhibit D) 27,373,767 28,183,324 Unrestricted (Exhibit C): i j
Designated 2,750,619 --
Undosignated 1,065,141 639,842
( Total current funds $95,347,655 3101,313,324 1
1 Loan funds: i Fund balances (Exhibit E) -
U. $. Government grants refundable 521,178,900 520,262,825 Univer ?ty funds - restricted j 10,944,994 10,261,503 I
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Total toen funds $32,123,894 53'],524,328 l
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Exhibit A (Continued)
THE PENNSYLVANI A STATE UN IVERSI TY DALANCE SHEET 5 JUNE 30, I984 ANO I983 ASSET $
June 30, 1984 June 30, 1983 Endo-ment and similar funds:
Cash and temporary investments 5 2,044,142 5 582,932 Accounts receivable 2,633 3,858 investments - et cost -
Land-Grant Fund - on deposit with state treasurer 517,000 5l7,000 Other (approximate market value, 532,894,000 ana $33,515,000) 36,250,374 32,855,936 Total endow:nent and similar f unds $38,774,149 533,959,726 Funds held in trust for The Milton S. Hershey Medical Center:
Cash and temporary investments 5 696,296 5 102,437 ccrued interest receivable 392,862 329,734
.nvestments - at cost (approximate market value, 522,399,000 and 122,091,000) 24,515,729 22.030,903 Total f unds held in trust 125,604,887 122,463,074 Picnt funds:
Unexpended -
Cash and temporary Investments $10,118,615 5 18,222 Accounts receivable 251,769 58,197 Accrued interest receivable 1,744,076 1,316,965 investments - at cost (approximate market value, 567,056,000 and 170,660,000) 71,039,618 70,410,335 Trustee account - 1st mortgage bonds 40,739 40,739 Funds held by Centre County Higher Education Authority ,
, 2,141, % 5 8.689.488 Total unexpended 85,336,782 80,533,946 See notes to financial statements.
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O BALANCE. 5HEETS, JUNE 30, 1984 AND 1983 LIASaLITBES AND FUND BALANCES June 30, 1984 June 30, 1983 Endowment and slalter funds:
Due to toen funds 5 755,499 5 755,499 Fund belances ,-
Endowment (Enhibit F) 28,899,372 24,084,949 Quest endowment - unrestricted 9.119,278 9,119,278 Total endowment and similar funds 538,774,149 $33,959,726 2
O~FundsheldintrustforTheMiltonS.HersheyMedicalCenter Principal belance (En>Ibit C) 525.604.887 $22,463,047 Total funds held in trust $25,604,887 $22,463,074 l- Plant funds:
i Unospended -
Bonds payable 5 2,182,704 5 8,730,227 Fund belances (Enhibit H) -
Restricted 2,241,411 3,175,072 Unrestricted 80,912,667 68.628,647 Total unexpended 85,336,782 80,533,946 l
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See notes to financlel statements.
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l Exhibit A (Concluded)
TNE PfNN$YLVANIA STATE UN I VERS 1TY 8ALANCE SHEETS, JUNE 30, 1984 AND 1983 ASSETS June 30, 1984 June 30, 1983 Plant funds - continued:
Renewals and replacements -
Cash and temporary investments 1 8,752,191 --
Accrued interest receivable 1,467,837 5 940,342 Investments - at cost (approximate market value, 558,125,000 and 156,885,0c3) 61,388,633 56,528,134 Total renewals and replacements 71,608,661 57,461,476 Retirement of Indebtedness -
Cash and temporary investments 378,768 --
Accrued Interest receivable 928,961 1,275,932 investments - at cost (approximate market value, 132,643,000 and 536,074,000) 36,n76.475 34,644,152 Total retirement of Indebtedness 38,I84,204 35,920,0 investment in plant -
Land 8,516,225 8,252,538 Dulldings 476,529,527 464,.160,647 Improvements other than buildings 55,628,301 54.843,028 Equipment 177,949,769 15L 283,896 Total Investment in plant 713,623,822 682,640,109 Total plant funds $908,753,469 5856,555,615 Agency funds:
Cash and temporary Investments 1 4,649,781 5 2,954,269 335,244 143,740 Accounts receivable 424,625 349,319 Accrued interest receivable investments - at cost (approximate market value, 111,825,000 and 512,949,000) 14.047,062 13,136,290 Total agency funds 1 19,456,712 5 16,583,618 See notes to financial statements.
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V 8ALANCE $HEET$. JUNE 30, I984 AND I983 L1A8iL1 TIES AND FUND DALANCE$
June 30, 1964 June 30, 1983 Plant f unds = continued:
Renevels and replacements -
Fund belences - unrestricted (Exhibit 1) 5 71,608,661 5 57,461,476 s
Total renewals and replacements 71,608,661 57,461,476 Retirmnent of indebtedness -
Fund belances - unrestricted (Enhibit JJ 38,I84,204 35,920.084 Total retirement of Indebtedness 38,184,204 35,920,084
.s investment in plant -
Bonds payable 56,109,296 55,130,773 Notes payable 1,977,000 2,465,000 Contractual payments withheld 451,675 200,893 Not investment in plant (Exhibit K) 655,090.851 624,843,443 Total Investment in plant 713,623,822 682,640,109 Total plant funds $908,753,469 5856,555,615 Agency funds: -
Federal temos withheld $ 4,228,552 $ 2,903,949 Life Insurance tonefIts f und 12,I48,704 11,165,896 Deposits held in custody for others 3.079,456 2,513,773 I
Total agency funds 1 19,456,712 5 16,583,618 See notes to financial statements.
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THE PENN$YLYANIA 5 TATE E=hibi-UNIVERSITY STATEMENT OF CURRENT FUNDS REVENUES, EXPENDITURES AND OTHER CHANGES FOR THE YEAR ENDED JUNE 30, 1984 o
Unrestricted 1954 1983 General Citts Total Restricted Total total Povenu es:
Educational and general -
Student tuition and fees $157,477,188 Governmental -
5157,477,188 --
1157,477,188 1145,748,63 Commonwealth of Pennsylvania -
Appropriations 149,368,000 --
149,368,000 Special contracts --
149,368,000 143,4 81,0C G United States Government -
5 7,216,101 7,216,101 5,516,9E Grants 50,000 --
50,000 31,695,899 Special contracts 31,745,899 30,862.6E Private gifts, grants, and contracts 39,294,116 39,294,116 36,934,42 679,845 679,845 22,967,076 23,646,921 Endowment income 21,833,92 21,950 -
21,950 2,155,846 2,177,796 Earnings of departments -
- 2,049,50 Sales and services 8,098,360 -
8,098,360 --
8,098,360 6,570,25 Organized activities related to educational departments 2,272,844 -
2,272.844 Other sources -
2,272,844 2,079,46 Investment income 9,226,024 -
9,226,024 802,195 10,028,219 8,155,01 income from Hershey Trust funds (Exhibit G1 1,983,710 --
1,983,7I0 -
1,983,710 1,916,59 Recovery of Indirect costs 18,868.456 -
13,868,456 --
11,868,456 10,260,41 Total educational and general 340,366,532 679,845 341,046,377 104,131,233 445,I77,610 415,388,88 Auxiliary enterprises 125,252,114 -
125,252,114 96,929 125,349,043 Hospital operations 118,258,01 68,868,782 -
68,868,782 333,352 69,202,134 Applied Research Laboratory 62,763,73
- -- - 15,168,239 15,168,239 15.050,01-Total revenues $34,487,428 679,845 535,167,273 I19,729,753 654,897,026 611.460,65.
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Educational and general -
u Instructice.
150,003,607 8,531 150,812,138 Organtred activities related to educational departments 16,232,646 167,044,784 5,211,644 160,267,C Orgentred research --
5,211,644 402,404 5,614,048 16,753,826 5,188,3-Public service -
16,753,826 53,426,719 70,180,545 17,291,627 65,014,2' Academic support -
17,291,627 50,078,558 27,370,185 42,489,537 26.989,4:
Student services -
42,489,537 415,977 42,905,514 18,742,008 16,618 4 0,4 23, lt lastitutional support 18,758,626 246,905 19,005,531 17,607,4t 35,913,971 592,605 36,506,576 Operation and maintenance of physical plant 178,957 36,685,533 33,336,6:
Student aid 41,274,050 28,230 41,302,280 2,052 41,304,332 37,369,7:
1,952,190 9,042 1,961,232 Mandatory transfers to/(from): 22,762,834 24,724,066 23,486,74 Loan funds - matching grants (Exhibit D 114,711 -
114,731 toan funds - other (Exhibit E) - -
114.711 132,1 Endo-ment funds (Exhibit F) 17,651 17,651 (3,67 Unexpended plant funds (Exhibit H) -
405,112 405,112 559,24 (10,718)
(10,718) (38,582) (49,3003 (5,05 Total educational and general 330,536,453 655,026 331,191,479 104,131,233 435,322,712 410,365,45 Auxillary enterprlses - '
Expenditures C 107,212,846 --
107,212,846 96,929 Mandatory transfers for principal and Interest (Exhibit J) 107,309,775 101,253,2E 3,798,878 - 3,798,878 HosDItal operations --
3,798,878 3,428,92 61,763,921 -
61,763,921 333,352 Appiled Research Laboratory - 62,097,273 59,245,95 Expenditures Mandatory transfer to retirement of Indebtedness funds 14,837,363 14,837,363 14,64t 47 (Exhibit J) - - --
330,876 330,876 405,54 Total expenditures and mandatory transfers 503,312,098 655,026 503,967,124 119,729,753 623,696,877 589,343,66 Other transfers and additions /(deductions):
Excess of restricted receipts over/(under) transfers to revenues - --
Ncq-mandatory transfers to: --
(809,557) (809,557) 4,527,70-Unexpended plant funds (Exhibit H) (2,861,293) -
(2,861,2931 --
(2,861,293)
Funds for senewals and replacements (Exhibit 1) (21,291,529) (2,304,37' (21,291,529) --
(23,291,529)
Retirement of Indebtedness funds (Exhibit J) (1,147,522) (18,088.888 (1,147,522) --
(1,147,522) (1,337,191 Funds held in trust (Exhibit G) (2,723,887) - (2,723,887) --
(2,723,887) (644,68t
- Total other transfers and additions /(deductions) (28,024,231) -
(28,024,231) (809,557) (28,833,788) (17,847.427 Net increase /tdecrease) la fund balances (Exhibits C and D) 5 3,151,099 $24,819 5 3,175,938 $(809,557) 5 2,366,361 5 4,269,56' See notes to financial. statements.
Exhibit C THE P[ NNSYLVAN IA STA1E UN IyERSI TY 5TATEMENT 0F CHANGES IN FUND 0ALANCES O
UNHE$iRICTED C U R R E"N T FUND 5 FOR THE YEAR ENDEO JUNE 30, !984
General----------
Deslanated undesignated Gifts Total Balance July I, 1983 --
5 (283,105) 1922,947 5 639,842 Revenuest Unrestricted cu'rrent funds --
534,487,428 679,845 535,167,273 Expenditures:
Educational and general' -- 330,432,460 655,026 331,087,486 Auxillary enterprises --
107,212,846 --
107,212,846 tusplial operations -- 61,763,921 -- 61,763,921
-- 499,409,227 655,026 500,064,253 Transf ers among f unds -
additions /(deductions):
Mandatory -
Loan funds - matching grants (Exhibit E) -- (184,711) --
(194,711)
Unexpended plant funds (Exhibit H) - 10,718 --
10,718 Principal and interest (Exhibit J) - (3,798,878) -- (3,798,878)
Non-mandatory -
Unexpended plant f unds (Exhibit H) -- (2,861,293) --
(2,861,293) ,
funds for renewals and replacements I (Exhibit I) -- (21,291,529) -- (21,2dl,529)
Retirement of Indebtedness funds (Exhibit J) -- (1,147,522) -- (1,147,522)
Funds held in trust (Exhibit G) -- (2,723,887) -- (2,723,887)
Designated for specific purposes 52,750,619 (2,750,619) - --
2,750,619 (34,677,721) - (31,927,102) e Not increase for the year - (Exhibit B) 2,750,619 400,480 24,819 3,175,918 Balance June 30, 1984 (Exhibit A) 12,750,619 5 117,375 5947,766 5 3,815,760 Soo notes to financist statements.
14 th
sy; 4
IHE P[NNSYLVANIA SIA1E UNIWLMSITY m
STATEMENT Di CHANGES IN FUND DALANCES RESTRICTED CURRENT FUNDS FOR THE YEAR ENDED JUNE 30, t984 i'- 8elence July 1, 1983 528,183,324 Revenues and other additions:
GIfis, grants, and contractr. 1127,830,611 Endonnent locame 2,I55,846 Inwestaent income 302,195
. t-130,788,652 Esponditures and other deductions: ,
Educeflonel and general 103,747,052 Auxi1Iery entorprises 96,929
~
Hospltal operations 333,352 Applied Research Laboratory 14,837,363
,, h Recovery of ladirect costs 11,868,456 130,883,I52 i
Mandatory transfers among f unds - additions /(deductions):
Loen funds (Exhibit E) (17,651)
Endomment funds (Exhibit F) (405,112)
Unexpended plant funds (Exhibit H) 38,582 Retirement of Indebtedness funds (Exhibit J) (330,876) l (715,057)
Not decrease for the year (Exhibit 8) (809,557)
I Betance June 30,1994 (Exhibit A) 527,373,767 ,
4 a
O . n.... ,. t,n.n .., s... n,s.
,s
, _ _ . . _.m.__.__ _. . . _ _ .. ._. _ , _ _ . _ _ _
THE PENN5YLVAN IA $ TATE UN 1 VER$1TY
$TATEMCNT OF CHANGCS IN FUNO O% LANCES L0AN FUNDS F0R IHE YEAR ENDED JUNE 30, I984 I
Belance July 1, 1983 130,524,328 Additions:
Endowment income 1 278,564 Interest on loans 526,475 investment income 174,910 CIffs and grants 179,002 National Direct Student Loar Program -
Federal capital contribution 1,029,118 Health Professions Student Loan Program -
Federal capital contribution 3.356 2,191,425 Deductions:
l Administrative and collection costs 217,044 Loan cancellations 502.430 719,474
( Transfers among funds - additions /(deductions):
l Mandatory -
Unrestricted current funds-matching grants (Enhibits B and C) 114,711 l Restricted current funds (Enhibits B and D) 17,651
- Endowment funds (Enhibit F) (4.7471 127,615 '
Not increase for the year 1.599.566 Balance June 30, 1984 (Exhibit A) 532,123,894 1
See notes to financial statements.
i 16 I
l _ _ _ _ _ _ _ _ . . _ _ _ _ _ _
U Exhibit F THE PENN5YLVANIA 5 TATE UNIVERSITY 5TATEMENT OF CHANGE $ iN FUND DALANCE$
ENOOWMENT FUND $
F0R THE YEAR ENDEO JUNE 30, 1984 Balance July 8, 1983 524,084,949 Additions:
Gifts, etc. ' 53,248,350 Not gain on sale of Investments 1,146,914 4,395,264 ;
Nandatory transfers omong funds - additions:
Restricted current funds (Exhibits 8 and D) 405,112 Loan f unds (Exhibit E) 4,747 Unexpended plant funds (Exhibit H) 9,300 419,159 Not incrosse for the year 4,814,423 i
I Balance June 30, 1984 (Exhibit A) 128,899,372 See notes to financial statemon+;.
I 17 4
9'
Euhicit G THE PENN$YLVANIA STATE UNIVERS1TY
$TATEHENT OF CHANGES 1N FUND 8ALANCES FUNOS HEL0 iN TRUST FOR THE Mi LT0N S. HERSHEY MEDl CAL CENTER FOR THE YEAR ENDED JUNE 30, t 984 Balance July 1, 1983 522,463,074 Additions:
lacome from investments 12,104,997 Less amount included in unrestricted current revenues (Exhibit B) 1,983,710 $ 121,287 Nut gain on sale of Investments 348,252 469,539 Deductior s - Custodial and other charges 51,613 Hon-mandatory transfer from unrestricted current funds (Exhibits B and C) 2.723,887 Not increase for the year 3,141,813 Balance June 30, 1984 (Exhlblt A) ,
125,604,887 See notes to financial statements.
O 18 ;
I i
1
Exhible H THE PENN$YLVAN iA STATL UN l VERSI iY
\
%d STAtEMEdi 0F CHANGES IN FUND DALANCES UNEXPENOED PLANT FUND $
FOR THE YEAR ENDED JUNE 30, I984 Belence July 1, 1983 571,803,719 Additions:
Gifts received 5 783,848 Ceapelgo receipts - unrestricted 199,394 Campoign receipts - restricted 503,252 income fromsinvesteents 7,302,320 use of feellltles 228,l72 Reimbursement for construction 332,082 Student parking fees 486,472 Other additions 35,596 9,871,136 3 Deductions:
Expenditures for construction capiteilzod (Exhibit K) 5,193,076 s Plant expenditures not capitalized 1,022,510 Perking eroe expense 107,009
- Other deductions 475 6,323,070 i
Transfers among funds - additions /(deductions):
l Mandatory -
Restricted current funds (Exhibits 8 and D) (38,582) ,
Evidowment Funds (Exhibit F) (9,300)
Unrestricted current funds (Exhibits B ond C) (10,718)
Non-mandefory -
Unrestricted current funds (Exhibits B and C) 2,861,293 Renewels and replacements (Exhibit 1) 5,082.116 Retirement of Indebtedness funds (Exhibit J) (82,5I6) 7,802,293 l
l Not increase for the year 11,350,359 i I
Balance June 30,1984 (Exhibit A) 583,154,078
(%
\ See notes to financial statements.
19
Exhibit i THE PENN5YLVANIA $ TATE U to i VERS 1TY STATEMEHf Of CHANGES IN FUND BALANCES
FUND $ FOR RENEWALS AND REPLACEMENTS FOR THE YEAR ENDED JUNE 30, I984 Balance July 1, 1983 557,46l,476 Additions: s income from investments 5 5,933,997 Deductions:
Expenditures for construction capitalized (Exhibit K) - 5,388,961 Plant expenditures not capitall2ed 2,233,264 7,622,225 Transfers among funds - additions /(deductions):
l Mandatory -
Retirement of Indebtedness funds (Exhibit J) (374,000)
Non-mandatory - '
Unrestricted current funds (Exhibits B and C) 21,291,529 Unexpended plant funds (Exhibit H) (5,082,116) 15,835,413 l
Not increase for the year 14,147,185 Betance June 30, 1984 (Exhibit A) 571,608,661 See notes to financial statements.
O 20
1 iHE PENN$YLVANIA 5 TATE UNIVERSITY STATEMENT 0F CNANGE5 i N FUND 8ALANCES RE7iREMENT OF 1NDE8TEONE5S FUND 5 FOR THE YEAR ENDED JUNE 30, I984 8elance July 1, 1983 535,920,084 Additions: s income froe investments 54,548,156 Deductions:
Bonds redeemed - at par value 2,505,000 Donds redeemed in advance (maturity value 53,064,000) 2,127,693 Bond Interest paid 3,335,749 Payments to bond trustee 15,445 A Other deductions 33,941 8,017,828 Transfers among funds - additions:
n Mandatory -
Unrestricted current funds (Exhibits B and C) 3,798,878 Restricted current funds (Exhibits 8 and D) 330,875 Renevels and replacements (Exhibit I) 374,000 Non e ndatory -
Unrestricted current funds (Exhibits 8 and C) I,147,522-Unexpended plant f uncs (Exhibit H) 82,516 5,733,792 Not increase for the year 7,264,120 Balance June 30, 1984 (Exhibit A) 538,184,204 See notes to financial statements.
21 y,,,,~y-rm-.---,w_- .v-.,,.,ow w, -
.---- --, y---- - - - - - - - - -, - - , - - - - -
- - - - , e- a , - - - . - - w --. ,- - . - - - -
I i
I Exhibit K THE PENN$YLVAN I A STA tE UN 1 YERSI i Y
$TATENEN I OF CHANGES 4N INVESTMINT I N PLANT ANO CHANGES IN NET INVESTNENT l N PLANT l FOR THE YEAR ENDED JUNE 30, 1984 l lavestment in plant, July 1, 1983 1682,640,109 .
Additions: i Expended from - +
[
Regular departmental budgets 5 8,835,053 '
Current operating budgets - special projects 3,253,661
[ Unexpended plant f unds (Exhibit H) 5,193,076 l
Funds for r9newals and repiscements (Exhibit il 5,388,961 Restricted funds + 3,733,274
! Department of General Services projects 4,032,945 Construction costs financed by bonds payable 3,044,438 i Other -
increase in contractual paymonts withhold 250,782 Appraised value of gif ts received 2.173,739 Estimated cost of other additions 153,763 36,059,672
.uctions:
Disposals, etc. 3,001.767 tirite-of f of equipment 7,074,192 5,075,959 .
t Not lacrease for the year i 30,983,713 l b IIvtstment in plant, June 30,1984 (Exhibit A) 5713,623,822 1
~
Net investment in plant, July 1, 1983 5624,843,443 Additions /(deductions): '
Not lacrease in plant assets (above) 530,983,713 Not increase in bonds payable (978,523)
Decrease in notes payable 493,000 increase in contractual payments withheld (250,782) 30,247,408
! Not investment in plant, June 30,1984 (Exhibit A) 5655,090.851 l
l See notes to financial statements.
I on
\ .
t
THE PENNSYLVANIA STATE UN IVERSiTY NOTE $ T0 FiNANCl A 1. 5TATEMENTS FOR THE YEAR ENDED JUNE 30 1984
- 1. SLHMARY OF SIGNIFICANT ACCOUNTING POLICIES The significant accounting policles followed by The Pennsylvania State University, as summarized below, are in accordance with the recommendations for account 1ng and reporting included in the Industry Audit Guide for Colleges and Universities issued by the American Institute of Certified Public Accountants. In addition, the University has utillzed the descriptions and classifications of current funds revenues, expenditures and other changes set forth in Part 5 of College and University Business Adelnistration - Administrative Service published by the National Association of College and University Susiness Officers.
s Accrual Basis The financial statements of the University, an Instrumentality of the Coasnonwealth of Pennsylvania, have been prepared on the accruel basis except that depreciation on physical plant and equlpment is not recorded. The statement of current funds revenues, expenditures and other changes is a statement of financial activities of current f unds related to the current reporting period. It does not purport to present the results of operations or the not income or loss for the period as would a statement of income or a statement of revenues and expenses.
I
( To the outent that current funds are used to finance plant assets, the amounts so provided are l
s accounted for as (1) expenditures, la the case of purchases of moveable equipment, minor leprovements or renovations and library books; (2) mandatory transfers, In the case of required provisions for debt amortization and Interest; and (3) as transfers of a nonmandatory nature for all other cases.
Fund Accounting To ensure observation of limitations and restrictions placed on the use of resources available to the University, the accounts of the University are maintained in accordance with the principles of
" fund accounting." This is the procedure by which resources for various purposes are classified for accounting and reporting purposes into funds that are in accordance w!th specified activities or I
objectives. Separate accounts are maintained for each fund; however, in the accompanying financial statements, f unds that have slaller characteristics have been combined into fund groups. Accordingly, all financial statements have been recorded and reported by fund group.
Within each fund group, fund balances restricted by outside sources are so Indicated and are distinguished f rom unrestricted f unds designated for specific purposes by action of the governing j~ board. Externally restricted f unds may only be utilized in accordance with the purposes established l by the source of such funds and are in contrast with unrestricted funds over vnich the governing board retalas f ull control to use in achieving any of its Institutional purposes. (he of the Uni-l versityas largest sources of revenue is its appropriation f rom the Coaumonwealth of Pennsylvania. The I
appropriation is included in unrestricted current funds revenues because the restrictions imposed by the Commonwealth are not so specific that they substantially reduce the University's flexibility in financial operations. Income derived f rom f unds held in trust for The Milton S. Hershey Medical l Center is also included in unrestricted current funds revenues because any portion of the fund prin-clpal and the income derived therefrom can be used for any maintenance, construction or operational purpose at the Medical Center.
23
l Indowment funds are subjuct to restrictions of gif t instrianonts ruquiring that the principal bo invested in perpetulty and incenms only bu utallred. While quasi endowment f unds have buen established by the governing board for tho same purposes as endowment funds, any portion of quasi endowment funds may be expended.
s All galns and losses arising from tho sale, collection, or other dispostflon of Investmonts and other non-cash assets are accounted f or in the f und which owned such assets. Ordinary income derived from investments, recolvables, and the liko, is accounted for In the fund owning such assets, except for income derived from investments of endowment and similar funds which income is accounted for in the f und to which it is restricted, or if unrestricted, as revenues in unrestricted current funds.
All other unrestricted revenuo is accounted for in unrestricted current funds. Restricted ,
gifts, grants, appropriations, endowmont income, and other restricted resources are accounted for 8 in the appropriate restricted funds. Rostricted current f unds are reported as revenues and expen- 4 ditures when expended for current operating purposes. '
Cash and Investments of the current, loan, quasi endowment, unexpended plant, renewal and replacement, retirement of indebtedness, and agency f unds are combined for investment purposes but allocated to the respective funds at year-end for financial statement presentation. Staff benefits are allocated to the respective categories in current f unds expenditures at year-end for financial statement presentation. Interdepartmental income of service departments and other institutional departments or of fices is not shown as current f unds revenues but rather as reductions of expenditures.
the Milton 5. harshey h dical Conter The financial statements include the accounts of the Medical Center, which the University operates as a successor trustee. As of June 30, 1984 the Center has net fixed assets of 1114,621,886 in addition to funds held in trust of $25,604,887 (Exhibit G).
Inventories inventorlos are stated at the lower of cost, generally on the first in, first out basis, or market.
Investments investments are stated at cost or fair market value at date of gift.
Investment in Plant Assets are stated at cost or fair market value at date of gift. Depreclation on physical plant and equipment is not recorded. t l As of June 30, 1984 the Pennsylvania Department of General Services has authorized the use of certain buildings and improvements which cost approximately $21,644,000; however, documentation to .
record such projects on the asset records of the University has not yet t.een received. Accordingly, such amounts have not been included with investment in plant.
- 2. POOLED ASSETS Certain eedowment f unds cash and Investments aggregating 125,682,036 are pooled on a market value basis, with each Individual fund subscribing to or disposing of units on the basis of the marnet value per unit at the beginning of the calendar quarter when the transaction tanes place. The following tabulatian sumnurizes changes in relationships between cost and market value of the pooled assets.
24
l pooi.d Asset s Not Calns Market Value Market Cost (Lossos) Per Unit End of year - 124,084,902 125,682,036 1(1,597,134) 5 9.95 Beginning of year- 22,865,735 22,093,746 771.989 10.71 Change in unreallred not gains and lossus (2,369,123)
'Aealized not gains 641.069 Total not losses for the year 1(1,728,054) 1 (.76)
The average annual earnings per unit, exclusive of not losses, amounted to 867 for the year.
- 3. CENTRE COUNTY HIGHER EDUCATION AUTHORITY Centre County Higher Education Authority (the " Authority") has provided long-term financing for the construction, esponsion, and equipping of certain f acilities at the University's main campus in Centre County and The Milton 5. Hershey Medical Center in Dauphin County (collectively the " Projects")
through the issuance of revenue bonds. At June 30, 1984 the Authority has outstanding $30,950,000 of revenue bonds with Interest rates ranging from 5.55 to 9.55. Principal payments are due annually on January 15 In increasing amounts ranging from 5775,000 to $3,165,000 through January 15, 2003.
The University has entered into Lease and Subloose Agreements with the Authority. The Agreements are for a term empiring May 1, 2003. The Lease provides for the leasing of the Projects by the University to the Authority f or rentals in amounts suf ficient to pay the cost of the Projects. The Sublesse provides for the subleasing by the Authority to the University of the Projects for rentals in -
amounts sufficient to pay the Authority's administration expenses and the principal of and Interest on the revenue bonds. Accordingly, the Agreements have been accounted for as direct financing leases and the Authority revenue bonds have been reflected as a liability of the University in plant funds.
4 LONG-TEfW DEBT The plant funds include a note payable for $1,972,000 with Interest at 51/45, payable in annual
- Installments of $493,004through January 1968.
Certain buildings and the gross revenue derived f rom them are pledged as collateral for 527,342,000 of first mortgage bonds payable, Series B through Series H (35 to 3.951, which are general obilgations of the University, Principal payments are due annually in amounts ranging from $275,000 to $1.905,000
- through July 1,1999 at which time the balance of $15,795,000 is due. As provided in the Indenture, beginning on July l,1984, bonds maturing July 1,1999 are entitled to the benefit of a sinking f und and are redeemable by lot at 100% of their principal amount, together with accrued Interest to the redemption date.
Maturities and sinking f und requirements on long-term debt (including Centre County Higher Education Authority bonds payable--see Note 3) for each of the next five fiscal years are summerlaed as follows:
L Annual Year Installments 1985 13,654,000 1986 3,630,000 1987 3,962,000 1988 3,795,000 1989- 3,290,000 25
..,-,e . - . , ,3.pw_q. , .-n.-.- y97 y_y. ...q_,
- > . 84 IINIMt Nf UCNU i f 5 the Univursity providus rutirumont benefits for sabitantially all regular umployoos. primarily undur a contributory plan administerud by it'o Comnunwealth of Pennsyl vania Employees' Retirement System. The University is billed f or lis share of the estimated actuarial cost of the plan. The University's cost' Included in expenditures for the year ended June 30, 1984 was $36,907,000.
The University uses an actuarial method for accruing the cost of continuing lif e ir.surance benefits for retired employees under the University's Group insurance Program, including emortization of prior service costs over 13 years f rom July 1, 1973. The cost of these benefits included in expenditures for the year ended June 30, 1984, including amortization of prior service costs, was
$282,189; the University's policy is to fund insurance cost accrued. The insurance fund balance encoeded the actuarlally computed present value of vested benefits to retired employees of the University as of June 30, 1984
- 6. CONTINGENCIES AND COMMITMENTS The University has contractual obligations for the construction of new buildings and for additions to salsting bulidings in the amount of 117, % 3,000, of which 17,444,000 has been paid or accrued to June 30, 1984 The contract costs are being financed from available resources and from borrowings.
The University has a program of self-insurance for medical malpractice at the Milton S. Hershey Medical Center and Is supplementing this program through participation in the Pennsylvania Health Care Services Malpractice Act ( Act Pb. lli of 1975), which provides limitec insurance coverage. As of June 30, 1984 no material claims in excess of available coverage have boon asserted. Additional claims could be asserted arising from services provided to patients; however, the amount of such claims, If any, cannot be estimated and no accrual has been made.
g various litigation has arisen in the course of conducting University bustress. The outcome of such litigation is not expected to have a material ef f ect on the financial position of the University.
)
t 0:i 8Qwel oppoqundy unweesdy 2G uco es m ,
SAFETY ANALYSIS REPORT
- fsv 1 TABLE OF CONTENTS
.5s_ ,/-
I. INTRODUCTION .................................................... I-1 II. SITE CHARACTERISTICS ............................................ II-1 A. Geography and Demography .................................... II-1
- 1. Reactor Site Access Control ............................. II-1 B. Nearby Industrial, Transportation, and Military Facilities .. II-6 C. Meteorology ................................................. II-6 D. . Geolo gy and Hydrology . . . . . . . . . . . . . . . . . . . . . . . . . . . . ....... II-7 E. Seismology .................................................. II-10 F. References .................................................. II-10 III. REACTOR DESIGN .................................................. III-1 A. Introduction ................................................ III-1 B. Mechanical Design ........................................... III-1
- 1. Reactor Bridge .......................................... III-1
- 2. Reactor Suspension Tower ................................ III-3
/,_s '
( / 3 Rea c to r Gr i d Pla te s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . III-3
- 4. Fuel-Moderator Elements ................................. III-6
- 5. Control Rods ............................................ III-6
- 6. Control-Rod Drives ...................................... III-9
- 7. Graphi te Reflector Elements . . . . . . . . . . . . .' . . . . . . . . . . . . . . . . III-16 C. Nuclear Design .............................................. III-16
- 1. Standard TRIGA Core ..................................... III-16
- 2. External Neutron Source ................................. III-23 D. Th e r mal De s i g n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . III-23 IV. R EACTOR POO L AND WATER SYSTEM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . IV-1 A. Reactor. Pool ................................................ IV-1 B. PSBR Water Handling System .................................. IV-2
-1. General ................................................. IV-2
- 2. Pool Recirculation Loop ................................. IV-2 3 Fission Product Monitor ................................. IV-2
- 4. Transfer of Pool Water .................................. IV-5
- 5. Heat Exchanger .......................................... IV-5 *
[~
\
- 6. Liquid Waste Evaporator ................................. IV-7
C. Water Quality Monitoring and Maintenance .................... IV-7 V. FACILITY CONSTRUCTION ............................ .............. V-1 A. Building .................................................... V-1 B. Heating and Ventilation ..................................... V-1 C. Utilities ................................................... V-6 D. Fire Protection ............................................. V-6 VI . . FACILITIES AND EXPERIMENTERS .................................... VI-1 A. Beam Ports .................................................. VI-1 B. D2 0 Thermal Column .......................................... VI-4 C. Central Thimble ............................................. VI-4
- 1. Central Thimble Oscillator .............................. VI-6 D. Vertical Tubes .............................................. VI-6
- 1. Jib Crane ............................................... VI-8 E. Pneumatic Transfer System ................................... VI-8
- 1. Pneumatic Transfer System I ............................. VI-8
- 2. Pneumatic Transfer System II ............................ VI-12 F. Instrument Bridge ........................................... VI-14 G. Hot Cells ................................................... VI-14 H. Co-60 Irradiation Facility .................................. VI-14 VII. CONTROL AND INSTRUMENTATION ..................................... VII-1 A. Control System Summary ...................................... VII-1 B. Steady State Mode ........................................... VII-1
- 1. General ................................................. VII-1
- 2. Start-up Channel ........................................ VII-1 3 Log Power and Period Channel ............................ VII-3
- 4. Linear Power Channel .................................... VII-6
- 5. Percent Power Channel ................................... VII-6
- 6. Temperature ............................................. VII-6
- 7. Manual Rod Drive Control ................................ VII-7 C. Automatic Mode .............................................. VII-7 D. Pulse Mode .................................................. VII-7
- 1. General ................................................. VII-7
- 2. Peak Pulse Indication ................................... VII-8 3 Temperature ............................................. VII-8 4 Rod Control ............................................. VII-8
- 5. Pulse Control ........................................... VII-8
g' N E. : Square Wave Mode ............................................ VII-9 s
\
- 1. General ................................................. VII-9
- 2. Log Power and Period Channel ............................ VII-9 3 Linear Power Channel .................................... VII-9
- 4. Manual Rod Drive Control ................................ VII-10
- 5. Automatic Control ....................................... VII-10 F. Control Room ................................................ VII-10
- 1. General ................................................. VII-10
- 2. Monitor Indications in the Control Room . . . . . . . . . . . . . . . . . VII-11 G. Minimum Safety Circuits & Interlocks ........................ VII-13 VIII. CONDUCT OF OPERATION ............................................ VIII-1 A. Organization and Responsibility ............................. VIII-1 B. Reactor Operating Safety Philosophy ......................... VIII-1 C. Training .................................................... VIII-3 D. Written Procedures .......................................... VIII-3 E .' Records ..................................................... VIII-3 F. Review and Audit of Records ................................. VIII-4
, ,s IX. SAFETY EVALUATION ............................................... IX-1 I / 1
( ,,/ A. Introduction ................................................ IX-1 B. TRIGA Fuel Temperature Analysis of the Penn State Breazeale Reactor ......................................... IX-2
- 1. Steady State Analyses ................................... IX-4
- 2. Pulsing Characteristics of the PSBR . . . . . . . . . . . . . . . . . . . . . IX-11
- 3. TRIGA Experiment to Measure Fuel Temperatures ........... IX-15
- 4. Evaluation of thegIt for Element I-14 .................. IX-17
- 5. Evaluation of the Pulse Data for Fuel Element I-14 ...... IX-22
- 6. Evaluation of the Fuel Element I-13 Temperature Data .... IX-26
- 7. Conclusion (Terperature Analysis) ....................... IX-28 C. Evnluation of the Limiting Safety System Setting ............ IX-29 i
D. Loss of Coolant Accident .................................... IX-30 E. Maximum Hypothetical . Accident (MHA) . . . . . . . . . . . . . . . . . . . . . . . . . IX-40 F. Reactivity Accident ......................................... IX-48 L G. Conclusion .................................................. IX-49 H. References .................................................. IX-51 o
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(' 'T SAFETY ANALYSIS REPORT
\_- LIST OF FIGURES l-Figure # Title 2-1 The Location of Centre County in Pennsylvania 2-2 The Location of the PSBR on The Pennsylvania State University Campus 2-3 The PSBR Site Boundary 2-4 Population Within a Five Mile Radius of the PSBR 2-5 The Physiography of Centre County 2-6 The Spring Creek Drainage Basin 3-1 The Location or the PSBR Core, Bridge, and Control Console 3-2 The Layout of the PSBR Grid Plates 3-3 The Arrangement of the PSBR Grid Plates and Safety Plate 3-4 A Standard TRIGA Fuel-Moderator Element
/s '3 An Instrumented TRIGA Fuel Element i \
\s_,/ 3-6 A Fueled Follower Control Rod with Respect to the PSBR Core 3-7 A Transient Control Rod 3-8 A Rack-and-Pinion Control Rod Drive
- 3-9 The Transient Rod Drive 3-10 Core Loading #1 Layout 3-11 Core Loading #4 Layout 3-12 - A Graph of Peak Power Versus Prompt Reactivity for the First Ninteen Pulses with Core Loading #4 3-13 A Graph of Peak Fuel Temperature Versus Prompt Reactivity for the First Nineteen Pulses with Core Loading #4 4-1 The PSBR Water Handling Systems 4-? The PSBR Fission Product Monitor 4 The PSBR Heat Exchanger 4-4 The PSBR Liquid Waste Evaporating System 5-1 The Location of the PSBR on The Pennsylvania State University Campus 5-2a The First Floor Plan of the Or131nal Reactor Bu'ilding (N~ ,/) 5-2b The Ground Floor Plan of the Orihinal Reactor Building
5-3 Location of the PSBR Electrical Supply Transformer 5-4a The First Floor Location of Fire Extinguishers and Fire Alarm Boxes 5-4b The Ground Floor Location of Fire Extinguishers and Fire Alarm Boxes 6-1 The Location of 3 Number of PSBR Facilities for Experimenters 6-2 The Location of the Beam Hole Laboratory, Hot Cells and Co-60 Irradiation Facility 6-3 The D20 Thermal Column 6-4 The Central Thimble Oscillator 6-5 Pneumatic Transfer System I 6-6 Pneumatic Transfer System I Laboratory Terminus 6-7 Pneumatic Transfer System II 7-1 The PSBR Control Console Instrumentation Layout 7-2 The PSBR Control and Instruaentation Block Diagram 7-3 The PSBR Control and Instrumentation Block Diagram (continued) ,
7-4 The PSBR Control Room Instrumenta' tion Pedestals 7-5 The PSBR Area and Air Particulate Monitoring System 8-1 The PSBR Organization Chart 9-1 PSBR Core Loading #36 9-2 Comparing Highest Measured Fuel Temperatures During a Pulse with EQ(34) for Fuel Element I-14 9-3 The Time Dependence of Air-cooled Fuel Body and Exhaust-air Temperature for Center Element with 267 W Input 9-4 Summary of Equilibrium Data for Loss-of-coolant Simulation Showing the Fuel-element Clad Temperature Versus Power Input to the Element for All Seven Dummy Elements Heated with the Same Power Input 9-5 Maximum Fuel Temperature Versus Power Density Af ter Loss of Coolant for Various Cooling Times Between Reactor Shutdown and Coolant Loss 9-6 Strength and Applied Stress as a Function of Temperature, U-ZrH1 .65 Fuel, Fuel and Clad at Same Temperature O
I l
I-l i
['~'y I. INTRODUCTION N_.Y On 15 January 1954, The Pennsylvania State University's Board of Trustees announced that the construction of a nuclear reactor had been authorized. On 8 July 1955, the U.S. Atomic Energy Commission issued license number R-2 for the operation of the Penn State Reactor at power levels up to 100 KW. Criticality of. the reactor was reached on 15 August 1955.
Penn State University (PSU) (R-2) and North Carolina State University (NCSU) (R-1) entered into a cooperative agreement with Argonne National Laboratory in April 1956 to establish the International School of Nuclear Science and Engineering. It was now possible for scientists from other countries to spend a semester at either PSU or NCSU in resident training and then'go to Argonne for four more months of additional training.
It was a natural extension to turn the material developed for this international training program into an academic program for nuclear engineers.
Thus, it was that on 16 June 1959, a Department of Nuclear Engineering was
, fx . formed.at Penn State.
,) Because of ever increasing experimental and educational demands, the maximum operating power level was increased to 200 KW on 10 June 1960. Also, in 1960, the General State Authority authorized funds for the construction of additior.s to.the Nuclear Reactor Facility consisting of a Hot Laboratory containing two hot cells, additional research floor space around the reactor, added general office and classroom space and a sizeable radiochemistry -
-chemical engineering wing. Full occupancy of the new facilities occurred in Spring 1964; however, by that time plans were well underway for the addition of a Co-60 irradiation facility. The Co-60 faciilty was occupied in January of 1967.
Very early in 1965 it became apparent that the 200 KW MTR reactor was not adequate to handle the increased experimental and research demands. The TRIGA was installed, was operational, and was taken critical on 31 December 1965 During the past twenty years, the TRIGA reactor has been and continues to Jbe used extensively -to help educate students in nuclear engineering, to allow research in a wide variety of scientific and engineering fields, to instruct reactor operators, and to educate the public regarding the peacerel uses of
[v; nucle.ar technology. The Penn State Breazeale Reactor (PSBR) normally operates
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I-3 on an eight hour shift five days a week. Occasionally, the reactor operates for two shif ts to accommodate nuclear engineering laboratory experiments, conduct special reactor operations for instruction purposes or perform a special experiment.
Because the present R-2 operating license expires 30 June 1985, the license must be renewed; thus, this document is submitted to the NRC to request a renewal of the R-2 operating license.
A Physical Security Plan and an Emergency Preparedness Plan in support of this request have been previously submitted and accepted by NRC.
O l
1 0
II-1 k II. SITE CHARACTERISTICS A. Geography and Demography The Penn State Breazeale Reactor (PSBR), a 1 MW(th) TRIGA reactor, is locate'd on the University Park Campus of The Pennsylvania State University in central Pennsylvania in the county of Centre (see Figures 2-1 and 2-2).
The campus is bordered by the commercial and residential areas of the North of the Borough of State College to much of the east, west, and south.
campus are university owned athletic fields and farms. The reactor site boundary is defined by the chain-link fence surrounding the PSBR as shown in Figure 2-3 The campus has a student population of about 32,000 and State College Porough has a population of about 36,000 (1980 census). Campus or borough residential areas are located approximately 300 yards from the reactor building in all directions except the north, where the residential area is more than 300 yards distance.
[ Four townships surround University Park Campus and State College
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. Borough (see Figure 2-4). Population figures (1980 census) for Patton, Ferguson, Harris, and College Townships are 7409, 8105, 3415, and 6239 residents, respectively. The great majority of the population of these four townships are within a five mile radius of the reactor building. The nearest town of significant population other than State College is Bellefonte (1980 census population is 6300) 10 miles to the north.
- 1. Reactor Site Ac' cess Control The land adjacent to all sides of the reactor site boundary is owned and controlled by the University. The only entrance into the site is through the chain-link gate at the southern end of the site. The
!- gate is normally open during working hours (0730 to 1700) and is locked thereafter. Police Service personnel make periodic random checks of site security af ter working hours.
A sign at the site entrance directs all personnel to check into the PSBR through the lobby. Entry through the lobby during working hours is always under the control of a receptionist who issues appropriate I personnel dosimetry. If the receptionist is temporarily away from the lobby, the lobby door is locked. All other outside doors to the PSBR
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II-6 building are locked and the possession of keys is governed by an O= Administrative Policy (AP-2), Regulation for Reactor Facility Keys.
B. Nearby Industrial, Transportation, and Military Facilities No heavy industry or military facilities exist in the University Park or State College areas.
University Park Airport is located about one and one-half miles north of the reactor building. This airport serves small private and small commercial commuter aircraft.
There is no rail service to University Park or State College.
C. Meteorology The records of the Penn State University weather station for the period 1887-1983 provide the following information. Average annual rainfall is 38.53 inches, the maximum 24-hour rainfall is 4.71 inches, and the maximum 48-hour rainfall is 7 73 inches. The most rainfall in one month is 12.82 O. inches. Average annual snowfall is 45.6 inches, the maximum 24-hour snowfall is if.6 inches, the maximum 48-hour snowfall is 30.5 inches, and the maximum snow accumulation in one winter is 98.2 inches.
According to the United States Weather Bureau, thunderstorms average about 40 per year (25 during the summer months) and are occasionally severe enough to cause property damage from hail, wind, lightning, and local flash flooding.
4 The Penn State University weather station gives the following campus wind information. The fastest wind speed on record is approximately 55 miles per hour (noted on several occasions). The fastest gust of wind on record during the past 30 years is 88 miles per hour. During the past 20 years, the number of gusts over 80,70 and 60 miles per hour was 1, 3 and 13 respectively. Wind direction varies with the season, but on an annual basis winds are from the southwest 38% of the time, from the northwest 21% of the time, from the northeast 13% of the time, from the southeast 10% of the time, and f nm the east, west, north and south 18% of the time.
According to Penn State Univer'sity meteorologists, Pennsylvania is not ff in the normal hurricane path. The most recent storm to be classified as a
II-7 hurricane as it entered Pennsylvania was Hazel in 1954. This storm was less AdQh severe in the University Park area than normally occurring thunderstorms.
Central Pennsylvania is an area of low tornado activity. According to United States Weather Bureau information, during the 1050-1981 period, only six tornadoes have been recorded in a twenty-five mile radius of the reactor building. This is an average of one tornado every five years in the twenty-five mile radius. No deaths and four injuries resulted, and some light to moderate property damage occurred. One tornado was a category one
(< 72 mi/hr wind speed), two tornadoes were category two (73-112 mi/hr wind speed), and two were category three (113-157 mi/hr wind speed). Information is not available for the sixth tornado. The closest tornado was a category
\' two about six miles southwest of the reactor building.
D. Geology and Hydrology The PSBR is located in a limestone valley in an area called the Valley and Ridge Province. This province is characterized by broad limestone valleys interrupted by steep, forested sandstone ridges (see Figure 2-5).
The PSBR site is in the Spring Creek Drainage Basin, an area of about 175 square miles (see Figure-2-6). The basin is underlaid primarily by limestone. The outlet for the drainage basin is at the Milesburg Gap, about ten miles north of the PSBR. The PSBR is not located in an area of the basin with a past history of floods.
The hydrology of the area is typical of limestone terrains, with water entering the underground aquifer system by way of sinkholes, caves, fissures in rocks, and percolation through stream bottoms and soil covers. In the State College Borough and University Park areas, the water table is typically 225 feet below ground level, well below the reactor building foundations.
The University water supply is served by wells located two to three miles north of the PSBR. The State College Borough and many surrounding areas are served by well fields located 2 miles southeast and four miles southwest of the PSBR, and a spring fed mountain reservoir four miles southeast of the PSBR. Within a five-mile radius of the PSBR, other private wells and small public water systems using both wells and mountain springs can be found.
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'~'g The PSBR is served by the University water system. An adequete water Ns _s/ supply exists for PSBR needs, which are not large.
E. Seismology Twenty-four earthquakes occurred in Pennsylvania from 1728-1908, a period for which records may not be complete. We have no information regarding recorded earthquakes between 1908 and 1925; twenty-five earthquakes have occurred since 1925. Only two earthquakes have occurred with a Mercalli intensity as high as VII (an intensity of VII is noticed by all as a shaking of trees, waves on ponds, and quivering suspended objects but causes negligible damage to buildings of good design and construction)
- and these occurred at distances of 120 and 170 miles from University Park.
Only one earthquake, of'very low intensity (Mercalli intensity not known),
has been recorded in Centre County, that one occurring in 1941. Two earthquakes have occurred in adjoining Blair County (a Mercalli intensity VI in 1938 and a Mercalli intensity V in 1964).
fsg Mathematical predictions for the frequency and intensity of earthquakes 4
,, ) in Pennsylvania in the future.are based on past historic records for the area of concern. Depending on the equations used and the time base of the past records used, the predicted ' average occurrence period for-a Mercalli intensity VII earthquake is every 50 to 60 years, for a Mercalli intensity VIII earthquake is every 136 to 200 years, and for a Mercalli inteneity IX earthquake is every 341 to 1000 years.
. Earthquakes in Pennsylvania are so infrequent that few of its citizens have ever felt one.
.F.
References Howell, B.F., Jr. , Earthquake Expectancy in Pennsylvania, Vol. 53, Issue 2.
Proceedings of the Pennsylvania Academy of Science: 1979, pp. 205-208.
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III-l s
III REACTOR DESIGN V
A. . Introduction As the original Pennsylvania State University (PSU) 200 KW reactor, with its MTR type fuel elements, approached its tenth year of operation, it became apparent that the replacement of certain basic components was desirable. . To fulfill that desire and to increase research, instruction, and continuing education capabilities, The Pennsylvania State University purchased from Gulf General- Atomic the components for converting to a TRIGA MARK III Reactor.
Dismantling of the 200 KW MTR reactor began on 26 November 1965 and five weeks later ab 1237 hours0.0143 days <br />0.344 hours <br />0.00205 weeks <br />4.706785e-4 months <br /> on 31 December 1965 the 1 fni Penn State Breazeale Reactor (PSBR) achieved criticality.
B.' Mechanical Design
- 1. --Reactor Bridge
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-( _,j The angle aluminum suspension tower, four control rod drive motors and a fission chamber drive motor, two area monitor ion chambers, a fuel handling tool, the dif fuser pump, the central thimble oscillator, a jib crane, an accumulator tank for transient rod air, bulk pool temperature sensors. . pool lights, a nitrogen -gas bottle, and a TV receiver are all items supported by a bridge assembly which_ spans the pool as shown in Figure 3-1. The. bridge is mounted on four wheels and can be moved on rails that are bolted to the top surface of the pool walls. Movement of the bridge assembly is controlled by hand and the speed of the movement is limited by a high gear ratio hand wheel.
'Two vises clamp the bridge to the rails and the hand wheel is chained and padlocked so that'the bridge assembly cannot be moved Juring operation. Electrical power and control circuit wiring are supplied to
-the bridge by a cable arrangement. The slack cable, which allows reactor bridge movement, is stored in a floor trough which runs parallel to the pool wall.
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- 2. Reactor Suspension Tower d The suspension tower is a welded assembly of 2 x 2 x 1/4 inch
- angle altainum. The upper end of the tower is bolted to the reactor bridge I-beams and extends 22'8" below the bridge floor. The lower end of .the tower has a 19 x 29 x 3 inch grid plate bolted to it. A partial list of equipment supported by the suspension tower follows: the core
-aseeably and grid plates, a central thimble, four detectors, the N-16 diffuser system plumbing and spray nozzle, an external neutron source, two pneumatic transfer system termini, vertical access tubes, and core
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1 j 3. -Reactor Grid Plates The grid plate arrangement is shown in Figure 3-2. The fuel element positioning holes are arranged in a hexagonal pattern. The bottom grid plat'e is an aluminum plate approximately 19 x 29 x 3 inches and is bolted to _ the vertical aluminum angles of the suspension tower structure. The bottom grid plate supports the weight of the fuel and has fuel element locating holes over. its entire surface. The top grid J. plateLis 5/8" thick, made of aluminum but covers only a portion-of the available area so that experiments can be conveniently mounted on the bottom grid plate immediately adjacent to the active core. Holes approximately IV" in diameter in the top aluminum grid plate position the fuel elements-and control rods. A 12 x 16 x 1 1/4 inch aluminum safety plate is suspended approximately 12 3/4" below the bottom grid plate to prevent the control rods from dropping out of the core should their mechanical connections fail (see Figure 3-3). Small holes at various positions in the top grid plate permit insertion of wires and other small devices into the reactor for in-core measuring purposes.
The following special in-core experimental facilities are available 3 in the top grid plate (see Figure 3-2):
- a. A central thimble 1 33" inside diameter
- b. A central ~ removable section of top grid plate provides a 4.12" diameter hole (15.85 sq.in. cross section).
! c. Two removable triangular sections of the top geld plate provide F -
circular openings with 2.4" inside diameters.
l , d.. A large removable side section of the top grid plate.
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- 4. Fuel-Moderator Elements The PSBR utilizes fuel-moderator elements in which zirconium hydride moderator is homogeneously combined with partially enriched uranium fuel. The fueled section of these elements is 15" long,1.43" in diameter, and contains uranium in 2 different weight percentages enriched to slightly less than 20% in U-235. Part of the elements are 8.5 weight % uranium in zirconium hydride and the remainder are 12 weight %. The hydrogen to zirconium atom ratio of the fuel-moderator material for the original 8.5 weight 5 fuel is 1.7 to 1.0 and 1.65 to 1.0 for the 12 weight % fuel. To facilitate hydriding, a 0.18" diameter hole was drilled through the center of the active fuel section. A zirconium rod was inserted in this hole after hydriding was completed.
Figure 3-4 shows a standard fuel element.
The weight of a fuel element is about TV pounds with the U-235 content between 36 and 39 grams in the 8.5 weight 5 elements and between 53 and 56 grams in the case of 12 weight % elements. Serial numbers scribed on the top end fixtures are used to identify individual fuel elements. Each element is clad with 0.02" thick stainless steel.
To measure fuel temperature during reactor operation, instrumented fuel elements are fabricated similar to standard elements but with three thermocouples embedded in the fuel region. One thermocouple is at the vertical centerline of the element and the other two are located 1" above and 1" below center. All three thermocouples are located 0.27" radially from the center of the fuel element. Figure 3-5 shows an instrumented fuel element. The thermocouple lead wire pasaes through a water tight seal contained in a 3/4" outside diameter stainless steel tube welded to the upper end-fixture. This stainless steel tube is extended to provide a watertight conduit carrying the lead wires above the water surface of the pool. Additionally, the stainless steel tube provides a means of handling the element.
- 5. Control Rods Three standard, motor-driven control rods: one safety, one shim, one regulating rad; and one electro pneumatic transient rod control reactor power during steady state operation. These control rods pass
III-7
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III-9 7s - through and are guided by the top and bottom grid, plates. The rod A
N__e) locations are shown in Figure 3-2. The stainless steel clad control rods are 43" long and 1 3/8" in diameter. A standard control rod is shown in the withdrawn and inserted positions in Figure 3-6.
The upper section of a standard rod is graphite; the next 15" is graphite impregnated with powdered boron carbide which provides neutron absorption; the follower section consists of 15" of uranium zirconium hydride fuel (about 31 grams or U-235); the bottom 'section is 6T" of graphite.
The fourth control rod, shown in Figure 3-7, is the transient rod.
It has two functions; (1) it acts as a safety and/or control rod in the steady state mode of operation and (2) it is pneumatically driven from the core for the square wave and pulse modes of operation. The transient rod is 37" long and is contained in a 1 1/4" diameter aluminum tube. The borated graphite section is 15" long. Unlike the standard control rods, the transient rod has an air filled follower that is 21" long. The transient rod is guided laterally in the core by a thin fs walled aluminum guide tube that passes through the upper and lower grid
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-( ,,/ . plates and screws into the safety plate. All four control rods have a
. stroke of approximately 15".
- 6. Control Rod Drives Rack-and pinion drives are used to position the shim rod, the regulating rod, and the safety rod. Each drive consists of a single phase, reversible motor; a magnetic rod-coupler; a rack-and pinion gear system; and a ten turn potentiometer used to provide an indication of rod position (see Figure 3-8). The pinion gear engages a rack attached to a draw tube supporting an electromagnet. The magnet engages an iron arcature attached above the water level to the end of a long connecting rod that terminates at the lower end in the poison rod. The magnet, its draw tube, the armature, and the upper portion of the connecting rod are housed in a tubular barrel. The
. barrel extends below the pool water level with the lower end of the
. barrel serving as a mechanical stop to limit the downward travel of the control rod assembly. Part way down the upper portion of the connecting rod, i.e., just below the armature, there is a piston that travels (OV)
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III-13
. within the - barrel assembly. Because ' t.v upper portion of the barrel is
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well ventilated by large ' slotted openings, une piston moves freely in
-this range; but when the piston is within 2 inches of the bottom, its movement is restrained by the dashpot action provided by the graduated
~ vents in the. lower end of the barrel. This dashpot action reduces '
bottoming impact when the rods are dropped by a scramming action.
The control rod is withdrawn from the core by the rotation of the motor- shaf t when the electromagnet la energized. When the reactor is scrammed, the electromagnet is de-energized and the control rod drops into -the core by gravitational force.
- The drive motors for the shim and safety rods are nonsynchronous, single phase, and electrically reversible; they will insert or withdraw the rods at a rate of approximately 19 inches per minute. Electrical dynamic and static breaking of these motors is used to provide fast stops and-to limit coasting or over travel. The regulating rod drive
. motor is _a variable-speed servosotor with a tachometer generator for -
rate feedback; it inserts and withdraws the rod at a maximum rate of 24
[ inches per minute.
Limit switches mounted on each drive assembly stop the rod drive motor at the top and bottom of travel and provide contact for console indicator lights (see Figure 7-1), which indicates
- a. when'the magnet is in the UP position
'b. - when the magnet (and thus the control rod) is in the DOWN
- position
- c. ~ when 'the magnet is in contact with the control rod armature l- To allow transient operation with the fourth control rod, use was made of a pneumatic-electro-mechanical drive system to eject a predetermined amount of the transient rod from the core (see Figure L.
L3-9).
'E The pneumatic ' portion of the pneumatic-electrc-mechanical drive, referred to herein as the " transient rod drive," is basically a ;
single-acting pneumatic cylinder. A piston within the cylinder is attached to the transient rod by means of a connecting rod. The piston
- irod passes through an air seal at the lower end of the cylinder.
Compressed air.is admitted at the lower end of the cylinder to drive the piston upward. As the piston rises, the air being compressed above the e
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III-15
- (7 piston is. forced out through vents at the upper end of the cylind sr. At the end of its stroke, the piston strikes the anvil of a shock absorber.
This piston is thus decelerated at a controlled rate during its final inch of travel. This action minimizes rod vibration uhen the piston reaches its upper limit stop.
An accumulator tank mounted on the movable reactor bridge stores the compressed air that operates the pneumatic portion of the transient rod drive. A three-way solenoid valve, located in the piping between the accumulator tank and the cylinder, controls the air supplied to the pneumatic cylinder. De-energizing the solenoid valve interrupts the air supply and relieves the pressure in the cylinder so that the piston drops to its lower limit by gravity. With this operating feature, the transient rod is inserted in the core except when air is supplied to the cylinder. Applying air to the transient rod cylinder by depressing the TRANSIENT ROD FIRE button prior to moving the cylinder allows the transient rod to be used as an ordinary control rod. Pre positioning the transient rod cylinder and then applying air in accordance with PSBR
.o Standard Operating Procedures allows the transient rod to be used for square wave and pulse operation.
The electromechanical portion of the transient rod drive consists of an electric motor, a ball-nut drive assembly, and the externally threaded air cylinder. During electromechanical operation of the transient rod, the threaded section of the air cylinder acts as a screw in the ball-nut assembly. These threads engage a series of balls contained in a ball-nut assembly in the drive housing. The ball-nut assembly. is in turn connected through a worm gear drive to an electric
, motor. The cylinder may be raised or lowered independently of the
! piston and control rod by means of the electric drive. Adjustments of
- the position of the cylinder controls the upper limit of the piston travel, and hence controls the amount of reactivity inserted for a pulse or square wave.
A system of limit switches similar to that used with the standard control rod drives is used to indicate the position of the air. cylinder and the transient rod. Two of these switches, the Drive Up and Drive Down, are actuated by the cylinder. A third limit switch, the Rod Down switch, is actuated when the piston reaches its lower limit of travel.
C/
III-16
- 7. Graphite Reflector Elements Graphite reflector elements may occupy the grid positions not filled by fuel-moderator elements and other core components. The graphite reflector elements are canned in aluminum and have aluminum end fixtures and spacer blocks. These elements are of the same dimensions as the fuel-moderator elements, but are filled entirely with graphite.
Each graphite reflector element weighs 2.8 pounds and is anodized af ter assembly. The spacer blocks have a blue anodized finish to make the graphite dummy elements easily distinguishable from fuel-moderator elements. When properly installed in the core, the top of the triangular spacer block of the fuel and graphite elements are level with the top of the top grid plate.
C. Nuclear Design
- 1. Standard TRIGA Core The design and operating characteristics of standard TRIGA cores are well known as is the inherent safety charactcristic of this class of reactor. The first PSBR standard TRIGA core loading reached criticality in December 1965 (see Figure 3-10). Table 3-1 lists operating characteristics that were observed for core loading #4, one of the first, widely used, PSBR configurations of standard 8.5 wt% TRIGA elements (see Figure 3-11). Most of the fuel in core loading #4 received little burnup during this period of the PSBR operation. Table 3-1 also lists the core characteristics for the present loading #36, a core composed of a mixture of 8.5 wt% and 12 wt% fuel elements.
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\j - Operating Characteristics of PSBR Loading Core #4 Steady State Power Level 1 MW Core. Mass 3219 grams U-235 Minimum Critical Mass (Core #1) 2537 grams U-235 Total Control' Rod Worth $11. 63 Excess Reactivity $6.64 Power Defect (1MW) $3.50 Prompt Negative Temperature Coefficient - 1.4 x 10-4Ak/k/'C Maximum Pulse Temperature 912=C Maximum Pulse Reactivity Insertion $3 40 Operating Characteristics of PSBR Loading Core #36 Steady State Power Level 1 MW Core Mass 3190 grams U-235 Minimum Critical Macs (Core #1) ---
f Total Control Rod Worth $11.14 Excess Reactivity $6.39 Power Defect -(IMW) $4.00 Prompt Negative Temperature Coefficient - 1.4 x 10-4Ak/k/*C Maximum Pulse Temperature- 1067*C
- Maximum Pulse Reactivity Insertion $3.40 The Safety Evaluation section in this document will show that pulsing to $3.40 will not exceed the Limiting Safety System Setting
-(LSSS) of 700*C. The temperature and power characteristics during the first 19 pulses of core loading #4 are shown in Figures 3-12 and 3-13 Greater than 50% of the prompt negative temperature coefficient of a standard TRIGA core comes from the " cell effect" or temperature dependent disadvantage factor, and approximately 20% each from Doppler broadening of the U-235 resonances and temperature dependent leakage from the core. The current loading that is being used routinely at the PSBR is a mixture _ of 12 wt$ and 8.5 wt1 fuel (mixed core) as shown in Figure 3-14. Loading #36 exhibits many of the characteristics of Loading #4, i.e., similar excess reactivity, similar operating
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x_j mixed cores behave in a manner very similar to cores loaded with only 8.5 wt% fuel. The section on Safety Evaluation gives a detailed analysis of TRIGA core characteristics, i
- 2. _E_xternal Neutron Source The start-up source used in the PSBR is a 25 curie antimony-beryllium neutron source clad with 0.01 inches of type 347 stainless steel. The source is contained in an aluminum source holder which has outside dimensions similar to standard fuel-moderator elements so that the holder can be positioned in any fuel element position in the core. Movement is accomplished manually by a cable attached to the top of the source holder.
Another external source available for use at the PSBR is a 0.235mg Californium-252 neutron source. The Californium oxide source in its platinum matrix is doubly encapsulated in zircaloy-2. It was fabricated r~ at Savannah River Laboratories and is on loan to the University from the i Department of Energy.
D. Thermal Design The PSBR operates at 1 MW thermal steady state and is cooled by natural convection. Cooling water enters the core from the perimeter of the core region immediately above the bottom grid plate and to a smaller extent through holes provided in the bottom grid plate. The triangular shaped spacer blocks on the upper end of the fuel elements are used for positioning as well as providing a means for water to flow up through the core and out the top grid plate. Because of the natural convection cooling, the PSBR can be operated in a water filled pool with no direct coupling between the core and the heat removal system in the pool. Since no connection to a force.
circulation system is required, the reactor can be operated at any position in the pool.
, - ~
/
IV-1
- ,__ IV. REACTOR POOL AND WATER SYSTEMS
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I' A. Reactor Pool
- The reactor pool is approximately 30 feet long,14 feet wide, and 24
' feet deep. The pool is constructed of steel reinforced concrete. The pool walls are IV feet thick below the level of the reactor oay floor and one foot thick above the reactor bay floor. The inside of the pool wall is coated with epoxy to form an epoxy pool liner. The pool is surrounded by earth fill with the exception of the south side. A room (Beam Hole Laboratory) exists outside the south side of the pool at the elevation of the reactor core. Additional shielding is provided by 3V feet of high density concret e on the outside of the pool wall in this room.
The total pool volume is approximately 71,000 gallons. The pool can be divided in two by a removable gate which permits draining either part of the pool while maintaining the reactor under water in the other part of the pool.
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^~' Seven beam ports penetrate the pool wall from the Beam Hole Laboratory to provide access to reactor radiation. The pool is equipped with two floor drains; one for each side of the pool when it is divided with the removable gate. Four other pool wall penetrations exist, two located approximately 10 feet above the pool floor to serve the pool recirculation loop and two located approximately 17 feet above the pool floor for the heat exchanger.
Several sources of water. are available for adding water to the pool they are (a) the distillate from the waste evaporator, which is stored in an underground tank, can be pumped into the recirculation loop for normal pool water make-ups (b) water from the University water system can be added to the pool through the demineralizer; (c) water from the University water system can be added to the pool at a high flow rate through the pool drain lines, af ter connecting a fire hose that is maintained in location for this purpose (d) water from the secondary side of the heat exchanger can be diverted directly to the pool through a fire hose for emergency make-up water.
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IV-2 B. PSBR Water Handling System
- 1. General For the proper and safe operation of the reactor and related equipment, certain properties of the water such as temperature and mineral content must be controlled. Figure 4-1 shows all of the water handling systems for the PSBR facility and in the following sections the individual systems are discussed in detail. '
- 2. Pool Recirculation Loop The recirculation loop normally recirculates pool water continuously through filters and a demineralizer to maintain water quality.
The flow rate through the system is about 40 gallons per minute.
Water enters the system through a skimmer arrangement at the south end of the pool, flows through the recirculation pump, a filter, a mixed bed demineralizer and then flows back into the pool. In addition, a small portion of the water is diverted through a fission product monitor..
3 Fission Product Monitor Pool water is withdrawn from the pressure side of the pool water recirculation pump, passed through the fission product monitor, and returned to the suction side of the pump. Figure 4-2 shows a diagram of the fission product monitor. Water first passes through a filter and then a cation exchange bed where the cations are removed. The water is then sent through the anion exchange bed where anions are removed from the water. The water which emerges from the anion bed then passes through the flowmeter-alarm and re-enters the recirculation loop. The flowmeter gives an annunciator indication to the reactor operator when the flow rate is not within a specified range.
The radioactivity in the anion bed is monitored with a scintillation counter with a discriminator setting to detect and quantify the high energy gamma rays (above 0.8 MeV) of the I-135 O
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Figure 4-2 The PSBR Fission Product Monitor 1
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IV-5 fission product. This scintillation counter output is displayed in the reactor control room.
- 4. Transfer of Pool Water Occasionally, for maintenance purposes, it is necessary to drain the reactor pool. In order to provide shielding for the reactor core and any stored fuel elements, only one half of the pool is drained at a time. The pool is divided into two parts by inserting an aluminum structure, or gate, into the gate support structures.
The water from either half of the pool can be stored in a 48,000 gallon aluminum hold-up tank located behind the facility. The water leaves the pool via large drains located in the pool floor along the west pool wall near the partial concrete divider. The storage tank transfer pump (capacity - 300 gal / min) located in the beam hole laboratory is used to pump the water. Water can be returned to the pool from the storage tank using the same pump and floor drains (see Figure 4-1).
By opening the valves on either of the pool drain lines one could release pool water to the storm sewer. However, to minimize the release of radioactivity and to conserve the demineralized pool water, the pool water is transferred to the storage tank when it is necessary to drain the pool. To prevent any accidental release, valves #65 and #66 are secured with a padlock (see Figure 4-1). Keys to the lock are issued only to those with an NRC Senior Reactor Operator License.
- 5. Heat Exchanger The PSBR heat exchanger limits the tempersture of the PSBR pool water (see Figure 4-3). Maintaining lower pool temperatures decreases pool water evaporation losses and temperatures below 110*F are needed to prevent damage to the demineralizer anion resins.
The system is composed of two loops. In the primary loop, pool water is pumped through the baffled shell side of two double pass heat exchangers connected in series. In the secondary loop, cooling water is pumped from Thompson Pond, a spring fed pond located approximately 650 h k
IV-6
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IV-7 yards southeast of the PSBR, through the tube bundle side of the two heat exchangers and then to a storm sewer which returns the water to r Thompson Pond. The lower quality secondary side water is passed through the tube bundles since by removing the ends of the two heat exchangers, the tube bundles can be cleaned more easily than the baffled shell side.
When the measured pressure difference shows the secondary outlet pressure to be less than 1.5 psig greater than the primary inlet pressure, a HEAT EXCHANGER SECONDARY PRESSURE LOW control room annunciator panel signal is activated.
Since the source of cooling water is Thompson Pond, which is fed by a 3 million gallon per day spring, relatively small year round variations in cooling water temperature are noted (55'F + 2*F).
- 6. Liquid Waste Evaporator Radioactive liquid waste is collected in either an underground holding tank just outside the evaporator building or in a holding tank
, below floor level inside the evaporator building. Liqui.d waste from either of these tanks (see Figure 4-4), can be pumped to the evaporator feed tank. This feed tank supplies the liquid to the evaporator during evaporator operation. The evaporator is a low pressure, low temperature unit which uses hot water as a heat source. The distillate from the evaporator is collected in a distillate tank. An overflow line installed in this tank directs the distillate to an underground holding tank where it is stored for later use as pool make-up water. The residue from the evaporation operation is removed from the evaporator, solidified and disposed by the Health Physics Office.
C. Water Quality Monitoring and Maintenance Specific conductivity, pH and gross radioactivity are used as a measure of water quality. Conductivity is monitored by conductivity cells in the pool water recirculation loop (see Figure 4-1). pH and gross radioactivity are measured in the laboratory using a grab sample of pool water.
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IV-9 l All three parameters: conductivity, pH, and gross radioactivity are controlled by filters and a demineralizer in the pool water recirculation loop. Abnormal levels in any of these parameters may indicate that this system is in need of service.
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V-1
/'~'\ V. FACILITY CONSTRUCTION V
A. Building The reactor is housed in the Breazeale Nuclear Reactor Building on the University Park Campus (Main Campus) of The Pennsylvania State University (see Figure 5-1). The original portion of the building, which contains the reactor and reactor control system, reactor pool, and the reactor pool water handling system was constructed in 1954. This original portion of the building is shown cross-hatched in Figures 5-2a and 5-2b. In 1961, an addition to the building added offices, laboratories, a classroom and two hot cells. A 1965 addition added a pool and more laboratory space to accommodate a large cobalt-60 source. The building is constructed of concrete blocks; bricks; insulated steel and aluminum panels; structural steels and re-enforced concrete and is, in general, fireproof in nature.
'The entire building is constructed on two levels with the beam hole
! /s laboratory (room 17) being at a lower level than the remainder of the ground
'\s -) floor. The reactor is housed in the pool in the reactor bay (room 123) located in the central portion of the building. The reactor controls are located in room 119 adjacent to the reactor bay. The water maintenance equipment (demineralizer, filters, heat exchanger and associated pumps) is located in room 9 directly below the control room. The pool water transfer equipment (pump and valves) to drain and refill the pool is located in room 17, the beam hole laboratory. Except for the hot cells (in rooms 15 and 16) and the cobalt-60 facility (rooms 20 and 130), the remainder of the building is utilized as research laboratories, teaching laboratories and faculty and staff offices.
B. . Heating and Ventilation The reactor bay is maintained at a negative pressure with respect to the atmosphere and the remainder of the building by one of two separate exhaust systems. Ventilating air to the room is supplied by leaks around .
doors,.etc. Normal ventilation of the reactor bay is accomplished by roof i fans which exhaust air to the atmosphere at roof level. This system is V
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( ,/ second system, an emergency exhaust system, is automatically started when the building evacuation alarm is sounded. This emergency system exhausts the reactor bay through roughing filters, absolute filters and charcoal filters all in series before discharging to the atmosphere three feet above the reactor bay roof.
A control / status panel for the emergency exhaust system is located in the reception area for the cobalt-60 facility. This area is designated RECEPT A in Figure 5-2a. This instrument panel is located such that the instruments on it could, in an emergency, be read from outside the building.
Instruments on the panel consist of four differential pressure gauges, three of which indicate pressure drop across, the prefilter, the absolute filter and the charcoal filter. The fourth differential pressure gauge indicates the velocity pressure in the stack. Also located on the panel are two pilot lights; one indicates that the system is energized, the other indicates that there is air flow in the system. The second light gets a signal from a flow switch in the duct. A switch on this panel permits manual activation of the
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() receipt of an evacuation alarm.
The alc in the reactor bay and control room is heated and cooled by a dedicated reactor bay air conditioner. This unit recirculates, heats, cools or dehumidifies reactor bay air as required. No air is interchanged with any other part of the building or outside of the building by this unit.
Heating is supplemented by steam unit heaters as needed. The condensate from the reactor bay air conditioner is piped into the reactor pool as makeup water to help compensate for pool water evaporation. A typical evaporation rate is 25.5 gal / day for the period November 1983 through April 1984. The beam hole laboratory (room 17) has a separate air conditioner to provide cooling to that area. Heat is supplied to this room by steam unit heaters located near the ceiling. No heating or cooling is provided for the domineralizer room (room 9). Steam for the heating system is supplied from University power plants located at the east and west ends of the campus (see Figure 5-1).
f3
V-6 C. Utilities Electric power is supplied to the facility through a dedicated three phase transformer located inside the reactor site boundary fence (see Figure 5-3). The power is supplied by the West Penn Power Company. An uninterruptible power supply (UPS) system is maintained in the reactor bay.
A battery bank in this system is maintained in a charged state by normal building power. The battery condition is checked daily and monitored continuously by a meter. The battery bank in turn supplies pcwer to the following facility devices so that in the event of a power failure, the device continues to operate normally:
- 1. Control room annunciator panel
- 2. Reactor bridge East and West radiation monitors 3 Reactor bay East and West air monitors (except for pumps)
- 4. Cobalt-60 bay area monitor
- 5. Beam hole laboratory area monitor
- 6. Several evacuation alarms
- 7. Building intrusion alarm system In the event of a power failure, emergency lighting is provided in twelve places throughout the PSBR building by individual battery packs.
Water is supplied to the facility from the University's water tupply from University owned wells located on University property.
Liquid propane gas is supplied to the laboratories from a tank located outside of the building adjacent to storage room 12.
Compressed air is supplied by two air compressors. A 1V horsepower compressor is dedicated to supply compressed air to the reactor transient rod drive. A 20 horsepower compressor supplies compressed air for general use throughout the building.
Both of these compressors are located in an equipment room (room 18) located under the loading dock adjacent to room 8.
D. Fire Protection The reactor building is equipped with an internal (local alarm only) fire alarm system. Fire extinguishers of either the CO2 type or compressed air and water type are located at strategio locations throughout the
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VI-1
~s VI. FACILITIES FOR EXPERIMENTERS v
A. Beam Ports Located in the south end of the reactor pool are seven beam hole penetrations through the pool wall as shown in Figur 6-1. The beam holes extend through the pool wall plus additional high density concrete which provides a total wall thickness of five feet. The pool side of each beam hole is sealed to prevent water leakage with a gasket and a blank flange.
The other end of the beam holes terminatos in a beam hole laboratory (BHL) which is available for dry irradiation purposes (see Figure 6-2 for location of BHL).
Doors constructed of 1/4" stainless steel plate and poured full of lead are hung on wheel and track arrangements on the BHL wall. These doors plus two aluminum shells, 2T' and 3' long, which are poured full of high density concrete, are used to provide shielding for each of the seven beam ports when they are not in use.
) The inside diameter measurement of each of the beam holes is as follows (see Figure 6-1):
Beam Ports #2, #4, #6 are 7" Beam Ports #1, #7 are 4" Beam Ports #3, #5 are 3" Because of mechanical limitations, the reactor cannot be moved closer thin 27" to beam port #4 Two methods of coupling the core to the beam ports are provided. The first is to position an air void (air filled cylinder) between the core and the beam port to displace the pool water.
The other is to use the D 02 tank in cases where thermal neutrons are desired. Of course, a third option of a combination of air void and thermal column exists.
Administrative control of the use of the beam ports to provided by PSBR Special Procedures.
Inflatable test plugs are available in the beam hole laboratory which can be used as an emergency repair in the event of a water leak through a beam port. Area monitoring is provided by a 0-H area monitor. This monitor
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VI-4 readout in the reactor control room. A TV camera is located in the BHL which can scan all areas of the lab. A remote TV receiver and the controls to operate the camera are located in the control room. A reactor power indicator is available to the experimenters in the BHL. Also available are two remote manual scram buttons on the east and weut walls of the BHL.
Finally, a light beam alarm (photo-sensitive relay) is energized when any one of the seven beam hole doors is opened. The light beam is located in such a way that it warns personnel of the potential hazard of walking in front of beam port #4. Both a local alarm and an alarm in the control room are provided if the light beam is broken.
B. D2 0 Thermal Column The D2 0 thermal column in use at the PSBR is constructed of 6061-T6 aluminum and the tank section measures 34" in diameter and is 27" long.
Although it is movable, its normal position is adjacent to beam port #4 (see Figures 6-1 and 6-3). The air filled flux trap that is built into the D2 0 tank is 9" long and 10" in diameter. The air filled flux trap creates an 18" thickness of D20 between the reactor core and the trap, which provides an optimum flux for the rabbit thimble and beam port #4.
Since mechanical interference prohibits the D 20 tank from contacting the beam port, an aluminum air filled extension is bolted to the beam port flange (see Figure 6-3) to remove pool water from the path of the neutron beam.
To protect the pool wall directly behind the thermal column from neutron and gamma radiation, a boral shield is fitted to the beam port end of the D20 tank. A second boral shield on the reactor end of the D 20 tank keeps the scattered neutron interactions with the reactor instrumentation detectors to a minimum. Additional gamma shielding is provided for the pool wall by positioning a 2' x 2' x 1" lead shield around beam port #4.
C. Central Thimble The central thimble, located in the radial center of the core, provides space for the irradiation of samples at the point of maximum neutron flux.
The thimble is an aluminam tube 1.5" 0.D. and 1 33" I.D. It extends from
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VI-6 -
the reactor bridge through the central hole of the removable hexagonal section in the top grid plate, through the bottom grid plate and is supported at its lower end by the safety plate situated below the bottom grid plate (see Figures 3-2 and 3-3).
The central thimble contains water since there are four 1/4" holes 3/4" above the safety plate which communicate to the pool water. Thus, there is no streaming of -neutrons or gammas to the reactor bridge as a result of using the thimble. A cutaway section in the central thimble allows irradiated samples to be removed below the pool surface, thus minimizing personnel radiation exposure.
- 1. Central Thimble Oscillator A vertical oscillator assembly is mounted on the reactor bridge and is used in conjunction with some central thimble sample exposures. The oscillator eliminates any flux gradient which would occur if multiple samples are vertically stacked and placed in a stationary position in the central thimble. By means of an electric drive motor and a roller chain moving over a set of sprockets, a nylon cord is made to oscillate vertically through the core at a maximum rate of 1.25 inches /sec. The stroke of the oscillator is 34 inches (see Figure 6-4).
D. Vertical Tubes Certain experiments cannot be submerged in water to be irradiated, e.g., electronic circuits. For this type of irradiation, vertical tubes are available. A vertical tube is simply an air filled aluminum tube which extends from the reactor core level to just above the reactor bridge floor level. The vertical tubes are weighted so that they do not float and in some cases the lower ends are designed to plug into the bottom gr.3 plate.
A variety of sizes of vertical tubes is available ranging up to 6" I.D.
Another type of vertical tube available has a 2" x 3" rectangular cross section.
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VI-8
- 1. Jib Crane A 1/2 ton hand operated jib crane is mounted on the floor of the reactor bridge. The jib cranc is used to support shielding plugs of lead and paraffin filled steel shells. The shield plugs are lowered into the open ended vertical tubes and extended below the pool surf ace to minimize radiation dose rates above the tubes.
E. Pneumatic Transfer Systems
- 1. Pneumatic Transfer System I This system provides a means of rapidly transferring samples between the laboratory wing of the facility and the reactor core.
The system is a closed loop design, with the major components being two core termini assemblies, a laboratory terminus and a blower and filter assembly with connecting tubing between these units. Carbon dioxide is used as the working fluid to reduce production of radionuclides. Four solenoid-operated valves control the fluid Flow, whereby the system operates on a pressure differential drawing the
" rabbit" into and out of the core by vacuum (see Figure 6-5). Two different core termini assemblies are provided in the system, one a bare aluminum tube,the other an aluminum tube lined with cadmium.
Components of the system considered most prone to leakage and most difficult to seal are enclosed in a gas-tight steel container. Since a surge volume is considered essential to the operation of the closed loop system, the containment is constructed large enough to serve as a reservoir for this surge volume. The electrically operated switching valves, and the system filter are enclosed in this containment.
Over pressure caused by any system malfunction would be released through a "U" tube filled to a predetermined level so that if the system pressure exceeds two psig the liquid is expelled from the "U" tube into a baffled receiver tank. The gas from the system escapes through a vent in the top of the receiver tank and subsequently through an absolute filter and a charcoal filter before being discharged to the atmosphere.
Positioning and design of this tank is such that the liquid will flow back into the "U" tube, thus, resetting the over pressure release automatically after the system pressure is reduced to normal.
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VI-10 The carbon dioxide working fluid is supplied from a high pressure cylinder through a standard two-stage regulator at about 20 psig to a fixed output regulator mounted on the containment box which further reduces the pressure to a few inches of water. When a pressure sensor between these two regulators senses a pressure less than 15 psig a LOW GAS PRESSURE light is lit on the reactor console.
With a closed carbon dioxide system, contamination with atmospheric air would lead to argan radioactivity in the system. The laboratory terminus is designed to minimize the entry of air into the system. The terminus is cylindrical and uses "0" ring seals on a sliding internal piston to minimize gas leakage (see Figure 6.6).
Measurements indicate that about one liter of gas per minute is lost through leaks in the system. Concentration of radioactivity in this gas, with the reactor operating at one megawatt, has been measured to be 1.5 x 10-3 pci/ml resulting in approximately 1.5 uci/ min escaping from the system during its operation.
Primary control of the system is from the reactor console. A MASTER I switch supplies power to a master relay which in turn permits the transfer fan to be operated by the FAN ON switch, opens an electrically operated valve which supplies CO 2 to the system, and turns on a chart recorder that records the output of the radiation monitor.
The AUDIO I & II switch allows the operator to hear when a sample enters and leaves the core via microphones attached to the system tubing near the core terminus. Verbal communication between the experimenter and the reactor operat0c is via an intercommunication system.
The experimenter in the laboratory selects either MANUAL or AUTO mode of operation. In automatic mode, a preset timer controls the length of time the sa.nple remains in the core.
A G-M tube in the containment box is connected to a monitor located on the Rabbit I panel in the reactor bay. An alarm on this radiation monitor will (a) prevent blower operation and (b) open an electrically operated valve to bypass the pressure relief manometer so that any pressure in the system is relieved thus minimizing leakage to the O
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building. An ALARM OVER-RIDE switch on the console allows operation of the fan with a radiation alarm.
- 2. Pneumatic Transfer System II This pneumatic transfer system provides a means of rapidly transferring samples between the ventilated hood in the reactor bay southeast annex and the reactor core or D 20 tank.
Electro pneumatic valves control the flow of nitrogen gas in the system to push samples to and from the core (see Figure 6-7). A high pressure cylinder supplies nitrogen through a regulator to a surge tank providing gas to the " sample in" side of the system. Another high pressure cylinder supplies nitrogen through a regulator to a second surge tank providing gas to the " sample out" side of the system.
Irradiation termini consist of a stainless steel reactor core terminus, a cadmium lined aluminum reactor core terminus and an aluminum terminus that fits into the D 20 tank thimble.
Primary control of the system is from the reactor console. A MASTER II switch supplies power to all the system components. A delay relay prohibits operating for 10 minutes af ter energizing this switch so the recirculation blower can cool the core terminus before system use begins. The cooling provided by this blower is necessary to prevent e sof tening of the polyethylene capsules when using the stainless steel core terminus at higher power levels. After the 10 minute delay, the reactor operator can depress the FIRE PERMIT switch to allow the experimenter to send the samples. Depressing the AUDIO I & II switch will allow the operator to hear (via microphones attached to the system tubing near each terminus) when the sample enters and leaves a core terminus. Verbal communication between the experimenter and the reactor operator is via an intercommunication system.
The send station, the receive station,and all valves which release gas from the system are housed in hoods equipped with a ventilating system which exhausts to the outside through an absolute filter in the reactor bay.
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VI-14 F. Instrument Bridge The instrument bridge is a device which provides three dimensional positioning of experiments in the vicinity of the reactor core. This bridge, which is mounted on wheels that roll on the same track on the pool wall as the reactor bridge, is constructed of steel I-beams. A tower constructed of aluminum tubing extends from this bridge down into the pool to below the level of the reactor core. The instrument bridge can be easily disassembled and removed from the pool by using the overhead crane or reassembled on the other side of the reactor bridge if needed. A typical location of the instrument bridge is shown in Figure 6-1.
G. Hot Cells Two hot cells are available at the PSBR for the safe handling of radioactive materials (see Figure 6-2 for location). The cells are constructed of high density concrete 2' thick and each cell has inside dimensions of 5' x 7V' x 13'. E'ch a cell has a 20V" x 30V" x 27" lead glass window and is designed to accomodate a 100 curie Co-60 source or its equivalent. Master-slave manipulators are provided in each of the cells along with a y ton remotely operated crane.
Access to the hot cells is provided by locked concrete doors or by overhead shield plugs that can be removed by a crane in the reactor bay area.
Access to and utilization of the hot cells are governed by PSBR Auxiliary Operating Procedures.
H. Co-60 Irradiation Facility An 18,000 curie gamma source composed of 70 aluminum clad pencils of Co-60 and 80 stainless steel encapsulated pencils of Co-60 is housed in a stainless steel lined pool which measures 10' x 16' x 17' deep (see Figure 6-2 for location). A variety of vertical tubes is available for dry irradiations. Exposure rates up to 500,000 R/hr are available through various arrangements of the pencils around the tubes.
O
F VII-l VII. CONTROL AND INSTRUMENTATION
(~~}
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A. Control System Summary The control and instrumentation equipment described in the following sections (see Figure 7-1 for physical placement) was designed to provide the PSBR with safe and reliable operation in five different modes. Selection of AUTOMATIC, STEADY STATE, SQUARE WAVE, PULSE HI, or PULSE LO mode of operation is made by positioning the MODE SELECTOR switch to the desired mode at the appropriate time in accordance with PSBR Standard Operating Procedures.
The various signals, scrams, interlocks and readouts described in the following sections are obtained through the use of four detecting channels; namely the Start-up Channel, the Log Power and Period Channel, the Linear Power Channel, and the Percent Power Channel. The detectors used in the PSBR instrumentation and control system are: two compensated ionization chambers, one fission chamber and one gamma ionization chamber.
With the four detectors and the four detecting channels, a range of
~3 s_, ) power from 10-3 watts to 2 x 109 watts is covered with appropriate overlap from channel to channel.
B. Steady State Mode
- 1. General Steady state mode allows operation at power levels up to 1 megawatt.
This mode is used for start-up and is used for traversing the sub-critical to critical region regardless of what mode is ultimately to be used.
- 2. Start-up Channel The start-up channel consists of a fission chamber which is movable.
When the reactor power reaches 1 kilowatt, the fission chamber is automatically withdrawn about 14" from the reactor core and subsequently.
returns to its start-up position when power drops below 200 watts.
The fission chamber signal ultimately gets to the log countrate
() amplifier by passing through a preamplifier then through a pulse
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, amplifier where discrimination against gamma background takes place.
v From the log countrate amplifier, three outputs are provided. The first output goes to a log countrate meter which indicates 1 count /see to 105 counts /sec. The second output drives a low countrate bistable which prevents rod movement if the countrate drops below 2 counts /sec. The I bistable resets at 3 counts /sec and allows rods to be moved again. This rod block feature in no way interfers with the scram functions for the reactor. The third output goes to the red pen of a dual pen chart-recorder where the log of the countrate is displayed.
An optional piece of equipment located in the control room, a scaler, receives a signal from this channel in the control room (see Figures 7-2, 7-3 for block diagram).
3 Log Power and Period Channel The signal from a compensated ionization chamber (CIC) provides the
-input signal to a log power amplifier in the log power and period channel. Three outputs are provided by the log power amplifier. The g first output is time-shared on the red pen with the log countrate output.
A COUNTRATE RECORD switch allows the operator to select either the log countrate or the log power signal to be displayed on the recorder. The log power scale goes from 0.3 watts to 2 x 106 watts. The second output goes to a 1 kilowatt bistable. It has already been explained that one function of the 1 kilowatt bistable was to drive the fission chamber between its upper and lower operating positions. A second function of the 1 kilowatt bistable is to prevent pulse operation from power levels greater than 1 kw. Finally, the 1 kw bistable provides a low countrate interlock defeat to the low countrate bistable in the start-up channel.
This interlock defeat permits the operator to move control rods even if the countrate is less than 2 c/s. Such a low countrate could occur due to the above mentioned fission chamber movement. The third output of the log power amplifier passes through a differentiator network to provide the input to the period channel.
From the period amplifier in the period channel there are four outputs. The first two outputs go to a period meter with a range of -30 to +3 seconds and a decade / minute meter which has a range of -1 to 8 d/m.
O v
The third output goes to another bistable that provides a scram signal
Figure 7-2 Block Diagram of the PSBR Control Console Instrumentation H
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VII-6 if the reactor period ever becomes shorter than +3 seconds. The last output of the period amplifier goes to the automatic controller (servo system).
- 4. Linear Power Channel The output of a second C.I.C. is amplified by a multirange linear amplifier. A 15 position REACTOR POWER switch provides range selection from 0.1 watt to 106 watts. The output of the linear amplifier goes to three places. First, the signal goes to the blue pen of the dual pen chart-recorder where linear power is recorded as 0 to 110% of range.
The second output goes to a 1.1 range bistable. This bistable provides a scram signal if the linear power indication ever exceeds 110% of any range. The third output of the linear amplifier goes to the automatic controller.
- 5. Percent Power Channel The output of a gamma ionization chamber (G.I.C.) is fed through a linear amplifier to a 5 power meter which indicates 0 to 110% of rated power (106 watts). The prime function of this channel is to scram the reactor at 10% over rated power. This scram is accomplished by an optical type sensing relay on the % power meter.
- 6. Temperature Two temperature meters are provided on the reactor control console.
One of these meters, with a range of O'C to 700*C, receives a signal from a thermocouple in an instrumented fuel element. This meter provides a high fuel temperature scram by means of an optical relay in the meter. The second meter receives a signal from a resistance bulb thermometer suspended in the pool from the reactor bridge and displays bulk pool temperature on a range from O'C to 60*C. A continuous bulk pool temperature monitor receives a signal from a second resistance bulb thermometer in the pool and provides an alarm at a preset high pool water temperature.
O
VII-7 7s 7. Manual Rod -Drive Control
( ,) The three rack and pinion operated drives are controlled by UP and DOWN momentary contact push buttons, which become illuminated when the full up or down positions are reached. The cylinder of the pneumatic rod can be similarly controlled. If air pressure is applied to the pneumatic rod system to hold the rod in the UP position relative to the cylinder, it can then be used as a manually adjustable control rod by using the push buttons to drive the cylinder up and down. It is intended to be withdrawn in this fashion for use as an ordinary control rod in the steady state mode. Under these conditions, it is dropped by a signal from the scram circuit which releases the air pressure in the cylinder. The solenoid valve which controls the air pressure is interlocked so that in the steady state mode, air pressure can be applied only with the cylinder in the DOWN position. This eliminates any possibility of accidentally producing a step increase in reactivity.
,_s C. Automatic Mode
/\
( ) The PSBR automatic controller system is designed to change power automatically or to maintain power automatically. A servo amplifier compares a signal from the % demand potentiometer with that from the linear channel and adjusts the regulating rod to maintain the desired power. A period signal is also fed into the servo amplifier and used to maintain a period greater than 10 seconds. Once the reactor has been taken critical under manual control, the multi-range linear amplifier can be set on the desired range, and the reactor switched to automatic control. The power will be increased to the demand level automatically. If in the course of this start-up, the regulating rod approaches 25% or 75% of its travel, the shim rod will be automatically moved to assist the regulating rod.
D. Pulse Mode
- 1. General Pulsed operation is produced by the rapid removal of a predetermined portion of the transient rod. This results in a step
'N insertion of positive reactivity. There are two modes of pulse
[G
VII-8 operation available to the operators of the PSBR. PULSE LO mode is used for pulsed operation up to 400 megawatts. PULSE HI mode is used for pulsed operation up to 2000 megawatts; the selection being made with the five position MODE SELECTOR switch. The only difference in the two pulse modes is the range of the power readout. The transient rod is automatically scrammed by a preset timer following the initiation of a pulse.
- 2. Peak Pulse Indication When either of the two pulse modes of operation is selected, the start-up channel, the log power and period channel, and the linear C.I.C. are deactivated. The output of the G.I.C. from the % power channel is fed through an amplifier-memory circuit and the linear amplifier of the linear power channel. From there it is displayed on the blue pen of the chart recorder af ter a short time delay.
3 Temperature Fuel and water temperature indicators are the same as for steady state operation except that a second thermocouple in the instrumented fuel element is connected to the red pen of the chart-recorder providing a readout from 0 to 700*C on the linear chart paper.
- 4. Rod Control Once the mode selector switch has been turned to either pulsing mode, interlocks prevent movement of any rods except the transient rod.
Movement of the transient rod cylinder is also prevented.
- 5. Pulse Control A pulse is initiated by first taking the reactor critical and to some power level less than 1000 watts in the steady state mode. The 1 kw bistable prevents pulsed operation from power greater than 1 kw. The mode selector switch is then turned to one of the pulse positions. At this time, the TRANSIENT ROD FIRE button is depressed causing air to be admitted to the transient rod cylinder. Until a pulse mode is selected and the range switch is turned to the 1 MW (PULSE) position, the fire button does not illuminate nor does it perform its function; thus, g
VII-9 b N ' eliminating the possibility of an inadvertent pulse. At the end of a pre-set period of time,.the air pressure is automatically removed, allowing the rod to fall back into the core. The time period is variable from 0.15 to 15 seconds, normally a 3 second delay is used.
E. Square Wave Mode
- 1. General For some irradiation applications it is desirable to bring the reactor power level rapidly to a predetermined value, hold it there for
> a short duration,then terminate the exposure by a manual scram action.
This can be accomplished by selecting the square wave mode of operation.
Square wave is a combination of the features of steady state,
. pulse, and automatic modes of operation. In steady state mode, the reactor power is stabilized at a level of 1 KW or less. The pulsing feature of the transient rod is then used to insert a step function of reactivity not to exceed $1.00. Finally-when the desired square wave power is reached (as a result of the reactivity insertion), the O automatic control system is automatically activated to maintain the desired power. level.
The. control configuration which is operative in square wave mode is essentially the-same as steady state, with the exceptions noted in the-following sections:
- 2. Log Power and Period Channel The' input to the period amplifier'is connected to ground when the
- mode selector switch is in the square wave position. There is, therefore, no period information displayed on the console and the period scram circuit is not active in this mode of operation.
.3 Linear Power Channel An administrative control prohibits square wave initiation from a power level greater than 1 kilowatt.
e
VIl-10
- 4. Manual Hod Drive Control For square wave mode of operation, the transient rod drive cylinder is adjusted so that when air pressure is applied to the cylinder, the movement of the rod will insert that amount of reactivity for the j reactor to attain the desired power level. At the end of the desired exposure time, the reactor is shut down by manual scram. Fine adjustment of power level is made by the automatic controller system.
It is not until square wave mode is selected on the mode switch that the transient rod fire button illuminates, allowing the application of compressed air to the cylinder. This prevents inadvertent firing of the transient rod.
- 5. Automatic Control Because a rapid rise in power is desired in square wave mode of operation, the automatic controller system is placed in a standby (ready) condition until the desired power level is attained. Closed loop power control by the automatic controller system is initiated when the linear power (signal from the linear channel) reaches the preset value (as determined by the percent demand potentiometer).
F. Control Room
- 1. General Most of the equipment and all of the readouts described in sections A through E are located in the control room. The control room is located on the west side of the reactor bay. A safety glass window is provided between the control room and the reactor bay such that an operator seated at the reactor controls can observe personnel movement in the reactor bay. A closed circuit TV system is also provided in the control room so that the operator can observe personnel movement in the beam hole laboratory.
Three internal communication systems and a commercial telephone are available to the control room operator. The internal communication systems allows a) two way conversation with anyone on the reactor bridge and the experimenters using the pgeumatic transfer systems at any of the sending stations; b) two way conversation between twenty-four L
VII-ll offices or laboratories; and c) the uee of a page system that has
}r s (s_,/ speakers in all parts of the building.
A plexiglass window between the control room and a public corridor allows visitors to conveniently view the reactor controls; it allows
- Police Services to observe any unusual indications on their inspection tours; and it allows the reactor staff to observe instrumentation from outside the reactor bay.
- 2. Monitor Indications in the Control Room There l's an instrumentation pedestal located at either end of the control console in the control room. Figure 7-4 shows the equipment that is mounted in each of the pedestals. Table 7-1 lists the monitors
.that are equipped with alarms, their detectors, settings, and ranges.
An alert (as used in Table 7-1) results in an amber warning light coming on. Table 7-2 lists the various parameters that are represented on the control room annunciator panel.
Monitor #1 and Monitor #2 on the left pedestal in Figure 7-4 monitor the condition of the equipment used to transmit information to 7-University Police Services.
I
\
TABLE 7-1 CONTROL ROOM
- ALARMED RADIATION MONITORS Monitor Detector Range Setting Reactor Bridge Ionization 0.1 to 2 x 103 mR/hr Alert: 15 mR/hr East Chamber Alarm: 50 mR/hr Reactor Bridge Ionization 0.1 to 2 x 103 mR/hr Alert: 30 mR/hr West Chamber Alarm: 200 mR/hr Reactor Bay Thin End 10 to 105 c/m Alert: '4000 c/m Air East G-M Tube Alarm: 5000 c/m Reactor Bay Thin End 10 to 105 c/m Alert 4000 c/m Air West G-M Tube Alarm: 5000 c/m Co-60 Bay G-M Tube 0.1 to 104 mR/hr Alarm: 6 mR/hr Beam Hole G-M Tube 0.1 to 104 mR/hr Alarm: 6 mR/hr Laboratory
- Alarmed is defined as scramming the reactor and sounding the building evacuation alarm.
.f"m Figure 7-4 Control Room crumentetion Pedratals LEFT PEDESTAL RIGHT PEDESTAL
?
Y M
D/M Meter Beam Hole Co-60 Bay Monitor #1 Monitor #2 Lab Monitor Monitor Control Room Annunciator Panel (See Table 7-2 for Detail)
Reactor Bay Reactor Bay East Monitor West Monitor SCALER East Air West Air Monitor Monitor TIMER Fission Product Activity Monitor i
l l Beam Hole Laboratory l Remote TV Camera Controls i
O O O
VII-13
,-s All of the instrumentation referred to in this section is checked (v) for operability weekly in accordance with PSBR Standard Operating Procedures. The alarmed monitors are calibrated at least annually according to internal Checks and Calibration Procedures (see Figure 7-5).
Table 7-2 CONTROL ROOM ANNUNCIATOR PANEL INDICATIONS
- Reactor Bay Radiation Level High
- Reactor Bay East Air Activity High
- Reactor Bay West Air Activity High
- Co-60 Bay Radiation Level High
- Beam Hole Laboratory Radiation Level High
- Reactor Pool Level Low
- Co-60 Pool Level Low
- Fission Product Activity High
- Waste Tank Liquid Level High Fission Product Monitor Flow Low Fission Product Monitor Flow High Reactor Pool Conductivity High Light Beam Alarm Beam Hole Laboratory Pump Room Radiation Level High Heat Exchanger Secondary Pressure Low Transient Rod Air Pressure Low
/~'N Building Air Pressure Low
( ,)
Reactor Pool Water Temperature High
- Alarmed as defined in Table 7-1
- Signal to Police Services G. Minimum Safety Circuits and Interlocks Tables 2a and 2b in the Technical Specifications section list the minimum, safety circuits and the minimum interlocks for the PSBR.
m
m Reactor Bay Thin End Local Light <
Air East G-M Tube _
Meter and U Recorder h
Cobalt-60 [ _
Control Room Bay G-M Tube L Annunciator Q Local Light and Alarm Control Room Remote Readout Panel l
Reactor Ion Module ~
Police Service -
Bridge East Chamber 7 ] Alarm Panel Evacuation Alarm Reactor Ion _
Emergency Exhaust Bridge West Chamber _
Alarm Light System - ON Local Light West Stairwell Roof Fans - OFF and Alarm Panel Beam Hole _
Laboratory G-M Tube Reactor Bay Thin End _
Local Light Air West G-M Tube Meter and IAS Recorder I
Control Room Control Room (Fission Product NaI Remote Readout Annunciator Monitor) Crystal Module Panel Demineralizer f Room Local Light G-M Tube and Alarm Control Room (Area Monitor) '
l Annunciator Panel Hot Cell G-M Tube Local Light
" Hot" Side and Alarm Figure 7-5 Hot Cell G-M Tube Local Light Area and Air Particulate
" Clean" Side and Alarm Monitoring System O O O
VIII-1 VIII. CONDUCT OF OPERATION A. Organization and Responsibility The Penn -State Breazeale Reactor (PSBR) is operated by the Director for the Nuclear Engineering Department whose Head reports to the President of the University through the Dean of the College of Engineering and the Vice President for Research and Graduate Studies. An organization chart is presented in Figure 8-1.
Responsibility for the safe operation of the PSBR lies with the Director and the operations staff which is composed of permanent cumulative University employees. Presently, the permanent staff has a cumulative total of 186 years of experience operating the PSBR.
The Penn State Reactor Safeguards Committee, an independent group with technically experienced members from both within and outside the University, advises the Director on all matters or policy pertaining to safety. Members are appointed by the Dean of the College of Engineering, acting for the Vice O President for Research and Graduate Studies.
The University Health Physics staff, also independent of the reactor administration,' provides "onsite" advice concerning personnel and radiological safety and provides technical assistance and review in the area.
of radiation protection. The Health Physcia office is one of the Intercollege Research Programs and reports to the office of the Vice President for Research and Graduate Studies.
B. Reactor Operating Safety Philosophy All operations involving the PSBR shall be conducted in compliance with
' pertinent existing local, state, and federal regulations. The reactor shall be operated within the limits established by the operating license and the
, technical specifications. An ALARA program (as low as reasonably achievable) will be in effect to minimize radiation exposures to the public, the staff, and the environment.
10 i
k _
s VIII-2 I
l President '
The Pennsylvania State University i
Vice President for Research and Graduate Studies Dean College of Engineering Nuclear Engineering Department - Head Administrative Aide Low-Level Radiation Monitoring Laboratory University _____
PSBR __
Penn State Reactor Health Physics Director, Deputy Director Safeguards Committee Radionuclear Secretarial Staff Applications Laboratory I I I I Instrumentation Training Operations Supervisor Machine Shop and Control & Bldg. Main.
Senior Reactor Operators Reactor Operators Figure 8-1 ORGANIZATION CHART
VIII-3 7-~g C. Training k,_,) The competence of the PSBR operators and senior operators is maintained by the PSBR requalification program. This progran keeps the reactor staff cognizant of features of facility design, reactor principles and operating characteristics, reactor instrumentation and control, safety and emergency systems, emergency and standard operating procedures, administrative and special procedures, radiation control and safety, and other facility information necessary for a licensed individual to perform his or her duty.
Competency of licensed individuals to respond appropriately to the emergency plan is maintained by an annual review of the emergency plan for the reactor staff by the Emergency Director, and by participation of licensed individuals in drills required by the emergency plan.
D. Written Procedures The practical application of the philosophy of safe reactor operation set forth in section B above is augmented by detailed written procedures.
These procedures fall into four general categories, Emergency Procedures O)
( (fire, civil disorder, loss of pool water, etc.): Standard Operating Procedures (operation of the reactor, instrumentation checkout, fuel handling, etc.): Special Procedures (beamport utilization, cooling systems, etc.): Administrative Policies (personnel requirements for operation, facility keys, reactor safeguards committee procedures, etc.).
The written procedures are approved by either the Director or Deputy Director of the PSBR. Further, these procedures govern the activity of all staff personnel, experimenters, and visitors while in the PSBR. The procedures are reviewed annually and the review is documented.
l i
E. Records A daily reactor operations log is maintained by reactor operating personnel and contains such information as core loading and changes, experiments in the reactor including time in and out, power level, startup and shutdown times, control rod positions, and calibrations and maintenance notations.
O L -) t
VIII-4 Separate and more complete files are kept on reactor instrumentation readings, checkouts, calibrations, maintenance and other items concerned with operational aspects of the facility. All unscheduled shutdowns are reviewed and any corrective action necessary is performed before restart.
Records are maintained which indicate the review, approval, and conditions necessary for the production of radioisotopes and/or the performance of irradiation experiments.
F. Review and Audit of Records The Penn State Reactor Safeguards Committee (PSRSC) acts as a review panel for any reactor experiments which, by their unusual nature and/or potential hazard or unprecendented complexity, could endanger health, life, and property in and about the PSBR.
The PSRSC also provides for an annual outside, independent audit of the operation of the PSBR facility.
O O
IX-1
/
'~ ') IX. SAFETY EVALUATION i
'J A. Introduction Fuel management studies at the Penn State Breazeale Reactor (PSBR) performed in 1972(1,2) showed both experimentally and analytically that refueling the PSBR with 12 wt$ U-ZrH TRIGA fuel instead of 8.5 wt5 U-ZrH f " a '. reduces the fuel costs by a large f actor. The optimum core position for replacing used 8.5 wt$ fuel with the 12 wt$ fuel is in the center-most ring, the B-ring. The power density in the 12 wt5 fuel is 35% greater than that in the 8.5 wt% fuel when both are in the B-ring. Thus, the 12 wt% fuel produces correspondingly higher fuel temperatures. As the core depletes,
.the partially depleted 12 wt1 fuel is moved outward replacing 8.5 wt% fuel.
The new 12 wt% fuel is placed in the B-ring. The maximum power density in the new 12 wt5 fuel is the same at the beginning of cycle (BOC) as in the first case (35% higher than the 8.5 wt1 fuel).
g'~'S Since 1972, the PSBR has continued to be refueled with 12 wt$ fuel.
\s_s/ Several studies have been made(3-7) regarding the reactor physics and heat transport characteristics of the 12 wt% and 8.5 wt% fueled PSBR cores.
These studies show that the calculational techniques agree closely with the experimental data.
The principal computen programs used to perform the calculations-are PSU-LEGPARD(8), EXTERIMINATOR-2(9), MCRAC(10), and SCRAM (11). PSU-LEOPARD incorporates the standard LEOPARD (12) computer program as originally received and adds additional subroutines. LEOPARD and PSU-LEOPARD calculate the group constants of the core as a function of burnup.
MCRAC is an automatic, multicycle, two-dimensional depletion code that gives the power distribution, kerr, and isotopic inventory of the core at each burr.up step. It is based on the flux and keer calculation performed by EXTERMINATOR-2, a multigroup two-dimensional dif fusion theory code.
SCRAM is a multicycle depletion code created specifically for TRIGA reactors and adapted to the PSBR lattice design. It uses analytical equations to compute the power distribution, kerr, and isotopic inventory
,-s for each cycle. The analytical equations are based on diffusion theory and l
- (~.-) the empirically fitted constants are derived using the PSU-LEOPARD, t
IX-2 EXTERMINATOR-2, and MCRAC codes. In general, the calculations give good agreement with the measured power distributions and neutron fluxes.
The fact that we calculate the power distribution with good accuracy is important to calculating the safety margin in PSBR operation. The calculations identify the fuel element having the highest power density in the core, and thus the one which will produce the highest fuel temperatures.
This is true for both steady state and pulse operations.
The experimental and analytical studies which have been performed to show the safety margin in the operation of the PSBR are described in this section. In particular, the maximum measured fuel temperatures during steady state operation and pulse operation are mathematically related in a unique way to provide an accurate prediction of their values during the PSBR operation. All predictions indicate that the design and construction of the PSBR 13 such that excessive temperatures cannot be attained during steady state operation. Further, any pulse temperature of too large a magnitude can be prevented by reviewing the steady state temperature measurements prior to pulsing a large excess reactivity into the core. This is also true for abnormal operating conditions. The following accidents are analyzed:
(1) The loss of coolant accident.
(2) The design basis accident which includes cladding rupture.
(3) A reactivity accident.
The results of these analyses demonstrate that the reactor can continue to be operated safely with negligible environmental impact or effect on the public health and safety.
B. TRICA Fuel Temperature Analysis of the Penn State Breazeale Reactor TRIGA fuel elements are considered damaged and no longer useable if their cladding has been ruptured or their dimensions change to where the transverse bend exceeds 0.125 inches over the length of the cladding or its length increases 0.125 inches. There are two limiting conditions for establishing maximum allowed fuel temperatures. When the reactor is operating in the pool, the maximum allowed fuel temperature at any location is 1150'C. Under these conditions, the fuel cladding temperature is less than 500*C and the cladding will not be ruptured by the internal hydrogen '
pressure.(13) During a loss of coolant accident (LOCA), the fuel is not
. . ~ . . ~ ~ ~ . - - . - - - - . - - - .. -. -.-
IX-3 covered with water and must be air cooled. Under these conditions, the maximum fuel temperature allowed at any location is 900*C. When the fuel is air cooled, the cladding temperature will go above 500'C, where the strength of the cladding decreases. Below a fuel temperature of 900'C the hydrogen pressure will not rupture the cladding under air cooled conditions.(13)
Flux gradients across the fuel produce uneven temperature distributions.
Pulsing a TRIGA fuel element to high power densities produces sudden expansion and contraction.- During the rapid expansion phase, a large temperature gradient ~ in the radial direction can cause uneven axial
. expansion producing a transverse bend. Experience with the TRIGA fuel elements has shown that they can receive thousands of pulses without being damaged'provided their temperature limits are not exceeded. The temperature distribution in a single TRIGA fuel element is a function of its fuel and fission product distribution and content and is different for pulse operation than for steady state operation.
During' steady state operation, the maximum fuel temperature is at the central fuel-zirconium rod interface. Since the thermocouple'is placed near this-juncture, the measured fuel temperature is close to maximum fuel temperature.(D This is not true during a pulse. During a pulse, the maximum fuel temperature is near the fuel-cladding interf ace and the measured fuel temperature is about 60-655 of the maximum fuel temperature.(D To prevent-fuel damage, it is.important to understand the TRIGA fuel temperature distribution in a fuel element during steady state and pulse operation and to relate.the measured fuel temperature to the maximum fuel temperature.
An. instrumented TRIGA fuel element is built with three thermocouples ,
placed 0.69 cm radially from the center, but spaced. vertically I' inch apart.
The middle thermocouple is in the midplane of the' fuel region of the TRIGA fuel element. The thermocouple measures the fuel temperature at a specific e
- point within the fuel element which is not the maximum fuel temperature for j pulse operation. During a pulse, the temperature distribution is the same as that of,the volumetric thermal source strength, q(r), so that the peak j fuel temperature is near the fuel cladding interface. As a consequence, the.
i
[ measured fuel temperature can be significantly lower than the maximum fuel temperature, particularly'if the thermal neutron flux self-shielding is high.
t On the other hand, the peak fuel temperature during steady state operation E____________________.___.________________._
g IX-4 is at the inner boundary of the fuel; thus, the measured fuel temperature is slightly less than the maximum fuel temperature. The measured fuel temperature in an 8.5 wt% fuel element is closer to the maximum temperature than it is in a 12 wt% fuel element because the self-shielding of the 12 wt%
is greater than the 8.5 wt% producing a q(r) with a steeper gradient.
The temperature distribution within the fuel can be calculated from a knowledge of fuel geometry, heat transport parameters, and q(r) as shown by Haag and Levine.(3) The volumetric thermal source strength in a fuel element is a function of the core power, the core configuration, and the element's position within the core. For any core configuration, q(r) can be determined by neutronic analysis and then used to determine the peak temperature during a pulse or during stealy state operation. The steady state fuel temperature is determined primarily by the boundary conditions at the cladding water interface. Studies peaformed by Haag and Levine have shown that subcooled boiling takes place in the PSBR when the TRIGA core exceeds 200 KW, This helps limit the temperature rise of the cladding surface temperature, t e, because when boiling occurs te increases proportional to approximately (q")0.33.(14) Hence, the heat flux, q", must increase by a factor of 8 to increase the difference between t o and the water saturation temperature by a factor of 2.
It is important to recognize that the q(r_) produced in a fuel element for a particular core configuration is the heat source that establishes the fuel temperature for both steady state operation and pulse operation. Hence, there is a direct relation between the measured fuel temperature at steady state and during pulse operation for the same core configuration and fuel element. The fuel temperature measured in the fuel element having the highest power density in the core during steady state operation can be used to determine the maximum fuel temperature in the core during a pulse. This is described in this section.
- 1. Steady State Analyses Standard heat transport calculations are used to analyze the steady state fuel temperatures for the PSBR.
Let gj(r) = volumetric thermal source strength at position r,within the jth region
IX-5
, Pj = power generated by the jth fuel element
"\-%
Then Pj =
fyqj(r)dV (1 ) 5 where the integral is'over the fuel volume of the jth fuel element, V.
The average power, P, produced by a fuel element in the core, and the normalized power for the jth fuel element, NPj, are related to Pj by the expression Pj =
F NPj (2)
A core of Nofuel elements producing a total power of Q megawatts or BTU per hour can be expressed as Q.- No P = No qV (3a)
~
and P- -
Q/N o = qV (3b) where q is the volumetric thermal source strength averaged over all fuel in the core, and No is the number of fuel elements in the core.
It can be immediately observed that operating at 1 megawatt, F is inversely related to No. Thus, for a core where Ne = 90, P =
0.0111 MW, and for a core with No = 100, F = 0.010 MW which is 10%
less.
Using Goodwin's(15) measured q(r,z) as gj(r,z) = (Ao + Be r2)q(z) (4) where for the jth fuel element ;
'h V
IX-6 jyqj'(r,z)dV = 4ev, (5) z is the axial position along the jth fuel element, and Ao and Bo are constants. A thermocouple is located at the fuel midplane where z =
7.5". The length of the fuel is 15". Thus, in general, gj ' ' ' (r ,7. 5" ) =
goj(r) =
ioj(Ao+Br) o 2 (6) and goj =
fa9j
- f aNPjg =
f aNPjP/V (7) where fa is the axial hot channel factor. The following definitions are used for the jth fuel element in these equations:
qj(r,z) = point volumetric thermal source strength gj = volumetric thermal source strength averaged over the radial direction gj = volumetric thermal source strength averaged over the fuel volume When the jth subscript is missing, q and q refer to the fuel element producing an average power in the core.
For the jth fuel element Pj =
6jV (8a) and q = P/V (8b)
Eq. (8a) can be written, using Eq. (2), in the following form:
. IX-7
- O The temperature rise between the fuel and the cladding at the fuel element midplane during steady state operation is directly dependent on goj(r). Dropping the j subcript for convenience but remaining in the fuel midplane fh krf
= q (r) g
= q (A + B r2), r < r > R (10) where rz = radius of Zr rod in the center of the fuel rod R = radius of the fuel rod
. Integrating Eq. (10) gives t(r) - ts . (32 - p2) o- zA r2 Inf+f(R4-r)- o#2 4 1" (11) i All variables in Eq. (11) are described and given in Table 9-1.
Substituting the values for the above constants from Table 9-1 into Eq.
D (11) gives for the thermocouple temperature, ttc.
(
tto - ts = 6.039 x 10-5 go (12a)
Substituting Eq. (7) into Eq. (12a) and returning to the -jth fuel element gives (tgo - ts)J =
6.039 x 10-5 NPj (12b)
Using Q =
1 y in Eq. (3b).
E- 3.412 x 106 BTU /hr (13)
The volume of fuel in a TRIGA fuel element is V =
w(R2~Pz )H s
IX-8 Table 9-1 Parameters for the 12 wt% U-ZrH TRIGA Fuel Elements (Enriched to less than 20% 235u)
Thermocouple radius, Rtc 0.0226 ft Fuel mean radius, R 0.0596 ft Zr rod radius, r z 0.0079 ft Fuel element radius, R+C 0.06125 ft Conductivity cladding, k e 9.5 Btu /hr ft*F Conductivity fuel, kr 10.5 Btu /hr ft*F Core average volumetric 2.49 x 108 Btu thermal source strength, It hr ft3 )
AO (12 wt% new)(Reference 4) 0.6534 Bo (12 wt% now)(Reference 4) 202 ft-2 Number of elements, No (Loading 36) 94.6 Axial hot channel factor, f a 1.35 Prompt temperature coeff., a (Ref 19) - 1.4 x 10-4 6k/k/*C Cr (Loading 36) Correction Factor 0.98 O
IX-9 O
ch where H is the fuel height. Using H = 1.25 ft and the value from Table 9-1 for the other parameter l V =
1.37 x 10-2 ft3 (14)
Substituting Eqs. (13) and (14) into Eq. (8b) gives q -=
Cr Nc BTU /hr-ft3 (15)
'where Cr is a correction factor for setting the linear recorder to read 1 MW when the actual power is reduced by the factor Cr. This reduction is presently made to prevent accidentally exceeding 1.1 MW to provide a safety margin to compensate for an uncertainty in calibration or scram setting.
Eq. (12b) can now be written as:
b (tte - ts)j =
1.5 x 104 Cr fP3 'F (16) c (tte - ts)J =
83.3 x 102 C fhD c
'C The measured fuel temperature, ttc, depends on the temperature at the cladding surface in the fuel midplane, te . Because of subcooled boiling above 200 KW, this temperature rises very slowly. The At is proportional to (q)0.33, where at is the difference between toand the saturation temperature.(14) As a consequence, it is assumed that the surface cladding is superheated by a fixed at degrees and thus at 1 MW, to = 140*C. This should be correct within + 10*C at 1 MW for all NPj's greater than approximately 1 and less than 3 We may write ttoj =
(tto - ts)J + (ts - t )jo + te (17) where the first term on the right hand side of Eq. (17) is evaluated using Eq. (16). .The second term, the temperature change between the fuel cladding and the surface of the fuel rod, is derived assuming a
IX-10 gap, g, between the cladding and the fuel. Solution of the standard heat equation gives for (t 3 -t)j c =
(t3 - t g)j + (tg - tc)j (18a) the following equation:
(t8 - t o)3 =
M2nk gin b R
+ M2wk in *E' R+C (18b) e where t3 -
temperature of the fuel at the fuel cladding interface tg -
temperature of the cladding f acing the gap kg = conductivity of the gap kc = conductivity of the cladding C = thickness of the cladding g - thickness of the gap gej = linear heat generation rate and the other parameters are as previously defined.
Thus (t3 - t g)j = In (19a) and (t8 - b in *E* b In t c)3 =
2nk R+g 2nk R (19b) e e Equation (19a) will be evaluated experimentally as described later, whereas Eq. (19b), the temperature drop across the cladding, can be evaluated from the physical values of the parameters.
By definition q'cj =
w(R2 -pg 2) q0j ese =
w(R2 _pz2)f NPj 3 c"[' ' ' (20)
Assuming an average core temperature drop across the gap, Itg , Eq.
(19a) becomes
IX-11 i
t
~ U (t3 - t g)j =
Cfa f NPjItg (21a) where tg= 2k g R
(
Also, ' Eq. (19b) becomes, using Eq. (20) .
(tg - t c)j
- 2k f aNPj IIn t (21c)
Using the values of Table 9-1, Eq. (21c) reduces to (tg -t)j c = 1.248 x 103 Crf P3 *F (21d)
(tg -t)j c = 6.936 x 102 CrfnN 3 *C A Substituting Eqs. (21a) and (21d) into Eq (18a), and using the result in Eq. (17), it follows that ttoj * -( 9 * -+ t g ) C faf NPj + 140 *C (22)
Eq. (22) is the equation used to calibrate an instrumented 12 wt% fuel element to provide a measured fuel temperature, tto, during steady state operation.
- 2. Pulsing Characteristics of the PSBR The temperature distribution in a TRIGA fuel element during a pulse has.the same distribution as expressed in Eq. (4) up to 89% of the fuel radius.(15) It has been found that adiabatic conditions hold up to 0.07 sec. during which time the maximum fuel temperature is reached.(15)
Using the values of Table 9-1, it is found(4) that the maximum fuel temperature during a pulse is 1.6 times that measured by the thermocouple. Thus, during the pulse, the shape of the temperature
/ distribution in a fuel element remains constant, but the magnitude v
IX-12 i quickly rises. What we are concerned with here is the maximum fuel temperatures reached during the pulse. To prevent confusion, we use the term highest maximum fuel temperature to refer to the highest temperatures reached at any point within the fuel element during the pulse. The highest maximum fuel temperature is thus the maximum fuel temperature reached during a pulse and must remain below 1150*C.
However, the highest measured fuel temperature is 1/1.6 or 0.625 times the highest maximum fuel temperature which corresponds to a measured fuel temperature of 720*C. Thus, setting the Limiting Safety System Setting (LSSS) at 700'C, corresponds to a maximum fuel temperature of 1120*C. Thus, the LSSS as defined in the Technical Specifications provides a safety margin from reaching or exceeding the 1150*C.
A semiempirical equation, Eq. (29), is developed using the definition of the prompt temperature coefficient. The large negative prompt temperature coefficient, a, provides the TRIGA core with its pulsing capability. When excess kerr, 6kex = kerr - 1, is inserted into the reactor, the reactor will go on a prompt period, provided 6k p = 6k ex -8 (23) is positive, i.e., 6k p >0. 8 is the effective delayed neutron fraction (0.007).
Let:
It p = maximum fuel temperature rise averaged over the total core fuel volume for a pulse.
6tpoj = maximum fuel temperature rise averaged over the radius of the jth fuel element at its midplane, a = prompt temperature coefficient of reactivity of the TRIGA Core.
The prompt temperature coefficient is defined as O
IX-13 TD
(,/
a- - 6kn it pp or 5t pp =-0 P (24b) where It pp = average maximum rise in core fuel temperature due to the prompt excess reactivity insertion.
Eq. (24b) does not include the average core temperature rise due to a pulse ' insertion of $1 excess reactivity, 6t pi. Thus, the total average fuel temperature rise during a pulse, Itp , is 6t p =
6tpp + 6tp3 -(25a)
For'the jth fuel element, its corresponding temperature increase in U the midplane is 1
It poj =
f aNPj6tp (25b) or 5tpoj =
f aNPj (- P) +
f aNPj6tp1 (25c)
Initially the core fuel temperature is that of the pool water, T o, and during the pulse an adiabatic increase in temperature, 6tpoj(r), is assumed. Hence, 6tpoj(r) =
It ojp ( Ao + Bor2) (26a)
Eq. (26a) expresses the maximum temperature rise above room temperature for the jth fuel element as a function of fuel radius. For convenience 6tpoj(r) =
It pojf(r) (26b)
IX-14 where f(r) -
Ao + Bo r2 (26c)
The temperature in the midplane of the fuel element at any r position, tpoj(r), can be expressed as:
t poj(r) =
Itpojf(r) + To (27a)
Let t poj =
maximum pulse temperature measured by the thermocouple in the jth fuel element Then tpoj =
Itpoj f(rte) + To (27b)
Using the values of Table 9-1 t poj =
0.7566 itpoj + T o (27c)
Substituting Eq. (25c) into Eq. (27c) t poj =
0.7566 [faNPj ( ) + f aNPj 5tp j] + To (28)
Equation (28) is used as the basis for developing the semiempirical equation, Eq. (29), to fit the actual pulse data as a function of NPj ,
i.e.,
tpoj =
+ f aNPj 5t po + To (29) where K111 and Epo are empirical constants to be determined experimentally. The experimental data may also be represented by:
IX-15 1
tpoj - To -
Mj 6kp + bj (30)
'd where-K 1hf iNPj MJ .
a and bj =
f aNPjIt po (31b)
NPj and It po are determined by fitting the experimental data to Eq.
(30) where NPj, fa , and a are known In Eq. (30), MJ 6kp represents the temperature rise during a pulse due to the prompt 6kp excess reactivity insertion and bj represents the corresponding temperature rise due to the $1 excess reactivity insertion.
3 TRIGA Experiment to Measure Fuel Temperatures Using the analyses of the previous sections, a calibration was made to determine fuel temperatures for steady state and pulse modes of operation. This section describes the calibration techniques.
A series of fuel temperature measurements was made using the 12 wt%
instrumented fuel elements in core configuration loading 36 as shown in Fig. 9-1. One instrumented fuel element, I-13, had been in the core since September 1977 and the other, I-14, had never been used. The first series of measurements was taken with I-13 in the G-8 and I-14 in the G-10 core positions. The core position of a fuel element is identified in Figure 9-1 by a letter for the vertical axis position and a number for the horizontal axis position. The instrumented fuel elements have been numbered sequentially with an I prefix. After rotating both fuel elements at 0.5 MW steady state operation to obtain maximum temperature readings, the reactor power was increased in steps to 0.7MW, 0.9 MW, and 1 MW. The actual values of the power are 0.98 of that read on the recorder because the readout on the linear recorder is adjusted to read 0.4 MW when the actual power, as determined by a
(" thermal power calibration, was approximately 390 KW.
E.k O
- _C D_E G_H_I _ J.K_
F .
4I 6 Y 3 5 g 4
n i
d
- 28. a o
- L
'1 1 :
n l.
. o v i O t I 0 1 a
. r
- u g
9 9 i f
l
. .k n O
i '
~
e 8 o 8 C 7
'.7 r e
o
- C 6 6 R B
- S P
5 1
4 4 -
9 Y .
3 g i
Y F 2
e
!CDE F G_H_l _ J_K_
O
IX-17
- Af ter completing the series of steady state runs, the reactor was pulsed sequentially with 2, 2.25, 2.50, and 2.75 dollar pulses. Before each pulse, the reactor was made suboritical to allow the temperature to reach equilibrium.
The'above experiment was then repeated to determine reproducibility and measure the effect of the gap in I-14 created by the 4 pulses. The above measurements were again repeated with I-13 positioned in G-10 and
.I-14 positioned in G-8 to again study the reproducibility of the data and obtain another measurement on the temperature drop across the fuel cladding gap, 4t g.
The steady state and pulse measurements were again repeated, first with I-14 in F-10 and then with I-14 in H-11: I-13.was in position in G-8 for both sets of these measurements.
The data for all measurements are summarized in Table 9-2. Both L the chart recorder and the meter were used to measure the fuel temperature of I-13 ac shown in Table 9-2. The chart recorder is connected to the thermocouple at the midplane of the fuel element, whereas, the meter is connected to the thermocouple located 1" below.
E
.The banked control rods during a pulse causes the-position of the highest power density in the fuel element to be displaced slightly downward from the midplane of the fuel. This causes the meter readings to be approximately 24*C higher than those read on the chart recorder.
l'
'4. Evaluation of It gfor Fuel Element I-14 The unused TRIGA 12 wtl instrumented fuel element, I-14, has been placed in the core configuration of Fig. 9-1, Core Loading 36, and experiments performed to evaluate Eq. (22). It is assumed that before pulsing the instrumented fuel element I-14, it had a It gequal to 0 l as the fuel would be in contact with the cladding. When the fuel lI element is first pulsed, the cladding is stretched introducing a gap which increases the Itg . After a number of pulses Itg reaches a maximum value and does not increase with further pulsing.(4) The increase in Itg after pulsing I-14 several times is now determined by comparing the steady state temperatures for the same condition after j each set of pulses, p n L-.
r IX-18 Table 9-2 Fuel Temperature Measurement Data for Loading 36 To - 21 *C Fuel Core ttc(*C) tnn(*C) Recorder / Meter Element Position SS 1MW Pulse $2.00 Pulse $2.25 Pulse $2.50 Pulse $2.75 I-13 0-8 412 353/379 392/421 436/467 478/509 I-13 0-8 411 --- --- --- ---
I-13 G-8 411 343/381 387/421 431/461 478/511 I-13 G-8 411 350/381 389/419 435/466 478/511 I-14 G-8 445 389 427 468 517 I-14 G-8 466 395 434 482 518 I-13 G-10 381 323/333 359/371 399/412 430/453 I-13 G-10 382 311/332 357/373 400/416 439/453 I-14 G-10 372 339 375 415 456 I-14 G-10 41 8 --- --- --- ---
I-14 F-10 450 348 391 425 466 I-14 H-11 433 342 373 411 449 I-13 is a 12 wt1 fuel element burned to 2.2 Megawatt days I-14 is a fresh 12 wt% fuel element O
IX-19
(~N At 1 MW, I-14 measured tt o " 372*C in core Loading 36 and position G-10 before any pulsing began. Using Eq. (20 372 0.98 90.2 '
94 6 x 1 35 NPj + 140 it follows that NP - 372 - 140 - 1.84 J 128.8 x 0.98 After 4 pulses to t
- 418'C and thus 418 -
0.98 (95 39 + It g) (1.35)(1.84) + 140 p It g - 19'c These data and analysis show the initial (few pulses) increase in the temperature across the gaps this increase diminishes to zero with successive pulses (see Table 9-2, lines 9 and 10). Element I-14 was then moved from position G-10 to position C-8 in the B-ring where the NPj is different from that in G-10. Assuming Atg - 19'c and using tt e -
445'c as measured in its new position at 1 MW, it follows that 445 - 0.98 (114.4) 1 35 NPj + 140 Therefore, NPj - 2.02 After 8 pulses,the tga for element I-14 was measured again in position 0-8. This time tte was 466*C at 1 MW. Therefore, using Eq. (22),
(96.4 + It )(1.35)(2.02) 0.98+ 140
( 466 -
g
IX-20 and It g = 26.6*C It can be observed that af ter 8 pulses, Itg = 26.6'C. Past studies have shown that additional pulses do not alter the Atg significantly.
For I-14, it is assumed that af ter many more pulses, the Itg increase will be 1 *C, hence It g =
27.6'c and Eq. (22) becomes for Loading 36 at Q = 1 MW (t tc)J =
163 Cf NPj + 140 (32)
Eq. (32) can now be used to determine the NPj for I-14 anywhere in Core Loading 36 at 1 MW power. To generalize Eq. (32) for any core configuration, it is only necessary to account for N o. If this is done, Eq. (32) becomes (tto)J "
No r j+ 0 (33)
The steady state data of Table 9-2 has been evaluated using Eq.
(33) and the results are given in Table 9-3 The tte for the G-10 position was not measured af ter all pulsing had ceased and, therefore, is not listed in Table 9-3 Eq. (33) is used to evaluate the measured fuel temperature during steady state operation. During steady state operation, the measured fuel temperature is close to the maximum fuel temperature. In this case, the LSSS of 700*C is extremely conservative because under steady state conditions, the maximum fuel temperature is close to 700'C and thus, is well b(low the safety limit-fuel temperature of 1150'C. For loading 36, an upper limit for the measured maximum fuel temperature can be determined by setting NP = 2.2. Extensive calculation have been performed (1,2,5,7) to study the maximum power distribution produced by L
IX-21 Table 9-3 Evaluation of NPj's From I-14 Data 2
4 NPj NPj(Ave) l tto Steady State Pulse Core Position ('C) Eq.(32) Eq.(34) 0-8 467 2.01 2.07
- F-10 450 1.90 1.85 H-11 433 1.80 1.78
- G-10 1.84 1.80 4
I 1
4 Table 9-4 t
Pulse Parameter Characteristics of Fuel Element I-14 Mj Mj/NPj 1 X104 X104 bj bj/NPj ;
- Core Position Ng 'C/6k/k 'C/6k/k 'C .C * ;
G-10 1.84 2.23 1.21 162 88 0-8 2.01 2.44 1.21 197 98 G-8 2.01 2 34 1.17 210 104 i
F-10 1. 90 2.25 1.18 170 89
- H-11 1.80 2.04 1.13 178 99 4 ,
c t
l
.)
IX-22 different core configurations with fresh 12 wt% fuel in the B-ring and the other core configurations containing a mixture of both 12 wt$ fuel and 8.5 wt% fuel. In some configurations, one or two fuel elements were removed from a configuration. In no case was it possible to obtain a power distribution, wherein, the NP was greater than 2.2 provided the core configurations were within the limitations set forth in Section 3.4, CORE CONFIGURATION LIMITATION, of the Technical Specifications.
' The maximum power produced by a fuel element operating at 1 MW in loading 36 is thus 23.2 kw. Substituting NP = 2.2 into Eq. (32), tt e -
499'C when the fuel element produces 23.2 kw. Hence, from a practical point of view, future maximum steady state measured fuel temperature will be below the 700*C LSSS.
- 5. Evaluation of the Pulse Data for Fuel Element I-14 Each series of pulse data using I-14 is fitted to a straight line to determine Mj and bj of Eq. (30). Table 9-4 summarizes the results.
The constants Kjg and It po are determined to be 1.22 and 71 respectively using Eq. (31). Substituting these values and those from Table 1 into Eq. (29) the result is t poj = 1.177 x 104 NPj okp + 95.8 NPj + To (34)
Eq. (34) is now used to evaluate NPj for the various core positions.
The results are shown in Table 9-5 and the average NPj's in Table 9-3.
The highest measured fuel temperatures in fuel element I-14 are compared to that using Eq. (34) in Fig. 9-2. It can be observed that the measured temperatures are in good agreement with Eq. (34).
The Penn State in-core fuel management codes were employed to determine the power distribution and NPj's for Core Loading 36. In a recent Ph.D. thesis,(16) the group constants of the individual fuel elements were evaluated as a function of their burnup using the SCRAM code. The SCRAM code provides a simple but reasonably accurate method of depleting the PSBR as has occurred since December 1965. These constants were input into the EXTERMINATOR-2 code to obtain the NPj's for core Loading 36. The NPj's for G-9 and H-10 with new 12 wt% fuel
f IX-23 O l l
i NP = 2.07 500 I E _
I E NP = 1.85
,o m,, 9 NP = 1.80
- NP = 1.78 l
3 a l*CM
.e O
400 a
l l
01.N
-e l-E PULSE DATA [
M NP = 1.78 (M-11)
O NP = 1.80 (G-10)
& NP = 1.85 (F-10) 300 5 NP = 2.07 (G-8) f l
I Nb :
?
2.00 2.25 2.50 2.75 O Fig. 9-2.
STEP REACTIVITY INSERTION IN DOLLARS COMPARING HIGHEST MEASURED FUFL TEMPERATURES DURING A PULSE WITH IQ(34) POR FUEL ELEMENT I-14 l i
IX-24 Table 9-5 Table of NPj Determined for I-14 Using Pulse Data in Eq (34) 6k,x 6k p CORE POSITIONS Dollars Dollars G-10 G-8 G-8 F-10 H-11 2.00 1.00 1.79 2,06 2.10 1.84 1.80 2.25 1.25 1.79 2.04 2.08 1.86 1.78 2.50 1.50 1.79 2.04 2.10 1.85 1.78 2.75 1.75 1.81 2.07 2.07 1.86 1.78 Ave NPj 1.80 2.05 2.09 1.85 1.78 O
O
IX-25
_p varied between 2.03 and 2.10. This is to be compared with the measured V values of 2.01 steady state and 2,07 by pulse. In general, the steady state and pulse NP's agree with + 35.
It is now possible to eliminate the NPj from Eq. (32) and Eq. (34) to give for I-14 tpoj - To =
Cr~1(72.2 6kp + 0.588)(ttej - 140)*C (35)
Eq. (35) is an equation developed for fuel element I-14. It can be used anywhere in the core to predetermine the highest measured fuel temperature as a function of pulse prompt excess reactivity insertion (6kp ). It will require using the I-14 measured temperature when operating at 1 MW with I-14 in the same core position. Using a maximum value of NP=2.2 and corresponding t te = 499*C, see Eq. (32), the maximum value for t poj is 667*C for a pulse reactivity insertion of $3.40 assuming To = 21 *C and Cr = 1.0. If the pulse reactivity insertion is increased to $3.70, the resulting t poj is 720*C. Using the same condition, this corresponds to the highest maximum fuel temperature of
( 1150*C.
Thus, in the future, Eq. (28) can be used to evaluate and predict t poj . This requires placing I-14 in the hottest spot in the core and running at 1 MW to evaluate tge. Then starting with a $2 pulse, verify Eq. (28) and predict tp oj for the high values of 6kp . The tp oj related to a $2 pulse will be more than 100*C below tp oj for a $2.75 pulse and even much lower than that for a $3.40 pulse. Hence, these initial pulses will produce maximum measured fuel temperatures well below 700*C and allow determining the maximum fuel temperature attainable at the maximum allowable reactivity insertion pulses. A highest fuel temperature measured by the thermocouple in a 12 wt5 fuel element as 700*C corresponds to a highest maximum fuel temperature of 1120*C. This is below the maximum allowed 1150*C.
i n.
y
IX-26
- 6. Evaluation of the Fuel Element I-13 Temperature Data (Pulse and Steady State)
Table 9-2 shows the temperature data taken for I-13 and I-14. As expected, a review of these data shows that the measured temperature of I-13 during 1 MW steady state operation, tte, is significantly lower than the tto of I-14 for the same core positions. On the other hand, the I-13 measured pulse temperature data is not significantly lower than that of I-14 for the corresponding conditions. This is because the A o and Bo constants of Eq. (14) are not the same for I-13 and I-14. The depletion of the outer rim of fuel in I-13 during burnup, in addition to the burnup of 235 0 in all or the fuel element, lowers the selfshielding of thermal neutrons. As a consequence, the q(r) distribution for I-13 is much flatter than that or I-14. Using Eq. (4) for I-13, and setting Ao =
0.9267 Bo = 40 yields the results of the I-13 data as shown in Table 9-6. The equivalent of Eq. (33) and Eq. (34) for I-13 are Eq. (36) and Eq. (37) respectively.
(ttc)J
- No rN j + 0 (36) t poj =
1.475 x 104 NPj6kp + 80 NPj + To (37)
It can be observed that the agreement between the pulse data and steady state data for determining NPj is not as good as that for I-14.
This is due to the approximations made in deriving Eqs. (33) and (34),
namely, (1) The Ao + Bor2 shape of q(r) approximates the excess burnup of U-235 at the perimeter of the U-ZrH fuel in I-13 (2) The It gfor I-13 is probably different from that or I-14.
IX-27 ,
)
i l
Table 9-6 Evaluation of NPj's From I-13 Data {
I-13 (1.2 MWD Depleted) l l
I NPj NPj(Ave) tga Steady State Pulse ;
Core Position (*C) Eq.(36) Eo.(37) !
l G-8 411 1.56 1.62 !
G-10 382 1 39 1.54 !
I i
For comparison see I-14 data in Table 9-3 l
s f
9 ;
r i
f f
f I
I i
c.
I i
I t
T j.
IO '
f e l i.; ,
i e t
r IX-28 However, temperatures measured by I-13 are consistent for the purposes of monitoring the core fuel temperatures.
- 7. Conclusion (Temperature Analysis)
A major conclusion of this section is that an unused 12 wt1 U-ZrH TRICA fuel element can be calibrated and used to monitor the maximum fuel temperatures in the core. Once calibrated, the fuel element will only be used to measure maximum fuel temperatures in new core configurations. The maximum NPj that any PSBR core can obtain in configurations permitted by the Technical Specifications is approximately 2.2. Thus, the measured fuel temperature of a new TRIGA fuel element will not exceed, under any conditions, approximately 700*C.
For steady state operations a measured fuel temperature of 700*C results in a maximum fuel temperature well below 1150'C. Under these conditions, the measured fuel temperature is close to the maximum fuel temperature. For pulse operation, a measured 700*C fuel temperature corresponds to a maximum fuel temperature of 1120*C which is below 1150'C. From a safety point of view, the safety limit-fuel temperature is not exceeded.
Once a fuel element has been depleted, its maximum temperature decreases. The fuel temperatures measured during steady state operation with a depleted fuel element are related to the fuel temperature of a new fuel element by a simple ratio. Hence, this ratio can be used to assess the maximum fuel temperature during steady state in a new fuel element. The maximum fuel temperatures measured with a depleted fuel element during a pulse are close to that in a new fuel element.
This is because the preferential depletion of the periphery of the fuel element causes the power distribution and hence, the temperature distribution during a pulse, to be flatter than that of a new fuel element. Thus, the measured fuel temperature in a depleted fuel element corresponds to a lower maximum fuel temperature. It is also closer to the average fuel temperature. The core average fuel temperature rise for a given okp insertion is the same for all cores. The lower NP for a e
IX-29 t
):
depleted fuel element compensates for its being closer to the average core fuel temperature.
C. Evaluation of the Limiting Safety System Setting (LSSS)
The limiting safety system setting is a measured fuel temperature of 700'C as defined in the Technical Specifications.
If the core power were at 1.15 MW (15% over power) steady state, the measured fuel temperature using Eq. 32 is 552*C fer the hottest fuel element in the core, i.e., NP-2.2 and P=26.7 KW per fuel element for loading 36.
The measured 552*C fuel temperature is close to the maximum fuel temperature (within approximately 10%) dut to the radial temperature distribution. A sudden insertion of reactivity, clots to but less than $1, into the core will initially increase the reactor power exponentially at a period faster than one second (for this to occur, the period scram must fail as it is set for three seconds). Using a neg1tive temperature coef ficient of 1 x 10'4 7' 's ok/'C', the increase in average core fuel temperature is less than,
.007 ok/k =
1 x 10"9 6k/R 'C 70*C And for an NP = 2.2, the maximum fuel temperature increase is 154*C (2.2 x 70*C = 154*C). Adding this increased fuel temperature in the hottest fuel element to the 552*C steady state temperature results in 706*C, much less than the maximum limit of 1150*C. For this to occur at power levels above the power level scram set point will require that both power level scrams fail. The temperature scram will be initiated when the measured temperature exceeds its set point. The equilibrium temperature of 706*C will be achieved at least within two to three periods (seconds) af ter reactivity insertion. A control rod drop time less than one second assures an early decrease in reactivity and fuel temperature. At this point, the control rods moving into the core will begin to decrease the reactor power in less
- The temperature coefficient during a fast period is slightly less than the 7s prompt temperature coefficient.
( )
O
IX-30 than a second af ter the scram. Control rods are checked semiannually to assure their rod drop time is less than one second. The kinetics of the reactor cause the reactor power to decrease as soon as the control rods move a few inches into the core. Thus, the maximum fuel temperature will remain well below 1150'C sinco the measured fuel temperature is close to the maximum fuel temperature for these quasi-static conditions.
The maximum allowed pulse reactivity of $3.40 is established to prevent the measured fuel temperature from exceeding the LSSS. When the core produces a $3.40 pulse, the maximum measured temperature is, using Eq. (34) and NP = 2.2 for loading 36, 667'C. This corresponds to a maximum fuel temperature of 1067'C. The temperature scram will not lower the maximum fuel temperature attained during a pulse once the pulse is initiated:
however, it does protect the core from high temperatures during steady state operation.
D. Loss of Coolant Aeoident The PSBR pool contains 71,000 gallons of water. For a loss of coolant accident to occur, a break in the pool wall or break in a connecting pipe must occur below the bottom of the core. A series of alarms will occur as the water level drops more than 26 cm. Just below 26 cm, alarms will notify the reactor operator in the control room and the University Police Services.
If the reactor is operating at 1.0 MW the area radiation alarms will sound and initiate a scram of the control rods and activate the evacuation alarm.
It will also provide information to operating personnel that high levels of radiation exist in the reactor bay before the water drops nnother 100 cm.
In case of a leak, there exists a moveable gate that can be used to isolate either side of the pool within I hour af ter the leak is noticed. PSBR Emergonoy procedures call for moving the reactor to the non-leaking side of the pool and isolating that side of the pool with the gate to prevent the water level from dropping below the reactor core.
If a leak occurs, it will take time for the pool level to drop below the bottom of the fuel. With the reactor cooled by water for at least two minutes after the scram, which is produced by radiation monitors or operator, the maximum fuel temperature will drop more than 350*C (17) and
t IX-31 p) three minutes af ter the scram the maximum fuel temperature is within 20*C of V the water temperature.(17)
The largest conceived LOCA is one wherein a break in the 6" pipe connected to the bottom of the pool occurs. For this LOCA, it will take more than 1360 seconds (22.6 min) before the water falls below the bottom of the reactor core. Therefore, the minimum time before air convective cooling occurs is about 23 minutes arter a LOCA occurs.
This 23 minutes is much longer than the 3 minutes it takes the fuel to reach a temperature within 20*C of the water temperature.
As soon as the water falls below the reactor core, the fuel temperature will begin to rise, because the natural air covection cooling is less effective than water. The rate of rise of. the fuel temperature will depend on the previous operating history of the reactor, and the effectiveness of natural air convection to cool the fuel elements. Thus, the total time, t, it takes, starting when the LOCA occurs until the fuel reaches its maximum temperature, is the sum of two times tj and2 t . The time it takes for the water to fall below the bottom of the core once LOCA occurs is it . The time it takes once air cooling begins until the fuel temperature reaches its (j maximum temperature is t 2+
General Atomic conducted a set of LOCA experiments for TRIGA reactors.(10) In these experiments dummy TRIGA fuel elements were electrically heated in a grid to determine the rate of temperature increase of TRIGA fuel elements when cooled by natural air convection. The dummy fuel elements were wound with resistance wire to simulate a cosine distribution similar to that produced in the core. The standard TRIGA grid plate assembly pitch for a circular (non-hexagonal) core was used with a seven element assembly to mock-up the central portion of a standard core.
The PSBR does not have a central fuel element in the core to block the air flow in the hottest part of the core. In addition, the hexagonal pitch of the P3BR is less likely to produce hot spots on the cladding. When the core is uncovered, the central part of the PSBR core will allow more ef ficient cooling of the fuel elements in the B-ring increasing the safety factor associated with these calculations.
The total time it takes for the fuel to reach its maximum temperature during a LOCA, t is equal to tj plus t ,2The time tt may be computed
( I
- _)
9
_ . - - . _ _ _ - - _ _ _ _ _ . . _ _ _ . _ _ _ _ _ _ . . - . . _ _ - - _ _ . . . . . - _ _ . . . _ - _ _ - - _ . _ _ _ - - - - _ _ _ _ - - _ _ ._._____-______.______-__a
IX-32 assuming the 6" drain pipe at the bottom of the pool ruptures. In this case, t3 -
1360 sec.
To calculate the time t 2 , it is necessary to review GA's results as summarized in Figs. 9-3 and 9-4 Fig. 9-3 shows that with a constant cosine shape power input of 267 watts, it takes approximately t -45 2 minutes before the maximum fuel temperature reaches 360*C or 60% of its final or equilibrium temperature of 600'C. The maximum. fuel temperature attainable, i.e., equilibrium maximum fuel temperature as a function of the source power in watts, is given in Fig. 9-4. The thermal time constant of the fuel after a LOCA is approximately the same for all values of decay power. Thus, it will take 45 minutes once air cooling beings before the fuel temperature reaches 60% of its equilibrium fuel temperature. Assuming a steady input of decay power, 45 minutes is, thus, a conservative value to use for t 2-Before the maximum fuel temperature reached during a LOCA is determined, the core operation history and maximum fuel element powers must be established. As shown in Fig. 9-4, the maximum fuel temperature reached during a LOCA is directly related to the decay power and the decay power is a function of the pre-LOCA reactor operating history.
The following assumptions are conservative and are used to determine the decay power. The PSBR is licensed for a maximum steady state power of 1 MW and normally operates on a 40 hr/wk schedule. Even when the reactor operates on a two shif t schedule for reactor operator instruction or for laboratory experiments, the average power per 40 hr week is much less than 1 MW. The PSBR operated for a total of 11,766 MW hours during the 15 year period 1967-81 or an average of 15 MW hr/wk. Since that time, the average power per week has been reduced. Assuming the PSBR operates for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> at 1 MW during the week establishes an upper limit for its operation. However, to cover any future increase in operational activities, it will *>e assumed that the reactor is operated for 70 MW hours in one week.
It is thus assumed that the PSBR had been operating at 1 MW f or 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> continously when a LOCA occurs. With this assumption, the fission product decay power can be determined in the following manner.
El-Wakil(14) gives the following equation for decay power:
O
O -
0 0
~ . , , , , , , , , . . .
( + 6-1/2 IN.
\
500 _ _
U
- - 400 _
EXHAUST AIR TEMPERATURE _
S!
B
=c
$300 _ _
5 w
200 - .
100 -
0 ' ' ' 8 i i i i i :
0 20 40 60 80 100 120 140 160 180 200 220 240 ELAPSED TIME (MINUTES) l s
Figure 9-3 The Time Dependence of Air-cooled Fuel Body and Exhaust-air Temperatures for Center Y Element with 267 W Input U l
l
1 l
IX-34 O
1000 , , , , , ,
s
,s*
m s
5 _ -
M M
- b
~
M
/
/
M 100 -
G -
,s
- ,/
5 . - _
B
.~
/
10 ' ' ' ' ' '
0 100 200 300 400 500 600 700 CLAD TEMPERATURE (*C)
Fig. 9-4. Summary of Equilibrium Data for Loss-of-coolant Simulation Showing the Fuel-element Clad Temperature versus Power Input to the Element for All Seven Dummy Elements Hested with the Same Power Input
1 l
IX-35 l
Ps/Po = [0.1(0 3+10)-0.2 - 0.087(0 3+2x107 )-0.23 (38) g l 1
-[0.1 (e s+0o+10)-0.2 - o,og7(Os+0o+2x107 )-0.2]
where J
P3 is the power after shutdown produced by the fission product decay Po is the steady state power before LOCA, i.e. 1 MW 03 is the time after LOCA, i.e., 1360 sec Oo is the time of operation before LOCA, i.e., 70 hrs x 3600 sec/hr - 2.52 x 105 3ee Using these values
% P3/Po -
1.53 x 10-2 or for Core Loading 36 and an NP - 2.2 which corresponds to a Po - 23.2 KW, the maximum power due to fission product decay, Po maxe 18 Po 0
max -
P3 /Po x 2.2 x ,
=
1.53 x 10-2 x 23.2 KW/ fuel element 355 watts This decay power is higher than any achieved in the PSBR as the average power history is about 15 mwhr/wk, instead of the 70 mwhr/wk used in this calculation. Assuming the core is uncovered and reaches equilibrium fuel temperature 1360 sec (a 23 min.) af ter a LOCA, the hottest fuel element in the core will have less than 355 watts of power which will continue to decay.
Figure 9-4 shows the equilibrium fuel cladding temperature as a function of fuel element power input. The data of Figure 9-4 have been fit with a straight line equation, i.e.,
a
. - , - - , - . . _ ,, - , ,,, ..m.--__,.-,_,y,< - . . . . , ,
IX-36 Pp =
0.5498t - 62.89 (39) where Pp is the decay power (watts) producing temperature in 'C.
When Pp = 355 watts then t = 761'C, and This shows, that once air cooling begins and the average decay power remains constant at 355 watts for approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (see Fig. 9-4), the maximum fuel temperature is 761 *C. The decay power does not remain constant but continues to reduce exponentially. Therefore, for the maximum conceived LOCA for the PSBR, the maximum fuel temperature remains well below the 900*C limit throughout the LOCA.
It should also be stated that the experimental data of Fig. 9-3 shows that because of the long time it takes to heat up a TRIGA fuel element, i.e.
l t 2, once air cooling begins, the time t1 is less critical. The time t1 is the time it takes for the water to fall below the reactor core once LOCA begins.
The LOCA for a TRIGA core may also be analyzed analytically instead of by the results of experiments used above. General Atomic used one of their own two dimensional transient-heat transport computer codes to calculate the TRIGA fuel element temperature af ter the loss of pool water.(19) It was assumed that the reactor was operating for an infinite period of tim .
Their results are plotted in Figure 9-5 for several cooling or delay times showing maximum fuel temperatures in the TRIGA fuel element as a function of its operating power. It can be observed that a fuel element having approximately 23 2 KW before the LOCA, will attain a 870'C maximum temperature, 1360 seconds after the LOCA.
The PSBR can withstand a LOCA more severe than the maximum conceived LOCA. This is because the PSBR is limited to a maximum of 70 MW-hrs per week. The reduced power history, i.e., 70 MW-hrs per week instead of infinite time at 1 MW, causes a corresponding reduction in decay power. The reduction in decay power is a function of the cooling time after LOCA. The 22 KW per fuel element before LOCA (see Fig. 9-5); which would give a 900*C maximum fuel temperature for instantaneous loss of water from the pool, can be increased from 22 KW per fuel element to 1.1 x 22 KW per fuel element or
IX-37 l
2000 -
1800 -
/
1600 .
0
'1400 -
103 1200 .
1000 - 104 o
,3 800 -
{W 105 600 -
l 106 400 -
200 .
l 1
0 - - - . l - - . . .
0 5 10 15 20 25 30 35 40 45 OPERATING POWER OENSITY-KW/ ELEMENT Figure 9-5 Maximum Fuel Tenperature Versus Power Density after Loss of Coolant for Various Cooling Times Between Reactor
- Shutdown and Coolant Loss
IX-38 24.2 KW per fuel element. For normal PSBR operation, the maximum power density is 23 2 KW per fuel element which is less than the allowed 24.2 KW per fuel element.
The 900aC maximum fuel temperature is important when analyzing the TRIGA core for a LOCA. During a LOCA, the fuel element is uncovered producing cladding temperatures greater than 500*C. Under these conditions, if 900aC fuel temperature is reached or exceeded, the TRIGA cladding could be ruptured.(13) Below a fuel temperature of 900'C the cladding remains intact. The strength of the fuel element cladding is a function of its temperature. The yield strength of the stainless steel cladding under LOCA conditions (heated in air), is shown in Figurc 9-6. Also shown in Figure 9-6 is the cladding stress produced by any gas in the gap. This gas pressure consists of the hydrogen gas pressure plus the pressure of the volatile fission products plus the pressure of the trapped air. It can be observed that the cladding stress equals the cladding yield strength at approximately 900'C. This is different from the safety limit (1150aC) which is the case when the cladding is in water and cladding temperatures remain below 500*C.
As a consequence, the following results are applicable to the PSBR experiencing a LOCA:
(1) The maximum temperature that the PSBR TRIGA fuel element can have during a LOCA without damage to the cladding is 900'C. As long as the 900*C temperature limit is not exceeded, there will be no stress sufficient to rupture the cladding thereby allowing the escape of fission products.
(2) Reviewing the complete history of the TRIGA cores at Penn State, the PSBR has never been operated for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> at 1 MW during a week with the 12 wt% fuel. In addition, no core configuration had an NP greater than approximately 2.2 or an equivalent maximum power per fuel element of 23.2 KW. Assuming 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> of PSBR operation at 1 MW during a week with the 12 wt% fuel producing 23.2 KW (NP = 2.2), a LOCA will not cause the fuel element to heat up to 900aC under any condition.
O
IX-39 5
10 ,
ULTIMATE STRENGTH
~~~~,,'s
's N N
N YIELD STRENGTH 4
\
10 - N N
N
- \
G \
E , N
'a N E N
\
STRESS IMPOSED ON CLAD 10 2
l'0 , , , , , , , .
400 600 800 1000 1200 TEMPERATURE (*C)'
Figure 9-6 Strength and Applied Stress as a Function of Temperature, U-ZrH Fuel, Fuel and Clad at Same Temperature l.65 O
i-
IX-40 In conclusion, a LOCA with the PSBR will not result in damaged fuel and, thus, fission product containment within the fuel is assured.
E. Maximum Hypothetical Accident (MHA)
The maximum hypothetical accident (MHA) is one in which it is assumed that a fuel element cladding ruptures in an air cooled core releasing volatile fission products to the reactor bay. The MHA is defined as a postulated accident with potential consequences greater than those from any event that can be mechanistically postulated. The assumptions create conditions f ar more severe than is actually possible. Nevertheless, the accident exhibits an insignificant environmental impact or effect on the health and safety of the public.
The potential hazards associated with the MHA are related to the escape of fission products from the ruptured cladding of a fuel element into the reactor bay and then from the bay to the environment. The fission product buildup in a fuel element is a function of the core power history, the position of the fuel element in the core, and its fuel content. It is assumed in these calculations that the fuel element is a 12 wt% U-ZrH fuel element operated in the B-ring. The core is assumed to operate 70 hr/wk at 1 MW throughout its life. A review of the operating history of the 12 wt%
U-ZrH fuel element I-13 shows that it began operating at 1 MW with a maximum NP 5 2.1 (fuel temperature ' 460*C) and af ter operating from October 10, 1977 to the present, its NP has dropped to a little less than 1.5 (fuel temperature at 1 MW now reads - 380*C). Since all of the volatile fission products reach their maximum af ter a few weeks of operation, a value of NP =
2, i.e., 21.1 KW/ fuel element, is used to compute the fission product activity, Arp, in a fuel element. It is assumed that the rupture occurs when the reactor is just completing 70 hrs of 1 MW operation. The saturated activity of the core, R, for one fission product (fp) nuclide is R =
1 MW x Yo x curies (40a) where
fission product cummulative yield Yo 1 MW
3.1 x 1016 fission /see l-
IX-41 1 Ci =
3 7 x 1010 distingration/seo v
A continuous operation of 70 hrs per week produces fission product activity in the core which is an accumulation of fp activity from the previous weeks operation provided the half life is greater than approximately a day. The contribution to the fp activity at the time of the rupture is as follows:
R(1-e I) Ci is produced during the week the rupture occurs,
~
R(1-e 1) e~ 1+ 2) Ci is produced during the week before the rupture occurs, and R(1-e- 1) e~ " 1+ 2 Ci is produced during the nth week before the rupture occurs.
The total activity of this fp in the core at the time of the rupture is the sum of the above activities. It can easily be shown that the core activity is I ~ ' ATj Co- R -
+T)2 (40b) 1-e
( \ and NP 1-e -T1 Arp = R
_ -A(Tj+T 2)
- where T1
- 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> T2
98 hours0.00113 days <br />0.0272 hours <br />1.62037e-4 weeks <br />3.7289e-5 months <br /> Tj + T2
number of hours in one week = 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> The PSBR presently operates on a 40 hr/wk schedule. The actual operation produces much less than 40 MW hours (MWhr) per week. During the 18 year period, 1967-1984, the PSBR operated for 12,324 MWhr or an average of 685 MWhr/yr (13 2 MWhr/wk). The 70 MWhr/wk, therefore, establishes a.
large safety factor for this calculation.
Table 9-7 lists the activity of each of the important gaseous fission products. The fission product yields were taken from the Katcoff, et.al., report.(20) These yields include direct production plus precursor
IX-42 Table 9-7 Fission Product Data l l l l Activity (C1) l l l l Fission l l Co l (NP=2.0)l Bay (C b ) l l Isotope l Yield (%) l Half-life l Core l Element l (uC1/ml) l l Kr-83m l 0.544 l 1.86h l 4560 l 96 l 8.10x10-7 l l Kr-85m l 1.01 l 4.48h l 8460 l 180 l 1.50x10-6 l l Kr-85 l 0.293 l 10.5y l 1020 l 22 l 1.82x10-7 l l Kr-87 l 2.76 l 76 min l 23120 l 490 l 4.10x10-6 l l Kr-88 l 4.38 l 2.84h l 36700 l 770 l 6.51 x10-6 l l Kr-89 l 5.47 l 3.15 min l 45830 l 970 l 8.10x10-6 l l Kr-90 l 5.00 l 32.3sec l 41890 l 890 l 7.40x10-6 l l Xe-133m l 0.16 l 2.19d l 907 l 19 l 1.61x10-7 l l Xe-133 l 6.62 l 5.25d l 29370 l 620 l 5.20x10-6 l l Xe-135m l 1.83 l 15.3m l 15330 l 320 l 2.72x10-6 l l Xe-135 l 6.3 l 9.09h l 52530 l 1110 l 9.30x10-6 l l Xe-137 l 6.17 l 3.86m l 51700 l 1090 l 9.20x10-6 l l Xe-138 l 5.49 l 14.2m l 46000 l 970 l 8.20x10-6 l l Br-83 l 0.51 l 2.40h l 4270 l 90 l 7.60x10-7 l l Br-84 l 0.01 9 l 6.om l 160 l 3.5 l 2.83x10-8 l l Br-85 l 1.1 l 2.87m l 9220 l 195 l 1.64x10-6 l l I-131 l 3.1 l 8.041 d l 12740 l 270 l 2.26x10-6 l l I-132 l 4.38 l 2.29h l 36700 l 780 l 6.50x10-6 l l I-133 l 6.9 l 20.8h l 52390 l 1110 l 9.30x10-6 l l I-134 l 7.8 l 52.6 min l 65350 l 1380 l 1.16x10-5 l lI-135 l 6.1 1 6.585h l 51080l 1080 l 9.10x10 6l O
IX-43 i
( decay from fission products. Only a fraction of these fission products k escape from the fuel element into the reactor bay. The fraction that escapes from the fuel element is called the release fraction, fr .
The fission product release fraction, determined experimentally at General Atomic, is a function of the fuel temperature.(21) However, it is important to note that the release fraction in accident conditions is characteristic of the normal operating temperature and not the temperature during the accident. This is because the only fission products that escape are those accumulated in the gap between the fuel ard cladding during the operation of the reactor.
A review of the operating history of the instrumented fuel element I-13 in the PSBR B-ring shows that its maximum measured temperature during steady state operation at 1 MW was less than 460*C, After the first year of operation, its measured temperature at 1 MW dropped to approximately 400*C.
During this period, it's burnup was approximately 0.65 MWD. This temperature drop occurs because as burnup increases, the 235U core inventory decreases with a corresponding drop in NP and temperature. Operation af ter one year lowers the maximum measured fuel temperature in the B-ring to 400*C or less. Thus, it is conservative to use an average maximum measured temperature of 466*C, corresponding to NP-2 in Eq (32), to compute the release fraction. Studies have shown that as long as core configurations have only single vacancies within the central 5 in, radius of the core, the maximum possible NP is below 2.2 using the mixture of 12 and 8.5 wt$ U-ZrH TRIGA fuel elements.
Experiments demonstrated at General Atomic (21) generated release fraction data for the U-ZrH fuel under various conditions. Below 400*C the release fraction, f ,p is a constant,1.5 x 10-5, and above 400*C the following equation is used:
fr -
1.5 x 10-5 + 3.6 x 103 exp (-1 34 x 104/To) (41 )
where To is the fuel temperature in 'K
- l l
IX-44 For low temperature results, i.e. , below 400'C, the release fraction for a typical TRIGA fuel rod is constant, independent of operating history or details of operating temperatures. Averaging Eq. (41) over the volume of the fuel element gives a release fraction of 2.1 x 10-5 Applying the release fraction of 2.1 x 10-5 to a single element operating at 1 MW yields, with each fission product activity in Table 9-7 a release to the reactor bay of Co curies. The bay concentrations, C b , are based on a free air bay volume of 2500m3 The concentration in the unrestricted area (outside the reactor building) is obtained by dividing the activity release rate through the emergency exhaust system by the dilution rate. The release rate for the emergency exhaust system is equal to the flow rate (3100 cfm) times the bay concentration C b . The dilution rate is the wind velocity (1 m/s) times the cross-sectional area of the building (200 m2) or 2 x 108 mf./sec. Thus, at the instant the fuel element cladding ruptures, the maximum Concentration in the unrestricted area -
Cb x 7.30 x 10-3 In the reactor bay (restricted area), the concentration in air of airborne activity decreases with time because of radioactive decay and the removal of air by the emergency exhuast system.
Consider a model in which activity, released from the single fuel element, instantly mixes completely with air in the bay. The resulting uniform concentration is C b. We have, then, CT
= C b e ( d + Ar)t (42) where i
=
CT bay air concentration of activity at elapses time t after release from the single fuel element Ad
decay constant Ap
ventilation rate constant O
i
E IX-45 l'.46m3/se O P ventilation rate bay volume 2500m3
- 5.84 x 10'4 see 1 The dose is proportional to the integral of CT over the period T, labeled ICT*
Ch _g"(ld + Ap)T)
ICT" (43)
- The doses calculated for each of the. nuclides in Table E-7 of Ref. 22 are shown in Table 9-8. The exposure time is calculated for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in the reactor bay and for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the unrestricted area. The value of A'p, the ventilation' decay constant, is 5.84 x 10-4 sec 1 which is equivalent to a time of 1166 seconds or 19.3 minutes to remove 50% of the activity in the reactor bay. More than 955 is removed in the first hour.
Thus, essentially all activity is released to the unrestricted area within a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and doses in both the restricted and unrestrict'ed areas have essentially reached their maximum values. Release of the activity from
.the fuel element over an extended period of time would significantly reduce the calculated doses because of the additional decay of those nuclides with half-lives'of several hours, which contribute a large part of the total
. dose.
Dose in the reactor bay and in the unrestricted area depends on the type of radiation' emitted by each nuclide and the distribution of the nuclides in space. In these calculations, the nuclides are assumed to be a homogeneous distribution'in a semi-infinite cloud.
.For the noble gases, Kr and Xe, the beta dose to the skin and the gamma dose to the body is calculated using the nuclides and their conversion
. factors for doses given in Table B-1 of Ref. 21. Using appropriate units.
.the dose, D, in area is D(area) =
ICT x Conversion Factor ( [ ' ) (44)
Results of these calculations for both the reactor bay ~ and the unrestricted area are given in Table 9-8.
s
IX-46 Table 9-8 MHA Doses Adult Male (Breathing Male Adult)
Skin (mrem) Body (mrem) Thyroid (mrem)
Bay Unrestricted Bay Unrestricted Bay Unrestricted Isotope 1 hr 1 hr 24 hr 1 hr 1 hr 24 hr 1 hr 1 hr 24 hr Kr-83m 0 0 0 0 0 0 Kr-85m 0.18 0.001 0.001 0.079 0 0 0.079 0 0 Kr-85 0.12 0 0 0 0 0 0 0 0 Kr-87 2.5 0.01 9 0.020 0.96 0.007 0.007 0.96 0.007 0.007 Kr-88 4.8 0.035 0.039 4.2 0.030 0.033 4.2 0.030 0.033 Kr-89 1.6 0.012 0.012 1.0 0.007 0.007 1.0 0.007 0.007 Kr-90 0.24 0.002 0.002 0.17 0.001 0.001 0.17 0.001 0.001 Sub Total 9.4 0.069 0.074 6.4 0.045 0.048 6.4 0.045 0.048 Xe 133m 0.009 0 0 0.002 0 0 0.002 0 0 Xe 133 0.15 0.001 0.001 0.072 0 0 0.072 0.001 0.001 Xe 135m 0.24 0.002 0.002 0.20 0.001 0.001 0.20 0.001 0.001 Xe 135 1.6 0.011 0.013 0.77 0.006 0.006 0.77 0.006 0.006 Xe 137 1.1 0.008 0.008 0.11 0 0 0.11 0 0 Xe 138 2.4 0.017 0.017 1.6 0.012 0.012 1.6 0.012 0.012 Sub Total 5.5 0.039 0.041 2.8 0.019 0.019 2.8 0.019 0.019 Br 83
- 0.001 0 0 0.001 0 0 h Br 84
- O O O O O O Br 85 * *
- 0 0 0 0 0 0 Sub l Total .001 0 0 0.001 0 0 0.001 0 0 I 131 * *
- 2.9 0.021 0.024 1660 12. 14.
I 132 * *
- 0.42 0.003 0.003 42 0 30 0.33 I 133 * *
- 2.6 0.01 9 0.021 1220 8.9 10.
I 134 * *
- 0.34 0.003 0.003 17 0.12 0.13 I 135 * *
- 1.4 0.010 0.011 242 1.8 1.5 Sub Total 7.7 .056 .062 7.7 0.056 0.062 3181. 23 26.
GRAND TOTAL 23 0.16 0.18 17. 0.12 0.13 3190. 23 26.
Notes: 1.
- indicates no data for dose equivalent conversion in Regulatory Guide 1.109.
- 2. Dose equivalent values less than 0.001 mrem are listed as 0.
3 Thyroid dose for other than radioiodine is assumed to be equal to body dose.
- 4. Skin dose equivalent is sum of skin plus penetrating (body) dose equivalent.
O
_ _ _ _ _ . _ . _ _ - = _ _ __ _ .. _ _ _ _ - --- _ _ _ _ __ .. _ _ _ _
l IX-47 i
L For the Br and I nuclides, the internal dose is a function of the inhalation-rate. For radioiodine the thyroid dose is the controlling factor.
1 The internal dose is calculated as shown in equaition 45, D(arem) = ICT x Breathing Rate x Conversion Factor ( "" ) (45) :
where the breath'ing rate is 20 liters / minute (3.33 x 10-4 m3/sec)(23) and !
the conversion factor is taken from Table E-7 of Ref. 22. Table 9-8 also ;
shows'the doses for these nuclides. It can be observed that in the unrentricted area, the 1 he and 24 he dose to the thyroid is 23 3 mrem and 26.3 mrem,~respectively. The dose to the oody and the skin is less. Ref.
18_ did not include data for calculating the skin dose due to the bromine and t - iodine nuclides. However, a review of the radiation emitted by these ,
isotopes shows that they will produce skin doses less than that of the noble l
gaser. The total skin dose is less than the body dose and is, therefore,
- ' not significant when determining the hazards. The principle risk is the i dose absorbed by the thyroid which is not significant.~
l- In the bay, the dose to the thyroid is also controlling since it is ,
3195 mrem for a 1 hr exposure.
l - Since the emergency exhaust system removes the bay air through a.
- charcoal filter, the lodine reaching the unrestricted area should be less f than 10% of that'given in Table 9-8'. Hence, the dose to the thyroid in the unrestricted area should be less than 2.3 mrom and 2.6 mrem for the 1 he and
- j. -24 hr exposures, respectively.
The above calculations assumed an NP=2.0 (21.1 KW per fuel element),
- - which corresponds to a maximum measured fuel temperature of 460*C. The 21.2 KW 'per fuel element condition corresponds to a core configuration, which has all positions within the core occupied by fuel elements. If it is-assumed that the fuel element has an NP=2.2 (23.2 KW per fuel element and a
- i. maximum fuel temperature of 499'C); which corresponds to a core
- configuration with an empty position (no fuel) in the B-ring, all of the j
above exposures (Table 9-8) are increased by 255. Ten percent of this
- increase in the exposure is due to the higher fission product inventory. -
l which results from the higher power density, i.e., 23 2 KW per fuel element t
i versus 21'.1 KW per fuel element. _ The remaining increase is due to a larger release fraction, fr. of 2.4 x 10-5 as opposed to the 2.1 x 10~5 due to the
!(
t i
6
- . - .~- . m .- _ . . . _ . _.-._,. ,. _ . _ __.,___._.-.. _ ,. _ .. ..._ _ _ . _ _ _,m___c.,-__~.m_m.
l
(
IX-48 I i
i slightly higher ftal temperature. In this case, the 1 hr. and 24 hr. dose to the thyroid is 29 neem and 33 mrem, respectively.
The MHA creates conditions f ar more severe than are actually possible.
Therefore, the fact that the MHA produces no significant hazard outside the reactor bay in the unrestricted area shows that the PSBR's operation is safe to the public. By limiting time in the reactor bay during the MHA, personnel can also avoid any significant hazard from the MHA.
F. Reactivity Accident In this accident, it is assumed that the reactor is taken to a 1.15 MW power level with the transient rod inserted in the core and then the reactor is pulsed with a $3 reactivity insertion. This accident requires a breakdawn in the PSBR Standard Operating Procedures, the overpower scrams, and a failure of the interlocks.
When the core is operating at 1.15 MW, its total reactivity has been reduced by $4. Its measured fuel temperature in the B-ring using Eq. (32) and NP - 2.2 will be 552*C in a new 12 wt% fuel element. The maximum temperature will be slightly higher, but the fuel temperature near the cladding will be approximately half this temperature.
The maximum allowed core reactivity of $7 leaves $3 available for pulsing. Should a $3 pulse occur while the reactor is at 1.15 MW, the measured fuel temperature will rise from 552*C to 1146*C as calculated using Eq. (34) and setting NP = 2.2 and To = 21'C. In this case, when the core is pulsed from an initial power of 1.15 MW, the maximum fuel temperature is the measured fuel temperature. This is because the temperature rise during a pulse has a different radial shape than that attained during steady state operation. During a pulse, the increase in fuel temperature is a maximum near the edge of the fuel. Superimposing this shape of the fuel temperature on that attained at a steady state power of 1.15 MW produces, at the end of the pulse, a relatively flat radial temperature distribution at approximately 1150*C. However, since the negative temperature coefficient acts immediately as the transient rod moves upward, the final maximum temperature will be less than 1150*C.
O
I IX-49 m If the reactor power is accidently raised to 1.25 MW, the excess reactivity available for pulsing will be less than $3 As a consequence the maximum fuel temperature attainable remains less than 1150*C.
Hence the reactivity accident will not violate the safety limit.
G. Conclusion There are two limits which, if not exceeded, will prevent rupture of the cladding of a TRIGA fuel element. They are:
(1) Limit the fuel temperature to a maximum 1150*C when the cladding temperature remains below 500*C, i.e., when the fuel is covered with water.
(2) Limit the fuel temperature to a maximum 900*C when the cladding temperature is above 500'C, i.e., during a LOCA.
The Technical Specifications for the PSBR are established to prevent reaching these two limits. The 1150'c temperature limit is not reached as the fuel temperatures are limited during pulse mode operations. Eq. (34)
Q provides a direct method for determining the maximum fuel temperature based on the measured fuel temperature during a pulse. Using this equation and the maximum possible NP - 2.2 and core loading 36, the following limits are established:
(1) The maximum allowed reactivity insertion for the pulse mode is
$3.40 and corresponds to a maximum peak fuel temperature of 1067'C and a measured peak temeprature of 667'C.
(2). The maximum allowed worth of the pulse rod is $3.70. A sudden insertion of $3 70 excess reactivity results in a maximum peak fuel temperature of 1150*C and a measured peak fuel temperature of 720*C.
(3) The maximum allowed excess reactivity of the core is $7. Thus, when the core is operating at 1.15 MW steady state, only $3 of excess reactivity is available for pulsing, $4 of excess reactivity is needed to reach 1.15 MW. A pulse insertion of $3 excess reactivity starting at 1.15 MW limits the maximum fuel temperature to 1150*C.
\
l
9 IX-50 (4) Core configuration limitations are also established to prevent a fuel element from producing too much power relative to the other fuel elements. An NP = 2.2 cannot be attained by any allowed core configuration limiting the maximum power of a fuel element in core loading 36 to 23.2 KW. Setting NP = 2.2 establishes a fuel temperature upper limit attainable during a pulse.
Limits set for steady state operation prevent the maximum fuel temperatures from coming close to 1150aC. Limits imposed here prevent the fuel temperature during a LOCA from reaching 900*C.
If operated at 1 MW for 70 hrs. during each week, a single fuel element could operate above its max power level of 23.2 KW, as high as 24.2 KW, and still not have its fuel temperature exceed 900*C during any conceived LOCA. In addition, these same limitations on steady state operatien limit the release of fission products to the environment to very low values if the cladding ruptures.
The maximum hypothetic accident (MHA) analyzes the effect of a fuel element cladding rupture in air at the end of 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> of reactor operation at 1 MW. In addition, the reactor was assumed to have operated for 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> each week during the previous year.
Under these extreme conditions, the maximum exposure to a person in the unrestricted area is 23 3 mrem to the thyroid after 1 hr.
and 26.3 mrem to the thyroid after 24 hrs.
In conclusion, the analysis described in this section shows that under no possible accident conditions will the regulations in either 10CFR20 or 10CFR100 be violated. Thus, the PSBR can be operated without exposing the public to any significant radiation risk, i.e., the PSBR can continue to operate safely without adverse environmental impact and without undue risk to the public.
O
IX-51 p H. REFERENCES U
- 1. Naughton, W.F. , Cenko, M.J. , Levine, S.H. , and Witzig, W.F. , "TRIGA Core Management Model," Nucl. Technology. vol. 23, p. 256 (Sept.1974).
- 2. Naughton, W.F. , Cenko, M.J. , Levine, S.H. , and Witzig, W.F. , Increasing
- TRIGA Fuel Lifetime with 12 wt5 U TRIGA Fuel, TOC-5. TRIGA Owner's Confereace III (February 1974).
3 Haag, J. A., and Levine, S.H, " Thermal Analysis of The Pennsylvania State University Breazeale Nuclear Reactor," Nucl. Technology, vol. 19, p. 6 (July 1973).
- 4. Levir.e, S.H., Geisler, G.C., and Totenbier, R.E., Temperature Behavior of 12 wt$ U TRIGA Fuel TOC-5, TRIGA Owners' Conference III (February 1974).
- 5. Levine, S.H. , Totenbier, R.E. (Penn State Univ. ), and Ahmad T. Ali (PPAT Ismail - Malysia), Fourteen Years of Fuel Management of the Penn State TRIGA Breazeale Reactor (PSBR), ANS Trans, vol. 33 (November 1979).
- 6. Levine, S.H. and H. Ocampo, "The k.,-Meter Concept Verified via
,m Subcritical/Critcal TRIGA Experiments," Proceedings of the International Symposium on the Use and Development of Low and Medium Flux Research Reactors, MIT, Cambridge, MA (October 1983).
- 7. Kim, S.S. and S.H. Levine, " Verifying the Asymmetric Multip'l e Position Neutron Source ( AMPNS) Method Using the TRIGA Reactor," Ninth TRIGA User's Conference, Anaheim, CA (March 1984).
- 8. Levine, S.H., " Module 5 - In-core Fuel Management," Nuclear Fuel Cycle Educational Module Series, N.D. Eckhoff, gen.ed. , Kansas State University (July 1980).
- 9. Fowler, T.B., et.al., " EXTERMINATOR-II: A FORTRAN IV Code for Solving Multigroup Neutron Diffusi.on Equations in Two Dimensions," ORNL-4078, Oak Ridge National Laboratory (April 1967).
- 10. Huang, H.Y. and S.H. Levine "An Automated Multiple-Cycle PWR Fuel Management Code," ANS Trans. (November 1978).
-11. Cenko, M.J., " Comparison of PSBR Operation's History with the TRIGA Core Management Model," M.S. Thesis, The Pennsylvania State University (1972).
- 12. Barry, R.F. " LEOPARD - A Spectrum Dependent Non-Spatial Depletion Code for IBM-7094," WCAP-3269-26, Westinghouse Electric Corporation (September p .
1963).
(v)
IX-52 13 Simnad, M. T. , F. C. Foushie, and G. B. West, " Fuel Elements for Pulsed Reactors," GA Report E-117-393 (January 1975).
- 14. El-Wakil, M.M., " Nuclear Heat Transport," ANS (May 1978).
- 15. Goodwin, W.A., "The Measurement of Radial Power Distribution in a TRIGA Fuel Element During Reactor Excursion," Ph.D. Thesis, University of Illionis (1967).
- 16. Kim, S. S., " Development of an Asymmetric Multiple Position Neutron Source
( AMPNS) Method for Monitoring the Criticality of the Degraded Reactor Core," Ph.D. Thesis, The Pennsylvania State University (1984).
- 17. PSBR Log Book 37, page 265 (November 21, 1984).
- 18. Shoptaugh, J. R. , Jr. , " Simulated Loss-of-Coolant Accident for TRIGA Reactors," GA-6596 ( August 1965).
- 19. West, G. B. , " Safety Analysis Report for the Torrey Pines TRIGA Mark III Reactor," GA-9064 (January 5,1970).
- 20. Katcoff and Seymour, Nucleonics, vol.18, p. 201 (November 1960).
- 21. Foushee and R. H. Peters " Summary of TRIGA Fuel Fission Product Release Experiments," GULF-EES-A10801 (September 1971).
- 22. Regulatory Guide 1.109, " Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance With 10CFR Part 50, Appendix I."
23 International Commission on Radiation Protection Report #23 O
-~
TECHNICAL SPECIFICATIONS
('- 'sb TABLE OF CONTENTS 1.0 DEFINITIONS .................................................................. 1 1.1 ABNO RM A L OCCU RR ENC E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.2 ALARA .................................................................. 1 13 AUTOMATIC MODE ......................................................... 1 1.4 C H AN N EL C ALI BR AT IO N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.5 C H AN N EL C H EC K . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.6 C H ANN EL T EST . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.7 CO LD C R I T I C AL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.8 C LOS E P AC K ED ARR AY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.9 COR E L ATTIC E POS ITIO N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.10 EXPERIMENT ............................................................. 2 1.11 EX PERIMENTAL FAC I LITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.12 FUEL ELEMENT .......................................................... 2 1.13 IN STRUMENTED EL EM ENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.14 LIMITING CONDITIONS FOR OPER ATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.15 ' LIMITING S AFETY SYSTEM SETTING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.16 M E ASU R ED V ALU E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.17 M E ASU R I N G C H ANN EL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.18 ~ MINIMUM SHUTDOWN MARG IN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.19 MOVABLE EXPERIMENT .................................................... 3
- 1. 2 0 - NORMALI Z ED PO WER ( N P ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
('s - .1.21 OfERABkE .............................................................. 3 1.22 OPERATIONAL CORE ...................................................... 3 1.23 PU LS E M OD E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.24 f. EAC TOR I NTER LOC K . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
- 1. 2 5 ~ R EACTOR SAF ETY SYST EMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.26 R EACTO R SECU R ED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.27. R EAC TOR SHU TDO WN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.28 R EPO R TA B LE OCC U R R ENC E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.29 SAFETY CHANNEL ........................................................ 5 1 30 SAFETY LIMITS ......................................................... 5 1.31. SECURED EXPERIMENT .................................................... 5 1.32 SECURED EXPERIMENT WITH MOVABLE P ARTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1.33 SHIM , R EGU LATING , S AFETY ROD S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1.34 SQU AR E W AVE M OD E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1 35 STANDBY'............................................................... 5 1.36 STEADY STATE MODE ..................................................... 5 1.37 TRANSIENT ROD ......................................................... 5 2.0 S AFETY LIMIT AND LIMITING S AFETY SYSTEM SETTING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 2.1 ' SAFETY LIMIT-FUEL ELEMENT TEMPERATURE .................................. 6 2.2 LIMITING SAFETY SYSTEM SETTING ( LSSS ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6i 3.0 LIMITING CONDITIONS FOR O PER ATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
'3 1 STE AD Y ST ATE OP ER ATIO N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 3.2 R EACTI VITY LIM ITATIO N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3.3 PU LS E MOD E O P E R ATIO N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 IN) 3.4 COR E CONFIGUR ATION LIMITATIO N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
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TECHNICAL SPECIFICATIONS TABLE OF CONTENTS 3.5 CONTRO L A ND S AFETY SYST EM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 3 5.1 Scram Time .................................................... 11 3 5.2 Reactor Control System ........................................ 11 3.5 3 Reactor Saf ety System and Interlocks . . . . . . . . . . . . . . . . . . . . . . . . . . 12 3.6 R ADI ATION MONITORING SYSTEM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 3.7 EV ACU AT IO N AL A RM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 3.8 A RGO N- 41 D I SC H AR G E L IM IT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 5 39 ENGINEERED F 4FETY FEATURE - FACILITY EXHAUST SYSTEM . . . . . . . . . . . . . . . . . . . 16 3 10 LIMITATIONT JF EXPERIMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 3.11 ALARA ................................................................ 18 4.0 SU RVEILL ANC E R EQU IR EM E NTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 9 4.1 R EACTOR CO R E P AR AM ET ERS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 9 4.2 R EACTOR CONTRO L AND S AFETY SYSTEMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 4.2.1 R EACTOR CONTRO L SYST EMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 0 4.2.2 REACTOR SAFETY SYSTEMS ........................................ 21 4.2.3 FU EL T EM P E R ATU R E ~ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 4.2.4 R EAC TOR I NTER LOC KS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2 4.2.5 REACTOR FUEL ELEMENTS ......................................... 23 4.3 C OO LA NT S YST EMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 4 4.4 FAC ILITY EXH AUST SYST EM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 4 4.5 R ADI ATION MCNITORING SYSTEM AND EV ACU ATION ALARM . . . . . . . . . . . . . . . . . . . . . . 2 4.6 POO L W ATE R LEVEL AL ARM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 5.0 DESIGN FEATURES ............................................................. 26 5.1 R EA C TOR FU EL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 6 5.2 R EAC TOR C O R E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 7 5.3 CO NTRO L RO D S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 7 5.4 FU EL STOR AG E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 8 5.5 R EACTOR B AY AND EXH AUST SYSTEM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 9 5.6 R EACTOR POO L WAT E R SYST EMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - . . . . . . . . . . . 3 0 6.0 ADMINI STR ATI VE CONTRO LS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 0 6.1 ORGANIZATION .......................................................... 30 6.2 R EV I EW AND AUD I T . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 6.3 ACTION TO BE TAKEN IN THE EVENT A SAFETY LIMIT IS EXCEEDED . . . . . . . . . . . . 32 6.4 ACTION TO BE TAKEN IN THE EVENT OF A REPORTABLE OCCURRENCE . . . . . . . . . . . . 33 6.5 OPERATING PROCEDURES .................................................. 33 6.6 FAC ILITY OPER ATING R ECORDS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 4 6.7 R EPOR TI NG R EQU IR EM ENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 5 1
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1 TECHNICAL SPECIFICATIONS FOR in!E
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s PENN STATE BREAZEALE REACTOR (PSBR)
FACILITY LICENSE NO. R-2 Included in this document are the Technical Specifications and the bases for the Technical Specifications. These bases, which provide the technical support for' the individual technical specifications, are included for information purposes only. They are not part of the Technical Specifications, and they do.not constitute limitations or requirements to which the licensee must adhere.
1.0 DEFINITIONS' 1.1 ABNORMAL OCCURRENCE An abnormal occurrence is defined for the ' purposes of .the reporting requirements of Section 208 of the Energy Reorganization Act of _1974 (P.L.93-438) as an unscheduled incident or event which the Nuclear Regulatory Commission determines is significant from the standpoi.7t of public health or safety. A more detailed description is given in ANS 15.1 Section 1.1. ,
1.2 ALARA k.) The ALARA program (As Low As Reasonably Achievable) is a program for maintaining occupational exposures to radiation and release of radioactive effluents to the environs as low as reasonably achievable.
13 AUTOMATIC MODE Automatic mode shall mean operation of the reactor with the mode selector switch in the automatic mode position.
1.4 CHANNEL' CALIBRATION A channel calibration is an adjustment of the channel such that its output responds, with acceptable range and accuracy, to known values of the parameter which the channel measures.
1.5 - CHANNEL CHECK A channel check is a . qualitative verification of acceptable performance by observation of channel behavior.
1.6. CHANNEL- TEST A channel ' test is the introduction of a signal into the channel to verify that it is operable. '
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1.7 COLD CRITICAL The reactor is in the cold critical condition when it is critical with the fuel and bulk water temperatures both below 100*F (43.3aC).
1.8 CLOSE PACKED ARRAY An arrangement of fuel elements wherein no empty grid positions are completely surrounded by fuel elements.
1.9 CORE LATTICE POSITION The core lattice position is that region in the core over a grid plate hole used to position a fuel element. It may be occupied by a fuel element, an experiment, an experimental facility, or a reflector element.
1.10 EXPERIMENT Experiment shall mean (a) any apparatus, device, or material which is not a normal part of the core or experimental f acilities, but which is inserted in these facilities or is in line with a beam of radiation originating from the reactor core; or (b) any operation designed to measure reactor parameters or characteristics.
1.11 EXPERIMENTAL FACILITY Experimental facility shall mean beam port, including extension tube with shields, thermal column with shields, vertical tube, central thimble, in-core irradiation holder, pneumatic transfer system, and in pool irradiation facility.
1.12 FUEL ELEMENT A fuel element is a single TRIGA fuel rod of standard type, either 8.5 wt% U-ZrH in stainless steel cladding or 12 wt5 U-ZrH in stainless steel cladding.
1.13 INSTRUMENTED ELEMENT An instrumented element is a fuel element in which sheathed chromel-alumel or equivalent thermocouples are embedded in the fuel.
1.14 LIMITING CONDITIONS FOR OPERATION Limiting conditions for operation of the reactor are those administratively established constraints required for safe operation of the facility.
1.15 LIMITING SAFETY SYSTEM SETTING A limiting safety system setting is a setting for an automatic protective device related to a variable having a significant safety function.
3 em 1.16 MEASURED VALUE
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The measured value is the magnitude of a variable as it appears on the output of a measuring channel.
1.17 MEASURING CHANNEL A measuring channel is the combination of sensor, interconnecting cables or lines, amplifiers, and output device which are connected for the purpose of measuring the value of a variable.
1.18 MINIMUM SHUTDOWN MARGIN Minimum shutdown margin shall mean the minimum shutdown reactivity necessary to provide confidence that the reactor can be made suboritical by means of the control and safety system, starting from any permissible operating conditions and that the reactor will remain suboritical without further operator action.
1.19 MOVABLE EXPERIMENT A movable experiment is one where it is intended that the entire experiment may be moved in or near the core or into and out of the reactor while the reactor is operating.
1.20 NORMALIZED POWER (NP)
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The normalized power, NP, is the ratio of the power of a fuel element to the average power per fuel element.
1.21 OPERABLE A system, device, or component shall be considered operable when it is capable of performing its intended functions in a normal manner.
1.22 OPERATIONAL CORE An operational core is one for which the core parameters of shutdown margin, fuel temperature, power calibration, maximum excess reactivity, and maximum reactivity insertion rate have been determined to satisfy the requirements of the Technical Specifications. It may contain 8.5 wt$ U-ZrH or a mixture of 8.5 wt$ U-ZrH and 12 wt1 U-ZrH TRIGA fuel elements.
1.23 - PULSE MODE Pulse mode operation shall mean operation of the reactor with the mode selector switch in a pulse position.
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4 1.24 REACTOR INTERLOCK A reactor interlock is a device which prevents some action, associated with reactor operation, until certain reactor operation conditions are satisfied.
1.25 REACTOR SAFETY SYSTEMS Reactor safety systems are those systems, including their associated input circuits, which are designed to initiate a reactor scram.
1.26 REACTOR SECURED The reactor is secured when all the following conditions are satisfied:
- a. The reactor is shut down.
- b. The console key switch is in the off position and the key is removed from the console and under the control of a licensed operator or stored in a locked storage area,
- c. No work is in progress involving in-core fuel handling or refueling operations, or insertion or withdrawal of in-core experiments.
1.27 REACTOR SHUTDOWN The reactor is shut down when the reactor is subcritical by at least one dollar of reactivity.
1.28 REPORTABLE OCCURRENCE A reportable occurrence is any of the following which occurs during reactor operation:
- a. Operation with the safety system setting less conservative than specified in Section 2.2, limiting safety system setting.
- b. Operation in violation of a limiting condition for operation,
- c. Failure of a required reactor safety system component which could render the system incapable of performing its intended safety function.
- d. Any unanticipated or uncontrolled change in reactivity greater than one dollar.
- e. An observed inadequacy in the implementation of either administrative or procedural controls which could result in operation of the reactor outside the limiting conditions for operation.
- f. Release of fission products from a fuel element.
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, ) 1.29 SAFETY CHANNEL '
A safety channel it a measuring channel in a reactor safety system.
1.30 SAFETY LIMITS Safety limits are limits on imoortant process variables which are found to be necessary to reasonably protect the integrity of certain of the physical barriers which guard against the uncontrolled release of radioactivity.
1 31 SECURED EXPERIMENT A secured experiment is an experiment held firmly in place by a mechanical device or by gravity such that it cannot be moved by a force of less than sixty pounds.
1.32 SECURED EXPERIMENT WITH MOVABLE PARTS A secured experiment with movable parts is one that contains parts that are intended to be moved while the reactor is operating.
1 33 SHIM, REGULATING, SAFETY RODS A shim, regulating, or safety rod is a control rod having an
('~N electric motor drive and scram capabilities. It may have a fueled
(,,j) follower section.
~1 34 SQUARE WAVE MODE Square wave mode operation shall mean operation of the reactor with the moae selector switch in the square wave position.
1.35 STANDBY Standby is that condition when the reactor is suberitical but has approximately $3.50 of control rod worth removed from the core.
1 36 STEADY STATE MODE Steady state mode operation shall mean operation of the reactor with the mode selector switch in the steady state position.
i 1.37 TRANSIENT ROD The transient rod is a control rod with scram capabilities that can be rapidly ejected from the reactor core to produce a pulse. It may have a voided or an aluminum follower.
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6 2.0 SAFETY LIMIT AND LIMITING SAFETY SYSTDK SETTING 2.1 SAFETY LIMIT-FUEL ELEMENT TEMPERATURE Applicability The safety limit specification applies to the maximum temperature in the reactor fuel.
Obj ective The objective is to define the maximum fuel element temperature that can be permitted with confidence that no damage to the fuel element and/or cladding will result.
Specifications The temperature in a standard TRIGA fuel element shall not exceed 1150*C under any operating condition.
Basis The important parameter for a TRIGA reactor is the fuel element temperature. This parameter is well suited as a single specification especially since it can be measured at a point within the fuel element. The measured fuel temperature is directly related to the maximum fuel temperature of the region. A loss in the integrity of the fuel element cladding could arise from a build-up of excessive pressure between the fuel-moderator and the cladding if the maximum fuel temperature exceeds 1150*C. The pressure is caused by the presence of air, fission product gases, and hydrogen from the dir.sociation of the hydrogen and zirconium in the fuel-moderator.
The magnitude of this pressure is determined by the fuel-moderator temperature, the ratio of hydrogen to zirconium in the alloy, and the rate change ir. the pressure.
The safety limit for the standard TRIGA fuel is based on data, including the large mass of experimental evidence obtained during high performance reactor tests on this fuel. These data indicate that the stress in the cladding doe to the increase in the hydrogen pressure from the dissociation of zirconium hydride will remain below the ultimate stress provided that the temperature of the fuel does not exceed 1150*C (2102*F) and the fuel cladding is water cooled. See Safety Analysis Report, section IX.
2.2 LIMITING SAFETY SYSTEM SETTING (LSSS)
Applicability The LSSS specification applies to the scram setting which prevents the safety limit from being reached.
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7 Objective l
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The objective' is to prevent the safety limit (1150*C) from being
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reached.
Specifications The limiting safety system setting shall be a maximum of 700*C as measured with a 12 wt$ U-ZrH instrumented fuel element. The
' instrumented fuel element shall be located in the B-ring and l adjacent to an empty fuel position when an empty fuel position i exists in the B-ring.
Basis The limiting safety system setting is a temperature which, if reached shall cause a reactor scram to be initiated preventing the safety limit ffcm being arceeded. Experiments and analyses described in-the
- Safety Analysis Report,:Chapl6r IX - Safety Evaluation, show that the measured fuel temperature at steady state pcuer has a simple linear
' relationship to the normalized power or power of the imighest powered .
fuel element in the core. Maximum fuel temperature occurs when a new 12 wt5 U-ZrH fuel element-is placed in the B-ring of the core. The measured fuel temperature during steady state operation is close to
.the maximum fuel temperature. Thos, 450*C of safety margin exists before the 1150*C safety limit is reached.- This safety margin provides adequate compensation for using a depleted instrumented 12 wt%~U-ZrH fuel element instead of.unieradiated one to measure the fuel temperature. See Safety Analysis Report, section IX.
LIn the. pulse mode-of operation, the same limiting safety system setting shall apply. However, the. temperature channel will have no effect on limiting the peak power generated, because of its relatively
, long time constant . (seconds), compared with the width of the pulse
! (milliseconds). In this mode, however, the temperature trip will act
! to reduce the amount of energy generated in the entire pulse transient, by cutting the " tail" of the power transient Lif the pulse rod remains stuck -in the fully withdrawn position -with ~enough -
' reactivity to exceed the temperature-limiting safety system setting.
3.0 LIMITING CONDITIONS FOR OPERATION 31 STEADY STATE OPERATION
. Applicability This specification applies to the maximum power generated during
-steady state operation.
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8 Obj ective The objective is to assure that the safety limit (fuel temperature) will not be reached during steady state operation by providing a set point to automatically limit the maximum neutron flux produced in the core and to limit the energy produced in any seven (7) consecutive days to that used in the LOCA analysis in the Safety Analysis Report.
Specification
- a. The reactor steady state power level shall not exceed 1.15 megawatts. The normal steady state operating power level of the reactor shall be limited to 1.0 + 0.05 megawatts. However, for purposes of testing and calibration, the reactor may be operated at power levels not to exceed 1.15 megawatts for times not to exceed one hour during a test period,
- b. The reactor shall not be operated to produce more than 70 megawatt hours of energy in any seven (7) consecutive days.
Basis Thermal and hydraulic calculations and operational experience indicate that TRIGA fuel may be safely operated up to power levels of at least 1.5 megawatts with natural convective cooling. Steady state operation at 1.15 megawatts will not produce fuel temperatures which exceed 600*C using any allowed core configuration giving a large safety measure at this power of operation. See Safety Analysis Report, section IX.
This specification limits the energy output of the PSBR to that used for the Safety Analysis heport analysis of a LOCA.
3.2 REACTIVITY LIMITATION Applicability These specifications apply to the reactivity condition of the reactor and the reactivity worth of control rods, experiments, and experimental facilities. They apply for all modes of operation.
Obj ective The objective is to assure that the reactor can be shut down at all times and to assure that the safety limit will not be exceeded.
Specifications
- 1. The maximum excess reactivity above cold, clean, critical plus samarium poison of the core configuration with experiments and experimental facilities in place shall be 4.9% ak/k ($7.00).
- 2. The reactor shall not be operated unless the shutdown margin provided by control rods is greater than 0.25 dollar with:
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, a. All movable experiments, experiments with movable parts, and experimental facilities in their most reactive state.
- b. The highest reactivity worth control rod fully withdrawn.- !
- c. The reactor in the cold condition without xenon.
1 Basis Limiting the excess . reactivity of the core to $7.00 prevents the fuel temperature in'the core from exceeding 1150*C under any assumed accident condition as described in the Safety' Analysis Report, Chapter IX - Safety Evaluation. The value of the shutdown margin assures that the reactor can be made subcritical from any operating condition even if the highest worth control rod should remain in the -
fully-withdrawn position.
. 33 PULSE MODE OPERATION
' Applicability This specification applies to the energy generated in the reactor as
.a result of a _ pulse insertion of reactivity.
j; Obj ective The objective is to assure that the safety limit will not be exceeded during pulse mode operation.
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The stepped reactivity insertion for pulse operation shall not -
exceed 2.385 Ak/k ($3.40) and the maximum worth of the poison section of the transient rod with respect to water shall be limited to 2.595 ak/k ($3.70).
. Basis i' .
Experiments and analyses described in the SAR show that .the peak i pulse temperatures can be predicted for new 12 wtl fuel placed in '
the B-ring. . These experiments and analyses show that the maximum L
allowed pulse reactivity of $3.40 ' prevents the maximum measured fuel temperature from reaching 700*C for any allowed core configuration.
The maximum worth of the pulse rod is limited to $3 70 to prevent
- exceeding the safety limit (1150*C). Accidental insertion of the i
' transient rod during cold clean conditions will limit the maximum '
measured temperature to 720*C and the maximum fuel temperature to 1150*C. See Safety Analysis Report, section IX.
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34 CORE CONFIGURATION LIMITATION
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Applicability This specification applies to a core configuration with water holes producing large power peaking in some fuel elements.
Obj ective The objective is to assure that the safety limit will not be exceeded due to power peaking effects in the various core geometries.
Specifications
- a. The critical core shall be an assembly of either 8.5 wt%
stainless steel clad or a mixture of 8.5 wt% and 12 wt%
stainless steel clad TRIGA fuel-moderator elements placed in water with a 1.7 inch center line grid spacing.
- b. The fuel and fueled follower control rods shall be arranged in a close packed array except for single positions and core lattice positions the centers of which are greater than 5 inches from the center of the core where flux peaking and corresponding power densities produce fuel temperatures less than in the B-ring.
- c. When the ek rr of the core is less than or equal to 0.99 with all control rods at their upper limit, the fuel may or may not be arranged in a close packed array. The source and detector shall be arranged such that the ke rr of the subcritical assembly shall always be measured and monitored to assure compliance with ke rr
< 0.99 wnen all control rods are fully withdrawn from the core.
,B, a,, s,1 s, Calculations and experiments performed with the PSBR have shown that with only one empty fuel position in the central region of the core defined as lattice positions with centers less than 5 inches from the core center, the power peaking for any mixture of 12.0 wt% and 8.5 wt% fuel remains less than 23.2 kw per fuel element, i.e., an NP < 2.2 (see Safety Analysis Report, Chapter IX - Safety Evaluation). This automatically limits the maximum fuel temperature, which always occurs in the B-ring, to well below 700*C for any mode of operation. The maximum fuel temperature always occurs in a new 12 wt% fuel element in the B-ring adjacent to a water filled fuel position also in the B-ring.
When theek rr of the core is less than 0.99 with all control rods at their upper limit, the core can not be taken critical. Hence, the requirement for close packed arrays is not necessary to prevent the core from attaining high fuel temperatures.
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4 11 3.5 CONTROL AND SAFETY SYSTEM
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3 5.1 Scram Time i
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Applicability i This specification applies to the time required to fully insert any control rod to a full down position from a full up position.
Objective The objective is to achieve rapid shutdown of the reactor to prevent
, fuel damage.
Specification
- The time from scram initiation to the full insertion of any control rod from a full up position shall be less than 1 second.
Basis This specification assures that the reactor will be promptly shut down when a scram signal is initiated. Experience and analysis (see Safety Analysis Report, Chapter IX - Safety Evaluation) have indicated that for the range of transients anticipated for a TRIGA reactor, the specified scram time is adequate to assure the safety of the reactor.
3.5.2 Reactor Control System Applicability This specification applies to the information which must be available to the reactor operator during reactor operation.
Obj ective The objective is to require that sufficient information is available to the operator to assure safe operation of the reactor.
Specification The reactor shall not be operated in a given mode unless the measuring channels listed in Table 1 for that mode are operable.
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Table 1 Measuring Channels Min. No. Effective Mode Measuring Channel Operable SS Pulse SW Fuel Element Temperature 1 X X X Linear Power 1 X X Percent Power 1 X X Pulse Peak Power 1 X Count Rate 1 X Log Power 1 X X Basis Fuel temperature displayed at the control console gives continuous information on this parameter which has a specified safety limit.
The power level monitors assure that the reactor power level is adequately monitored for both steady state and pulsing modes of operation. The specifications on reactor power level indication are included in this section since the maximum fuel temperature is related to the power level.
3.5.3 Reactor Safety System and Interlocks Applicability This specification applies to both the reactor safety system channels and the interlocks.
Obj ective The objective is to specify the minimum number of reactor safety system channels and interlocks that must be operable for safe operation.
Specification The reactor shall not be operated unless all of the channels and interlocks described in Table 2a and Table 2b are operable.
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r 13 Table 2a Minimum PSBR Safety Channels Number Effective Mode
' Safety Channel Operable Function .SS Pulse SW Fuel Temperature 1 SCRAM i 700'C X X X High Power- 2 SCRAM i 1155 of 1 MW X X Detector Power- 1 SCRAM on failure of X X Supply supply voltage
. Scram Bar on Console-' 1 Manual scram X X X Preset Timer- 1 Transient rod scram X 15 seconds or less after pulse Table 2b Minimum PSBR Safety Interlocks Number Effective Mode Safety Interlocks Operable Function SS, Pulse SW Source Level 1 Prevent rod withdrawal X with less than two neutron induced counts per second on the startup channel Log Power 1 Prevent' pulsing from 'X levels above 1 -kW
-Transient Rod 1 Prevent applications X of air unless cylinder is fully. inserted Shim, Safety, and. 1 Movement of any rod X Regulating Rod except transient rod Simultaneous Rod 1 Prevents simultaneous X X
' Withdrawal-manual. withdrawal of-two rods Basis A temperature scram and two power level scrams provide protection to assure that the reactor can be shut'down before the safety limit on O. the fuel element temperature will be exceeded. The manual scram I
14 allows the operator to shut down the system in any mode of operation if an unsafe or abnormal condition occurs. In the event of failure of the power supply for the safety chambers, operation of the reactor without adequate instrumentation is prevented. The preset timer insures that the reactor power level will reduce to a low level af ter pulsing.
In the pulse mode, movement of any rod except the transient rod is prevented by an interlock. This interlock action prevents addition of excess reactivity over that in the transient rod. The interlock
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to prevent startup of the reactor with less than 2 cps assures that sufficient neutrons are available for proper startup. The interlock to prevent the initiation of a pulse above 1 kW is to assure that the magnitude of the pulse will not cause the safety limit to be exceeded. The interlock to prevent application of air to the transient rod unless the cylinder is fully inserted is to prevent pulsing the reactor in the steady state mode.
3.6 RADIATION MONITORING SYSTEM Applicability This specification applies to the radiation monitoring information which must be available to the reactor cp9eator during reactor operation.
Objective The objective is to assure that sufficient radiation monitoring information is available to the operator to assure safe operation of the reactor.
Specification The reactor shall not be operated unless the radiation monitoring channels listed in Table 3 are operable.
Table 3 Radiation Monitoring Channels Radiation Monitoring Channels Function Number Area Radiation Monitor Monitor radiation levels 1 in the reactor bay Air Radiation Monitor 1 9
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-Basis l O- The radiation monitors provide information to operating personnel of any impending or existing danger from radiation so that there will be sufficient time to evacuate the facility and take the necessary steps to prevent the spread of radioactivity to the surroundings.
3.7 EVACUATION ALARM Applicability This specification applies.to the evacuation alarm which must be audible to personnel within the PSBR building when activated by the
' radiation monitoring channels in table 3 or a manual switch.
" Objective The objective is to assure that all personnel are alerted to evacuate the PSBR building when a potential radiation hazard exists within this building.
Specification
'The reactor shall not be operated unless the evacuation alarm is operable. .
Basis The evacuation alarm produces a loud pulsating sound throughout the PSBR building when there is any impending or existing danger from radiation. The sound notifies all personnel within the PSBR building to evacuate the building as prescribed by the PSBR
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emergency procedures.
4 3.8 ARGON-41 DISCHARGE lLD4IT Applicability This specification applies to the concentration of argon-41 that may be discharged from the PSBR. l
. Objective l..
l~ To _ insure that the health and safety of the public is not endangered by the discharge of argon-41 from the PSBR.
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Specification
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, The concentration in the unrestricted aree of argon-41 in the
. effluent gas from the facility as diluted by atmospheric air due to the turbulent wake effect shall not exceM 4.0 "x 10-8 pCi/ml averaged over one year in the unrestricte ' area.
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'T-" wy e wrur .yet4WMw--.ww,-yy+-~ ge e7--y-gy,u--g-myo-pm gr - gmg4ggmy,.gga,.mpge-e p +qwwmym+ g vgegy,-ev q.m m gi--97 &-it--g-'tvMW*D tW"u8D-Tet e w'M W7er-WNTw"5-eet"--'PT*-
16 j
The maximum allowable concentration of argon-41 in air in unrestricted areas as specified in Appendix B, Table II of 10 CFR 20 is 4.0 x 10-8pC1/ml.
I 3.9 ENGINEERED SAFETY FEATURE - FACILITY EXHAUST SYSTEM Applicability This specification applies to the operation of the facility exhaust system.
Objective The objective is to mitigate the consequences of the release of radioactive materials resulting from reactor operation.
I Specification The facility exhaust system shall be maintained in an operable condition except for periods of time less than 48 hrs. when it is necessary to permit maintenance and repairs.
Basis During normal operation, the concentration of airborne radioactivity in unrestricted areas is below MPC as described in the Safety Analysis Report,Section IX - Safety Evaluation. In the event of a substantial release of airborne radioactivity, an air radiation monitor and/or an area radiation monitor will sound a building l
evacuation alarm which will automatically cause the exhausted air to be passed through the system of filters before release. This reduces the radiation within the building and the filters will reduce by 90% all of the fission products but noble gases that escape to the atmosphere. Radiation monitors within the building independent of the facility exhaust system will give warning of high levels of radiation that might occur during operation with the L facility exhaust system out of service.
3.10 LIMITATIONS OF EXPERIMENTS Applicability This specification applies to experiments installed in the reactor and its experimental facilities.
Objective The objective is to prevent damage to the reactor and to prevent excessive release of radioactive materials in the event of an experiment failure.
17
[ '\ Specifications
\~-
The reactor shall not be operated unless the following conditions governing experiments exists
- a. A movable experiment and/or movable portions of a secured experiment shall have a reactivity worth less than 15 Ak/k '
( $1. 43 ) .
- b. The reactivity worth of any single experiment shall be less than 1.4% ak/k ($2,00).
- c. Explosive materials in quantities greater than 250 milligrams shall not be allowed within the PSBR facility. Irradiation of explosive materials shall be restricted as follows:
Explosive materials in quantities greater than 25 milligrams shall not be irradiated. Explosive materials in quantities less than 25 milligrams may be irradiated provided the pressure produced upon detonation of the explosive has been calculated
+-
and/or experimentally demonstrated to be less than the design pressure of the container.
- d. Experiment materials, except fuel materials, which could off gas, sublime, volatilize, or produce aerosols under (1) normal operating conditions of the experiment and reactor, (2)
(N-S) credible accident conditions in the reactor, or (3) possible
- accident conditions in the experiment shall be limited in activity such that the airborne concentration of radioactivity averaged over a year would not exceed the limit of Appendix B Table II of 10 CFR Part 20.
When calculating activity limits, the following assumptions shall be used:
(1) If an experiment fails and releases radioactive gases or aerosols to the reactor bay or atmosphere,100% of the gases or aerosols escapes.
(2) If the effluent from an experimental facility exhausts through a holdup tank which closes automatically on high radiation level, at least 10% of the gaseous activity or aerosols produced will escape.
(3) If the effluent from sa experimental facility exhausts through a filter installation designed for greater than 995 efficiency for 0.3 r.icron particles, at least 10% of these vapors can escape.
(4) For materials whose boiling point is above 130*F and where vapors formed by bollang this material can escape only through an undisturbed column of water above the core, at (s_--) least 10% of these vapors can escape.
1
- . . . ,- . - - . . , , - - - . - . - - . - - - - - . , . _ , - - - . , - - . , - . ,e - - - , -..w -
18
- e. Each fueled experiment shall be controlled such that the total inventory of iodine isotopes 131 through 135 in the experiment is no greater than 1.5 curies,
- f. If a capsule fails and releases material which could damage the reactor fuel or structure by corrosion or other means, physical inspection shall be performed to determine the consequences and need for corrective action. The results of the inspection and any corrective action taken shall be reviewed by the Director or a designated alternate and determined to be satisfactory before oper& tion of the reactor is resumed.
Basis
- a. This specification is intended to provide assurance that the worth of a single movable experiment or movable part of a secured experiment will be limited to a value such that the safety limit will not be exceeded if the positive worth of the experiment were to be suddenly inserted,
- b. The maximum worth of a single experiment is limited to 1.4% Ak/k so that its removal from the cold critical reactor will not result in the rssctor achieving a power level high enough to exceed the core te.1perature safety limit. Since experiments of such worth are secured in place, its removal from the reactor operating at full power vould result in a relatively slow power increase such that the reactor protective systems would act to prevent high power levela from being attained.
- c. The failure of an experiment involving the irradiation of up to 25 milligrams of properly contained explosive material in a reactor irradiation facility will not result in damage to the reactor or the reactor pool containment structure.
This specification is also intended to prevent damage to vital equipment by restricting the quantity of explosive materials to less than 250 milligrams within the reactor building.
- d. This specification is intended to reduce the likelihood that airborne activities in excess of the limits of Appendix B Table II of 10 CFR Part 20 will be released to the atmosphere outside the facility boundary.
- e. The 1.5 curie limitation on iodine-131 through 135 assures that, in the event of failure of a fueled experiment leading to total release of the iodine, the exposure dose at the exclusion area boundary will be less than that allowed by Appendix B of 10 CFR Part 20 for an unrestricted area.
- f. Operation of the reactor with the reactor fuel or structure damaged is prohibited to avoid release of fission products.
3 11 ALARA
19 Applicability This specification applies to all reactor operations that could result in significant personnel exposures.
. Objective To maintain all exposures to ionizing radiation to the staff and the general public as low as reasonably achievable.
Specification As part of the review of all operations, consideration shall be
- given to alternative operational profiles that might reduos staff L exposures, release of radioactive materials to the environment, or l both.
L I
Basis Experience has shown that experiments and oNeational requirements can, in many cases, be satisfied with a variety of combinations of
, facility options, core positions, power levels, time delays, and l
other modifying factors. Many of these can reduce radioactive effluents or staff radiation exposures. Similarly,'overall reactor scheduling achieves significant reductions ,in staff exposures.
Consequently, ALARA must be a part of both the overall reactor scheduling and the detailed experiment planning.
O 4.0 -~ SURVEILLANCE REQUIREMENTS 4.1 REACTOR CORE PARAMETERS
' Applicability These specifications apply to the surveillance requirements for the excess reactivity of the core, and the reactivity control of experiments and systems affecting reactivity.
Objective The objective is to measure and verify the core reactivity, and the worth, performance, and operability of those systems affecting the reactivity of the reactor.
Specifications
- a. The reactivity of a core shall be measured to determine that its excess a reactivity does not exceed $7.00.
- b. The reactivity worth of each control rod and the shutdown margin for p the. core loading in use shall be determined annually but at intervals Q not to exceed 15 months.
20
- c. The reactivity worth of an experiment shall be estimated or measured as appropriate, before reactor power operation with an experiment, the first time it is performed.
- d. The control rods shall be visually inspected for deterioration at intervals of 2 years, not to exceed 26 months,
- e. On each day that pulse mode operation of the reactor is planned, a functional performance check of the transient rod system shall be performed. Semiannually, at intervals not to exceed 8 months, the transient rod drive cylinder and the associated air supply system shall be inspected, cleaned, and lubricated as necessary,
- f. The reactor shall be pulsed semiannually not to exceed 8 months to compare fuel temperature measurements and peak power levels with those of previous pulses of the same reactivity value or the reactor shall not be pulsed until such comparative pulse measurements are performed.
Basio Standard Operating Procedures have been established for the PSBR which provide adequate surveillance of the core reactivity to assure compliance with the $7.00 excess reactivity limit.
The reactivity worth of the control rods is measured to assure that the required shutdown margin is available and to provide an accurate means for determining the reactivity worths of experiments inserted in the core.
Past experience with TRIGA reactors gives assurance that measurements of the reactivity worth, on an annual basis, are adequate to insure that no significant changes in the shutdown margin have occurred. Visual inspection of the control rods is made to evaluate corrosion and wear characteristics caused by operation in the reactor. Functional checks along with periodic maintenance assure repeatable performance. The reactor is pulsed at suitable intervals and a comparison made with previous similar pulses to determine if changes in fuel or core characteristics are taking place.
4.2 REACTOR CONTROL AND SAFETY SYSTEMS 4.2.1 REACTOR CONTROL SYSTEMS Applicability These specifications apply to the surveillance requirements for reactor control systems.
Objective The objective is to verify the operability of systems that affect the safe and proper control of the reactor.
i 21
('~'N Specification 1
v] The control rod drop times shall be measured semiannually, but at intervals not to exceed 8 months.
Basis Measurement of the scram time on a semiannual basis is a verification of the scram system, and is an indication of the capability of the control rods to perform properly.
4.2.2 REACTOR SAFETY SYSTEMS Applicability aneo, opool.~1.oti.... opply tu cne surveillance requirements for measurements, tests, and calibrations of the reactor safety systems.
Objective The objective is to verify the performance and operability of the systems and components that are directly related to reactor safety.
Specifications
- a. A channel test of the scram function of the high-flux safety
[\ _/) channels shall be made on each day that the reactor is to be operated, or prior to each operation that extends more than one day.
- b. Channel calibration shall be made of the linear power level-monitoring channels annually, at intervals not to exceed 15 months.
Basis TRICA system components have proven operational reliability. Daily channel tests insure accurate scram functions and insure the detection of possible channel drift or other possible deterioration of operating characteristics. The channel checks will assure that the safety system channel scrams are operable on a daily basis or prior to an extended run. The power level channel calibration will assure that the reactor is to be operated at the authorized power levels.
4.2 3 FUEt, TEMPERAnlRE These specifications apply to the surveillance requirements for the safety channel measuring the fuel temperature.
Objective
/^) To insure operability of the fuel temperature measuring channel.
22 Specifications
- a. A channel test of the fuel temperature scram shall be made on each day that the reactor is operated or prior to each operation that extends more than one day.
- b. A calibration of the fuel temperature measuring channel shall be made annually, at intervals not to exceed 15 months.
- c. If a reactor scram caused by high fuel element temperature occurs, an evaluation shall be conducted to determine whether the fuel element temperature actually exceeded the safety limit.
Basis Operational experience with the TRIGA system assures that the thermocouple measurements have been proven sufficiently reliable as an indicator of fuel temperature. The daily channel test assures operability. The daily scram check assures scram capabilities.
4.2.4 REACTOR INTERLOCKS Applicability This specification applies to the surveillance requirements for the reactor control system interlocks.
Objective To insure performance and operability of the reactor control system interlocks.
Specifications
- a. A channel check of the source interlock shall be performed each day that the reactor is operated or prior to each operation that extends more than one day,
- b. A channel test shall be performed semi-annually, not to exceed 8 months on the log power interlock which prevents pulsing from power levels higher than one kilowatt.
- c. A channel test shall be performed semi-annually, not to exceed 8 months, on the transient rod interlock which prevents application of air to the transient rod unless the cylinder is fully down.
- d. A channel test shall be performed semi-annually, not to exceed 8 months, on the rod drive interlock which prevents movement of any rod except the transient rod in pulse mode.
- e. A channel test shall be performed semi-annually, not to exceed 8 months, on the rod drive interlock which prevents simultaneous manual withdrawal of more than one rod.
23
[~'N Basis These channel tests and checks will verify operation of the reactor interlock system. Experience at the PSBR indicates that the prescribed frequency is adequate to insure operability.
4.2.5 REACTOR FUEL ELEMENTS Applicability This specification applies tu the surveillance requirements for the fuel elements. ,
Obj ective The objective is to verify the continuing integrity of the 1 fuel element cladding.
Specifications All fuel elements and control rods with fueled followers shall be inspected visually for damage or deterioration and measured for length and bend at intervals not to exceed the sum of 3,500 dollars in pulse reactivity or two years, not to
~ exceed 26 months, whichever comes first. The reactor shall L [ 'h not be operated with damaged fuel. A fuel elenent shall be
\_,/ removed from the core if
- a. In measuring the transverse bend, the bend exceeds 0.125 inch over the length of- the cladding.
- b. In measuring the elongation, its length exceeds its original length by 0.125 inch.
- c. A clad defect exists as indicated by release of fission products.
Basis The frequency of inspection and measurement schedule is based on the parameters most likely to affect.the fuel cladding of a pulsing reactor operated at moderate pulsing levels and utilizing fuel elements whose characteristics are well known.
The limit of transverse bend has been shown to result in no difficulty in disassembling the core. Analysis of the removal of heat from touching fuel elements shows that there will be no hot spots resulting in damage to the fuel caused by this touching.
Experience with TRIGA reactors has shown that fuel element l'\s_,/
h bowing that could result in touching has occurred without deleterious effects. This is because (1) during steady state
34 !
l operation, the maximum fuel temperatures are several hundred degrees below 1150'C (the safety limit), and (2) during a ,
pulse, the cladcing temperatures remain well below their stress limit due to the adiabatic nature of the fuel I temperature rise. The elongation limit has been specified to assure that the cladding material will not be subjected to stresses that could cause a loss of integrity in the fuel containment and to assure adequate coolant flow.
4.3 COOLANT SYSTEMS Applicability This specification applies to the surveillance requirements for monitoring the pool water and the water conditioning system.
Objective The objective is to assure the integrity of the water purification system, thus maintaining the purity of the reactor pool water, eliminating possible radiation hazards from activated impurities in the water system, and limiting the potential corrosion of fuel cladding and other components in the primary water system.
Specifications
- a. An alarm shall annunciate and corrective action shall be taken if during operation the bulk pool water temperature reaches 100*F.
- b. The conductivity of the water at the output of the purification system or near the surface of the open pool shall be measured weekly, not to exceed 8 days, whenever operations are planned.
Basis .
Based on experience, observation at these intervals provides acceptable surveillance of limits that assure that fuel clad c0 erosion and neutron activation of dissolved materials will not occur.
4.4 FACILITY EXHAUST SYSTEM Applicability This specification applies to the facility exhaust system.
Objective The objective is to assure the proper operation of the exhaust system in controlling releases of radioactive material to the uncontrolled environment.
25 Specification
( >
It shall be verified weekly, not to exceed 8 days, whenever
. operation is planned that the exhaust system is operable within its design specification.
-Basis Experience accumulated over a few years of operation has demonstrated that a test of the exhaust _ system on a weekly basis is sufficient to assure the proper operation of the system. This provides reasonable assurance on the control of the release of radioactive material.
4.5 RADIATION MONITORING SYSTEM AND EVACUATION ALARM
~ Applicability
.This specification applies to surveillance requirements for the area radiation monitor, the air radiation monitor, and the evacuation alarm.
Obj ective The objective is to assure that the radiation monitors and evacuation alarm are operating and to verify the appropriate alarm settings.
Specification The area radiation monitor, the air radiation monitor, and the evacuation alarm system shall be channel tested weekly, not to
- exceed 8 days, when operation is planned. They shall'be verified to be operable by a channel l check daily when the reactor is in operation, and shall be calibrated annually, not to exceed 15 months.
Basis Experience has shown that weekly verification of area radiation -
monitor, air. radiation monitor set points and operability and the evacuation alarm operability is adequate to correct for any variation in the system due to a change of operating characteristics.
Annual channel calibration insures that units are within the specifications.
4.6 POOL WATER LEVEL ALARM Applicability This specification applies to the surveillance requirements for the pool water level alarm. !
I 26 Obj ective The objective is to verify the operability of the pool water level alarm. I Specification The pool water level channel shall be channel checked weekly to assure its operability.
Basis Experience has shown that weekly checks of the pool water level alarm assure operability of the system during the week.
5.0 DESIGN FEATURES 5.1 REACTOR FUEL Applicability This specification applies to the fuel elements used in the reactor Core.
Obj ective The objective is to assure that the fuel elements are of such a design and fabricated in such a manner as to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics.
Specifications The individual unirradiated TRIGA fuel elements shall have the i following characteristics:
(1) The uranium content shall be a maximum of either 9.0 wt5 or 12.7 wt% enriched to a nominal 20% uranium-235.
(2) The hydrogen-to-zirconium aton ratio (in the ZrH x) shall be a nominal 1.65 H atoms to 1.0 Zr atom.
(3) The cladding shall be 304 stainless steel with a nominal 0.020 inch thickness.
Basis A maximum uranium content of 9 wt5 or 12.7 wt% in a TRIGA element is about 6% greater than the design value of 8.5 wt% or 12 wt%,
respectively. Such an increase in loading would result in an increase in power density of less than.6%. An increase in local power density of 6% reduces the safety margin by at most 10%. The hydrogen-to-zirconium ratio of 1.65, as presented in the Safety
27 4
4 '
Analysis Report,Section IX - Safety Evaluation, will produce a
(~ maximum pressure within the clad during any possible accident well below the rupture strength of the clad. Such accidents are analyzed in the Safety Analysis Report, Chapter IX - Safety Evaluation.
5.2 REACTOR CORE-Applicability
. This specification applies to the configuration of fuel and in-core
.- experiments.
, Objective i
The objective is to assure that provisions are made to restrict the arrangement of fuel elements and experiments so as to provide assurance that excessive power densities will not be produced.
Specifications
- a. The core shall be an arrangement of TRIGA uranium-zirconium hydride fuel-moderator elements positioned in the reactor grid plate.
- b. The reflector, excluding experiments and experimental facilities, shall be water or a combination of graphite and
> water, k
Basis.
- a. 'TRIOA cores using 8.5 wtl and 12 wt5 fuel elements have been in
- use for years and their characteristics are well documented.
The analyses of experimental data have shown that the mixed ~12 wt5 and 8.5 wt5 fuel satisfy all operational requirements.
- b. The core will be assembled in the reactor grid plate which is located in a pool of light water. Water in combination with graphite reflectors can be used for* neutron economy and the enhancement of experimental facility radiation requirements.
53 CONTROL RODS Applicability 2
- - This specification applies to the control rods used in the reactor core.
, Objective 1
. The objective is to assure that the control rods are of such a l design as to permit their use with a high degree of reliability with respect to their physical and nuclear characteristics. -
l l
4 i
.-. - - - _ . . . . ~ .
/
28 Specification
- a. The shim, safety, and regulating control rods shall have scram capability and contain borated graphite, B C4 powder, or boron and its compounds in solid form as a poison in aluminum or stainless steel cladding. These rods may incorporate fueled followers which have the same characteristics as the fuel region in which they are used.
- b. The transient control rod shall have scram capability and contain borated graphite, B C4 powder, or boron and its compounds in a solid form as a poison in an aluminum or stainless steel clad. When used as a transient rod, it shall have an adjustable upper limit to allow a variation of reactivity insertions. This rod may incorporate a voided or an aluminum follower.
Basis
- The poison requirements for the control rods are satisfied by using neutron absorbing borated graphite, B 4C powder, or boron and its compounds. These materials must be contained in a suitable clad material, such as aluminum or stainless steel, to insure mechanical stability during movement and to isolate the poison from the pool water environment. Control rods that are fuel followed provide additional reactivity to the core and increase the worth of the control rod. Scram capabilities are provided for rapid insertion of the control rods which is a primary safety feature of the reactor.
The transient control rod is designed to provide pulsing capabilities.
5.4 FUEL STORAGE Applicability This specification applies to the storage of reactor fuel at times when it is not in the reactor core.
Objective The objective is to assure that fuel which is being stored will not become critical and will not reach an unsafe temperature.
Specifications
- a. All fuel elements shall be stored in a geometrical array where the ke rr is less than 0.8 for all conditions of moderation.
- b. Irradiated fuel elements and fueled devices shall be stored in an array which will permit sufficient natural convection cooling by water or air such that the fuel element or fueled device temperature Will not reach design values.
O
29 j Basis V
An array of fuel elements having a ke rr less than 0.8 provides a large safety margin from accidentally reaching a ek rr of 1.0. The fuel elements are stored in vertical positions with sufficient space between them that natural convection cooling by water or air will prevent them from reaching temperatures capable of damaging the fuel elements.
5.5 REACTOR BAY AND EXHAUST SYSTEM Applicability This specification applies to the reactor bay which houses the reactor.
Obj ective The objective is to assure that provisions are made to restrict the amount of release of radioactivity into the environment.
Specifications
- a. The reactor shall oe housed in a room (reactor bay) designed to restrict leakage. The minimum free volume in the reactor bay
,-, shall be 2500 m3
( i
\s ' b. The reactor bay shall be equipped with two exhaust systems.
Under normal operating conditions, one of these systems exhausts unfiltered reactor bay air to the environment releasing it at a point at least 30 feet above ground level. Upon initiation of a building evacuation alarm, the previously mentioned system is automatically secured and an emergency exhaust system automatically starts. The emergency exhaust system is designed to discharge reactor bay air at a point at least 30 feet above ground level after passing it through a 3-stage filter system.
Basis Detection by either an area radiation monitor or an air radiation monitor of radioactivity above their respective alarm set points, will automatically secure the normal unfiltered, exhaust system and start the emergency, filtered, exhaust system. The emergency exhaust system will continue to maintain a negative pressure in the reactor bay drawing intake air only through leaks in the building.
Filters in the emergency exhaust system are designed to remove particulate matter and minimize the gaseous releases of airborne radioactivity. Controls to operate the emergency exhaust system (will not over-ride automatic start) are located in the cobalt-60 reception area. Flow indicators and gauges indicating pressure drops across each filter stage are also displayed in this area.
Monitoring of the operation of the emergency exhaust system during
(' 'N an emergency can be conducted from the cobalt-60 reception area with
\s_,) a minimum of radiation exposure to operating personnel. The filters
30 and blower for this emergency system are located on a roof adjacent to the reactor bay. These could be changed and serviced during an emergency with minimal radiation exposure to operating personnel.
An electrical disconnect is provided on the roof near the fan so that auxiliary electrical power could be connected in the event of a power failure during an emergency.
5.6 REACTOR POOL WATER SYSTEMS Applicability This specification applies to the pool containing the reactor and to the cooling of the core by the pool water.
Obj ective The objective is to assure that water shall be available to provide adequate cooling of the reactor core and ade.quate radiation shielding.
Specifications
- a. The reactor core shall be cooled by natural convective water flow.
- b. A pool level alarm shall be activated if the pool level drops approximately 26 cm below operating level.
Basis
- a. This specification is based on thermal and hydraulic calculations which show that the TRIGA core (mixed 8.5 wt% and 12 wt%) can operate in a safe manner at power levels which produce 30 kw in the hottest fuel element with natural convection flow of the coolant water.
- b. In the event of pipe failure and siphoning of pool water through the skimmer, the pool water level will not drop below the top of the core.
- c. A water loss which causes the pool level to drop 26 cm from the top of the pool requires corrective action. This alarm is located in the reactor control room and at the University Police Services panel, the latter providing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> monitoring.
6.0 ADMINISTRATIVE CONTROLS 6.1 ORGANIZATION
- a. The facility shall be under the direct control of the Director or a designated licensed senior operator. The Director shall be responsible to the Nuclear Engineering Department Head for safe
31 (7 operation and maintenance of the reactor and its associated i,
V
) equipment. The Director or appointee (s) shall review and approve all experiments and experimental procedures prior to their use in the reactor. The Director shall enforce rules for the protection of personnel from radiation.
- b. The operation of the PSBR is conducted under the following University Administration organization:
Vice President Research and Graduate Studies Dean, College of Engineering
,_s Nuclear Engineering
/
Department Head D )'
University Health Penn State Reactor Physics Safeguards Committee i I l l L_____ oirector Penn State Breazeale Reactor
_.___J I
Reactor Staff 6.2 REVIEW AND AUDIT
- a. The Penn State Reactor Safeguards Committee (PSRSC) shall
/^5 review, evaluate, and approve items concerning the safety
() associated with the operation and use of the facility. The PSBR
- 1 l
1 32 l Director shall be an ex-officio member of the PSRSC. The jurisdiction of the PSRSC shall include all nuclear reactor operations and experiments utilizing special nuclear material and source material in the f acility,
- b. The operations of the PSRSC shall be in accordance with a written charter, including provisions for (1) Heeting frequency, (2) Voting rules, (3) Quorums ,
(4) Records of meetinF.
- c. The PSRSC shall audit reactor operations at least quarterly, but at intervals not to exceed four months.
- d. The primary domain of jurisdiction of the PSRSC shall be the safety evaluation of in-core nuclear reactor experiments and the periodic review and evaluation of the physical integrity of the core. Expressly:
(1) The PSRSC shall be concerned with those reactor experiments which by their unusual nature, potential hazard or unprecedented complexity, could endanger health, life, and property in and about the PSDR.
(2) Reactor experiments which contain unusual dangers in addition to those listed above shall also be submitted to the PSRSC for evaluation. Favorable review of an experiment by the PSRSC shall not obligate the staff of the PSBR to carry out the experiment.
(3) Review and approval of all proposed changes to the Technical Specifications.
(4) Review of the operation and operational records of the facility.
(5) Review of unusual or abnormal occurrences and incidents which are reportable under 10 CFR Part 20 and 10 CFR Part 50.
(6) The PSRSC may aid the reactor staff with the review and evaluation of any indications that changes are taking place in the core which influsnoe its integrity.
63 ACTION TO DE TAKEN IN THE EVENT A SAFETY LIH!T IS EXCEEDED In the event the safety limit (1150*C) is exceeded f
r 33 t'~'% a. The reactor shall be shut down and reactor operation shall not
'd
) be resumed until authorized by the U.S. Nuclear Regulatory Commission.
- b. An immediate report of the occurrence shall be made to the Chairman, PSRSC and reports shall be made to the USNRC in accordance with Section 6.7 of these specifications.
- c. A report shall be prepared which shall include an analysis of the causes and extent of possible resultant damage, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence. This report shall be submitted to the PSRSC for review and then submitted to the USNRC.
6.4 ACTION TO BE TAKEN IN THE EVENT OF A REPORTABLE OCCURRENCE In the event of a reportable occurrence, the following action shall be taken:
- a. The Director or a designatec alternate shall be notified and corrective action taken with respect to the operations involved,
- b. The Director or a designated alternate shall notify the Nuclear Engineering Department Head who, in turn, Will notify the office of the Dean of the College of Engineering and the office of the
/T Vice , President for Research and Graduate Studies.
! l
'~~
- c. The Director or a designated alternate shall notify the Chairman of the PSRSC.
- d. A report shall be made to the PSRSC which shall include an analysis of the cause of the occurrence, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence.
- e. A report shall be made to the USNRC in accordance with Section 6.7 or these specifications.
6.5 OPERATING PROCEDURES Written operating procedures shall be adequate to assure the safe operation of the reactor, but shall not preclude the use of independent judgement and action should the situation require such.
Operating procedures shall be in effect for the following items:
- a. Evacuation
- b. Loss of pool water
- c. Gaseoua release
(} d. Reactor operating procedure
()
34 I
- e. Daily checkout procedure
- f. Core loading and fuel handling
- g. Radiation, evacuation and alarm chseks
- h. Experiment evaluation and authorization Changes to the procedures that do not change their original intent may be made by the Director or a designated alternate.
6.6 FACILITY OPERATING RECORDS To fulfill the requirements of applicable regulations, records and logs shall be prepared of at least the following items and retained for a period of at least five years for items a through g and indefinitely for items h through k.
- a. Log of reactor operation.
- b. Checks and calibrations procedure file.
- c. Electronic maintenance log.
- d. Experiment authorization file.
- e. Event evaluation forms.
- f. Maintenance records of associated reactor equipment.
- g. Facility radiation and contamination surveys.
- h. Radiation exposure for all f acility personnel.
- 4. Fuel inventories and transfers.
J. Environmental suaveys,
- k. Liquid radioactive waste released to the environs.
For th3 past several years, the radioactive liquid release from the facility has been near zoro. The only routine gaseous releases from the facility are argon-41 and tritium released from the water into the reactor room. These releases have been measured to be a function of the reactor power history. Since the dilution factor from the bay to the outside environs is 7 x 10-3, the gaseous radiation release is negligible and calculated based on the power history of the reactor. All pneumatio systems transporting samples to and from the reactor core operate with non-activatable gases (N2 and CO 2 ) and release no significant gaseous radioactive products to the room during their operation.
O i
l
35 6.7 REPORTING REQUIREMENTS To fulfill the requirements of applicable regulations, reports shall be made to the NRC Region I, Office of Inspection and Enforcement as follows:
- a. A report within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and telegraph of:
(1) Any accidental release of radioactivity above permissible limits in unrestricted areas whether or not the release resulted in property damage, personal injury, or exposure.
(2) A violation of the safety limit (fuel temperature >
1150*C).
(3) Any reportable occurrences as defined in Section 1.1 of these specifications,
- b. A report within 10 days in writing of:
(1) Any accidental release of radioactivity above permissible limits in unrestricted areas whether or not the release resulted in property damage, personal injury, or exposure.
The written report (and, to the extent possible, the preliminary telephone or telegraph report) shall describe,
, analyze, and evaluate safety implications, and outline the
( i .
corrective measures taken or planned to prevent
\ recurrence of the event.
(2) A violation of the safety limit. 1 (3) Any reportable abnormal occurrence as defined in Section 1.1 of these specifications.
, c. A report within 30 days in writing of:
)
(1) Any significant variation of measured values from a corresponding predicted or previously measured value of safety-connected operating characteristics occurring during operation of the reactor.
(2) Any significant change in the transient or accident analysis as described in the Safety Evaluation Report.
(3) Any changes in facility Director or Deputy Director.
(4) Any observed inadequacies in the implementation of administrative or procedural controls,
- d. An annual report covering the operation of the unit during the previous fiscal year will be submitted prior to September 15 of each year, I
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- e. Routine operating reports on radioactive waste as required by the state and federal regulatory agencies, l
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ENVIRONMENTAL IMPACT APPRAISAL d Table of Contents 1.0 Facility, Environmental Effects of Construction ..................... 1 2.0 Environmental Effects of Facility Operation ......................... 1 2.1 Thermal Discharges ............................................. 1 2.2 Radioactive Discharges ......................................... 3 i
2.2.1 Argon-41 ................................................ 3 2.2.2 Tritium ................................................. 3 2.2 3 Nitrogen-16 ............................................. 4 2.2.4 Radioactive Waste ....................................... 5 23 ALARA .......................................................... 5 2.3.1 Personnel Training ...................................... 6 2.3 2 Administrative Policy ................................... 6 2.4 Radiation Control .............................................. 7 I 2.4.1 Environmental Radiation Monitoring ...................... 7 2.4.2 Fixed Radiation Monitoring System ....................... 7
\s 2.4.3 Por ta ble Survey Ins tr ument s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 2.4.4 Personnel Monitoring .................................... 9 30 Environmental Effec'ts of Accidents .................................. 9 j 4.0 Unavoidable Ef fects of Facility Construction and Operation . . . . . . . . . .
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5.0 Alternatives to Ccustruction and Operation of the Facility .......... 10 6.0 Long-Term Effects. Ccsts and Benefits, and Alternatives of Facility Construction and Operation .......................................... 10 1
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ENVIRONMENTAL IMPACT APPRAISAL
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e 1.0 Facility, Environmental Effects of Construction This Environmental Impact Appraisal (EIA) is submitted for the continued operation of The Penn State Breazeale Reactor (PSBR) under the license R-2. Because the PSBR has a power level less than 2 MW(th), an Environmental Impact Statement (EIS) is not needed.
-The PSBR core is a mixture of 8.5 wt5 and 12 wt5 standard TRIGA fuel
- and is supported on a movable bridge structure in an open concrete pool of water containing 71,000 gallons of domineralized water.
The building housing the reactor is a standard industrial-type steel building. When the evacuation alarm sounds, the normal roof fan exhaust system closes and an emergency filtered exhaust system is activated automatically.
Supporting facilities include a Cobalt-60 irradiation facility, two hot. cells, various laboratories, evaporator building, pool water storage tank, and liquid waste storage tanks.
No plans for construction exist that would affect the environment, since the reactor is located in an existing building at a developed site.
- ' 2.0 Environmental Effects of Facility Operation
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2 .1 Thermal Discharges ,
The PSBR heat: exchanger limits the temperature of the PSBR pool
' water. Maintaining a low pool temperature decreases pool water evaporation losses and a temperature below 110*F is needed to !
prevent deterioration of-the domineralizer anion resins.
The system is comprised of two loops. In the primary loop, pool water _is pumped through the baffled shell side of two double'
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-pass heat exchangers connected in series. In the secondary loop, I cooling water is pumped from Thompson Pond (650 yards from the PSBR)
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. through the tube bundle side of the two heat exchangers and then to l
a storm sewer which returns the water to Thompson Pond. l Relatively small variations in cooling water -temperatures are 7 h lnoted (55' + 2*F) but other factors such as flow rates, pool-
2 temperature, and system cleanliness affect the system's heat removal capacity.
A check on the efficiency of the heat exchanger was performed in August 1984. During this check, the pool recirculation system was secured and two mixing pumps were installed on the pool divider wall to provide a more uniform pool temperature. The pool was cooled from 99'F to 73 5"F and periodic measurements were taken of the heat exchanger pool in and pool out temperatures. Calculations show that with a pool inlet temperature of 99'F the heat exchanger is removing 1 MW of heat.
Thompson Pond is fed by a spring whose output averages 4.5 x 106 gal / day on a yearly basis, with a dry season low of 3 x 106 gal / day and a rainy season high of 7 x 106 gal / day. Assuming a 1 MW heat input from the heat exchanger, the temperature rise of the 4.5 x 106 gal / day would be 2.2*F (3 3*F in the dry season and 1.4*F in the wet season). However, the reactor pool seldom reaches a temperature as high as 99'F where the heat exchanger heat removal capacity is 1 MW. Records for the 1983 calendar year show the highest recorded pool temperature as 80.6*F, resulting in a heat removal rate of 0.6 MW. So in the worst case dry season, the temperature rise of the 3 x 106 gal / day would be about 2*F at a pool temperature of 80.6*F.
The above discussion assumes that the heat exchanger effluent contributes its heat directly to Thompson Pond. As earlier noted, the heat exchanger effluent travels to Thompson Pond by way of a storm sewer where some heat loss would take place making the calculated temperature rise even less than calculated.
l PSBR thermal discharge presents no significant hazard to the environment.
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2.2.1 Argon-41 Argon-41 is produced by the thermal neutron activation of argon-40, which comprises about 1% of air. The major source of argon-41 released from the PSBR is the air dissolved in the pool water. As the water passes through the core the dissolved air containing argon is irradiated, producing argon-41. At the same time the water is heated, reducing the solubility of the gases, and gas bubbles are formed and rise to the surface releasing argon-41. The argon-41 mixes with the air in the reactor bay and is released through the exhaust system.
Monitoring of the reactor bay exhaust for argon-41 has shown an average production of 1.4 + .5 mci of argon-41 par HW-hr of reactor operation. The amount released is very dependent upon the pool water temperature and convection currents, therefore it varies with the power level and
., diffuser operation. The irradiation of the air in dry tubes
( and air which leaks into the carbon dioxide and nitrogen in the pneumatic sample changers also produces small amounts of argon-41.
Because argon is a noble gas, there is essentially no uptake in the body and the exposure limits are based upon the external dose to the body. Thus, external radiation monitors include the exposure to argon-41 in their readings along with other gamma, x-ray and beta radiation sources. Over the past several years the annual argon-41 releases from the PSBR are estimated to be about 600 mci. The calculated dose from this release at ground level outside the reactor building is estimated to be less then 1 mrem / year. The inert nature of the gas and the short 108 minute half-life of this material insure against any accumulation in the environment.
2.2.2 Tritium Tritium is found in the reactor pool at a concentration !
of about 2 x 10-4 pCi/ml. There are several possible sources l CN of the tritium, including neutron activation of the naturally h
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occurring deuterium in the water, leakage of tritium contaminated heavy water from the thermal column, and leakage of fission product tritium from the reactor fuel. The tritium concentration in the pool has changed significantly over the years since replacement of the original aluminum clad MTR reactor core with the TRIGA core, decreasing from 8 x 10-4 pCi/ml in 1972 to about 2 x 10-4 pCi/ml at present.
This indicates that the primary source of tritium was probably diffusion of fission product tritium from the MTR fuel into the pool. The MTR fuel was used from 1955-1965.
The evaporation of pool water puts tritiated water vapor into the reactor bay air. The air is then released through the exhaust fans. For an average pool water evaporation rate of 25 gallons / day with a tritium concentration of 2 x 10-4 pCi/ml and an exhaust rate of 2000 cfm, the tritium concentration in the exhaust would be 2 x 10-10 pCi/ml. This is 0.1% of the MPC for unrestricted areas for tritiated water vapor (2 x 10-7 pCi/ml) at the point of release. The concentration at ground level would be about 2 x 10-12 pC1/ml assuming a dilution of 100 m3/s. The release is further reduced by the return of dehumidifier condensate from the bay air conditioning system to the reactor pool.
2.2.3 Nitrogen-16 During 1 MW(th) steady-state operation, the gamma level directly above the reactor core near the pool surface is 25 mR/hr. Just over the edge of the pool wall, the maximum gamma level is 15 mR/hr. Neither of these areas are ones whore personnel spend much time. The radiation levels on the i reactor bay floor areas at shoulder level range from 0.2 to 1.5 mR/hr. Control room radiation levels are much less than 1 mR/hr.
The main source of the radiation in the reactor bay is the production of radioactive nitrogen-16 from the action of fast neutrons on the oxygen in the pool water. The problem is minimized by a diffuser pump that prolongs the amount of time for the nitrogen-16 to reach the surface, thus allowing
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<~s for decay of much of the nitrogen-16 (7.14 second half-life).
\s_) This short half-life also makes the' release of nitrogen-16 to the environment negligible.
2.2.4' Radioactive Waste Radioactive material produced by the PSBR comes under the control of the University Isotopes Committee and the broad byproduct material license number 37-185-4 upon removal from the reactor pool. This includes samples and activated components and equipment, excluding fuel elements.
Radioactive waste from these materials is disposed of by the Health Physics office in the same manner as radioactive waste from other campus laboratories. The disposal techniques include shipment to licensed disposal facilities, storage of short-lived radioisotopes for decay, and release of low-concentration liquid and gaseous material. Volume reduction techniques include compaction, incineration, and
,s evaporation.
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The largest volume of liquid waste from the PSBR is from the regeneration of the PSBR pool recirculation system demineralizer. This liquid, plus waste from some floor drains and pump gland leakage, is collected in holdup tanks.
Although the liquid could be released with little or no dilution, most of it is evaporated and the distillate used as makeup water for the reactor pool. The evaporator bottoms are solidified for disposal. The remaining solid dry waste is compacted into 55 gallon drums and shipped to commercial burial sites for disposal. Approximately 6 fif ty-five gallon drums are collected for disposal each year.
2 3 ALARA It is the policy of the University that environmental releases of radioactive material and exposure of individuals to ionizing radiation be kept as far below regulatory limits as is reasonably achievable (reference: Rules and Procedures for the Use of O
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6 Radioactive Material at The Pennsylvania State University by The University Isotopes Committee - Jie f 1980).
2.3 1 Personnel Training PSBR edministrative procedures call for all students, staff, f aculty, or other users working independently in the PSBR f acility to participate in a health physics orientation program. Participants must pass a written exam at the end of the program.
Before individuals can receive radioisotopes released from the reactor pool, they or their supervisor must possess an authorization issued by the University Isotopes Committee.
Authorization approval requires the completion of the above-mentioned health physics orientation program.
Authorized limits of the isotopes and their quantity depend upon the individual's qualifications and experience in handling radioactive materials as determined by the University Isotopes Committee.
Proper training of personnel in health physics procedures and licensing of individuals to possess radioisotopes are effective ways of minimizing personnel exposure within the facility and limiting the release of radioactive materials to the environment.
2.3 2 Administrative Policy Where practical, steps are taken to minimize radiation levels in both restricted and unrestricted areas. Several years ago, a pneumatic transfer system used to transport samples between laboratories and the reactor core was converted from air to CO2 as the working fluid to minimize Argon-41 production and release to the environment. Recently a newly designed nozzle on the Nitrogen-16 diffuser pump caused a decrease in radiation levels due to Nitrogen-16 in the reactor bay. Floor drains have been clearly marked as to their destination to minimize chances of accidentally releasing radioactivity to the environment.
The Health Physics office uses appropriate methods to limit radiation workers internal and external exposure to I
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7 s radiation. Experimenters using loose radioactive material or
( performing beashole laboratory irradiations work under guidelines developed by the Health Physics office. Public areas and l'aboratories are monitored for transferrable
]' contamination by periodic smear surveys. Monitoring for airborne activity is used if needed. Bioassays are used in appropriate ' situations. Radioisotopes Laboratory Rules are posted in all areas where radioactive materials are used.
Disposal of solid and liquid radioactive wastes are handled under the strict procedures of the Health Physics office to comply with 10CFR20 reCulations.
2.4 Radiation Control l The r'adiation monitoring devices at the PSBR help to ensure compliance with the radiation limits in 10CFR20 and help to keep the radiation exposure of individuals in both restricted and unrestricted areas as low as reasonably achievable.
2.4.1 Environmental Radiation Monitoring An environmental radiation monitoring program is
,,_ ) . conducted continuously by the Health Physics Office.
Integrated radiation measurements are made for successive 90-day periods using thermoluminescent dosimeters (TLD's).
2.4.2 Fixed Radiation Monitoring System Thefprincipal~PSBR fixed' radiation monitoring system consists of six devices. Ionization chambers on the east e and west sides of the reactor bridge have alarms at 30 mR/hr and 200 mR/hr,-respectively.- Geiger-Mueller detectors are located in the reactor beashole laboratory and in the Cobalt facility . and both devices alars at 6 mR/hr. Particulate air monitors are located along the east and west walls of the
- - reactor bay.. Both monitors are assembled from various manufacturer's components and use Geiger-Mueller detectors.
. Both monitors alarm at 5000 CPM, a setting determined by 10CFR20 MPC considerations and detector efficiencies. All six devices provide readouts to' modules located in the !
I reactor control room. A building evacuation alarm and j- reactor scram occur when any monitor reaches its alarm point.
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8 The reactor demineralizer room, hot cell clean and control areas, and the Cobalt facility basement have Geiger-Mueller detectors which provide local alarms at 5 mR/hr. In addition, the Cobalt f acility basement monitor provides an alarm to the Cobalt bay and the demineralizer room provides an alarm to the enunciator panel in the reactor control room.
The lobby, lunch room, and various laboratories are equipped with rate meters with appropriate Geiger-Mueller probes.
All f acility fixed monitors are checked for proper operation weekly and calibrated as a minimum annually.
2.4.3 Portable Survey Instruments A variety of portable survey instruments are available for general radiation surveys and emergency situations. A typical inventory is listed below.
Instrument Range Location Victoreen 440-IC 0-300 mR/hr Reactor Bay Victoreen 440-IC 0-300 mR/hr Reactor Bay Eberline RO-2-IC 0-5 R/hr Reactor Bay Eberline R0-2-IC 0-5 R/hr Reactor Bay Eberline R0-2-IC 0-5 R/hr Reactor Bay Eberline E510G-GM 0-1 R/hr Cobalt Bay Eberline E500B-GM 0-2 R/hr Emergency Case Eberline E520-GM 0-2 R/hr Emergency Case Eberline E520-GM 0-2 R/hr Emergency Support Cntr.
0-250,000 cpm Eberline E120(4)-GM 0-50 mR/hr Health Physics Office at 0-50,000 cpm PSBR and Room 120 O
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9 Instrument Range Location Dosimeters (2) 0-100 R/hr Emergency Case (2) 0-200 mR/hr Emergency Case (9) 0-200 mR/hr Lobby (9) 0-200 mR/hr Room 120 (2) 0-100 & 0-600 mR/hr Reactor Bay These instruments are checked for proper operation weekly and calibrated as a minimum annually.
A variety of similar portable survey instruments are also available in the Academic Projects Building at the Health Physics office.
2.4.4 Personnel Monitoring Faculty, staff, students, and other facility users are issued either film badges or thermoluminescent dosimeters by the Health Physics office to monitor their radiation exposure. Other facility visitors such as
, public tours, are issued pocket dosimeters.
-3 0 Enviromental Effects of Accidents Accidents ranging from fail"re of experiments to core damage and fission product release are considered in the PSBR Safety Analysis Report. Effects are considered negligible with respect to the environment.
.4.0 Unavoidable Effects of Facility Construction and Operation The unavoidable effects of construction and operation of the facility involve the materials used in construction that cannot be recovered and the fissionable material used in the reactor. No adverse impact on the environment is expected from either of the unavoidable effects.
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10 5.0 Alternatives to Construction and Operation of the Facility Some of the activities conducted at the PSBR could be done by using particle accelerators, radioactive sources, or other means; but these alternatives are more costly and less efficient. Much of the research and educational activities using the PSBR can not be done by other suitable or economic means. Therefore, there is no reasonable alternative to a research reactor such as the PSBR for conducting the wide variety of research and education activities presently done.
6.0 Long-Term Effects, Costs and Benefits, and Alternatives of Facility Construction and Operation Since the facility is an existing one, the capital costs are low.
Environmental impact due to the facility's operation is minimal, and no long-term environmental damage is anticipated.
Beneficial long-term effects are numerous and include (1) the education of students, the public, and power plant personnel; and (2) research activities, including neutron radiography, activation analysis, and isotope production, which serve the national interest in areas of health, nuclear power production, national defense, and many i other areas.
No reasonable alternatives exist to the wide versatility of research reactors such as the PSBR in contributing to education and scientific knowledge.
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OPERATOR AND SENIOR OPERATOR REQUALIFICATION
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Table of Contents A. PURPOSE .............................................................. 1 B. ADMINISTRATION ....................................................... 1 C. WRITTEN EXAMINATION .................................................. 1 D. ORAL EXAMINATIONS .................................................... 2 E. ON-THE-JOB TRAINING .................................................. 2 F. EVALUATION OF OPERATING PERFORMANCE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 G. LICENSEE REQUALIFICATION REQUIREMENTS FOLLOWING AN ABSENCE FROM DUTY OF MORE THAN ONE C ALEND AR QUARTER . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 H. RECORDS .............................................................. 3 I. REFERENCE ............................................................ 4 l P
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OPERATOR AND SENIOR OPERATOR REQUALIFICATION s_- 1 l
1 A. PURPOSE To assure that all operators and senior operators maintain competence.
B. ADMINISTRATION The Director or Deputy Director shall appoint a facility tra.aing officer for each two year requalification program. The training officer shall be responsible for conducting the requalification program, shall write and/or edit all exams, correct all exams, assure that the examinations are not compromised, and chall maintain records required by the program. The training officer may ask other staff members to assist in the preparation of the written exam in areas of their expertise. No more than the training officer and one other person shall prepare and be exempt from the written exam in any one subject area. The training gs s officer shall conduct or ask other staff members to conduct the two
( ) required oral exams. Only the person conducting the oral exam shall be exempt from it.
The appointment of the training officer and assignment of other persons to conduct the oral exams and portions of the written exam shall rotate so that no examining responsibility is duplicated during successive two year requalification programs.
C. WRITTEN EXAMINATION
- 1. During a period not to exceed two calendar years, a comprehensive written exam shall be given in the subject areas listed below:
- a. Principles of Reactor Operation
- b. Features of Facility Design
- c. General Operating Characteristics
- d. Instrumentation and Control
- e. Safety and Emergency Systems
- f. Procedures, Technical Specifications, and Government Regulations
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- g. Radiation Control and Safety
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- 2. A licensee receiving a grade of less than 70% in any subject shall be given instruction in that subject until the licensee can pass an examination in that subject with a grade of 70% or more.
3 A licensee receiving a grade of less than 70% (overall) shall, in the next two working weeks, be given an accelerated program to remedy deficiencies. At the end of this period, the licensee shall be given an oral examination. If the results of this examination are negative, removal from licensed duties shall occur until the requirements of C2 are met. If the oral examination results are satisfactory, the licensee may continue performing licensed duties while receiving additional instruction until the requi snts of C2 are met.
D. ORAL EXAMINATIONS
- 1. During each two year period, a " walk around" type examination shall be given to each licensee to ensure familiarf t3 with facility changes.
- 2. During each calendar year, each licensee is expected to revi o all PSBR procedures, including abnormal and emergency procedures.
Subsequent to this, an oral examination shall be given to ensure familiarity with these procedures.
3 The oral examinations shall be graded on a pass-fail basis. If a negative result is obtained, the next two working weeks shall be devoted to an accelerated remedial program. A retest shall be given at the end of this period. If the results are still negative, the licensee shall be removed from licensed duties until the deficiency is remedied.
E. ON-THE-JOB TRAINING
- 1. Each licensed senior operator shall make or supervise at least one reactivity change each calendar quarter and at least ten reactivity changes during the two year requalification program. Each licensed operator shall make at least one reactivity change each calendar quarter and at least ten reactivity changes during the two year requalification program.
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- 2. Meetings shall be held as deemed necessary at which facility design, license, and procedure changes are explained and discussed.
3 Written material concerning facility design, license, and procedure changes shall be distributed and/or circulated.
F. EVAi-UATION OF OPERATING PERFORMANCE
- 1. The Director, Deputy Director, or training officer shall evaluate operator and senior operator control console performance once during each two year period.
- 2. A performance checklist shall be filled out by the evaluator for all licensees and graded on a pass-fail basis.
3 In case of a failure, the licensee shall be removed from licensed duties, shall be given instruction at the control console in areas of deficiency, and shall be returned to duty upon permission of the evaluator.
4 The evaluator shall be exempt from this exam.
G. LICENSEE REQUALIFICATION REQUIRDIENTS FOLLOWING AN ABSENCE FROM DUTY OF
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' MORE THAN ONE CALENDAR QUARTER (NRC Reinstatement ' Required)
- 1. The licensee shall meet any reactivity requirement deficiencies of E1.
- 2. The licensee shall be evaluated at the control console as per F.
-3 The licensee shall pass any written or oral exams as per C and D that were given during the absence from duty.
- 4. The licensee shall be informed of any facility design, licent.e, or procedure changes during the absence.
H. RECORDS
- 1. Records of the requalification program shall be maintained for each complete two year program until the completion of the following complete two year program.
- 2. The records shall contain copies of written exams administered, the answers given by the licensees, and documentation and results of additional training and examinations in areas where licensee deficiencies were noted.
3 Console performance checklists shall be maintained in each licensee's file.
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4 Reactivity manipulation checklists shall be maintained in each licensee's file.
- 5. Records shall be maintained for actions taken to requalify licensees who have been absent from duty as described in G.
I. REFERENCE ANSI /ANS-15.4-1977 Selection and Training of Personnel for Research Reactors O
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'Special Nuclear-Materials and Neutron Source Requirements 1.
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- Special Nuclear Materials Requirements .................................... 1 i Neutron Source' Requirements ............................................... 2 ;
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' Special Nuclear Materials and Neutron Source Requirments
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for the Penn State Breazeale Reactor The quantity of Special Nuclear Material requested to be licensed under this application is the lowest acceptable quantity necessary to sustain current and projected operation of the PSBR.
The PSER operates using standard 8.5 wt5 and 12 wt1 TRIGA fuel enriched to less than 20% in the isotope uranium-235. This SNM is of low strategic significance as defined in 10 CFR 73 2.(y).(2), and is exempt from certain physical protection requirements pursuant to 10 CFR 73.6.(a).
The MTR type fuel - 935 enriched in the isotope uranium-235 possessed by the PSBR is SNM of low strategic significance as defined in 10 CFR 73 2.(y).(1),
and is protected by the PSBR Physical Security Plan as required by 10 CFR 73.67.(f).
We specifically request that this license permit The Pennsylvania State
,/~'\ University to receive, possess and use up to nine (9) kilograms of contained
'- uranium-235 in TRIGA fuel in connection with operation of the reactor and 0.900 kilograms of contained uranium-235 in MTR type fuel elements. Current inventory of Special Nuclear Materials, which should be licensed under this R-2 license, follows:
-PSBR Fuel Element Inventory Number of Elements TRIGA Fuel Location 8.5 wt% 12 wt% MTR Fuel In Core 81 14 -
In pool racks 49 20 -
In locked safe -- --
7 Totals 130 34 7 A
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PSBR Fuel-Uranium Content (grams)
Element (U) Isotope (235u)
TRIGA Fuel 33,414 6,150 MTR Fuel 716 668 Totals (as of 6/30/84) 34,130 6,818 Other special nuclear material; contained in fission counters, fission foils, neutron sources, etc., required to sustain operation of the PSBR are authorized by license number SNM-95 issued June 18, 1981 to The Pennsylvania State University.
NEUTRON SOURCE REQUIREMENTS We hereby request that pursuant to title 10, CFR, Chapter I, Part 30,
" Licensing of Byproduct Material," that this license permit The Pennsylvania State University to receive, possess, and use two (2) fif ty (50) curie sealed antimony-beryllium neutron sources, either or both of which may be used for reactor start-up, or to use a 0.235 milligram Californium-252 neutron source for operation of the reactor and to possess, but not to separate such byproduct material as may be produced by operation of the reactor.