ML083510754
| ML083510754 | |
| Person / Time | |
|---|---|
| Site: | National Bureau of Standards Reactor |
| Issue date: | 09/12/2008 |
| From: | Richards W US Dept of Commerce, National Institute of Standards & Technology (NIST) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TAC MD3410 | |
| Download: ML083510754 (74) | |
Text
UNITED STATES DEPARTMENT OF COMMERCE National Institute of Standards and Technology Gaithersburg, Maryland 20899-Documet Control Desk September 12,2008 U.S. Nuclear Regulatory Commisson Washington, D.C. 20555
Subject:
NRC Requed for Additional Infomation (TAC No. MD3410) dated July 1,2008.
Docket No. 50.184 Please find the attached rep*onse to NRC Request for Additional Information (TAC No. MD3410), t*e revised NBSR Technical Specifications and the NBSR Plm Plan.m Quetions conc i we reponss should be *ected to Dr. Wade J.
Richads, Chief Reactor Opavtions and (301-975*6260) or
- Sincady, Wade J. Richards Chief Reactor Operations and NIST Center for Neutron Reseauh I certify under pamlty of perjury that the following is true and correct.
Executed on:_____________-Cb t by.-U CC.
Mr. William B. Kenn U.S. Nuclear Regulatory Commission MS 012-015 Washington, D.C. 20555 N LST Requalification Program for the NBSR REQUALIFICATION PROGRAM FOR THE NBSR LICENSE TR-5 DECEMBER 2008
TABLE OF CONTENTS 1.0 INT R O D UC TIO N............................. ............. ............................................. 1 2 .0 D E F INIT IO NS ............... ................................................................................ .. 1 3 .0 G E NE RA L .................................................. .......................... ................................. 2 3.1 Administration ............................................... 2 3.2 Description............................................... ...2 3.2.1 Refresher Training .................................... ....... )2 3.2.2 W ritten Exam ination ................................................................................... 2 3.2.2.1 Examination Administration and Evaluation ............................................ 3 3.2.2.2 Evaluation and Retraining ..................................................................... 3 3.2.3 Medical Exam ination ............................................................ ............. 3 3.2.3.1 :G eneral Requirem ents .......................................................................... 4 3.2.3.2 Disqualifying Conditions ......... ;.......................... 4 3.2.3.3 Specific Minimum Capacities Required for Medical Qualification ....... 5 3.2.3.4 Additional Examination. .................................... 6 3.2.4 Reactivity Control Manipulations. ................................ 6 3.2.5 Operating Test or Evaluation ................................... 7 3.2.5.1 Examples of tasks to be performed under normal and abnormal circum stances ............................................................................................. . .7 3.2.6 D ocum ent Review .......................................................................................... 7 3.2 .7 R elicensing .......................................................................................... ... . 7 3.2.8 Absence from Licensed Functions ....................... ... .. ... 7 3.2.9 Exemptions. ................... ......................... ...8 4.0 DOCUMENTATION AND RECORDS ................................... 8 4 .1 D ocum e ntation .............................................................................................. . . 8 4 .2 R e co rd s ................................................ .............................................. . . 8
1.0 INTRODUCTION
The NBSR operator requalification program is designed to provide refresher training to the licensed operator in areas of infrequent operation, to review facility and procedural changes, to address subject manner not reinforced by direct use, and to improve in areas
.ofperformance weakness. The program is designed to evaluate an operator's knowledge and proficiency to performntheir duties and to retrain where necessary. Emphasis is placed on those subjects necessary for the continued proficiency. Successful completion of the program is required for the operator to continue licensed activities. The program conforms to the applicable content of ANSI/ANS-15.4-2007, "Selection and Training of Personnel for Research Reactors."
2.0 DEFINITIONS controls. When used with respect to a nuclear reactor, means apparatus and mechanisms the manipulation of which directly affects the reactivity or power level of the reactor.
designated medical examiner. A licensed medical practitioner familiar with the medical provisions of this standard.
disqualifying or disqualifying conditions. Something which precludes unconditional medical approval for research reactor operator licensing.
license. The written authorization; by the U.S. Nuclear Regulatory Commission (NRC),
for an individual to carry out the duties and responsibilities associated with a position requiring licensing.
licensed. See licensee.
licensee. An individual or organization holding a license.
licensing. The confirmation by the NRC of the experience, education, medical condition, training, and testing pertinent to a specific job assignment.
on-the-job training. A systematic, structured method using a qualified person to provide the required job-related knowledge and skills to a trainee, usually in the actual work place, with proficiency documented.
reactor operator. An individual who is licensed to manipulate the controls of a reactor.
reactor supervisor. An individual as described in the NBSR Technical Specifications and responsible for an operating crew or shift.
research reactor. A research reactor is defined as a device designed to support a self-sustaining neutron chain reaction for research, developmental, educational, training or experimental purposes, and which may have provisions for the production of radioisotopes.
research reactor facility. Those facility-designated areas within which the owner or operator directs authorized activities associated with the reactor.
senior reactor operator. An individual who is licensed to direct the activities of reactor operators. Such an individual is also a reactor operator.
1
shall, should, and may. The word "shall" is used to denote a requirement; the word "should" to denote a recommendation; and the word "may" to denote permission, neither a requirement nor a recommendation.
significant power change. A tenfold increase in flux indication above critical or a change in thermal power of at least 1 MW.
solo operation. Operation of the controls, including monitoring of instrumentation, during reactor operation with no other person in the vicinity of the controls.
3.0 GENERAL 3.1 Administration. Responsibility for the administration of the requalification program rests with the Chief, Reactor Operations. The program shall be administered over a normal period of 24 months, to be followed by successive 24 month periods, with no period to exceed 30 months. The maximum period is to provide administrative flexibility and is not to be used to reduce frequency; the 24 month period shall be maintained over the long term.
3.2 Description. During the requalification period, the following shall be provided or accomplished:
(a) Refresher training (b) Written examination (c) Medical evaluation (d) Reactivity manipulations (e) Operating test or evaluation (f) Document Review 3.2.1 Refresher Training. This training shall be provided in critical areas not routinely used by the operator such as emergency planning, response to abnormal conditions, selected topics in radiation protection and reactor operation principles, and changes to facility design and procedures.
3.2.2 Written Examination. Written examinations shall be operationally oriented, practical, and objective. The interval between any two successive exams shall not exceed 30 months. The number and form of questions shall be selected to best evaluate a particular examinee, but the number of questions successfully answered shall not preclude attainment of the minimum acceptance scor6 specified in section 3.2.2.1. The exam
- should be of a multiple choice type, composed of multiple categories, with 20 questions per category, and with four answers per question. The following categories should comprise the exam:
(a) Theory. Topics include nuclear theory, principles of reactor operations, general and specific facility operating characteristics, and-applicable thermodynamics.
(b) Procedures and Radiological Controls. Topics include normal procedures, abnormal procedures, emergency procedures, radiation protection principles and procedures, administrative rules, and technical specifications.
2
(c) Systems. Topics include plant systems, radiation protection systems, instrumentation and controls, and facility protection and engineered safety features.
3.2.2.1 Examination Administration and Evaluation. The minimum acceptance score shall be 70% for each category of the written examination. Individuals who did not achieve passing scores in one or more of the categories listed in section 3.2.2 may be re-examined following retraining in the deficient areas. The Chief, Reactor Operations may waive re-examination in categories with passing scores providing the candidate has demonstrated proficiency in those portions of an examination.
3.2.2.2 Evaluation and Retraining. Additional requalification training in the form of formal lectures, tutoring, self-study or on-the-job training shall be based on the results of the requalification examination. The following considerations should be used:
(a) A score on the written examination equal to or greater than the acceptance criterion may require no additional training.
(b) A score on the written examination below the acceptance criteria in section 3.2.2.1 shall require additional training in those topics where weakness or deficiencies are indicated. This retraining and retesting shall be completed prior to the candidate being relicensed.
(c) An overall score on the written examination of less than 60% shall require that an evaluation by Chief, Reactor Operations or designated representative be performed. The evaluation shall determine if the deficiencies require that the indiVidual's license be withdrawn pending completion of an accelerated retraining effort. The evaluation shall take into account the individual's past performance record, supervisor's evaluation and past test scores, as well as current deficiencies. Additional oral or operational examinations may also be given to aid in the evaluation. In any case, the individual shall be removed from licensed activities within four months if the candidate cannot achieve passing scores after re-examination.
(d) Regardless of the score, if the evaluation indicates a deficiency in a critical area that affects safety, training shall be administered to promptly correct the critical deficiency.
3.2.3 Medical Examination. Each licensed individual shall undergo medical examination and evaluation as part of the requalification program and shall meet the requirements of section 3.2.3.1. The primary responsibility for assuring that qualified personnel are on-duty rests with the Chief, Reactor Operations. The health requirements set forth herein shall be considered to determine the physical condition and general health of the individual in order to perform certain assigned duties as determined by the Chief, Reactor Operations. Each requirement should be considered in the context of the certain assigned duties of the individual as related to the consequences of health-induced operational errors endangering public health and safety. The designated medical examiner shall be conversant with the requirements.
3
a The interval between any two successive medical evaluations shall not exceed 30 months.
More frequent examinations may be required if conditions warrant as determined by the Chief, Reactor Operations or upon the recommendation of the designated medical examiner. The physical condition and the general health of an operator shall be such that they are capable of properly carrying out licensed activities under normal, abnormal and emergency conditions and able to perform the associated tasks. Conditions that can cause sudden incapacitation such as coronary heart disease, stroke, epilepsy, mental disorder, diabetes, fainting spells, impaired hearing, or vision, and effects of medication, shall be considered. Many of the conditions indicated above may be accommodated by restricting the activities of the individual, requiring close surveillance of the condition, imposing a medical regime, or requiring a second individual to be present when the individual in question is performing certain assigned duties. As a minimum, the second individual shall be able to shut down the reactor and summon competent help.
3.2.3.1 General Requirements (a) Capacity. The examinee shall demonstrate stability and capacity for all of the following:
(1) Mental alertness and emotional stability; (2) Acuity of senses and ability of expression to allow accurate communications by spoken, written, or other audible, visible, or tactile signals; (3) Stamina, motor power, range of motion, and dexterity as needed to allow ready access to and safe execution of certain assigned duties.
(b) Freedom from incapacity. The examinee shall be free of any of the following conditions that are considered by the designated medical examiner and the Chief, Reactor Operations as predisposing'to incapacity for duty:
(1) Mental or physical impairments; (2) Any medical, surgical, or other professional treatment; (3) Any condition, habit, or practice which might result in sudden or unexpected incapacitation.
3.2.3.2 Disqualifying Conditions. The presence of any of the following conditions, that have a high probability of sudden or unexpected incapacitation, unless adequately compensated shall disqualify the individual for unsupervised operation except as noted: Laboratory tests such as ECG, blood and urinalysis, x-rays and other tests should be used to rule out disqualifying conditions identified in this section.
(a) Respiratory Condition (1) Frequent severe uncontrolled attacks of asthma within the previous two years.
(2) Tracheostomy or laryngectomy if they severely impair speech or cause shortness of breath.
(3) Severe chronic pulmonary disease.
(b) Cardiovascular Condition (1) Ischemic heart disease, myocardial infraction, coronary insufficiency or angina pectoris unless thorough history, physical examination, 4
electrocardiogram (ECG), and other test procedures indicate satisfactory cardiac function and reserve.
(2) Heart failures.
(3) Arrhythmia other than benign extra systoles.
(4) Valve replacement.
(5) Pacemaker.
(6) Implantable defibrillator.
(7) Peripheral vascular insufficiency.
(8) Arterial aneurysm.
(c) Endocrine, Nutritional, Metabolic Conditions (1) Diabetes mellitus. Uncontrolled diabetes, ketoacidosis, diabetic coma,.
or insulin shock within the previous two years.
(i) Stable diabetics adequately controlled by diet or oral medication may be qualified for solo operation.
(ii) Insulin dependent stable diabetics may also be qualified for solo operation providing adequate provisions are made to guard against insulin shock as certified by the designated medical examiner.
(d) Neurological Condition (1) History of epilepsy, unless the examinee has remained seizure-free for at least the previous five years with medication or has remained seizure-free during the previous two years without medication.
(e) Mental Condition. An established history or clinical diagnosis of any of the following:
(1) Any psychological or mental condition that could cause impaired alertness, judgment or motor ability. Clinically significant emotional or behavioral problems shall require thorough clinical evaluationthat may include psychological testing and psychiatric evaluation.
(2) A personality disorder that is severe enough .to have repeatedly manifested itself by overt bizarre, disruptive or similar acts, unless the condition has been relieved and certified. Otherwise the disorder shall be disqualifying for all operations.
(3) History or threat of suicide attempt shall be disqualifying for all operations.
(4) History of a psychotic disorder shall be disqualifying for all operations.
(5) Alcohol abuse or dependence, unless treated and corrected, shall be disqualifying for all operations.
(6) Abuse of drugs other than alcohol, tobacco, or ordinary caffeine containing beverages, as evidenced by non-prescribed habitual use of the drug, unless the condition is treated and corrected. Otherwise, abuse shall be disqualifying for all operations.
(f) Medication. Any medication taken in such a dosage that the taking or temporary delay of taking might be expected to result in high probability of sudden incapacitation.
3.2.3.3 Specific Minimum Capacities Required for Medical Qualification (1) Ears. Puretone audiometric threshold average better than 30 dB, for speech frequencies 500, 1000, 2000 Hz in better ear with or without the use of a 5
b hearing aid. If audiometric scores are unacceptable, qualification may be based upon onsite demonstration to the satisfaction of the facility operator of the examinee's ability to safely detect, interpret, and respond to speech and other auditory signals.
(2) Eyes (a) Near and distant visual acuity 20/40 in better eye, corrected or uncorrected. Corrective lenses may be used only as needed to correct a specific vision deficiency.
(b) Peripheral vision fields by confrontation to 120 degrees or greater.
(c) Color vision adequate to distinguish among red, green, and orange-yellow signal lamps, and any other unique coding if required for safe operation of the particular facility as defined by the facility operator.
(d) Adequate depth perception, either by steropsis or secondary clues as demonstrated by practical test.
(3) Respiratory. Free of disqualifying conditions enumerated in 3.2.3.2(a).
(4) Cardiovascular. Normal configuration and function including normal blood pressure with tolerance to postural changes and capacity for exertion during emergencies. The examining physician shall report whether asymmetrical neck and peripheral pulses or resting pulse rates less than 50 or more than 100 beats per minute are normal for the individual and of no significance. If the examination reveals significant cardiac arrhythmia, murmur, untreated hypertension (over 160/100 mm Hg) intolerance to postural changes, cardiac enlargement or other evidence to cardiovascular abnormality, a report of an evaluation shall accompany the medical examination report. This evaluation shall include, but is not limited to, an interpretation of an ECG and chest x-ray to indicate whether condition will cause sudden incapacitation.
(5) Musculo-skeletal. Normal symmetrical structure, range of motion and power. If any impairment exists, the applicant shall demonstrate ability to effectively perform certain assigned duties.
(6) Hematopoietic. Normal function.
(7) Lymphatic. Normal function.
(8) Neurological. Normal central and peripheral nervous system function.
Tactile discrimination (Stereognosis) sufficient. to distinguish among various shapes of control knobs and handles by touch.
3.2.3.4 Additional Examination. If the results of the examination including medical history are inconclusive, more comprehensive examination and testing as indicated by the designated medical examiner should be performed in order to determine whether or not the individual meets the requirements of section 3.2.3 and is free of disqualifying conditions.
3.2.4 Reactivity Control Manipulations. The licensed individual shall perform a number of reactivity manipulations in any combination of reactor startups, shut-downs, and significant power changes. The recommended number is 10 with the individual having primary responsibility for at least 5 of those reactivity manipulations. For senior reactor 6
operators, direct supervision of these operations may be considered equivalent to actual performance.
3.2.5 Operating Test or Evaluation. For the first and second 12 month intervals of the requalification period the licensed individual shall complete an operating test or evaluation. The interval between any two successive operating test or evaluations shall not exceed 15 months. At least five tasks selected from Section 3.2.5.1, including a reactor startup and shutdown, shall be performed and evaluated. The performance of the task may be actual or simulated.
3.2.5.1 Examples of tasks to be performed under normal and abnormal circumstances:
(a) Normal Circumstances (1) Operations and procedures tasks including pre-startup, restart, and shutdown checklists, reactor startup, shutdown, reactivity manipulations to change power, and application of administrative rules such as tagging of equipment and radiation work permits.
(2) Other reactivity tasks including fuel movements, insertion and removal of experiments, and rod exchange or movements without power change.
(3) Maintenance and monitoring tasks including verifying operability of equipment for the purpose of technical specification compliance, routine inspection of the facility, surveillance tests, water chemistry analysis, and demonstrating knowledge of reactor system and auxiliary systems' controls and indications.
(b) Abnormal Circumstances (1) Response to alarms and trips such as scrams, rundowns, high radiation, low water level, loss of coolant, loss of flow, and loss of electrical power.
(2) Emergency action, such as initiation of emergency response to radioactive releases, contaminated personnel, failed fuel, loss of confinement, fire, security violations, and natural disasters.
3.2.6 Document Review. For the first and second 12 month intervals of the requalification period the licensed individual shall review the contents of abnormal and emergency procedures. All licensed individuals shall be cognizant of facility technical specifications, design, and procedure changes in a timely manner.
3.2.7 Relicensing. Licenses may be renewed prior to their expiration upon application and successful completion of the requalification program and medical certification.
3.2.8 Absence from Licensed Functions. Licensed individuals who have not actively performed the functions of an operator or senior, operator for a minimum of four hours per calendar quarter shall perform a minimum of six hours of licensed functions under the direction of a qualified individual holding the same or higher level license prior to being reinstated.
7
3.2.9 Exemptions. At the discretion of the NRC, any portion of the requalification examination and operating evaluation may be waived for the Chief, Reactor Operations or for individuals preparing the requalification examination.
4.0 DOCUMENTATION.AND RECORDS 4.1 Documentation. The qualifications of licensed personnel shall be appropriately documented. The documentation should include the following:
(a) Medical/physical evaluation; (b) Copy of the currently valid license; (c) Records of requalification program including examinations.
4.2 Records. Records of the qualification, training, retraining, examinations, and evaluations of each licensed individual in the organization shall be retained for the duration of the currently valid license.
8 Technical Specification for the NIST Test Reactor (NBSR)
As a result of our conversation of November 26, 2008, the following changes have been made to the Technical Specifications for the NIST Test Reactor (NBSR).
- 1. TS 3.1.2: Specification (3) will be removed since this specification is redundant with the definition jof the shutdown margin.
Response
Technical Specification (3) has been removed (see attached copy).
- 2. TS 3.1.4: The specification will be changed to the wording proposed in your submission of the technical specifications dated August 14, 2007, to better reflect the intended fuel utilization during the period of the renewed license.
Response
The wording of TS 3.1.4 has been changed (see attached copy) to better reflect our discussion of November 26, 2008.
- 3. TS 6.1.3: Specification (1)(c) will be revised to include "recovery from a significant reduction in power," as required by 10 CFR 50.54(m)(1).
Response
TS 6.1.3 has been revised (see attached copy) to be in accordance with ANS 15.1(2007) Section 6.1.3(3)(d) which includes the 10CFR 50.54(m)(1) requirement.
- 4. TS figure 6.1: The legend will be revised to correct a typographical error.
Response
TS figure 6.1 legend has been revised (see attached copy).
Appendix A License No. TR-5 Technical Specifications for the NIST Test Reactor (NBSR)
Table of Contents 1.0 Intro d uctio n ..................................................................................................................... 5 1.1 Sc o p e .......................................................................................................................... 5 1.2 A pp licatio n ..................... .............................................................................................. 5 1.2.1 Purpose ..................................... ................................................................... 5 1.2.2 Format. ......................................... ................ 5 1.3 D efi n itio ns .......................................................................................................... ........ 5 1.3.1 ALARA .......................................... ........... ............................... ......... 5 1.3.2 Channel ...................................................................................................... 5 1.3.2.1 Channel Calibration ........... ........................................................................ 6 1.3.2.2 Channel Check............................................................................................. 6 1.3.2.3 Channel Test ................ :.............................................................................. 6 1.3.3 Confinem ent .................................................................................................... 6 1.3.4 Core Configuration ........................................................................................ 6 1.3.5 Excess Reactivity ............................................................................................. 6 1.3.6 Em ergency D irector ......................................................................................... 6 1.3.7 Experim ent ........... ........................................... 6 1.3.7.1 In-Reactor Vessel ........................................................................................ 6 1.3.7.2 Beam Tubes ................. ................................................... 6 1.3.7.3 M ovable Experim ent ................................................................................... 7 1.3.7.4 Secured Experim ent .................................................................................... 7 1.3.8 L ic en se ............................................................... .............................................. 7 1.3.9 M easured Value ............................ .................. 7 1.3.10 M oderator Dump .................................................................................... :............ 7 1.3.11 N atural Convection Cooling .......................................................................... 7 1.3.12 Operable .... ........ .................. 7 1.3.13 Operating ................................................................................................... 7 1.3.14 Protective Action ............................................................................................. 7 1.3.15 Reactor Operating ........................................................................................... 8 1.3.16 Reactor Operator ............................................................................. ................. 8 1.3.17 Reactor Safety System ...................................................................................... 8 1.3.18 Reactor Secured .......................................................................................... :........ 8 1.3.19 Reactor Shutdown ................................................................................................ 8 1.3.20 Reactor Shutdown M echanism s ....................................................................... 8 1.3.21 Reference Core Condition ................................................................................ 9 1.3.22 Reactor Rundown ........................................................................................... 9 1.3.23 Rod, Control ................ ................................ 9 1.3.24 Rod Drop Mode ................................................................................................ 9 1.3.25 Rod, Regulating .............................................................................................. 9 1.3.26 Scra m ............................................ ....................................................................... 9 1.3.26.1 M ajor Scram ........................................................................................... 9 1.3.27 Scram Tim e .................................................................................................. 9 1.3.28 Senior Reactor Operator .................................................................................. 9 1.3.29 Shall, Should and M ay ......................................................................................... 9 1.3.30 Shutdown M argin .............................................................................................. 10
1.3.31 , Surveillance Activities .................................................................................. 10 1.3.32 Surveillance Intervals ..................................................................................... 10 1.3.33 Unscheduled Shutdown .................................................................................. 10 2.0 Safety Limit and Lim iting Safety System Settings ................................................... 11 2.1 Safety Lim it ................................................................................................................ 11 2.2 Limiting Safety System Settings .............................................................................. 11 3.0 Lim iting Conditions for Operations ........ I................................................................. 13 3.1 Reactor Core Param eters ............................................................................. ......... 13 3.1.1 Reactor Pow er ............................................................................................... 13 3.1.2 Reactivity Limitations ..................................................................................... 13 3.1.3 Core Configuration ......................................................................................... 14 3.1.4 Fuel Burnup ................ ................................ 14 3.2 Reactor Control and Safety System s .................................................................... 15 3.2.1 Shim Arm s ................................................................. 15 3.2.2 Reactor Safety System Channels ....................................................................... 16, 3.3 Coolant System ........................................................................................................... 17 3.3.1 Prim ary and Secondary .................................................................................. 17 3.3.2 Emergency Core Cooling ................. ........................................ 19 3.3.3 M oderator Dum p System ............................................................................. 19 3.4 Confinem ent System ............................................................................................ 20 3.4.1 Operations that Require Confinem ent ........................................................... 20 3.4.2 Equipm ent to Achieve Confinem ent ...................................... ............................. 21 3.5 Ventilation System ............................................................................................... 22 3.6 Em ergency Power System .................................................................................... 23 3.7 Radiation M onitoring System s and Effl uents ........................................................ 24 3.7.1 M onitoring System s and Effluent Lim its ........................................................ 24 3.7.2 Effluents ........................................................ 26 3.8 Experim ents .............................................................................................................. 27 3.8.1 Reactivity Lim its .......................................................................................... 27 3.8.2 M aterials .......................................................................................................... 28 3.9 Facility Specific ................................................................................ .................... 29 3.9.1 Fuel Storage ................................................................................................. 29 3.9.2 Fuel Handling ................................................................................................ 30 3.9.2.1 W ithin the Reactor Vessel .......................................................................... 30 3.9.2.2 All Other Conditions ............................. ........... 30 4.0 Surveillance Requirem ents ........................................................................... .......... 32 4.1 Reactor Core Parameters .......................................... 32 4.1.1 Reactor Pow er ................... .......................................................................... 32 4.1.2 Reactivity Lim itations .................................................................................. 33 4.2 Reactor Control and Safety System s ..................................................................... 33 4.2.1 Shim Arm s .................................................................................................... 33 4.2.2 Reactor Safety System Channels ................................................................... 34 4.3 Coolant System s ................................................................................................... 35 4.3.1 Primary and Secondary .................................................................................. 35 4.3.2 Em ergency Core Cooling System ................................................................. 35 4.3.3 M oderator Dump System .............................................................................. 36 3
4.4 Confinement System ................................................................................................... 36 4.5 Ventilation System .............................................. 37 4.6 Emergency Power System ................................................................................... 38 4.7 Radiation M onitoring System and Effluents .......................................................... 39 4.7.1 M onitoring System ........................................................................................ 39 4 .7 .2 Effl u ents ............................................................................................................ 41 4 .8 E xperim ents ......................................................................... .....................4 1 5 .0 D esig n F eatu res ............................................................................................................. 43 5.1 Site D escription .................................................................................................. . . 43 5.2 Reactor Coolant System ................ ........................... 43 5.3 Reactor Core and Fuel ........................................................................................... 44 6.0 Administrative Controls ................................................................................................. 45 6 .1 Organ izatio n ................................................................................................... .......... 4 5 6 .1.1 S tru ctu re ........................................................................................................... 45 6 .1.2 R esp o nsib ility .................................................................................................... 45 6 .1.3 S taffing .............................................................................................................. 45 6.1.4 Selection and Training of Personnel ............................................................... 46 6.1.4.1 Selection of Personnel ............................................................................... 46 6.1.4.2 Training of Personnel ................................................................................ 47 6.2 Review and Audit .................................................................................................. 47 6.2.1 Composition and Qualifications .................................................... 48 6.2.2 Safety Evaluation Committee Charter and Rules ........................................... 48 6.2.3 SEC Review Function ................................................................................. 48 6.2.4 SEC Audit Function ....................................................................................... 49 6.2.5 Safety Assessment Committee (SAC) ............................................................ 49 6.3 Radiation Safety .................................................................................................. .. 49 6 .4 Pro cedu res .......................................................... .................................................... 50 6.5 Experiment Review and Approval ..................... 50 6.6 Required Actions ..... .. .............................. .................. 51 6.6.2 Actions to Be Taken in the Event of an Occurrence of the Type Identified in Section 6.7.2 other than a Safety Limit Violation ...................... 51 6 .7 Rep o rts ............................... ...................................................................................... 52 .
6.7.1 Annual Operating Report ............. :................................................................ 52 6 .7.2 Sp ecial R ep orts 52 5...........................
6 .8 Reco rd s .................................................... ................................................................ 53 6.8.1 Records to be Retained for a Period of at Least Five Years or for the Life of the Component Involved if Less than Five Years ............................................... 53 6.8.2 Records to be Retained for at Least One Operator Licensing Cycle ............... 54 6.8.3 Records to be Retained for the Life of the Reactor Facility ............... 54 4
1.0 Introduction These technical specifications apply to the National Institute of Standards and Technology (NIST) Test Reactor (NBSR) license TR-5.
1.1 Scope The following areas are addressed: Definitions, Safety Limits (SL) and Limiting Safety System Settings (LSSS), Limiting Conditions for Operation (LCO),
Surveillance Requirements, Design Features, and Administrative Controls.
1.2 Application The dimensions, measurements, and other numerical values given in these specifications may differ slightly from actual values as a result of the normal construction and manufacturing tolerances, or normal accuracy of instrumentation.
1.2.1 Purpose These specifications are derived from NISTIR 7102 (NBSR 14 Safety Analysis Report). They consist of specific limitations and equipment requirements for the safe operation of the reactor and for dealing with abnormal situations.: These specifications represent a comprehensive envelope of safe operation. Only those operational parameters and equipment requirements directly related to verifying and preserving this safety envelope are listed.
1.2.2 Format The format of these specifications is as described in ANSI/ANS 15.1- 2007.
1.3 Definitions The following terms are sufficiently important to be separately defined:
1.3.1 ALARA As Low As is Reasonably Achievable. The practice of making every reasonable effort to maintain exposures to radiation as far below dose limits as is practicable, consistent with the purpose and benefits of licensed activities and the mission of the NBSR.
1.3.2 Channel The combination of sensor, line, amplifier, and output devices which are connected for the purpose of measuring the value of a parameter.
5
1.3.2.1 Channel Calibration The adjustment of the channel such that its output corresponds with acceptable accuracy to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip and shall be deemed to include a channel test.
1.3.2.2 Channel Check A qualitative verification of acceptable performance by observation of channel behavior, or by comparison of the channel with other independent channels or systems measuring the same variable.
1.3.2.3 Channel Test The introduction of a signal into the channel for verification that it is operable.
1.3.3 Confinement An enclosure of the C wing of the NCNR that is designed to limit the release of effluents between the enclosure and its external environment through controlled or defined pathways.
1.3.4 Core Configuration The number, type, or arrangement of fuel elements, reflector elements and regulating or control rods occupying the core grid.
1.3.5 Excess Reactivity That amount of reactivity that would exist if all reactivity control devices were moved to the maximum reactive condition from the point where the reactor is critical.
1.3.6 Emergency Director The functions of the Emergency Director are defined in the NBSR Emergency Plan.
1.3.7 Experiment 1.3.7.1 In-Reactor Vessel Any operation, hardware, or target (excluding devices such as detectors and foils), that is designed to investigate non-routine reactor characteristics or that is intended for irradiation within the reactor vessel.
1.3.7.2 Beam Tubes Any sample or hardware placed in a beam tube that has an unobstructed view of the reactor vessel or any materials placed in a 6
beam tube, such as filters and shields for which accident mitigation credit is taken.
1.3.7.3 Movable Experiment Any experiment in which all or part of the experiment may be moved in or near the core or into and out of the reactor while the reactor is operating 1.3.7.4 Secured Experiment Any experiment, experimental apparatus, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means. The restraining force must be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces which are normal to the operating environment of the experiment, or by forces which can arise as a result of credible malfunctions.
1.3.8 License The written authorization, by the Nuclear Regulatory Commission, for an individual or organization to carry out the duties and responsibilities associated with a facility requiring licensing.
1.3.9 Measured Value The value of a parameter as it appears on the output of a channel.
1.3.10 Moderator Dump An action which drops the water level to approximately one inch (2.5 cm) above the reactor core, thereby ensuring a subcritical state for an emergency shutdown under all reactor operating conditions.
1.3.11 Natural Convection Cooling That flow of primary water between the reactor core and a heat exchanger with no pumps operating.
1.3.12 Operable The condition of a system or component when it is capable of performing its intended function, as determined by testing or indication.
1.3.13 Operating The condition of a component or system when it is performing its intended function.
1.3.14 Protective Action The initiation of a signal or the operation of equipment within the reactor safety system in response to a variable or condition of the reactor facility having reached a specified limit.
7
1.3.15 Reactor Operating The condition of the reactor when it is not secured or shutdown.
1.3.16 Reactor Operator An individual licensed by the U.S. Nuclear Regulatory Commission to manipulate the controls of the NBSR.
1.3.17 Reactor Safety System Those systems designated in these technical specifications, including their associated input channels, which are designed to initiate automatic reactor protection or to provide information for initiation of manual protective action.
1.3.18 Reactor Secured The condition of the reactor when (a), (b), or (c) is true.
(a) (1) The Control Power key switch or the Rod Drive Power key switch is in the off position with the key removed and under the control of a licensed operator; and (2) The condition of the shim arms is per the specification of Section 3.1.2(3); and (3) No work is in progress involving core fuel, core structure, installed shim arms, or shim arm drives, unless the shim arm drive shafts are mechanically fixed; and (4) No experiments in any reactor experiment facility, or in any other way near the reactor, are being moved or serviced if the experiments have, on movement, reactivity worth exceeding the maximum value allowed for a single experiment or $1.00, whichever is smaller.
(b) There is insufficient fissile material in the reactor core or adjacent experiments to attain criticality under optimum available conditions of moderation and reflection.
(c) The reactor is in the rod drop test mode, and a senior reactor operator is in direct charge of the operation.
1.3.19 Reactor Shutdown When the reactor is subcritical by at least one dollar ($1.00) in the Reference Core Condition with all installed experiments in their most reactive condition.
1.3.20 Reactor Shutdown Mechanisms Mechanisms that can place the reactor in a shutdown condition, and include:
(a) Rundown (b) Scram (c) Major Scram (d) Moderator Dump 8
1.3.21 Reference Core Condition The condition of the core when it is at ambient temperature and the reactivity worth of xenon is negligible.
1.3.22 Reactor Rundown The electrically driven insertion of all shim arms and the regulating rod at their normal operating speed.
1.3.23 Rod, Control A device, also known as a shim arm, fabricated from neutron absorbing material that is used to establish neutron flux changes and to compensate for routine reactivity losses. The shim arms, when coupled to their drives, provide reactivity control and therefore flux control. When the shim arm becomes decoupled from its drive mechanism it provides a safety function by rapidly introducing negative reactivity into the reactor core.
1.3.24 Rod Drop Mode Any combination of control systems and mechanical systems that allows for the movement of only a single shim arm and ensures the reactor remains shutdown, when sufficient fissile material for criticality is present.
1.3.25 Rod, Regulating A low worth control rod used primarily to maintain an intended power level that need not have scram capability. Its position may be varied manually or automatically.
1.3.26 Scram The spring assisted gravity insertion of all shim arms.
1.3.26.1 Major Scram A scram accompanied by the immediate activation of the confinement isolation system.
1.3.27 Scram Time The elapsed time between the initiation of a scram signal and a specified movement of a control or safety device.
1.3.28 Senior Reactor Operator An individual licensed to direct the activities of reactor operators. Such an individual is also a reactor operator.
1.3.29 Shall, Should and May The word "shall" is used to denote a requirement; the word "should" to denote a recommendation; and the word "may" to denote permission, neither a requirement nor a recommendation.
9
1.3.30 Shutdown Margin The minimum shutdown reactivity necessary to provide confidence that the reactor can be shutdown by means of the control and safely systems starting from any permissible operating condition, with the most reactive shim arm in the most reactive position and the regulating rod fully withdrawn, and that the reactor will remain shutdown without further operator action.
1.3.31 Surveillance Activities Those tests, checks and calibrations done to predict the operability of the equipment described in Section 4.0.
1.3.32 Surveillance Intervals Maximum intervals are established to provide operational flexibility and not to reduce frequency. Established frequencies shall be maintained over the long term. The surveillance interval is the time between a check, test or calibration, whichever, is appropriate to the item being subjected to the surveillance, and is measured from the date of the last surveillance. Surveillance intervals are:
(a) Five Year Interval not to exceed six years.
(b) Biennial Interval not to exceed two and half years.
(c) Annual Interval not to exceed 15 months.
(d) Semi-annual Interval not to exceed seven and a half months.
(e) Quarterly Interval not to exceed four months.
(f) Monthly Interval not to exceed six weeks.
(g) Weekly Interval not to exceed ten days.
1.3.33 Unscheduled Shutdown Any unplanned shutdown of the reactor caused by actuation of the reactor safety system, operator error, equipment malfunction, or a manual shutdown in response to conditions that could adversely affect safe operation, not including shutdowns that occur during testing or equipment operability checks.
10
2.0 Safety Limit and Limiting Safety System Settings 2.1 Safety Limit Applicability: Fuel temperature Objective: To maintain the integrity of the fuel cladding and prevent the release of significant amounts of fission products.
Specification The reactor fuel cladding temperature shall not exceed 842°F (450°C) for any operating conditions of power and flow.
Basis Maintaining the integrity of the fuel cladding requires that the cladding remain below its blistering temperature of 842 0 F (450°C). For all reactor operating conditions that avoid either a departure from nucleate boiling (DNB), or exceeding the Critical Heat Flux (CHF)), or the onset of flow instability (OFI), cladding temperatures remain substantially below the fuel blistering temperature. Conservative calculations have shown that limiting combinations of reactor power and reactor coolant system flow and temperature will prevent DNB and thus fuel blistering.
2.2 Limiting Safety System Settings Applicability: Power, flow, and temperature parameters Objective: To ensure protective action if any combination of the principal process variables should approach the safety limit.
Specifications (1) Reactor power shall not exceed 130% of full power.
(2) Reactor outlet temperature shall not exceed 147TF.
(3) Forced coolant flow shall not be less than 60 gpm/MW for the inner plenum and not less than 235 gpm/MW' for the outer plenum.
(4) Reactor power, with natural circulation cooling flow, shall not exceed 500 kW.
Basis At the values established above, the Limiting Safety. System Settings provide a significant margin from the Safety Limit. Even in the extremely unlikely event that 11
reactor power, coolant flow, and outlet temperature simultaneously reach their Limiting Safety System Settings, the critical heat flux ratio (CHFR) is at least 2. For all other conditions the CHFR is considerably higher. This will ensure that any reactor transient caused by equipment malfunction or operator error will be terminated well before the safety limit is reached. Overall uncertainties in process instrumentation have been incorporated in the Limiting Safety System Settings.
Steady state thermal hydraulic analysis shows that operation at 500 kW with natural circulation results in a CHFR and OFI ratio greater than 2. Transient analysis of reactivity insertion accidents shows that the fuel cladding temperature remains far below the safety limit.
12
3.0 Limiting Conditions for Operations 3.1 Reactor Core Parameters 3.1.1 Reactor Power Applicability: Reactor power Objective: To ensure that licensed power is not exceeded and the safety limit is not exceeded through initiation of protective action at a specified power.
Specification The nominal reactor power shall not exceed 20 MW thermal. The reactor scram set point for a reactor power level safety channel shall not exceed 125% of full power.
Basis Operational experience and thermal-hydraulic calculations demonstrate that the fuel elements may be safely operated at these power levels. The operating limits developed here are based upon well tested correlations, are conservative, and provide ample margin to ensure that there will be no damage to fuel during normal operation. In addition, the operating conditions provide ample margin for all credible accident scenarios to ensure that there will be no fuel damage.
3.1.2 Reactivity Limitations Applicability: Core reactivity and shim arm worth Objective: To ensure that the reactor can be placed in a shutdown condition at all times and that the safety limit shall not be exceeded.
Specifications (1) The maximum available excess reactivity for reference core conditions shall not exceed 15% Ap (approximately $20).
(2) The reactor shall not be operated unless the shutdown margin provided by the shim arms is greater than 0.757% Ap ($1.00) with:
(a) The reactor in 'any core condition, and (b) All movable experiments in their most reactive condition.
13
Basis (1) An excess reactivity limit provides adequate excess reactivity to override the xenon buildup and to overcome the temperature change in going from zero power to 20 MW, without affecting the required shutdown margin. In addition, the maximum reactivity insertion accident at startup, which assumes the insertion of 0.5% Ap into a critical core, is not affected by the total core excess reactivity.
(2) These, specifications ensure that the reactor can be put into a shutdown condition from any operating condition and remain shutdown even if the maximum worth shim arm should stick in the fully withdrawn position with the regulating rod also fully withdrawn.
3.1.3 Core Configuration Applicability: Core grid positions Objective: To ensure that a failed shim arm does not adversely affect core reactivity and cooling flow is maintained.
Specification The reactor shall not operate unless all grid positions are filled with full length fuel elements or thimbles.
Basis The NBSR employs shim arm stops to prevent a broken shim arm from dropping from the reactor core. The proper operation of these stops depends on adjacent fuel elements or experimental thimbles being in place to prevent the broken shim arm from falling from the core lattice. Furthermore, core grid positions shall be filled to prevent coolant flow from bypassing the fuel elements.
3.1.4 Fuel Burnup Applicability: Fuel Objective: To remain within allowable limits of burnup 14
Specification The average fission density shall not exceed 2 x 1027 fissions/m3 .
Basis Fuel elements in the NBSR are burned for seven (7) or eight (8) cycles. An eight (8) cycle fuel element has an average fission density of approximately 1.9 x 10 27 fissions/m 3 . The 'U3 0 8 - Al dispersion MTR fuels have been in widespread use for over 40 years. Extensive testing of fuel plates has been performed to determine the limits on fission density as a function of fuel loading. Several measurements of swelling in fuel plates show that NBSR fuel, which is moderately loaded at 18% is well below the curve that represents the allowable limit of burnup.
3.2 Reactor Control and Safety Systems 3.2.1 Shim Arms Applicability: Shim arms and shim arm worth Objective: To ensure proper shim arm reactivity insertion.
Specifications The reactor shall not be operated unless:
(1) All four shim arms are operable.
(2) The scram time shall not exceed 240 msec for a shim arm insertioni of 5 degrees.
(3) The reactivity insertion rate for the four shim arms shall not exceed 5 x 104 Ap/sec.
Basis (1) Although the NBSR could operate and maintain a substantial shutdown margin with less than the four installed shim arms, flux and shim arm worth distortions could occur by operating in this manner. Furthermore, operation of the reactor with one shim arm known to be inoperable would further reduce the shutdown margin that would be available if one of the remaining three shim arms were to suffer a mechanical failure that prevented its insertion.
15
(2) and (3) A shim arm withdrawal accident for the NBSR was analyzed using the maximum reactivity insertion rate, corresponding to the maximum beginning-of-life shim arm worths with the shim arms operating at the design speed of their constant speed mechanisms. The analysis shows that the most severe accident, a startup from source level, will not result in core damage.
3.2.2 Reactor Safety System Channels Applicability: Required instrument channels Objective: To provide protective action for nuclear and process variables to ensure the LSSS values are not exceeded.
Specifications The reactor shall not be operated unless the channels described in Table 3.2.2 are operable and the information is displayed in the reactor Control Room.
16
Table 3.2.2 Reactor Safety System Channels Minimum Nuclear and Process Channels Required Channel Scram Major Scram Rundown (1) High Flux level 2 (2) Short period below 5% rated power 2 (3) Low reactor vessel D2 0 level 1,3 2 (4) Low flow reactor outlet 2,3 1 (5) Low flow reactor inner or outer plenum 2 ' 3 1 (6) Manual (outside of the Control Room) 1 (7) Manual 1 1 (8) Reactor Outlet Temperature I (9) Gaseous Effluent Monitors4 2 1 One (1) of two (2) channels may be bypassed for tests or during the time maintenance involving the replacement of components and modules or calibrations and repairs are actually being performed.
2 One (1) of these two (2) flow channels may be bypassed during tests, or during the time maintenance involving the replacement of components and modules or calibrations and minor repairs are actually being performed. However, outlet low flow may not be bypassed unless both inner and outer low-flow reactor inlet safety systems are operating.
3 May be bypassed during periods of reactor operation (up to 500 kW) when a reduction in Limiting Safety System Setting values is permitted per the specifications of Sections 2.2 and 3.3.1.
4 See specifications of Section 3.7.1 Basis The nuclear and process channels of Table 3.2.2 initiate protective action to ensure that the safety limit is not exceeded. With these channels operable, the safety system has redundancy.
3.3 Coolant System 3.3.1 Primary and Secondary Applicability: Primary fluid systems Objective: To prevent degradation of primary systems' materials.
17
Specifications The reactor shall not be operated unless:
(1) The reactor vessel coolant level is no more than 25 inches below the overflow standpipe.
Exception: To permit periodic surveillance of the effectiveness of the moderator dump, it is necessary to operate the reactor without restriction on reactor vessel level.
(2) The D2 concentration in the Helium Sweep System shall not exceed 4% by volume.
(3) All materials, including those of the reactor vessel, in contact with the primary coolant shall be compatible with the D20 environment.
Basis (1) The limiting value for reactor vessel coolant level is somewhat arbitrary because the core is in no danger so long as it is covered with water.
However, a drop of vessel level indicates a malfunction of the reactor cooling system and possible approach to uncovering the core. Thus, a measurable value well above the minimum level is chosen in order to provide a generous margin of approximately 7 feet (2.13 m) above the fuel elements. To permit periodic surveillance of the effectiveness of the moderator dump, it is necessary to operate the reactor without restriction on reactor vessel level. This is permissible under conditions when forced reactor cooling flow is not required, such as is permitted in the specifications of Section 2.0.
(2) Deuterium gas will collect in the helium cover gas system because of radiolytic disassociation of D2 0. Damage to the primary system could occur if this gas were to reach an explosive concentration (about 7.8% by volume at 77°F (25'C) in helium if mixed with air). To ensure a substantial margin below the lowest potentially explosive value, a 4% limit is imposed.
(3) Materials of construction, being primarily low activation alloys and stainless steel, are chemically compatible with the primary coolant. The stainless steel pumps are heavy walled members and are in areas of low stress, so they should not be susceptible to chemical attack or stress corrosion failures. A failure of the gaskets or valve bellows would not result in catastrophic failure of the primary system. Other materials should be compatible so as not to cause a loss of material and system integrity.
18
3.3.2 Emergency Core Cooling Applicability: Emergency Core Cooling System Objective: To ensure an emergency supply of coolant.
Specifications The reactor shall not be operated unless:
(1) The D20 emergency core cooling system is operable.
(2) A source of makeup water to the D2 0 emergency cooling tank is available.
Basis (1) In the event of a loss of core coolant, the emergency core cooling system provides adequate protection against melting of the reactor core and associated release of fission products.
(2) The emergency core cooling system employs one sump pump to return spilled coolant to the overhead storage tank. Because only one sump pump is used, it must be operational whenever the reactor is operational. There is sufficient D2 0 available to provide approximately 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of cooling on a once-through basis. In the event that the sump pump fails and the D20 supply in the overhead storage tank is exhausted, domestic water or a suitable alternative would be used to furnish water for once-through cooling. The water makeup capacity must be in excess of 25 gpm, which was found adequate in cooling calculations to prevent fuel damage.
3.3.3 Moderator Dump System Applicability: Moderator dump Objective: To provide a backup shutdown mechanism.
Specification The reactor shall not be operated unless the reactor moderator dump system is operable.
Basis In the unlikely event that the shim arms cannot be inserted, an alternate means of shutting down the reactor is provided by the moderator dump. The moderator dump provides a shutdown capability for any core configuration.
19
Hence, it is considered necessary for safe operation. It has been shown that the moderator dump provides sufficient negative reactivity to make the normal startup (SU) core subcritical even with all four shim arms fully withdrawn.
3.4 Confinement System 3.4.1 Operations that Require Confinement Applicability: Reactivity changes within the vessel and fuel movements outside of the vessel Objective: To provide an additional barrier to fission product releases.
Specifications Confinement shall be maintained when:
(1) The reactor is operating.
(2) Changes of components or equipment within the confines of the thermal shield, other than rod drop tests or movement of experiments, are being made which could cause a significant change in reactivity.
(3) There is movement of irradiated fuel outside a sealed container or system.
(4) The reactor has been shutdown for shorter than the time specified in the specification of Section 3.9.2.2.
Basis (1) The confinement system is a major engineered safety feature. It is the final physical barrier to mitigate the release of radioactive particles and gasses to the environment following accidents. Confinement is stringently defined to ensure that the confinement building shall perform in accordance with its design basis. Confinement is not required when the reactor is shutdown and experiments are to be inserted or removed.
(2) Changes in the core involving such operations as irradiated fuel handling or shim arm repairs affect the reactivity of the core and could reduce the shutdown margin of the reactor. Confinement shall be required when these changes are made because they affect the status of the core.
The reactor is normally shutdown by a substantial reactivity margin.
Experiments are usually inserted and removed one at a time; hence, the' total reactivity change in any single operation shall be limited to the specified maximum worth of 0.5% Ap for any single experiment 20
(including "fixed" experiments). Under this circumstance, the shutdown margin would be substantial.
(3) Even when the reactor is shutdown, irradiated fuel contains fission product inventories sufficient to allow the'specification of Section 3.7.2 to be exceeded should the element fail. This fuel poses a potential hazard in that its cladding could be damaged when it is not contained in a closed system, such as during transit or during sawing of aluminum end pieces.
Confinement integrity is not required when irradiated fuel is contained within a closed system, such as the reactor vessel, the transfer lock of the refueling system, or a sealed shipping cask, that serves as a secondary barrier of fission product release.
(4) The specification of Section 3.9.2.2 restricts fuel movement for a specified period. Maintenance that would disable the confinement is prohibited during that period. Building doors could be opened, however, provided that confinement can be rapidly re-established. Confinement integrity is no longer required after the waiting period, because a loss of all water to fuel in a sealed container or system will not cause fuel damage.
3.4.2 Equipment to Achieve Confinement Applicability: Confinement system Objective: To ensure that TS 3.4.1 can be met.
Specifications Confinement shall mean that:
(1) All penetrations of the confinement building are either sealed or capable of being isolated. All piping penetrations within the reactor building are capable of withstanding the confinement test pressure.
(2) All automatic isolation valves in the ventilation, process piping and guide tubes are either operable or can be closed.
(3) All automatic personnel access doors can be closed and sealed.
(4) Except during passage, at least one set of the reactor building vestibule doors for each automatic personnel door is closed or attended, or the automatic door is closed and sealed.
(5) The reactor building truck door is closed and sealed.
21
Exception to (1) - (5): In order to provide for prompt remedial action, reactor confinement effectiveness may be reduced for a period of no longer than 15 minutes when specifications (1) - (5) are not met or do not exist.
Basis (1) and (2) The confinement building is designed to be automatically sealed upon indication of high activity. To attempt to operate the reactor with any of these conditions unmet is a violation of the confinement design basis.
Although tests have shown that the confinement building can continue to operate with one or more of these closures failed, its margin of effectiveness is reduced. If a closure device is placed in its closed or sealed condition, then operability of the automatic closure device is not required.
(3) and (4) Tests performed on the confinement building have shown that even if one of the automatically closing personnel doors fails to operate properly, confinement design capability can be met if one set of building vestibule doors per vestibule are closed. By specifying that these doors remain closed except when they are being used or attended, a backup to the normal confinement closure is provided.
(5) The reactor building truck door is not provided with automatic closure devices. Tests have shown that the confinement building can continue to operate properly, although at reduced efficiency, if the truck door seal were to fail. Confinement cannot be established if the truck door is open.
3.5 Ventilation System Applicability: Emergency and normal ventilation Objective: To minimize exposures outside of the confinement building Specifications The reactor shall not be operated unless:
(1) The building emergency recirculation system and emergency exhaust systems, including both fans, are operable, and both the absolute and charcoal filter efficiencies are at 99% or greater.
(2) The reactor building ventilation system can filter exhaust air and discharge it above the confinement building roof level.
Exception to (1) and (2): In order to provide time for prompt remedial action, reactor ventilation may be inoperable for a period of no longer than 15 minutes when the 22
specifications are not met or do not exist. Minor maintenance which disables a single fan and can be suspended without affecting the operability of the system may be performed during reactor operation.
Basis The potential radiation exposure to staff personnel and persons at the site boundary and beyond has been calculated following an accidental release of fission product activity. These calculations are based on the proper operation of the building recirculation system and the emergency exhaust system to maintain the confinement building at a negative pressure and to direct all effluents through filters and up through the reactor building stack. The emergency exhaust system is a redundant system to ensure its operation. Because of its importance, this redundancy should be available at all times so that any single failure would not preclude system operation when required.
The emergency exhaust system is designed to pass reactor building effluents through high-efficiency particulate filters capable of removing particles of 0.3 gim or greater with an efficiency of at least 99% and the charcoal filters are capable of removing greater than 99% of the Iodine from the air. All discharge of the effluents is above the reactor building roof level. This system ensures filtering and dilution of gaseous effluents before these effluents reach personnel either onsite or offsite. The system can properly perform this function using various combinations of its installed fans and the building stack.
3.6 Emergency Power System Applicability: Emergency electrical power supplies Objective: To ensure emergency power for vital equipment.
Specification The reactor shall not be operated unless at least one (1) of the diesel-powered generators and the station battery are operable, including associated distribution equipment, and the nuclear instrumentation and emergency exhaust fans can be supplied with electrical power from the diesel generator or the battery.
Exception: In order to provide time for prompt remedial action, the Emergency Power may be inoperable for a period of no longer than 15 minutes when the specification is not met or does not exist.
Basis One diesel-powered generator is capable of supplying emergency power to all necessary emergency equipment. The second diesel-powered generator is provided to permit outages for maintenance and repairs.
23
The station battery provides an additional source of emergency power for the nuclear instruments and the emergency exhaust fans. These fans may be powered from AC or DC power supplies. The battery is capable of supplying this emergency load for a minimum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. By allowing this amount of time and by requiring operability of.
at least one diesel and the station battery, adequate emergency power sources shall always be available.
3.7 Radiation Monitoring Systems and Effluents 3.7.1 Monitoring Systems and Effluent Limits Applicability: Radiation monitoring systems Objective: To detect abnormal levels or locations of radioactivity.
Specifications The reactor shall not be operated unless:
(1) Two of three gaseous effluent monitors are operable for normal air, irradiated air, and stack air.
(2). One fission products monitor is operable or sample analysis for fission product activity, is conducted daily.
(3) One secondary coolant activity monitor is operable or a D2 0 storage tank level monitor is operable.
(4) Two area radiation monitors are operable on floors C-100 and C-200.
(5) The primary tritium concentration is less than or equal to 5 Ci/l.
(6) An environmental monitoring program shall be carried out and shall include as a minimum the analysis of samples from surface waters from the surrounding areas, vegetation or soil and air sampling.
When required monitors are inoperable, then portable instruments, survey or analysis may be substituted for any of the normally installed monitors in specifications (1) - (4) for periods of one (1) week or for the duration of a reactor run.
Basis (1) The requirements of 10 CFR 20.1502(b) (2007) are met by regular monitoring for airborne radionuclides and bioassay of exposed personnel.
The two primary airborne radionuclides present at the NBSR are 41Ar and 24
3H. The normal air exhaust system draws air from areas supplied by conditioned air, such as the first and second floors of the confinement building. The irradiated air exhaust system draws air from areas most likely to have contaminated air, such as waste sumps and penetrations in the biological shield. Normal and irradiated air are monitored continuously with G-M detectors sensitive to 03and y emissions and the combined air is exhausted through the stack. The stack release is monitored with a G-M detector.
(2) A fission products monitor located in the helium sweep gas will give an indication of a "pin-hole" breach in the cladding so that early preventive measures can be taken. When this monitor is not functional, daily testing will ensure that the fuel cladding is intact. These two measures ensure that there are no undetected releases of fission products to the primary coolant.
(3) Monitoring for primary water leakage into the secondary coolant is done by a secondary water monitor that is sensitive to radionuclides in the primary water. Leakage of primary to secondary would also be detected by a change in the D2 0 storage tank level (4) Fixed gamma area radiation monitors are positioned at selected locations in the confinement building. Typical alarm setting are less than 5 mrem/hr and adjusted as needed for non-routine activities, generally with the objective of identifying unusual changes in radiation conditions.
(5) At the end of the term of the NBSR license the maximum tritium concentration in the primary coolant is estimated to be 5 Ci/l. This value and reliable leak detection ensures that tritium concentrations in effluents shall be as low as is practicable.
(6) Area vegetation and soil samples are collected for analysis. Grass samples are, collected during the growing season, April through September, and soil samples during the non-growing season, October through March.
Thermoluminescent dosimeters or other devices also are placed around the perimeter of the NBSR site to monitor direct radiation. The continuation of this environmental monitoring program will verify that the operation of the NBSR presents no significant risk to the public health and safety.
Since 1,969, when the NBSR began routine power operation, the environmental monitoring program has revealed nothing of significance, thereby confirming that operation of the NBSR has had little or no effect on the environment.
A report published in March 2003 supports the findings of previous studies conducted on the hydrology and geology of the NIST site and vicinity. No significant changes in the hydro-geologic systems or ground 25
water use were identified. This report further verifies the assumptions and techniques developed in 1964.
3.7.2 Effluents Applicability: Annual releases Objective: To minimize exposures to the public.
Specification The reactor shall not be operated unless:
The total exposure from effluents from the reactor facility to a person at the site boundary shall not exceed 100 mrem per calendar year, less any external dose from the facility. The limit shall be established at the point of release or measurement using accepted diffusion factors to the boundary. For halogens and particulates with half-lives longer than 8, days, a reconcentration factor shall be included where appropriate.
Basis The criteria for determination of concentration limits specified above ensure that 10 CFR 20 (2007) limits are not exceeded at the site boundary. The allowance for dilution from the reactor building stack to the nearest site boundary is 1,000. This value of 1,000 from the diffusion view point is the minimum expected at the nearest site boundary under the least favorable meteorological conditions. This number could be increased by one or two orders of magnitude if normal variations in wind speed and direction were considered. Because these variations are not considered, a one or two order of magnitude margin is inherent in this limit.
In specifying the limits on particulates and long lived (longer than 8 days) halogens, consideration was given to the possibility of biological reconcentration in food crops or dairy products. Using available information (Soldat, J.D., Health Physics 9, p. 1170, 1963), a conservative (both the COMPLY and CAP88 codes indicate that 700 is at least an order of magnitude higher than needed) reconcentration factor of 700 is applied. Thus, the limits for those isotopes are the Effluent Concentration Limits as specified in Appendix B, Table II of 10 CFR 20 (2007) multiplied by the 1,000 dilution factor divided by the 700 reconcentration factor; that is, 1.4 times the Effluent Concentration Limit.
For the purpose of converting concentrations to dose, the values of 10 CFR 20, Appendix B, Table 2 (2007), represent an annual dose of 50 mrem, except for submersion gases where they represent an annual dose of 26
100 mrem. It should be taken into consideration that the values for submersion gases are based on an infinite hemisphere geometry which is rarely achievable and therefore tends to overestimate the dose.
3.8 Experiments 3.8.1 Reactivity Limits Applicability: Reactivity of experiments Objective: To limit reactivity excursions.
Specifications The reactor shall not be operated unless:
(1) The absolute reactivity of any experiment shall not exceed 0.5% Ap.
(2) The sum of the absolute values of reactivity of all experiments in the reactor and experimental facilities shall not exceed 2.6% Ap.
(3) No experiment malfunction shall affect any other experiment so as to cause its failure. Similarly, no reactor transient shall cause an experiment to fail in such a way as to contribute to an accident.
Basis (1) The individual experiment reactivity limit is chosen so that the failure of an experimental installation or component shall not cause a reactivity increase greater than can be controlled by the regulating rod. Because the' failure of individual experiments cannot be discounted during the operating life of the NBSR, failure should be within the control capability of the reactor. This limit does not include such semi-permanent structural materials as brackets, supports, and tubes that are occasionally removed or modified, but which are positively attached to reactor structures. When these components are installed, they are considered structural members rather than part of an experiment.
(2) The combined reactivity allowance for experiments was chosen to allow sufficient reactivity for contemplated experiments while limiting neutron flux depressions to less than 10%. Included within the specified 2.6% Ap is a 0.2% Ap allowance for the pneumatic irradiation system, 1.3% Ap for experiments that can be removed during reactor operation, and the remainder for semi-permanent experiments that can only be removed during reactor shutdown. Even if it were assumed that one experiment with the maximum allowable reactivity of 0.5% Ap for movable 27
experiments was removed in 0.5 seconds, analysis shows that this ramp insertion into the NBSR operating at 20 MW would not result in any fuel failure leading to the release of fission products. The 0.2% Ap for the combined pneumatic irradiation systems has been shown to be bounded by the ramp insertion of 0.5% Ap and is well below this referenced accident as well as being within the Ap capability of the regulating rod.
(3) In addition to all reactor experiments being designed not to fail from internal gas buildup or overheating, they shall be designed so that their failure does not affect either the reactor or other experiments. They shall also be designed to withstand, without failure, the same transients that the reactor itself can withstand without failure.
3.8.2 Materials Applicability: All materials used in experiments Objective: To prevent damage to the reactor or a significant release of radioactivity.
Specifications (1) Explosive or metastable materials capable of significant energy releases shall be irradiated in double walled containers that have been satisfactorily tested.
(2) Each experiment containing materials corrosive to reactor components or highly reactive with the reactor or experimental coolants shall be doubly contained.
(3) All experiments performed at the NBSR shall be reviewed and authorized in accordance with the specifications of Section 6.5.
Basis (1) In addition to all reactor experiments being designed not to fail from internal overheating or gas buildup, they shall also be designed to be compatible with their environment in the reactor. Specifically, their failures shall not lead to failures of the core structure or reactor fuel, or to the failure of other experiments. Also, reactor experiments shall be able to withstand the same transients that the reactor itself can withstand, such as loss of reactor cooling flows and startup accident.
The detonation of explosive or metastable materials within the reactor is not an intended part of the experimental procedure for the NBSR, but the 28
possibility of a rapid energy release shall be considered when these materials are present. Full testing of the container design shall be done.
(2) Experiments containing materials corrosive to reactor components or highly reactive with reactor or experimental coolants shall have an added margin of safety to prevent the release of these materials to the reactor coolant system. This margin of safety is provided by the double encapsulation, each container being capable of containing the materials to be irradiated.
(3) An independent technical review of experiments ensures the experiment will not reduce the reactor safety margin.
3.9 Facility Specific 3.9.1 Fuel Storage Applicability: Fuel element storage Objective: To prevent inadvertent criticality and maintain fuel element cladding integrity.
Specifications (1) All fuel elements or fueled experiments shall be stored and handled in geometry such that the calculated- kfr shall not exceed 0.90 under optimum conditions of water moderation and reflection.
(2) The water chemistry, level, and temperature in the spent fuel storage pool shall be maintained so as to ensure the integrity of the fuel elements.
Basis (1) To ensure that no inadvertent criticality of stored fuel elements or fueled experiments occurs, they shall be maintained in a geometry that ensures an adequate margin below criticality exists. This margin is established as a k*ff of no greater than 0.90 for the storage and handling of fuel or fueled experiments.
(2) The cooling of spent fuel elements in storage at the NBSR depends upon the decay heat of the elements, the volume of water in a storage pool, and any additional cooling, such as the use of pumps and heat exchangers. A storage pool is a stable environment, where water chemistry, temperature and level are easily monitored and the fuel is adequately shielded.
29
3.9.2 Fuel Handling 3.9.2.1 Within the Reactor Vessel Applicability: Fuel element latching Objective: To ensure that all fuel elements are latched between the reactor grid plates.
Specifications Following handling of fuel within the reactor vessel, the reactor shall not be operated until all fuel.elements that have been handled are inspected to determine that they are locked in their proper positions in the core grid structure. This shall be accomplished by one of the following methods:
(1) Elevation check of the fuel element with main pump flow.
(2) Rotational check of the element head in the latching direction only.
(3) Visual inspection of the fuel element head or latching bar.
Basis Each NBSR fuel element employs a latching bar, which shall be rotated to lock the fuel element in the upper grid plate. Following fuel handling, it is necessary to ensure that this bar is properly positioned so that an element cannot be lifted out of the lower grid plate, which could lead to a reduction in flow to the element after pump flow is initiated. Any of the three methods above may be used to verify bar position. Tests have shown that flow from a primary pump will raise an unlatched element above its normal position and thus will be detected by the pickup tool under flow conditions. The efficacy of rotational checks has been confirmed by visual inspections.
3.9.2.2 All Other Conditions Applicability: Refueling system Objective: To ensure the integrity of the fuel element cladding.
30
A fuel element shall not be removed from water in the reactor vessel unless the reactor has been shutdown for a period equal to or longer than one hour for each megawatt of operating power level.
Basis To ensure that a fuel element does not melt and release radioactive material, a time limit is specified before a fuel element may be removed from the vessel following reactor shutdown. Measurements carried out during reactor startup showed that for the hottest element placed dry in the transfer chute, 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after shutdown from 10 MW, the maximum temperature was only 550°F without auxiliary cooling. Extrapolation of these measurements shows that 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after shutdown from 20 MW, the maximum temperature for the hottest element would be less than 800NF without auxiliary coolant. For all other power levels below 20 MW the specified waiting time would result in even lower temperatures. This provides a substantial margin of safety from the safety limit.
31
4.0 Surveillance Recuirements Introduction The Surveillance frequencies denoted herein are based on continuing operation of the reactor.
Surveillance activities scheduled to occur during an operating cycle which can not be performed with the reactor operating may be deferred to the end of that current r~eactor operating cycle. Jf the reactor is not operated for a reasonable time, a reactor system or measuring channel surveillance requirement may be waived during the associated time period. Prior to reactor system or measuring channel operation, the surveillance shall be performed for each reactor system or measuring channel for which surveillance was waived. A reactor system or measuring channel shall not be considered operable until it is successfully tested. Surveillance intervals shall not exceed those defined in these Technical Specifications. Discovery of noncompliance with any of the surveillance specifications below shall limit reactor operations to that required to perform the, surveillance.
4.1 Reactor Core Parameters 4.1.1 Reactor Power Applicability: Reactor Safety System channels Objective: To ensure operability of the safety system channels.
Specifications (1) The reactor safety system channels shall be channel tested before each reactor startup, following a reactor shutdown that exceeds 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or quarterly.
(2) The reactor safety system channels shall be channel calibrated annually.
(3) A channel check of power range indication, with flow multiplied by AT, shall be performed weekly when the reactor is operating above 5 MW.
(4) Following maintenance on any portion of the reactor control or reactor safety systems, the affected portion of the system shall be tested before the system is considered operable.
Basis The channel tests, calibrations and flow AT comparison will ensure that the
'indicated reactor power level is correct. The power level channel calibration is performed by comparison of nuclear channels with the thermal power measurement channel (flow times AT). Because of the small AT (about 15'F at 20 MW), these calibrations will not be performed below 5 MW.
32
4.1.2 Reactivity Limitations Applicability: Core reactivity and shim arm worth.
Objective: To ensure that the reactor can be placed in a shutdown~condition at all times and that the safety limit shall not be exceeded.
Specifications (1) The excess reactivity (reference core conditions) shall be verified annually or following any significant changes in the core or shim arm configuration.
(2) The total reactivity worth of each shim arm and the regulating rod, and the shutdown margin shall be verified annually as described in these Technical Specifications, or following any significant change in the core or shim arm configuration.
Basis (1) Determining the core excess reactivity annually will ensure that the critical shim arm positions do not change unexpectedly.
(2) Measurements of reactivity worth of the shim arms and regulating rod over many years of operation have shown rod worths vary slowly as a result of absorber burnup, and only slightly with respect to operational core loading and experimental changes. An annual check shall ensure that adequate reactivity margins are maintained.
4.2 Reactor Control and Safety Systems 4.2.1 Shim Arms Applicability: Shim arm motion Objective: To ensure proper shim arm reactivity insertion.
Specifications (1) The withdrawal and insertion speeds of each shim arm shall be verified semiannually.
(2) Scram times of each shim arm shall be measured semi-annually.
33
Basis The shim arm drives are constant speed mechanical devices. A reactor scram is aided by a spring that opposes drive motion during shim arm withdrawal.
Withdrawal and insertion speeds or scram time should not vary except as a result of mechanical wear. The surveillance frequency is chosen to provide a significant margin over the expected failure or wear rates of these devices.
4.2.2 Reactor Safety System Channels Applicability: Required instrument channels Objective: To ensure reliability of protective action for nuclear and process variables.
Specifications The Scram and Confinement Channels shall have the surveillance requirements shown in Table 4.2.2.
Table 4.2.2 Surveillance Requirements for the Scram and Confinement Channels Channel Action Required Surveillance Required (1) High Flux level Scram X,A (2) Short period below 5% rated power Scram X,A (3) Low reactor vessel D2 0 level Scram X,A (4) Low flow reactor outlet Scram X,A (5) Low flow reactor inner or outer plenum Scram X,A (6) Manual (outside of the Control Room) Scram X,A (7) Manual Scram X,A (8) Normal Air Exhaust Activity High Major Scram X,A (9) Irradiated Air Activity High Major Scram .X,A (10) Stack Air Activity High Major Scram X,A (11) Reactor Coolant Outlet Temperature Rundown X,A X - Channel test before startup after a shutdown of longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or quarterly.
A - Annual Channel Calibration.
Basis To ensure that instrument failures do not go undetected, frequent surveillance of the listed channels is required and operating experience has shown these frequencies to be adequate to ensure channel operability.
34
4.3 Coolant Systems 4.3.1 Primary and Secondary Applicability: Primary fluid systems Objective: To prevent degradation of primary system materials.
Specifications (1) The primary cooling system relief valve shall. be tested annually.
(2) Major additions, modifications, or repairs of the primary cooling system or its connected auxiliaries shall be tested before the affected portion of the system is placed into service.
(3) The D2 concentration in the helium sweep gas shall be verified every five (5) years.
Basis (1) The frequency for testing the pressure at which the relief valve opens is consistent with industry practices on this type of valve for clean water service conditions.
(2) Major additions, modifications, or repairs of the primary system shall be either pressure tested or checked by X-ray, ultrasonic, gas leak test, dye penetrants or other methods.
(3) Recombination of deuterium and oxygen is accomplished primarily by the reactor. Operational experience and data suggests that the specified frequency is appropriate for verifying D2 levels.
4.3.2 Emergency Core Cooling System Applicability: Emergency Core Cooling System Objective: To ensure an emergency supply of coolant.
Specifications (1) Control valves in the emergency core cooling system shall be exercised quarterly.
(2) The operability of the emergency sump pump, using either heavy or light water, shall be tested annually.
35
(3) The light water injection valves shall be exercised semi-annually.
Basis The equipment in this system is not used in the course of normal operation, so its operability shall be verified periodically. The frequencies are chosen so that deterioration or wear would not be expected to be an important consideration. Moreover, the frequency should be sufficient to ensure that the pumps and valves will not fail because of corrosion buildup or other slow acting effects during extended periods of standby operation. Control and injection valves specified are those leading to or from the D20 emergency cooling tank.
4.3.3 Moderator Dump System Applicability: Moderator dump valve Objective: To provide a backup shutdown mechanism.
Specification The Moderator Dump valve shall be cycled annually.
Basis The moderator dump valve is of proven dependable design. Operating the dump valve annually is and has been a reliable predictor of performance.
4.4 Confinement System Applicability: Confinement building and components Objective: To ensure the continued integrity and reliability of the confinement building.
Specifications (1) A test of the operability of the confinement closure system shall be performed quarterly. The trip feature shall be initiated by each of the radiation monitors that provides a signal for confinement closure, as well as by the manual major scram switch. A radiation source shall be used to test the trip feature of each of the radiation monitors annually.
(2) An integrated leakage test of the confinement building shall be performed annually at a gauge pressure of at least 6.0 inches of water and a vacuum of at 36
least 2.0 inches of water, with a maximum allowable leak rate of 24 cfm/inch of water.
(3) Any additions, modifications, or maintenance to the confinement building or its penetrations shall be tested to verify that the building can maintain its required leak tightness.
Basis (1) The confinement closure system is initiated either by a signal from the confinement building gaseous effluent radiation detectors or manually by the major scram switch and each of these signal sources is used to initiate the test. In addition, each radiation detector is tested for proper response to ionizing radiation.
(2) A preoperational test program was conducted to measure the representative leakage characteristics at values of a gauge pressure of +7.5 inches of water and
-2.5'inches of water. The specified test pressures and vacuums are acceptable because past tests have shown leakage rates to be linear with applied pressures and vacuums.
(3) Changes in the building or its penetrations shall be verified to withstand specified test pressures; therefore, tests shall be performed before the building Confinement System can be considered to be operable.
4.5 Ventilation System Applicability: Normal and Emergency ventilation system Objective: To ensure the operability of the ventilation system.
Specifications (1) An operability test of the emergency exhaust system, including the building static pressure controller and the vacuum relief valve, shall be performed quarterly.
(2) An operability test of the controls in the Emergency Control Station and an inspection to determine that all instruments in the Emergency Control Station are indicating normally shall be made monthly.
(3) The efficiency of the absolute filters in both normal and emergency exhaust systems shall be verified biennially. It shall be verified that the absolute filters remove 99% of particles with diameters of 0.3 gm and greater.
(4) It shall be verified biennially that the charcoal filter banks in the emergency exhaust and recirculation systems have a removal efficiency of 99% for Iodine.
37
Basis (1) The emergency ventilation system depends on the proper operation of the emergency exhaust system fans, valves, and filters, which are not routinely in service. Because they are not continuously used, their failure rate as a result of wear shouldbe low. Since they are not being used continuously, their condition in standby shall be checked sufficiently often to ensure that they shall function properly when needed. An operability test of the active components of the emergency exhaust system quarterly will ensure that each component will be operable if an emergency condition should arise. The quarterly frequency is considered adequate since this system receives very little wear and since the automatic controls are backed up by manual controls..
(2) The Emergency Control Station instrumentation must be operable to monitor the reactor's condition in the event the Control Room becomes uninhabitable.
Therefore, monthly checks of the instrumentation have been shown to be adequate to ensure operability.
(3) The biennial verification of the absolute filter efficiency has been shown to be appropriate for filters subject to continuous air flow. Because the absolute filters in the emergency exhaust system will be idle except during brief periods of fan operation, deterioration should be much less than for filters subjected to continuous air flow where dust overloading and air breakthrough are possible after long periods of use. Therefore, a biennial frequency should be adequate to detecting filter deterioration.
(4) Biennial verification of filter banks, which are subjected to flow only during brief periods of fan operation ensures that the filters will perform as analyzed.
4.6 Emergency Power System Applicability: Emergency electrical power supply equipment Objective: To ensure emergency power for vital equipment after the reactor is shutdown.
Specifications (1) Each diesel generator shall be tested for automatic starting and operation quarterly.
(2) Should one of the diesel generators become inoperative, the operable generator shall be started monthly.
(3) All emergency power equipment shall be tested under a simulated complete loss of outside power annually.
38
(4) The voltage and specific gravity of each cell of the station battery shall be tested once every annually. A discharge test of the entire battery shall be performed 5 years.
Basis (1) The NBSR is equipped with two diesel power generators, each capable of supplying full emergency load; therefore, only one of the generators shall be required. The diesel generators have proven to be very reliable over decades of service. The quarterly test frequencies are consistent with industry practice and are considered adequate to ensure continued reliable emergency power for emergency equipment.
(2) This testing frequency of the operable generator will ensure that at least one of the required emergency generators will be operable.
(3) An annual test of the emergency power equipment under a simulated complete loss of outside power will ensure the source will be available when needed.
(4) Specific gravity and voltage checks of individual cells are the accepted method of ensuring that all, cells are in satisfactory condition. The annual frequency for these detailed checks is considered adequate to detect any significant changes in the ability of the battery to retain its charge. During initial installation, the station battery was discharge tested to measure its capacity. Experience has shown that repeating this test at the specified interval is adequate to detect deterioration of the cells.
4.7 Radiation Monitoring System and Effluents 4.7.1 Monitoring System Applicability: Radiation monitoring equipment Objective: To operability of radiation monitors.
Specifications (1) The gaseous effluent monitors for normal air, irradiated air and stack air shall be channel tested before startup, after a shutdown of longer than twenty-four (24) hours, or quarterly. Each of the above air monitors shall be channel calibrated annually.
(2) The fission products monitor shall be channel tested monthly and channel calibrated annually.
39
(3) The secondary coolant activity monitor shall be channel tested monthly and channel calibrated annually. Analysis of the secondary water for tritium shall be conducted monthly. Should the secondary cooling water activity monitor be inoperable, analysis for tritium shall be performed daily.
(4) The Area Radiation Monitors shall be channel tested monthly and channel calibrated annually.
(5) For primary tritium concentrations of less than or equal to 4 Ci/l, the primary water shall be sampled annually. For tritium concentrations of greater than 4 Ci/l, the primary water shall be sampled quarterly.
Basis (1) A channel test ensures the monitoring systems will respond correctly to an input signal. An annual channel calibration ensures the detection and response capability of the channels.
(2) A channel test monthly is considered reasonable for a device of this type.
A channel calibration annually is considered adequate to ensure that a significant deterioration in accuracy from its normal setting does not occur.
(3) The secondary cooling water activity monitor usually gives the first indication of a primary-to-secondary leak. This monitor employs a simple radiation detector, the operability of which has been shown to be very good. Therefore, a monthly channel test is considered reasonable. An annual channel calibration frequency is considered adequate to ensure that a significant deterioration in accuracy from its normal settings does not occur. Assuming operation of the secondary cooling water activity monitor and no detectable loss of primary coolant, a monthly sampling for tritium should be adequate to detect small tritium leaks. If the secondary cooling water activity monitor is out of service, then sampling is the primary means of leak detection and more frequent sampling is required.
A daily frequency is judged adequate since large leaks would still be detected by a decreasing level in the D2 0 storage tank.
(4) The area radiation monitors (ARM) may give the first indication of a radioactive release resulting from an experiment or reactor malfunction. A monitor employs a simple radiation detector, the operability of which has been shown to be very good over many years. Therefore, a monthly channel test is considered reasonable. These monitors are primarily used to detect an increase in activity over that which has previously existed, so they are normally set at some reasonable value above background and their absolute accuracy is not critical. Hence, the annual calibration 40
frequency is considered adequate to ensure that a significant deterioration in accuracy does not occur.
(5) The primary tritium concentration can be carefully monitored by annual analysis of the primary water. All new water is tested prior to addition to the system. Operational experience and well established neutron activation principles provide a good basis for predicting tritium buildup in the primary. Increasing the sampling frequency after concentrations exceed 4 Ci/l will ensure that the tritium concentration limit is not exceeded.
4.7.2 Effluents Applicability: Environmental monitoring sampling program Objective: To minimize radiation exposures outside of the confinement building.
Specifications (1) Water, soil and vegetation samples shall be collected quarterly.
(2) Thermoluminescent dosimeters shall be collected quarterly.
(3) Air sampling shall be done quarterly.
Basis (1) Collecting and analyzing the water, soil and vegetation samples on a quarterly basis will provide information that environmental limits are not being exceeded.
(2) Collecting and analyzing the thermoluminescent dosimeters on a quarterly basis will provide information that radiation limits are,not being exceeded.
(3) Sampling the air on a quarterly basis will provide information that release limits are not being exceeded.
4.8 Experiments Applicability: Irradiation Experiments Objective: To ensure that experiments conform to the limits of the specifications of Section 3.8.
41
Specification The reactivity worth of any experiment installed in a pneumatic transfer tube, or in any other NBSR irradiation facility insidethe thermal shield shall be estimated before reactor operation with said experiment.
Basis Estimation. of the reactivity worth based either on calculation or on previous or similar measurements ensures that the experiment is within authorized reactivity limits.
42
5.0 Design Features 5.1 Site Description Specifications (1) The NBSR complex is located within the National Institute for Standards and Technology grounds and access to the reactor shall be controlled.
(2) The reactor shall have a minimum exclusion radius of 400 meters, as measured from the reactor stack.
Basis The location and government ownership of the NBSR site ensures auxiliary services including fire and security are available. The exclusion radius of 400 meters is the distance on which all unrestricted doses are calculated. Should this value decrease for any reason, a recalculation of the unrestricted doses would be necessary. Access to the reactor complex is controlled either by the facility staff or by NIST Police.
5.2 Reactor Coolant System Specifications (1) The reactor coolant system shall consist of a reactor vessel and a single cooling loop containing heat exchangers, pumps, and valves.
(2) The reactor vessel shall be designed in accordance with Section VIII of the American Society of Mechanical Engineers (ASME) Code for Unfired Pressure Vessels. The vessel shall be designed for 50 psig and 250°F. The heat exchangers shall be designed for 100 psig and a temperature of 150TF. The connecting piping shall be designed for 125 psig and a temperature of 150 0 F.
Basis (1) The reactor coolant system has been described and analyzed as a single cooling loop system containing heat exchangers, pumps and valves.
(2) The design temperature and pressure of the reactor vessel and other primary system components provide adequate margins over operating temperatures and pressures. The reactor vessel was designed to Section VIII, 1959 Edition of the ASME Code for Unfired Pressure Vessels. Any subsequent changes to the vessel should be made in accordance with the most recent edition of this Code.
43
5.3 Reactor Core and Fuel Specifications (1) The 20 MW reactor core consists of 30, 3.0 x 3.3 inch (7.6 x 8.4 cm) MTR curved plate-type fuel elements. The NBSR MTR-type fuel element shall be such that the central 7 inches of the fuel element contains no fuel. The middle 6 inches of the aluminum in the unfueled region of each plate shall have been removed.
(2) The side plates, unfueled outer plates, and end adaptor castings of the fuel element shall be aluminum alloy.
(3) The fuel plates shall be U30 8 dispersed in a matrix of aluminum, clad in aluminum alloy Basis (1) The neutronic and thermal hydraulic analysis was based on the use of 30 NBSR MTR-type thirty-four (34) plate fuel elements. The NBSR fuel element has a 7 inch centrally located unfueled area,. in the open lattice array. The middle 6 inches of aluminum in the unfueled region has been removed. The analysis requires that the fuel be loaded in a specific pattern. Significant changes in core loading patterns would require a recalculation of the power distribution to ensure that the CHFR would be within acceptable limits.
(2) and (3) The aluminum clad dispersion fuels used in the MTR fuel elements have a 50 year record of reliability at many research reactors.
44
6.0 Administrative Controls 6.1 Organization The Director, NIST Center for Neutron Research shall be the licensee for the NBSR.
The NBSR shall be under the direct control of the Chief, Reactor Operations and Engineering. The Chief, Reactor Operations and Engineering shall be accountable to the Director, NCNR for the safe operation and maintenance of the NB SR.
.6.1.1 Structure The management for operation of the NBSR shall consist of the organizational structure as shown in Figure 6.1.
6.1.2 Responsibility Responsibility for the safe operation of the NBSR shall be with the chain of command established in Figure 6.1. Individuals at the various management levels shall be responsible for the policies and operation of the NBSR, for safeguarding the public and facility personnel from undue radiation exposures, and for adhering to all requirements of the operating license and technical specifications.
6.1.3 Staffing (1) The minimum staffing when the reactor is not secured shall be:
(a) A Reactor Operator in the Control Room.
(b) A Reactor Supervisor present within the reactor exclusion area.
(c) An SRO present in the facility whenever a reactor startup is performed, fuel is being moved within the reactor vessel, experiments are being placed in the reactor vessel or a recovery from an unplanned or unscheduled shutdown or a significant power reduction.
(2) A list of reactor facility personnel by name and telephone number shall be available to the reactor operator in the Control Room. This list shall be updated annually. The list shall include:
(a) Management personnel.
(b) Health Physics personnel.
(c) Reactor Operations personnel.
45
6.1.4 Selection and Training of Personnel The selection, training and requalification of operations personnel shall meet or exceed the requirements of the American National Standard for Selection and Training of Personnel for Research Reactors (ANSI/ANS 15.4-2007.).
Qualification and'requalification of licensed reactor operators shall be performed in accordance with a Nuclear Regulatory Commission (NRC) approved program.
6.1.4.1 Selection of Personnel Minimum educational and experience requirements for those individuals who have line responsibility and/or authority for the safe operation of the facility are as follows:
(1) Chief, Reactor Operations and Engineering The Chief, Reactor Operations and Engineering shall have an advanced college degree in engineering or a science related field, or equivalent experience and training. Equivalent experience for this position requires five years experience in a responsible position in reactor operations or reactor engineering, including one year experience in senior reactor facility management or supervision.
(2) Chief, Reactor Operations The Chief, Reactor Operations shall have a college degree in engineering or a science related fields or a combined seven years of college level education and nuclear reactor experience. Three years of reactor operations experience is required. The individual shall demonstrate the capability to be an SRO at the NBSR.
(3) Reactor Supervisor (a) Four years experience in reactor operations, including experience in the operation and maintenance of equipment and in the supervision of technicians and/or senior reactor operators.
(b) A high school diploma or equivalent and formal training in reactor technology and reactor operations. An additional two years of experience may be substituted for education and formal training.
46
(c) Shall have been a licensed as a Senior Reactor Operator at the NBSR.
(4) Senior Reactor Operator A Senior Reactor Operator shall have a high school diploma or equivalent and one year experience in reactor operations. The individual shall be licensed as a Senior Reactor Operator.
(5) Reactor Operator A Reactor Operator shall have a high school diploma or equivalent and six months of technical training. The individual shall be licensed as a Reactor Operator.
(6) Auxiliary Operator An Auxiliary Operator shall have a high school diploma or equivalent.
6.1.4.2 Training of Personnel (1) A training program shall be established to maintain the overall proficiency of the Reactor Operations organization. This program shall include components for both initial licensing and requalification, consistent with ANSI/ANS 15.4-2007.
(2) The training program shall be under the direction of the Chief, Reactor Operations and/or the Chief, Reactor Operations and Engineering.
(3) Records of individual reactor operations staff members' qualifications, experience, training, and requalification shall be maintained as described the specification of Section 6.8.2.
6.2 Review and Audit The NCNR Safety Evaluation Committee (SEC) is established to provide an.
independent review of NCNR reactor operations to ensure the facility is operated and maintained in such a manner that the general public, facility personnel and property shall not be exposed to undue risk.
The NCNR Safety Assessment Committee (SAC) is established to provide an independent review or audit of NCNR reactor operations. This audit is to ensure that safety reviews and reactor operations are being performed in accordance with regulatory requirements and public safety is being maintained.
47
6.2.1 Composition and Qualifications The Director, NCNR, upon recommendation of the Chief, Reactor Operations and Engineering, shall appoint all members and alternates to the SEC. The SEC shall be composed of no less than four members and membership terms are indefinite and at the discretion of the Director. Members and alternates shall be selected on their ability to provide independent judgment and to collectively provide a broad spectrum of expertise in reactor technology and operation. At least two members shall be from the NCNR and one from Health Physics. Unless otherwise designated by the Director, the SEC shall include the following ex officio members: the Chief, Reactor Operations; Chief, Reactor Engineering; and the Senior Supervisory Health Physicist.
6.2.2 Safety Evaluation Committee Charter andRules The SEC shall conduct its review functions in accordance with a written charter and the charter shall be consistent with ANSI/ANS 15.1-2007. This charter shall include provisions for:
(1) Meeting frequency.
(2) Voting rules.
(3) Quorums.
(4) Method of submission and content of presentation to the committee.
(5) Use of subcommittees.
(6) Review, approval and dissemination of minutes.
6.2.3 SEC Review Function The responsibilities of the SEC, or a designated subcommittee thereof, shall include but are not limited to the following:
(1) Review proposed tests or experiments significantly different from any previously reviewed or which involve any questions pursuant to 10 CFR 50.59 and determine whether proposed changes or reactor tests or experiments have been adequately evaluated, documented, approved and recommendations sent to the NCNR director for action.
(2) Review the circumstances of all events described in this section and the measures taken to preclude a recurrence and provide recommendations to the NCNR director for action.
(3) Review proposed changes to the NBSR facility equipment or procedures when such changes have safety significance, or involve an amendment to the facility license, a change in the Technical Specifications incorporated 48
a in the facility license, or questions pursuant to 10 CFR 50.59 and provide recommendations to the NCNR director for action. Review SAC reports.
(4) The SEC shall on a biennial basis review its charter and recommend to the NCNR director any changes necessary to ensure the continued effectiveness of the charter.
6.2.4 SEC Audit Function The responsibility of the SEC, or a designated subcommittee thereof, shall include but not be limited to the following audits:
(1) Facility operations at a frequency of once per calendar year, not to exceed fifteen (15) months.
(2) Results of actions taken to correct deficiencies that affect reactor safety at a frequency of once per calendar year, not to exceed fifteen (15) months.
(3) Requalification program at a frequency of once every other calendar year, not to exceed thirty (30) months.
(4) NBSR Emergency Plan at a frequency of once every other calendar year, not to exceed thirty (30) months.
6.2.5 Safety Assessment Committee (SAC)
The Safety Assessment Committee (SAC) shall be composed of at least three senior technical personnel who collectively provide a broad spectrum of expertise in reactor technology. The Committee members shall be appointed by the Director, NIST Center for Neutron Research. Members of the SAC shall not be regular employees of NIST. At least two members shall pass on any report or recommendation of the Committee. The SAC shall meet annually and as required. The Committee shall review or audit the NCNR reactor operations and the performance of the SEC. The SAC shall report in writing to the Director, NIST Center for Neutron Research.
6.3 Radiation Safety, The NIST Reactor Health Physics Group shall be responsible to support the licensee in the implementation of the radiation protection and ALARA program at the reactor using the guidelines of the American National Standard for Radiation Protection at Research Reactor Facilities, ANSI/ANS 15.11-2004. The NIST Reactor Health Physics Group leader shall report to the Director, NIST Center for Neutron Research for radiological matters concerning the NB SR.
49
0 6.4 Procedures Written procedures shall be prepared, reviewed and approved prior to initiating any of the activities listed in this section. The safety significant changes (determined by the Chief, Reactor Operations and Engineering or the Chief, Reactor Operations) to operating procedures shall be reviewed by the SEC and approved by the Chief, Reactor Operations and Engineering or the Chief, Reactor Operations. Such reviews and approvals shall be documented in a timely manner. Activities requiring written procedures are:
(1) Startup, operation, and shutdown of the reactor.
(2) Fuel loading, unloading, and fuel movement within the reactor vessel.
(3) Surveillance checks, calibrations, and inspections of equipment required by the technical specifications that may have an effect on reactor safety.
(4) Personnel radiation protection, consistent with applicable regulations or guidelines. The procedures shall include management commitment and programs to maintain exposures and releases as low as is reasonably achievable in accordance with the guidelines of ANSI/ANS 15.11-2004.
(5) Conduct of irradiations and experiments that could affect reactor safety or core reactivity.
(6) Implementation of required plans such as emergency or security plans.
(7) Use receipt, and transfer of byproduct material, if appropriate.
Substantive changes to the procedures listed above shall be made effective only after documented review by the SEC and approval by the Chief, Reactor Operations and Engineering or the Chief, Reactor Operations. Minor modifications or temporary deviations to the original procedures which do not effect reactor safety or change their original intent may be made by the Reactor Supervisor in order to deal with special or unusual circumstances or conditions. Such changes shall be documented and reported within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the next working day to the Chief, Reactor Operations and Engineering or the Chief, Reactor Operations.
6.5 Experiment Review and Approval Experiments shall be carried out in accordance with established and approved procedures. The following provisions shall be implemented:
(1) All new experiments or class of experiments shall be reviewed by the SEC and approved in writing by the Director, NCNR.
50
(2) Substantive changes to previously approved experiments shall be made only after review by the SEC and approved in writing by the Director, NCNR. Minor changes that do not significantly alter the experiment safety envelope may be made in accordance with the SEC charter.
6.6 Required Actions 6.6.1 Actions to Be Taken in the Event the Safety Limit is Exceeded (1) The reactor shall be shutdown and reactor operations shall not be resumed until authorized by the NRC.
(2) An immediate notification of the occurrence shall be made to the Chief, Reactor Operations and Engineering and the Chief, Reactor Operations.
The Chief, Reactor Operations and Engineering shall inform the NCNR director.
(3) Reports shall be made to the NRC in accordance with the specifications of Section 6.7.2. A written report 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. The report shall be prepared by the Chief, Reactor Operations and Engineering and submitted to the SEC for review. The SEC shall review the report and submit it to the Director, NIST Center for Neutron Research director for approval. The Director shall then submit the report to the NRC.
6.6.2 Actions to Be Taken in the Event of an Occurrence of the Type Identified in Section 6.7.2 other than a Safety Limit Violation (1) The reactor shall be secured and the Chief, Reactor Operations and Engineering and the Chief, Reactor Operations notified.
(2) Operations shall not resume unless authorized by the Chief, Reactor Operations and Engineering.
(3) The SEC shall review the occurrence at their next scheduled meeting.
(4) Where appropriate and in addition to the initial notification, a report shall be submitted to the NRC in accordance with the specifications of Section 6.7.2.
51
a 6.7 Reports 6.7.1 Annual Operating Report A report shall be submitted annually to the NRC and include:
(1) A brief summary of operating experience including the energy produced by the reactor and the hours the reactor was critical.
(2) The number of unscheduled shutdowns, including reasons therefore.
(3) A tabulation of major preventative and corrective maintenance operations having safety significance.
(4) A brief description, including a summary of the safety evaluations, of changes in the facility or in procedures and of test and experiments carried out pursuant to 10 CFR 50.59 (2007).
(5) A summary of the nature and amount of radioactive effluents released or discharged to the environs and the sewer beyond the effective control of the licensee as measured at or prior to the point of such release or discharge.
(6) A summary of environmental surveys performed outside the facility.
(7) A summary of significant exposures received by facility personnel and visitors.
6.7.2 Special Reports In addition to the requirements of applicable regulations, and in no way substituting therefore, reports shall be made by the Director, NCNR or the Chief, Reactor Operations and Engineering, to the NRC as follows:
(1) There shall be 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, facsimile, or other NRC approved method, to the NRC Operations Center and confirmed in writing by facsimile or similar conveyance, to be followed by a written report within 14 days that describes the circumstances associated with any of the following:
(a) Accidental release of radioactivity above applicable limits in unrestricted areas, whether or not the release resulted in property damage, personal injury, or exposure.
(b) Violation of the safety limit.
52
(c) Operation with a safety system setting for required systems less conservative than the Limiting Safety System Setting values.
(d) Operation in violation of a Limiting Condition for Operation (LCO) established in the technical specifications unless prompt remedial action is taken as permitted by exception statements.
(e) A reactor safety system component malfunction which renders or could render the reactor safety system incapable of performing its intended safety function. If the malfunction or condition is caused by maintenance, then no report is required.
Where components or systems are provided in addition to those required by the technical specifications, the failure of the extra components or systems is not considered reportable.
(f Any change in reactivity greater than one dollar ($1.00) that could adversely affect reactor safety.
(g) An observed inadequacy in the implementation of either administrative or procedural controls, such that the inadequacy could have caused the existence or development of conditions which could result in operations of the reactor outside the safety limit.
(h) Abnormal and significant degradation in reactor fuel, cladding, coolant boundary, or confinement boundary (excluding minor leaks) where applicable.
(2) There shall be a report submitted in writing within 30 days to the NRC, Document Control Desk, Washington D.C. 20555, of:
(a) Permanent changes in the facility organization involving the Director, NCNR, or the Chief, Reactor Operations and Engineering.
(b) Significant changes in the accident analyses as described in the Safety Analysis Report.
6.8 Records 6.8.1 Records to be Retained for a Period of at Least Five Years or for the Life of the Component Involved if Less than Five Years Records of this section may be in the form of logs, data sheets, or other retrievable forms. The required information may be contained in single or multiple records, or a combination thereof Annual reports as described in the 53
specifications of Section 6.7.1, to the extent the reports contain all of the required information, may be used as a record of the following:
(1) Normal reactor operation logs, not including supporting documents such as checklists and log sheets. (Supporting documents shall be retained for a period of at least one year.)
(2) Principal maintenance activities.
(3) Special Reports.
(4) Surveillance activities required by these Technical Specifications.
(5) Solid radioactive waste shipped off-site.
(6) Fuel inventories and transfers.
6.8.2 Records to be Retained for at Least One Operator Licensing Cycle Records of retraining and requalification of licensed operations personnel shall be maintained for the period the individual is employed or until the license is renewed.
6.8.3 Records to be Retained for the Life of the Reactor Facility (1) Gaseous and liquid radioactive effluents released to the environs.
(2) Off-site environmental monitoring surveys required by these Technical Specifications.
(3) Radiation exposure for all personnel monitored.
(4) Drawings of the reactor facility.
54
Director,-
NIST Center for Neutron Research
I I
Chief, Reactor-Operations and Engineering' I
hief,. ;[ Chief, .
I~Cl Operations Reactor Engineering Reactor i H Supervisor - NBSR Health Physics ... I Reactor I
Senior Rea Ctor Operator Reactor )perator Auxiliar Operator I---------------------- ----------------
Safety Assessment Committee (SAC)
Recommendations and Technical Advice Administrative Reporting Channels Figure 6.1 55'-