ML19345C850
| ML19345C850 | |
| Person / Time | |
|---|---|
| Site: | National Bureau of Standards Reactor |
| Issue date: | 11/30/1980 |
| From: | NATIONAL INSTITUTE OF STANDARDS & TECHNOLOGY (FORMERL |
| To: | |
| Shared Package | |
| ML19345C846 | List: |
| References | |
| NUDOCS 8012080491 | |
| Download: ML19345C850 (28) | |
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Proposed Changes To NBSR Technical Specifications License #TR-5 Docket #50-184 November 1980 l l 80120804T
. . _ . - . _ . . - ._. . - - . . . . . . . . _. - - ..__
s * .. 1 i Proposed Changes to NBSR Technical Specifications License No. TR-5 1 1 , _
-
- 1. Introduction Change 10'Mwt to 20 Mwt..
'
- 2. Section 2.1
- Delete entire section and replace by attached new section.
- 3. Section 2.2 1
Delete entire section and replace by attached new section.
- 4. Section 3.1
. -
- Delete entire section and replace by attached new section.
- 5. Section 3.2 e
+ Revise basis as per attached.
- 6. Section 3.4 Change specification 3.4b to read as follows:
l The reactivity insertion rate, using all four shim safety arms does not exceed 5.0 x 10-4 Ao/sec. Delete High Differential Temperature from Table 1. l Revise basis as per attached. t
- 7. Section 3.5 i
l, Change references in basis as per attached.
- 8. Section 3.6 Revise basis as per attached.
'
- 9. Section 3.7 Delete entire section and replace by attached new section.
- 10. Section 3.8
, Revise basis as per attached.
- 11. Section 3.9
, Revise basis as per attached.
- 12. Section 3.11 Revise basis as-per attached.
4 i t f e M --y -My - t -"+g- e y wr ---y * .-+-p+ m------??4w---'w--7+*- w
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- 13. Section 4.0 Revise basis as per attached.
- 14. Section 5.1 Specification 5.lc, change 6.5" to 6.0".
Specification 5.1d, delete last sentence. Revise basis as per attached.
- 15. Section 5.4 Revise basis as per attached.
- 16. Section 5.7 Specification 5.7c, change 99.9% to 99%.
Specification 5.7d, change 0.1% to 1%. Revise basis as per a.tached.
- 17. Section 6.1 Add reference 5 to references 2, 3, and 4.
(5) Final Safety Analysis Report, NBSR 9, Addendum 1, Sections 2 and 3, November 1980. ! 18. Section 6.3 Typographical' error, uranium-aluminum oxide should read aluminum-uranium oxide.
- 19. Section 7.0 Replace abnormal by reportable wherever it appears.
- 20. -tion 7.2 l Change monthly to quarterly.
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l t . I 2.1 Safety Limits .
-Applicability: Applies to reactor power, reactor coolant system -
flow and temperature. i
,
Objective: To maintain the integrity of the fuel cladding and i i prevent the release of significant amounts of fission products. Specifications: Reactor power, coolant system flow, and inlet ! temperature shall not exceed the limits shown in figures 2.1 and 2.2. (Note 1) Note 1 The reactor may be operated at power levels of up to 10kw with reduced flow (including no flow) if decay heat is insufficient to cause significant heating of the reactor coolant. Basis Maintaining the integrity of the fuel cladding requires that the - cladding remain below its melting temperature. For all plant
!
operating conditions which avoid a departure from nucleate boiling, cladding temperatures remain substantially below the melting temperature. . Conservative calculations ( } have shown that limiting + combinations of reactor power, reactor coolant system flow and temperature to values less than the Safety Limits will prevent cladding burnout. References (1) Final Safety Analysis Report, NBSR 9, Addendum 1, Section 3.2.2, 1 November 1980.
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- 2.2 Limiting Safety System Settings Applicability: Applies to maximum settings for instruments monitoring' safety limit parameters.
Obj ective: To assure protective action if any of the principal process variables should approach a safety limit. Specification: The maximum or minimum safety system trip settings shall be as follows: ! Maximum Safety System Setting i Thermal Power, % 130 Reactor Inlet 127 (rundown) Temperature F Coolant Flow (Minimum Setting) - (GPM) (Note 1) Inner Plenum Outer Plenum
- For Operation up to 1GMw 650 2400 4
For Operation up to 20Mw 1250 4700 Note 1 4 May be bypassed during periods of reactor operation (up to 10kw) when a reduction in safety limit values is permitted (Section 2.1 of these specifications).
.
Basir. At the values established, the safety system settings provide a significant margin from the safety limits. Even in the extremely unlikely-event that all three parameters, reactor power, coolant flow and inlet temperature, simultaneously reach their safety system settings, the burnout-ratio is at least 1.38. For all other 4 conditions the burnout ratio is considerably higher ( . This will assure that any reactor transient caused by equipment malfunction- . or operator error will be terminated well before the safety limits are reached. Overall uncertainties in process instrumentation have been incorporated in Limiting Safety System Settings values. References (1) Final Safety Analysis Report, NBSR 9, Addendum 1, Section 3.2.2, - November 1980.
- .
_
. - _ . .
. _ . . .
,- 3.1 Confinement System Applicabilitv: Applies to the operating status of the confinement
' building.-
' Objective: To ensure confinement integrity when it is required.
Specifications: Confinement integrity shall be maintained when any of the following conditions-exist: 4
- a. 'The reactor is operating.
.
- b. Changes of components or equipment within the confines of the .
~
thermal shield, other than rod drop tests or movement of , single experiments, are being made which could cause a change in reactivity. i c. Movement of irradiated fuel, which contains significant fission product inventories outside a sealed container or system is being conducted. In addition:
- d. No maintenance which causes a breach in confinement shall be performed unless the reactor decay heat is below the value specified in Section 3.7(b) of these specifications.
Basis
, -The confinement system is a major engineering safeguard since it serves as the final physical barrier'to confine radioactive
, particles and gasses following accidents.(1,2) Changes in the core involving such operations as fuel handling I or control rod repairs affect the reactivity of the core and could reduce the shutdown margin of the reactor. Since these changes affect the status of the core, confinement integrity is required. When the reactor is shut down and single experiments are to be inserted or-removed, confinement integrity is not required. The
! reactor is normally shut.down by a substantial reactivity margin (calculated to be at least 14% 40).( ) Experiments will be inserted
' and removed one at a time; hence, the total reactivity change in any single operation will be limited to the specified maximum worth of 0.5% Ao for any single experiment.( ) Even-if the sequential movement of all experiments (including " fixed" experiments) were postulated, .the maximum potential reactivity insertion would not , exceed the 2.6% Ao (see Section 4.0 of these specifications) worth of all experiments permitted in the reactor at any time. Even under this circunstance, the shutdown margin would still be substantial.
-
!
, .. - -
.- .- . - . ~ . .
o-
'
2, Even when the reactor is shut down, irradiated fuel which , contains~significant fission product inventories (sufficient to allow specification 3.11 to be exceeded should the element fail)
-
i poses a _ potential hazard should its cladding be violated when it is not otherwise. contained (e.g.,'during transit-or during sawing of
~
aluminum end pieces). When irradiated fuel is -ontained within a closed system, such as the reactor vessel, transfer lock of the refueling system, sealed shipping cask, etc., these serve as a secondary barrier to fission product release and confinement integrity is not required. Should. it be necessary to perform maintenance which prevents normal rapid closing of the confinement, then all possibilities of' fission product release from fuel melting must be precluded. For this reason, no such maintenance should be permitted unless the .< reactor has been shutdown a specified time. (See Section 3.7b)~
References:
'(1) Final Safety Analysis Report, NBSR 9, Section 13.6, page 13-15,
. April 1966.
(2) Final.. Safety Analysis Report, NBSR 9, Addendum 1, Sections 2 and 3, November 1980. ( Final Safety Analysis Report, NBSR 9, Section 4.6.6.1,
- page 4-18, April 1966.
!
(4) Final Safety Analysis Report, Supplement A. NBSR 9A, Response No. 22, page 22-2, December 16, 1966. (5) Final Safety Analysis Report, NBSR 9, Section 4.6.5, page 4-17, April 1966. 1 , "
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. ___ ,__ , _'
. . . . .-. .. . . - . .
I Section 3.2 Readtor Coolant System Basis Loss of flow accidents have been analyzed for the NBSR assuming a single shutdown cooling pump is operable. Under this condition, the hot spot of the hottest place remains below 160 F.( The NBSR has also been analyzed, assuming no shutdown cooling flow.( } For this case the maximum fuel plate temperature would be less than 500 F, well below the temperature that would cause any damage. Requiring a shutdown pump to be operable will assure that fuel place temperatures following loss of flow will be near or below normal operating temperatures. The effect of leakage through the heat exchangers from the primary to the secondary system was analyzed. Calculations show that tritium releases off-site to be below the concentrations allowed by 10CFR-20.(3) Limits on such leakage have been established in Section 3.6 of these specifications. To minimize the amount of any such leakage, the heat exchanger isolation valves must be operable and means for detecting the leakage must be provided. l The limiting value for reactor vessel coolant level is somewhat arbitrary, since 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 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 (i.e., about 7 ft.) above the fuel element. In order to permit periodic surveillance of the effectiveness of the moderator dump, it is necessary to operate the reactor without restrictions on reactor vessel level. This is permissible under conditions when forced reactor cooling is not required such as is permitted in specification 2.1. Because of radiolytic disass,ciation of D20, deuterium gas will collect in the helium cover gas system. If this gas were to reach an explosive concentration (about 7.8% by volume at 25 C in helium),(') , damage to the primary system cot.id occur. To assure a substantial , margin below the lowest potentially explosive value, a 4% limit is imposed.
- - , -
. . ._ . _ -. __ - _ . . _ ._. . . _ - . _ _ ...
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References:
(1) Final Safety Analysis-Report,.NESR 9, Addendum 1, Section 3.3.3.1, November 1980. 4 ( 4 Final Safety. Analysis Report, NBSR 9,. Addendum 1,-Section 3.3.3.2,
'
. November 1980.
(3) Final Safety Analysis Report, NBSR 9, Addendum 1, Section 2.6.4, ., November 1980. i
-
" Flammability of Deuterium in Oxygen-Helium Mixtures,".USAEC Report No. TID-20898,' Explosives Research Center, Bureau of Mines, June 15, 1964..
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._ _ _ _ _ _ _ .
. . l 1 1 1 Section 3.4 Reactor Control and Safety Systems Basis Although the NBSR could operate and could maintain a substantial shutdown margin with less than the four installed shim safety arms, flux and rod worth distortions could occur by operating in this manner.(1) 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 which prevented its insertion. Rod withdrawal accidents for the NBSR were analyzed using a maximum withdrawal rate of 5 x 10- ao/sec.( ) This rate corresponds to the maximum beginning of life rod worths with the rods operating at the design speed of their constant speed mech.'nisms.( ) These analyses showed that for the most severe accident (startup from source level), the resultant energy of 4.8 MW-seconds is significantly below the 34 MW-seconds required to adiabatically heat the core to the point of fuel failure. The parameters listed in Table I are monitored by the reactor safety system. This system automatically initiates action to assure that appropriate safety limits and minimum conditions for operation are
,
not violated. With the channels operable as required by Table I, the safety system meets the reliabilite requirements (including testing and maintenance provisions as suggest.d by the appropriate IEEE standard for these systems).( Where only 1ngle channels are required in Table I, parameters measured by other crinnels combine to provide the necessary redundancy. In the unlikely event that the shim safety 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 calculated to be at least 4% ao. (0} Hence, it is also considered necessary for safe operation. References ( ) Final Safety Analysis Report, Supplement A, NBS. 9A, Response No. 7, page 7-4, October 1, 1966.
. . . . . . .- . . . . .-... - . -
. w a
, i ( Final Safety Analysis' Report, NBSR.9, Addendum 1, Section 3.3.1,
, ~ November 1980.
' (3) Final Safety Analysis Report, Supplement A, NBSR 9A, Response No. 7, page_7-4; October 1,~1966. ( ) Standards for Nuclear Power Plant Protection Systems, IEE2/NSG/ Reactor
,
7 strumentation and Control, 8th Rev., September 13, 1966. J Final Safety' Analysis Report on the NBSR, NBSR 9, Section 4.6.9, ' page 4-19, April 1966. (6) Final Safety Analysis. Report on the NBSR, Supplement A, NBSR 9A, ' Response No. 7,.page 7-7, October 1, 1966. 1 1 j- + h k 9 i 1
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.o. s Section 3.5 Reactor Emergency Cooling System
- Basis
References:
I ' (1) Final Safety Analysis Report on the NBSR, Supplement B, NBSR 9B, Response No. 14, page 14-1, December 16, 1966, i (2) Final Safety Analysis Report, NBSR 9, Addendum 1, Section 3.3.4, October.1980. (3) Final Safety Analysis Report on the NBSR, NBSR 9, Section 7.1.1,
+
page 7-1, April 1966.
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l 2 Section 3.6 Secondary Cooling System I 4 Basis At the end of the term of the NBSR license (2001), the tritium , 4 concentration in the primary coolant is calculated to be 5 mci /mL. 4 Using.this value, the above criteria assure that tritium concentrations ' in effluents will be as low as practicable and below concentrations allowed by 10 CFR 2.0303 for liquid effluents and 10 CFR 20.106 for. gaseous. effluents.(l} The specified daily and week?.y leakage rates ' represent the lowest limits of positive detection of D 02 losses under both reactor operating ! and shutdown conditions. The specified yearly leak rate represents i an estimat* of the smallest size leak enat can be positively located i and repaired.
References:
( ) Final Safety Analysis Report, N3SR 9, Addendum 1, Section 2.6.4, November 1980. i 0 , k c 3 4 ,
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~ 3.7 Fuel Handling and Storage.
Applicability: ' Applies-to the handling and storage of fuel elements 4^ or fuel experiments outside of the reactor vessel. Objective: .To' prevent fuel element overheating or inadvertent criticality outside of the reactor vessel. 4 ~ Speci~ications: ,
- a. All fuel elements or fueled experiments shall be stored and handled in :a geometry- such that the calculated k,gf. is less
,
i than 0.9_-under optimum-conditions of water moderation and reflection.
'
- b. A fuel-element shall not be.placed in the fuel transfer chute ,
' or be otherwise removed from the reactor vessel unless the reactor has been shut down for a period equal to or greater l , than'one hour for each megawatt of operating power level
- tasis In order to assure that no inadvertent criticality of stored i or handled fuel' elements occurs, they shall be maintained in a j.
geometry which assures an adequate margin below criticality. This margin is established as a k,ff of 0.9 for normal storage facilities or for handling outside of normal storage facilities. To assure that a fuel element which may become stuck in the i fuel transfer chute does not melt and release radioactive material, j a time limit is specified before a fuel element may be removed t.sa the vessel following reactor shutdown. Measurements carried out during reactor startup showed that i, for the hottest element placed dry in the transfer chute, 8 hours after shutdown from 10Mw, the maximum temperature is only 550 F without auxiliary cooling. Extrapolation of these measurements shows'that 20 hours after shutdown from 20Mw the maximum temperature for the. hottest element would be less than 800 F without aux 111ery r cooling. For al1~other power levels below 20Mw, the specified i a, waiting time would result in even lower temperatures. This provides
- a'significant margin from the melting temperature of 1200 7. These values are confirmed by fuel temperature tests carried out at the Oak Ridge Research Reactor. Therefore the waiting cimes specified
. will preclude'any fuel elementadamage or fission product release. E 4 s -- , v . ,, , , . . --v e r,,--,. - - . - - - -
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- . . .. . . . . . -. - . . . . . .-
-
, - p .. .: ? 4 - I Section 3.8 Fuel Handling-Within Reactor Vessel Basis 2 Each NBSR fuel element employs a latching bar which must be
- rotated-to lock the fuel element in the upper grid. Following
, fuel handling, it is necessary to insure that this bar is properly positioned so that each element which has been moved cannot'" wash 1
out" when flow is initiated.. Either of two. inspection methods may be employed. A periscope can be used for visual inspection, or ,
'
a pick-up tool can be positioned approximately 12 inches above each ' t element, one at a time. When a primary pump is started and flow established, fuel elements will rise approximately 30 inches if not
. latched-in place and will strike the pick-up tool.
-
Alternately, flow is established with s primary pump, then the pick-up tool is lowered
~
to near the top er the element. If no contact is made, then the element-is p.operly latched.
!
Reference:
, (1) Final Safety Analysis 'Aeport, NBSR 9,-Section 7.2.1.2, page 7-3, April 1966. 4 l d I <
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. . Section 3.9 Normal and Post-Incident Exbaust Systems Basis The potential radiation exposure to persons at *ha 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 emergency exhaust system to maintain the confinement building at a negative pressure and to direct all etiluents through filters and up the reactor building stack. The emergency exhaust system has been made redundant to assure its operation. Because of its Laportance, this redundancy should be available at all times so that any single failure would not preclude system operat;on when required.( The emergency cleanup system, which internally circulates confinement building air thcuugh riiters, can reduce short-term (two-hour) activity release from the building by nearly 50% and long-term (30-day) releases by a factor of five. Even though calculations of off-site releases, for the Design Basia Accident, assumes that the emergency cleanup system is incherable, its use will minimize the releases and is therefore, required to be operable. The normal reactor building exhaust is designed to pass reactor building effluents through high efficiency particulate filters at least capable of removing particles of 0.3 microns or greater with an efficiency of at least 99% and discharge them above the reactor building roof level. This system assures filtering and dilution of gaseous effluents before these effluents reach personnel cither on-site or off-site.( } It can properly perform this. function using various conbinations of its installed fans and building stack. Gaseous effluent monitors are required by Section 3.4 of these specifications. References ! ( Final Safety Analysis Report, NBSR 9, Addendum 1, Section 3.4, November 1980. (2) Final Safety Analysis Report on the NBSR, Supplement B, NBSR 9B, Response No. 5, page 5-1, December io, 1966. (3) Final Safety Analysis Report on the NBSR, NBSR 9, Section 3.6.4, page 3-15, April 1966.
-
_ . _
- __ -. __ . . . __ _ _ . __.
w " . , 'Section 3.11 Miscellaneous' Systems q Basis
.
A' fission products monitor located in the helium sweep gas or the i primary coolant will give indication of a " pin-hole" breach in the clad so that early~ preventive measures can be taken. Since this monitor is not redundant, periodic sampling and analysis of the helium sweep gas must.be en'stituted;for b periods when it is undergoing ~ maintenance.(1) , The frequency chosen (daily)-is adequate to assure early detection of , any.small failures before they would be expected to grow significantly. Larger failures would occur only after an accidental teactor transient which would be followed Fy a reactor shutdown. Parc of the post-incident evaluation would include a helium sweep gas sample,.so the existence of aa actual failure would be detected prior to continuation of operation. The concentration limits specified assure that 10 CFR 20 limits 4 (MPC) are not exceeded at the site boundary. An allowance for dilution from the reactor building stack to the nearest site boundary of 1000 (as , justified in the FSAR, page 2-7) is given.( } This value of 1000 from a } - diffusion view point is the minimum expected at the nearest site boundary
- under the least favorable meteorologica1' conditions. This number could i
j be increased by one or ctwo orders of magnitude if normal variations in . wind speed and direction were considered. Since these variations are not considered, a one or two order of magnitude margin is inherent in , this limit. The instantaneous release limit assures that the average release is not obtained by a small number of very large releases with the attendent possibility of high local concentrations of released effluents. This < specification, although more restrictive than 10.CFR 20, provides
- additional assurance that, releases to off-site personnel are minimized.
i-In specifying the limits on particulates and long half-lived !
- -(greater than 8 days) halogens, consideration was given to the oc , bility
.
of biological reconcentration in food crops or dairy products. Us1ng available information,( ) a conservative reconcentration factor of 700 't i is applied. Thus, the limit for .those isotopes is the Maximum Permissible Concentration (as specified in Appendix B, Table II of 10 CFR 20) times the 1000 dilution factor divided by the 700. reconcentration factor l -(i.e., 1.4 MPC), i 1
- . . ._. _- , . _ ,- , _ . , . _ , _ . . _ _ ~ . . _ , . - _ _ , _ . , . . _ _ _ ,
-. .
References:
Final Safety Analysis Report on the NBSR, NBSR 9, Section 9.7.3.2, page 9-53, April 1966. Final Safety Analysis Report on the NBSR, NBSR 9, Section 2.3.7, page 2-7, April 1966. ( J. D. Soldat, Health Physics 9, 1170, 1963.
. . - . .- .. .
. .
;
Section 4.0 Experiments Basis
- The individual experiment reactivity limit is chosen such that the failure of an experimental installation or component will not cause a reactivity increase greater than can be controlled by the regulating rod. (l) Since the failure of individual experiments cannot be discounted
-
during the operating life of the NBSR, failure should be within the reactor's control capability. This limit does not include such semi-permanent structural materials as brackets, supports, and tubes which 7 are occasionally removed or modified, but which are positively attached . to core structures. Whet: these components are installed, they are considered structural members rather_than part of an experiment. 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% ao is a 0.2% ao allowance for the pneumatic irradiation system, 1.3% 40 for experiments which 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 all of the 1.3% ao for removable experiments moved in 0.5 Sec, analysis has shown that this ramp insertion into the NBSR operating at 20MW would not result in any core damage.( The 0.2% ao for the combined pneumatic irradiation systems is well below this reference accident as well as
!
being within the 0.5% ap capability of the regulating rod. In addition to all reactor experiments being designed not to fail ' from internal overheating or gas buildup, they must also be designed to be compatible with their environment in the reactor.( } Specifically, ' their failures must not lead to failures of the core structure or fuel, or to the-failure of other experiments. Also, they must be able to withstand without failure the same transients which the reactor itself , can withstand without failure (i.e. , loss of reactor cooling flows, startup accident and others where the reactor's safety system provides the ultimate protection.)(5) i The detonation.of explosive or metastable materials within the reactor is not an intended part of the experimental procedure for the
- _ -- _ _ _ -
. _ _ _ . . . ..
Y ' i l NBSR; however, the possibility of a-rapid energy release must be considered when these materials are present.( Since the analytical methods used for designing containers for very rapid energy releases are not well developed, full prototype testing of the containment design is specified. The requirement for testing twice the amount of material to actually be irradiated provides a safety margin of at least a factor of two to allow for possible experimental uncertainties. Experiments containing material.s corrosive to reactor components or highly reactive with reactor or experimental coolants, although limited by item C of this specification, provide the potential for reducing the integrity of the fuel elements. For this reason, an added margin of safety is required 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 material to be irradiated. Since the failure of experiments must be considered, the radiological consequences cf these failures must also be considered. Consistent with the Commission's regulations, these failures which are possible must not result in on-site personnel exposures or off-site concentrations in excess of those permitted by 10 CF2 20. The analysis of the cryogenics facility at temperatures below the freezing point of D 0(39 2 F) has not been completed. Accordingly, its use is not permitted.(7)
References:
(1) Final Safety Analysis Report, Supplement A, NBSR 9A, Response No. 22, page 22-2, October 1, 1966. ( ! Final Safety Analysis Report, NBSR 9, Section 4.6.5, page 4-17, April 1966. ( } Final Safety Analysis Report, NBSR 9, Addendum 1, Section 3.3.2, November 1980. (4) Final Safety Analysis Report, NBSR 9, Section 8.3.1, page 8-5, t April 1966. (5) Final Safety Analysis Report, NBSR 9, Section 8.3.2, page 8-6, i April 1966.
- .. ._ - _ _ _ . - - . _
,
- . (6) Final -afety Analysis Report, Supplement A, NBSR 9A, Response No. 22, page 22-2, ~etober 1, 1966. (7) Final Safety Analysis Report, Supplement A, NBSR 9A, Response No. 22, page 22-12, October 1, 1966.
o . Section 5.1 Confinement System Basis The confinement closure system is _.iitiated either by a signal from the confinement building radiation detectors or manually by the Major Scram button.( } In order to assure complete surveillance, the system is tested by using these same devices to initiate the test. In addition, checks of both the trip features and the ability of the radiation detectors to respond to ionizing radiation are made. A preoperational test program was conducted to measure the representative leakage characteristics at values of +7.5" H,0 and
~
-2.5" H20.(2) The specified test pressures and vacuums are acceptable
.
since past tests have shown leakage rates to be linear with applied pressures and vacuums. Test frequencies have been specified to be consistent with current practices and proposed guides.( Changes in the building or its penetrations =ust be verified to be at least equal to the original building design; therefore, tests must be performed before the building can be considered to be operable.
References:
(1) Final Safety Analysis Report, NBSR 9, Section 9.5.4.4, page 9-49, April 1966. ( Final Safety Analysis Report, NBSR 9, Section 3.7.2, page 3-16, April 1966.
" Proposed Technical Safety Guide III, Reactor Containment Leakage Testing and Surveillance Requirements," USAEC, December 15, 1966.
- . . . - . . . - .. .._ .
l . , .; '1 i l J Section'5.4 Reactor Control and Safety System
^ Basis Measurements of reactivity worths of the shim arms have shown over many years of operation to vary slowly due to absorber burnup and only
- slightly with respect to operational core loading and experimental changes. An annual check will ensure adequate reactivity cargins.
The shim arm-drives are constant speed mechanical devices.(1 Scram is aided by a spring which opposes drive motion during arm withdrawal. Withdrawal'and insertion speeds'or scram time should not vary except due to mechanical wear. The surveillance frequency is chosen to provide a significant margin over the expected failure or wear rates of these
'
devices. Because of its importance, the scram times are tested twice as often as the driven withdrawal and insertion times. The shim arms are considered operable for scram if they drop 5 degrees within 220 msec. This value is consistent with the amount and rate of reactivity insertion assumed in analyzing the accident requiring the most rapid scram.( } Since redundancy of all important safety channels is provided, random failures should not jeopardize the ability of these systems to perform their required functions.( ' ) However, in order to assure that failures do not go undetected, frequent surveillance is required and specified. Because various experiments require precise operating conditions, the NBSR has been designed to assure that accurate recalibration of 1 power level' channels can be easily and frequently achieved. The cali-bration is performed by comparison of nuclear channels with the thermal
, power measurement channel (flow-AT product). Because of the small AT in the NBSR (15 at 20 Mw) these calibrations will not be performed below 5 Mw for 10 Mw operation or below 10 Mw for 20 Bhi operation; however to 2
assure that no gross discrepancies between nuclear instruments and flow-AT indicators occur, comparisons (but not necessarily calibrations) are
=ade above 1 Mw.
Annual.recalibration of power range detectors by a separate independent technique, one which utilizes a standard source and geometry, is adequate
- -to show detector variations due- to boron burnup in the chambers or other
~
long-range change of char'cteristics. a . 4
, ~ _
. . - , - . . - . . . , . . . . _ ~ , .- -
,
o
References:
(1) Final Safety Analysis Report, NBSR 9, Section 4.6.6.1, page 4-18, April 1966. ( Final Safety Analysis Report, NBSR 9, Section 13.2, page 13-2, April 1966. ( 4 Final Safety Analysis Report, NBSR 9, Addendum 1, Section 3.3.2, November 1980. (4)NBS Information Letter, Item 3, June 1967. ,
- - .-
- ,- .-
l Section 5.7 Post-Incident and Gaseous Waste Systems Basis The post-incident gaseous waste system depends on the proper
- operation of the emergency exhaust system fans, valves, and filters, -
which are not routinely in service. Since they are not continuously used, their failure rate due to wear or loading should be low. On the other hand, since they are not being used, their condition in standby must be checked sufficiently often to assure that they will function properly when needed. The functional checking of the fans and valves on a monthly schedule (Section 5.1.a of these specifications) is consistent with the test frequency of other important confinement building equipment in the NBSR. An operability test of the active components of the emergency exhaust system is performed quarterly to assure that each component will be operable if an emergency condition required use of the system. The quarterly frequency is considered adequate since this system receives very little wear and since the automatic controls are
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backed up by manual control provisions. The test frequency of the absolute filters has been established as at least annually.( ) This is the same frequency as used at the Savannah River Laboratory for filters subject to continuous air flow. Since the NBS absolute filters in the emergency exhaust system and the internal recirculation system will be idle except during testing, deterioration - should be much less critical than for filters subjected to continuous air flow where dust overloading and air breakthrough are possible after long pe.riods of use. Therefore, an annual testing frequency should be adequate in detecting filter deterioration.
- The test requirement for the charcoal filters in the emergency enhaust and internal recirculation systems is basically a physical integrity test. It is prudent to verify that the NESR filters are not installed or operated in such a way as to be damaged or bypassed.(
Therefore, a Freon gas in-place leakage test is required annually to detect leakage paths resulting from charcoal settling and deterioration of the filter seals. Experience t the Savannah River plant has demon-
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strated Freon gas to be an acceptable means for determining the leakage characteristics of charcoal filter installations. The 17. acceptability l
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- limit is specified to give assurance that a high overall iodine filter efficiency significantly above the 95% used in the DBA will Ina maintained.(5)
References:
( ' Final Safety Analysis Report, Supplement A, NBSR 9A, Response No. 12,. pages 12-10 through 12-17, October 1, 1966. (2) Final Safety Analysis Report, Supplement B, NBSR 9B, Response No. 5, pages 5-1 and 5-2, December 16, 1966. ( J Final Safety Analysis Report, Supplement B, NBSR 93, Response No. 1, page 1-1, December 16, 1966. (4) Final Safety Analysis Report, Supplement B, NBSR 9B, Response No.1, page 1-1, December 16, 1966. l (5) Final Safety Analysis Report, NBSR 9, Addendum 1, Section 3.4.2, l- November 1980. 1
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