ML20042F465
| ML20042F465 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 05/04/1990 |
| From: | SYSTEM ENERGY RESOURCES, INC. |
| To: | |
| Shared Package | |
| ML20042F464 | List: |
| References | |
| NUDOCS 9005080354 | |
| Download: ML20042F465 (40) | |
Text
{{#Wiki_filter:' [ Attachment 21to-
'AECM-90/0003 ,! 'A. SUBJECT
- 1. NL 89/01 Load Handling Controls'- f
- 2. Revised Technical Specifications: ;
- a. Isolstion Actuation Instrumentation, Table 3.3.2 Page 3/4 3-14.
- b. Isolation Actuation Instrumentation Surveillance Requirements,
. Table 4.3.2.1 Page 3/4 3-26.=
- c. Radiation Monitoring Instrumentation, Table 3.3.7.1 Page 3/4'3-60.
- d. Radiation Monitoring Instrumentation Surveillance Requirements' .,-
Table.4.3.7.1 Page 3/4 3-62.
- e. Secondary Containment Integrity, 3.6.6.1 - Page 3/4 6-48.
. . . I
- f. Secondary Containment Automatic Isolation Dampers / Valves, 3.6.6.2 - Page 3/4 6-49.
- g. Standby Gas Treatment System, 3.6.6.3.- Page'3/4 6-55. . ;
- h. Standby Service Water System, 3.7 1.1 - Page-3/4 7-1.
- i. Ultimate Heat Sink, 3.7.1.3 - Page 3/4 7-4.
- j. Control Room Emergency Filtration System, 3.7.'2 - Page 3/4 7-5.
- k. A. C. Sources - Shutdown, 3.8.1.2 - Page 3/4 8-9,
- 1. D. C. Sources - Shutdown, 3.8.2.2 - Page' 3/4 8-14. -
- m. Distribution - Shutdown, 3.8.3.2 - Pages 3/4 8-17 and 3/4' 8-18.
- n. Secondary Containment Bases,-3/4.6.6 - Page B 3/4 6-8. ;
l- B. DISCUSSION
- 1. In July, 1988 System Energy Resources, Inc. (SERI) identified an apparent discrepancy between the Grand Gulf Nuclear Station (GGNS)
Technical Specifications (TS) and the UFSAR accident analysis. Analyses discussed in UFSAR sections 15.7.4, 15.7.6 and 15A.6.5.3 assume secondary containment isolation capability and standby gas treatment system (SGTS) operability during a postulated fuel. . handling accident (FHA). The TS require secondary containment integrity and SGTS operability when handling irradiated fuel in the e primary or secondary containment but do not apply such requirements-when handling loads that if postulated to drop on irradiated fuel assemblies could result in excessive offsite releases. A9001041/SNLICFLR - E 900508G354 900504 PDR ADOCK 05000416 P PDC
~
a Attachment 2 to AECM-90/0003-
~
- 2. -Further evaluation of this discrepancy led to the concern being.
identified that certain loads could be handled over' irradiated fuel assemblies.in the' primary and secondary containment during~. Cold Shutdown or Refueling when' secondary containment ' integrity and SGTS ' operability may not be required. Thus, there may be conditions
/ under which the drop of such loads could result in offsite doses in- J excess of~25% of 10CFR100 dose. limits. !
- 3. As reported in Licensee Event Report:88-016-00 (LER) (AECM-88/0204, - l dated October 21, 1988)=and supplemented in LER 88-016-01' .
(AECM-89/0025, dated February 1, 1989) pursuant to 10CFR50 73(a)(2)(ii)(A), SERI determined'on September 23, 1988 that there existed situations during Cold Shutdown and Refueling where-certain loads could have been handled over irradiated-fuel. ' assemblies without secondary containment--integrity which were not - bounded by events currently analyzed. This is because TS and; . administrative controls did not control such: load movements during conditions when secondary containment was notLoperable, j
- 4. After review,.SERI considers the'following to be contributing factors to the above condition:
- a. The GGNS TS were developed using previously licensed BWR TS m
' which did not address handling loads over irradiated fuel without. secondary containment integrity. ;
- b. The GGNS administrative controls were developed to implement !
the TS-and did not contain the necessary controls'to handle. loads without secondary containment integrity. [ ,
- 5. In order to corract the discrepancy between the TS and the UFSAR and '
r'esolve the concern described above, the following TS changes are. proposed: ] '
- a. Revise ACTION 21 and NOTE
- of Table 3.3.2-1 to require the .
operability of the isolation actuation instrumentation related to secondary containment integrity when " suspended loads"'are-being handled over irradiated fuel.
- b. Revise the
- footnote of' Table 4.3.2.1-1 to require the surveillance of the isolation actuation instrumentation related to secondary containment integrity.when " suspended loads" are {
being handled over irradiated fuel.
- c. Revise the ** footnote of Table 3.3.7.1-1 to require the operability of the radiation monitoring instrumentation related ;
to secondary containment integrity when " suspended loads" are a being handled over irradiated fuel. ' L
- d. Revise the ** footnote of Table 4.3.7.1-1 to. require the l
, surveillance of the radiation monitoring instrumentation related to secondary containment integrity when " suspended-loads" are being handled over irradiated fuel. l l -A9001041/SNLICFLR - 6 : q
Attachment 2 to, AECM-90/0003
- e. Revise ACTION b and the
- footnote of TS 3.6.6.1 to require secondary containment integrity be maintained or suspend handling " suspended loads" over irradiated fuel.
- f. Revise the ACTION and the
- footnote of TS 3.6.6.2 to require the operability of the secondary containment automatic isolation dupers / valves or suspend handling " suspended loads"'
over irradiated fuel.
- g. Revise ACTIONS a.2 and b and the
- footnote of TS 3.6.6.3 to require the $GTS be operable or suspend handling " suspended loads" over irradiated fuel.
- h. Revise the
- footnote'of TS 3.7.1.1 to require the Standby Service Water (SSW) subsystems associated with secondary containment integrity be operab'e when " suspended loads" are being handled over irradiated fuel,
- i. Revise the
- footnote of TS 3.7.1.3 to require the operability of the ultimate heat sink basin associated with the SSW subsystems required operable to maintain secondary containment integrity when " suspended loads" are being handled over irradiated fuel.
- j. Revise ACTION b.2 and the
- footnote of TS 3.7.2 to require the control room emergency filtration system be operable or suspend handling " suspended loads" over irradiated fuel.
- k. Revise ACTION a and the
- footnote of TS 3.8.1.2 to require the operability of the A.C. electrical power sources or suspend handling " suspended loads" over irradiated fuel.
- 1. 'levise ACTION a and the
- footnote of TS 3.8.2.2 to require the {
operability of the Division 1 and 2 D.C. electrical power j i sources or suspend handling " suspended loads" over irradiated fuel. I' m. Revise ACTIONS a.1, b.1 and the
- footnote of TS 3.8.3.2 to L require the operability of the A.C. and D.C. power distribution systems during shutdown or suspend handling " suspended loads" ;
over irradiated fuel. '
- n. Revise Bases 3/4.6.6 to include a discussion of why secondary containment integrity is required when handling " suspended loads" over irradiated fuel.
- 6. Attached to this proposed TS change are the revised marked up TS pages, s i
A9001041/SNLICFLR - 7 l 1
Attachment 2 to AECM-90/0003 l C. JUSTIFICATION
- 1. Currently, the TS do not require secondary containment integrity when handling " suspended loads" directly over irradiated fuel assemblies othee than when handling irradiated fuel assemblies in the primary or secondary containment. This is not a radiological problem, unless a " suspended load" is being handled which, if dropped onto irradiated fuel assemblies, produces enough damage to the fuel cladding to cause a release of radioactive material which could result in offsite doses in excess of 25% of 10CFR100 dose limits.
- 2. For " suspended loads" which could drop, TS should require secondary containment integrity be maintained whenever those " suspended loads" are to be handled directly over a source term large enough to produce offsite doses in excess of 25% of the 10CFR100 dose limits.
Evaluation has shown that " suspended loads" may be handled directly over irradiated fuel assemblies af ter 20 days of decay time in either the primary or secondary containment without secondary containment integrity. 'The decay time is assumed to commence at the time the fuel assemblies are placed in a subcritical condition (e.g., reactor shutdown). Secondary containment integrity is not-required after 20 days because if a " suspended load" were to drop onto irradiated fuel assemblies following 20 days of decay, the resultant unfiltered release of radioactive material would not be sufficient to result in offsite doses in excess of 25% of 10CFR100 dose limits. " Suspended loads" which must be postulated to drop are inads which are moved directly over irradiated fuel assemblies by a lifting device (e.g., polar crane, jib crane) and are not handled in accordance with commitments implementing the NUREG-0612, Section 5.1.1 guidelines.
- 3. Thus, in terms of the proposed TS amendment, a " suspended load" is defined as a load:
- a. suspended by a lifting device (e.g., polar crane, jib crane) directly over irradiated fuel assemblies, and
- b. postulated to drop onta irradiated fuel assemblies, and
- c. If dropped onto irradiated fuel assemblies could cause offsite releases in excess of 25% of 10CFR100 dose limits.
- 4. Using the dose consequence of not exceeding 25% of the 10CFR100 dose limits, as the bounding value for the amount of fuel damage, is considered acceptable for two reasons. First, Standard Review Plan i 15.7.4, " Radiological Consequences of Fuel Handling Accidents", has !
been utilized by the NRC as a basis for determining the acceptability of a licensee's capability te successfully mitigate a fuel handling accident. An acceptance criterion for the , radiological release is to have the offsite doses be no greater than ; 25% of the 10CFR100 dose limits. Second, the NRC in NUREG-0612, "The Control of Heavy Loads at Nuclear _ Power Plants", uses 25% of the 10CFR100 dose limits as a criterion to determine whether the consequences of a potential load drop are acceptable or if further , preventative load handling measures are necessary, f A9001041/SNLICFLR - 8
Attachment 2 to AECM-90/0003
- 5. To implement the proposed TS changes, a GGNS plant procedure incorporates administrative' controls governing handling of
" suspended loads" directly over irradiated fuel. The procedure specifies the conditions for which secondary containment integrity i is required in terms of " suspended load" height over irradiated !
fuel, load weight, and irradiated fuel decay time. The procedure ; conditions triggering establishment of secondary containment are subject to change. For example, if a different type fuel assembly
- were to be used at GGNS the fuel asserbly parameters assumed in. the calculation (i.e., clad strength, weight, number of pins) upon which: ;
the administrative controls are based may change which may require procedure revisions. However, the use of.25% of the 10CFR100 dose ' limits as the bounding value for the amount of fuel damage allowed will not change. The administrative controls are for. conditions when secondary containment integrity is not maintained. The procedure is used to determine the maximum height a particular weight load could be handled such that-if dropped onto irradiated fuel of a specified decay time.the offsite dose consequences would . not exceed 25% of the 100FR100 dose limits. The administrative controls were developed using methodology and assumptions similar to that described in the GGNS VFSAR Section 15.7.4 and 15.7.6 for the design basis fuel handling accident. D. NO SIGNIFICANT HAZARDS CONSIDERATIONS The proposed amendment would revise the Technical Specifications (TS) by adding a requirement to establish secondary containment integrity when - certain " suspended loads" are being handled nyer irradiated fuel. In-addition, action statements which require suspension of handling irradiated fuel assemblies would be revised to require suspension of
- handling " suspended loads" over irradiated fuel. The proposed changes would ensure that secondary containment integrity is maintained when certain loads are handled over irradiated fuel assemblies. Should the load be dropped, the offsite radiation dose consequences would be within the limits of 10CFR100.
The following paragraphs are an evaluation of the proposed amendment against the standards of 10CFR50.92.
- 1. No significant increase in the probability or consequences of an accident previously evaluated results from this change,
- a. The relevant accident to consider for this change is damage to irradiated fuel rods and release of fuel rod gap activity as a result of a dropped load. The previously analyzed accident was a dropped fuel assembly accident, assuming filtration of gaseous activity released from damaged fuel rods through the SGTS. The planned number and type of load movements has not been altered by the proposed TS change. Therefore, the probability of a load' drop has not been increased.
A9001041/SNLICFLR - 9
L i Attachment 2 to AECM-90/0003
- b. Although the consequences of an accident tould be greater than the consequences of a previously evaluated accident-(i.e., l damaged fuel rods), they are not significantly greater because the potential offsite doses would remain within the applicable .
regulatory acceptance criteria.
- c. Therefore, there is no significant increase in the probability or consequences of previously analyzed accidents due to the proposed change, j
- 2. This change would not create the possibility of a new or different kind of accident from any previously evaluated,
- a. The hopping of loads _ onto fuel assemblies has been previously evalw ted as documented in.the GGNS VFSAR Sections 15.7.4 and 15.7.6.
- b. Therefore, the possibility of a new or different kind of accident from any previously evaluated is not created.
- 3. This change would not involve a significant reduction in the margin of safety,
- a. The TS do not currently address the. issue of handling non-irradiated fuel loads of less than 1140 pounds over spent fuel or loads over irradiated fuel in reactor vessel. The proposed change will add a requirement that secondary
- ontainment integrity be maintained when handling " suspended loads" over irradiated fuel which, if postulated to drop, could result in offsite dose consequences e weeding 25%.of 10CFR100 dose limits. The addition of such 0 r.,quirement will ensure that should an accident occur, accident doses'will be within regulatory acceptance criteria.
- b. Therefore, this proposed cha % e will not involve a significant reduction in the margin of safety.
Therefore, based upon the above evaluation, operation in accordance with the proposed amendment involves no significant hazards consideration. : i A9001041/SNLICFLR - 10 -
NL M/C4 I
,1NSTRUMENTATION TABLE 3.3.2-1 (Continued) '
ISOLATION ACTUATION INSTRUMENTATION ACTION ACTION 20 Be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN ' within the next 24 hours. ~ ACTION 21 - Close the affected system isolation valve (s) within one hour or: ;
- a. In OPERATIONAL CONDITION 1, 2, or 3, be in at least HOT .
SHUTDOWN within the next 12 hours and in COLD SHUTDOWit within the following 24 hou n (Maut $
- b. In OPERATIONAL CONDITION *, suspend CORE ALTERATIONS, handling of irradiated fuel n in the primary containment and l operations with a potential for draining the reactor vessel. ;
ACTION 22 - Restore the manual initiation function to OPERABLE status within i 48 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. ACTION 23 - Be in at least STARTUP with the associated isolation valves closed within 6 hours or be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours. !' ACTION 24 - Be in at least STARTUP within 6 hours. ACTION 25 - Establish SECONDARY CONTAINMENT. INTEGRITY with the standby gas treatment system operating within one hour.
- ACTION 26 -
Restore the manual initiation function to OPERABLE status within 8 hours.or close the affected system isolation valves within the next hour and declare the affected system inoperable. ACTION 27 - Close the affected system isolation valves within one hour and declare the affected system inoperable.- r ACTION 28 - Within one hour lock the affected system isolation volves closed.
- or verify, by remote indication, that the valve is closed and electrically disarmed, or isolate the penetration (s) and declare the affected system inoperable.
ACTION 29 - Close the affected system isolation valves within one hour and declare the affected system or component inoperable or:
, a. In OPERATIONAL CONDITION 1, 2 or 3 be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the fcilowing 24 hnurs.
- b. In OPERATIONAL CONDITION # suspend CORE ALTERATIONS and opera-tions with a potential for draining the reactor vessel.
ACTION 30 - Dec_lare the affected SLCS pump inoperable. [7nsectk NOTES When handling irradiated fuel in the primary or secondary containment and g during CORE ALTERATIONS and operations with a potential for draining the l reactor vessel. The lo condenser vacuum MSIV closure may be manually bypassed during reactor SHUTDOWN or for reactor STARTUP when condenser vacuum is below the trip set-point to allow opening of the MSIVs. The manual bypass shall be removed when condenser vacuum exceeds the trip setpoint. During CORE ALTERATIONS and operations with a potential for draining the reactor vessel. ,
## With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
(a) See Specification 3.6.4, Table 3.6.4-1 for valves in each valve group. t GRAND GULF-UNIT 1 3/4 3-14 Qmendmed No _
f NL 89/01 INSERT A , l 1 or suspended loads over irradiated fuel i t P
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O g IABLE 4.3.2.1-1 (Contiaued) 5 t o ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIRLMENTS I r"- 00
& CHANNEL 3 OPERATIONAL N CHANNEL FUNCTIONAL CHANNEL COM0lil0NS IN nellCH - - - -
TRIP EUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED
- 6. RHR SYSTEM ISOLATION (Continued)
- e. Drywell Pressure - High 5 M R ICI 1.2.3
- f. Manual Initiation MA M(a) NA 1. 2, 3
*When handling irradiated fuels,in the primary or secondary containment and during CORE ALTERATIONS and T
operations with a potential for draining the reactor vessel. l w
**1he low condenser vacuum M51V closure may be manually bypassed during reactor SHUIDOWN or for reactor STARIUP when condenser vacuum is below the trip setpoint to allow opening of the M51Vs. The manual E
- bypass shall be removed when condenser vacuum exceeds the trip setpoint.
#During CORE ALTERATION and operations with a potential for draining the reactor vessel.
##With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
(a) Manual initiation switches shall be tested at least once per 18 months during shutdown. All other circuitry associated with manual initiation shall receive a CHANNEL FUNCTIONAL TEST at least once per 31 days as part of circuitry required to be tested for automatic system isolation. (b) Each train or logic channel shall be tested at least every other 31 days. 3 (c) Calibrate trip unit at least once per 31 days. 2 ~ SL 3 o 3
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NL 89/01 ; INSERT A : r or suspended loads over irradiated fuel . f L L
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4 I L 4 t A9001041/SNLICFLR - 11
_ _ . _ . _. _ = _ _ _._ . _ _ _ _ _ _ _ ._. E TABLE 3.3.7.1-1 (Continued) RA01ATION MONITORING INSTRUMENTATION f MINIMUM CHANNELS APPLICABLE ALARM / TRIP
] INSTRUMENTATION OPERABLE MEASUREMENT 00 CONDITIONS SETPOINT RANGE ACTION E 10. Area Monitors
% a. Fuel Handling Area
~ Monitors
- 1) New Fuel 1 (e) 10
-2 3 Storage Vault -<2.5 mR/hr/NA to 10 d/hr 72
- 2) Spent Fuel 1 (f) -2 3 Storage Pool 52.5 mR/hr/NA 10 to 10 mR/hr 72 w 3) Dryer Storage 1 (g) -2
'1 Area 52.5 mR/hr/NA 10 to 103mR/hr 72 e, b. Control Room 1 At all times <0.5 mR/hr/NA -2 3 Radiation Monitor - 10 to 10 mR/hr 72
('Ch5edh With RHR heat exchangers in operation. When irradiated fuelf4e-being handled in the primary or secondary containment.
# Initial setpoint. l Final Setpoint to be determined during startup test program. Any required change to this setpoint shall be submitted to Commission within 90 days after test completion.
With ADHR heat ~exchangers in operation. (a) Trips system with 2 channels upscale-Hi Hi Hi, or one channel upscale Hi Hi Hi and one channel downscale or 2 channels downscale. E (b) Isolates containment /drywell purge penetrations.
@ (c) With irradiated fuel in spent fuel storage pool.
& (d) Also isolates the Auxiliary Building and Fuel Handling Area Ventilation Systems.
(e) With fuel in the new fuel storage vault. (f) With fuel in the spent fuel storage pool.
~
(g) With fuel in the dryer storage' area.
~
j (h) Two upscale Hi Hi, one upscale Hi Hi and one downscale, or two downscale signals from the same trip system actuate the trip system and initiate isolation of_the associated isolation valves.
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i TABLE 4.3.7.1-1 o RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS . m j 5: r 1 o OPERATIONAL se g, INSTRUMENTATION CHANNEL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH SURVEILLANCE y CHECK TEST . CALIBRATION
' REQUIRED
- 1. Component Cool'ing Water Radiation
.E Monitor 5 M O 2. Standby Service Water System '
A At all times
- Radiation Monitor 5 M
- 3. Plant Service Water System
- A 1, 2, 3, and*
Radiation Monitor S M A
- 4. [0ELETED] #
- 5. Carbon Bed Vault Radiation Monitor S M A 1, ' 2
- 6. Control Room Ventilation Radiation Monitor S M II
- 7. A 1, 2, 3, 5 and**
Containment and Drywell Ventilation Exhaust Radiation Monitor S M
- 8. Fuel Handling Area Ventilation A At all times Y Radiation Monitor S M A O 9. Fuel Handling Area Poci Sweep 1, 2, 3, 5 and**
Exhaust Radiation Monitor S M A
- 10. Area Monitors (b)
- a. Fuel Handling Area Monitors
- 1) New Fuel Storage Vault 5 M
- 2) Spent Fuel Storage Pool S R (c)
M R (d)
- 3) Dryer Storage Area S M R
- b. Control Room Radiation Monitor 5 (e)
M R At all times With RHR heat exchanners in operation. I 3 When irradiated fuel ps being handled in the primary or secondary containment. 2 g
& (a) The CHANNEL conditions exist.. FUNCTIONhl TEST shall demonstrate that control room annunciation occurs if any of the following
% 1. Instrument indicates measured levels above the alarm / trip setpoint.
a 2. Circuit failure. z 3. Instrument indicates a downscale failure.
- 4. Instrument controls not in Operate mode.
u, (b) With irradiated fuel in the spent fuel. storage pool.
& (c) With fuel in-the new fuel storage vault. -
) (d) With fuel in.the spent fuel storage pool.
(e) With fuel in the dryer storage area.
# With ADHR heat exchangers in operation. - , _ ~, . , _ , _ , .. . . ._ .. . _. _ _ . .
NL 89/01 INSERT B-i or suspended loads over irradiated fuel are j t i i L A9001041/SNLICFLR - 13
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= 5 .y. # y;;;3." ::~~ " ",c,," ~ ' ~ ~ ~ ~ ~ ' * " "~ ~ ~ ~ ~ ~ - ' ~ ~ ~ ~~~ ~ " ~~ ~~ - . .. .. , .. - . ~ . - - N L 39/01 CONTAINMENT SYSTEML 3/4.6.6 SECONDARY CONTAINMENT SECONDARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.6.1 SECONDARY CONTAINMENT INTEGRITY shall be maintained. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and *. ACTION: Without SECONDARY CONTAINMENT INTEGRITY:
- a. In OPERATIONAL CONDITION 1, 2 or 3, restore SECONDARY CONTAINMENT INTEGRITY within 4 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.4ns.er
- b. In Operational Condition *
, suspend handling of irradiated fuel n in the primary or secondary containment, CORE ALTERATIONS and operations l with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.-
SURVEILLANCE REQUIREMENTS 4.6.6.1 SECONDARY CONTAINME,NT INTEGRITY shall be demonstrated by:
- a. Verifying at least once per 31 days that:
- 1. All Auxiliary Building and Enclosure Building equipment hatches and blowout panels are c1csed and sealed.
- 2. The door in each access to the Auxiliary Building and Enclosure Building is closed, except for routine entry and exit.
- 3. All Auxiliary Building and Enclosure Building penetrations not capable of being closed by OPERABLE secondary containment automatic isolation dampers / valves and required to be closed during accident conditions are closed by valves, blind flanges, rupture discs or deactivated automatic dampers / valves secured in position,
- b. At least once per 18 months:
- 1. Verifying that one standby gas treatment subsystem will draw down tne secondary containment to greater than or equal to 0.25 inches of vacuum water gauge in less than or equal to 120 seconds, and
- 2. Operating one standby gas treatment subsystem for one hour and maintaining greater than or equal to 0.266 inches of vacuum water gauge in the secondary containment at a flow rate not exceeding 4000 CFM.
hseEtB) Mien irradiated fuelais. being handled in the primary or secondary containment l and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel. 1 GRAND GULF-UNIT 1 3/4 6-48 Amendment No. 42 l .
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NL SW ! l CONTAINMENT SYSTEMS l SECONDARY CONTAINMENT AUTOMATIC ISOLATION DAMPERS / VALVES LIMITING CONDITION FOR OPERATION 3.6.6.2 The secondary containment ventilation system automatic isolation dampers / valves shown in Table 3.6.6.2-1 shall be OPERABLE with isolation times less than or equal to the times shown in Table 3.6.6.2-1. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and *. ACTION: With one or more of the secondary containment ventilation system automatic isolation dampers / valves shown in Table 3.6.6.2-1 inoperablo, maintain at i least one isolation damper / valve OPERABLE in each affected penetration that is open, and within 8 hours either: ,
- a. Restore the inoperable damper / valve (s) to OPERABLE status, or
- b. Isolate each affected' penetration by use of at least one deactivated automatic damper / valve secured.in the isolation position (the provi-sions of Specification 3.0.4 are not applicable for entry into condi-tion
- for a maximum of 10 inoperable secondary containment isolation dampers / valves #),or
- c. Isolate each affected penetration by use of at least one closed manual valve or blind flange (the provisions of Specification 3.0.4 are not applicable for entry into condition
- for a maximum of 10 inoperable secondary containment isolation dampers / valves #). .;
Otherwise, in OPERATIONAL CONDITION 1, 2 or 3, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
~7 Otherwise, in Operational Condition
- suspend handling of irradiated
- ~
~'db t*
fuellin the primary or secondary containment, CORE ALTERATIONS and opera-tions with a potential for draining the reactor vessel. The provisions l of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS ' 4.6.6.2 Each ' secondary containment ventilation system automatic isolation damper / valve shown in Table 3.6.6.2-1 shall be demonstrated OPERABLE:
- a. Prior to returning the damper / valve to service after maintenance, repair or l
replacement work is performed on the damper / valve or its associated actuator, [ control or power circuit by cycling the damper / valve through at least one i cQmplete cycle of full travel and verifying the specified isolation time, b. During COLD SHUTDOWN or REFUELING at least once per 18 months by verifying that'on a containment isolation test signal each isolation damper / valve actuates to its isolation position. I
- c. By verifying the isolation time to be within its limit when tested pursuant to Specification 4.0.5. 6n se,.4' g "When irradiated fuel n 4e being handled in the primary or secondary containment l and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.
#This exceptio'n is applicable until startup from the third refueling outage. >
GRAND GULF-UNIT 1 3/4 6-49 Amendment No. 58,
- , . - . - - , , . . , . .._,,...,w- , , , , - . , - - . . , - . . . , ~ . , , ,
t NL 89/01 . INSERT A t or suspended loads over' irradiated fuel e f f I e P
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F l NL 89/01 JRSERT B or suspended loads over irradiated fuel are - r 4 5 t e t t 1 1 1 i A9001041/SNLICFLR - 13 :
NL 89/01 INSERT A or suspended loads over irradiated fuel A9001041/SNLICFLR - 11
i V NL 89/01 INSERT B i or suspended loads over irradiated fuei are i t Y k l
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1 F 1 A9001041/SNLICFLR - 13 k
NL 99/0i CONTAINMENT SYSTEMS STANDBY GAS TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.6.6.3 Two independent standby gas treatment. subsystems shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and *. ACTION:
- a. With one standby gas' treatment subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days, or:
1 ' 1.- In OPERATIONAL CONDITION 1, 2 or 3, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- 2. In Operational Condition *
, suspend handling of irradiated fueln in the primary or secondary containment, CORE ALTERATIONS and operations with a potential for draining the reactor vessel. - -
The provisions of Specification'3.0.3 are not applicable. [Tnserij
- b. With both standby gas treatment subsystems inoperable in Operati.qn.gl Condition *, suspend handling of irradiated fuellin the primary or l secondary containment, CORE ALTERATIONS or operat' ions with.a potential for draining the r'eactor vessel. The provisions of Specification 3.0.3. :
are not applicable. J SURVEILLANCE REQUIREMENTS s 4.6.6.3 Each standby gas treatment' subsystem shall be demonstrated OPERABLE:
- a. At least once per 31 days by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that t i
' the subsystem operates for at least 10 continuous hours with the heaters OPERABLE. ' I hserf "When irradiated fue1A being handled in the primary or secondary containment l and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.
'l GRAND GULF-UNIT 1 3/4 6-55 Amendment No. 42 j
i NL 89/Ol 3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEriS STANDBY SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION I i 3.7.1.1 Each of the following independent standby service water (SSW) system subsystems shall be OPERABLE with each subsystem comprised of: ; I
- a. One OPERABLE SSW pump, and I
- b. An OPERABLE flow path capable of taking suction from the associated I SSW cooling tower basin and transferring the water through the RHR i heat exchangers and to associated plant equipment, as required, shall i be OPERABLE as follows:
- 1. In OPERATIONAL CONDITIONS 1, 2, and 3: two subsystems; and .
- 2. In OPERATIONAL CONDITIONS 4, 5, and *: the subsystems associated with the systems and components required to be OPERABLE by Specifications 3.4.9.2, 3.B.2, 3.8.1.2, 3.9.11.1 or 3.9.11.2. ;
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, 5 and *. ACTION:
- a. In OPERATIONAL CONDITION 1, 2 or 3:
- 1. With one SSW subsystem inoperable, restore the inoperable sub- ,
system to OPERABLE status within 72 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
- 2. With both SSW subsystems inoperable, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN ** within the following 24 hours,
- b. In OPERATIONAL CONDITION 3 or 4 with the SSW subsystem, which is asso-ciated with an RHR loop required OPERABLE by Specification 3.4.9.1 or 3.4.9.2, inoperable, declare the associated RHR loop inoperable and take the ACTION required by Specification 3.4.9.1 or 3.4.9.2, as appli-cable. The provisions of Specification 3.
entryintoOPERATIONALCONDITION4from5.g.4'arenotapplicablefor -
- c. In OPERATIONAL CONDITION 4 or 5 with the SSW subsystem, which is associated with an ECCS pump required OPERABLE by Specification 3.5.2, inoperable, declare the associated ECCS pump inoperable and i
take tha ACTION required by Specification 3.5.2. The provisions of Specificatio water level.p 3.0.4 are not applicable for lowering reactor cavity nSehh When handling irradiated fuel in the primary or secondary containment. h l Whenever both SSW subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods. This exception is applicable until startup from the third refueling outage. GRAND GULF-UNIT 1 3/4 7-1 Amendment No. 58 3 (
i NL 89/01 l INSERT A 1
)
or suspended loads over irradiated fuel 1 e i 5 r L 9 A9001041/SNLICFLR - 11
NL 24/t>l PLANT SYSTEMS ULTIMATE HEAT SINK ] LIMITING CONDITION FOR OFCRATION 'I 3.7.1.3 At least the following independent SSW cooling tower basins, each with: a.
~ A minimum basin water level at or above elevation 130'3" Mean Sea Level, USGS datum, equivalent to an indicated level of >87".
- b. Two OPERABLE cooling tower fans,#
shall be OPERABLE:
- a. In OPERATIONAL Condition 1, 2 and 3, two basins,##
b. In OPERATIONAL Condition 4, 5 and *, the basins ## associated with . and components required OPERABLE by Specifications 3.7.1.1 and 3.7.1.2. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, 5 and *. - ACTION:
- a. >
In OPERATIONAL CONDITION 1, 2, 3, 4. 5 and
- with one~SSW cooling tower basin inoperable, declare the associated SSW subsystem inoper- '
able and, if applicable, declare the HPCS service water system inoper-able, and take the ACTION required by Specifications-3.7.1.1 and 3.7.1.2, as applicable. The provisions of Specification 3.0.4 are not applicable for entry into OPERATIONAL CONDITIONS 4 and 5.** ,
- b. ,
In OPERATIONAL CONDITION 1, 2, 3, 4 or 5 with both nW cooling tower basins inoperable, declare the SSW system and the ffPCS service water system inoperable and take the ACTION required by Specifications 3.7.1.1 and 3.7.1.2. . c. In Operational Condition
- with both SSW cooling tower' basins inoperable, declare the SSW system inoperable and take the ACTION required by Specificatior.3.7.1.1, not applicable.
The provisions of Specification 3.0.3 are SURVEILLANCE REQUIREMENTS I 4.7.1.3 At least the above required SSW cooling tower basins shall' be deter-mined OPERABLE at least once per: a. l 24 hours by verifying basin water level to be~ greater than or equal to 87". b. 31 days by starting from the control room each SSW cooling tower fan not already in operation and operating each fan for at least 15 minutes.
- c. 18 months by verifying that each SSW cooling tower fan starts auto-matica11y when the associated S_SW subsystem is started.
(T.nsertA) When handling irradiated fue1 in the primary or secondary containment. [
# 7 The basin cooling tower fans are not required to be OPERABLE for HPCS service water system OPERABILITY.
i
##A n OPERABLE basin shall have a 30 day supply of water either self-contained by means of an OPERABLE siphon. **This exception is applicable until startup from the third refueling outage.
GRAND GULF-UNIT 1 3/4 7-4 Amendment No. 58j
NL 89/01
' INSERT A or suspended loads over irradiated -fuel I
e i
-i I'
L l l 1 1' l I A9001041/SNLICFLR - 11
NL' 99/61 ! PLANT SYSTEMS 3/4.7.2 CONTROL ROOM EMERGENCY FILTRATION SYSTEM LIMITING CONDITION FOR OPERATION i ' 3.7.2- Two independent control room emergency filtration system subsystems shall be OPERABLE. t APPLICABILITY: All OPERATIONAL CONDITIONS and *. ACTION:
- a. In OPERATIONAL CONDITION 1, 2 or 3 with one control room emergency
' filtration subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours,
- b. In OPERATIONAL CONDITION 4, 5 or *:
- 1. With one control room emergency filtration subsystem inoperable, ,
restore the inoperable subsystem to OPERABLE status within 7 days r i or initiate and maintain operation of the OPERABLE subsysten in the isolation mode of operation. The provisions of Specifica-
- tion 3.0.4 are not applicable for entry into OPERATIONAL CONDI- _.
TION 4 from 5.# QgtQ
- 2. With both control room emergency filtration subsystemi inoperable. I suspend CORE ALTERI.TIONS, handling of irradiated fuelnin tie l 9
primary or secondary containment and operations with a potential for draining the reactor vessel.
- c. 'The provisions of Specification 3.0.3 are not applicable in Operational Condition *.
SURVEILLANCE REQUIREMENTS 4.7.2 Each control room emergency filtration subsystem shall be demonstrated OPERABLE:
- a. At least oncs per 31 days on a STAGGERED TEST BASIS by initiating, from the cont.*ol room, flow through the HEPA filters and charcoal adsorbers and verifying that the subsystem operates for at least 10 continuous hoars with the heaters OPERABLE.
- b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the subsystem by:
- 1. [ DELETED) p@ se d 8)
"When irradiated fuel g ie being handled in the primary or secondary containment. #This exception is applicable until startup from the third refueling outage.
GRAND GULF-UNIT 1 3/4 7-5 Amendment No, 53, F
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NL 89/01 [ INSERT A or suspended loads over irradiated fuel . 6 i P F m d b I e A9001041/SNLICFLR - 11 u , , . w
m ~ s; NL 89/01 I!UiERI B or suspended loads over irradiated fuel are e 1 i l l A9001041/SNLICFLR - 13
NL' 89/01 ELECTRICAL POWER SYSTEMS-A.C. SOURCES - SHUTDOWN , 1
.LIMlllHG CONDITION FOR OPERATION i 3.8.1.2 As a minir.um, the.following A.C. electrical power sources shall be OPERABLE:
- a. One cirwit between the offsite transmission. network'and'the onsite Class 1E distribution system, and
- b. Diesel generator 11 or 12, and diesel generator 13.when the HPCS system is required to be OPERABLE, with each diesel generator having: ;
"I
- 1. A day tank.containing :a minimum.of-220 gallons of fuel.
- 2. A fuel storage system:containing.a minimum of:
a) 57,200 gallons of fuel for each OPERABLE diesel generator:11 3 or 12. 1 b) 39,000 gallons of fuel for die'sel generator 13. [
- 3. A fuel transfer pump.
APPLICABILITY: OPERATIONAL CONDITIONS 4, 5:and *. ACTION:-
- a. With all offsite circuits' inoperable and/or with' diesel generators 11 i and 12 inoperable,, suspend CORE ALTERATIONS, handling'of irradiated
] fuelain the primary or secondary containment, operations with a potential for draining the reactor-vessel and crane. operations over-l the spent fuel storage pool and the upper containment pool when fuel assemblies are stored therein.: In: addition, when in OPERATIONAL CONDITION 5 with the water level less than 22 feet 8 inches above the L reactor pressure vessel flange, immediately initiate corrective action to restore the required power sources to OPERABLE status as'soon as practical. .
- b. With diesel generator 13 inoperable, restore the inoperable. diesel generator 13 to OPERABLE status within 72 hours-or declare the HPCS system inoperable and take the ACTION required by Specification-3.5.2 and 3.5.3.
- c. The provisions of Specification 3.0.3 are not applicable. ,
SURVEILLANCE REQUIREMENTS , 4.8.1.2 At least the above required A.C. electrical power sources shall be demonstrated OPERABLE per Surveillance Requirements 4.8.1.1.1, 4.8.1.1.2 and-
- 4. 8.1.1.3, except for the requirement ' of 4 8.1.1. 2. a. 5..
*When handliEg irradiated fuel n in the primary or secondary contcinment.
l 5 GRAND GULF-UNIT 1 3/4 8-9 Amendment No. y S-l
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p NL89[DI-ELECTRICAL POWER SYSTEMS i D.C. SOURCES - SHUTDOWN j LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum Division 1 or Division 2,'and, when the HPCS system 'W is. required to be OPERABLE, Division 3, of the D.C. electrical power sources shall be' OPERABLE with:
- a. Division 1 consisting of:.
- 1. 125 volt battery 1A3. ;
- 2. 125 volt full. capacity charger IA4 or 1A5.
- b. Division 2 consisting of:
- 1. 125 volt. battery 1B3.
2, 125 volt' full-capacity charger 1B4 or 185.
- c. Division 3 consisting of: -
- 1. 125. volt battery 1C3. -
- 2. 125. volt full capacity charger 104, g APPLICABILITY: OPERATIONAL CONDITIONS 4, 5 and
- ACTION: nEs
- a. With both Division 1 battery and Division 2 battery of the above required D.C. electrical power sources-inoperable suspend CORE ALTERATIONS, handling of irradiated fuelAin the primary or secondary l containment and operations with a potential for draining the reactor vessel,
- b. With Division 3 battery of the above required D.C. electrical power sources inoperable, declare.the HPCS system inoperable and take the ,
ACTION required by Specification 3.5.2 and 3.5.3. 'l
- c. With any of the above required full capacity chargers inoperable,
! demonstrate the OPERABILITY of its assoolated battery bank by performing Surveillance' Requirement 4.8.2.1.a.1 within one hour and at least once per 8 hours thereafter. If any Category A limit in l
- Table 4.8.2.1-1 is not met, declare the battery inoperable.
- d. The provisions of Specification 3.0.3 are not applicable.
-l SURVEILLANCE REQUIREMENTS 4.8.2.2 At least the above required battery and charger shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.1.
x ^$ When handling irradiated fuel n in the primary or secondary containment. 1 GRAND GULF-UNIT 1 3/4 8-14 Arnendme.nf No, l
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ELECTRICAL POWER SYSTEMS '
. DISTRIBUTION - SHUTDOWN. t LIMITING CONDITION FOR OPERATION-3;8.3.2 As a minimum, the following power distribution system divisions shall be energized:
- a. For A.C. pour. distribution, Division 1 or Division 2, and when the HPCS system is required to be OPERABLE',; Division 3, with: .
, 1. Division 1 consisting of: a) 4160 volt A.C. bus 15AA. b) 480 vo1t'A.C. MCCs 15811, 15B21,~15B31,-15841,J15B51 and 15B61. c) 120 volt A.-C. distribution panels in 15P11', 15P21, 15P31, 15P41, 15P51 and 15P61, d)~ LCCs 15BA1,: 15BA2,;15BA3,15BA4,15BA5 and 15BA6.
- 2. Division 2 consisting of: ,
a) 4160 volt A.C. bus 16AB. . s
'b) 480 volt A.C. MCCs 16B11,<16B21- 16831, 16B41, 16B51 and 16B61. ,
c) 120 volt A.C. distribution' panels in 16P11, 16P21, 16P31, 16P41, 16P51 and 16P61. d) LCCs 16BB1, 16BB2, 16BB3, 16BB4, 16BB5 and 16BB6.
- 3. Division 3 consisting of: .
a). 4160 volt A.C.' bus 17AC. b) 480 volttA.C. MCCs 17B01 and.17B11 c) 120 volt A.C. distribution panels 17P11,'
- 4. The OPERABLE load shedding and sequencing panel associated with the division (s) required to.be energized, i b. For D.C. power distribution, Division 1 or Division 2, and when the HPCS system is required to be OPERABLE, Division 3, with:
- 1. Division 1 consisting of 125 volt D.C. distribution panel 1DA1 L and IDA2. !
- 2. Division 2 consisting of 125 volt D.C. distribution panel 10B1
~
- and 10B2.
- 3. Division 3 consisting of 125 volt D.C. distribution panel 1DC1.
p APPLICABILITY: OPERATIONAL CONDITIONS 4, 5 and *. A Ase a i l When handling irradiated fuel g in the primary or secondary containment. l I L l L GRAND GULF-UNIT 1. L 3/4 8-17 kmendment NO.
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s Nf A)L M/bl - - ELECTRICAL POWER SYSTEMS: pM(flNGCONDITIONFOROPERATION(Continued) r AC,f,(ON,:
- a. For A.C. power distribution:
- 1. With_both Division 1 and Division 2 of the above required A.C.-
distribution system not energized and/or with the load shedding' , and sequencing panel associated with the. division (s) required j Cna-to J
' ""*" "" '" "' '"'"*"# ' "" ^"^ "S of irradiated fuelAin the primary or secondary containment and l'
operations with a potential-for draining the reactor' vessel.
<2. With Division 31of the above required A.C. distribution' system not energized, declare the HPCS system inoperable and take the .:
ACTION required by Specification 3.5.2 and 3.5.3. .
~
, b. For D.C. power distribution:
, 1. With<both Division 1-and Division 2 of the above required D.C. distribution system-not energized, suspend CORE ALTERATIONS, (InseriA] nandling of irradiated fuelfin the primary or secondary con- l tainment and operations with a potential for draining the'
- reactor vessel.~ q
- 2. With Division 3 of the-above required D.C._ distribution sys. tem not energiz6d, declare the HPCS system inoperable and take-the ACTION required by Specification 3.5.2'and 3.5.3.
- c. The provisions of Specification 3.0.3 are not applicable.
- SURVEILLANCE REQUIREMENTS
[ 4.8.1.2.1 At least the above required power distribution' system divisions shall be determined energized at least once per 7 days by verifying correct ; breaker alignment on the busses /LCs/MCCs/ panels- and-voltage on the busses /LCs. g- . 4. H. 3. 2. 2 The above required load shedding and sequencing panel (s) shall be ! demonstrated OPERABLE:
- a. At least once per 12 hours by determining that the. auto-test' system is operating and is not indicating a faulted condition.
- b. At least once per 31 days by performance of a manual test and .
verifying response within the design criteria to the following test ( inputs: + a) LOCA. b) Bus undervoltage, c) Bus undervoltage followed by LOCA. d) LOCA followed by bus undervoltage. I I . GRAND GULF-UNIT 1 3/4 8-18 AmenclmenY No, b l
J NL 89/01; INSERT A' or suspended loads over irradiated fuel- t s
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1 x ' NL iM/of CONTAINMENT SYSTEMS-SASES 9 DRWELL VACUUM RELIEF'(Continued) ! tems associ,ated with three 10-inch lines penetrating the drywell.-T - , post-LOCA vacuus relief subsystems are in a parallel arrangement connected toTwo drywell
.one of the three 10-inch vacuus relief lines penetrating the drywell.
well post-LOCA vacuum relief. subsystem consists of a motor-operated isolationEach dry-valve in series with a check valve. OPERASILITY of either drywell post-LOCA relief line' penetrating the drywell. vacuum relief subsystem-assures 0PERABILITY ! Each of the two remaining 10-inch vacuum l relief lines penetrating the drywell contains a drywell purge vacuum relief. i subsystem. Each drywell purge vacuum relief subsystes consists of a series , arrangement of a motor-operated isolation valve and two check valves.- Vacuum j reliefor than initiates equal to at1apsi. differential pressure across.the check valves of less ;
- loadings to essential equipment and systems in the drywell abo ;
' Drywell negative pressure analysis for rapid weir wall overflow in a large-l break LOCA assumes a vacuum breaker capability of A//K = 0.38 ft8' thus re-quiring a minimum of two 10-inch drywell vacuum relief paths.'
j .
- OPERASILITY requirements for the four drywell vacuus relief subsystems in
- relationship to continued plant' operation are based on maintaining at least two p' of the three 10-inch drywell vacuum relief paths OPERABLE.-.4owever to ensure i that essential equipment is returned to service in a timely manner,, continued i
plant operation is lietted with only one 10-inch drywell vacuum relief Ifne out of service. Plant operation is further limited when two of the.three 10-inch Ifnes are out of service to ensure prompt response to restore equipment to l required.or to place the plant in a condition where the equipment.is not service l breaker Plant operation is also limited with a drywell isolation va:uum l age is nin the open position to help ensure that design drywell bypass leak-ot potentially exceeded. Position indication is required to be OPERA 8LE on all drywell vacuum breakers and motor-operated isolation valves to help identify potential drywell bypass leakage paths.. Surveillance requiressnts and intervals were chosen to reflect the impor-tance associated with the drywell vacuum relief function and are based on good 1 engineering judgement using previous accepted testing methods.
~
3/4.6.6 SECONDARY CONTAINMENT i l Secondary containment is designed to minimize any ground level release of i radioactive material which may result from an accident. ! and Enclosure Building provide secondary containment during normal operationThe Auxiliary I. when the containment is sealed and in service. When the reactor is in COLD I SHUTDOWN or REFUELING, the containment may be open and the Auxiliary Building and Enclosure Building then become the only containment. i The maximum isolation times for secondary containment automatic isolation l-dampers valves'with / valves are the analytical times closing used in the FSAR aJ:ident analysis for dampers / times. l-For automatic isolation valves not having L dRANDGULF-UNIT.1 8 3/4 6-8 \ Amendment No. 21) l F l
W , 1: NL 89/01 INSERT TO BASES 3/4.6.6
- The purpose of the requirement'to establish SECONDARY' CONTAINMENT-INTEGRITY when handling suspended loads directly over irradiated fuel
. assemblies is to limit the offsite, radiological consequences of'a postulated load drop' accident to no greater than 25% of 10CFR100 dose limits. . A suspended load if topped onto' irradiated . fuel' assemblies could produce enough.- l
' damage to the fuel cladding to cr.use a release of radioactive material which could result'in cffsite doses in excess of 25% of 10CFR100' dose limits.
~
However, suspended loads may be handled directly over irradiated fuel _ assemblies after 20 days. of decay time .in either the primary or secondary ~ containment without SECONDARY CONTAINMENT-INTEGRITY. The' decay time is
~ assumed to commence at the time-the fue1~ ass'emblies.are placed~in a subcritical condition (e.g., re' actor shutdown). SECONDARY CONTAINMENTL
-INTEGRITY is not required after 20 days because'if a' suspended load were to- -
drop onto irradiated fuel assemblies'following 20 dcys of' decay,-the: resultant ,I unfiltered release of radioactive material would not be sufficient to result . in offsite doses in excess.of 25% of 10CFR100 dose limits. j A suspended load is defined as a load: f
- 1) suspended by a lifting device (e.g., polar crane, jib crane) directly over irradiated fuel assemblies, and s ,
l
- 2) postulated to drop onto irradiated fuel assemblies, and i
- 3) if dropped onto irradiated' fuel assemblies could cause offsite }
releases in caxcess of 25% of 10CFR100 dose limits, j i Suspended loads which must be' postulated to drop 'are loads which are not ] handled in accordance with commitments implementing the NUREG-0612, Section 5.1.1 guidelines. 1 l
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