ML20248J882
| ML20248J882 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 04/06/1989 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20248J877 | List: |
| References | |
| NUDOCS 8904170132 | |
| Download: ML20248J882 (7) | |
Text
.
{15, .
-y .
V :'
- TS.3.8-3t. .I
=N ;/n/C
~
D. Spent Fuel Pool Special Ventilation Svstem l l L1. Except as specified in Specification 3.8.D.3 below, both trains of the Spent' Fuel' Pool Special Ventilation' System and the diesel generators j!
required for their operation shall be operable at all times. I
- 2. a. The;results of in-place DOP and halogenated hydrocarbon tests at design flows on HEPA filters and charcoal adsorber banks respec- !
I- ,tively'shall-show 1 99% DOP removal for particles'having a mean.
diameter of 0.7 microns and 1 99% halogenated hydrocarbon removal,
~
q
- b. The results of laboratory carbon sample analysis shall show 1 90% radioactive methyl-. iodide removal' efficiency (130 C,.95% RH)..
't
.c. The Spent Fuel Pool Special Ventilation System fans shall operate i within i 10% of 5200 cfm psr train.
]
- 3. From and'after.the date'that'one train of the Spent Fuel Pool Special l: . Ventilation System is made or found inoperable for any reason, fuel l . handling operations are permissible only during the succeeding seven j days'(unless such train is made operable) provided that the redundant ,
train is verified to be. operable daily. !
- 4. If the conditions for operability of the Spent Fuel Pool.Special'
~. Ventilation System cannot be met, fuel handling operations'in the Auxiliary Building shall be terminated immediately.
M 6 e rt 6 - **
~
1 Basis The equipment and general procedures to be utilized during refueling are dis- I cussed in,the FSAR. Detailed instructions, the precautions specified.above, )
and the design of the fuel-handling equipment incorporating built-in inter- t locks and safety features, provide assurance that no incident could occur during the refueling operations that would result in a hazard to public. health ~
and safety.(1) Whenever changes are not being made in core geometry, one flux monitor is sufficient. This permits maintenance of the instrumentation.
Continuous monitoring of radiation levels (B. above) and neutron flux provides ;
immediate indication of an unsafe condition. The residual heat removal pump
.i is used to maintain a uniform boron concentration.
Under rodded and unrodded conditions, the Kegg of the reactor must be < 0.95 ,
and the boron concentration must be 1 2000 ppm as indicated in A.4. Periodic i checks of refueling water boron concentration insure that proper shutdown ')
margin is maintained. A.9 above' allows the control room operator to inform the manipulator operator of any impending unsafe condition detected from the
. main control board indicators during fuel movement.
_. f -
\ l
%[$
P J
' '('
. l l
Insert A:
l l
Storage of_ Low Burnup Fuel '
E. l
- 1. The following restrictions shall apply whenever fuel with an average assembly burnup less than 5,000 MWD /MTU is stored in the spent fuel pool (except as specified in 3.8.E.2 and 3.8.E.3 below):
- a. The boron concentration in the spent fuel pool shall be maintained greater than or equal to 500 ppm, and
- b. Fuel with an average assembly burnup less than 5,000 MWD /MTU shall not be stored in more than three storage locations of every four by four storage rack array.
- 2. If the conditions in 3.8.E.1.a above are not met, verify that the spent fuel pool storage configuration meets the requirements of specification 3.8.E.1.b and suspend all actions involving the movement of fuel in the spent fuel pool until the boron concentration is increased to 500 ppm or greater.
- 3. If the conditions in 3.8.E.1.b above are not met, suspend all actions involving movement of fuel in the spent fuel pool, verify the spent fuel pool boron concentration to be greater than or equal to 500 ppm and initiate corrective actions. Mis-positioned fuel assemblies shall be moved to acceptable locations prior to the resumption of other fuel movement in the spent fuel pool.
N- K Q
l TS.3.8-5
- -Rg" 7 4 0/20/03 1
l
(/ to the environment via one of the Shield Building exhaust stacks. Two l )
completely redundant trains are provided. The exhaust fan and filter of each I train are shared with the corresponding train of the Containment In-service Purge System. High efficiency particulate absolute (REPA) filters are installed before the charcoal adsorbers to prevent clogging of the iodine !
adsorbers in each SFPSVS filter train. The charcoal adsorbers are installed I to reduce the potential release of radioiodine to the environment. The in-place test results should indicate a HEPA filter leakage of less than 1%
through DOP testing and a charcoal adsorber leakage of less than 1% through halogenated hydrocarbon testing. The laboratory carbon sample test results should indicate a radioactive methyl iodide removal efficiency of at least 90% under test conditions which are more severe than accident conditions.
The satisfactory completion of these periodic tests combined with the quali-fication testing conducted on new filters and adsorber provide a high level of assurance that the emergency air treatment systems will perform as predicted in the accident analyses.
During movement of irradiated fuel assemblies or control rods, a water level ;
of 23 feet is maintained to provide sufficient shielding. l The water level may be lowered to the top of the RCCA drive shafts for latching and unlatching. The water level may also be lowered below 20 feet for upper internals removal / replacement. The bases for these allowances are (1) the refueling cavity pool has sufficient level to allow time to initiate repairs or emergency procedures co cool the core, (2) during latching /unlacching and upper internals removal / replacement the level is closely monitored because the
( activity uses the2 level as a reference point, (2) the time spent at this level is minima uirements for the storage of low burnup fuel in the spent fuci pool
,(, ensure that the spent fuel pool will remain suberitical during fuel storage.
The maximum enrichment of fuel stored in the spent fuel pool will be limited to a batch average of 4.25 weight percent U-235. It has been shown by criticality analysis that the use of the three out of four storage configuration. will assure that the K,ff will remain less than 0.95 when fuel with a maximum batch average enrichment of 4.25 weight percent U-235 and average assembly burnup of less than 5,000 MWD /MTU is stored in the spent fuel pool.
/
The requirement for maintaining the spent fuel pool boron concentration greater than 500 ppm whenever fuel with average assembly burnup of less than 5,000 MWD /MTU is stored in the spent fuel pool ensures that K.f, for the spent
, fuel pool will remain less than 0.95 even if a fuel assembly is inadvertently inserted in the empty cell of the 3 out of 4 storage configuration.
\
I s
References
+'" J .J. A _J~ J~ J (1) FSAR Section 9.5.2 (2) FSAR Section 14.2.1 (3) FSAR Section 9.6 (4) FSAR Page 9.5-20a (5) Exhibit C, NSP License Amendment Request Dated December 21, 1984 l 1
L- ,
i .
TS.5.3-1 l Z'l 00 11/14/00 l
. 5.3 REACTOR I
f A. Reactor Core
- 1. The reactor core contains ;ppscxieetely 40 ;;tric t;;; cf uranium in the form of slightly enriched uranium dioxide pellets. The pellets are encapsulated in Zircaloy-4 tubing to form fuel rods.
The reactor' core is made up of 121 fuel assemblies. Each fuel mbi cont _ain erence 1).
2.
e boLi~
..: cv......n.......... .... r:Is;; c... .. :
n::g 2.
.....~w......
LN 23 Ei.
-per cent of U 235. O.: high::t Uraniu 235 le::!ing i: : n::inal '
l O') grass of U-235 per exial centimeter of fuci :::::bly ( v:::g;) .
- 3. In the reactor core, there'are 29 full-length RCC assemblies that contain a 142-inch length of silver-indium-cadmium alloy clad with stainless steel (Reference 2). l 1 B. Reactor Coolant System l
l-
- 1. The design of the reactor coolant system complies with all appli- l cable code requirements (Reference 3). .j l t
- 2. All high pressure piping, components of the reactor coolant {
system and their supporting structures are designed to Class I j requirements, and have been designed to withstand: 'j
. a
.h
- a. .The design seismic ground acceleration, 0.06g acting in the horizontal and 0.04g acting in the vertical planes simultane- l ously, with stresses maintained within code allowable working i stresses.
- b. The maximum potential seismic ground acceleration, 0.12C, acting in the horizontal and 0.08g acting in the vertical planes simultaneously with no loss of function. j
- 3. The nominal liquid volume of the reactor coolant system, at rated operating conditions, is 6100 cubic feet.
i C. Protection Systems The protection systems for the reactor and engineered safety features are designed to applicable codes, including IEEE-279, dated 1968. The design includes a reactor trip for a high negative rate of change of neutron flux as measured by the excore nuclear instruments (Reference
- 4) ,, The system is intended to trip the reactor upon the abnormal dropping of more than one control red (Reference 4). If only one control rod is dropped, the core can be operated at full power for a short time, as permitted by Specification 3.10.
References
- f. USAR, Section 7.1
\, 2. USAR, Section 3.5.2 4.
TS.5.6-1 RE" 00 11/14/00 =
( 5.6 FUEL HANDLING A. Criticality Consideration The new and spent fuel pit structures are designed to withstand the ,
anticipated earthquake loadings as Class I (seismic) structures. The l spent fuel pit has a stainless steel liner to ensure against loss of j j
water (Reference 1). l l
The new and spent fuel storage racks are designed so that it is impossible to insert assemblies in other than the prescribed locations, hurt 6 The feel ie etered verticelly in on array with th; : nt r-: -: nt r distan:c between aasc;blies sufficient to assure K 1 webeseted w:ter ;;;; used Oc fill the pit. Inadd!!!ca,0.03cvenif fu;l in the a;erage poci shell he n e U-205 leeding of ,100.0 cr e Of U-235 pt exici centic ::: cf fuel ::::bly (eve-eate4.
The criticality considerations as they relate to the dropping of a spent fuel cask (i.e. , heavy load) drop onto the racks has been evalu- ;
ated. The maximum K has been> calculated to be 0.949 at a water /UO ratioofa2.0withaboronconcentrationof1800 ppm. I B. Spent Fuel Storage Structure The spent fuel storage pool is enclosed with a reinforced concrete building having 12- to 18-inch thick walls and roof (Reference 1). l
{ The pool and pool enclosure are Class I (seismic) structures that afford protection against loss of integrity from postulated tornado missiles. The storage compartments and the fuel transfer canal are connected by fuel transfer slots that can be closed off with pneumatically sealed gates. The bottoms of the slots are above the tops of the active fuel in the fuel assemblies which will be stored vertically in specially constructed racks.
The spent fuel pool has a reinforced concrete bottom slab nearly 6 feet thick and has been designed to minimize loss of water due to a dropped cask accident. In addition, the spent fuel cask will have an impact limiter attached or a crash pad will be in place in the pool which will have the capability to absorb energy of impact due to a cask drop. This will result in no structural damage taking place to the pool which would result in significant leakage from the pool.
Piping to,the pool is arranged so that failure of any pipe cannot drain the pool below the tops of the stored fuel assemblies.
C. Fuel Handling The fuel handling system provides the means of transporting and handling fuel from the time it reaches the plant in an unirradiated condition until it leaves after post-irradiation cooling. The system consists of the refueling cavity, the fuel transfer system, the spent fuel storage pit, and the spent fuel cask transfer system. .
L _ _ _ _ _ _ . - _ _ _ _ . _ _ _ . _ _ .
Insert B:
The design of the new fuel storage pit and racks (Reference 1) ensures a new fuel pit K ,f of less than or equal to 0.95 even if unborated water were used to fill the pit. The new fuel rack configuration also ensures
~
K.,, less than or fqual to 0.98 even if the new fuel racks were accidentally filled with a low density moderator which resulted in optimum low density moderation conditions. The maximum batch average enrichment of fuel to be stored in the new fuel storage racks will be 4.25 weight percent U-235.
The spent fuel storage rack design (Reference 1) and the limitations on the storage of low burnup fuel contained in Technical Specification Section 3.8.E ensure a spent fuel pool K ,, of less than or equal to O.95.
The maximum batch average enrichmer.t of fuel to be stored in the spent fuci pool will be 4.25 weight percent U-235.
A l
l l
l i
a l
l l
I l
i s .
- - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ __ i
TS.S.6-2 ,
ROJ 00 11/11/S$ j l
( Major components of the fuel handling system are the manipulation crane, the spent fuel pool bridge, the auxiliary building crane, the fuel transfer system, the spent fuel storage racks, the spent fuel cask, and the rod cluster control changing fixture. The reactor vessel stud tensioner, the reactor vessel head lif ting device, and the reactor internals lifting device are used for preparing the reactor for refueling and for assembling the reactor after refueling.
Upon arrival in the storage pit, spent fuel will be removed from the transfer system and placed, one assembly at a time, in storage racks ,
using a long-handled manual tool suspended from the spent fuel pit bridge crane. After sufficient decay, the fuel will be loaded into shipping casks for removal from the site. The casks will be handled by the auxiliary building crane.
The load drop consequences of a spent fuel cask for Prairie Island have been evaluated. It is not possible, due to physical constraints, for a cask to be dropped into the large pool (pool no. 2). A load path has been defined which provides for safe movement of the cask. Travel interlocks and mechanical stops prevent cask movement outside of this ,
path. The only safety-related equipment that ca'n be impacted directly ,
during a cask drop along this path is the fuel stored in the small pool
'(pool no. 1). The consequences of this drop have been evaluated and i
found to meet the NRC staff criteria contained in NUREG-0612 if at least 50 days have elapsed since reactor shutdown for fission gas release considerations and the pool water contains at least 1800 ppm
( boron for criticality considerations. While 50 days was determined adequate, a minimum decay period of 5 years has been incorporated into these technical specifications to provide additional margin in meeting the criteria specified in NUREG-0612 for fission gas releases, while not restricting the plant's operational flexibility. A cask impact limiter or crash pad prevents significant structural damage to the pool floor.
The spent fuel cask will be lowered 66 feet from the auxiliary building to the railroad car for offsite transportation. Specification 3.8 will limit this loading operation so that if the cask drops 66 feet, there will not be a significant release of fission products from the fuel in the task.
D. Spent Fuel Storage Capacity The' spent fuel storage facility is a two-compartment pool that, if completely filled with fuel storage racks, provides up to 1582 storage locations. The southeast corner of the small pool (pool no. 1) also serves as the cask lay down area. During times when the cask is being used, four racks are removed from the small pool. With the four ;
storage racks in the southeast corner of pool I removed, a total of i 1386 storage locations are provided. To allow insertion of a shipping cask, total storage is limited to 1386 assemblies, not including those assemblies which can be returned to the reactor. '
i Reference 1.
USAR,Section10.P@
o--_