ML20086E216
| ML20086E216 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 07/03/1995 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20086E213 | List: |
| References | |
| NUDOCS 9507110276 | |
| Download: ML20086E216 (8) | |
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11 UNITED STATES E
NUCLEAR REGULATORY COMMISSION If WASHINGTON, D.C. 20666 4 001
,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION i
RELATED TO AMENDMENT NOS.119 AND 112TO FACILITY OPERATING LICENSE NOS. DPR-42 AND DPR-60 I
l' NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT NOS. 1 AND 2 DOCKET NOS. 50-282 AND 50-306
1.0 INTRODUCTION
By application dated December 5,1994, as supplemented January 9,1995, and 4
May 15, 1995, the Northern States Power Company (the licensee), requested amendments to the Prairie Island Nuclear Generating Plant, Units 1 and 2.
The proposed amendments would revise Technical Specification (TS) 3.8 to allow containment airlock doors and containment isolation valves to remain open during core alterations provided certain conditions are met.
The licensee subsequently withdrew isolation valves from the scope of the application in its letter dated May 15, 1995.
The January 9 and May 15, 1995, letters provided updated TS pages and clarifying information in response to discussions with the staff during various teleconferences conducted during the review process.
This information was within the scope of the original submittal and did not change the staff's initial propossd no significant hazards consideration determination.
2.0 DISCUSSION AND EVALUATION 2.1 Ob.iectives of the Proposed Amendment During a refueling outage there are approximately 250 personnel movements through the primary containment airlocks each day.
During periods of core alterations, TS 3.8 requires that at least one of the two doors in each airlock be closed at all times.
As a result of these TS requirements, each personnel movement into or out of the containment requires that each airlock door be cycled open and shut.
The airlock doors were not designed for such an amount of cycling and as a result, receive excessive wear and damage.
In addition to the accelerated damage and wear, the TS requirement to have one door closed during core alterations impedes personnel access and thus worker performance. The proposed amendments would increase the amount of outage time during which both doors in each airlock are allowed to be continuously open with the interlocks disabled.
This would expedite i
personnel movements into and out of containment and reduce airlock door corrective maintenance requirements. The proposed TSs would require that at least one of the i
airlock doors in each airlock be operable (i.e., in good working order and capable I
of being closed within 30 minutes notice) and under direct procedural controls
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i that assure that it can be closed as soon as practical following an accident.
Hoses, cables, platforms, or other devices which must be removed to close an airlock door would have to be capable of rapid disconnect and removal.
9507110276 950703 PDR ADOCK 05000282 P
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ns to be met if.
Included in the proposed amendments are certain additional conditip/ min) purge the airlocks doors are open. The containment high flow (33,000 ft t.be isolated by at least one valve or blind flange, and the low flow system mug/ min) purge system must be isolated or capable of automatic isolation.
(6,000 ft In addition, at least two containment cooling system fans must be operable (i.e.,
t fans normily operating at low speed and switchable to high speed in the event of a fuel handling accident) to ensure mixing.
l 2.2 Core Alterations - Applicability h
I The proposed changes would apply during shutdown operations when core alterations are taking place. A core alteration, as defined in the facility TSs, is the j
movement or manipulation of any component within the reactor vessel with the vessel head removed and fuel in the vessel, which may affect core reactivity.
During core alterations the following conditions, ir addition to those applicable to the airlocks and piping penetrations, must be met The equipment hatch must be closed.
Radiation levels in the fuel handling area of the containment must be continuously monitored.
i The core neutron flux must be continuously monitored if core geometry l
is being changed.
At least 23 feet of water must be maintained over the level of the vessel flange during fuel movement or control rod outward movement.
At least one residual heat removal [RHR] pump must be operable and running (pump may be shut down for up to I hour).
Both shall be running if the water level above the top of the vessel flange is less than 20 feet except for control rod latching / unlatching operations or upper internals removal / replacement.
The reactor must have been subcritical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to movement of irradiated fuel in the reactor.
Direct communication between the control room and operating floor of the containment must be available, and The radiation monitors that initiate isolation of the containment purge system must be tested and verified operable.
These are existing TS requirements and would remain unaffected by the proposed amendments.
1 2.3 Descriptions of Affected Systems and Eauipment A description of the Prairie Island containment system is provided below. Also, a schematic diagram is attached.
This diagram was taken from the Final Safety Analysis Report (FSAR) Figures 5.1-1 and 5.2-10.
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. 2.3.1 Primary Containment and Airlocks The primary containment is a freestanding, cylindrical steel pressure vessel enclosed within a concrete shield building. The primary containment is designed for a maximum internal pressure of 46 paig and temperature of 268'F.
Each primary containment at Prairie Island is provided with two airlocks, a personnel airlock and a maintenance airlock. The maintenance airlock receives the greater usage during outages.
Each airlock consists of a chamber with manually operated, interlocked doors at each end of the chamber. The airlocks permit personnel to enter and exit the primary containment without creating an open release pathway between the interior of the containment and the outside environment. The personnel airlock connects the primary containment wiih the auxiliary building as shown in FSAR Figures 12.1-6/7.
The maintenance airlock connects the primary containment with a staging area known as the " basketball court." This arrangement is depicted in FSAR Figures 12.1-11/12.
2.3.2 Secondary Containment (Shield Buildina)
A shield building surrounding each primary containment serves as a secondary containment providing an annulus space for the collection of primary containment fission product leakage effluent.
It is constructed of reinforced concrete and provides radiation shielding for operational radiation protection.
In the event of a design basis accident-loss-of-coolant accident (DBA-LOCA) during power operation, the shield building vent system will produce a negative pressure in the annulus and recirculate the annulus atmosphere through a filter system prior to release. The secondary containment has no accident mitigation function for refueling operations and need not be operable during refueling.
2.3.3 Purae and Filtration Systems 5/ min high flow system Two containment purge systems are provided, a 33,000 f}/ min low flow system havin having 36-inch containment penetrations and a 4,000 ft 3
18-inch penetrations.
(NOTE: The low flow system was designed as a 4,000 ft / min 3
system but has a measured flow of 6,000 ft / min.) The 36-inch lines of the high i
flow system are normally blanked-off using blind flanges.
The 18-inch lines of the low flow system contain provisions for spool-pieces in the annulus.
Blank
)
flanges are normally installed in the 18-inch lines but are removed and replaced with the spool-pieces during outages.
The Prairie Island facilities also contain a containment internal clean-up system which is installed for the nonsafety purpose of permitting expedited personnel entry into containment at power. Also, the fuel handling building and the shield building are each provided with safety-grade ventilation supply and exhaust systems having charcoal filters.
No credit is taken for the fission product removal capability of these systems as they are not required to be operable during refueling.
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2.4 Radioloaical Dose Conseauences l
The controls established by the TSs during core alterations are based on the need to ensure acceptable dose consequences in the event of a fuel handling accident (FHA) in containment.
The FHA is the limiting DBA postulated to occur during core alterations and serves as the basis for the design of any fission product control and cleanup systems required to be operable during core alterations to ensure acceptable dose consequences. Due to the low reactor coolant temperature and decay heat level during core alterations, the high containment pressure associated with a DBA-LOCA cannot occur.
In calculations of the dose consequences of an FHA, the design primary containment leakage rate is thus not a factor. The release flow rate assumed for the purpose of dose calculations is determined by the flow rate of any operating purge or vent systems and the tested exfiltration rate of any pressurized secondary containment or other fission product control barrier.
The staff analyzed the potential radiological consequences of an FHA.
In the new calculationthea}mosphericdispersionfactor(x/Q)hasbeenfurtherincreased 3
from 4.7E-4 sec/m to 6.U-4 sec/m to reflect historical meteorological data and the fuel assembly fission product inventories in the new dose calculation which encompasses the effects of extended fuel burn-up. This results in a considerably increased thyroid dose. However, the effects of limited mixing and dilution in the containment building have been included. Standard Review Plan (SRP) 15.7.4 allows the radiological consequences of an FHA to be reduced by the degree of mixing and dilution occurring in containment prior to containment isolation.
The dose equation includes a containment retention factor (C) based on the containment free volume and the containment purge flow rate that reduces the offsite dose to the public due to the mixing and dilution taking place in containment.
Because the containment airlop/ min.
ks are open, the cop / min leakage rate is based on continued tainment will be assumed to be leaking at a rate of 6,000 ft The 6,000 ft operation of the operating train of the containment low flow purge system.
(This system, which has charcoal filters, would actually isolate on containment high radiation.) The credit for mixing considers immediate mixjng of the pool effluent (fission product gases leaving the pool) with 1,000,000 ft of the containment volume. The mixing credit is based on operation of one train of contajnment fan coolers which provides for forced mixing of approximately 1,000,000 ft of the free volume pf the primary containment.3 The licensee has verified that over 1,000,000 ft of the total 1,320,000 ft of containment volume are freely contiguous to the fan-cooled areas of the containment.
The staff computed the offsite doses for Prairie Island using the above assumptions and NRC computer code ACTICODE. Control room operator doses were determined using the methodology in SRP Section 6.4.
The computed offsite doses and control room operator doses are within the acceptance criteria given in SRP Section 15.7.4 and General Design Criterion (GDC) 19. The assumptions used in calculating those doses and the resulting calculated values are provided in attachments 1 and 2.
As shown in attachment 2, the calculated doses are within the staff's acceptance criteria.
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- 2.5 Conclusion Analyses of radiological consequences of a fuel handling accident, with the primary containment airlocks open, confirm that dose acceptance criteria for the FHA are met.
Accordingly, the staff finds the licensee's proposal to revise the TSs acceptable.
In addition, the staff grants the licensee's request in its letter of May 15, 1995, to withdraw all aspects of the application concerning opening of containment penetrations during core alterations.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Minnesota State official was notified of the proposed issuance of the amendments.
The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types. of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration and there has been no public comment on such finding (60 FR 6306). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22 c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
5.0 CONCLUSION
The staff has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the prcposed manner, (2) such activities will be conducted in compliance with the Connission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors:
W. Long D. Carter Date:
July 3, 1995 Attachments:
1.
Staff Assumptions Used for Calculating Radiological Consequences 2.
Calculated Radiological Consequences 3.
FSAR Schematic Diagram
6 STAFF ASSUMPTIONS USED FOR CALCULATING RADIOLOGICAL CONSE0VENCES
' Parameters Quantity Power Level, Mwt 1,650 Number of Fuel Rods Damaged (1 assembly) 179 Total Number of Rods-(121 assemblies) 21,659 Shutdown time, hours 100 Power Peaking Factor 1.65 Fission Product Release Duration, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Core Fission Product Inventories per TID-14844 Fuel Burnup (MWD /MTU) 60,000 Release Rate from Containment 6000 scfh i
Effective Containment Mixing Volume 1.0 x 10' Receptor Point Variables Exclusion Area Boundary 3
Atmospheric Relative Concentration, X/Q (sec/m )"
0-2 hours 6.5 x 10
Low Population Zone 3
Atmospheric Relative Concentration, X/Q (sec/m )"
0-2 hours 1.77 x 10
8-24 hours 3.99 x 10'3 1-4 days 7.12 x 10
4-30 days 1.04 x 10
Control Room 3
Atmospheric Relative Concentration, X/Q (sec/m )"
5.58 x 10
Control Room Volume, cubic fept 4.42 x 10' Maximum Infiltration Rate, ft / min 44 Geometry Factor 31.56 1
Iodine Protection Factor 64.8 Recirculation Air Flow" 3
Flow Rate, ft / min 3000 ESF Filter Efficiency Elemental Iodine 95%
Organic Iodine 95%
Particulate Iodine 95%
Note: Dose conversion factors from ICRP-30 were utilized for all calculations Regulatory Guide 1.25 t
- Prairie Island FSAR
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CALCULATED RADIOLOGICAL CONSE0VENCES i
(rem) l Exclusion Area Boundarv Qgig SRP 15.7.4 Acceptance i
Criterion Whole Body 0.6 6
Thyroid 61 75-l Control Room Operator Daig GDC-19 Accentingg j
Criterion l
Whole Body
<0.1 5
Thyroid 1.6 Equiyalent to 5 rem whole~
body Section 6.4 of the Standard Review Plan defines the thyroid dose acceptance l
criterion as 30 rem.
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