ML20154A997
| ML20154A997 | |
| Person / Time | |
|---|---|
| Issue date: | 08/24/1988 |
| From: | Liza Cunningham Office of Nuclear Reactor Regulation |
| To: | Ebneter S, Norelius C, Stohr J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I), NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II), NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| NUDOCS 8809130167 | |
| Download: ML20154A997 (15) | |
Text
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MEMORANDUM FOR:
Stewart D. Ebneter, Director, DRSS, Region I J. Philip Stohr, Director, DRSS, Region II Charles E. Norelius, Director, DRSS, Region III Richard L. Bangart, Director, DRSS, Region IV Ross A. Scarano, Director, DRSS, Region V FROM:
LeMoine J. Cunningham, Chief Radiation Protection Branch Division of Radiation Protection and Emergency Preparedness Office of Nuclear Reactor Regulation
SUBJECT:
TECHNICAL
SUMMARY
OF TI 2500/23 RESPONSES On March 13, 1987 Temporary Instruction 2500/23 was issued requesting an evaluation of radiological controls for drywell work during spent fuel movement.
We have received responses from all operating (34) BWR plants.
These responses (as well as supplementary information received via phone from regional inspectors) formed the basis for the enclosed Evaluation of Responses to Temporary Instruction 2500/23.
We appreciate the efforts and the support we received from all regions and from individual inspectors.
No further generic action is needed.
If you have questions or would like to suggest revision of a specific plant response item, please contact Frank Skopec PRPB, x23153.
/5/
LeMoine J. Cunningham, Chief 8009130167 000024 Radiation Protection Branch PDR ORG NRRDPNV Division of Radiation Protection and Emergency Preparedness Office of Nuclear Reactor Regulation
Enclosures:
1.
Evaluation of Responses to TI 2500/23 2.
TI 2500/23 CONTACT:
Frank C. Skopec, NRR 492-3153 Distribution:
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ENCLOSURE 1 EVALUATION OF RESPONSES TO TEMPORARY INSTRUCTION 2500/23, "EVALUATE BWR LICENSEE RADIOLOGICAL CONTROLS FOR DRYWELL DURING SPENT FUEL MOVEMENT."
1.
INTRODUCTION It has been recognized for some time that the spent fuel transfer from reactor vessel into fuel storage could cause potentially high transient radiation levels in upper elevations of the drywell in Mark I, II, and III BWRs.
Drywell workers could receive near fatal radiation exposure, if during such transfer, the spent fuel assembly would fall (fuel drop accident).
For this reason, the General Electric Company issued Service Information Letters (in 1973 and 1980) to applicable BWR licensees, informing them of this potential problem and recommended corrective actions.
To ensure that all BWR licensees have instituted effective radiological controls and to protect drywell workers from the high radiation exposures, the NRC issued (March 13,1987) Temporary Instruction 2500/23, "Evaluate BWR Licensee Radiological Controls for Drywell During Spent Fuel Movement," Enclosure 2.
The purpose of this memo is to integrate the regional responses to the Temporary Instruction (TI) 2500/23 and to summarize and evaluate radiological controls taken by the licensees, initiate improvements where necessary, and to ensure that drywell workers will be protected from this potentially hazardous condition in the future.
Y J
2.
DISCUSSION / EVALUATION Regional inspectors incorporated the TI 2500/23 with their routinely scheduled inspections, and reported the findings on Technical Summary Form, Attachment 4 to TI 2500/23.
All applicable BWR plants were inspected. The' Attachment 4 (a part of Enclosure 2) consist of 9 questions and procedure review items, which were reviewed with the licensees and the inspection findings were reported by the regional inspectors.
8 l
1
The regional responses to the Attachment 4 questions and procedure review items form the basis of this evaluation of BWR licensee radiological controls for drywell during the spent fuel transfer.
The following Table is a summary of the regional findings.
Table 1 Summary of TI 2500/23 responses Fuel chute Replies to items Areas in which shielding (inches) of Attachment 4:
improvements were lower made as result of Plant name bottom side 123456789 TI 2500/23 Browns Ferry,1,2,3 6 Pb 6 Pb YYYYYNYYY N None Brunwick, 1,2 6 Pb 0
YYYYYNYYY Y Fuel Chute Design and Shielding Improvements Clinton, 1 4 Pb (Permanent)
YYYYYNYYN Y Personnel Training Cooper SPb+11/2 e 0
NYYYYYYYY Y Radiological Controls F
Dresden, 2,3 SPb+11/2 e 0
YYYYYNNY0 Y Procedures, Controls F
Duane Arnold 6 Pb 0
Y0YYYYNNN Y Training, Procedures FermiQ 0
0 000000000 0 Will Meet Objectives Of TI 2500/23 Fitzpatrick 6 Pb 0
YYYYYNYYY N None Grand Gulf, 1 6Pb+1/4 e 31/2 e YYYYYYYYY Y Procedures, Controls F
F Hatch, 1,2 6Pb+1/2 e 4Pb+1/2 e YYYYYYNYY Y Oper. Procedures F
F Hope Creek, 1**
6 Pb 21/4 Pb YYYYYNYYY Y Licensee Is Aware Of High Dose Rates LaSalle, 1,2 6 Pb 8 Fe YYYYYYYNN Y Training, Procedures mimerick, 1 SPb+11/2 e 0
YYYYYNYYY N None F
- Millstone,1 6 Pb 0
YYYYYNYYY N None Monticello 6 Pb 0
YYYYYNYYY N None Nine Mile, 1,2 6 Pb 0
YYYYYNYYY Y Radiological Controls Oyster Creek, 1 9 Fe 6 Fe YNYYYNYYY N None Peach Bottom, 2,3 6Pb+1Fe 0
YYYYYNNNN Y Traininr, Procedures Porry, 1 6 Pb 1/2Pb+21/2 e YYYYY?.YYY Y Procedu'res F
Pilgrim, 1 6Pb+3/ Fe 0
NY0YYYYNY Y Training, Licensee 4
Awareness Quad Cities, 1,2 6 Pb 0
YYYYY0YYN Y Training, Controls River Bend, 1 51/s+11/sSS 6t/4 Fe YYYYYNYYN Y Training, Controls Susquehanna, 1,2 8 Pb (Permanent)
YYYYYNYNY N None Vermont Yankee SPb+11/2Fe 0
YYYYYYYNY Y. Training, Procedures Washington (WNP) 91/2 e 6 Fe Y0YYYNYYY N None F
i NOTE:
Y=yes; N=no; 0=information not received
- FERMI's First fuel movement is planned in 1989; presently fuel chute is not available, licen ee is developing policy and procedures.
- First refueling and fuel shielding is in planning.
From the Table 1, it can be seen that all applicable BWR facilities have available shielding to protect personnel in drywell from the fuel drop accident.
Most plants have removable shielding.
At nearly all facilities, the 2
equivalent of six inches of lead is used to protect the personnel in drywell from the spent fuel above.
Shield thickness on the side of fuel chute is usually tapered and therefore more complex to describe and probably therefore it was not described in most TI 2500/23 responses.
In some cases, regions also reported radiation dose rate measurements, or the results of calculations, at the accessible locations in the drywell from the spent fuel above.
Following is the evaluation of the replies to questions and procedure review items.
QUESTIONS REPLIES (%)
1.
Did Licensee fully understand drywell hazards from spent fuel YES:
94 movements?
If no, briefly explain on the back of this page.
NO :
6 2.
Did licensee have GE information letters (1973, 1980)?
YES:
97 NO :
3 3.
When did licensee first implement effective drywell controls YES: 100 (e.g., at first refueling)?
NO :
0 4.
Is drywell access allowed during fuel movements? If answer YES: 100 is no, then stop:
You are done.
If yes, answer the NO :
0 remaining questions.
5.
Briefly describe on the back of this page shielding for fuel-YES: 100 chute area (e.g., how much Pb).
NO :
O PROCEDURE REVIEW REPLIES (%)
6.
Describe on the back of this page any significant weaknesses YES:
28 in implementing procedures for fuel handling and HP drywell NO:
72 control point.
/
7.
Is overall procedural control adequate to protect w'rkers YES:
80 o
from drywell hazards? If no, describe on the back of this NO:
20 page licensee commitments to upgrade / correct.
8.
Is there specific training to provide all affected workers an YES:
74 adequate understanding of the drywell hazards?
NO:
26 9.
Does this training provide clearly defined worker actions, YES:
73 NO:
27 responsibilities during normal and emergency) operations l
(e.g., in event of drywell evacuation signal ?
Table 1 indicates that improvements were (or will be) made at approximately 70%
of all applicable BWR's as result of the TI 2500/23 issuance.
3 l
Most of the required improvements were made in the following areas:
Radiological controls at 11 plants, Procedures at 10 plants, Personnel training at 9 plants, Licensee awareness at 2 plants, Design improvements at 1 plant.
Some plants improved in more than one area.
The regional inspectors have confirmed that approximately 30% of BWR plants had previously implemented radiological programatic drywell controls that fully address the TI specific concerns during the spent fuel transfer, so that improvement were not necessary.
3.
CONCLUSION From the 34 inspected BWR plants, 24 were judged to need improvements in one or more areas.
The most frequent improvements were made in radiological controls,'
procedures and personnel training, in that order.
In two cases, the licensee was not fully aware of the magnitude of the personnel exposure due to the pos-sible fuel drop accident.
In one case fuel chute design and shielding needed substantial improvement.
~
Most common radiological control deficiencies included inadequate HP corbuni-cation between fuel operators and drywell control point; instructions on alarm set point; need for revision of HP procedures; need for improved access control to the upper drywell area; personnel accountability within drywell prior to fuel movement; verification that drywell radiation monitors are operational with appropriate warning lights and alarms; verification that posting and positive controls are established in drywell; and coordination of fuel movement activities between HP and operations department.
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> 7, Most' common procedural deficiencies included operating procedures that did not require HP to be notified prior to fuel movement, nor did they address communication requirement with drywell access point. Also, procedures did not address the potential hazard associated with fuel moved too close to the reactor pressure vessel wall; the procedures did not specify a distance (in feet) that the fuel must remain away from the reactor vessel wall; nor did they address personnel access restrictions to protect drywell personnel during fuel transfer.
TheobjectiveoftheTI 2500/23 is to ensure that all BWR licensees have instituted lasting and effective programatic radiological controls to protect drywell workers from the potentially high transient dose rates caused by spent fuel r.avement or from a possible fuel drop accident.
Based on the regional responses summarized above, we conclude that the objective of the Temporary Instruction 2500/23 was achieved.
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NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT f
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INSPECTION AND ENFORCEMENT MANUAL DEPER TEMPORARY INSTRUCTION 2500/23
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EVALUATE BWR LICENSEE RADIOLOGICAL CONTROLS FOR DRYWELL DURING SPENT FUEL MOVEMENT 2500/23-01 PURPOSE To evaluate the radiological controls of BWR licensees for access to and work in the drywell during spent (irradiated) fuel movements.
2500/23-02 OBJECTIVE To ensure that all BWR licensees have instituted lasting, effective programmatic radiological controls to protect drywell workers from the This,
potential high, transient dose rates caused by spent fuel movements.
regional inspection is the first generic ef fort to systematically focus on this potentially hazardous area.
2500/23-03 RESPONSIBILITIES Establishes the 03.01 Director. Office of Inspection and Enforcement.
inspection policies and procedures.
03.02 Reaional Administrator.
Manages the implementation of the radiation protectTon inspection program.
/
2500/23-04 BACKGROUND 04.01 Introduction.
During certain spent fuel handling operations (such as those described below) very high dose rates can exist in BWR drywells.
All drywell-containment types (Mark I, !!, III) require compensation for a lack of design shie4 ding (see Attachment 1).
Certain fue.1 handling con-trols must be iristituted to prevent acute and potentially f atal exposure to drywell workers from mishaps with irradiated fuel.
to 10s R/hr Unshielded irradiated fuel can create radiation fields of 104 With dose rates at these high levels, the exact dose rate (at 1 foot).
largely is an academic question.
Assuming an LDso of 500 rads, an exposure R/hr field will deliver a f atal time as short as one minute in a 3x104 To date, the Trojan spent fuel transfer tube event (described in IE Bulletin 78 06) and the Indian Point 2 fuel pool-diver event (described in dose.
IE Information Notice 82 31) are the only incidents where spent fuel has been the radiation source for personnel overexposures at NRC-licensed LWRs.
Issue. late:
In response to the Trojan event, the NRC's Standard Review Plan (SRP) was modified to include a detailed evaluation that focuses on th This generic SRP item in-from this overexposure incident are addressed.
cludes PWR and applicable BWR plants (Mark III containments have an a sible fuel transfer tube).
This IN provided event (not involving spent fuel), IE IN 84-61 was issued. s manage underwater work to help prevent further overexposures.
Recognizing the potential radiation hazards in the drywell from spent fuel movements, General Electric issued two generic information letters on this subject in 1973 and 1980.
It would seem then that appropriate NRC and industry generic communications warning of spent fuel hszards have been sent to the licensees. White we can be sure that the NRC notices were received by the licensees, it is unclear whether the licensees received and/or took appropriate actions as a result Licensees were not required until January 1, of the two vendor advisories.
1981 to have a system to ensure that generic operating experience notices are properly reviewed and appropriate followup actions are tak threatening radiation doses, this TI will provide a forma 0737, Ites !.C.5.).
programs to prevent overexposures in the drywell during fuel movements.
The vendor warnings, discussions with vendor per nical descriptions and associated inspector requirements and guidance.
04.02 Description of Oryve11 Hazards.
Spent fuel movements can present four basic exposure conditions as a function of position (see Attachment 2 The tabulated pro-
"BWR Drywell Dose Rates During Spent Fuel Moveme These dose rates can vary as extrapolated from actual field measurements.a function of fuel ass Therefore, these dose rate history, and shielding /drywell configuration.
The data are values should be used as "order of magnitude" appro days) and early 1970's-vintage fuel.
history would be greater for fuel made more recently (given the same powerfto o For example, and decay times).
BWR 6 fuel, multiply the dose rate values by 1.3.
During refueling, a
Lack of Dest _an Shieldina in Fuel-Shute Area.
spent fuel bundle f( transferred from the reactor pressure vessel, up a 04.03 over the RPV flange ~, over the drywell, through the fuel-Shute areas, and For all types of BWR model/ containments, no l
into the fuel storage pool.
water) would protect a worker in the drywell assu Assuming free worker access to the (fuel configuration C in Attachment 2). fuel movements, this shielding deficiency during spent drywell core to-fuel pool "shute" area should be corrected.
As a backfit solution, portable fuel-shute shields (FSS) w keted in the 1970's. lead shielding at the bottom and lesser amounts on the l
ment 3, typical portable F55).
sno/-
Issue Date:
03/13/87
Dui f ng normal fuel transfer, Cithout an FSS, high transitory dose rates Given dose rates for fuel exist in the drywell for short time periods.
configuration 8 (Attachment 2) and an assumed time of 20 seconds to tra-verse the vessel-to-pool span over the drywell, workers a few feet away (at dose rate point 1 at the top of drywell) could receive a maximum dose of However, if the refueling platform / bridge happened to stop in-transit (as a result of power failure or equipment problems), then high 270 mrea.
dose rates would be sustained in the upper levels of the drywell and With an effective FSS in workers would have to promptly exit the area.
place, dose rate increases would probably go unnoticed (few mR/hr).
Unless fuel movement pro-Peripheral Spent Fuel Bundle Hazards.
spent fuel moved to or from the core's outer areas 04.04 cedures are modified, could be within a couple of feet of the RPV inside walls (Attachment 2, When spent fuel travels above the sacrificial bio-fuel configuration A).I W :a1 shielding (585) during fuel passage, a worker above the i
,fd be subject to dose rates increased by 10 R/hr Mn a few feet of
-:e outer reactor pressure vessel (RPV) wall and up to 50 R/hr on contact Assuming a fuel hoist speed of 15 ft/ min during a nor-with the RPV wall.
mal peripheral spent fuel bundle transfer, a worker could receive approxi-mately 250 mrem during the 90 second vertical If f t (or insertion).
Again with a bundle within a few feet of the RPV, dose rates while working at unshielded piping penetrations in the SBS (see fuel con-drywell If the fuel handling operations ceased (power figuration D, Attachment 2). failure), these high dose rates at the penetration w fuel handling procedures to avoid close Plants have typically modifiedthe unshielded RPV; once the spent fuel bundle clear proximity tocore, the fuel handler operator moves a bundle toward the middle of the This increased core, providing a minimum of 4 feet of water shielding. shield tude.
Plants using in vessel, out-of-core fuel sipping (a differential-pressureD method) have additional concerns.
moved to less than I foot of the inner RPV wall, causing sustained dose rate increases up to approximately 100 R/hr in the upper drywell levels.
$s'sultant The fuel-to-sipping container transfer can last 1 to 5 minutes l
l the shielded, SBS areas) can range up to several hundred mR/hr.
l 2500/23 05 INSPECTION REQUIREMENTS As part of the next scheduled inspection covering external 4xpos (either IP 83524 or 83724),
evaluated to ensure that licensees have implemented effective control programs for the drywell area during spent fuel handling movements.
[
Evaluate licensee's written procedures, tr'aining program materials, and organization interfaces using the guidance provided in Section 06.
05.01 05.02 If access to the drywell is allowed during spent fuel movements, review the licensee's written technical evaluation of the fuel shu effectiveness to protect workers against a worst case fuel drop accident. -
I Issue Date:
03/13/87 2500/23 l
O 2500/23-06 INSPECTION GUIDANCE The central focus of this TI is to ensure licensees programmatic controle Workers must be protected from spent fuel hazardt prevent overexposures.in all configurations during various operations described in Sec through 04.04.
Procramatic Control to Protect Workers From Fuel Oroo in Fuel-Shute 06.01 Ares Without an FSS (with approximately 4-5 inch Pb equivalents) in-stalled, the licensee should not allow workers to enter the drywell.
s.
Dose rates 40 feet away from a dropped spent fuel bundle can exceed 3 R/ min without adequate shielding.
With adequate shielding in place in the fuel-shute area, work can be accomplished in the drywell consistent with the recomended essen-b.
tial program elements of Section 06.02 in place.
Procrammatic Controls to Allow Work in Drywell Ourina Fuel Movements 06.02 (F55 Shute Shield in Place)
Fuel handling procedures should ensure that fuel is not moved within approximately 4 feet of the RPV inner wall when raised above (orFu a.
Iowered on reload) the 585.
contain appropriate caution notices clearly stressing' potential Communica.
hazards of fuel handling when the drywell is occupied.
tions between the fuel handlers, control room, and HP drywell con-trol point should be established and periodically tested, b.
Training and retraining (e.g., prior to each refueling outage) programs for drywell workers, health physics, and operations /
refueling workers should be formalized to explain the potential
- hazards, protective work controls / limitations, and evacuation signals and routes, Workers near the unshielded RPV (at or above the 185) in the upper drywell areas need to be quickly warned and evacuated in the event c.
fuel is inadvertently moved close to the vessel wall or if fpel was This should include work ardund SBS dropped in the fuel-shute area.
Workers can be issued self a'1 arming dosimeters to penetrations. alert them to rapidly changing radiation fields. High range, local reading gamma detectors can be installed in the alarming, remote upper drywell with reacte readout / alarm at the HP access control point.
2500/23 07 REPORTING REQUIREMENTS Complete the Technical Summary Form (Attachment 4) for each facility in-spected and send the completed form to ORP8, attention priate generic action can be considered (e.g., issue infor l
2500/23
- sw ' a.i:
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4-
2500/23-08 EXPIRATION This TI will remain in effect untb February 26, 1988.
2500/23-09 It CONTACT Wigginton Questions regarding this TI should be addressed to James E.
(492-4967) or t.. J. Cunningham (492 9664).
2500/23-10 MODULE TRACKING SYSTEM INPtJT in fulfilling the requirements of this instruction should be Time spent charged to 25023.
END Attachments l
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03/13/87 I
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TECHNICAC SUMKARY FORM -
TI 2500/23 EVALUATE BWR LICENSEE RADIOLOGICAL CONTROLS FOR ORYWELL DURING SPENT FUEL MOVEMENT
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CONTAINMENT T)/E:
MARK '
INSPECTOR (5):
INSPECTION DATE(5):
INSPELTION REPORT NO.
QUESTION _$
Did licenses fully understand drywell hazards from apent fuel movements?
{.
If no, briefly explain on the back of this page.
Did licensee have GE information letters (1973,1980)?
2 When did licensee first iniplement affective drywell con *rols (e.g., at first 3.
refueling)?
If answer is no, then stop; 4.
Is drywell access allowed during fuel movements?
)
You are done.
If yes, answer the remaining questions.
Briefly describe on the back of this page shielding for fuel-shute ares (e.g.,
i 3.
how much Pb).
PROCEDURE REVIEW Describe on the back of this page any significant weaknesses in implementing CA procedures for fuel handling and NP drywell control point.
Is over.11 procedur:1 control adequate to protect workers from drywell hazaros?
7 If no, describe on the back of this page licensee commitments to upgrade / correct, af l
8.
Is there specific training to provide all affected workers an adequate under-I standing of the drywell hazards?
Does this training provide clearly defined worker actions, responsib,111 ties i
9.
during normal and soargency operations (e.g., in event of drywell evacuation l
signal)?
s l,2500/23 A4-1 Issue Date:
03/13/87
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