ML20202F527
| ML20202F527 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 11/28/1997 |
| From: | Collins S NRC (Affiliation Not Assigned) |
| To: | Jun Lee AFFILIATION NOT ASSIGNED |
| References | |
| NUDOCS 9712090115 | |
| Download: ML20202F527 (4) | |
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NUCLEAR REGULATORY COMMISSION o
WASHINGTON. D.C. soMH001 November 28, 1997 Ms. Jane Lee 183 Valley Road Etters. PA 17319
Dear Hs. Lee:
This letter responds to the letter you sent to the NRC on November 7. 1997, concerning the exemption from requirements for criticality alarms at the Three Mile Island Nuclear Generating Station. Unit 1 (THI).
It seems that there has been a misunderstanding in that the exemption in question neither eliminates systems to warn the public nor constitutes special treatment for TMI.
Criticality alarms are not designed or intended to warn or alert the off site community.
The relevant regulation.10 CFR 70.24 requires alarm systems to warn plant workers in case of an accidental criticality.
The regulation is important for worker protection in non-reactor facilities that handle substantial quantities of s)ecial nuclear material.
in nuclear nower plants however, the manner in whic1 the special nuclear materials are h b dled and the presence of other monitorin systems, tend to obviate the need for these special alarms.
According y where licensees have shown that an exemption from this recuirement would have no significant impact, exemptions have been granted.
GPL did request an exemption and showed that granting it would have no significant safety impact.
Accordingly, the NRC staff reviewed the request, determined the GPU position was valid, and granted the requested exemption.
A copy of the staff's safety evaluation is enclosed for your information.
In regards to D. C. Kocher's 1980 publication. Dave Kocher has served as a consultant to the NRC and we find no significant difference between his dose rate conversion factors and the NRC values for the same quantities. Of course, there have been some advances in this area since 1980. but the changes are minor.
Certainly, some radiation is released from nuclear power plants, ust as there are radiation doses from the sun and from every human activity, he radiation doses from natural sources range from about 0.15 to about 1.5 rem annually, with the average being about 0.3 rem.
To ensure that nuclear power plant releases do not significantly increase natural doses, releases u a limited so the maximally exposed individual off the reactor site does not receive an annual dose of more than 0.005 rem.
The "recent incident" to which you refer is believed to be the " unusual event" of June 21. 1997. That event was properly reported to the NRC, the licensee actions were monitored by the NRC staff, and the releases were quite small, even when com ared to the NRC 's stringent limits.
I can assure you that no
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J. Lee 2-The NRC takes its responsibility for safety very seriously, and follows the rules and regulations that implement the Atomic Energy Act as well as the National Environmental Policy Act. We hope this information meets your needs.
Sincerely.
/ w a 41 1 4; Samuel J. Collins. Director Office of Nuclear Reactor Regulation
Enclosure:
Safety Evaluation
J. Lee e The NRC takes its responsibility for safety very seriously, and follows the rules and regulations that implement the Atomic Energy A:t as well as the National Environmental Policy Act.
We hope this information meets your needs.
Sincerely.
Original signed by Brian W. Sheron for Samuel J. Collins. Director Office of Nuclear Reactor Regulation
Enclosure:
Safety Evaluation DISTRIBUTION See next page DOCUMENT NAME: G:\\BUCKLEY\\G970812 *SEE PREVIOUS CONCURRENCE To rGceive a copy of this document, indicate in the box:
"C" - Copy without attachment / enclosure "E" = Copy with attachment / enclosure "N" - No copy
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RAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION Rd,&ljNGToREQUESTFOREXEMPTIONFROM10CFR70.24 GPU NUCLEAR CORPORATION THREE MILE ISLAND UNIT 1 DOCKET N0. 50-289
1.0 INTRODUCTION
By submittal dated fegruary 7,1997, as supplemented March 26 and June 5,1997,irements of subsection (a) of 10 CFR 70.24 *(Criticality Acc GPU Nuclear Corporation (the licensee) requested an Exemption 7
from the requ Requirements.)" The staff evaluation of this above cited request is i
delineated below.
2.0 EVALUATION The Code of federal Regulations at subsection (a) of 10 CFR 70.24,
' Criticality Accident Requirements," requires that each licensee authorized to possess special nbelear material shall maintain in each area where such material is handled, used, or stored, a criticality monitoring system 'using gamma-or neutron-sensitive radiation detectors which will energize clearly audible alarm signals if accidental criticality occurs." Subsection (a)(2) of 10 CFR 70.24 specifies the detection, sensitivity, and coverage capabilities l
of the monitors required by 10 CFR 70.24(a) for persons licensed prior to December 6,1974.
The specific requirements of Subsection a)(2) are to
' maintain a monitoring system capable of detecting a critica(lity which i
l generates radiation levels of 300 rems per hour one foot from the source of the radiation." Subsection a)(3) of 10 CFR 70.24 requires that the licensee shall maintain emergency proc (edures for each area in which this licensed 4
special nuclear material is handled used, or stored and provides (1) that the proceduresensurethatallpersonnelwithdrawtoanareaofsafetyuponthe sounding of a criticality monitor alarm, (2 that the procedures must include drills to familiarire personnel with the eva)cuation plan, and (3) that the procedures designate respot.sible individuals for detemining the cause of the alam and placement of radiation survey instruments in accessible locations for use in such an emergency. Subsection licensee who believes tant there is good ca(d) of 10 CFR 70.24 states that any use why he should be granted an Enclosure
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exemption from all or part of 10 CFR 70.24 may apply to the Commission for i
such an exemption and shall specify the reasons for the exemption requested.
The purpose of 10 CFR 70.24(a), (a)(2)d that action is taken to protect
, and (a)(3) is to ensure that any inadvertent criticality is detected an personnel and correct the problem. By letter dated February 7, 1997, as supplemented March 26 1997, and June 5, 1997, the licensee requested an exemption from the req,uirements of 10 CFR 70.24 a).
handle and store unirradiated fuel and other spe(cial nuclear material withoutThe lice having either the criticality monitoring system or the emergency procedures specified in 10 CFR 70.24. The licensee believes that procedures and design i
features make an inadvertent criticality unlikely, in accordance with General Design Criterion 62.
Sacial nuclear material, as nuclear fuel, is stored in the spent fuel pool or tse new (unirradiated) fuel storage racks. The spent fuel pool is used to store irradiated fuel under water after its discharge from the reactor, and new fuel prior to loading into the reactor. The new fuel racks are used to store new fuel in a dry condition upon arrival on site.
Special nuclear material is also present in the form of fissile material incorporated into nuclear instrumentation. The small quantity of special nuclear material present in these items precludes an inadvertent criticality.
Consistent with Technical S>ecifications (TS) Section 5.4.1, the spent fuel pool is designed to store tte fuel in a geometric array that precludes criticality. The spent fuel racks are designed such that the effective neutron multiplication factor, k Will remain less than or equal to 0.95 under all normal and accident co$ltions for fuel of maximum enrichment of 5.0 wtX U-22<.
The staff has found this design adequate.
The new fuel storage racks may be used to receive and store new fuel in a dry condition upon arrival on site and prior to loading in the reactor or spent fuel pool.
The spa:ing between new fuel assemblies in the storage racks is sufficient to maintain the array in a suberitical condition even under accident conditions assuming the presence of moderator. The maximum enrichment of 5.0 wtX U-235 for the new fuel assemblies results in a maximus km of less than 0.95 under conditions of accidental flooding by unborated water, and k less than 0.98 under conditions of low-density optimum moderation.,Yhe staff has found the design of the licensee's new fuel storage racks to be adequate to store fuel enriched to no greater than 5.0 wtX U-235.
Nuclear fuel is moved between the new fuel storage racks, the reactor vessel, and the spent fuel pool to accommodate refueling operations.
In addition, fuel is moved into the facility and within the reactor vessel, or within the-spent fuel pool.
In all cases, fuel movements are procedurally controlled and designed to preclude conditions involving criticality concerns.
Fuel handling procedures and the design features of the fuel handling system are discussed in the licensee's Final Safety Analysis Report.
procedures and controls handling;-nevertheless, prevent an inadvertent criticality during fuel radiation monitoring, as required by General Design Criterion 63, is provided for handling new fuel prior to being placed into the t
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spent fuel pool.
In addition, handling of fuel in the spent fuel pool is monitored by TS-required radiation monitors on the fuel handling bridges.
These required radiation monitors have alarm response procedures which provide instructions to the operators upon receipt of alarus.
The licensee conducts training of fuel handlinfi bridge operators and maintenance personnel.
In addition all indiv< duals who have access to radiologically controlled areas are, required to complete initial training, and annual refresher training thereafter, which includes proper response to area l
radiation monitor alarms.
The purpose of 10 CFR 70.24 is to ensure that if a criticality were to occur during the handling of special nuclear material, personnel would be alerted to that fact and would take appropriate action. The staff has determined that such an accident is highly unlikely to occur; furthermore, the licensee has radiation monitors, as required by General Design Criterion 63, in fuel storage and handling areas. These monitors will alert personnel to excessive radiation levels and allow them to inttinte appropriate safety actions. The low likelyhood of an inadvertent criticality together with the licensee's t
adherence to General Design Criterion 63 and radiation worker training constitute good cause for granting an exemption to the requirements of 10 CFR 70.24(a).
3.0 CONCLUSION
Based upon the information provided, there is reasonable assurance that irradiated and unirradiated fuel will remain subcritical during handling and storage. The circumstances for granting an exemption to 10 CFR 70.24(a) are met because criticality is precluded with the present design configuration, TS requirements, administrative controls, and the fuel handling equipment and procedures. Therefore, the staff concludes that the licensee's request for an exemption from the requirements of 10 CFR 70.24(a) is acceptable and should be granted.
Principal Contributor:
L. Kopp Dated: July 3,1997 t
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