IR 05000309/1983008
| ML20024A858 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 06/13/1983 |
| From: | Gallo R, Eugene Kelly, Lazarus W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20024A849 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.2, TASK-TM 50-309-83-08, 50-309-83-8, NUDOCS 8307010047 | |
| Download: ML20024A858 (7) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-309/83-08 Docket No.
50-309 License No.
DPR-36 Priority Category C
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Licensee: Maine Yankee Nuclear Power Station Inspection At: Wiscasset, Maine and Framingham, Massachusetts Inspection Conducted: May 3-5, 1983
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[ Mzart(1/ Project Engineer Inspectors:
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' 4 ate E. Kelly,ReactyEngineer dite Approved by:
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[6 R. M. Gallo, Chief, Reactor Projects Section
'dat6 1A, Division of Project and Resident Programs Inspection Summary:
Inspection on May 3-5, 1983 (Inspection Report No. 50-309/
83-08)
Areas Inspected: Special safety inspection by two region based inspectors (30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />) of licensee actions taken to comply with requirements discussed in NUREG-0737, Item II.B.2, Design Review of Plant Shielding.
Results: No violations were identified.
- 8307010047 830617
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PDR ADOCK 05000309 Q
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DETAILS 1.
Persons Contacted Yankee Atomic Electric Company
- R. Groce, Senior Licensing Engineer A Hodgdon, Radiological Engineer o
J. McCumber, Senior Systems Engineer
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- R. Turcotte, Systems Engineer Maine Yankee Atomic Power Company
- J. Brinkler, Technical Support Department Head
- J. Garrity, Senior Director of Nuclear Engineering and Licensing i
- E. Wood, Plant Manager i
- denotes those present at exit interview on May 5, 1983.
2.
Plant Shielding Design Review a.
Background and Scope As discussed in Item II.B.2 of NUREG-0737, " Clarification of TMI Action Plan Requirements," each power reactor licensee was required to perform a radiation and shielding design review of spaces around systems that may, as a result of an accident, contain highly radio-active materials. The design review was intended to identify the location of vital areas * and equipment in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by the radiation fields during post-accident operations of these systems.
Additionally, each licensee was required to provide for adequate access to vital areas and prctection of safety equipment for design changes, increased permanent or temporary shielding, or post-accident procedural controls. The design review was to determine which types of corrective actions were needed for vital areas throughout the facility.
These requirements were discussed in Item 2.1.6.b of NUREG-0578,
"TMI-2 Lessons Learned Task Force Status Report and Short-Term Requirements"; were issued by NRC letters dated September 13 and October 30, 1979 to all operating nuclear power plants; and were incorporated into NUREG-0660, "TMI-2 Action Plan." Significant changes in requirements or guidance were described in an NRC letter to all licens6es of operating plants dated September 5, 1980, and were subsequently described in Item II.B.2 of NUREG-0737.
Lastly,
- Any area which will or may require occupancy. to permit an operator to aid in the mitigation of or recovery from an accident is designated as a vital area.
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l an NRC letter to all licensees of operating power reactors dated March 17, 1982 (Generic Letter No. 82-05) requested reconfirmation of the schedule for completing Item II.B.2 of NUREG-0737.
With respect to operating power reactor licensees, the October 30, 1979 NRC letter indic.ated that licensees' plant shielding design reviews were among those items for which post-implementation NRC review is acceptable. Although prior NRC approval was not required, licensees were to document their methods of implementation by the required completion date, e.g., design review by January 1, 1980 and plant modifications January 1,1981.
With respect to documentation specified by NUREG-0737 for vital area access, operatirg license applicants were to provide to the NRC a summary of the shielding design review, a description of the results of this review, and a description of the modifications made or to be made to implement the results of the review. The submittals were to include:
(1) Specification of source terms used in the evaluation, including time after shutdowr. that was assumed for source terms in systems; (2) Specification of systems assumed in the analysis to contain high levels of radioactivity in a post-accident situation; (3) Specification of areas where access is considered necessary for vital system operation after an accident; and, (4) The projected doses to individuals for necessary occupancy times in vital areas and a dose rate map for potentially occupied areas.
NUREG-0737 did not state that licensees of operating reactors were to submit the above documentation to the NRC.
Rather, they were to have available for review the final design details of the implementa-tion * of the Item II.B.2 position and clarification.
(Information equivalent to that submitted by operating license applicants is expected to be available for review as documentation of the design review that provided the bases for final design details.) If devia-tions to that position and clarification were necessary, licensees were to provide detailed explanation and justification for the deviation by January 1, 1981.
The licensee's plant shielding design review and corrective actions were reviewed during this inspection.
This review included licensee submittals to the NRC, a sampling verification of the shielding
- In. addition to providing clarificstion of requirements, NUREG-0737 revised the completion date for modification resulting from the plant shielding design review to January 1, 198.
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design review methodology and representative calculations, a review of selected emergency procedures to determine if the vital areas where personnel must gu are safely accessible, and a review of cor-rective actions taken or planned by the licensee (including plant modifications).
b.
Licensee Submittals to the NRC and Previous Staff Evaluation In the case of Maine Yankee Nuclear Power Station (Maine Yankee),
the shielding design review and planned corrective actions were discussed by Maine Yankee Atomic Power Company (MYAP Co. or the licensee) in a letter to the NRC dated March 5, 1980. The licensee's shielding study and planned actions were evaluated by the NRC staff to meet the Category "A" Lessons Learned requirements for this item (NUREG-0578, Item 2.1.6.b) as discussed in the NRC letter to the licensee dated April 29, 1980. The licensee subsequently submitted a revised shielding design review dated November 6, 1980, which super-seded the original submittal and included details of proposed design changes. A subsequent letter dated May 10, 1982 reported that the modifications described in their November 6, 1980 letter had been completed except for installation of reach rods on two valves in the Residual Heat Removal System. A submittal dated November 30, 1982 indicated that this last modification would be completed by December 31, 1987.. A letter to the NRC dated March 30, 1983 described a proposed modification to install additional shielding at the contain-ment purge valve room entrance which had not been addressed in previots submittd s.
Curing this inspection the licensee indicated that ern rs had been identified in the accuracy of their previous submittals dated Nove.nber 6,1980 and May 10, 1982, in that modifi-cations had not becn completed at described. The inspectors were provided with portions of a draft letter being prepared to correct those inaccuracies. This letter was subsequently submitted to the NRC dated May 18, 1983.
The above licensee and NRC letters were reviewed during this inspec-tion to determine the licensee actions completed or to be taken and the extent of previous staff evaluation, regarding the plant shield-ing design review for Maine Yankee. The licensee's statements and commitments in these letters provided the bases, in part for inspector's verification that plant modifications have been adequately identified and implemented (or scheduled for implementation), to allow access to required vital areas, as discussed in this report.
c.
Shielding Design Calculations and Dose Estimates The inspector reviewed details of the licensee's shielding calcula-tions with various licensee representatives. These details included the mathematical models, assumptions, source terms, dose rates and doses to personnel encountered as part of post-accident access to vital areas. The licensee employed four in-house computer codes to calculate the doses: DIDOS-III, SKYSHINE, RASCAL, and ORIGEN.
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DIDOS-III is a three dimensional point-kernel shielding code used to determine direct gamma radiation from a cylindrical, geometrical source. This code was used to calculate doses from radioactive fluid-carrying pipes.
l SKYSHINE estimates indirect (scattered) radiation from atmospheric reflections. The mathematical model incorporates a modified single-scatter approach that includes absorption by, and buildup within, the medium. The code was used to calculate "skyshine" (air scattered)
doses at the facility.
The RASCAL computer program calculates gaseous source term activity and estimates radiation dose rate levels and integrated doses (cloud immersion) in various areas of the plant. This program was used to estimate the immersion dose rates and integrated doses to personnel within the control room and other vital areas.
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l ORIGEN models production, decay and release of fission and activation products in the core, and calculates source terms at the end of a typical fuel cycle. The results are then appropriate, after adjust-i ment for release fractions and dilution volumes required by NUREG-0737
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Item II.B.2, as input for the other three computer codes utilized in the shielding calculations. No removal mechanisms, such as sprays, plateout or natural deposition were assumed.
The inspector reviewed the licensee's benchmark documentation which compared the results obtained using the above computer codes with results from standard accepted industry codes and practices.
These documents indicated that the licensee's in-house computer codes employ mathematical models that are consistent with recognized methods for shielding calculations. The inspector compared the results obtained from the licensee's calculations with those developed by NRC consultants for a similar plant configuration, as well as with his own independent calculational methods, and the dose rates were consistent.
The inspector also reviewed licensee documents which demonstrated how doses to personnel during post-accident access to vital areas were maintained within the guidelines of GDC-19 and NUREG-0737, Item II.B.2.
No discrepancies were identified.
d.
Corrective Actions The inspector reviewed the licensee's assessment of vital areas and shielding study methodology as described in the licensee's letters to the NRC dated November 6, 1980, March 30, 1983 and May 18, 1983 and internal memoranda dated November 8, 1979, February 7, 1980, May 22, 1980,- Janua ry 27, 1981, February 13, 1981, March 13, 1981 and February 14, 1983.
Based on the results of the licensee's vital
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area assessment, a listing of 8 operations which required access to several different plant areas. This assessment led to the following design changes / procedure revisions:
--Install new bypass piping with valves and reach rods for long term core cooling, hot leg injection alignment (EDCR 80-37)
--Relocation of Hydrogen monitoring / containment air grab sampling equipment to the lower PAB penetration area in lieu of additional shielding.
(EDCR 80-06)
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--Installation of reach rods and additional sheilding in the steam generator auxiliary feedwater pump room for Hydrogen purge or Hydrogen recombiner operation.
(EDCR 80-44).
The inspector revie', red the above Engineering Design Change Requests l
(EDCR), and verified that they had been completed with the exception of the installation of the containment purge valve room shield door and a reach rod on a containment purge valve for post-LOCA Hydrogen control. The licensee has committed to complete this EDCR by May 20, 1983. This installation will be reviewed following completion (309/83-08-01). With the completion of this modification, the licen-see's corrective acticns to assure access to vital areas are acceptable.
e.
Vital Area Accessibility - Procedure Review The inspector reviewed the emergency procedures that would be imple-mented by the licensee in the event of a loss of coolant acciaent.
The review inclued (1) a plant walkdown of the procedures to deter-mine the ability to perform the procedural steps and the accessi-bility of manual valves or breakers that may' require local operations,
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and (2) an assessment of potential exposures to plant personnel based j
on the results of the licensee's shielding design review. The procedrues reviewed included Emergency Operating Procedures (EOP)
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2-70-0, " Emergency Shutdown from Power", Rev. 10, 2-70-1, " Safety
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Injection" Rev. 5, 2-70-2, " Loss of Coolant Accident," Rev. 12, and
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A0P 2-17, " Post Accident Hydrogen Control," Rev. 2.
Based on this review the inspector determined that (1) the procedures could be performed from the vital areas identified by the licensee's shielding design review, (2) the procedures contained appropriate provisions to assure controlled access to vital areas for post-accident operations, and (3) post-accident doses to plant personnel would be within the guidelines of NUREG-0737.
The inspector had no further questions regarding the accessibility of vital areas associated with these procedures.
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3.
Exit Interview
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The inspectors met with licensee representatives on May 5, 1983 (see detail 1 for attendees) to discuss _the scope and findings of this inspec-tion as detailed in this report.
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