CY-97-121, Provides Addl Info W/Respect to Proposed Defueled Emergency Plan for Plant.Revised Pages for Proposed Defueled Emergency Plan Encl
| ML20197B685 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 12/18/1997 |
| From: | Mellor R CONNECTICUT YANKEE ATOMIC POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20197B689 | List: |
| References | |
| CY-97-121, NUDOCS 9712240022 | |
| Download: ML20197B685 (12) | |
Text
C CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK PLANT 362 INJUN HOLLOW ROAD e EAST HAMPTON, CT 06424-3099 December 18,1997 Docket No. 50-213 CY-97-121 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Haddam Neck Plant AdditionalInformation For The Proposed Defueled Fmeraency Plan The purpose of this letter is for Connect l cut Yankee Atomic Power Company (CYAPCO) to provide the NRC with additional information with respect to the proposed Defueled Emergency Plan for the Haddam Neck Plant (HNP).
Completlen Of Spent Fuel Movement in a letter dated September 26,1997,W CYAPCO provided the NRC with assessments that determined that once the fuel is stored in a configuration consistent with the loss of spent fuel poo, vater analysis, there are no significant off-site consequences to a postulated beyond design basis event of a loss of all spent fuel pool water after October 1,1997.
In a letter dated October 7,1997,m CYAPCO discussed the processes that would be implemented prior to and during the movement of the spent nuclear fuel into a configuration consistent with the analysis. On October 23,1997, this spent nuclear fuel configuration change was completed. Therefore, at this time, there are no significant off-site consequences to a postulated beyond design basis event of a loss of all spent fuel pool water, fo*
NR AO O o 13
}
F PDR (1)
CYAPCO Letter CY-97-066, from T. C. Feigenbaum, to the U. S. Nuclear Regulatory Commission, " Assessment Of The Beyond Design Basis Event Loss Of All Water in The Spent Fuel Pool," dated September 2E.1997.
(2)
CYAPCO Letter CY-97-110, from T. C. Feigenbaum, to the U. S. Nuclear Regulatory Commission, " Movement Of Fuel in The Spent Fuel Pool," dated
' '"' 7 ' ** 7'
.l!I.Ll.l!.!il. il.l!..l
U. S. Nucle:r Regul: tory Commission CY-97_-121/Page 2 ScatteLDas_e_Aasnamant in order to bound what the dose consequences would be for a postulated accident onsite, a non-mechanistic, beyond design basis event of a loss of all pool water was considered. The loss of all spent fuel pool water is regarded as a non-mechanistic, beyond design basis event since there is sufficient time to counteract small leaks (via piping or fire hoses, using electric or diesel pumps with water from tanks or the Connecticut River). Also, the spent fuel poolis designed not to leak even due to a Safe Shutdown Earthquake (SSE) postulated for the Haddam Neck site. A catastrophic loss of all water would require an earthquake well beyond the SSE.
The dose consequences were determined by performing a scatter dose rate analysis
- due to gamma [y] radiation stemming from a complete loss of shielding that would be provided by the spent fuel pool water.
The scatter dose analysis assumed that there was no water in the spent fuel pool. The analysis also assumed that other than the Spent Fuel Building, no other structures at the site provided any shielding. This assumption is conservative. Finally, as shown in Table 1 of CYAPCO's letter dated May 30,1997,W there are 1019 HNP spent fuel assemblies stored on site with another 83 HNP spent fuel assemblies stored off-site.
While it is not expected that the 83 assemblies will be returned to the HNP site, the scatter doae assessment looked at the scatter dose both for one spent fuel assembly and for 1102 (i.e.,1019 + 83) spent fuel assemblies.
Provided in Table 1 are the resuhs of both the analysis and the Monte Carlo check calculation (the analysis has been OA reviewed while the Monte Carlo check calculation has not). The analysis is conservative while the Monte Carlo methodology provides a more exact representation of the results, in both cases, the results indicate that the hourly dose at the Exclusion Area Boundary is a fraction of the 10CFR100 dose limits and less than the EPA Protective Action Guides (PAGs). Hence, even in postulating this non mechanistic beyond design basis event, and using the mos' conservative dose assessment, it would take more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to reach the EPA PAG limit (using the Monte Carlo results, it would take over 11 days), in addition, the analysis result for the house nearest the plant (directly adjacent to CYAPCO property) indicates that it would take more than 2 '/ days to reach the EPA PAG limit.
2 (3)
Radiological Assessment Branch Calculation," Scatter Dose Rates Due To CY Spent Fuel Pool Draindown," Revision 0, dated December 8,1997.
(4)
CYAPCO Letter CY-97-006, from T C. Feigenbaum, to the U. S. Nuclear Regulatory Commission, " Proposed Revision To Operating License And Technical Specifications Defueled-Ooerating License And Technical Specifications," dated May 30,1997.
U. f iluclear RegulClory Commission CY-121/Page 3 Therefore, the immediate response required by the current Emergency Plan is no longer necessary Nevertheless,in the proposed Defueled Emergency Plan, CYAPCO plans to maintain communication channels with the State of Connecticut and local emergency service organizations to apprise them of onsite conditions in the event of an emergency being declared at the Haddam Neck site.
Zitcaloy-Clad FuelDose ASSeAnmeAt in SECY-96-256* and in SECY-97-120,* the NRC discusses the consequences of a non-mechanistic, beyond design basis event of a loss of all the water in a spent fuel pool. The concern is that if, for Zircaloy-Clad fuel assemblies, the clad temperature exceeds 565* C, the fuel assemblies could start to experience clad failure which would lead to the release radioactive gas. There is no similar concern for the Stainless Stecl-Clad fuel assemblies.
In Attachment 5 of the May 30,1997 letter, CYAPCO also provided a calculation
- which discussed the dose consequences of both a cask drop, which damaged 157 fuel assemblies, and a fuel assembly drop, which damaged one fuel assembly. The results of that calculation are provided in Table 2 of this letter.
As discussed in Table 1 of the May 30,1997 letter, there are only 161 Zircaloy-Clad fuel assemblies in the spent fuel pool (the remaining fuel assemblies are Stainless Steel-Clad). Since the dose result for more than one assembly is a multiple of the dose from one assembly, Table 2 of this letter also provides the dose for 161 fuel assemblies (i.e., the cask drop plus four single fuel assemblies).
The results of the release doses, which are provided in Table 2 of this letter indicate that, even if all the Zircaloy-Clad fuel assemblies in the spent fuel pool were to fail, the dose at the Exclusion Area Boundary is a fraction of the 10CFR100 dose limits and less than the EPA Protective Action Go, des (PAGs).
(5)
SECY-96-256, " Changes To The Financial Protection Requirements For Permanently Shutdown Nuclear Power Reactors, 10CFR50.54(w) And L'CFR140.11," dated December 17,1996.
(6)
SECY-97-120, "Rulemaking Plan For Emergency Planning Requirements Fcr Permanently Shutdown Nuclear Power Plant Sites 10 CFR Part 50.54(q) And (t);
10 CFR 50.47; And Appendix E To 10 CFR Part 50," dated June 16,1997.
(7)
Radiological Assessment Branch Calculation XX-XXX-60RA, " Radiological Assessment Of A Spent Fuel Shipping Cask Drop in The CY Spent Fuel Pool,"
Revision 1, dated March 13,1997.
l l
I U. S. Nucle:r Regul: tory Commission CY-97-121/Page 4 Spent Resin Handling in Attachment 5 of. the May 30,1997 letter, CYAPCO also provided a calculation
- which discussed the dose consequences of a spent resin accident. The results of that-analysis, at the Site Boundary, are:
o Committed Effective Dose Equivalent 4.33E-01 rem 3
(DBA Analysis) 9tal Effective Dose Equivalent 9.60E-01 rem EP Analysis)
These results indicate that the doses at the Exclusion Area Boundary are a fraction of the 10CFR100 dose limits and less than the EPA Protective Action Guides (PAGs). A description of the conservatism of this analysis is provided in Table 3.
The process of handling and dewatering resins is proceduralized and personnel who handle resins are trained and experienced. During the dewatering process, the resin temperature is monitored. If the temperature exceeds either 130 F or increases at a rate greater than 25 F/ hour, the dewatering process is stopped and the resin is flooded with water. CYAPCO has successfully processed over 35 resin liners usirg essentially these procedures without experiencing an exothermic reaction.
The Radiation Protection Manual (RPM) procedures that are used in the handling of spent resins, are as follows:
RPM 3.4-1 Receipt, inspection, Set-Up And Handling Of
- High Integrity Containers (HIC) e RPM 3.4-2 Set-Up Of High Integrity Container For Resin Slurry e
RPM 3.4-4 Dewatering Of High Integrity Containers in The Spent Resin Facility e
RPM 3,4-5 Use Of Remote Lifting Device (RLD) e RPM 3.4-6 Spent Resin System Operation RPM 3.4-8 Resin Slurry To Spent Resin System These procedures are available for NRC Staff review.
(8)
Radiological Assessment Branch Calculation CYRESIN-01578-RY," Radiological Consequences From A Resin Accident," Revision 0, dated May 30,1997.
U. S. Nucle r Regul: tory Commission CY-97-121/Page 5 Emergency Plan Training A Medical Drill was conducted on April 16,1997 which demonstrated the following; The ability of on-shift personnelincluding the nurse, the Security Shift Supervisor o
(SSS) and Health Physics (HP) to respond to an injured person and prepare the patient for transfer to the hospital; The ability cithe Operations Shift Manager (OSM) to classify the event and to o
perform direct notifications; The ability of the Shift Manager's Staff Assistant (SMSA) to request offsite assistance, notify the hospital, and make notifications to the state; The ability of Security to resnond to the accident, provide communications, c! ear the ambulance for arrival and departure, and escort the ambulance and Emergency Medical Technicians (EMTs) to the location for pickup of the patient; The ability of the Ambulance Service to respond to the event, wear dosimetry o
and Protective Clothing (PCs) as necessary, obtain information on the patient's condition, prepare the patient for transfer, and brief the hospital on patient's condition via ambulance radio; and The ability of HP Tech to survey the patient, control contamination, provide e
radiological status to the control room, station nurse and ambulance personnel, assist in radiological control at the hospital, collect and maintain cont 31 of contaminated materials for decontamination and release or disposal, an'. keep the OSM informed of the patient's status.
In addition, an HP Drill was conducted as part of this Medical Drill.
Additional SERO (Station Emergency Response Organization) training drills were conducted in March and April of 1997.
Participants were the OSM, SMSA, Assistant Director of Technical Support (ADTS), and the Assistant Director of the EOF (ADEOF).
Each crew was required to perform two tabletop scenarios.
Areas of concentration included emergency classification, onsite and offsite protective action decision making, and offsite notifications. Adjuncts of the drills included plans for corrective actions (plant repairs), activation of the EOF and TSC, and coordination of the event. In addition, three crews were required to perform two tabletop scenarios for the NRC during the week of April 21,1997. Each crew responded to two scenarios and were evaluated on the criteria stated above. Training was determined to be effective.
U. S. Nuclear Regul: tory Commission CY-97-121/Page 6 There has been limited training for key SERO positions (i.e., Director of Station Emergency Operations, Manager of Control Room Operations, Assistant Director Technical Support and Assistant Director Emergency Operations Facility) in 1997 with respect to the current Emergency Plan.
The DERO (Defueled Emergency Response Organization) initial training has been conducted.
Additionally, many of the individuals who have been trained in the proposed Defueled Emergency Plan and implementing Procedures are members of the SERO. Several DERO tabletop training drills have also been conducted.
Therefore, based on the above and in anticipation of the approval of the proposed Defueled Emergency Plan, no SERO training is envisioned for 1998. However, if the Defueled Emergency Plan is not approved by the NRC for implementation on January 1,1998, CYAPCO is requesting a 3-month extension in conducting SERO training for the existing Emergency Plan in 1998.
NRC Conference Call On December 11, 1997 a conference call was held between the NRC Staff and CYAPCO. During this call several enhancements to the proposed Defueled Emergency Plan were suggested. Attachment 1 to thi.= 'etter provides the revised pages to the proposed Defueled Emergency Plan,C '" which reflect the proposed enhancements suggested by both the NRC Staff and CYAPCO that have been developed since the original submittal.
A request was also made to revise the exemption request for 10CFR50.47 and 10CFR50, Appendix E that were submitted in the October 21,1997 letter.
These changes are provided in Attachment 2 and do not change the basis of the previously submitted exemption request.
(9)
CYAPCO Letter CY-97-047, from T. C. Feigenbaum, to the U. S. Nuclear Regulatory Commission,"Defueled Emergency Plan And Request For Exemption From 10CFR50.54(q) For Offsite Response," dated May 30,1997.
(10)
CYAPCO Letter CY-97-103, from T. C. Feigenbaum, to the U. S. Nuclear Regulatory Commission, " Submittal Of The Emergency Action Levels For The Defueled Emergency Plan," dated September 19,1997.
(11)
CYAPCO Letter CY-97-109, from T. C. Feigenbaum, to the U. S. Nuclear Regulatory Commission, " Additional Information On The Proposed Defueled Emergency Plan And The Request For Exemption From 10CFR50.54(q)," dated October 21,1997.
U. S. Nucle:r_Regul tory Commission CY 97-121/Dage 7
- Summary 4
As indicated above, the radiological consequences of a ' non mechanistic, beyond design basis loss of spent fuel pool waterf releases from fuel assemblies and releases from spent resin at the Exclusion Area Boundary are a fraction of the 10CFR100 dose limits and less than the EPA Protective Action Guides (PAGs). Since it would take more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to reach the EPA PAG limit for a loss of all spent fuel pool water
- (using the Monte Carlo results, it would take over 11 days), and the analysis result for the house nearest the plant (directly adjacent to CYAPCO propertv) indicates toat it would take more than 2 '/ days to reach the EPA PAG limit, the immediate response 2
required by the current Emergency Plan is no longer necessary.
A significant consequence of maintaining the current Emergency Plan for a full power 4
operating plant is a need to maintain offsite liability insurance and onsite property insurance at levels equivalent to those for a full power operating plant. Maintaining the current Emergency Plan and paying insurance premiums for an operating plant is an inappropriate use of limited resources and operating funds.
Presently, CYAPCO. expends approximately an additional $ 78,000/ month to maintain the Offsite Emergency Plan for a full power operating plant and approximately
$ 253,000/ month in additional premiums to pay for the financial protection required of a full power operating plant. This large expenditure of funds (over $ 330,000/ month) for plant conditions that no longer exist, adversely affects the decommissioning process for CYAPCO and/or its ratepayers.
Therefore, CYAPCO respectfully requests the NRC e expedite its revh..v of the proposed Defueled Emergency Plan and the exemption request for a reduction in the
+
financial protection requirements.02)
Commitments
- There are no commitments in this letter. Statements within this letter are provided for information only.
(12)
CYAPCO Letter CY-97-065, from T. C. Feigenbaum, to the U. S. Nuclear Regulatory Commission, " Request For Exemptions From The Financial Protection Requirement Limits Of 10CFR50.54(w) And 10CFR140.11," dated October 7,1997.
I U. S. Nucle r Regulatory Commission CY-97-121/Page 8 If the NRC should have any questions,- please-contact Mr. G. P. van Noordennen at (860) 267-3938.
' Very truly yours,.-
CONNECTICUT YANKEE ATOMIC POWER COMPANY
\\
C
~
c>
_.K %
R. A. MelIo' r.
T Vice President-_- Operations and Decommissioning Attachments-cc:
H. J. Miller, NRC Region I Administrator M. B. Fairtile, NRC Senior Project Manager, Haddam Neck Plant W. J. Raymond, NRC Senior Resident inspector, Haddam Neck Plant D. Galloway, Acting Director, CT DEP Monitoring and Radiation Division
U. S. Nucle:r Regulat:ry Commission
- CY 97-121/Page 9 Ithle_1 Scatter Dose (mrem /hr)
Results Outside The EOF )
Nearest House (2)
U From One From 1102 From One From 1102 -
Assembly Assemblies (3)
Assembly Assemblies (3)
Analysis 6.49E-02 7.15E+01 1.51 E-02 1.64E+01
")
M)
- Monte Carlo-3.38E-03 3.72E+00 Check Calculation The EPA PAG limits are:
Whole Body 1 rem e
Thyroid 5 rem o
o Skin 50 rem Nolen:(1)
Exclusion Area Boundary (2)
Directly adjacent to CYAPCO property (3) 83 assemblies stored offsite included in the total of 1102 assemblies (4)
The Monte Carlo Check Calculation did not did not extend its computations to the nearest house
U. S. Nucle:r Regul: tory Commission CY-97-121/Page 10 Iable.2 Release Dose From Fuel Assemblies At The Exclusion Area BoundaIy (rem)
Whole Body Thyroid Thyroid Skin (Assuming DF=100)
(Assuming DF=1) 1 Assembly 2.60E-03 2.28E-04 2.28E-02 2.16E-01 (Fuel Drop) 157 Assemblies 4.08E-01 3.58E-02 3.58E+00 3.40E+01 (Cask Drop) 161 Assemblies 4.18E-01 3.67E-02 3.67E+00 3.49E+01 (Zircaloy-Clad)
Whole Body 1 rem Thyroid 5 rem o
Skin 50 rem
- U S. Nuclear Regulatory Commission CY-97-121/Page 11 Table 3 Qualitative Conservatism in Calculation CYRESIN-01578-RY QIiginal Calculation The normal CY 10CFR61 resin analysis was compared with observed changes in the normal radionuclide mix at Yankee Rowe and conservatively assumed 24 r244 increased transuranic loading of the resin. The content of Pum2', Cm and Am were projected to be 2.9,4.4 and 5.9 times the " normal" 10CFR61 2
resin concentrations respectively. These are the radionuclides that are the major contributors to the dose result.
The resin activity loading was maximized to the point where the resin would remain suitable for burial in a solidified configuration.
Decontamination Scope it was assumed that a full system decontamination would be performed resulting e
in a total resin source term of approximately 8000 Curies and would result in transport of approximately 20 liners near the maximum loading (488 Cillinert Current plans call for a partial system decontamination to be performed for a total of approximately 1200 Curies and that approximately 12 liners will be shipped in a dewatered state with maximum loading of approximately 100 Curies each. The resin loading assumed in the safoty evaluation (488 Ci/ liner) was approximately fiva times this value.
Process Controls -
During the decontamination process the resin is monitored for radioactivity to ensure that the radioactivity loading assumed in the calculation is not exceeded.
The resin is dewatered in a heavily shielded cask with monitoring that would indicate the existence of an exothermic reaction.
In the event the temperature exceeds either 130 F or increases at a rate greater than 25' F/ hour, indicatir.g an exothermic reaction, water will be reintroduced to the resin, e,
Once dewatered, the resin is transferred in its liner to a concrete vault. The vault serves as a radiation shield and isolates the resin liner from external sources of damage or ignition.
U. S. Nucle r Regul: tory Commission CY-97-121/Page 12 Iahlal 3
Qualitative Consstyatism in Calculation CYRESIN-01578-RY (continued)
Assumed Release In A Plume There is no industry history of a significant radiological release from a fire in dewatered resin.
The industry has a great deal of experience with handling and dewatering resins.
e In the few instances, where an exothermic reaction has occurred, temperatures have been slightly above the boiling point of water.
CYAPCO has successfully processed over 35 resin liners using the existing process without exothermic reaction. Installed plant systems are planned to be used in the system decontamination at the Haddam Neck site.
Contrary to this experience and as a measure of conservatism, the hypothetical release assumes that the fire could occur and has sufficient energy to release 1% of the resin radionuclide content into the plume. Various references list release fractions from 0.001% to 0.5% Thus the assumed release fraction is conservative.
Offsite Dose Calculation e
In the hypothetical release, the 95* percentile X/Q (i.e. worst case meteorological conditions) was assumed in the offsite dose assessment, No additional credits were taken for the vault containing the resin or the lack of e
an ignition source.
The nearest resident to the site is beyond the site boundary and further than 500 meters from the source. The calculation provided dose assessments at the site boundary.
The dose postulated at the site boundary from the source of the fire is less than e
the EPA Protective Action Guides.
l Conclusion The assumptions in the series of calculations that hypothetically result in a dose e
of 960 mrem at the site boundary are very conservative.