ML17254A687
| ML17254A687 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 12/16/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17254A686 | List: |
| References | |
| NUDOCS 8512230235 | |
| Download: ML17254A687 (25) | |
Text
gp,fl RPQQ (4
+4 po A.0 DQ Wp*~4 UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0.12 TO FACILITY OPERATING LICENSE NO.
DPR-18 ROCHESTER GAS AND ELECTRIC CORPORATION R.
E.
GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244
1.0 INTRODUCTION
By letter dated February 27, 1985 as supplemented on June 10, June 26 and July ll, 1985, Rochester Gas and Electric Corporation (RGSE) requested a
Technical Specification change to allow the storage of consolidated fuel in the spent fuel pool at Ginna Nuclear Power Plant.
The proposed storage of consolidated spent fuel involves placing spent fuel containing, at most, all the rods from two standard spent fuel assemblies into one canister.
- Also, a
new Technical Specification,
- 3. 11.5, is to be established allowing the movement of canisters containing consolidated fuel rods over stored spent fuel which has decayed for at least 60 days since reactor shutdown.
2.0 DISCUSSION AND EVALUATION
- 2. 1 Criticalit Considerations The consolidation process could theoretically double the capacity of the spent fuel storage facility, however, due to heat removal limitations the maximum total storage planned is 1,253 fuel assemblies by the year 2014.
Consolidation in the context of this application means that the fuel rods of two fuel
. assemblies (358 rods) are stored in a stainless steel canister (in a packed triangular array) capable of being stored in a storage cell.
The canister will accept only undamaged fuel rods.
Bowed, broken or otherwise failed fuel rods will be stored first in a stainless steel tube of.75 inch outer diameter.
Each canister will accommodate 110 such tubes (Ref. 1).
The design of the Ginna fuel storage racks has been described in an earlier submittal (Ref. 2) and it consists of two regions.
Region I is for the storage of unirradiated or low burnup fuel assemblies and Region II is for assemblies which satisfy certain minimum burnup criteria.
Region II has been designed as a high density configuration.
Rochester Gas and Electric does not have nor do they conte'mplate to use fuel assemblies with burnable poisons,
- hence, the fresh assemblies which are the most reactive are to be stored in Region I (Ref. 3).
851223023'512ib PDR ADOCK 05000244 P
The previous criticality analyses for the Ginna spent fuel pool utilized the LEOPARD (Ref. 4),
PD(07 (Ref.
- 5) and CINDER (Ref.
- 6) codes.
The Boraflex absorbers on the racks were treated with blackness theory.
A similar metho-dology has been employed for the analysis of the consolidated fuel storage.
The validation was based on experiments performed by Babcock and Wilcox Company for this purpose (Ref. 7).
The tightly packed fuel rods present two modeling difficulties, namely the high metal/water and the related dis-tortion of the neutron spectrum (compared to the Wigner Wilkins model) and the "rubber bond" method to define the amount of water to be included in the fuel region.
This method provides a consistent definition of the fuel region for the analyses of the critical experiments as well.
The critical experiments analyzed constituted an adequate test of the component-codes of the model.
The results of the analyses were compared to the experiment as well as to KENO-IV calculations from Reference 7.
The comparison indicated that the LEOPARD-PD(07-CINDER model results in a closer prediction than KENO-IV, a previously approved model.
Therefore, the LEOPARD-PD(07-CINDER model is considered validated and acceptable.
The analytical model described above was supplemented with burnup capability in order to be used for spent fuel.
The burnup code was part of the spent fuel pool modification resubmittal (Ref. 2) and has been approved.
An allowance was made for geometrical manufacturing and thermal deviations.
The licensee took advantage of existing calculations for 3. 13 w/o U-235 Exxon fuel to support conservative criticality estimates for the West Valley New York fuel assemblies which had an initial enrichment of 2.8 w/o U-235.
The calculation yields K
vs burnup which is everywhere less than 0.8.
This is
'true for either 358 rod/canister (i.e.,
2 x 179 where 179 is the number of rods/assembly) or a reduced loading of 350 rods/canister (2 x 175).
For the fuel rods stored in West Valley the average burnup is about 15,000 YiWD/NTU and the K for consolidated fuel is 0.63.
In view of the estimated low values for K it was not deemed necessary to calculate the specific uncer-tainties, therefore the corresponding estimates of the fuel pool storage capacity increase of April 2, 1984, were adopted (Ref. 2).
The total reactivity adjustment to the calculated value for the consolidated rods is
.056 zK.
The above results do not include the effect of the 2,000 ppm of dissolved boron in the pool water.
The results of these estimates assure that for fuel designs satisfying the enrichment-burnup criteria of Figure 5.4-2 of the Ginna Technical Specifications, the K f criteria of 0.95 is satisfied and, therefore, the proposed consolidatel fuel storage is acceptable.
2.1. 2 Accident Anal si s The accident analysis performed in the submittals of April 2, 1984 (Ref. 2) and February 23, 1983 (Ref. 8) are also applicable for the case of the con-solidated fuel storage with respect to criticality.
These submittals analyzed configurations that are analogous to those found in a fully packed canister and demonstrate that criticality cannot occur.
In addition,
- however, the loss of containment of all of the rods in a canister and their subsequent relocation on a uniform and optimum pitch has been analyzed.
In the particular case of the Exxon 3. 13 w/o U-235 fuel, assuming the presence of 2,000 ppm of boron, the optimum pitch is.632 inches.
On such a pitch a square array would be 12 inches on the side with a K
of about 0.95.
When the 2,000 ppm boron and a
15,000 YiWD/MTU burnup is taken into account, the K
reduces to 0.55.
This particular fuel rod arrangement is thought to be extremely unlikely.
The staff also has considered the case of a canister being not completely filled with the fuel rods since damaged fuel rods will be stored in a separate canister in individual steel tubes and the remaining undamaged rods may then not fill a canister.
However, the neutron leakage from a partially filled canister will be even greater than that of a completely filled canister (i.e.,
2 x 179
= 358 rods);
hence, for a partially filled canister, the Y
g 0.55.
In summary, none of the postulated accidents could result in criticality.
In the analysis of the postulated accidents it was assumed that 2,000 ppm boron is diluted =in the pool cooling water.
Credit for this boron is allowed through the use of the double contingency principle.
In view of the above, the accident analysis presented for the Ginna consolidated spent fuel storage is acceptable.
2.1.3 Conclusions The staff concludes that the spent fuel pool, including the consolidated fuel, meets the General Design Criterion 62 as regarding criticality.
This conclusion is based on the following considerations:
1.
Acceptable calculational methods which have been verified by comparison to critical experiments have been used.
2.
Assumptions regarding the enrichment of the fuel rods which have been analyzed are conservative.
3.
A series of credible accidents has been considered.
4.
Allowance for uncertainties in the estimation of the applicable multiplication factor is conservative.
5.
The estimated final value of the multiplication factor meets the NRC acceptance criterion.
2.2 Naterials The staff has reviewed the materials compatibility and the corrosion degradation aspects of the storage canisters.
The canisters are designed to store the equivalent number of fuel rods from two fuel assemblies and can be placed in either Region I or Region II rack locations of the spent fuel storage pool.
2.2.1 Evaluation The canisters are made of stainless steel type 304, the same materials used to construct the spent fuel pool racks that hold the canisters.
In a safety evaluation dated September 10, 1984, the staff concluded:
( 1) that the corrosion that will occur in the spent fuel storage pool environment should
4-be of little significance during the life of the plant.
Components in the spent fuel storage pool are constructed of alloys which have a low differential galvanic potential between them and have a high resistance to general corrosion, localized corrosion, and galvanic corrosion; (2) that the environmental compat-ibility and stability of the materials used in the spent fuel storage pool is adequate based on the test data and actual service experience in operating reactors; and (3) that the selection of appropriate materials of construction bv the licensee meets the requirements of General Design Criterion 61 in Appendix A to 10 CFR Part 50.
2.2.2 Conclusion Based on the above evaluation, the staff concludes that the proposed canisters for storage of consolidated fuel rods will have little significant corrosion degradation during the life of the plant, provide adequate material compat-ibility and stability with the environment in which they will be used, and meet the requirements of General Design Criterion 61, as related to fuel storage systems designed with appropriate confinement, and are, therefore, acceptable.
2.3 S ent Fuel Pool Coolin and Load Handlin 2.3. 1 S ent Fuel Pool Coolin and Deca Heat Load In 1981, the staff reviewed and approved a proposed spent fuel pool cooling system modification for R.
E.
Ginna (Ref. 9).
This modification will be implemented in 1986, and will consist of the addition of a new cooling loop in parallel with the existing loop which is sized to accommodate the maximum normal and abnormal heat loads.
Further, the licensee has stated that the decay heat load on the spent fuel pool cooling system resulting from storage of consolidated fuel will remain below the ~reviously approved spent fuel pool cooling system design capability of 16 x 10 Btu/hr.
Additionally, the maximum pool water temperature will not exceed the Technical Specification limit of 150'F.
Since the present capability of up to 1016 fuel assemblies will not be increased, the staff concludes that the previously approved spent fuel pool cooling system will acceptably handle the maximum normal and abnormal heat loads for the proposed storage of consolidated fuel.
As indicated
- above, the decay heat loads will not exceed those previously considered and approved during the pool cooling system modification review in 1981.
Therefore, the staff concludes that the associated boiloff rate also will not exceed that which was previously accepted.
Similarly the staff concludes that demands on pool water makeup will not exceed those previously reviewed and approved and, therefore, the makeup capability is acceptable.
The canisters containing consolidated fuel are considered a heavy load per NUREG-0612 criteria and will be transported within the pool using a special tool suspended from a 5 ton hook of the 40 ton auxiliary building crane.
In a safety evaluation report dated October 1,
1984, the staff reviewed and approved modifications to the auxiliary building crane in order to meet the crane single-failure criteria of NUREG-0612 and NUREG-0554.
Therefore, handling of consolidated fuel will be performed in accordance with the guide-lines of NUREG-0612 with regard to limiting the chance of an unacceptable heavy load drop.
2.3.3 Conclusions Based on the above review, the staff concludes that the proposed change to R.
E. Ginna Technical Specification 5.4.4 regarding storage of consolidated fuel is in accordance with the applicable guidelines of SRP Section
- 9. 1.2,
- 9. 1.3 and 9. 1.5 and is, therefore, acceptable.
2.4 Occu ational Radiation Ex osure The staff has estimated the increment in onsite occupational dose during normal operations considering the proposed storage of consolidated spent fuel.
This estimate is based on information supplied by the licensee for occupancy times and for dose rates in the spent fuel area from radionuclide concentrations in the SFP water.
The spent fuel assemblies themselves contribute a negligible amount to dose rates in the pool area because of the depth of water shielding the fuel.
Based on present and projected operations in the spent fuel pool
- area, we estimate that the proposed storage of consolidated spent fuel should add less than one percent to the total annual occupational radiation exposure at the plant.
The small increase in radiation exposure should not affect the licensee's ability to maintain individual occupational dose to as low as is reasonably achievable levels (ALARA) and within the limits of 10 CFR Part 20.
Thus, the staff concludes that storing additional fuel in the SFP will not result in any significant increase in dose received by workers.
2.5 Radiolo ical Conse uences of Accident Involvin Postulated Mechanical ama e to ent ue For evaluation of offsite radiological consequences of accidents involving
.consolidated spent fuel stored in the spent fuel pool, four types of accidents were considered; a cask drop or tip, a tornado missile impact, a standard fuel assembly
- drop, and a fully-loaded consolidated spent fuel canister (classified as a heavy load) drop while handling standard fuel assemblies and/or spent fuel canisters.
These are discussed below.
2.5. 1 Cask Dro /Ti Accidents The staff, in its Safety Evaluation of November 14,
- 1984, has judged that the auxiliary building crane meets the intent of Guideline 7 of NUREG-0612, Section
- 5. 1. 1.
The staff, therefore, does not postulate a cask drop or tip accident which could damage stored spent fuel.
2.5.2 Tornado Nissile Accidents The design tornado missile, established in the staff review of Systematic Evaluation Program (SEP) Topics III-2, Wind and Tornado Loadin s, and III-4.A, Tornado Missiles, is a
1490 lb. wooden po e, 35 eet in engt and 13.5 inches h
1dip h
k h
i 1
1 ly f7oft/
The minimum decay time of consolidated spent fuel rods contained in storage canisters located in the pool is five years.
The staff judges that the worst position for impact of this missile would be that centered on a fuel storage location where a total of nine fuel storage cells could be damaged in reracked pool sections (staff SER of November 14, 1984), or two total assemblies in unreracked pool sections.
In either case, the offsite radio)ggical consequences due to the release of volatile gap activity (almost totally Kr) due to missile impact are bounded, due to the five year minimum decay time of consolidated fuel assemblies, by the consequences determined in the November 14, 1984 staff SER.
After long decay time periods, such as 5 years or more, th~5volatile gap radio-nuclides have decayed to insignificant levels, except for Kr which has a 10.8 year half-life.
These (0-2 hr.) bounding offsite radiological consequence values are:
1) 63 rem thyroid* and 0. 1 rem whole body for impact with stored assemblies in the unreracked section of the pool; and 2) 2 rem thyroid* and
- 0. 1 rem whole body for impact with stored assemblies in the reracked section of the pool.
Both limiting sets of consequences are well within the guideline values of 10 CFR Part
- 100, and are, therefore, acceptable.
2.5.3 Standard Fuel Assembl Dro The offsite radiological consequences of the drop of a standard fuel assembly are
- bounded, due to the five year minimum decay time of consolidated fuel assemblies, by the consequences determined for the Fuel Handling Accident in the Staff SER of November 14, 1984.
These (0-2 hr.) bounding offsite radio-logical" consequence values are:
- 1) 44 rem thyroid* and 0. 1 rem whole body in the unreracked section of the pool; and 2) 1 rem thyroid* and 0. 1 rem whole body in the reracked section of the pool.
Both limiting sets of consequences are well within the guidelines value of 10 CFR Part 100.
2.5.4 Consolidated S ent Fuel Canister Dro The movement of'anisters of consolidated spent fuel over spent fuel stored in the pool requires a change in the Technical Specifications because the fully loaded canister weight will be approximately 2300 lbs.
This exceeds the 2000 lb weight of a standard fuel assembly and its handling tool.
The fully loaded consolidated spent fuel canister must thus be classified as a heavy load.
131
- The key radionuclide producing thyroid dose, I, with an 8.05 day half-life, has decayed to negligible concentrations at 5 years cooldown time.
Thus the bounding thyroid doses are far beyond any expected thyroid doses resulting from the accident.
Present Technical Specification 3. 11.3 states that "A load in excess of one fuel assembly and its handling tool shall never be stationed or permitted to pass over storage racks containing spent fuel."
A new Technical Specification is proposed (TS 3. 11.5) which states that "The restriction of 3. 11.3 above shall not apply to the movement of canisters containing consolidated fuel rods if the spent fuel racks beneath the transported canister contain only spent fuel that has decayed for at least 60 days since reactor shutdown."
This proposed Technical Specification allows canisters containing consolidated fuel rods to be transported over either standard stored spent fuel assemblies with 60 days decay or stored canisters containing spent fuel with at least five years decay and will result in very small (0-2 hr ) thyroid and whole body doses
( 0. 1 rem).
In the case of a canister dropped onto a standard stored spent fuel
- assembly, which will have at least 60 days decay (new T.S. 3. 11.5), the staff judges that the (0-2 hr) offsite radiological consequences due to the postulated release of the volatile gap activities of both the canister (at least five years decay) and the stored assembly are bounded by the consequences of the tornado missile impact onto nine standard assemblies in the reracked section of the pool, as discussed in the staff SER of November 14, 1984.
These bounding consequences are 2 rem thyroid and 0. 1 rem whole body, both well within the guidelines of 10 CFR Part 100.
2.5.5.
Conclusions Since the staff has concluded that the auxiliary building crane meets the intent of Guideline 7 of NUREG-0612, Section
- 5. 1. 1, a cask drop or tip accident which could damage stored spent fuel is sufficiently unlikely that it need not be evaluated.
The staff also concludes that a tornado missile accident resulting in damage to either two standard and/or consolidated stored spent fuel assemblies in unreracked pool sections (staff SER of November 14, 1984), or nine stan-dard and/or consolidated assemblies in reracked pool sections, will result.
in atmospheric radionuclide releases with (0-2 hr) offsite radiological consequences which are well within the guidelines of 10 CFR Part 100.
The staff concludes, additionally, that the (0-2 hr) offsite radiological consequences of the drop of a standard fuel assembly are bounded, due to the five year minimum decay time of consolidated spent fuel assemblies, by the consequences determined for the Fuel Handling Accident in the staff SER of November 14,
- 1984, and are, therefore, well within the guideline values of 10 CFR Part 100.
Finally, the staff judges that the (0-2 hr) offsite radiological consequences due to the postulated release of the volatile gap activities of both a dropped fully loaded consolidated spent fuel canister and a stored standard or con-solidated fuel assembly are bounded by the (0-2 hr) offsite radiological con-sequences of the tornado missile impact onto nine standard stored spent fuel assemblies in the reracked section of the pool, as discussed in the staff SER of November 14, 1984.
These bounding consequences are well within the guide-lines of 10 CFR Part 100.
2.5.6 Structural Evaluation References 2 and 10 document the structural analysis performed and the staff evaluation for the Ginna spent fuel storage racks under the loads due to storage of consolidated fuel.
This evaluation determined that the structural integrity of the racks would be maintained under a seismic event.
The cannisters will be fabricated from SS304.
All Welding will be in accordance with ASME Section 3, subsection NF requirements.
The design loads will satisfy the criteria for a seismic category 1 component.
Based on the above, the staff concludes the proposed change to be acceptable.
3.0 OVERALL CONCLUSION Based on the review, the staff concludes that the licensee's proposed storage of consolidated fuel assemblies is acceptable.
In addition, the proposed Technical Specifications are acceptable.
The staff concludes, 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 proposed manner; and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
4.0 ACKNOWLEDGEMENT This Safety Evaluation was prepared by the following NRC staff:
C. Miller, J. Kelly, M. Wohl, J.
Wing, L. Lois, and A. Singh.
5.0 REFERENCES
1.
R.
W. Kober (RG&E) to J.
A. Zwolinski (NRC),
SUBJECT:
Response
to NRC Staff Questions, dated June 26, 1985.
2.
R.
W. Kober (RG8E) to Director, NRR, dated April 2, 1984.
3.
R.
W. Kober (RG8E) to J.
A. Zwolinski (NRC),
SUBJECT:
Responses to NRC Questions on the February 27, 1985 Submittal, dated July 2, 1985.
4.
R.
F. Barry, "LEOPARD - A Spectrum Dependent Non-Spatial Depletion Code for the IBM-7094," WCAP-3269, (September 1963).
5.
W.
R. Caldwell, "PDQ-7 Reference Manual," WAPD-TM-678, (January 1967).
6.
Electric Power Research Institute, "Fission Product Data for Thermal
- Reactors, Part 1 and Part 2:
Data Set for EPRI-CINDER and Users Manual for EPRI-CINDER Code and Data,"
EPRI NP-356, Final Report (1976).
7.
G.
S. Hoovler, et al., "Critical Experiments Supporting Underwater Storage of Tightly Packed Configurations of Spent Fuel Pins,"
BAW-1645-4, (November 1981).
8.
J.
E. Maier (RG&E) to H.
R.
Denton (NRC), dated February 23, 1983.
9.
D.
M. Crutchfield (NRC) to J.
E. Maier (RG&E),
SUBJECT:
Spent Fuel Pool Cooling System Modifications (Ginna), dated November 3, 1981.
- 10. J.A. Zwolinski (NRC) to R.
W. Kober(RG&E),
SUBJECT:
Increase of the Spent Fuel Pool Capacity, dated November 14, 1984.
Dated:
December 16, 1985
ATTACHMENT TO LICENSE AMENDMENT NO.
FACILITY OPERATING L'ICENSE NO.
DPR-8 DOCKET NO. 50-244 Revise Appendix A Technical Specifications by re oving the pages identified below and inserting the enclosed pages.
The re ised pages are identified by the captioned amendment number and contain mar inal lines indicating the area of change.
REMOVE l-~
3.11-2 3.11-4 5.,4-2 to 5.4-4 INSERT ip I
- 3. 11-2
'3.11-4 5.4-2 to 5.4-4 5.4-4a
l 1
h 'g N
I'!
t N
a The previous criticality analyses for the Ginna spent fue pool utilized the LEOPARD (Ref. 4),
PD(07 (Ref.
- 5) and CINDER (Ref.
- 6) cod s.
The Boraflex absorbers on the racks were treated with blackness theo y.
A similar metho-dology has been employed for the analysis of the conso idated fuel storage.
The validation was based on experiments performed by abcock and Wilcox Company for this purpose (Ref. 7).
The tightly pac d fuel rods present two modeling difficulties, namely the high metal/w er and the related dis-tortion of the neutron spectrum (compared to the igner Wilkins model) and the "rubber bond" method to define the amount of water to be included in the fuel region.
This method provides a consistent definition of the fuel region for the analyses of the critical experiments well.
The critical experiments analyzed constituted an adequate test of the omponent-codes of the model.
The results of the analyses were compared t the experiment as well as to KENO-IV calculations from Reference 7.
Th comparison indicated that the LEOPARD-PD(07-CINDER model results in a c)oser prediction than KENO-IV, a previously approved model.
Therefore, tHe LEOPARD-PD(07-CINDER model is considered validated and acceptable.
The analytical model described above as supplemented with burnup capability in order to be used for spent fuel.
The burnup code was part of the spent fuel pool modification resubmittal (Ref. 2) and has been approved.
An allowance was made for geometrica manufacturing and thermal deviations.
The licensee took advantage of exist ng calculations for 3. 13 w/o U-235 Exxon fuel to support conservative cr'ticality estimates for the West Valley New York fuel assemblies which had an initial enrichment of 2.8 w/o U-235.
The calculation yields K
vs burn p which is everywhere less than 0.8.
This is true for either 358 rod/cani ter (i.e.,
2 x 179 where 179 is the number of rods/assembly) or a reduced loading of 350 rods/canister (2 x 175).
For the fuel rods stored in West V lley the average burnup is about 15,000 MWD/MTU and the K for consolida d fuel is 0.63.
In view of the estimated low values for K it was no deemed necessary to calculate the specific uncer-tainties, therefore th corresponding estimates of the fuel pool starage capacity increase of A ril 2, 1984, were adopted (Ref. 2).
The total reactivity adjustmen to the calculated value for the consolidated rods is
.056 aK.
The above results do not include the effect of the 2,000 ppm of dissolved boron in the pool water.
The results of these estimates assure that for fuel des'gns satisfying the enrichment-burnup criteria of Figure 5.4-2 of the Gin a Technical Specifications, the K f criteria of 0.95 is satisfied
- and, herefore, the proposed consolidatel fuel storage is acceptable.
2.1.2 Accid t Anal sis C.
The accide analysis performed in the submittals of April 2 j1984 (Ref.
2) and Febru ry 23, 1983 (Ref. 8) are also applicable for the Vase of the con-solidate fuel storage with respect to criticality In addition,
- however, the loss of containment of all of the rods in a ca >ster and their subsequent relocation on a uniform and optimum pitch has bee analyzed.
In the particular case of the Exxon 3. 13 w/o U-235 fuel, assuming he presence of 2,000 ppm of
L
~
J l
ll p4 1
A
/1 I
t'
boron, the optimum pitch is.632 inches.
On such a pitch a square array would be 12 inches on the side with a K
of about 0.95.
When the $,000 ppm boron and a 15,000 NWD/MTU burnup is taken into account, the K
ry6uces to 0.55.
This particular fuel rod arrangement is thought to be extrq&ely unlikely.
The aitase of a acid'anister being not completely filled uitb he fuel rods ~
pl
~~ damaged.guel rods will be tor in a separate canister n individual steel tubes~e remaining nu
~
then not fill a he+)'- canister.
However the leakage from a n s er will be even reater tha a ~hob canister~ (i.e.,
2 x 179
= 358 rods); hence, e
.55.
In summary, none of the postulated accidents could result in cubi icality.
In the analysis of the postulated. accidents it was assumed that 2,0 0 ppm boron is diluted in the pool cooling water.
Credit for this boron is llowed through the use of the double contingency principle.
In view of the above, the accident analysis presented for the Ginna consolidated spent fuel torage is acceptable.
2.1. 3 Conclusions The staff concludes that the spent fuel pool including the consolidated fuel, meets the General Design Criterion 62 s regarding criticality.
This conclusion is based on the following consi rations:
1.
Acceptable calculational methods whic have been verified by comparison to critical experiments h ve been used.
2.
Assumptions regarding the enrichme t of the fuel rods which have been analyzed are conservative.
3.
A series of credible accidents s been considered.
4.
Allowance for uncertainties i
the estimation of the applicable multiplication factor is con rvative.
5.
The estimated final value the multiplication factor meets the NRC acceptance criterion.
2.2 Materials The staff has reviewed the materials compatibility and the corrosion degradation aspects of the storage ca isters.
The canisters are designed to store the equivalent number of fue rods from two fuel assemblies and can be placed in either Region I or Regi n II rack locations of the spent fuel storage pool.
2.2.1 Evaluation The canisters are de of stainless steel type 304, the same materials used to construct the ent fuel pool racks that hold the canisters.
In a safety evaluation dated eptember 10, 1984, the staff concluded:
(1) that the corrosion that ll occur in the spent fuel storage pool environment should
i%,
U
be of little significance during the life of the plant.
Compon nts in the spent fuel storage pool are constructed of alloys which have a low d'fferenti galvanic potential between them and have a high resistance ~ general corrosion, localized corrosion, and galvanic corrosion; (2) that the enysronmental compat-ibility and stability of the materials used in the spent fup'l storage pool is adequate based on the test data and actual service experiepce in operating reactors; and (3) that the selection of appropriate materials of construction by the licensee meets the requirements of General Design Criterion 61 in Appendix A to 10 CFR Part 50.
2.2.2 Conclusion Based on the above evaluation, the staff concludes at the proposed canisters for storage of consolidated fuel rods will have li tie significant corrosion degradation during the life of the plant, provide adequate material compat-ibility and stability with the environment in wh'ch they will be used, and meet the requirements of General Design Criteri n 61, as related to fuel storage systems designed with appropriate conf'nement, and are, therefore, acceptable.
2.3 S ent Fuel Pool Coolin and Load Handle n 2.3. 1 S ent Fuel Pool Coolin and Deca eat Load In 1981, the staff reviewed and approv~
a proposed spent fuel pool cooling system modification for R.
E.
Ginna
( ef. 9).
This modification will be implemented in 1986, and will consis of the addition of a new cooling loop in parallel with the existing loop ich is sized to accommodate the maximum normal and abnormal heat loads.
F rther, the licensee has stated that the decay heat load on the spent fuel pool cooling system resulting from storage of consolidated fuel will remain elow the ~reviously approved spent fuel pool cooling system design capabi lit of 16 x 10 Btu/hr.
Additionally, the maximum pool water temperature ill not exceed the Technical Specification limit of 150'F.
Since the pr sent capability of up to 1016 fuel assemblies will not be increased, the s ff concludes that the previously approved spent fuel pool cooling system wi acceptably handle the maximum normal and abnormal heat loads for the propose storage of consolidated fuel.
As indicated above, the cay heat loads will not exceed those previously considered and approved uring the pool cooling system modification review in 1981.
Therefore, thj staff concludes that the associated boiloff rate also will not exceed hat which was previously accepted.
Similarly the staff concludes that demand on pool water makeup will not exceed those previously reviewed and approve and, therefore, the makeup capability is acceptable.
2.3.2 Load Handli The caqjsters co taining consolidated fuel are considered a heavy load per NUREG-$012 crit ria and will be transported within the pool using a special tool suspended from a 5 ton hook of the 40 ton auxiliary building crane.
In a safety evaluation report dated October 1, 1984, the staff reviewed and
I II fi l
lt I r
P I
l lt l
)
I Pf gt r
1 tg
= tl l
Il lk r I
I' t
'I
'I 1
l
~ '
P I
~ 1 IP t I
P 4
I I'tt I
lt tk I it ~
It t
il Ik l
tr I
fk I
1 tt l I
li
-=lit P
1 b
k k
tp t
P It II
~l I'
II
,, P II
'I lf I
C iM l
2.5.2 Tornado Nissile Accidents The design tornado missile, established in the staff review of Sy tematic Evaluation Program (SEP) Topics III-2, Wind and Tornado Loadin
, and III-4.A,
~
~
~
~
~
~
~
~
~
~
~
~
Tornado Missiles, is a
1490 lb. wooden po e, eet sn engt and 13.5 inches h
id'M h
k
'h i
1 1 iy f7Df/
a4-+he minimum decay time of conso idated spent fuel rods contained in storage canisters located in the pool is ive years.
The staff judges that the worst position for impact of this m'ile would be that
'centered on a fuel storage location where a total of nin fuel storage cells could be damaged in reracked pool sections (staff SER o
November 14, 1984), or two total assemblies in unreracked pool sections.
In ither case, the offsite radiologj~al consequences due to the release of vola ile gap activity (almost totally Kr) due to missile impact are bounded, e to the five year minimum decay time of consolidated fuel assemblies, by th consequences determined in the November 14, 1984 staff SER.
After long de y time periods, such as 5
years or more, the glatile gap radionuclides ve decayed to insignificant levels, except for Kr which has a 10.8 year alf-life.
These (0-2 hr.)
, bounding offsite radiological consequence v
ues are:
1) 63 rem thyroid* and
- 0. 1 rem whole body for impact with stored semblies in the unreracked section of the pool; and 2) 2 rem thyroid* and 0.
rem whole body for impact with stored assemblies in the reracked sectio of the pool.
Both limiting sets of consequences are well within the gui eline values of 10 CFR Part 100, and are, therefore, acceptable.
2.5.3 Standard Fuel Assembl Dro The offsite radiological consequ ces of the drop of a standard fuel assembly are bounded, due to the five ye r minimum decay time of consolidated fuel assemblies, by the consequence determined for the Fuel Handling Accident in the Staff SER of November 4,
1984.
These (0-2 hr.) bounding offsite radio-logical consequence values e:
- 1) 44 rem thyroid* and 0. 1 rem whole body in the unreracked section the pool; and 2) 1 rem thyroid* and 0. 1 rem whole body in the reracked sect on of the pool.
Both limiting sets of consequences are well within the guid ines value of 10 CFR Part 100.
2.5.4 Consolidated S
nt Fuel Canister Dro The movement of can ters of consolidated spent fuel over spent fuel stored in the pool requires change in the Technical Specifications because the fully loaded canister w ight will be approximately 2300 lbs.
This exceeds the 2000 lb weight of a standard fuel assembly and its handling tool.
The fully loaded consolidated s 4nt fuel canister must thus be classified as a heavy load.
131
- The key rad onuclide producing thyroid dose, I, with an 8.05 day half-life, has decayed to negligible concentrations at 5 years cooldown time.
Thus the bounding t yroid doses ar far beyond any expected thyroid doses resulting from the accid nt.
~ i
'l ll F I'
e ia
~
c s)s 7-Present Technical Specification
- 3. 11.3 states that "A load in excess of one fuel assembly and its handling tool shall never be stationed or permit ed to pass over storage racks containing spent fuel."
A new Technical Spec fication is proposed (TS 3. 11.5) which states that "The restriction of 3. 11.
above shall not apply to the movement of canisters containing consolidatpd fuel rods if the spent fuel racks beneath the transported canister contain Pnly spent fuel that has decayed for at least 60 days since reactor shutdown."
This proposed Technical Specificagjgg allows canisters containing cphsolidated fuel rods to be transported over>standard stored spent fuel assemb ies with 60 days decay or stored canisterS be4h-containing spent fuel with at least five years deca will result in very small (0-2 hr) thyroid and whole body doses
( 0. 1 rem).
In t e case of a canister dropped onto a standard stored sfent fuel assembly, which will have at least 60 days decay (new T.S. 3. 11.5) g the staff judges that the (0-2 hr) offsite radiological consequences due/to the postulated release of the volatile gap activities of both the canister (at least five years decay) and the stored assembly are bounded by he consequences of the tornado missile impact onto nine standard assemblie in the reracked section of the pool, as discussed in the staff SER of Nove ber 14, 1984.
These bounding consequences are 2 rem thyroid and 0. 1 rem whole body, both well within the guidelines of 10 CFR Part 100.
2.5.5.
Conclusions Since the staff has concluded that the aux'ary building crane meets the intent of Guideline 7 of NUREG-0612, Sect'on
- 5. 1. 1, a cask drop or tip accident which could damage stored spent fuel is sufficiently unlikely that it need not be evaluated.
The staff also concludes that a tom o missile accident resulting in damage to either two standard and/or conso dated stored spent fuel assemblies in unreracked pool sections (staff ER of November 14, 1984), or nine stan-dard and/or consolidated assembli s in reracked pool sections, will result in atmospheric radionuclide rele ses with (0-2 hr) offsite radiological consequences which are well wit in the guidelines of 10 CFR Part 100.
The staff concludes, additio lly, that the (0-2 hr) offsite radiological consequences of the drop of a standard fuel assembly are bounded, due to the five year minimum decqy time of consolidated spent fuel assemblies, by the consequences determi kd for the Fuel Handling Accident in the staff SER of November 14, 198
, and are, therefore, well within the guideline values of 10 CFR Part 00.
Finally, the staff udges that the (0-2 hr) offsite radiological consequences due to the postulated release of the volatile gap activities of both a dropped fully loaded con 6lidated spent fuel canister and a stored standard or con-solidated fuel ssembly are bounded by the (0-2 hr) offsite radiological con-sequences of t e tornado missile impact onto nine standard stored spent fuel assemblies i
the reracked section of the pool, as discussed in the staff SER of November 14, 1984.
These bounding consequences are well within the guide-lines of 10 CFR Part 100.
~
~
g
~
l 1
II 1
For further details with respect to this action see (1) the application for amendment dated February 27, 1985 as supplemented on, June 10, June 26, and July ll, 1985, (2) Amendment No. 12 to Facility Operating License No. DPR-18, and (3)rthe Commission's related Safety Evaluation.
All of these items are available for public inspection at the Commission's Public Document
- Room, 1717 H Street, N.W., Washington, D.C.
and at the Rochester Public Library, 115 South Avenue, Rochester, New York 14610.
A copy of items (2) and (3) may be obtained upon request addressed to the U.S. Nuclear Regulatory Commission, Washington, D.C.
20555, Attention:
Director, Division of PWR Licensing-A, NRR.
Dated at Bethesda, Maryland, this 16th day of December 1985.
FOR THE NUCLEAR REGULATORY COMMISSION ORIGIML SIGNED BY George E. Lear, Director Project Directorate 0'1 Division of PWR Licensing DISTRIBUTION NRC PDR Local PDR CMiller JKel ly PShuttleworth PWR 81 Reading GLear
/@0
)Pl PWRt1: DPWR-A BWRgl/DBbfg BTPgl/DBWR E
D>
CMi 1 1er: jg JKel ly 34 CJWmerson z'~'I
'2//~/85 12/ CQ/85 12/ /¹0/85 12/tm/85 BWR¹1/DB~W PWR¹1/OPWR-A JZwolinski GLear 2 /l~/85 (g /l y/85
r J
~
'h
(~
I H
DISTRIBUTION:
(~n NRCPDR LPDR NSIC BWRk'1 R/F PShuttleworth CMiller JKelly JZwolinski PWR81 R/F GLear OELD ACRS (10)
EJordan JPartlow BGrimes HThompson LHarmon TBarnhart (4)
WJones OPA RDiggs (w/TAC)
VBenaroya MWohl JWing LLois ASingh
- 0
'T
~ q/