ML24142A444
| ML24142A444 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 05/20/2024 |
| From: | Public Service Enterprise Group |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML24142A428 | List:
|
| References | |
| LR-N24-0029, LAR H24-02 | |
| Download: ML24142A444 (1) | |
Text
ENCLOSURE 2
VOLUME 16
HOPE CREEK GENERATING STATION
IMPROVED TECHNICAL SPECIFICATIONS CONVERSION
ITS CHAPTER 4.0 DESIGN FEATURES
Revision 0 LIST OF ATTACHMENTS
- 1. ITS 4.0, Design Features
ATTACHMENT 1
ITS 4.0, Design Features
Current Technical Specifications ( CTS) Markup and Discussion of Changes ( DOCs)
SECTION 5.0
DESIGN FEATURES A01 ITS 4.0 ITS 5.0 DESIGN FEATURES
4.1 5.1 SITE LOCATION
Hope Creek Generating Station is located in Salem County, New Jersey along the eastern shore of the Delaware River approximately 8 miles southwest of Salem, New Jersey and 18 miles south of Wilmington, Delaware.
5.2 CONTAINMENT
CONFIGURATION
5.2.1 The primary containment is a steel structure composed of a spherical lower portion, a cylindrical middle portion, and a hemispherical top head which form a drywell. The drywell is attached to the suppression chamber through a series of downcomer vents. The suppression chamber is a steel pressure vessel in the shape of a torus. The drywell has a nominal free air volume of 169,000 cubic feet. The suppression chamber has an air volume of 137,000 cubic LA01 feet and a water region as described in Technical Specification Bases 3/4.6.2, Depressurization Systems.
DESIGN TEMPERATURE AND PRESSURE
5.2.2 The primary containment is designed and shall be maintained for:
- a. Maximum internal pressure 62 psig.
- b. Maximum internal temperature: drywell 340oF.
suppression pool 310oF.
- c. Maximum external differential pressure 3 psid.
5.2.3 The secondary containment consists of the Reactor Building, and a portion of the main steam tunnel and has a free volume of 4,000,000 cubic feet.
HOPE CREEK 5-1 Amendment No. 230
Intentionally Left Blank
HOPE CREEK 5-2 Amendment No. 230 A01 ITS 4.0 ITS
Intentionally Left Blank
HOPE CREEK 5-3 Amendment No. 230
DESIGN FEATURES
4.2 5.3 REACTOR CORE
4.2.1 FUEL ASSEMBLIES
5.3.1 The reactor core shall contain 764 fuel assemblies. Each assembly shall consist of a matrix of Zircalloy or ZIRLO fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material and water rods. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with NRC staff approved codes and methods and have been shown by tests or analyses to comply with all safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in non-limiting core regions.
A maximum of twelve GE14i Isotope Test Assemblies may be placed in non-limiting core regions., beginning with Reload 16 Cycle 17 core reload, with the purpose of obtaining surveillance data to verify that the GE14i cobalt Isotope Test Assemblies perform satisfactorily in service (prior to evaluating a future license amendment for use of these design features on a L01 production basis). Each GE14i assembly contains a small number of Zircaloy -2 clad isotope rods containing Cobalt-59. Cobalt-59 targets will transition into C obalt-60 isotope targets during cycle irradiation of the assemblies. Details of the GE14i assemblies are contained in GE -Hitachi report NEDC-33529P, Safety Analysis Report to Support Introduction of GE14i Isotope Test Assemblies (ITAs) in Hope Creek Generating Station, Revision 0, dated December 2009.
4.2.2 CONTROL ROD ASSEMBLIES
5.3.2 The reactor core shall contain 185 cruciform shaped control rod assemblies. The control material shall be boron carbide powder (B 4C) and/or hafnium metal. The absorber material has LA02 a nominal absorber length of 143 inches. as approved by the NRC
DESIGN PRESSURE AND TEMPERATURE
5.4.1 The reactor coolant system is designed and shall be maintained:
- a. In accordance with the code requirements specified in Section 5.2 of the FSAR, LA01 with allowance for normal degradation pursuant to the applicable Surveillance Requirements,
- b. For a pressure of:
- 1. 1250 psig on the suction side of the recirculation pump.
- 2. 1500 psig from the recirculation pump discharge to the jet pumps.
- c. For a temperature of 575°F.
HOPE CREEK 5-4 Amendment No. 184 A01 ITS 4.0 ITS DESIGN FEATURES
5.4 REACTOR COOLANT SYSTEM (continued)
VOLUME LA01
5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 21,970 cubic feet at a nominal steam dome saturation temperature of 547°F.
5.5 DELETED
4.3 5.6 FUEL STORAGE
4.3.1 CRITICALITY
4.3.1.1 5.6.1 The spent fuel storage racks are designed and shall be maintained with:
Add proposed ITS 4.3.1.1.a M01 4.3.1.1.b a. A keff equivalent to less than or equal to 0.95 when flooded with unborated water, A01 including all calculational uncertainties and biases as described in Section 9.1.2 of the FSAR.
4.3.1.1.c b. A nominal 6.308 inch center-to-center distance between fuel assemblies placed A02 in the storage racks.
4.3.1.2.b 5.6.1.2 The keff for new fuel for the first core loading stored dry in the spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is assumed. cons, which il A01 under optim lowanctainti as 3 described in S 4.3.2 DRAINAGE Add proposed ITS UFSAR M01 4.3.1.2.aand c 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 199' 4". in ft 4.3.3 CAPACITY is designed and shall be maintained with A01
5.6.3 The spent fuel storage pool shall be limited to a storage capacity of no more than 4006 fuel assemblies.
5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT See ITS 5.5
5.7.1 The components identified in Table 5.7.1-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7.1-1.
HOPE CREEK 5-5 Amendment No. 234
DISCUSSION OF CHANGES ITS 4.0, DESIGN FEATURES
ADMINISTRATIVE CHANGES
A01 In the conversion of the Hope Creek Generating Station (HCGS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1433, Rev. 5.0, "Standard Technical Specifications - General Electric BWR/4 Plants" (ISTS).
These changes are designated as administrative changes and are acceptabl e because they do not result in technical changes to the CTS.
A02 CTS 5.6.1.b states that spent fuel storage racks are designed with a nominal center-to-center distance between fuel assemblies placed in the storage racks as 6.308 inches. ITS 4.3.1.1. c also provides the nominal center to center design distance between fuel assemblies place in the storage racks and specified as 6.3 inches. This is a presentation change to a nominal value specified in the CTS.
The purpose of the CTS requirement is to ensure the spent fuel storage racks are designed and maintained to comply with 10 CFR 50.68, Criticality accident requirements. Since the center to center distance between fuel assemblies placed in the storage racks is stated as a nominal value, it is unnecessary to provide an accuracy to one thousandth of an inch. ITS 4.3.1.1 continues to require spent fuel storage racks be designed to comply with 10 CFR 50.68 and continue to assure control of fuel storage in the spent fuel storage pool, including limiting the center to center distance between fuel assemblies to a nominal value.
This is a presentation preference change and is designated as administrative.
A03 CTS 5.6.1.2 requires, in part, keff for new fuel shall not exceed 0.98 when aqueous foam moderation is assumed. ITS 4.3.1.2.b is revised to state that the new fuel storage racks shall be design and maintained with keff 0.98 under optimum moderation conditions and includes the detail on the assumptions; which includes an allowance for uncertainties as described in Section 9.1 of the UFSAR. This changes the CTS by including administrative detail related to aqueous foam moderation.
The purpose of the CTS requirement is to ensure the fuel storage racks are designed and maintained to comply with 10 CFR 50.68, Criticality accident requirements. 10 CFR 50.68 requires the k eff not to exceed 0.98 in the event of optimum moderation, at a 95 percent probability, 95 percent confidence level.
The added administrative detail ensures the uncertainties necessary to meet the regulatory requirements are described in a plant licensing basis document.
These changes represent administrative detail necessary to maintain compliance with regulations and are acceptable because they do not result in technical changes to the CTS.
Hope Creek Page 1 of 4 DISCUSSION OF CHANGES ITS 4.0, DESIGN FEATURES
MORE RESTRICTIVE CHANGES
M01 CTS 5.6.1 and 5.6.1.2 require the spent fuel and new fuel storage racks to be maintained within specific criticality requirements; however, CTS does not specify U-235 enrichment for the fuel assemblies. ITS 4.3.1.1.a and 4.3.1.2.a require fuel assemblies in the spent and new fuel storage racks to be limited to a maximum nominal U-235 enrichment of 4.9 weight percent. This changes the CTS by adding a specific fuel assembly U-235 enrichment limit.
CTS 5.6.1.2 requires spent fuel storage racks, for new fuel for the first core loading stored dry, be designed not to exceed 0.98 when aqueous foam moderation is assumed. ITS 4.3.1.2.b requires new fuel storage racks to be designed and maintained with a k eff 0.98 under optimum moderation conditions, including allowance for uncertainties described in Section 9.1 of the UFSAR. ITS 4.3.1.2.c also requires that new fuel storage racks are designed and maintained to have a minimum 7 inch center to center distance between fuel assemblies placed in each storage rack row and a minimum 12.25 inch center to center distance between rows in the storage racks. This changes the CTS by adding specific reactivity conditions and fuel assembly spacing requirements for the design and maintenance of new fuel storage racks.
The purpose of ITS 4.3.1.1 and 4.3.1.2 is to prevent criticality in the spent fuel storage racks and new fuel storage vault, respectively. UFSAR Section 9.1.1.1 states that the new fuel storage racks are designed, in part, with sufficient spacing between new fuel assemblies to ensure that when racks are loaded to their administrative limits, the array satisfies the subcriticality requirements.
UFSAR Section 9.1.2.1 states that the spent fuel storage pool facility is designed, in part, to store fuel assemblies (new and spent fuel assemblies) in a subcritical array so that the keff of fuel in the spent fuel storage racks, when flooded with pure water, shall not exceed 0.95, at a 95 percent probability, 95 percent confidence level under normal and abnormal storage conditions. Specifying the maximum nominal U -235 enrichment of fuel assemblies stored in the spent fuel pool and new fuel storage vault, and the establishment of minimum spacing design requirements and clarifying the keff limit for the new fuel storage racks, assists in meeting this design basis and provides assurance that no incident could occur that would result in a hazard to public health and safety. U-235 enrichment of the fuel assemblies and center to center distance between fuel assemblies are important inputs in spent fuel pool and new fuel storage vault criticality analyses. The specific new fuel keff criterion text is revised consistent with the new fuel storage vault criticality analysis and 10 CFR 50.68(b)(3).
This change is acceptable because it provides appropriate limits for the new and spent fuel storage racks and is designated more restrictive because general design requirements in CTS have been added or replaced with specific requirements in ITS consistent with the ISTS and 10 CFR 50.68.
RELOCATED SPECIFICATIONS
None
Hope Creek Page 2 of 4 DISCUSSION OF CHANGES ITS 4.0, DESIGN FEATURES
REMOVED DETAIL CHANGES
LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 5.2 and 5.4 provide description details of the reactor coolant system (RCS) and the containment systems, respectively. This changes the CTS by moving the description details of these systems to the UFSAR.
The removal of these details, which are related to system description, from the Technical Specifications, is acceptable because this type of information is not considered a design feature requirement as described in 10 CFR 50.36(c)(4);
and therefore, is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS retains requirements on RCS OPERABILITY in ITS Section 3.4, Reactor Coolant Systems (RCS), and containment systems OPERABILITY in ITS Section 3.6, Containment Systems. Also, this change is acceptable because the information will be adequately controlled in the U FSAR. Any changes to the UFSAR are made per the provisions of 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications.
LA02 (Type 1 - Removing Details of System Design and System Description, Inc luding Design Limits) CTS 5.3.2 requires, in part, the absorber material of the control rod assemblies to have a nominal absorber length of 143 inches. I TS 4.2.2 does not include this level of detail associated with control rod assemblies. This changes the CTS by moving details of the nominal absorber length of a control assembly to the UFSAR.
The removal of these details, which are related to system design, from the CTS is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS continues to specify that the control material of the control rod assemblies shall be boron carbide or hafnium metal as approved by the NRC.
ITS also retains operational requirements associated with core reactivity in Section 3.1, Reactivity Control Systems, and Section 3.2, Power Distribution Limits, to ensure core parameters are maintained within the limits specified i n the CORE OPERATING LIMITS REPORT and appropriate remedial actions are provided in the event core parameters are discovered not within limits. ITS retains sufficient control rod assembly information, which, if altered or modified, would have a significant effect on safety. These requirements provide adequate assurance the control rod assemblies are of a design previously approved by the NRC. Removed information is more appropriately contained in the UFSAR. The inclusion of the details of control rod design and material in the UFSAR is consistent with the content and purpose of the UFSAR. The UFSAR is controlled under 10 CFR 50.59 which ensures that changes to the information contained in the UFSAR are properly evaluated. Therefore, the removal of this information from the C TS and placement in the UFSAR is acceptable. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the CTS.
Hope Creek Page 3 of 4 DISCUSSION OF CHANGES ITS 4.0, DESIGN FEATURES
LESS RESTRICTIVE CHANGES
L01 CTS 5.3.1 states, in part, that a limited number of lead test assemblies that have not completed representative testing may be placed in non-limiting core regions.
CTS 5.3.1 further describes the details of a limited number of lead test assemblies associated with GE14i isotope test assem blies (ITAs). ITS 4.2.1 does not include detailed information related to use of the GE14i ITAs. This changes the CTS by deleting detail of specific lead test assemblies that have not completed representative testing since the subject ITAs are no longer being placed in the HCGS reactor core.
The purpose of the CTS requirement was to limit loading of ITAs into the reactor core beginning with the fall 2010 refueling outage (HCGS Reload 16 Cycle 17) along with specifying other design details related to the ITAs. The modified fuel assemblies (i.e., GE14i ITAs) were part of a joint pilot program with Global Nuclear Fuel - Americas, LLC (GNF) and GE Hitachi Nuclear Energy, LLC (GEH). The purpose of the pilot program was to obtain data to verify that the modified fuel assemblies perform satisfactorily in service prior to use on a production basis.This initiative has since been terminated and the ITAs have been removed from the HCGS reactor core. PSEG Nuclear LLC does not anticipate future installation of these type of ITAs in the reactor core. Therefore, this requirement is no longer applicable and is deleted. ITS 4.2.1 will continue to retain the allowance that a limited number of lead test assemblies that have not completed representative testing may be placed in non-limiting core regions,
consistent with the ISTS.Deleting information that is no longer applicable since the subject lead test assemblies are no longer in the reactor core has no impact on safety. This change is considered acceptable since the requirements of 10 CFR 50.59 regarding conducting special tests or modifications remain applicable and are sufficient to ensure that a limited number of lead test assemblies placed in nonlimiting core regions will be properly evaluated prior to use. This change is designated as less restrictive because it deletes requirements associated with specific lead test assemblies no longer in use at HCGS.
Hope Creek Page 4 of 4 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)
Design Features 4.0
CTS 4.0 DESIGN FEATURES
5.1 4.1 Site Location Hope Creek Generating Station is located in Salem County, New Jersey along the eastern shore of the Delaware River 1 approximately 8 miles southwest of Salem, New Jersey and 18
[ Text description of site location. ] miles south of Wilmington, Delaware.
5.3 4.2 Reactor Core
5.3.1 4.2.1 Fuel Assemblies 764
The reactor shall contain [560] fuel assemblies. Each assembly shall consist of a 1 matrix of [Zircalloy or ZIRLO] fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO 2) as fuel material and [water rods]. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used.
Fuel assemblies shall be limited to those fuel designs that have been analyzed with NRC staff approved codes and methods and have been shown by tests or analyses to comply with all safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.
5.3.2 4.2.2 Control Rod Assemblies 185 The reactor core shall contain [137] cruciform shaped control rod assemblies. 1 The control material shall be [boron carbide, hafnium metal] as approved by the NRC. or
5.6 4.3 Fuel Storage
4.3.1 Criticality
5.6.1 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:
DOC M01 a. Fuel assemblies having a maximum [k-infinity of [1.31] in the normal reactor core configuration at cold conditions] [average U-235 enrichment of [4.5] weight percent], nominal 4.9 5.6.1.a b. keff 0.95 if fully flooded with unborated water, which includes an 1 allowance for uncertainties as described in [Section 9.1 of the U FSAR], and 6.3
5.6.1.b c. A nominal [6.5] inch center to center distance between fuel assemblies placed in the storage racks.
5.6.1.2 4.3.1.2 The new fuel storage racks are designed and shall be maintained with:
General Electric BWR/4 STS 4.0-1 Rev. 5.0 Amendment XXX 2
Hope Creek Design Features 4.0
CTS 4.0 DESIGN FEATURES
5.6 4.3 Fuel Storage (continued)
DOC M01 a. Fuel assemblies having a maximum [k-infinity of [1.31] in the normal reactor core configuration at cold conditions] [average U-235 enrichment of [4.5] weight percent], nominal 4.9
- b. keff 0.95 if fully flooded with unborated water, which includes an 3 allowance for uncertainties as described in [Section 9.1 of the FSAR], 1 under optimum moderation conditions 5.6.1.2 c. keff 0.98 if moderated by aqueous foam, which includes an b U allowance for uncertainties as described in [Section 9.1 of the FSAR], and c minimum 7.0 DOC M01 d. A nominal [6.50] inch center to center distance between fuel assemblies placed in storage racks.
in each storage rack row and a minimum 12.25 5.6.2 4.3.2 Drainage inch center to center distance between rows 2
The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation [185 ft]. 199 ft 4 in
5.6.3 4.3.3 Capacity 4006 1 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than [2845] fuel assemblies.
General Electric BWR/4 STS 4.0-2 Rev. 5.0 Amendment XXX 1 Hope Creek JUSTIFICATION FOR DEVIATIONS ITS 4.0, DESIGN FEATURES
- 1. The ISTS contains bracketed information and/or values that are generic to all General Electric BWR/4 vintage plants. The brackets are removed, and the proper plant specific information/value is provided. This is acceptable since the information/value is changed to reflect the current licensing basis.
- 2. Changes are made (additions, deletions, and/or changes ) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
- 3. ISTS 4.3.1.2.b is not included in ITS. The new fuel dry storage rack and vault design is precluded from being fully flooded with unborated water. Therefore, this requirement is not applicable to Hope Creek Generating Station. Subsequent numbering is revised to support this deviation.
Hope Creek Page 1 of 1 Determination of No Significant Hazards Considerations (NSHCs)
DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 4.0, DESIGN FEATURES
10 CFR 50.92 EVALUATION FOR LESS RESTRICTIVE CHANGE L01
PSEG Nuclear LLC (PSEG) is converting Hope Creek Generating Station (HCGS) to the Improved Technical Specifications (ITS) as outlined in NUREG-1433, Rev. 5, "Standard Technical Specifications, General Electric BWR/4 Plants." The proposed change involves making the Current Technical Specifications (CTS) less restrictive. Below is the description of this less restrictive change and the determination of no significant hazards considerations for conversion to NUREG-1433.
CTS describes details of a limited number of lead test assemblies associated with GE14i isotope test assemblies (ITAs). This change deletes detail of specific lead test assemblies that have not completed representative testing since the subject ITAs are no longer being placed in the HCGS reactor core.
The purpose of the CTS requirement was to limit loading of ITAs into the reactor core along with specifying other design details related to the ITAs. The modified fuel assemblies (i.e., GE14i ITAs) have been removed from the HCGS reactor core and PSEG does not anticipate future installation of these type of ITAs in the reactor core.
Therefore, this requirement is no longer applicable and is deleted. The Technical Specifications will continue to retain the allowance that a limited number of lead test assemblies that have not completed representative testing may be placed in non-limiting core regions, consistent with the NUREG-1433 and the guidance in Supplement 1 of Generic Letter (GL) 90-02, Alternative Requirements for Fuel Assemblies in Design Features Section of Technical Specifications, which were issued by the NRC. The requirements of 10 CFR 50.59 regarding conduc ting special tests or modifications remain applicable and are sufficient to ensure that a limited number of lead test assemblies placed in nonlimiting core regions will be properly evaluated prior to use.
PSEG has evaluated whether or not a significant hazards consideration is involved with the proposed change by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates detail of lead test assemblies that are no longer in use in the reactor core. Deleting information that is no longer applicable since the subject lead test assemblies are no longer in the reactor core does not impact probability or consequences of an accident previously evaluated. Fuel assembly design is not considered an accident initiator for any previously analyzed accident, and therefore, the change does not involve a significant increase in the probability of an accident previously evaluated. The requirements of 10 CFR 50.59 regarding conducting special tests or modifications remain applicable and are sufficient to ensure that a limited number of lead test assemblies placed in nonlimiting core regions will not result in an increase in probability or consequences of an accident previously evaluated.
Hope Creek Page 1 of 2 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 4.0, DESIGN FEATURES
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed change eliminates detail of lead test assemblies that are no longer in use in the reactor core. Deleting information that is no longer applicable since the subject lead test assemblies are no longer in the reactor core cannot create a new or different kind of accident. Additionally, the requirements of 10 CFR 50.59 regarding conducting special tests or modifications remain applicable, and are sufficient to ensure that a limited number of lead test assemblies placed in nonlimiting core regions will not create the possibility of a new or different kind of accident.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The proposed change eliminates detail of lead test assemblies that are no longer in use in the reactor core. Deleting information that is no longer applicable since the subject lead test assemblies are no longer in the reactor core has no impact on safety. Additionally, the requirements of 10 CFR 50.59 regarding conducting special tests or modifications remain applicable and are sufficient to ensure that a limited number of lead test assemblies placed in nonlimiting core regions will not result in a reduction in a margin of safety.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, PSEG concludes that the proposed change does not involve a significant hazards consideration under the s tandards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
Hope Creek Page 2 of 2