ML18102A159
| ML18102A159 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 06/04/1996 |
| From: | Dubay T Public Service Enterprise Group |
| To: | |
| Shared Package | |
| ML18102A157 | List: |
| References | |
| S96-049, S96-049-R00, S96-49, S96-49-R, NUDOCS 9606110110 | |
| Download: ML18102A159 (11) | |
Text
FORM NC.NA-AP.ZZ-0059-3 10CFR50.59 SAFETY EVALUATION Page 1of11 Revision 0 ID Numbers/Reference/Revision: PSE&G Calculation S-C-FHB-:r\\1DC-1616, Rev.0
Title:
Fuel Handling Accident Analysis Radiological Evaluation Applicability:
Salem 1
___ Salem) (Gas Turbine)
Salem 2
___ Hope Creek X
Common to Salem 1 & 2
___ Common to Hope Creek & Salem COMPLETION AND APPROVAL Preparer:
Peer Reviewer:
Approval:
Safety Evaluation No. ------=S=9-=-6__.;-0:.....:4c::...9 _____ _
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SORC Review:
G.M. Approval:
Safety Evaluation and associated documentation sent to Offsite Safety Review (OSR):
SORC Presente~~ 4(;, Cv*~11">hA Date: ll/4\\ ~(.,
1.0 10CFR50.59 REVIEW - 10CFR50.59 applies because:
1.1 The proposal changes the facility as described in the SAR.
YES x
NO 9606110110 960604
- PDR ADOCK 05000272 p
FORM NC.NA-AP.ZZ-0059-3 10CFR50.59 SAFETY EVALUATION Page 2of11 Revision 0 ID Numbers/Reference/Revision: PSE&G Calculation S-C-FHB-MDC-1616, Rev.0
Title:
Fuel Handling Accident Analysis Radiological Evaluation Pursuant to 10CFR50.90, LCR 93-02 was submitted to the USNRC on April 28, 1993 to amend the facility license as related to increased spent fuel capacity for Salem Units 1 and
- 2. The LCR contained a Licensing Report which, among other things, provided a revised radiological evaluation of a postulated fuel handling accident. On May 4, 1994, LCR 93-02 was approved by the NRC in Amendments 151 and 131 to the facility operating license. Marked-up UFSAR pages were submitted to the NRC on April 10, 1996 in accordance with Amendments 151 and 131. The remainder of the Amendment 151/131 changes are scheduled for UFSAR update as part of the June 1996 update required by 10CFR50.71(e). As this safety evaluation and calculation S-C-FHV-MDC-1616 address clarifications of the radiological evaluation in the Licensing Report submitted with LCR 93-02, it constitutes a change to the facility as described in the UFSAR.
1.2 The proposal changes procedures as described in the SAR.
YES NO x
Neither calculation S-C-FHV-MDC-1616 nor this safety evaluation impact upon any procedures described or otherwise referenced in the UFSAR, including the Artificial Island Emergency Plan or Emergency Plan Implementing Procedures.
1.3 The proposal involves a test or experiment not described in the SAR.
YES NO x
Calculation S-C-FHV-MDC-1616 provides clarifications and amplifications of postulated fuel handling accidents. No test or experiment is involved.
2.0 LICENSING BASIS DOCUMENTATION 2.1 UFSAR REVISION DETERMINATION - Does the proposal require a UFSAR change?
YES x
NO UFSAR Change Notice No.:,_. ___
___:::_96=---=8-=-5 ___ _
FORM NC.NA-AP.ZZ-0059-3 10CFR50.59 SAFETY EVALUATION Page 3of11 Revision 0 ID Numbers/Reference/Revision: PSE&G Calculation S-C-FHB-l\\.IDC-1616, Rev.0
Title:
Fuel Handling Accident Analysis Radiological Evaluation 2.2 TECHNICAL SPECIFICATION REVISION DETERMINATION - Does the proposal require a Technical Specification change?
YES NO x
As further described in this safety evaluation, calculation S-C-FHV-l\\.IDC-1616 confirms the basis upon which LCR 93-02 was approved by the USNRC in Amendments 151 and 131.
Since the subject basis is confirmed herein, no changes to the Technical Specifications are needed.
The following Technical Specification sections were reviewed to make this determination:
3/4.3.3 3/4.9
3.0 DESCRIPTION
Radiation Monitoring Instruments Refueling Operations 3/4.9.3 Decay Time 3/4.9.12 Fuel Handling Area Ventilation System 3.1 Describe the modification or activity being evaluated and its expected effects.
In letter LR-N96147 dated May 29, 1996, PSE&G identified that the current licensing basis fuel handling accident (FHA) analysis for Salem 1 and 2 contained non-conservative assumptions when compared to NRC Safety Guide 25. Letter LR-N96155 dated May 31, 1996, documented a commitment to reanalyze the FHA to address this matter.
Calculation S-C-FHV-l\\.IDC-1616 was performed to satisfy this commitment. This safety evaluation reconciles differences between calculation S-C-FHV-l\\.IDC-1616, the Licensing Report submitted in LCR 93-02 (FHA reanalysis following spent fuel pool high density rack modification), and the NRC' s evaluation in SER Amendment 151 and 131. This proposal makes no physical changes to any plant system, equipment, or component and makes no changes to plant procedures or other controls.
PSE&G reanalyzed the FHA to support increasing spent fuel storage capacity at Salem Units 1 and 2.
That analysis (called the LCR 93-02 analysis throughout this safety
L FORM NC.NA-AP.ZZ-0059-3 10CFR50.59 SAFETY EVALUATION Page 4of11 Revision 0 ID Numbers/Reference/Revision: PSE&G Calculation S-C-FHB-1IDC-1616, Rev.O
Title:
Fuel Handling Accident Analysis Radiological Evaluation evaluation) was submitted to the NRC on April 28, 1993 as part of a license amendment request (LCR 93-02). PSE&G's request was reviewed and approved on May 4, 1994 as License Amendments 151 and 131 to the unit operating licenses. The spent fuel pool high density rack modifications have been completed at Salem Units 1and2.
The dose assessment associated with the revised FHA analysis was recently reviewed as part of an ongoing project to reconstitute and update specific dose assessments. That review identified that the spent fuel pool decontamination factor and charcoal filtration efficiencies assumed in the analysis were inconsistent with assumptions used in NRC Safety Guide 25 and UFSAR Section 15.
A thyroid dose at the site boundary of approximately 14.6 rem results when these input values are corrected to meet the values specified in the Safety Guide and UFSAR Section 15.
While this value remains well within the acceptance criteria specified in NUREG-0800 (75 rem thyroid dose), this dose exceeds the value calculated by the NRC in the Safety Evaluation Report for Amendments 151 and 131.
A new FHA analysis has been documented in PSE&G Calculation S-C-FHV-1IDC-1616 (called the new analysis throughout the remainder of this evaluation) to confirm that the plant remains in conformance with the NRC SER. Conservative assumptions incorporated to bound future changes to the plant design and licensing basis have been revised to reflect current licensing and design limits. Revised input assumptions have been generated and compared to NRC Safety Guide 25.
Deviations from the Safety Guide have been evaluated for conformance with the Salem licensing basis. The technical basis for each deviation has also been validated, as enumerated in this safety evaluation.
The final calculation continues to be conservative relative to assumed X/Q values, assumed versus actual charcoal filter efficiencies, and takes no credit for pool decontamination for organic iodine The new FHA analysis for Salem Units 1 and 2 documents a thyroid dose of 13.2 rem at the Exclusion Area Boundary. This value is consistent with that reported in the NRC SER for Amendments 151 and 131, and therefore is in compliance with the current licensing basis for the Salem Units. (Note: The NRC increased its original conservative analysis of 11 rem by 20% to account for extended fuel bum-up -- rounded down to 13 rem [ 11 x 1.2
= 13.2]. Hence, the actual licensing basis value is 13.2 rem.) In view of the above, and as further described herein, an unresolved safety question does not exist.
L FORM NC.NA-AP.ZZ-0059-3 10CFR50.59 SAFETY EVALUATION Page 5of11 Revision 0 ID Numbers/Reference/Revision: PSE&G Calculation S-C-FHB-MDC-1616, Rev.0
Title:
Fuel Handling Accident Analysis Radiological Evaluation 3.2 Identify the parameters and systems afTected by the change.
Calculation or analysis parameters which are changed and/or clarified by the PSE&G calculation are identified below:
Radial Peaking Factor Fuel Bum-up Pool WaterDF Charcoal Filter DF Reactor Power Average Specific Power Fuel Enrichment LCR 93-02 Analysis 1.70 65,000 Mwd/mtU 500 100 105.5%
40.45 kw/kgU 4.5 wt%
S-C-FHV-MDC-1616 1.65 60,000 Mwd/mtU 222 4.73 102%
18.03 Mw/Assembly*
5.0 wt%
- Average specific power was expressed in different units in the two analyses, but for the reactor power used in each, the two are equivalent.
3.3 Identify credible failure modes associated with the change.
No new failure modes are introduced by this proposal. This proposal makes no physical change to the system, system operation, maintenance, or testing.
3.4 Provide references to location of information used for the safety evaluation.
- a.
UFSARSection 9.1.2, Spent Fuel Storage
- b.
UFSAR Section 9.1.4, Spent Fuel Handling
- c.
UFSAR Chapter 15.4.6, Fuel Handling Accident Analyses
- d.
Technical Specification 3/4.9.3, Decay Time
- e.
Technical Specification 3/4.9.12, Fuel Handling Area Ventilation System f
10CFR50, Appendix A, Criterion 19, Control Room
- g.
Salem SER (including Supplements 1 through 6)
- h.
Artificial Island Emergency Plan (AIEP)
- 1.
Emergency Plan Implementing Procedure (EPIP)
J.
Regulatory Guide 1.25 (Safety Guide 25)
FORM NC.NA-AP.ZZ-0059-3 10CFR50.59 SAFETY EVALUATION Page 6of11 Revision 0 ID Numbers/Reference/Revision: PSE&G Calculation S-C-FHB-MDC-1616, Rev. 0
Title:
Fuel Handling Accident Analysis Radiological Evaluation 3.5 Other discussions, if applicable.
None.
4.0 USO DETERMINATION -Is an Unreviewed Safety Question (USQ) involved?
4.1 Which anticipated operational transients or postulated design basis accidents previously evaluated in the SAR are considered applicable to this proposal?
4.2 A fuel handling accident, as analyzed and discussed in UFSAR Chapter 15.4.6, the Salem SER, and SER Amendment 151/131 is the only design basis accident applicable to this proposal. Table 15.1 of the original USNRC SER provides the results for the Salem fuel handling accident analysis. These values were revised following the re-rack modification in 1995. These values are given in the USNRC SER (Amendment 151/131, 5-4-94), as shown below:
Two Hour Exclusion Low Population Boundary Zone Thyroid Whole Body Thyroid Whole Body Original Analysis 11 rem 1 rem 1 rem
<l rem Re-rack analysis 13 rem 1 rem 1.2 rem
<l rem NUREG800 Guidelines 75 rem 6rem 75 rem 6 rem 10CFRlOO Limits 300 rem 25 rem 300 rem 25 rem May the proposal:
- a.
Increase the probability of an accident previously evaluated in the SAR?
YES NO x
FORM NC.NA-AP.ZZ-0059-3 10CFR50.59 SAFETY EVALUATION Page 7of11 Revision 0 ID Numbers/Reference/Revision: PSE&G Calculation S-C-FHB-MDC-1616, Rev.0
Title:
Fuel Handling Accident Analysis Radiological Evaluation This safety evaluation documents the acceptability of a design basis calculation which predicts or estimates the consequences of a design basis fuel handling accident.
No physical changes are made to any system, equipment, or component. No changes or modifications are made to plant procedures. As such, the probability of an accident is in no way affected.
- b.
Increase the consequences of an accident previously evaluated in the SAR?
YES NO x
Calculation S-C-FHV-MDC-1616, after making changes to the analysis submitted with LCR 93-02 (as identified in Paragraph 3.2 above and further described in this section),
concludes thyroid dose at the Exclusion Area Boundary in a design basis fuel handling accident is 13.2 rem. This dose is the same value determined by the USNRC staff in Amendment 151/131 (the NRC increased its original conservative analysis of 11 rem by 20% to account for extended fuel bum-up -- rounded down to 13 rem [11x1.2 = 13.2]).
Since 13 (or 13.2) rem was identified in the NRC staff approval of Amendments 151 and 131, there are no increased accident consequences as a result of the new PSE&G analysis.
As listed in Paragraph 3.2 above, the PSE&G calculation made changes, modifications, or clarific~tions to the LCR 93-02 analysis as described below. The PSE&G calculation conforms to the current licensing basis of the plant and confirms the conservative analysis conducted by the USNRC staff.
- 1.
The radial peaking factor was changed from 1.70 to 1.65 because the 1.7 factor was used to bound future fuel management, while the 1.65 is the minimum allowed by Regulatory Guide 1.25 and still bounds current radial peaking.
- 2.
Fuel bum-up was changed from 65,000 Mwd/mtU to 60,000 Mwd/mtU. In a letter to the USNRC dated April 7, 1994, PSE&G withdrew its request to the NRC to extend fuel bum-up to 65,000 Mwd/MtU, reducing the request to 60,000 Mwd/mtU. The LCR 93-02 analysis, which was prepared in 1993, was performed at the higher bum-up.
FORM NC.NA-AP.ZZ-0059-3 10CFR50.59 SAFETY EVALUATION Page 8of11 Revision 0 ID Numbers/Reference/Revision: PSE&G Calculation S-C-FHB-MDC-1616, Rev.O
Title:
- 3.
- 4.
- 5.
6 Fuel Handling Accident Analysis Radiological Evaluation The LCR 93-02 analysis assumed an effective pool decontamination factor of 500 and referenced the justification provided in UFSAR Chapter 15.4.6. The implied assumption was that 500- was an overall effective decontamination factor.
Assuming the 500 decontamination factor is only applicable to elemental iodine, a more conservative decontamination factor of 222 was applied.
The LCR 93-02 analysis used a charcoal decontamination factor of 100. The Salem licensing basis requires 90% removal efficiency for inorganic iodine and 70% for organic iodine, which results in an overall decontamination factor of 4. 73.
The LCR 93-02 analysis used reactor power of 105.5%, anticipating future power up-rate for Salem. PSE&G used the present conservative value of 102%.
The LCR 93-02 analysis was performed at both 4.5 wt% enrichment and 5.0 wt%,
and determined the 4.5 wt% case to be bounding.
Recent analyses by Westinghouse have determined the 5.0 wt% case to be bounding. Consequently, the new PSE&G analysis used the 5.0 wt% value.
4.3 What malfunctions of equipment important to safety that were previously evaluated in the SAR are considered applicable to the proposal?
This safety evaluation supports a calculation which reanalyzes the Salem design basis fuel handling accident. It makes no physical changes to any plant systems, equipment, or components, and makes no changes to procedures or other controls. As such, there are no malfunctions of equipment important-to-safety which are applicable to this proposal.
Physical changes to the plant which resulted from the modification to increase spent fuel capacity at Salem and precipitated LCR 93-02 are not directly tied to fuel handling accidents (increased fuel bum-up already being included in the plant licensing basis). Also, physical changes from the rerack modification have been separately evaluated by the 10CFRS0.59 process.
4.4 May the proposal:
- a.
Increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR?
FORM NC.NA-AP.ZZ-0059-3 10CFRS0.59 SAFETY EVALUATION Page 9of11 Revision 0 ID Numbers/Reference/Revision: PSE&G Calculation S-C-FHB-MDC-1616, Rev.O
Title:
Fuel Handling Accident Analysis Radiological Evaluation YES NO x
Since there is no equipment important-to-safety applicable to this proposal, there can be no increase in the probability of failure of this equipment.
- b.
Increase the consequences of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR?
YES NO x
Since there is no equipment important-to-safety applicable to this proposal, there can be no increase in the consequences of failure of this equipment.
4.5 May the proposal:
- a.
Create the possibility of an accident of a different type from any previously evaluated in the SAR?
YES NO x
This safety evaluation supports a calculation which evaluates the design basis fuel handling accident described in UFSAR Chapter 15.4.6. There are no physical changes made to the plant of any kind. Consequently, no accident of a different type than any evaluated in the UFSAR can be created.
- b.
Create the possibility of a malfunction of a different type from any previously evaluated in the SAR?
YES NO x
This safety evaluation supports a calculation which evaluates the design basis fuel handling accident described in UFSAR Chapter 15.4.6. There are no physical changes made to the plant of any kind. Consequently, no malfunction of a different type than any evaluated in the UFSAR can be created
FORM NC.NA-AP.ZZ-0059-3 10CFR50.59 SAFETY EVALUATION Page 10of11 Revision 0 ID Numbers/Reference/Revision: PSE&G Calculation S-C-FHB-.MDC-1616, Rev.O
Title:
Fuel Handling Accident Analysis Radiological Evaluation 4.6 Does the proposal reduce the-margin of safety as defined in the basis for any Technical Specifications?
SER Amendments 151 and 131 make the following statement relative to the acceptability of the Salem FHA analysis:
"The staff concludes that the only potential increased doses resulting from the fuel handling accidents with extended bumup fuel is the thyroid doses; these doses remain well within the dose limits given in NUREG-0800 (Reference 5) and are, therefore, acceptable."
The margin of safety as defined in the bases of technical specifications associated with potential fuel handling accidents depends upon maintaining resulting doses (thyroid and whole body) well within the dose limits of 10CFR PartlOO, as further defined by the guidelines ofNUREG-0800. As indicated above, the USNRC staff found the Salem EAB and LPZ doses well within the Guidelines ofNUREG-0800.
Not only does calculation S-C-FHV-.MDC-1616 of this safety evaluation remain well within the Guidelines of NUREG-0800, but the calculation results essentially equal the NRC staff analysis of 13.2 rem at the exclusion area boundary. Consequently, the margin of safety, as defined in the bases of the technical specifications, has not been reduced.
5.0 CONCLUSION
If ALL answers to Section 4 are NO, the proposal does NOT involve a USQ.
If ANY answer to Section 4 is YES, the proposal involves a USQ.
Is a USQ Involved?
YES NO__x_
FORM NC.NA-AP.ZZ-0059-3 10CFR50.59 SAFETY EVALUATION Page 11of11 Revision 0 ID Numbers/Reference/Revision: PSE&G Calculation S-C-FHB-MDC-1616, Rev.O
Title:
Fuel Handling Accident Analysis Radiological Evaluation If a USQ is involved, refer to NC.NA-AP.ZZ-0035(Q) and obtain assistance from Licensing for additional processing.