RNP-RA/02-0118, Response to Request for Additional Information on Technical Specifications Change Re Selective Implementation of Alternative Radiological Source Term

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Response to Request for Additional Information on Technical Specifications Change Re Selective Implementation of Alternative Radiological Source Term
ML022310271
Person / Time
Site: Robinson Duke Energy icon.png
Issue date: 08/14/2002
From: Fletcher B
Carolina Power & Light Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RNP-RA/02-0118, TAC MB4632
Download: ML022310271 (40)
v • d • e


Text

10 CFR 50.90 S CP&L A ProgresmEnergy jVnpar,,

Serial: RNP-RA/02-0118 AUG 1 4 2002 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261/LICENSE NO. DPR-23 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON TECHNICAL SPECIFICATIONS CHANGE REGARDING SELECTIVE IMPLEMENTATION OF ALTERNATIVE RADIOLOGICAL SOURCE TERM (TAC NO. MB4632)

Ladies and Gentlemen:

On March 13, 2002, in accordance with the provisions of the Code of Federal Regulations, Title 10 (10 CFR), Part 50.90, Carolina Power & Light (CP&L) Company submitted a request for an amendment to the Technical Specifications (TS) for H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, that would modify the TS requirements for movement of irradiated fuel and performing core alterations.

The basis for the proposed change is a reanalysis of the limiting Fuel Handling Accident (FRA) using an alternative source term (AST) in accordance with 10 CFR 50.67 and Regulatory Guide 1.183, Revision 0, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," dated July 2000. Therefore, the submittal dated March 13, 2002, also requested NRC approval of the selective implementation of AST methodology for the HBRSEP, Unit No. 2, design basis FHA.

Discussions with the NRC staff, including conference calls on July 22 and July 26, 2002, have identified that additional information is needed to complete review of the proposed TS amendment. The purpose of this letter is to provide the requested information.

Attachment I provides an Affirmation as required by 10 CFR 50.30(b).

Attachment II provides the HBRSEP, Unit No. 2, responses to the NRC staff's requests for additional information.

Attachment III provides a supplement to the proposed license amendment, which incorporates changes based on guidance contained in the generic change Technical Specifications Task Force (TSTF)-51, Revision 2. This supplement also provides correction of the control room dose analysis and meteorological data as identified in response to NRC staff information requests in Attachment II to this letter.

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United States Nuclear Regulatory Commission Serial: RNP-RA/02-0118 Page 2 of 2 Attachment IV provides a revised markup of the proposed TS pages.

Attachment V provides the revised, retyped pages for the proposed TS.

A CD-ROM is provided, as an enclosure to this submittal, with the corrected meteorological data that was used for the ARCON96 dispersion factor calculations.

In accordance with 10 CFR 50.91(b), CP&L is providing the State of South Carolina with a copy of this response.

As described within the HBRSEP, Unit No. 2, submittal dated March 13, 2002, CP&L has requested approval of the proposed license amendment by September 1, 2002, with the amendment being implemented within 60 days of approval. This requested approval date was selected to support activities associated with Refueling Outage (RO) - 21, which is scheduled to begin on October 12, 2002.

If you have any questions concerning this matter, please contact Mr. C. T. Baucom.

Sincerely, B. L. Fletcher III Manager - Regulatory Affairs CTB/cac Attachments:

I. Affirmation II. Response to Request for Additional Information on Technical Specifications Change Regarding Selective Implementation of Alternative Radiological Source Term III. Supplement to Request for Technical Specifications Change Regarding Selective Implementation of Alternative Radiological Source Term IV. Markup of Technical Specifications Pages V. Retyped Technical Specifications Pages

Enclosure:

CD-ROM with corrected meteorological datasets c: Mr. L. A. Reyes, NRC, Region II Mr. H. J. Porter, Director, Division of Radioactive Waste Management (SC)

Mr. R. M. Gandy, Division of Radioactive Waste Management (SC)

Mr. R. Subbaratnam NRC Resident Inspector, HBRSEP Attorney General (SC)

United States Nuclear Regulatory Commission Attachment I to Serial: RNP-RA/02-0118 Page 1 of 1 AFFIRMATION The information contained in letter RNP-RA/02-0118 is true and correct to the best of my information, knowledge and belief; and the sources of my information are officers, employees, contractors, and agents of Carolina Power and Light Company. I declare under penalty of perjury that the foregoing is true and correct.

Executed on: AUG 1 4 2002 Vice President, RSEP, Unit No. 2

United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/02-0118 Page 1 of 4 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON TECHNICAL SPECIFICATIONS CHANGE REGARDING SELECTIVE IMPLEMENTATION OF ALTERNATIVE RADIOLOGICAL SOURCE TERM NRC Information Request

1. During a conference call on July 22, 2002, between NRC staff personnel and H. B.

Robinson Steam Electric Plant (HBRSEP), Unit No. 2, the following aspects of the proposed Technical Specifications (TS) changes for selective implementation of the alternative source term (AST) were discussed:

(a) The proposed deletion of TS 3.9.3, "Containment Penetrations," and TS 3.9.7, "Containment Purge Filter System," are not consistent with the approved Technical Specifications Task Force (TSTF) - 51, Revision 2.

(b) The proposed changes to TS 3.3.7, "Control Room Emergency Filtration System (CREFS) Actuation Instrumentation," would create the need for operator action in the event of a fuel handling accident that involves a radiological release.

Response

TSTF-5 1, Revision 2, states that the justification associated with this generic change is based on performing analyses that assume a longer decay period to take advantage of the reduced radionuclide inventory available for release in the event of a fuel handling accident (FHA). The March 13, 2002, license amendment submittal stated it was not appropriate for HBRSEP, Unit No. 2, to incorporate TSTF-5 1. After further review of TSTF-5 1, Revision 2, and discussions with NRC staff, it has been concluded that TSTF-5 1, Revision 2, can be utilized as guidance to modify the Technical Specifications changes proposed by the March 13, 2002, license amendment submittal. Attachment Inl provides the revised proposed Technical Specifications changes, as modified to incorporate TSTF-51, Revision 2.

Additionally, the Technical Specifications changes, as proposed herein, do not remove any TS operability requirements for automatic functions that would require substitution of manual operator action for the automatic function to mitigate design basis accidents and events.

United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/02-0118 Page 2 of 4 NRC Information Request

2. During a conference call on July 22, 2002, between NRC staff personnel and HBRSEP, Unit No. 2, analysis assumptions regarding unfiltered control room inleakage were discussed. During the postulated FHA's first hour, the analysis assumes an additional unfiltered air inleakage rate to the control room of 300 cfm. Following the switch to the emergency pressurization mode, the analysis assumes an unfiltered air inleakage rate of 230 cfm due to actions taken to preclude potential inleakage from the Hagan Room.

Clarification was sought regarding the basis for these analysis assumptions.

Response

By letter dated May 10, 2002, HBRSEP, Unit No. 2, submitted a license amendment request regarding full implementation of an alternative source term. Within this letter, CP&L committed to perform a leak rate test on the HBRSEP, Unit No. 2, Control Room envelope prior to implementation of the changes requested in the May 10, 2002, submittal.

Upon completion of this testing, CP&L will provide the results in a supplement to that submittal. In addition to the results of the testing, CP&L will also consider the following points in relation to this testing:

" A single value for unfiltered Control Room air inleakage will be established with a basis (tracer gas testing),

" In the event that the analyses contained in this submittal do not bound the new established value, the applicable analyses will be revised to bound the new established value,

" The testing and reanalysis, if required, is intended to demonstrate compliance with the Control Room dose acceptance criteria of 10 CFR 50, Appendix A, GDC-19. In the event compliance with the GDC-19 dose acceptance criteria cannot be supported by the current licensing basis, a comprehensive corrective action plan to restore compliance with the GDC-19 dose acceptance criterion will be developed.

NRC Information Reauest

3. During a conference call on July 26, 2002, between NRC staff personnel and HBRSEP, Unit No. 2, the NRC reviewer observed that the lower wind speeds were often substantially less than upper wind speeds. Clarification was sought regarding the site characteristics for the meteorological tower.

Response

As described in Updated Final Safety Analysis Report (UFSAR) Section 2.3.3.1, the meteorological tower is located approximately 0.53 miles north of the containment building. The base of the tower is at the plant grade level of approximately 225 feet above mean sea level. An environmentally controlled shelter, which houses recording instruments, data collection devices, and remote data access equipment, is located near the tower, perpendicular to the prevailing wind flow to minimize air trajectory deviations.

United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/02-0118 Page 3 of 4 The guyed, open-latticed tower supports upper and lower levels of instrumentation. The upper level instrumentation includes a wind sensor (for wind direction, wind speed, and wind variance) and redundant temperature sensors used in the differential temperature monitoring. The lower level instrumentation includes a wind sensor, redundant temperature sensors for the differential temperature monitoring and ambient temperature measurement, and a relative humidity sensor. The wind sensors are mounted on 12-foot booms oriented perpendicular to the general NE-SW prevailing wind flow to minimize tower shadow effects. The temperature sensors and the relative humidity sensor are housed in aspirated shields mounted on eight-foot booms. The other meteorological parameters monitored by the system (not located on the tower) are solar radiation, barometric pressure, and precipitation. These sensors are located near ground level, near or at the equipment shelter.

The meteorological tower is situated in a relatively flat field. Trees are situated at the edge of the field. The proximity of trees to the tower ranges from approximately 200 - 250 feet.

Trees near the tower are approximately 20 - 40 feet high. There is an approximately 50-foot high coal pile located approximately 300 feet north of the containment (i.e.,

approximately 2000 feet from the tower).

NRC Information Request

4. During a conference call on July 26, 2002, between NRC staff personnel and HBRSEP, Unit No. 2, it was identified that invalid data fields were blank in the submitted datasets. It was noted that a "9" value is more appropriate for these data fields to assure the proper results are obtained.

Response

The blank fields in the datasets for periods 1988 - 1991 and 1992 - 1996 have been filled with "9's" (i.e., wind speed - 9999, wind direction - 999, stability - 99) and the lower wind speed data was corrected in the dataset representative of 1992 - 1996 period, as discussed in response to NRC Information Request No. 5 (below). The datasets were then used to re-perform the on-site dispersion factor calculations using ARCON96. These changes, along with the correction of the data as described in response to NRC Information Request No. 5, have resulted in a very slight increase in the dispersion factor and resultant dose for the Control Room dose consequence analysis. The calculated Control Room doses remain within the applicable limits of 10 CFR 50.67.

NRC Information Request

5. During a conference call on July 26, 2002, between NRC staff personnel and HBRSEP, Unit No. 2, it was identified that lower wind speed data during the years 1992 through 1996 appears to not exceed 9.9 mph.

United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/02-0118 Page 4 of 4

Response

A formatting error occurred during the manipulation of the meteorological data to create the datasets needed for performing the calculations to determine the on-site dispersion factors using the ARCON96 computer code. The formatting error, which was the inadvertent deletion of the ten's digit from the lower wind speed data, has been corrected, including insertion of "9's" into the blank fields as discussed in response to NRC Information Request 4. The on-site dispersion factors were recalculated using ARCON96.

These changes have resulted in a very slight increase in the dispersion factor and resultant dose for the Control Room dose consequence analysis. The FHA dose calculations and analysis have been revised to incorporate the new on-site dispersion factors. The calculated Control Room doses remain within the applicable limits of 10 CFR 50.67.

The corrected meteorological datasets are also provided on a CD-ROM enclosed with this submittal.

NRC Information Request

6. During a conference call on July 26, 2002, between NRC staff personnel and HBRSEP, Unit No. 2, clarification was sought regarding special assumptions associated with the ARCON96 dispersion factor determinations.

Response

For the onsite receptor locations, the new dispersion factors were developed using the ARCON96 computer code ("Atmospheric Relative Concentrations in Building Wakes,"

NUREG/CR-6331, Rev. 1, May 1997, RSICC Computer Code Collection No. CCC-664).

New guidance, which supersedes the NUREG/CR-6331 recommendations for using certain default parameters as input, contained in NRC Draft Regulatory Guide DGI 111, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," December 2001, has been implemented.

Specifically, the following changes from the default values were made:

i. Surface roughness length, m: A value of 0.2 is used in lieu of the default value of 0.1.

ii. Averaging sector width constant: A value of 4.3 is used in lieu of the default value of 4.0.

In these calculations, the dispersion factors are based on ground-level, point-source releases. Diffusion-based dispersion factors were not used in these calculations.

United States Nuclear Regulatory Commission Attachment III to Serial: RNP-RA/02-0118 Page 1 of 10 SUPPLEMENT TO REQUEST FOR TECHNICAL SPECIFICATIONS CHANGE REGARDING SELECTIVE IMPLEMENTATION OF ALTERNATIVE RADIOLOGICAL SOURCE TERM Description of Current Condition The "Description of Current Condition" contained in the original submittal dated March 13, 2002, is not changed by this supplement.

Description of the Proposed Changes The "Description of the Proposed Changes" contained in the original submittal dated March 13, 2002, is revised in its entirety, as follows:

Carolina Power and Light (CP&L) Company proposes to revise the H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, licensing basis to selectively implement an alternative source term (AST), as described in Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," through reanalysis of the radiological consequences of a Fuel Handling Accident (FHA). As part of the selective implementation of this AST, the total effective dose equivalent (TEDE) acceptance criterion of 10 CFR 50.67(b)(2) replaces the previous whole body and thyroid dose guidelines of 10 CFR 100.11 and General Design Criterion (GDC)-19 of 10 CFR Part 50, Appendix A, which (based upon this application) is limited to the FHA only.

Changes to the HBRSEP, Unit No. 2, TS, which are based on the approved Standard Technical Specifications (TS) Change Traveler, Technical Specification Task Force (TSTF)-5 1, Revision 2, are proposed, as follows:

TS 3.3.6 is revised to delete "(a) During CORE ALTERATIONS" and "(c) During Purging" from the "Other Specified Conditions" associated with Table 3.3.6-1, "Containment Ventilation Isolation Instrumentation." The "Other Specified Conditions" stated as "During movement of irradiated fuel assemblies within containment" is being restated as "During movement of recently irradiated fuel assemblies within containment."

The notes for Table 3.3.6-1 are re-lettered, as appropriate. The Applicable MODES "1, 2, 3, 4," are also added to Function 3.a, "Containment Radiation Gaseous," and 3.b, "Containment Radiation Particulate."

These changes are based on TSTF-51, Revision 2, except for the deletion of the

"-(c) During Purging" from the "Other Specified Conditions." Deletion of "(c) During Purging" from the "Other Specified Conditions" is justified, as described in the "Safety Assessment," and establishes consistency of the HBRSEP, Unit No. 2, TS with the NUREG-1431, Revision 2, "Standard Technical Specifications - Westinghouse Plants,"

and TSTF-51, Revision 2.

United States Nuclear Regulatory Commission Attachment III to Serial: RNP-RA/02-01 18 Page 2 of 10

" TS 3.3.7 is revised to delete "During CORE ALTERATIONS" from the Applicability.

Following this change, TS 3.3.7 will be Applicable in MODES 1, 2, 3, 4, and during movement of irradiated fuel assemblies. Condition D of TS 3.3.7 is also modified to remove the Condition and Required Action pertaining to "CORE ALTERATIONS."

This change is based on TSTF-5 1, Revision 2, except the option to insert the word "recently" into the Applicability associated with the movement of irradiated fuel is not being used. The control room emergency filtration system (CREFS) instrumentation provides automatic operation of the CREFS. Therefore, the proposed change to the Applicability of TS 3.3.7 will maintain consistency with the proposed change to the Applicability of TS 3.7.9.

"* TS 3.7.9 is revised to delete "During CORE ALTERATIONS" from the Applicability.

Following this change, TS 3.7.9 will be Applicable during MODES 1, 2, 3, and 4, and during movement of irradiated fuel assemblies. Condition C of TS 3.7.9 is revised for consistency with the change in Applicability, and Required Action C.2.1 is deleted.

Required Action C.2.2 is renumbered as an editorial change. Condition D of TS 3.7.9 is revised for consistency with the change in Applicability, and Required Action D. 1 is deleted. Required Action D.2 is renumbered as an editorial change.

These changes are based on TSTF-5 1, Revision 2, except the option to insert the word "recently" into the Applicability associated with the movement of irradiated fuel is not being used. As described in the "Safety Assessment" for this license amendment request, the CREFS is used in the analysis that is provided to demonstrate compliance with 10 CFR 50, Appendix A, GDC-19, for the FHA. Therefore, the Applicability requirements for CREFS are not limited to recently irradiated fuel.

" TS 3.7.10 is revised to delete "During CORE ALTERATIONS" from the Applicability.

Following this change, TS 3.7.10 will be Applicable during MODES 1, 2, 3, and 4, and during movement of irradiated fuel assemblies. Condition C of TS 3.7.10 is revised for consistency with the change in Applicability, and Required Action C.2.1 is deleted.

Required Action C.2.2 is renumbered as an editorial change. Condition D of TS 3.7.10 is revised for consistency with the change in Applicability, and Required Action D. 1 is deleted. Required Action D.2 is renumbered as an editorial change.

These changes are based on TSTF-51, Revision 2, except the option to insert the word "recently" into the Applicability associated with the movement of irradiated fuel is not being used. The control room emergency air temperature control (CREATC) provides required temperature control for the control room. Therefore, the proposed change to the Applicability of TS 3.7.10 will maintain consistency with the proposed change to the Applicability of TS 3.7.9.

" TS 3.9.3 is revised to delete "During CORE ALTERATIONS" from the Applicability.

The Applicability stated as "During movement of irradiated fuel assemblies within containment" is being restated as "During movement of recently irradiated fuel assemblies within containment." Condition A of TS 3.9.3 is revised for consistency with the change in Applicability, and Required Action A. 1 is deleted. Required Action A.2 is

United States Nuclear Regulatory Commission Attachment III to Serial: RNP-RA/02-0118 Page 3 of 10 renumbered as an editorial change. These changes are based on TSTF-51, Revision 2.

" TS 3.9.6 is revised to delete "During CORE ALTERATIONS, except during latching and unlatching of control rod drive shafts" from the Applicability. Following this change, TS 3.9.6 will be Applicable during movement of irradiated fuel assemblies within containment. Condition A of TS 3.9.6 is revised by the deletion of Required Action A. 1.

Required Action A.2 is renumbered as an editorial change. These changes are based on TSTF-51, Revision 2.

" TS 3.9.7 is revised to delete "During CORE ALTERATIONS" from the Applicability.

The Applicability stated as "During movement of irradiated fuel assemblies in containment" is being restated as "During movement of recently irradiated fuel assemblies in containment." Following this change, TS 3.9.7 will be Applicable during movement of recently irradiated fuel assemblies in containment. Condition A of TS 3.9.7 is revised by the deletion of Required Action A.2. 1. Required Action A.2.2 is changed, based on the change to the Applicability, to suspend movement of recently irradiated fuel assemblies within the containment. Required Action A.2.2 is also renumbered as an editorial change.

HBRSEP, Unit No. 2, TS 3.9.7 does not correspond to any specification in NUREG 1431, Revision 2, "Standard Technical Specifications - Westinghouse Plants." At the time of conversion to Improved Standard Technical Specifications (ISTS), the HBRSEP, Unit No. 2, TS contained requirements associated with the Containment Purge Filter System. These plant-specific TS requirements were based on the conditions of the FHA analysis in effect at that time. The proposed changes to TS 3.9.7 are based on the application of standard technical specifications (STS) generic change TSTF-51, Revision 2.

The following summary describes the TSTF-5 1, Revision 2, changes that were determined to be not applicable to this license amendment request:

" The TSTF-51, Revision 2, identifies NUREG-1431, Section 3.3.8, "FBACS Actuation Instrumentation," as an affected section of the STS. There is no corresponding section in the HBRSEP, Unit No. 2, TS. The Fuel Building Air Cleanup System (FBACS) is required to be operable and in operation during the movement of irradiated fuel assemblies in the fuel building, in accordance with HBRSEP, Unit No. 2, TS 3.7.11.

Therefore, the changes described in TSTF-51, Revision 2, for NUREG-1431, Section 3.3.8, are not applicable.

"* The TSTF-51, Revision 2, identifies NUREG-1431, Section 3.7.13, "Fuel Building Air Cleanup System (FBACS)," as an affected section of the ISTS. The corresponding section in the HBRSEP, Unit No. 2, TS is 3.7.11, "Fuel Building Air Cleanup System (FBACS)." As described in the "Safety Assessment" for this license amendment request, the FBACS is used in the analysis that is provided to demonstrate compliance with the applicable dose acceptance criteria for the FHA. Therefore, the changes described in TSTF-51, Revision 2, for NUREG-1431, Section 3.7.11, are not applicable.

United States Nuclear Regulatory Commission Attachment III to Serial: RNP-RA/02-0118 Page 4 of 10

  • The TSTF-51, Revision 2, identifies NUREG-1431, Section 3.8.2, "AC Sources Shutdown," as an affected section of the ISTS. The corresponding section in the HBRSEP, Unit No. 2, TS is 3.8.2, "AC Sources - Shutdown." The AC Sources support operation of systems that are used for mitigation of the FHA. Therefore, the exceptions to TSTF-5 1, Revision 2, based on the use of mitigation features in the FHA analysis, such as CREFS and FBACS, make the changes described in TSTF-51, Revision 2, for NUREG-1431, Section 3.8.2, not applicable.
  • The TSTF-51, Revision 2, identifies NUREG-1431, Section 3.8.8, "Inverters Shutdown," as an affected section of the ISTS. The corresponding section in the HBRSEP, Unit No. 2, TS is 3.8.8, "AC Instrument Bus Sources - Shutdown." The AC Instrument Bus sources support operation of systems that are used for mitigation of the FHA. Therefore, the exceptions to TSTF-5 1, Revision 2, based on the use of mitigation features in the FHA analysis, such as CREFS and FBACS, make the changes described in TSTF-5 1, Revision 2, for NUREG-143 1, Section 3.8.8, not applicable.

9 The TSTF-5 1, Revision 2, identifies NUREG- 1431, Section 3.8.10, "Distribution Systems - Shutdown," as an affected section of the ISTS. The corresponding section in the HBRSEP, Unit No. 2, TS is 3.8.10, "Distribution System - Shutdown." The AC Instrument Bus sources support operation of systems that are used for mitigation of the FHA. Therefore, the exceptions to TSTF-5 1, Revision 2, based on the use of mitigation features in the FHA analysis, such as CREFS and FBACS, make the changes described in TSTF-51, Revision 2, for NUREG-1431, Section 3.8.10, not applicable.

Safety Assessment The "Safety Assessment" contained in the original submittal dated March 13, 2002, is revised as follows:

Proposed Changes to TS 3.3.6:

The reanalyzed FHA analysis for inside containment, provided with this submittal, is based on a decay time of 56 hours6.481481e-4 days <br />0.0156 hours <br />9.259259e-5 weeks <br />2.1308e-5 months <br />. The containment ventilation isolation instrumentation is not utilized as mitigating feature in that analysis. Therefore, the proposed change to insert the word "recently" in the Applicability associated with the movement of irradiated fuel is justified, in accordance with TSTF-51, Revision 2. The appropriate decay time (56 hours6.481481e-4 days <br />0.0156 hours <br />9.259259e-5 weeks <br />2.1308e-5 months <br /> in the analysis provided with this submittal) will be incorporated into the HBRSEP, Unit No. 2, TS Bases, consistent with bases changes described in TSTF-51, Revision 2.

The requirement for the containment ventilation system instrumentation, Functions 1, 2, and 3, to be operable during purging is based on requirements that existed in the HBRSEP, Unit No. 2, TS at the time of conversion to ISTS. TS 3.3.6 Bases states, "During purging is defined as opening the containment purge supply and exhaust penetrations with irradiated fuel assemblies within the containment and does not include opening the Containment Pressure and Vacuum Relief System." The change to add MODES 1, 2, 3, and 4 to the Applicability for Function 3a,

United States Nuclear Regulatory Commission Attachment III to Serial: RNP-RA/02-0118 Page 5 of 10 "Containment Radiation Gaseous," and Function 3b, "Containment Radiation Particulate,"

replaces the "During Purging" Applicability.

The current TS 3.3.6 Bases also states, "While in MODES 5 and 6 without fuel handling or Purging operations in progress, the containment ventilation isolation instrumentation need not be OPERABLE because the potential for radioactive releases is minimized and operator action is sufficient to ensure post accident offsite doses are maintained with the limits of Reference 1 [10 CFR 100.11]." There are no postulated accidents in MODE 5 that require operability of the Containment Ventilation Isolation Instrumentation. In MODE 6, the "Other Specified Conditions" associated with the proposed Note (a) of Table 3.3.6-1 establishes the appropriate Applicability.

Therefore, the Applicability in MODES 1, 2, 3, 4, and (a) During movement of recently irradiated fuel assemblies within the containment, provides the appropriate Applicability requirements for Functions 1, 2, and 3, as listed in TS Table 3.3.6-1, and the justification for the proposed changes to the Applicability of TS 3.3.6, Containment Ventilation Isolation Instrumentation, is consistent with the justification for generic change TSTF-51, Revision 2.

Deletion of "During CORE ALTERATIONS" from the Applicability for TS 3.3.7, TS 3.7.9, TS 3.7.10, TS 3.9.3, TS 3.9.6, and TS 3.9.7:

TS 3.3.7, TS 3.7.9, TS 3.7.10, TS 3.9.3, TS 3.9.6, and TS 3.9.7 are being revised to delete "During CORE ALTERATIONS" from the Applicability. The HBRSEP, Unit No. 2, TS definition of CORE ALTERATIONS states, "The movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel." Accidents postulated to occur during core alterations include the FHA (Updated Final Safety Analysis Report [UFSAR] 15.7.4), boron dilution (UFSAR 15.4.6), and inadvertent loading of a fuel assembly into the improper position (UFSAR 15.4.7). As described in the referenced UFSAR sections, the only accident postulated to occur during core alterations that has the potential to cause a radioactive release is the FHA. Therefore, an Applicability of "During CORE ALTERATIONS" is not required forTS 3.3.7, TS 3.7.9, TS 3.7.10, TS 3.9.3, TS 3.9.6, and TS 3.9.7, because these TS sections continue to contain Applicability requirements that pertain to the movement of irradiated fuel. This justification for the deletion of "During CORE ALTERATIONS" from the Applicability of TS 3.3.7, TS 3.7.9, TS 3.7.10, TS 3.9.3, TS 3.9.6, and TS 3.9.7 is consistent with the justification provided in TSTF-51, Revision 2.

Change to the Applicability for TS 3.9.3 and TS 3.9.7 to be applicable when moving recently irradiated fuel:

The reanalyzed FHA analysis for inside containment, provided with this submittal, is based on a decay time of 56 hours6.481481e-4 days <br />0.0156 hours <br />9.259259e-5 weeks <br />2.1308e-5 months <br />. The closure of containment penetrations (TS 3.9.3) and the containment purge filter system (TS 3.9.7) are not utilized as mitigating features in that analysis. Therefore, the proposed change to insert the word "recently" in the Applicability associated with the movement of irradiated fuel is justified, in accordance with TSTF-5 1, Revision 2. The appropriate decay time (56 hours6.481481e-4 days <br />0.0156 hours <br />9.259259e-5 weeks <br />2.1308e-5 months <br /> in the analysis provided with this submittal) will be incorporated into the HBRSEP, Unit No. 2, TS Bases, consistent with bases changes described in TSTF-5 1, Revision 2.

United States Nuclear Regulatory Commission Attachment III to Serial: RNP-RAI02-0118 Page 6 of 10 Defense-in-Depth The NRC staff has traditionally and conservatively required containment systems to be operable during core alterations and movement of irradiated fuel within containment as a defense-in-depth measure to mitigate the consequences of the postulated FHA. In previous amendments for similar relaxations at other facilities [Safety Evaluation dated April 16, 2001, associated with Duane Arnold Energy Center Amendment No. 237 (ADAMS Accession Number ML011070147)], the NRC staff has requested licensees make appropriate commitments to implement administrative controls to facilitate restoration of containment closure should a FHA occur, consistent with the commitment described in TSTF-51. TSTF-51 requires licensees incorporating the generic change to commit to NUMARC 93-01, Revision 3, Section 11.2.6, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants,"

subheading "Containment - Primary(PWR)/Secondary(BWR)." The commitment in TSTF-51 was based on a draft version of NUMARC 93-01, Revision 3. When NUMARC 93-01, Revision 3, was approved in July 2000, the guidelines referred to in TSTF-51 were designated as Section 11.3.6.5. Section 11.3.6.5 of NUMARC 93-01 states:

Maintenanceactivities involving the need for open containment should include evaluation of the capability to achieve containment closure in sufficient time to mitigate potentialfission product release. This time is dependent on a number of factors, including the decay heat level and the amount of RCS inventory available.

For BWRs, technical specificationsmay require secondary containment to be closed under certain conditions, such as duringfuel handling and operationswith a potential to drain the vessel.

In addition to the guidance in NUMARC 91-06, for plants which obtain license amendments to utilize shutdown safety administrativecontrols in lieu of Technical Specification requirements on primaryor secondary containment operabilityand ventilation system operabilityduringfuel handling or core alterations,the following guidelinesshould be included in the assessment of systems removedfrom service:

"During fuel handling/corealterations,ventilation systems and radiationmonitor availability(as defined in NUMARC 91-06) should be assessed, with respect to filtration and monitoring of releasesfrom the fuel. Following shutdown, radioactivityin the RCS decays fairly rapidly. The basis of the Technical Specification operabilityamendment is the reduction in doses due to such decay.

The goal of maintaining ventilation system and radiationmonitor availability is to reduce doses even further below that provided by the naturaldecay, and to avoid unmonitored releases.

" A single normal or contingency method to promptly close primary or secondary containmentpenetrations should be developed. Such prompt methods need not completely block the penetrationor be capable of resistingpressure. The purpose is to enable ventilation systems to draw the releasefrom a postulatedfuel handling accident in the properdirection such that it can be treatedand monitored.

United States Nuclear Regulatory Commission Attachment III to Serial: RNP-RA/02-0118 Page 7 of 10 CP&L commits to follow the guidelines in Section 11.3.6.5 of NUMARC 93-01, Revision 3, at HBRSEP, Unit No. 2, during movement of irradiated fuel assemblies within containment. Plant procedures will be revised, as appropriate, to implement these guidelines.

FHA Reanalysis The meteorological data used to calculate the X/Q factors has been corrected as described in the response to NRC Information Requests 4 and 5 in Attachment II. Specifically, the blank fields in the datasets for periods 1988 - 1991 and 1992 - 1996 have been filled with "9's" (i.e., wind speed - 9999, wind direction - 999, stability - 99) and lower wind speed was corrected in the dataset representative of 1992 - 1996 period. The datasets were then used to re-perform the on site dispersion factor calculations using ARCON96. These changes resulted in slightly higher calculated dose for the Control Room. The calculated Control Room doses remain within the applicable limits of 10 CFR 50.67.

The revised analysis results for the Control Room, based on the meteorological data changes described previously, are:

Control Room Item (REM TEDE)

FHA Inside Containment 4.46 FHA Inside Fuel Handling Building 0.55 5(4)

Regulatory Limit NOTES:

(4) 10 CFR 50.67 and 10 CFR 50, Appendix A, Criterion 19.

The revised onsite dispersion factors that were previously provided in Attachment V to the March 13, 2002, submittal, are provided as follows:

Table 5 Onsite Ground Level XIQ Factors (sec/m 3)

Time Period FHA in Containment IFHA in Fuel Handling Building New, ARCON96 Based X/Q Values (Occupancy Factors Not Included) 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 4.15E-03 1.24E-03 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 2.74E-03 8.97E-04 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.17E-03 3.62E-04 1 - 4 days 8.18E-04 2.58E-04 4 - 30 days 6.74E-04 2.14E-04

United States Nuclear Regulatory Commission Attachment III to Serial: RNP-RA/02-0118 Page 8 of 10 No Significant Hazards Consideration Determination The "No Significant Hazards Consideration Determination" contained in the original submittal dated March 13, 2002, is revised as follows:

The proposed amendment would revise the Technical Specifications (TS) for H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, to permit selective implementation of an alternative source term (AST). The proposed amendment would modify the TS requirements for movement of irradiated fuel and performing core alterations.

An evaluation of the proposed amendment has been performed in accordance with 10 CFR 50.91(a)(1) regarding no significant hazards considerations using the standards in 10 CFR 50.92(c). A discussion of these standards as they relate to this amendment request follows:

1. The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated.

Implementation of the alternative source term does not affect the design or operation of HBRSEP, Unit No. 2; rather, once the occurrence of an accident has been postulated, the new source term is an input to evaluate the consequences of the postulated accident. A review of the HBRSEP, Unit No. 2, Updated Final Safety Analysis Report (UFSAR) shows that the components and systems affected by the proposed changes are not initiators of any previously analyzed accident. Therefore, there is no significant increase in the probability of any previously analyzed accident.

The implementation of the alternative source term has been evaluated in a revision to the HBRSEP, Unit No. 2, Fuel Handling Accident. Based on the results of this analysis, it has been demonstrated that, with the requested changes to the Technical Specifications, the dose consequences of a postulated Fuel Handling Accident are within the regulatory guidance provided by the NRC for use with the alternative source term. This guidance is presented in 10 CFR 50.67 and Regulatory Guide 1.183. A review of the HBRSEP, Unit No. 2, UFSAR shows that the only accident resulting in dose consequences that is postulated to occur during core alterations is the Fuel Handling Accident. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Previously Evaluated.

The proposed changes are supported by the revised design basis Fuel Handling Accident analysis. The proposed changes do not introduce any new modes of plant operation and do not involve physical modifications to the plant.

Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

United States Nuclear Regulatory Commission Attachment III to Serial: RNP-RAI02-0118 Page 9 of 10

3. The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety.

The proposed changes are associated with the implementation of a new licensing basis for HBRSEP, Unit No. 2. The new licensing basis implements an alternative source term in accordance with 10 CFR 50.67 and the associated Regulatory Guide 1.183. The results of the revised Fuel Handling Accident analysis, revised in support of this submittal, are subject to revised acceptance criteria. This analysis has been performed using conservative methodologies in accordance with the regulatory guidance. The dose consequences of the limiting Fuel Handling Accident are within the acceptance criteria found in the regulatory guidance associated with alternative source terms.

The proposed changes continue to ensure that doses at the exclusion area and low population zone boundaries, as well as the control room, are within the corresponding regulatory limits. Specifically, the margin of safety for this accident is considered to be that provided by meeting the applicable regulatory limits, which are conservatively set below the 10 CFR 50.67 limits. With respect to control room personnel doses, the margin of safety continues to be satisfied based on the applicable limit in 10 CFR 50.67 being satisfied.

Therefore, because the proposed changes continue to result in dose consequences within the applicable regulatory limits, they do not involve a significant reduction in a margin of safety.

Based on the above discussion, CP&L has determined that the requested change does not involve a significant hazards consideration.

Environmental Impact Consideration 10 CFR 51.22(c)(9) provides criteria for identification of licensing and regulatory actions for categorical exclusion for performing an environmental assessment. A proposed change for an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed change would not (1) involve a significant hazards consideration; (2) result in a significant change in the types or significant increases in the amounts of any effluents that may be released offsite; (3) result in an increase in individual or cumulative occupational radiation exposure. Carolina Power and Light (CP&L) Company has reviewed this request and determined that the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in conmction with the issuance of the amendment. The basis for this determination follows:

Proposed Change The proposed amendment would revise the Technical Specifications (TS) for H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, to permit selective implementation of an alternative source term (AST). The proposed amendment would modify the TS requirements for movement of irradiated fuel and performing core alterations.

United States Nuclear Regulatory Commission Attachment III to Serial: RNP-RA/02-0118 Page 10 of 10 Basis The proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) for the following reasons.

1. As demonstrated in the No Significant Hazards Consideration Determination, the proposed change does not involve a significant hazards consideration.
2. As demonstrated in the No Significant Hazards Consideration Determination, the proposed change does not result in a significant increase in the consequences of an accident previously evaluated and does not result in the possibility of a new or different kind of accident. Therefore, the proposed change does not result in a significant change in the types or significant increases in the amounts of any effluents that may be released offsite.
3. The alternative source term does not affect the design or operation of the facility; rather, once the occurrence of a fuel handling accident has been postulated, the alternative source term is an input to evaluate the consequences. The implementation of the alternative source term has been evaluated in the reanalysis of the design basis Fuel Handling Accident. Based on the results of this reanalysis, it has been demonstrated that, with the requested Technical Specifications changes, the dose consequences of the limiting event are within the regulatory guidance provided by the NRC for use with the alternative source term. Therefore, the proposed change does not result in an increase in individual or cumulative occupational radiation exposures.

United States Nuclear Regulatory Commission Attachment IV to Serial: RNP-RA/02-0118 11 Pages H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 REQUEST FOR TECHNICAL SPECIFICATIONS CHANGE REGARDING SELECTIVE IMPLEMENTATION OF ALTERNATIVE RADIOLOGICAL SOURCE TERM MARKUP OF TECHNICAL SPECIFICATIONS PAGES

Containment Ventilation Isolation Instrumentation 3.3.6 Table 3.3.6-1 (page 1 of 1)

Containment Ventilation Isolation Instrumentation FUNCTION APPLICABLE REQUIRED SURVEILLANCE TRIP MODES OR OTHER CHANNELS REQUIREMENTS SETPOINT SPECIFIED CONDITIONS

1. Manual Initiation 1,2,3,4. 2 SR 3.3.6.6 NA (a),-()
2. Automatic Actuation Logic and 2 trains SR 3.3.6.2 NA Actuation Relays 1,2,3,4. SR 3.3.6.3 (a),--)-4e SR 3.3.6.5
3. Containment Radiation
a. Gaseous a),(b)-.(- 6 1 SR 3.3.6.1 44 SR 3.3.6.4 1,2,3,4 SR 3.3.6.7
b. Particulate (a),(-G)( *-* SR 3.3.6.1 SR 3.3.6.4 SR 3.3.6.7
4. Safety Injection Refer to LCO 3.3.2, "ESFAS Instrumentation." Functions 1.a-f. for all initiation functions and requirements.

Ca) __,,, ________,_____T__ recently (a) During CORE ALTE-RATIOS (b) During movement o rradiated fuel assemblies within containment.

(c) During Purging.

44)---Trip Setpoint shall be in accordance with the methodology in the Offsite Dose Calculation Manual.

HBRSEP Unit No. 2 3.3-39 Amendment No. 474

CREFS Actuation Instrumentation 3.3.7 3.3 INSTRUMENTATION 3.3.7 Control Room Emergency Filtration System (CREFS) Actuation Instrumentation LCO 3.3.7 The CREFS actuation instrumentation for each Function in Table 3.3.7-1 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, 4 During movement of irradiated fuel assemblies IDujMg C A*jATI ACTIONS

NOTE .......................

-NOTE- ------------

Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One automatic A.1 Place one CREFS train 7 days actuation train in emergency inoperable, pressurization mode.

B. Two automatic B.1 Place one CREFS train Immediately actuation trains in emergency inoperable, pressurization mode.

OR One radiation monitoring channel inoperable.

(continued)

HBRSEP Unit No. 2 3.3-40 Amendment No. 476

CREFS Actuation Instrumentation 3.3.7 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time for Condition A AND or B not met in MODE 1, 2, 3, or 4. C.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> D. Required Action and D.1 / Sus nd C E edatelý associated Completion A ERAT NS.

Time for Condition A or B not met during movement of irr ed fuel assemblie tD.2 Suspend movement of Immediately durKg CO / irradiated fuel AL7RATI S. assemblies.

SURVEILLANCE REQUIREMENTS

--.-.---........--------------------- NOTE ------------------------------------

Refer to Table 3.3.7-1 to determine which SRs apply for each CREFS Actuation Function.

SURVEILLANCE FREQUENCY SR 3.3.7.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.3.7.2 Perform COT. 92 days SR 3.3.7.3 Perform ACTUATION LOGIC TEST. 31 days on a STAGGERED TEST BASIS (continued)

HBRSEP Unit No. 2 3.3-41 Amendment No. 4Z6

CREFS 3.7.9 3.7 PLANT SYSTEMS 3.7.9 Control Room Emergency Filtration System (CREFS)

LCO 3.7.9 Two CREFS trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4 "j . ý" iated fuel assemblies, ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CREFS train A.1 Restore CREFS train 7 days inoperable, to OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A AND not met in MODE 1, 2, 3, or 4. B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> C. Required Action and C.1 Place OPERABLE CREFS Immediately associated Completion train in emergency Time of Condition A pressurization mode.

not met during OR movement of irradiated fuel assemblies, o C.2.1 Suspend CO Imme 'ately duriW L0 / ALTERATIOS.

ALT ATI

/AND Z

(continued)

HBRSEP Unit 2 3.7-22 Amendement No. 1176

CREFS 3.7.9 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2.i Suspend movement of Immediately irradiated fuel assemblies.

D. Two CREFS trains D.1 Sus nd C E edatel inoperable during A RATI S.

movement of irrad d c fuel assemblies, duri COR e .2 Suspend movement of ALT 0TIO. irradiated fuel assemblies.

E. Two CREFS trains E.1 Restore at least one 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> inoperable in MODE 1, CREFS train to 2, 3, or 4. OPERABLE status.

F. Required Action and F.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition E AND not met in MODE 1, 2, 3, or 4. F.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.7.9.1 Operate each CREFS train for > 15 minutes. 31 days (continued)

HBRSEP Unit 2 3.7-23 Amendement No. V46

CREATC 3.7.10 3.7 PLANT SYSTEMS 3.7.10 Control Room Emergency Air Temperature Control (CREATC)

LCO 3.7.10 Two CREATC Water Cooled Condensing Unit (WCCU) trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4 Durinq movement of irradiated fuel assemblies.

D *ng A SI ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CREATC WCCU train A.1 Restore CREATC WCCU 30 days inoperable, train to OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A AND not met in MODE 1, 2, 3, or 4. B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)

HBRSEP Unit No. 2 3.7-25 Amendment No. 4.76

CREATC 3.7.10 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Place OPERABLE CREATC Immediately associated Completion WCCU train in Time of Condition A operation.

not met during movement of irra d OR urC.2.1 Sus nd C E edate IALTtRATIOS A ERAT NS.

D C.2 Suspend movement of Immediately irradiated fuel assemblies.

D. Two CREATC WCCU trains D.1 / Susp AL RATnd CO I medihtelV inoperable during movement of irradiated fuel assemblies, o A idu, 1, OR ALT/ TIOD.2 Suspend movement of Immediately L// irradiated fuel assemblies.

E. Two CREATC WCCU trains E.1 Restore at least one 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> inoperable in MODE 1, CREATC WCCU train to 2, 3, or 4. OPERABLE status.

F. Required Action and F.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition E AND not met in MODE 1, 2, 3, or 4. F.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> HBRSEP Unit No. 2 3.7-26 Amendment No. 476

Containment Penetrations 3.9.3 3.9 REFUELING OPERATIONS 3.9.3 Containment Penetrations LCO 3.9.3 The containment penetrations shall be in the following status:

a. The equipment hatch closed and held in place by four bolts;
b. One door in the air lock closed; and
c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere either:
1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or
2. capable of being closed by an OPERABLE Containment Ventilation Isolation System.

APPLICABILITY: SC9rE ALT RATI9N S During movement oT irradiated fuel assemblies within containment.

ACTIONS A. One or more containment penetrations not in required status. /

Suspend movement of Immedi atel y

.irradiated fuel assemblies within containment.

HBRSEP Unit No. 2 Amendment No. 1-7-6 3.9-4

Refueling Cavity Water Level 3.9.6 3.9 REFUELING OPERATIONS 3.9.6 Refueling Cavity Water Level LCO 3.9.6 Refueling cavity water level shall be maintained Ž 23 ft above the top of reactor vessel flange.

APPLICABILITY: Duri g COR ALTERAIONS, */cept d/ing 1lchin*and t unla ching contrA rod Wive sh fts.

During movement of irradiated fuel assemblies within containment.

ACTIONS SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify refueling cavity water level is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

23 ft above the top of reactor vessel flange.

HBRSEP Unit No. 2 3.9-10 Amendment No. 176

Containment Purge Filter System 3.9.7 3.9 REFUELING OPERATIONS 3.9.7 Containment Purge Filter System LCO 3.9.7 The Containment Purge Filter System shall be OPERABLE and operating.

APPLICABILITY: Pujr C9RrALJ9TIpjS.I During movement of irradiated fuel assemblies in containmen/

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment Purge .1 Close each Immediately Filter System penetration providing inoperable, direct access from the containment OR atmosphere to the outside atmosphere by Containment Purge a manual or automatic Filter System not in valve, blind flange, operation. or equivalent method.

OR A.2.1 Sus nd C E med/atel A RATI S.

AND A.2 Suspend movement of Immediately irradiated fuel re l "assemblies within containment.

HBRSEP Unit No. 2 3.9-11 Amendment No. 476

United States Nuclear Regulatory Commission Attachment V to Serial: RNP-RA/02-0118 12 Pages H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 REQUEST FOR TECHNICAL SPECIFICATIONS CHANGE REGARDING SELECTIVE IMPLEMENTATION OF ALTERNATIVE RADIOLOGICAL SOURCE TERM RETYPED TECHNICAL SPECIFICATIONS PAGES

Containment Ventilation Isolation Instrumentation 3.3.6 Table 3.3.6-1 (page 1 of 1)

Containment Ventilation Isolation Instrumentation FUNCTION APPLICABLE REQUIRED SURVEILLANCE TRIP MODES OR OTHER CHANNELS REQUIREMENTS SETPOINT SPECIFIED CONDITIONS

1. Manual Initiation 1,2,3.4.(a) 2 SR 3.3.6.6 NA
2. Automatic Actuation Logic and 2 trains SR 3.3.6.2 NA Actuation Relays 1,2.3,4,(a) SR 3.3.6.3 SR 3.3.6.5 I
3. Containment Radiation
a. Gaseous 1,2,3,4,(a) 1 SR 3.3.6.1 (b)

SR 3.3.6.4 SR 3.3.6.7

b. Particulate 1,2,3,4.(a) 1 SR 3.3.6.1 (b)

SR 3.3.6.4 SR 3.3.6.7

4. Safety Injection Refer to LCO 3.3.2. "ESFAS Instrumentation," Functions 1.a-f, for all initiation functions and requirements.

(a) During movement of recently irradiated fuel assemblies within the containment.

(b) Trip Setpoint shall be in accordance with the methodology in the Offsite Dose Calculation Manual.

HBRSEP Unit No. 2 3.3-39 Amendment No. 116

CREFS Actuation Instrumentation 3.3.7 3.3 INSTRUMENTATION 3.3.7 Control Room Emergency Filtration System (CREFS) Actuation Instrumentation LCO 3.3.7 The CREFS actuation instrumentation for each Function in Table 3.3.7-1 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, 4 During movement of irradiated fuel assemblies ACTIONS

-NOTE-NOTE ------------------------------------

Separate Condition entry is allowed for each Function.

S..............................................................................

CONDITION REQUIRED ACTION COMPLETION TIME A. One automatic A.1 Place one CREFS train 7 days actuation train in emergency inoperable, pressurization mode.

B. Two automatic B.1 Place one CREFS train Immediately actuation trains in emergency inoperable, pressurization mode.

OR One radiation monitoring channel inoperable.

(continued)

HBRSEP Unit No. 2 3.3-40 Amendment No. 4

CREFS Actuation Instrumentation 3.3.7 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time for Condition A AND or B not met in MODE 1, 2, 3, or 4. C.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> D. Required Action and D.1 Suspend movement of Immediately associated Completion irradiated fuel Time for Condition A assemblies.

or B not met during movement of irradiated fuel assemblies.

SURVEILLANCE REQUIREMENTS Refer to Table 3.3.7-1 to determine which SRs apply for each CREFS Actuation Function.

SURVEILLANCE FREQUENCY SR 3.3.7.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.3.7.2 Perform COT. 92 days SR 3.3.7.3 Perform ACTUATION LOGIC TEST. 31 days on a STAGGERED TEST BASIS (continued)

HBRSEP Unit No. 2 3.3-41 Amendment No. 476

CREFS 3.7.9 3.7 PLANT SYSTEMS 3.7.9 Control Room Emergency Filtration System (CREFS)

LCO 3.7.9 Two CREFS trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4 During movement of i rradiated fuel assemblies.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CREFS train A.1 Restore CREFS train 7 days inoperable, to OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A AND not met in MODE 1, 2,

3. or 4. B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> C. Required Action and C.1 Place OPERABLE CREFS Immediately associated Completion train in emergency Time of Condition A pressurization mode.

not met during movement of irradiated OR fuel assemblies.

C.2 Suspend movement of Immediately irradiated fuel assemblies.

(continued)

HBRSEP Unit 2 3.7-22 Amendment No. 46

CREFS 3.7.9 ACTIONS (continued)

D. Two CREFS trains D.1 Suspend movement of Immediately inoperable during irradiated fuel movement of irradiated assemblies.

fuel assemblies.

E. Two CREFS trains E.1 Restore at least one 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> inoperable in MODE 1, CREFS train to 2, 3, or 4. OPERABLE status.

F. Required Action and F.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition E AND not met in MODE 1, 2, 3, or 4. F.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.9.1 Operate each CREFS train for Ž 15 minutes. 31 days SR 3.7.9.2 Perform required CREFS filter testing in In accordance accordance with the Ventilation Filter with VFTP Testing Program (VFTP).

SR 3.7.9.3 Verify each CREFS train actuates on an 18 months actual or simulated actuation signal.

SR 3.7.9.4 Verify one CREFS train can maintain a 18 months on a positive pressure of Ž 0.125 inches water STAGGERED TEST gauge, relative to the outside atmosphere BASIS and a positive pressure relative to adjacent building areas during the emergency pressurization mode of operation at a makeup flow rate of

  • 400 cfm.

HBRSEP Unit 2 3.7-23 Amendment No. 176

CREFS 3.7.9 Page is intentionally blank HBRSEP Unit 2 3.7-24 Amendment No. V46

CREATC 3.7.10 3.7 PLANT SYSTEMS 3.7.10 Control Room Emergency Air Temperature Control (CREATC)

LCO 3.7.10 Two CREATC Water Cooled Condensing Unit (WCCU) trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4 During movement of i rradiated fuel assemblies.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CREATC WCCU train A.1 Restore CREATC WCCU 30 days inoperable, train to OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A AND not met in MODE 1, 2, 3, or 4. B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)

HBRSEP Unit No. 2 3.7-25 Amendment No. 176

CREATC 3.7.10 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Place OPERABLE CREATC Immediately associated Completion WCCU train in Time of Condition A operation.

not met during movement of irradiated OR fuel assemblies.

C.2 Suspend movement of Immediately irradiated fuel assemblies.

D. Two CREATC WCCU trains D.1 Suspend movement of Immediately inoperable during irradiated fuel movement of irradiated assemblies.

fuel assemblies.

E. Two CREATC WCCU trains E.1 Restore at least one 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> inoperable in MODE 1, CREATC WCCU train to 2, 3, or 4. OPERABLE status.

F. Required Action and F.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition E AND not met in MODE 1, 2, 3, or 4. F.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> HBRSEP Unit No. 2 3.7-26 Amendment No. 14.6

Containment Penetrations 3.9.3 3.9 REFUELING OPERATIONS 3.9.3 Containment Penetrations LCO 3.9.3 The containment penetrations shall be in the following status:

a. The equipment hatch closed and held in place by four bolts;
b. One door in the air lock closed; and
c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere either:
1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or
2. capable of being closed by an OPERABLE Containment Ventilation Isolation System.

APPLICABILITY: During movement of recently irradiated fuel assemblies within containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more containment A.1 Suspend movement of Immediately penetrations not in recently irradiated required status. fuel assemblies within containment.

HBRSEP Unit No. 2 3.9-4 Amendment No. 1--6

Refueling Cavity Water Level 3.9.6 3.9 REFUELING OPERATIONS 3.9.6 Refueling Cavity Water Level LCO 3.9.6 Refueling cavity water level shall be maintained 2 23 ft above the top of reactor vessel flange.

APPLICABILITY: During movement of irradiated fuel assemblies within containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Refueling cavity water A.1 Suspend movement of Immediately level not within irradiated fuel limit, assemblies within containment.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify refueling cavity water level is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 2 23 ft above the top of reactor vessel flange.

HBRSEP Unit No. 2 3.9-10 Amendment No. 176

Containment Purge Filter System 3.9.7 3.9 REFUELING OPERATIONS 3.9.7 Containment Purge Filter System LCO 3.9.7 The Containment Purge Filter System shall be OPERABLE and operating.

APPLICABILITY: During movement of recently irradiated fuel assemblies in containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment Purge A.1 Close each Immediately Filter System penetration providing inoperable, direct access from the containment OR atmosphere to the outside atmosphere by Containment Purge a manual or automatic Filter System not in valve, blind flange, operation. or equivalent method.

OR A.2 Suspend movement of Immediately recently irradiated fuel assemblies within containment.

HBRSEP Unit No. 2 3.9-11 Amendment No. P

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