Text

Supplement to the Request for License Amendments Related to Application of Alternative Source Term, Dated July 14, 2003
	ML050400083
Person / Time
Site:	Peach Bottom
Issue date:	01/21/2005
From:	Braun R; Exelon Generation Co, Exelon Nuclear
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML050400083 (58)
	v • d • e

Exelkn.

Exelon Nuclear Peach Bottom Atomic Power Station Telephone 717.456.7014 www.exeloncorp.com Nuclear 1848 Lay Road Delta, PA 17314-9032 10CFR 50.90 January 21, 2005 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Peach Bottom Atomic Power Station, Units 2 & 3 Renewed Facility Operating Ucense Nos. DPR-44 and DPR-56 NRC Docket Nos. 50-277 and 50-278

Subject:

Supplement to the Request for License Amendments Related to Application of Alternative Source Term, dated July 14, 2003

References:

(1) Letter from M. P. Gallagher (Exelon Generation Company, LLC) to US NRC, dated July 14, 2003, PBAPS Application of Alternative Source Term (2) Letter from G. F. Wunder (U. S. Nuclear Regulatory Commission) to J. L.

Skolds (Exelon Generation Company, LLC), dated June 29, 2004, Request for Addition Information (3) Letter from R. C. Braun (Exelon Generation Company, LLC) to US NRC, dated December 8, 2004, Supplement to the Request for License Amendments Related to Application of Alternative Source Term (4) Letter from M. P. Gallagher (Exelon Generation Company, LLC) to US NRC, dated December 9, 2003, "Exelon/AmerGen 180-Day Response To NRC Generic Letter 2003-01, 'Control Room Habitability'"

(5) Letter from R. C. Braun (Exelon Generation Company, LLC) to US NRC, dated January 21, 2005, Control Room Envelope Unfiltered Air Inleakage Test Results, in Response to Generic Letter 2003-01, "Control Room Habitability" On December 8, 2004, Exelon responded to a Request for Additional Information (RAI) for the Peach Bottom Atomic Power Station Alternative Source Term (AST) License Amendment Request (Reference 2). As part of the response, Exelon identified three questions that could not be fully answered at that time. Two of those responses (numbered 2 and 5) were awaiting the final results of the Peach Bottom Main Control Room Tracer Gas In-leakage Test. This test has been performed and the results are now finalized. A third question (numbered 4.d) required additional analysis and subsequent review by corporate and station engineering. to this supplemental letter provides a complete response to these remaining questions (2,4.d, and 5) from the Reference (2) Request for Additional Information (RAI). Attachment 2 to this supplemental letter provides replacement "markups" of revised Technical Specification Bases pages resulting from the reference (2) RAI response. Additional changes to the Technical Specification Ac Bases have been included for clarification and readability. Attachment 3 provides the typed pages of the revised Technical Specification Bases pages.

Supplement to the Request for License Amendments Related to Application of Alternative Source Term January 21, 2005 Page 2 As a result of revisions per this supplement, the following Technical Specification pages are withdrawn from the original submittal:

For Units 2 & 3 Page 3.3-59 Page 3.7-7 Page 3.7-8 The following Technical Specification Bases pages, which had been provided for information, are to be removed from the submittal. This list includes the affected Bases pages associated with Electrical TS withdrawn in the Reference (3) letter.

For Units 2 & 3 B 3.3-182 B 3.8-42 thru 45 B 3.7-19 thru 20 B 3.8-70 B 3.8-22 B 3.8-72 thru 74 B 3.8-38 B 3.8-76 B 3.8-40 B 3.8-94 thru 96 There is no impact to the No Significant Hazards Consideration submitted in the Reference (1) letter.

There is one additional commitment contained within this letter regarding the control of secondary containment hatches.

If you have any questions or require additional information, please contact Doug Walker at (610) 765-5726.

I declare under penalty of perjury that the foregoing is true and correct.

Respectfully, Executed on A/2 /oS ?/

Robert C. Braun Site Vice President Peach Bottom Atomic Power Station Exelon Generation Company, LLC Attachments: 1. Supplement to License Amendment Request for UPBAPS Alternative Source Term Implementation"

2. Markup of Technical Specification Bases Pages

3. Typed Technical Specification Bases Pages

4. Referenced Drawings

5. Commitment Page cc: S. J. Collins, Regional Administrator, Region I, USNRC F. L. Bower, USNRC Senior Resident Inspector, PBAPS G. F. Wunder, Project Manager [PBAPS] USNRC R. R.Janati - Commonwealth of Pennsylvania CCN: 05-14010

ATTACHMENT 1 PEACH BOTTOM ATOMIC POWER STATION UNITS 2 AND 3 Docket Nos. 50-277 50-278 Renewed License Nos. DPR-44 DPR-56 Supplement to License Amendment Request for "PBAPS Alternative Source Term Implementation" Exelon Responses to the NRC's Requests for Additional Information

Response To Request For Additional Information Peach Bottom Atomic Power Station, Units 2 and 3 Attachment 1 Proposed Use of Alternative Source Term (AST) Methodology Page 1 of 5 Question 2

2. In Reference 1, page 4 of 18, the licensee does not provide an acceptable response to question 6. The licensee has not verified that no other potential unfiltered inleakage pathways could result in X/Q values higher than the control room intake values. In light of the control room habitability issues noted in Generic Letter (GL) 2003-01, the staff does not believe that the licensee has provided adequate assurance that the current habitability requirements will continue to be met. Please provide the information requested.

Response to Question 2:

Integrated Tracer Gas testing using ASTM E741-00 was completed for Peach Bottom in October 2004. Test results, including measurement tolerances, demonstrated that the measured unfiltered inleakage is within the AST analyzed value used in Control Room Operator dose consequence analyses. The test demonstrated that the control room pressure was greater than that in adjacent areas. It is concluded that any unfiltered inleakage that might enter the MCR would be through the negative pressure portions of the ventilation system downstream of the filter but upstream of the fan as demonstrated by the positive differential pressure. These components are in close proximity to the intake and the unfiltered inleakage is analyzed using the intake X/Q. The results of the Integrated Tracer Gas Test have been provided in a supplement to the Exelon response to Generic Letter 2003-01 (reference 5).

Question 4

4. In Reference 1, Attachment 1, page 16, the response to question 32 does not provide a complete analysis upon which to judge the adequacy of the response. The staff requests further clarification and justification of the analysis performed.

Regarding reference 1, Appendix 5, page 24:

d. The proposed change to Technical Specification 3.6.4.1 (Secondary Containment) will no longer require that the secondary containment be operable during the movement of fuel assemblies that have a decay period of at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The FHA analysis assumes the release to the control room intake and the environment is through the turbine building/reactor building (TB/RB) ventilation stack. Please justify that an FHA release through the TB/RB ventilation stack is an appropriately conservative assumption given that the secondary containment may be inoperable. Include general arrangement drawings in your response showing the potential release points.

Response to Question 4.d:

Secondary containment operability assures that any post-Fuel Handling Accident (FHA) releases from the fuel pool are captured by the reactor building ventilation system and directed via the Standby Gas Treatment System (SGT) to the main stack. With secondary containment operability no longer required, the ability to maintain a secondary containment negative pressure may be compromised. Therefore, alternative flow paths for the releases to the environment must be considered.

Response To Request For Additional Information Peach Bottom Atomic Power Station, Units 2 and 3 Attachment 1 Proposed Use of Alternative Source Term (AST) Methodology Page 2 of 5 Potential building door openings such as the Railroad Bay doors and reactor building personnel access doors were evaluated as potential alternative flow paths, with the entire release assumed to be through any one of these openings. The Units 2 and 3 reactor building roof scuttles (access hatches to the roof) were also evaluated. Releases through these openings result in higher X/Q values than the TB/RB vent stack.

Calculated Control Room FHA doses due to this release pathway are acceptable.

PBAPS has also evaluated FHA releases from other potential openings, including the following hatch plugs outside and to the west of the Reactor and /Radwaste Buildings (elevation 135' and 116'). Refer to PBAPS drawing M-2 and M-3.

Units 2 and 3 RHR Room Hatches (8 per unit, elev. 135')

Units 2 and 3 Torus Room Hatches (1 per unit, elev. 135')

Units 2 and 3 HPCI Room Hatches (1 per unit, elev. 116')

Various cases were evaluated to analyze the impact to control room dose. These cases are described below. Each case described below uses an artificially high release rate to ensure all activity is released over a two-hour period without mixing.

Case 1 - Ventilation Stack: This evaluates normal (unfiltered) exhaust through the TB/RB ventilation stack. It assumes no SGT credit, ground-level release, and no MCREV credit.

Case 2 - Roof Scuttle: This evaluates a release through the roof scuttle with the worst X/Q for either Unit 2 or 3. It assumes no SGT credit, ground-level release, and no MCREV credit.

Case 3- HPCI Room Hatch: This evaluates a release through one of the external hatches described above. The hatch closest to the Control Room intake (Unit 2 HPCI Room hatch) represents the bounding dose for any of these external hatches (10 per unit). It assumes no SGT credit, ground-level release, but does assume MCREV credit. Various decay periods were evaluated to determine the minimum decay time after shutdown for which irradiated fuel movement is permissible with this worst-case hatch removed and MCREV credit. A decay period of 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months (3.5 days) was calculated as the shortest decay period that produces an acceptable dose.

I

Response To Request For Additional Information Peach Bottom Atomic Power Station, Units 2 and 3 Attachment 1 Proposed Use of Alternative Source Term (AST) Methodology Page 3 of 5 Shown below are the results and dose acceptance criteria for the cases described above.

EAB LPZ Dose CR Dose Dose (Rem (Rem Case Release Point Decay Time MCREV (Rem TEDE) TEDE)

Cae RlaePit (Hours) Credit TEDE) [Limit = [Limit

[Limit = 6.3] 5.0]

._ 6.3]

Worst-Case TB/RB 1 Stack Release 24 No 1.16 0.132 2.35 Pathway Worst-Case 2 Reactor Building 24 No 1.16 0.132 3.85 Roof Scuttle Limiting External 3 Hatch Release 84 (3.5 d) Yes 0.714 0.081 4.56 Pathway Since the bounding control room dose (4.56 Rem TEDE) through the limiting external hatch was determined with MCREV credit, MCREV credit will be required for all FHA scenarios. As a result, the proposed revisions to the MCREV Technical Specifications and Bases are withdrawn. Although secondary containment is not required after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , these external hatches will not be opened concurrently with movement of irradiated fuel until 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months after shutdown.

These results indicate that the calculated consequences of a design basis Fuel Handling Accident at or after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of shutdown with releases through the closest (worst-case) roof scuttle release pathway are within regulatory limits without SGT credit.

Bounding FHA Doses Location Dose Acceptance Criteria (Rem TEDE) (Rem TEDE)

EAB 1.16 6.3 LPZ 0.132 6.3 1CR 4.56 5.0 Although an 84-hour decay period is required for Case 4, fuel movement would still be permitted without secondary containment integrity after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , provided these hatches remain in place. The 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months is a specific restriction for removal of these ground level hatches to ensure the release does not exit these openings prior to the appropriate decay period (84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months ). PBAPS is committing to have administrative controls of these hatches to ensure they remain closed while moving irradiated fuel until 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months decay after shutdown. PBAPS Technical Specification Bases will include this restriction.

Response To Request For Additional Information Peach Bottom Atomic Power Station, Units 2 and 3 Attachment 1 Proposed Use of Alternative Source Term (AST) Methodology Page 4 of 5 Question 5

5. In Reference 2 below, Attachment 1, page 10, the PBAPS response to question 17 does not provide a confirmation of the assumed inleakage value in the proposed amendment request.

Many licensees have found that walkdowns, while useful, do not alone provide a reliable method of determining the susceptibility of a control room to inleakage. PBAPS has also not confirmed that their facility's control room meets the applicable habitability regulatory requirements and that the control room habitability systems are designed, constructed, configured, operated, and maintained in accordance with the facility's design and licensing bases. Therefore, the staff believes that PBAPS has not shown that GDC 19 will be met with the proposed amendment. Please provide this confirmation as requested by question 1 of GL 2003-01 so that confirmation of your habitability requirements can be made. One method acceptable to the staff that may be used to provide this confirmation is Regulatory Guide 1.196, "Control Room Habitability at Light-Water Nuclear Power Reactors."

Response to Question 5:

Confirmation that the PBAPS Control Room meets the applicable habitability regulatory requirements and that the control room habitability systems are designed, constructed, configured, operated, and maintained in accordance with the facility's design and licensing bases has been provided as documented in the Exelon response to Generic Letter 2003-01 dated December 9, 2003.

Furthermore, Integrated Tracer Gas testing using ASTM E741 -00 was completed for Peach Bottom in October 2004. Test results, including measurement tolerances, demonstrated that the measured unfiltered inleakage is within the AST analyzed value used in Control Room Operator dose consequence analyses. Additionally, the results of the Integrated Tracer Gas Test have been provided in a supplement to the Exelon response to Generic Letter 2003-01 (reference 5).

ATTACHMENT 2 PEACH BOTTOM ATOMIC POWER STATION UNITS 2 AND 3 Docket Nos. 50-277 50-278 Renewed License Nos. DPR-44 DPR-56 Supplement to License Amendment Request for UPBAPS Alternative Source Term Implementation" Markup of Technical Specification Bases Pages (For Information Only)

UNITS 2 & 3 Revised Inserts Pages (Changes to Inserts annotated with Revision Bars)

B 3.6-73 thru 75 B 3.7-17 thru 18

PBAPS Units 2 and 3 Technical Specification Bases Markup Inserts INSERT A {pq. B 3.6-291 Total leakage through all four main steam lines must be < 150 scfh, and < 75 scfh for any one steam line, when tested at > 25 psig. The analysis in Reference 1 is based on treatment of MSIV leakage as secondary containment bypass leakage, independent of the primary to secondary containment leakage analyzed at La. The Frequency is in accordance with the Primary Containment Leakage Rate Testing Program.

INSERT B fta. B 3.1-391 The SLC System is also used to maintain suppression pool pH at or above 7 following a loss of coolant accident (LOCA) involving significant fission product releases. Maintaining suppression pool pH levels at or above 7 following a LOCA ensures that iodine will be retained in the suppression pool water.

INSERT C fpq. B 3.1-411 In MODES 1, 2, and 3, the SLC System must be OPERABLE to ensure that offsite doses remain within 10 CFR 50.67 (Ref. 3) limits following a LOCA involving significant fission product releases. The SLC System is designed to maintain suppression pool pH at or above 7 following a LOCA to ensure that iodine will be retained in the suppression pool water.

INSERT D fPq. B 3.3-1561

. Both channels are also required to be OPERABLE in MODES 1, 2, and 3, since the SLC System is also designed to maintain suppression pool pH above 7 following a LOCA to ensure that iodine will be retained in the suppression pool water. These INSERT E {na. B 3.6-731 The function of the secondary containment is to receive fission products that may leak from primary containment or from systems in secondary containment following a Design Basis Accident (DBA) and, in conjunction with the Standby Gas Treatment System (SGT) and closure of certain valves whose lines penetrate the secondary containment, to provide for elevated release through the Main Stack.

INSERT F tpa. B 3.6-761 The SGT System exhausts the secondary containment atmosphere to the environment through the elevated release point provided by the Main Stack.

To ensure that this exhaust pathway is used, SR 3.6.4.1.3 1

INSERT G {To. B 3.6-851 The primary function of the SGT System is to ensure that radioactive materials that leak from primary containment into the secondary containment following a Design Basis Accident (DBA) are discharged through the elevated release provided by the Main Stack.

INSERT H fpq. B 3.6-851 These filters are not credited in any DBA analysis.

INSERT I fpq. B 3.6-861 The design basis for the SGT System is to mitigate the consequences of a loss of coolant accident by providing a controlled, elevated release path. The SGT system also provides this function for OPDRVs. For all events where required, the SGT System automatically initiates to reduce, via an elevated release, the consequences of radioactive material released to the environment.

The HEPA filter and charcoal adsorber provided in the SGT System are not credited for any DBA analysis.

INSERT J {fp. B 3.6-901 The only credited safety function of the SGT System is to provide a secondary containment vacuum sufficient to assure that discharges from the secondary containment will be through the Main Stack. The VFTP test 5.5.7.d. provides verification that the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is acceptable. SR 3.6.4.1.3 and SR 3.6.4.1.4 provide assurance that sufficient vacuum in the secondary containment is established with the time period as used within the DBA LOCA analysis.

INSERT K fpq. B 3.7-161 Additionally, the MCREV System is designed to maintain the control room environment for a 30-day occupancy after a DBA without exceeding 5 rem TEDE.

INSERT L {Dq. B 3.7-161 The MCREV System is credited as operating following a loss of coolant accident or fuel handling accident. The MCREV System is not credited in the analysis of the main steam line break or the control rod drop accident, 2

INSERT M {Tn B 3.6-741 Movement of recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) requires that secondary containment be OPERABLE. After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following reactor shutdown (all rods in),

secondary containment operability is not required for movement of irradiated fuel. However, the eight external concrete plugs for the RHR rooms (el. 135'), the one external concrete plug for the torus room (el. 135'), and the one HPCI room hatch (el. 116') must remain in place to ensure MCR and offsite dose limits are met when moving irradiated fuel within the first 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months after shutdown.

Following 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months after shutdown (all rods in), there are no restrictions on removal or opening of the external concrete plugs and hatch described above when moving irradiated fuel.

INSERT N {na B 3.6-871 The SGT System is required to be OPERABLE during handling of recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

INSERT P {nx B 3.6-791 SCIVs are required to be OPERABLE during handling of recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

INSERT Q {na B 3.8-401 (This change has been withdrawn)

INSERT R fPG B 3.8-42.43 72,73,74-94. and 951 (This change has been withdrawn)'

INSERT S {nx B 3.8-941 (This change has been withdrawn)

INSERT T {na B 3.8-741 (This change has been withdrawn) 3

INSERT U {pc B 3.6-82, 3.6-88, 3.6-89. 1 I

, since the movement of recently irradiated fuel can only be performed in MODES 4 and 5.

I INSERT V frq B 3.8-44,741 (This change has been withdrawn)

INSERT W 1pq B 3.3-1741 The Functions are only required to be OPERABLE during handling of recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

I INSERTX foa B 3.3-1821 (This change has been withdrawn)

INSERT Y fIr B 3.1-401 The sodium pentaborate solution in the SLC System is also used, post-LOCA, to maintain suppression pool pH levels above 7. The system parameters used in the calculation are the I Boron-10 minimum mass of 162.7 Ibm, and an upper bound Boron-10 enrichment of 65%.

I INSERT Z tfa B 3.7-171 (This change has been withdrawn)

INSERTAA fMxq B 3.8-22. 3.8-38. 3.8-701 (This change has been withdrawn)

INSERT BB {na B 3.6-741 I Movement of recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) requires that secondary containment be OPERABLE.

After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following shutdown (all rods in), fuel moves may proceed without secondary containment OPERABLE provided that the eight external concrete plugs for the RHR rooms (el. 135'), the one external concrete plug for the torus room (el. 135'), and the one HPCI room hatch (el. 116') remain in place to ensure MCR and offsite dose limits are met. There are no restrictions to other secondary 4

containment hatches or penetrations. After 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months following shutdown (all rods in), Alternative Source Term calculations show that fuel moves may occur with no restrictions on removal or opening of the external concrete plugs and hatch described above.

INSERT CC {pq B 3.6-751 within the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following shutdown. After the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , secondary containment is not required to support movement of irradiated fuel. Note that the eight external concrete plugs for the RHR rooms (el. 135'), the one external concrete plug for the torus room (el. 135'), and the one HPCI room hatch (el. 116') must remain in place to ensure dose limits are met for the first 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months following shutdown (all rods in).

5

Secondary Containment B 3.6.4.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.1 Secondary Containment BASES The secondary containment is a structure that completely encloses the primary containment and those components that may be postulated to contain primary system fluid. This structure forms a control volume that serves to hold up and dilute the fission products. It is possible for the pressure in the control volume to rise relative to the environmental pressure (e.g., due to pump and motor heat load additions). To prevent ground level exfiltration while allowing the secondary containment to be designed as a conventional structure, the secondary containment requires support systems to maintain the control volume pressure at less than the external pressure. Requirements for these systems are specified separately in LCO 3.6.4.2, "Secondary Containment Isolation Valves (SCIVs)," and LCO 3.6.4.3,

'Standby Gas Treatment (SGT) System.'

APPLICABLE There are two principal accidents for which credit is taken SAFETY ANALYSES for secondary containment OPERABILITY. These are a loss of coolant accident (LOCA) (Ref. 1) and a fuel handling accident inside secondary containment (Ref. 2). The secondary containment performs no active function in response to each of these limiting events; however, its leak (continued)

PSAPS UNIT 2 B 3.6-73 Revision No. 0

Contairiient Secondary Secondary Containment B 3.6.4.1 BASES APPLICABLE tightness is required to ensure that fission products SAFETY ANALYSES entrtoed within the secondar containment structure will be (co by the SGT System discharge to the environmeno.

Secondary containment satisfies Criterion 3 of the NRC

- M~~'

~'- >olicy Statement.

LCO An OPERABLE secondary containment provides a control volume into which fission products that leak from primary containment, or are released from the reactor coolant pressure boundary components located in sec d r containment, can be -r___, o the environment. For the secondary containmert to be considered OPERABLE, it must have adequate leak tightness to ensure hat the required vacuum can be established and maintained.

APPLICABILITY In MODES 1, 2, and 3, a LOCA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, secondary containment OPERABILITY is required during the same operating conditions that require primary containment OPERABILITY.

In MODES 4 and 5, the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining secondary containment OPERABLE is not required in MODE 4 or 5 to ensure a control volume, except for other situations for which significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel (OPDRVs), during CORE ALTERATIONS, or during movement of irradiated fuel n5er assemblies in the secondary containment. if f ACTIONS A.1 If secondary containment is inoperable, it must be restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time provides a period of time to correct the problem that is commensurate with the importance of maintaining secondary containment during MODES 1, 2, and 3. This time period also ensures that the probability of an accident (requiring secondary containment OPERABILITY) occurring during periods where secondary containment is inoperable is minimal.

(continued)

PBAPS UNIT 2 B 3.6-74 Revision No. 0

Containment Secondary Secondary Containment B 3.6.4.1 BASES ACTIONS B.l and B.2 (continued)

If secondary containment cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions In an orderly manner and without challenging plant systems.

C.I. C:2. and C.3 Movement of irradiated fuel assemblies in the secondary containment, CORE ALTERATIONS, and OPDRVs can be postulated to cause fission product release to the secondary containment. In such cases, the secondary containment is the only barrier to release of fission products to the environment. CORE ALTERATIONS and movement of irradiated fuel assemblies must be immediately suspended if the secondary containment is ihoperable.

Suspension of these activities shall not preclude completing an action that involves moving a component to a safe position. Also, action must be immediately initiated to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

Required Action C.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

(continued)

PBAPS UNIT 2 B 3.6-75 Revision No. 0

MCREV System B 3.7.4 BASES BACKGROUND initiate an emergency shutdown of non-essential equipment (continued) and lighting to reduce the heat generation to a minimum.

Heat removal would be accomplished by conduction through the floors, ceiling s and walls to0adjacent rooms and t enM rvirulmen alO7da' the control room to prevent infiltration of air from surrounding buildings. NCREV System operation in maintaining control room habitability is discussed in the UFSAR, Chapters 7, 10, and 12, (Refs. 1, 2, and 3, respectively).

APPLICABLE The ability of the NCREV Systier to maintain the SAFETY ANALYSES habitability of the control room is an explicit assumption for the safety analyses presented in the UFSAR Chapter 10

-andd 2 (Refs. 2 and 3, respectively). l~ CE System si\

INSERTL s to Nperate!o wing a los o coolant cident, INSF oi_+hndlinaccident, Agi t ie a 'totrol.

break,

_ J ro do4-acidg tura-s-di-scussed in the uF AR, -'

Section 14.9.1.5 (Ref. 4). The radiological doses to control root personnel as a result of the various DBAs are 9 summarized in Reference 4. No single active or passive failure will cause the loss of outside or recirculated air from the control room.

The MCREV System -satisfies Criterion 3 of the. NRC Policy Statement.

LCO Two redundant subsystems of the MCREV System are required to be OPERABLE to ensure that at least one is available, assuming a single failure disables the other subsystem.

Total system failure could result in exceeding a dose f 5 rem to the control room operators in the event of a The ICREV System is considered OPERABLE when the individual ° components necessary to control. operator exposure are or OPERABLE in both subsystems. A subsystem is considered /

OPERABLE when its associated:

a. Fan is OPERABLE; (FHA^ )

PBAPS UNIT 2 B 3.7-16 Revision Nio. 0

MCREV System B 3.7.4 BASES LCO b. HEPA filter and charcoal adsorbers are not excessively (continued) restricting flow and are capable of performing their filtration functions; and

c. Ductwork, valves, and dampers are OPERABLE, and air flow can be maintained.

In addition, the control room boundary must be maintained, including the integrity of the walls, floors, ceilings, and ductwork. Temporary seals may be used to maintain the boundary. In addition, an access door may be opened provided the ability to pressurize the control room is maintained and the capability exists to close the affected door in an expeditious manner.

APPLICABILITY In MODES 1, 2, and 3, the MCREV System must be OPfMBLE to contro operator exposure during and following a since the/D could lead is oduct release.

In MODES 4 and 5, the probability and consequences of A-are reduced because of the pressure and temperature limitations in these MODES. Therefore, maintaining the MCREV System OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioactive releases can be postulated:

a. - During operations with potential for draining the reactor vessel (OPDRVs);

b. During CORE ALTERATIONS; and

c. During movement of irradiated fuel assemblies in the secondary containment.

ACTIONS A.1 With one MCREV subsystem inoperable, the inoperable MCREV subsystem must be restored to OPERABLE status within 7 days.

With the unit in this condition, the remaining OPERABLE MCREV subsystem is adequate to maintain control room temperature and to perform control room radiation protection. However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could (continued)

PBAPS UNIT 2 B 3.7-17 Revision No. 0

MCREV System B 3.7.4 BASES ACTIONS LA1 (continued) result in reduced MCREV System capability. The 7 day Completion Time is based on the low probability of a occurring during this time period, and that the remailig subsystem can provide the required capabilities.

B.1 and B.2 In MODE 1, 2, or 3, if the inoperable MCREV subsystem cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE that minimizes risk. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

C.]. C.2.1. C.2.2. and C.2.3 The Required Actions of Condition C are modified by a Note indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.

Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.

During movement of irradiated fuel assemblies in the secondary containment, during CORE ALTERATIONS, or during OPDRVs, if the inoperable MCREV subsystem cannot be restored to OPERABLE status within the required Completion Time, the OPERABLE MCREV subsystem may be placed in operation. This action ensures that the remaining subsystem is OPERABLE, that no failures that would prevent automatic actuation will occur, and that any active failure will be readily detected.

An alternative to Required Action C.1 is to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk.

(continued)

PBAPS UNIT 2 B 3.7-18 Revision No. 0

Secondary Containment B 3.6.4.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.1 Secondary Containment BASES 2Lmenor ha primary co0t4ainm2it is nrit requie fbk be

'lkeace oItside primaryppontai ment.

The secondary containment is a structure that completely encloses the primary containment and those components that may be postulated to contain primary system fluid. This structure forms a control volume that serves to hold up and dilute the fission products. It is possible for the pressure in the control volume to rise relative to the environmental pressure (e.g., due to pump and motor heat load additions). To prevent ground level exfiltration while allowing the secondary containment to be designed as a conventional structure, the secondary containment requires support systems to maintain the control volume pressure at less than the external pressure. Requirements for these systems are specified separately in LCO 3.6.4.2, "Secondary Containment Isolation Valves (SCIVs)," and LCO 3.6.4.3, "Standby Gas Treatment (SGT) System."

APPLICABLE There are two principal accidents for which credit is taken SAFETY ANALYSES for secondary containment OPERABILITY. These are a loss of coolant accident (LOCA) (Ref. 1) and a fuel handling accident inside secondary containment (Ref. 2). The secondary containment performs no active function in response to each of these limiting events; however, its leak (continued)

PBAPS UNIT 3 B 3.6-73 Revision No. 0

Secondary Containment B 3.6.4.1

' BASES APPLICABLE tightness is-required to ensure that fission products SAFETY ANALYSES entrapped within the secondar_containment structure will be (continued by the SGT System , discharge to the

' environment

/Secondary containment satisfies Criterion 3 of the NRC

(~ ___Policy Statement.

LCO An OPERABLE secondary containment provides a control volume into which fission products that leak from primary containment, or are released from the reactor coolant pressure boundary components located in secondary containment, can be I IsI the- ____,-'

environment. For the secondary containmen-f to be considered jY~Ew ~ OPERABLE, it must have adequate leak tightness to ensure that the required vacuum can be established and maintained.

APPLICABILITY In MODES 1, 2, and 3, a LOCA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, secondary containment OPERABILITY is required during the same operating conditions that require primary containment OPERABILITY.

In MODES 4 and 5, the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining secondary containment OPERABLE is not required in MODE 4 or 5 to ensure a control volume, except for other situations for which significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel (OPDRVs), during CORE ALTERATIONS, or during movement of irradiated fuel assemblies in the secondary containment.

0.

ACTIONS A.I If secondary containment is inoperable, it must be restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time provides a period of time to correct the problem that is commensurate with the importance of maintaining secondary containment during MODES 1, 2, and 3. This time period also ensures that the probability of an accident (requiring secondary containment OPERABILITY) occurring during periods where secondary containment is inoperable is minimal.

(continued)

PBAPS UNIT 3 B 3.6-74 Revision No. 0

Secondary Containment Secondary Containment B 3.6.4.1 BASES ACTIONS B.1 and B.2 (continued)

If secondary containment cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

C.1. C.2. and C.3 Movement of irradiated fuel assemblies in the secondary containment, CORE ALTERATIONS, and OPDRVs can be postulated to cause fission product release to the secondary containment. In such cases, the secondary containment is the only barrier to release of fission products to the environment. CORE ALTERATIONS and movement of irradiated

- ) fuel assemblies must be immediately suspended if the JLN.A T`EIC secondary containment is inoperablW -

Suspension of these activities shall not preclude completing an action that involves moving a component to a safe position. Also, action must be immediately initiated to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

Required Action C.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

(continued)

PBAPS UNIT 3 B 3.6-75 Revision No. 0

i I

MCREV System i B 3.7.4 I I

I BASES BACKGROUND initiate an emergency shutdown of non-essential equipment (continued) and lighting to reduce the heat generation to a minimum.

Heat removal would be accomplished by conduction through the floors, ceilings, and walls to adjacent rooms and to the environment. temfs dkiqna 3who bqdros lingle MCREV subsystem wi~llpressurize thi cont' room to prevent infiltration of air from surrounding buildings. MCREV System operation in maintaining control room habitability is discussed in the UFSAR, Chapters 7, 10, and 12, (Refs. 1, 2, and 3, respectively).

APPLICABLE The ability of the MCREV System to maintain the SAFETY ANALYSES habitability of the control room is an explicit assumption for the safety analyses presented in the UFSAR Chapters 10

_ I n asdscussed in the UFSAR Section 14.9.1.5 (Ref. 4). The radiological doses to control room personnel as a result of the various DBAs are summarized in Reference 4. No single active or passive failure will cause the loss of outside or recirculated air from the control room.

The MCREV System satisfies Criterion 3 of the NRC Policy Statement.

LCO Two redundant subsystems of the MCREV System are required to be OPERABLE to ensure that at least one is available, assuming a single failure disables the other subsystem.

Total system failure could result in exceeding a dose of 5 rem to the control room operators in the event of a MCREV System is considered OPERABLE when the individual or components necessary to control operator exposure are q £vel OPERABLE in both subsystems. A subsystem is considered OPERABLE when its associated:

a. Fan is OPERABLE; (continued)

PBAPS UNIT 3 B 3.7-16 Revision No. 0

MCREV System B 3.7.4 BASES LCO b. HEPA filter and charcoal adsorbers are not excessively (continued) restricting flow and are capable of performing their filtration functions; anc

c. Ductwork, valves, and dampers are OPERABLE, and air flow can be maintained.

In addition, the control room boundary must be maintained, including the integrity of the walls, floors, ceilings, and ductwork. Temporary seals may be used to maintain the boundary. In addition, an access door may be opened provided the ability to pressurize the control room is maintained and the capability exists to close the affected door in an expeditious manner.

APPLICABILITY In MODES 1, 2, and 3, the MCREV System must be OPERABLE to control operator exposure during and following a since the could ea f roduct release.

In MODES 4 and 5, o a y and consequences of a %

are reduced because of the pressure and temperature limitations in these MODES. Therefore, maintaining the MCREV System OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioactive releases can be postulated:

a. During operations with potential for draining the reactor vessel (OPDRVs);

b. During CORE ALTERATIONS; and

c. During movement of irradiated fuel assemblies in the secondary containment.

ACTIONS A.1 With one MCREV subsystem inoperable, the inoperable MCREV subsystem must be restored to OPERABLE status within 7 days.

With the unit in this condition, the remaining OPERABLE MCREV subsystem is adequate to maintain control room temperature and to perform control room radiation protection. However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could (continued)

PRAPS UNIT 3 B 3.7-17 Revision No. 0

MCREV System B 3.7.4 BASES ACTIONS A.1 (continued) result in reduced MCREV System capability. The 7 day Completion Time is based on the low probability of a occurring during this time period, and that the remaining subsystem can provide the required capabilities.

B.1 and B.2 In MODE 1, 2, or 3, if the inoperable MCREV subsystem cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE that minimizes risk. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

C.I. C.2.1. C.2.2. and C.2.3 The Required Actions of Condition C are modified by a Note indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.

Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.

During movement of irradiated fuel assemblies in the secondary containment, during CORE ALTERATIONS, or during OPDRVs, if the inoperable MCREV subsystem cannot be restored to OPERABLE status within the required Completion Time, the OPERABLE MCREV subsystem may be placed in operation. This action ensures that the remaining subsystem is OPERABLE, that no failures that would prevent automatic actuation will occur, and that any active failure will be readily detected.

An alternative to Required Action C.1 is to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk.

(continued)

PBAPS UNIT 3 B 3.7-18 Revision No. 0

ATTACHMENT 3 PEACH BOTTOM ATOMIC POWER STATION UNITS 2 AND 3 Docket Nos. 50-277 50-278 Renewed License Nos. DPR-44 DPR-56 Supplement to License Amendment Request for "PBAPS Alternative Source Term Implementation" Typed Technical Specification Bases Pages (For Information Only)

UNITS 2 & 3 B 3.1-39 B 3.1-40 B 3.6-73 B 3.6-74 B 3.6-74a B 3.6-75 B 3.6-79 B 3.6-87 B 3.6-90 B 3.7-16 B 3.7-17 B 3.7-18

SLC System B 3.1.7 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.7 Standby Liquid Control (SLC) System BASES BACKGROUND The SLC System is designed to provide the capability of bringing the reactor, at any time in a fuel cycle, from full power and minimum control rod inventory (which is at the peak of the xenon transient) to a subcritical condition with the reactor in the most reactive, xenon free state without taking credit for control rod movement. The SLC System satisfies the requirements of 10 CFR 50.62 (Ref. 1) on anticipated transient without scram using enriched boron.

The SLC System is also used to maintain suppression pool pH at or above 7 following a loss of coolant accident (LOCA) involving significant fission product releases. Maintaining suppression pool pH levels at or above 7 following a LOCA ensures that iodine will be retained in the suppression pool water.

Reference 1 requires a SLC System with a minimum flow capacity and boron content equivalent in control capacity to 86 gpm of 13 weight percent sodium pentaborate solution.

Natural sodium pentaborate solution is 19.8% atom Boron-10.

Therefore, the system parameters of concern, boron concentration (C), SLC pump flow rate (Q), and Boron-10 enrichment (E), may be expressed as a multiple of ratios.

The expression is as follows:

C a E x x 13% weight 86 gpm 19.8% atom If the product of this expression is 2 1, then the SLC System satisfies the criteria of Reference 1. As such, the equation forms the basis for acceptance criteria for the surveillances of concentration, flow rate, and boron enrichment and is presented in Table 3.1.7-1.

The SLC System consists of a boron solution storage tank, two positive displacement pumps, two explosive valves that are provided in parallel for redundancy, and associated piping and valves used to transfer borated water from the storage tank to the reactor pressure vessel (RPV). The borated solution is discharged near the bottom of the core shroud, where it then mixes with the cooling water rising through the core. A smaller tank containing demineralized water is provided for testing purposes.

(continued)

PBAPS UNIT 2 B 3.1 -39 Revision No.

SLC System B 3.1.7 BASES (continued)

APPLICABLE The SLC System is manually initiated from the main control SAFETY ANALYSES room, as directed by the emergency operating procedures, if the operator believes the reactor cannot be shut down, or kept shut down, with the control rods. The SLC System is used in the event that enough control rods cannot be inserted to accomplish shutdown and cooldown in the normal manner. The SLC System injects borated water into the reactor core to add negative reactivity to compensate for all of the various reactivity effects that could occur during plant operations. To meet this objective, it is necessary to inject a quantity of boron, which produces a concentration of 660 ppm of natural boron, in the reactor coolant at 688 F. To allow for potential leakage and imperfect mixing in the reactor system, an additional amount of boron equal to 25% of the amount cited above is added (Ref. 2). The minimum mass of Boron-10 (162.7 lbm) needed for injection is calculated such that the required quantity is achieved accounting for dilution in the RPV with normal water level and including the water volume in the residual heat removal shutdown cooling piping and in the recirculation loop piping. This quantity of borated solution is the amount that is above the pump suction shutoff level in the boron solution storage tank. No credit is taken for the portion of the tank volume that cannot be injected. The maximum concentration of sodium pentaborate listed in Table 3.1.7-1 has been established to ensure that the solution saturation temperature does not exceed 430 F.

The sodium pentaborate solution in the SLC System is also used, post-LOCA, to maintain suppression pool pH levels above 7. The system parameters used in the calculation are the Boron-10 minimum mass of 162.7 lbm, and an upper bound Boron-10 enrichment of 65%.

The SLC System satisfies Criterion 4 of the NRC Policy Statement.

LCO The OPERABILITY of the SLC System provides backup capability for reactivity control independent of normal reactivity control provisions provided by the control rods. The OPERABILITY of the SLC System is based on the conditions of the borated solution in the storage tank and the availability of a flow path to the RPV, including the OPERABILITY of the pumps and valves. Two SLC subsystems are required to be OPERABLE; each contains an OPERABLE pump, an explosive valve, and associated piping, valves, and instruments and controls to ensure an OPERABLE flow path.

(continued)

PBAPS UNIT 2 B 3.1-40 Revision No.

Secondary Containment B 3.6.4.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.1 Secondary Containment BASES BACKGROUND The function of the secondary containment is to receive fission products that may leak from primary containment or from systems in secondary containment following a Design Basis Accident (OBA) and, in conjunction with the Standby Gas Treatment System (SGT) and closure of certain valves whose lines penetrate the secondary containment, to provide for elevated release through the Main Stack.

The secondary containment is a structure that completely encloses the primary containment and those components that may be postulated to contain primary system fluid. This structure forms a control volume that serves to hold up and dilute the fission products. It is possible for the pressure in the control volume to rise relative to the environmental pressure (e.g., due to pump and motor heat load additions). To prevent ground level exfiltration while allowing the secondary containment to be designed as a conventional structure, the secondary containment requires support systems to maintain the control volume pressure at less than the external pressure. Requirements for these systems are specified separately in LCO 3.6.4.2, "Secondary Containment Isolation Valves (SCIVs)," and LCO 3.6.4.3, "Standby Gas Treatment (SGT) System."

APPLICABLE There are two principal accidents for which credit is taken SAFETY ANALYSES for secondary containment OPERABILITY. These are a loss of coolant accident (LOCA) (Ref. 1) and a fuel handling accident inside secondary containment (Ref. 2). The secondary containment performs no active function in response to each of these limiting events; however, its leak (continued)

PBAPS UNIT 2 B 3.6-73 Revision No.

Secondary Containment B 3.6.4.1 BASES APPLICABLE tightness is required to ensure that fission products SAFETY ANALYSES entrapped within the secondary containment structure will be (continued) collected by the SGT System for discharge to the environment via the Main Stack. I Secondary containment satisfies Criterion 3 of the NRC Policy Statement.

LCO An OPERABLE secondary containment provides a control volume into which fission products that leak from primary containment, or are released from the reactor coolant pressure boundary components located in secondary containment, can be discharged to the environment. For the I secondary containment to be considered OPERABLE, it must have adequate leak-tightness to ensure that the required vacuum can be established and maintained.

Movement of recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) requires that secondary containment be OPERABLE.

After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following shutdown (all rods in), fuel moves may proceed without secondary containment OPERABLE provided that the eight external concrete plugs for the RHR rooms (el. 135'), the one external concrete plug for the torus room (el. 135'), and the one HPCI room hatch (el. 116')

remain in place to ensure MCR and offsite dose limits are met. There are no restrictions to other secondary containment hatches or penetrations. After 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months following shutdown (all rods in), Alternative Source Term calculations show that fuel moves may occur with no restrictions on removal or opening of the external concrete plugs and hatch described above.

APPLICABILITY In MODES 1, 2, and 3, a LOCA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, secondary containment OPERABILITY is required during the same operating conditions that require primary containment OPERABILITY.

In MODES 4 and 5, the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining secondary containment OPERABLE is not required in MODE 4 or 5 to ensure a control volume, except for other situations for which significant releases of radioactive material can (continued)

PBAPS UNIT 2 B 3.6-74 Revision No.

Secondary Containment B 3.6.4.1 BASES APPLICABILITY be postulated, such as during operations with a potential (continued) for draining the reactor vessel (OPDRVs), during CORE ALTERATIONS, or during movement of irradiated fuel assemblies in the secondary containment.

Movement of recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the revious 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) requires that secondary containment by OPEdALE.

After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following reactor shutdown (all rods in),

secondary containment operability is not required for movement of irradiated fuel. However, the eight external concrete plugs for the RHR rooms (el. 135'), the one external concrete plug for the torus room (el. 135'), and the one HPCI room hatch (el. 116') must remain in lace to ensure MCR and offsite dose limits are met when moving irradiated fuel within the first 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months after shutdown.

Following 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months after shutdown (all rods in), there are no restrictions on removal or opening of the eternal concrete plugs and hatch described above when moving irradiated fuel.

ACTIONS A1 If secondary containment is inoperable, it must be restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time provides a'period of time to correct the problem that is commensurate with the importance of maintaining secondary containment during MODES 1, 2, and 3. This time period also ensures that the probability of an accident (requiring secondary containment OPERABILITY) occurring during periods where secondary containment is inoperable is minimal.

(continued)

PBAPS UNIT 2 B 3.6-74a Revision No. I

Secondary Containment B 3.6.4.1 BASES ACTIONS B.1 and B.2 (continued)

If secondary containment cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

C.1. C.2. and C.3 Movement of irradiated fuel assemblies in the secondary containment, CORE ALTERATIONS, and OPDRVs can be postulated to cause fission product release to the secondary containment. In such cases, the secondary containment is the only barrier to release of fission products to the environment. CORE ALTERATIONS and movement of irradiated fuel assemblies must be immediately suspended if the secondary containment is inoperable within the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following shutdown. After the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , secondary containment is not required to support movement of irradiated fuel. Note that the eight external concrete plugs for RHR rooms (el. 135'), the one external concrete plug for the torus room (el. 135'), and the one HPCI room hatch (el. 116') must remain in place to ensure dose limits are met for the first 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months following shutdown (all rods in).

Suspension of these activities shall not preclude completing an action that involves moving a component to a safe position. Also, action must be immediately initiated to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

Required Action C.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

(continued)

PBAPS UNIT 2 B 3.6-75 Revision No.

SCIVs B 3.6.4.2 BASES APPLICABLE established by SCIVs is required to ensure that leakage from SAFETY ANALYSES the primary containment is exhausted by the Standby Gas (continued) Treatment (SGT) System for elevated release to the environment via the main stack.

Maintaining SCIVs OPERABLE with isolation times within limits ensures that fission products will remain trapped inside secondary containment until discharged to the environment via the main stack.

SCIVs satisfy-Criterion 3 of the NRC Policy Statement.

LCO SCIVs form a part of the secondary containment boundary.

The SCIV safety function is related to control of offsite radiation releases resulting from DBAs.

The power operated isolation valves are considered OPERABLE when their isolation times are within limits and the valves actuate on an automatic isolation signal. The valves covered by this LCO, along with their associated stroke times, are listed in Reference 2.

The normally closed isolation valves or blind flanges are considered OPERABLE when manual valves are closed or open in accordance with appropriate-administrative controls, automatic SCIVs are de-activated and secured in their closed position, and blind flanges are in place. These passive isolation valves or devices are listed in Reference 2.

APPLICABILITY In MODES 1, 2, and 3, a DBA could lead to a fission product release to the primary containment that leaks to the secondary containment. Therefore, the OPERABILITY of SCIVs is required.

In MODES 4 and 5, the probability and consequences of these events are reduced due to pressure and temperature limitations in these MODES. Therefore, maintaining SCIVs OPERABLE is not required in MODE 4 or 5, except for other situations under which significant radioactive releases can be postulated, such as during operations with a potential for draining the reactor vessel (OPDRVs) or during movement of recently irradiated fuel assemblies in the secondary containment. SCIVs are required to be OPERABLE during handling of recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ). Moving irradiated fuel assemblies in the secondary containment may also occur in MODES 1, 2, and 3.

(continued)

PBAPS UNIT 2 B 3.6-79 Revision No.

SGT System B 3.6.4.3 BASES LCO For Unit 2, one SGT subsystem is OPERABLE when one fan (continued) (OAV020) and associated ductwork, dampers, valves, and controls are OPERABLE. The second SGT subsystem is OPERABLE I

when the other fan (OBV020) and associated ductwork, damper, valves, and controls are OPERABLE.

APPLICABILITY In MODES 1, 2, and 3, a DBA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, SGT System OPERABILITY is required during these MODES.

In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the SGT System in OPERABLE status is not required in MODE 4 or 5, except for other situations under which significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel (OPDRVs) or during movement of recently irradiated fuel assemblies in the secondary containment. The SGT System is required to be OPERABLE during handling of recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

ACTIONS AS With one SGT subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status in 7 days. In this Condition, the remaining OPERABLE SGT subsystem is adequate to perform the required radioactivity release control function. However, the overall system reliability is reduced because a single failure in the OPERABLE subsystem could result in the radioactivity release control function not being adequately performed. The 7 day Completion Time is based on consideration of such factors as the availability of the OPERABLE redundant SGT subsystem and the low probability of a DBA occurring during this period.

B.1 and B.2 If the SGT subsystem cannot be restored to OPERABLE status within the required Completion Time in MODE 1, 2, or 3, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within (continued)

PBAPS UNIT 2 B 3.6-87 Revision No.

SGT System B 3.6.4.3 BASES SURVEILLANCE SR 3.6.4.3.2 REQUIREMENTS (continued) This SR verifies that the required SGT filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The only credited safety function of the SGT System is to provide a secondary containment vacuum sufficient to assure that discharges from the secondary containment will be through the Main Stack. The VFTP test 5.5.7.d. provides verification that the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is acceptable. SR 3.6.4.1.3 and SR 3.6.4.1.4 provide assurance that sufficient vacuum in the secondary containment is established with the time period as used within the DBA LOCA analysis. Specific test frequencies and additional information are discussed in detail in the VFTP.

SR 3.6.4.3.3 This SR verifies that each SGT subsystem starts on receipt of an actual or simulated initiation signal. While this Surveillance can be performed with the reactor at power, operating experience has shown that these components will usually pass the Surveillance when performed at the 24 month Frequency. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.6.2, "Secondary Containment Isolation Instrumentation," overlaps this SR to provide complete testing of the safety function.

Therefore, the Frequency was found to be acceptable from a reliability standpoint.

REFERENCES 1. UFSAR, Section 1.5.1.6.

2. UFSAR, Section 14.9.

PBAPS UNIT 2 B 3.6-90 Revision No.

MCREV System B 3.7.4 BASES BACKGROUND initiate an emergency shutdown of non-essential equipment (continued) and lighting to reduce the heat generation to a minimum.

Heat removal would be accomplished by conduction through the floors, ceilings, and walls to adjacent rooms and to the environment. Additionally, the MCREV System is designed to maintain the control room environment for a 30-day occupancy after a DBA without exceeding 5 rem TEDE. A single MCREV subsystem will pressurize the control room to prevent infiltration of air from surrounding buildings. MCREV System operation in maintaining control room habitability is discussed in the UFSAR, Chapters 7, 10, and 12, (Refs. 1, 2, and 3, respectively).

APPLICABLE The ability of the MCREV System to maintain the SAFETY ANALYSES habitability of the control room is an explicit assumption for the safety analyses presented in the UFSAR, Chapters 10 At and 12 (Refs. 2 and 3, respectively). The MCREV System isNS credited as operating following a loss of coolanttzEimo"r fuel Jandling fccident. The MCREV System is not credited in the analysis of the main steam line break or the control rod drop accident, as discussed in the UFSAR, Section 14.9.1.5 (Ref. 4). The radiological doses to control room personnel as a result of the various DBAs are summarized in Reference 4. No single active or passive failure will cause the loss of outside or recirculated air from the control room.

The MCREV System satisfies Criterion 3 of the NRC Policy Statement.

LCO Two redundant subsystems of the MCREV System are required to be OPERABLE to ensure that at least one is available, assuming a single failure disables the other subsystem.

Total system failure could result in exceeding a dose of 5 rem TEDE to the control room operators in the event of a LOCA or a luel Oandling /ccident (FHA).

The MCREV System is considered OPERABLE when the individual components necessary to control operator exposure are OPERABLE in both subsystems. A subsystem is considered OPERABLE when its associated:

a. Fan is OPERABLE; (continued)

PBAPS UNIT 2 B 3.7-16 Revision No.

MCREV System B 3.7.4 BASES LCO b. HEPA filter and charcoal adsorbers are not excessively (continued) restricting flow and are capable of performing their filtration functions; and

c. Ductwork, valves, and dampers are OPERABLE, and air flow can be maintained.

In addition, the control room boundary must be maintained, including the integrity of the walls, floors, ceilings, and ductwork. Temporary seals may be used to maintain the boundary. In addition, an access door may be opened provided the ability to pressurize the control room is maintained and the capability exists to close the affected door in an expeditious manner.

APPLICABILITY In MODES 1, 2, and 3, the MCREV System must be OPERABLE to control operator exposure during and following a LOCA or FHA, since the LOCA or FHA could lead to a fission product release.

In MODES 4 and 5, the probability and consequences of a LOCA or FHA are reduced because of the pressure and temperature limitations in these MODES. Therefore, maintaining the MCREV System OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioactive releases can be postulated:

a. During operations with potential for draining the reactor vessel (OPDRVs);

b. During CORE ALTERATIONS; and

c. During movement of irradiated fuel assemblies in the secondary containment.

ACTIONS A.

With one MCREV subsystem inoperable, the inoperable MCREV subsystem must be restored to OPERABLE status within 7 days.

With the unit in this condition, the remaining OPERABLE MCREV subsystem is adequate to maintain control room temperature and to perform control room radiation protection. However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could (continued)

PBAPS UNIT 2 B 3.7-17 Revision No.

MCREV System B 3.7.4 BASES ACTIONS AA (continued) result in reduced MCREV System capability. The 7 day Completion Time is based on the low probability of a LOCA or FHA occurring during this time period, and that the remaining subsystem can provide the required capabilities.

B.1 and B.2 In MODE 1, 2, or 3, if the inoperable MCREV subsystem cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE that minimizes risk. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

C.1. C.2.1. C.2.2. and C.2.3 The Required Actions of Condition C are modified by a Note indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.

Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.

During movement of irradiated fuel assemblies in the secondary containment, during CORE ALTERATIONS, or during OPDRVs, if the inoperable MCREV subsystem cannot be restored to OPERABLE status within the required Completion Time, the OPERABLE MCREV subsystem may be placed in operation. This action ensures that the remaining subsystem is OPERABLE, that no failures that would prevent automatic actuation will occur, and that any active failure will be readily detected.

An alternative to Required Action C.1 is to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk.

(continued)

PBAPS UNIT 2 B 3.7-18 Revision No. I

SLC System B 3.1.7 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.7 Standby Liquid Control (SLC) System BASES BACKGROUND The SLC System is designed to provide the capability of bringing the reactor, at any time in a fuel cycle, from full power and minimum control rod inventory (which is at the peak of the xenon transient) to a subcritical condition with the reactor in the most reactive, xenon free state without taking credit for control rod movement. The SLC System satisfies the requirements of 10 CFR 50.62 (Ref. 1) on anticipated transient without scram using enriched boron.

The SLC System is also used to maintain suppression pool pH at or above 7 following a loss of coolant accident (LOCA) involving significant fission product releases. Maintaining suppression pool pH levels at or above 7 following a LOCA ensures that iodine will be retained in the suppression pool water.

Reference 1 requires a SLC System with a minimum flow capacity and boron content equivalent in control capacity to 86 gpm of 13 weight percent sodium pentaborate solution.

Natural sodium pentaborate solution is 19.8% atom Boron-10.

Therefore, the system parameters of concern, boron concentration (C), SLC pump flow rate (Q), and Boron-10 enrichment (E), may be expressed as a multiple of ratios.

The expression is as follows:

C 0 E x x 13% weight 86 gpm 19.8% atom If the product of this expression is 2 1, then the SLC System satisfies the criteria of Reference 1. As such, the equation forms the basis for acceptance criteria for the surveillances of concentration, flow rate, and boron enrichment and is presented in Table 3.1.7-1.

The SLC System consists of a boron solution storage tank, two positive displacement pumps, two explosive valves that are provided in parallel for redundancy, and associated piping and valves used to transfer borated water from the storage tank to the reactor pressure vessel (RPV). The borated solution is discharged near the bottom of the core shroud, where it then mixes with the cooling water rising through the core. A smaller tank containing demineralized water is provided for testing purposes.

(continued)

PBAPS UNIT 3 B 3.1-39 Revision No.

SLC System B 3.1.7 BASES (continued)

APPLICABLE The SLC System is manually initiated from the main control SAFETY ANALYSES room, as directed by the emergency operating procedures, if the operator believes the reactor cannot be shut down, or kept shut down, with the control rods. The SLC System is used in the event that enough control rods cannot be inserted to accomplish shutdown and cooldown in the normal manner. The SLC System injects borated water into the reactor core to add negative reactivity to compensate for all of the various reactivity effects that could occur during plant operations. To meet this objective, it is necessary to inject a quantity of boron, which produces a concentration of 660 ppm of natural boron, in the reactor coolant at 680 F. To allow for potential leakage and imperfect mixing in the reactor system, an additional amount of boron equal to 25% of the amount cited above is added (Ref. 2). The minimum mass of Boron-10 (162.7 lbm) needed for injection is calculated such that the required quantity is achieved accounting for dilution in the RPV with normal water level and including the water volume in the residual heat removal shutdown cooling piping and in the recirculation loop piping. This quantity of borated solution is the amount that is above the pump suction shutoff level in the boron solution storage tank. No credit is taken for the portion of the tank volume that cannot be injected. The maximum concentration of sodium pentaborate listed in Table 3.1.7-1 has been established to ensure that the solution saturation temperature does not exceed 430F.

The sodium pentaborate solution in the SLC System is also used, post-LOCA, to maintain suppression pool pH levels above 7. The system parameters used in the calculation are the Boron-10 minimum mass of 162.7 lbm, and an upper bound Boron-10 enrichment of 65%.

The SLC System satisfies Criterion 4 of the NRC Policy Statement.

LCO The OPERABILITY of the SLC System provides backup capability for reactivity control independent of normal reactivity control provisions provided by the control rods. The OPERABILITY of the SLC System is based on the conditions of the borated solution in the storage tank and the availability of a flow path to the RPV, including the OPERABILITY of the pumps and valves. Two SLC subsystems are required to be OPERABLE; each contains an OPERABLE pump, an explosive valve, and associated piping, valves, and instruments and controls to ensure an OPERABLE flow path.

(continued)

PBAPS UNIT 3 B 3.1-40 Revision No.

Secondary Containment B 3.6.4.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.1 Secondary Containment BASES BACKGROUND The function of the secondary containment is to receive fission products that may leak from primary containment or from systems in secondary containment following a Design Basis Accident (DBA) and, in conjunction with the Standby Gas Treatment System (SGT) and closure of certain valves whose lines penetrate the secondary containment, to provide for elevated release through the Main Stack.

The secondary containment is a structure that completely encloses the primary containment and those components that may be postulated to contain primary system fluid. This structure forms a control. volume that serves to hold up and dilute the fission products. It is possible for the pressure in the control volume to rise relative to the environmental pressure (e.g., due to pump and motor heat load additions). To prevent ground level exfiltration while allowing the secondary containment to be designed as a conventional structure, the secondary containment requires support systems to maintain the control volume pressure at less than the external pressure. Requirements for these systems are specified separately in LCO 3.6.4.2, "Secondary Containment Isolation Valves (SCIVs)," and LCO 3.6.4.3, "Standby Gas Treatment (SGT) System."

APPLICABLE There are two principal accidents for which credit is taken SAFETY ANALYSES for secondary containment OPERABILITY. These are a loss of coolant accident (LOCA) (Ref. 1) and a fuel handling accident inside secondary containment (Ref. 2). The secondary containment performs no active function in response to each of these limiting events; however, its leak (continued)

PBAPS UNIT 3 B 3.6-73 Revision No.

Secondary Containment B 3.6.4.1 BASES APPLICABLE tightness is required to ensure that fission products SAFETY ANALYSES entrapped within the secondary containment structure will be (continued) collected by the SGT System for discharge to the environment via the Main Stack. I Secondary containment satisfies Criterion 3 of the NRC Policy Statement.

LCO An OPERABLE secondary containment provides a control volume into which fission products that leak from primary containment, or are released from the reactor coolant pressure boundary components located in secondary containment, can be discharged to the environment. For the I secondary containment to be considered OPERABLE, it must have adequate leak tightness to ensure that the required vacuum can be established and maintained.

Movement of recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) requires that secondary containment be OPERABLE.

After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following shutdown (all rods in), fuel moves may proceed without secondary containment OPERABLE provided that the eight external concrete plugs for the RHR rooms (el. 135'), the one external concrete plug for the torus room (el. 135'), and the one HPCI room hatch (el. 116')

remain in place to ensure MCR and offsite dose limits are met. There are no restrictions to other secondary containment hatches or penetrations. After 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months following shutdown (all rods in), Alternative Source Term calculations show that fuel moves may occur with no restrictions on removal or opening of the external concrete plugs and hatch described above.

APPLICABILITY In MODES 1, 2, and 3, a LOCA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, secondary containment OPERABILITY is required during the same operating conditions that require primary containment OPERABILITY.

In MODES 4 and 5, the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining secondary containment OPERABLE is not required in MODE 4 or 5 to ensure a control volume, except for other situations for which significant releases of radioactive material can (continued)

PBAPS UNIT 3 B 3.6-74 Revision No.

Containment Secondary Secondary Containment B 3.6.4.1 BASES APPLICABILITY be postulated, such as during operations with a potential (continued) for draining the reactor vessel (OPDRVs), during CORE ALTERATIONS, or during movement of irradiated fuel assemblies in the secondary containment.

Movement of recently irradiated fuel (i.e., fuel that has occupied part 6f a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) requires that secondary containment be OPERABLE.

After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following reactor shutdown (all rods in),

secondary containment operability is not required for movement of irradiated fuel. However, the eight external concrete plugs for the RHR rooms (el. 135'), the one external concrete plug for the torus room (el. 135'), and the on HPCI room hatch (el. 116') must remain in place to ensure MCR and offsite dose limits are met when moving irradiated fuel within the first 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months after shutdown.

Following 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months after shutdown (all rods in), there are no restrictions on removal or opening of the external concrete plugs and hatch described above when moving irradiated fuel.

ACTIONS A.1 If secondary containment is inoperable, it must be restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time provides a period of time to correct the problem that is commensurate with the importance of maintaining secondary containment during MODES 1, 2, and 3. This time period also ensures that the probability of an accident (requiring secondary containment OPERABILITY) occurring during periods where secondary containment is inoperable is minimal.

(continued)

PBAPS UNIT 3 B 3.6-74a Revision No. I

Secondary Containment B 3.6.4.1 BASES ACTIONS B.1 and 8.2 (continued)

If secondary containment cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

C.1. C.2. and C.3 Movement of irradiated fuel assemblies in the secondary containment, CORE ALTERATIONS, and OPDRVs can be postulated to cause fission product release to the secondary containment. In such cases, the secondary containment is the only barrier to release of fission products to the environment. CORE ALTERATIONS and movement of irradiated fuel assemblies must be immediately suspended if the secondary containment is inoperable within the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following shutdown. After the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , secondary containment is not required to support movement of irradiated fuel. Note that the eight external concrete plugs for RHR rooms (el. 135'), the one external concrete plug for the torus room (el. 135'), and the one HPCI room hatch (el. 116') must remain in place to ensure dose limits are met for the first 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months following shutdown (all rods in).

Suspension of these activities shall not preclude completing an action that involves moving a component to a safe position. Also, action must be immediately initiated to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.

Required Action C.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

(continued)

PBAPS UNIT 3 B 3.6-75 Revision No.

SCIVs B 3.6.4.2 BASES APPLICABLE established by SCIVs is required to ensure that leakage from SAFETY ANALYSES the primary containment is exhausted by the Standby Gas (continued) Treatment (SGT) System for elevated release to the environment via the main stack.

Maintaining SCIVs OPERABLE with isolation times within limits ensures that fission products will remain trapped inside secondary containment until discharged to the environment via the main stack.

SCIVs satisfy Criterion 3 of the NRC Policy Statement.

LCO SCIVs form a part of the secondary containment boundary.

The SCIV safety function is related to control of offsite radiation releases resulting from DBAs.

The power operated isolation valves are considered OPERABLE when their isolation times are within limits and the valves actuate on an automatic isolation signal. The valves covered by this LCO, along with their associated stroke times, are listed in Reference 2. I The normally closed isolation valves or blind flanges are considered OPERABLE when manual valves are closed or open in accordance with app ropriate administrative controls, automatic SCIVs are de-activated and secured in their closed position, and blind flanges are in place. These passive isolation valves or devices are listed in Reference 2. I APPLICABILITY In MODES 1, 2, and 3, a OBA could lead to a fission product release to the primary containment that leaks to the secondary containment. Therefore, the OPERABILITY of SCIVs is required.

In MODES 4 and 5, the probability and consequences of these events are reduced due to pressure and temperature limitations in these MODES. Therefore, maintaining SCIVs OPERABLE is not required in MODE 4 or 5, except for other situations under which significant radioactive releases can be postulated, such as during operations with a potential for draining the reactor vessel (OPDRVs) or during movement of recently irradiated fuel assemblies in the secondary containment. SCIVs are required to be OPERABLE during handling of recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ). Moving recently irradiated fuel assemblies in the secondary containment may also occur in MODES 1, 2, and 3.

(continued)

PBAPS UNIT 3 B 3.6-79 Revision No.

SGT System B 3.6.4.3 BASES LCO For Unit 3, one SGT subsystem is OPERABLE when one fan (continued) (OCV020) and associated ductwork, dampers, valves, and controls are OPERABLE. The second SGT subsystem is OPERABLE I

when the other fan (OBV020) and associated ductwork, damper, valves, and controls are OPERABLE.

APPLICABILITY In MODES 1, 2, and 3, a DBA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, SGT System OPERABILITY is required during these MODES.

In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the SGT System in OPERABLE status is not required in MODE 4 or 5, except for other situations under which significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel (OPDRVs) or during movement of recently irradiated fuel assemblies in the secondary containment. The SGT System is required to be OPERABLE during handling of recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

ACTIONS With one SGT subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status in 7 days. In this Condition, the remaining OPERABLE SGT subsystem is adequate to perform the required radioactivity release control function. However, the overall system reliability is reduced because a single failure in the OPERABLE subsystem could result in the radioactivity release control function not being adequately performed. The 7 day Completion Time is based on consideration of such factors as the availability of the OPERABLE redundant SGT subsystem and the low probability of a DBA occurring during this period.

B.1 and B.2 If the SGT subsystem cannot be restored to OPERABLE status within the required Completion Time in MODE 1, 2, or 3, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within (continued)

PBAPS UNIT 3 B 3.6-87 Revision No.

SGT System B 3.6.4.3 BASES SURVEILLANCE SR 3.6.4.3.2 REQUIREMENTS (continued) This SR verifies that the required SGT filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The only credited safety function of the SGT System is to provide a secondary containment vacuum sufficient to assure that discharges from the secondary containment will be through the Main Stack. The VFTP test 5.5.7.d. provides verification that the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is acceptable. SR 3.6.4.1.3 and SR 3.6.4.1.4 provide assurance that sufficient vacuum in the secondary containment is established with the time period as used within the DBA LOCA analysis. Specific test frequencies and additional information are discussed in detail in the VFTP.

SR 3.6.4.3.3 This SR verifies that each SGT subsystem starts on receipt of an actual or simulated initiation signal. While this Surveillance can be performed with the reactor at power, operating experience has shown that these components will usually pass the Surveillance when performed at the 24 month Frequency. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.6.2, "Secondary Containment Isolation Instrumentation," overlaps this SR to provide complete testing of the safety function.

Therefore, the Frequency was found to be acceptable from a reliability standpoint.

REFERENCES 1. UFSAR, Section 1.5.1.6.

2. UFSAR, Section 14.9.

PBAPS UNIT 3 B 3.6-90 Revision No.

MCREV System B 3.7.4 BASES BACKGROUND initiate an emergency shutdown of non-essential equipment (continued) and lighting to reduce the heat generation to a minimum.

Heat removal would be accomplished by conduction through the floors, ceilings, and walls to adjacent rooms and to the environment. Additionally, the MCREV System is designed to maintain the control room environment for a 30-day occupancy after a DBA without exceeding 5 rem TEDE. A single MCREV subsystem will pressurize the control room to prevent infiltration of air from surrounding buildings. MCREV System operation in maintaining control room habitability is discussed in the UFSAR, Chapters 7, 10, and 12, (Refs. 1, 2, and 3, respectively).

APPLICABLE The ability of the MCREV System to maintain the SAFETY ANALYSES habitability of the control room is an explicit assumption for the safety analyses presented in the UFSAR, Chapters 10 and 12 (Refs. 2 and 3, respectively). The MCREV System is credited as operating following a loss of coolant accident or fuel handling accident. The MCREV System is not credited in the analysis of the main steam line break, or the control rod drop accident, as discussed in the UFSAR, Section 14.9.1.5 (Ref. 4). The radiological doses to control room personnel as a result of the various DBAs are summarized in Reference 4. No single active or passive failure will cause the loss of outside or recirculated air from the control room.

The MCREV System satisfies Criterion 3 of the NRC Policy Statement.

LCO Two redundant subsystems of the MCREV System are required to be OPERABLE to ensure that at least one is available, assuming a single failure disables the other subsystem.

Total system failure could result in exceeding a dose of 5 rem TEDE to the control room operators in the event of a LOCA or a fuel handling accident (FHA). I The MCREV System is considered OPERABLE when the individual components necessary to control operator exposure are OPERABLE in both subsystems. A subsystem is considered OPERABLE when its associated:

a. Fan is OPERABLE; (continued)

PBAPS UNIT 3 B 3.7-16 Revision No.

MCREV System B 3.7.4 BASES LCO b. HEPA filter and charcoal adsorbers are not excessively (continued) restricting flow and are capable of performing their filtration functions; and

c. Ductwork, valves, and dampers are OPERABLE, and air flow can be maintained.

In addition, the control room boundary must be maintained, including the integrity of the walls, floors, ceilings, and ductwork. Temporary seals may be used to maintain the boundary. In addition, an access door may be opened provided the ability to pressurize the control room is maintained and the capability exists to close the affected door in an expeditious manner.

APPLICABILITY In MODES 1, 2, and 3, the MCREV System must be OPERABLE to control operator exposure during and following a LOCA or FHA, since the LOCA or FHA could lead to a fission product release.

In MODES 4 and 5, the probability and consequences of a LOCA or FHA are reduced because of the pressure and temperature limitations in these MODES. Therefore, maintaining the I

MCREV System OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioactive releases can be postulated:

a. During operations with potential for draining the reactor Yessel (OPDRVs); and

b. During CORE ALTERATIONS; and

c. During movement of irradiated fuel assemblies in the secondary containment.

ACTIONS A.

With one MCREV subsystem inoperable, the inoperable MCREV subsystem must be restored to OPERABLE status within 7 days.

With the unit in this condition, the remaining OPERABLE MCREV subsystem is adequate to maintain control room temperature and to perform control room radiation protection. However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could (continued)

PBAPS UNIT 3 B 3.7-17 Revision No.

MCREV System B 3.7.4 BASES ACTIONS AJ1 (continued) result in reduced MCREV System capability. The 7 day Completion Time is based on the low probability of a LOCA or FHA occurring during this time period, and that the remaining subsystem can provide the required capabilities.

B.1 and B.2 In MODE 1, 2, or 3, if the inoperable MCREV subsystem cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE that minimizes risk. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

C.1. C.2.1. C.2.2. and C.2.3 The Required Actions of Condition C are modified by a Note indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.

Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.

During movement of irradiated fuel assemblies in the secondary containment, during CORE ALTERATIONS, or during OPDRVs, if the inoperable MCREV subsystem cannot be restored to OPERABLE status within the required Completion Time, the OPERABLE MCREV subsystem may be placed in operation. This action ensures that the remaining subsystem is OPERABLE, that no failures that would prevent automatic actuation will occur, and that any active failure will be readily detected.

An alternative to Required Action C.1 is to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk.

(continued)

PBAPS UNIT 3 B 3.7-18 Revision No.

ATTACHMENT 4 PEACH BOTTOM ATOMIC POWER STATION UNITS 2 AND 3 Docket Nos. 50-277 50-278 Renewed License Nos. DPR-44 DPR-56 Supplement to License Amendment Request for "PBAPS Alternative Source Term Implementation" Referenced Drawings (For Information Only)

M-1, Rev. 17 - General Arrangement Plan at El. 91' - 6" M-2, Rev. 20 - General Arrangement Plan at El. 116' - 0" M-3, Rev. 21 - General Arrangement Plan at El. 135' - 0" A-486, Rev. 6- Barrier Plans Elev. 135' 0" A-489, Rev. 4- Barrier Plans Elev. 234' 0"

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ATTACHMENT 5 PEACH BOTTOM ATOMIC POWER STATION UNITS 2 AND 3 Docket Nos. 50-277 50-278 Renewed License Nos. DPR-44 DPR-56 Supplement to License Amendment Request for "PBAPS Alternative Source Term Implementation" Commitment Page

Supplement to License Amendment Request for "PBAPS Alternative Source Term Implementation" Attachment 5 Commitment Page Page 1 of 1 PBAPS COMMITMENT The following table identifies those actions committed to by Exelon in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

Commitment Continuing Scheduled Compliance Completion Date PBAPS commits to incorporate administrative controls X Upon for secondary containment hatches to ensure these Implementation hatches remain closed while moving irradiated fuel without sufficient decay. The affected hatches for each unit consist of the eight external concrete plugs for the RHR rooms (el. 135'), the one external concrete plug for the torus room (el. 135'), and the one HPCI room hatch (el. 116'). This commitment shall be applicable for a period of 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months following shutdown.

ML050400083

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